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047626720	summary	BACKGROUND OF THE INVENTION A fast breeder reactor using liquid sodium as a coolant has a core comprising a core region charged with a fuel material prepared by enriching plutonium, a fissile material, and an external blanket region surrounding said core region and moreover charged with a fuel material (e.g. natural uranium or depleted uranium) whose main component is a fertile material (e.g. uranium-238). The output of the fast breeder reactor is controlled by moving control rods in and out of the core. Each control rod is provided with a plurality of neutron absorber rods packed inside with B.sub.4 C, a neutron absorber. The length of a region packed with the neutron absorber inside a neutron absorber rod is equal to that of the aforesaid core region. The control rods are moved vertically through control rod guide tubes installed in the core. The control rods are classified roughly into two kinds according to their functions. One is an output regulation control rod; the other is a reactor shutdown control rod. The output regulation control rod is inserted into the core region in the initial stage of operation of the fast breeder reactor to limit initial burnup reactivity. Then, it raised from the core region as burnup reactivity lowers as the operation proceed, and is raised completely out of the core region in the last stage of the operation. The output regulation control rod, however, is reinserted into the core region when the operation of the fast breeder reactor is stopped. Meanwhile, the reactor shutdown control rods are all held up outside the core region when the reactor is started, and are not inserted into the core region during the operation of the reactor. They are inserted thereinto when the operation of the reactor is stopped. In other words, the output regulation control rod regulates the output of the fast breeder reactor and stops the reactor, while the reactor shutdown control rod shuts down the reactor. The number of output regulation and reactor shutdown control rods is about equal. The neutron absorber rod in a control rod as disclosed in the Official Gazette on Japanese Patent Laid-Open No. 65794/1975 is provided, in the upper inside, with a gas plenum which accumulates He gas produced through the absorption of neutrons by B.sub.4 C. This gas plenum, however, does not increase core reactivity. SUMMARY OF THE INVENTION An object of the present invention is to furnish a fast breeder reactor which increases core reactivity. In a fast breeder reactor provided with a core comprising a core region packed with a fissile material and a blanket region surrounding the outside of said core region and formed mostly of a fertile material, and a plurality of control rods moved in and out of the core region by a control rod driving device, a fast breeder reactor characterized in that each of said control rods is constructed of a neutron absorber region packed with a neutron absorber, and a gas region disposed in the end portion on the side further separated from said control rod driving device than said neutron absorber region, can be obtained according to the present invention.
	description	This application is a continuation-in-part of application Ser. No. 12/891,317 filed Sep. 27, 2010 and titled “COMPACT NUCLEAR REACTOR WITH INTEGRAL STEAM GENERATOR”. The following relates to the nuclear reactor arts, steam generator and steam generation arts, electrical power generation arts, and related arts. Compact nuclear reactors are known for maritime and land-based power generation applications and for other applications. In some such nuclear reactors, an integral steam generator is located inside the reactor pressure vessel, which has advantages such as compactness, reduced likelihood of a severe loss of coolant accident (LOCA) event due to the reduced number and/or size of pressure vessel penetrations, retention of the radioactive primary coolant entirely within the reactor pressure vessel, and so forth. Disclosed herein are further improvements that provide reduced cost, simplified manufacturing, and other benefits that will become apparent to the skilled artisan upon reading the following. In one aspect of the disclosure, an apparatus comprises: a generally cylindrical pressure vessel defining a cylinder axis; a nuclear reactor core disposed in the generally cylindrical pressure vessel; a central riser disposed coaxially inside the generally cylindrical pressure vessel, the central riser being hollow and having an end proximate to the nuclear reactor core to receive primary coolant heated by the nuclear reactor core and an open end distal from the nuclear reactor core discharging the primary coolant; a once through steam generator (OTSG) disposed in an annular volume defined between the central riser and the generally cylindrical pressure vessel, the primary coolant discharged from the open end of the central riser flowing through the OTSG and heating secondary coolant also flowing through the OTSG, the primary coolant and the secondary coolant being disposed in separate flow paths in the OTSG; and internal primary coolant pumps arranged to circulate primary coolant within the pressure vessel, the internal primary coolant pumps having all mechanical and electromagnetomotive components including at least a motor and at least one impeller disposed inside the pressure vessel. In another aspect of the disclosure, an apparatus comprises: a pressurized water nuclear reactor (PWR) including a generally cylindrical pressure vessel defining a cylinder axis, a nuclear reactor core disposed in the generally cylindrical pressure vessel, and a hollow central riser disposed coaxially inside the generally cylindrical pressure vessel with an end proximate to the nuclear reactor core to receive primary coolant heated by the nuclear reactor core and an open end distal from the nuclear reactor core discharging the primary coolant; a once through steam generator (OTSG) disposed in the pressure vessel and including a primary coolant flow path and a secondary coolant flow path, primary coolant flowing in the primary coolant flow path heating secondary coolant flowing in the secondary coolant flow path to generate secondary coolant comprising steam; and a divider plate spaced apart from the open end of the central riser. The generally cylindrical pressure vessel includes a sealing portion cooperating with the divider plate to define an integral pressurizer volume that is separated by the divider plate from the remaining interior volume of the generally cylindrical pressure vessel. The integral pressurizer volume contains primary coolant at a pressure greater than a pressure of primary coolant disposed in the remaining interior volume of the generally cylindrical pressure vessel. In another aspect of the disclosure, an apparatus comprises: a generally cylindrical pressure vessel defining a cylinder axis; a nuclear reactor core disposed in the generally cylindrical pressure vessel; a central riser disposed coaxially inside the generally cylindrical pressure vessel, the central riser being hollow and having an end proximate to the nuclear reactor core to receive primary coolant heated by the nuclear reactor core and an open end distal from the nuclear reactor core discharging the primary coolant; a once through steam generator (OTSG) disposed in an annular volume defined between the central riser and the generally cylindrical pressure vessel, the primary coolant discharged from the open end of the central riser flowing through the OTSG and heating secondary coolant also flowing through the OTSG, the primary coolant and the secondary coolant being disposed in separate flow paths in the OTSG; neutron absorbing control rods; and an internal control rod drive mechanism (CRDM) configured to controllably insert and withdraw the control rods into and out of the nuclear reactor core, the internal CRDM having all mechanical and electromagnetomotive components including at least a motor and a lead screw disposed inside the pressure vessel. With reference to FIG. 1, a perspective partial sectional view of a illustrative nuclear reactor is shown. A nuclear reactor core 10 is disposed inside a generally cylindrical pressure vessel. In the illustrative embodiment the pressure vessel includes a lower pressure vessel portion or section 12 housing the nuclear reactor core 10, an upper vessel portion or section 14, and mid-flange region 16. This is merely an illustrative configuration, and the pressure vessel can in general be constructed of as few as a single portion or section, or two portions or sections, three portions or sections (as illustrated), four portions or sections (for example including a fourth upper “cap” portion or section separate from the upper portion or section), or so forth. The pressure vessel 12, 14, 16 contains primary coolant, which in the illustrative case of a light water reactor is water (H2O), optionally including other additives for reactivity control, such as a boron compound (e.g., “borated water”). In other contemplated embodiments the primary coolant may be another fluid, such as heavy water (D2O). The primary coolant fills most or all of the volume of the pressure vessel 12, 14, 16. A reactor inlet annulus 18 surrounds the reactor core 10 to enable primary coolant to flow to the reactor core 10. Optional shielding or shrouding 20 disposed in the reactor inlet annulus 18 provides additional radiation shielding from the reactor core 10. The illustrative reactor is a pressurized water reactor (PWR) in which the primary coolant is sub-cooled light water maintained under an elevated pressure at a temperature below the boiling point (saturation temperature) at the operating pressure; however, a boiling water reactor (BWR) in which the primary coolant operates at the saturation temperature at an elevated pressure, or another reactor configuration such as a configuration employing heavy water, is also contemplated. Reactor control is provided by upper internal neutron-absorbing control rods 22 and a control rod drive mechanism (CRDM) 24 that is configured to controllably insert and withdraw the control rods into and out of the nuclear reactor core 10. Diagrammatic FIG. 1 only identifies two illustrative control rods 22; however, in some embodiments the control rods may number in the dozens or hundreds, with insertion points spatially distributed across the reactor core area to collectively provide uniform reaction control. The CRDM 24 may be divided into multiple units, each controlling one or more control rods. For example, a plurality of control rods may be operatively coupled with a single CRDM unit via a connecting rod/spider assembly or other suitable coupling (details not illustrated). In some illustrative embodiments, a CRDM unit includes a motor driving a lead screw operatively connected with control rods via a connecting rod/spider assembly, such that motor operation causes linear translation of the assembly including the lead screw, connecting rod, spider, and control rods. Such CRDM units provide fine control of the precise insertion of the control rods into the reactor core 10 via the lead screw, and hence are suitable for “gray rod” operation providing fine incremental reaction control. In some illustrative embodiments, a CRDM unit may comprise a lifting piston that lifts an assembly including the connecting rod, spider, and control rods out of the reactor core 10, and during a SCRAM removes the lifting force to allow the control rods to fall into the reactor core 10 by gravity and optional hydraulic pressure force(s). Such CRDM units are suitably used for “shutdown rod” operation, as part of the reactor safety system. In yet other illustrative embodiments, the gray rod and shutdown rod functionality is integrated into a single CRDM unit, for example using a separable ball nut coupling with a lead screw such that the CRDM unit normally provides gray rod functionality but during a SCRAM the ball nut separates to release the control rods into the reactor core 10. Some further illustrative embodiments of CRDM units are set forth in application Ser. No. 12/722,662 titled “Control Rod Drive Mechanism for Nuclear Reactor” filed Mar. 12, 2010 and related application Ser. No. 12/722,696 titled “Control Rod Drive Mechanism For Nuclear Reactor” filed Mar. 12, 2010 are both incorporated herein by reference in their entireties. These applications disclose CRDM units providing gray/shutdown rod functionality, in which the connection between the motor and the lead screw is not releasable, but rather a separate latch is provided between the lead screw and the connecting rod in order to effectuate SCRAM. In these alternative configurations the lead screw does not SCRAM, but rather only the unlatched connecting rod and control rod SCRAM together toward the reactor core while the lead screw remains engaged with the motor. The diagrammatically illustrated CRDM 24 may include one or more CRDM units including various combinations of CRDM units of the described types or other CRDM unit configurations providing gray and/or shutdown rod functionality. The illustrative CRDM 24 is an internal CRDM in which all mechanical and electromagnetomotive components, including the motor, lead screw, connecting rod, and so forth are disposed inside the pressure vessel 12, 14, 16, with only electrical wires, hydraulic lines, or other power or control leads connecting with these components. In other contemplated embodiments, the CRDM may employ external CRDM units in which the motor is mounted outside the pressure vessel, for example above or below. With continuing reference to FIG. 1, the primary coolant may be circulated naturally, due to natural circulation set up by heating due of the primary coolant in the vicinity of the operating nuclear reactor core 10. Additionally or alternatively, the primary coolant circulation may be driven or assisted by optional reactor coolant pumps 26. The diagrammatically illustrated coolant pumps 26 are internal pumps having rotor and stator elements both located inside the pressure vessel 12, 14, 16. An advantage of the illustrated internal reactor coolant pumps 26 over external pumps or pumps employing an external motor (that is, in which the pump or the motor are located outside of the pressure vessel) is that the number, size, and complexity of vessel penetrations is substantially reduced. The only penetrations introduced by the internal reactor coolant pumps 26 are electrical feedthroughs, and these can be routed so as to minimize the likelihood and seriousness of a LOCA event. For example, the electrical lines can be routed upward so that the corresponding pressure vessel penetrations are at or near the top of the pressure vessel. Additionally or alternatively, if the electrical lines for driving the internal reactor coolant pumps 26 are bundled with other electrical lines such as those connecting with reactor core sensors or the like, then the internal reactor coolant pumps 26 do not introduce any additional pressure vessel penetrations. Alternatively, an external pump can be employed, for example having an external stator and a rotor coupled with the pressure vessel volume via a suitable conduit or tube, or the circulation pumps may be omitted entirely, as per natural circulation reactor embodiments. The use of external pumps has the advantage of easier access for pump repair or replacement. The nuclear reactor is further described with continuing reference to FIG. 1 and with further reference to FIGS. 2 and 3. FIG. 2 illustrates a side sectional view of the upper vessel 14 and selected components therein, while FIG. 3 shows Section D-D indicated in FIG. 2. As seen in FIG. 1, the illustrative nuclear reactor is an integral nuclear reactor, by which it is meant that a steam generator 30 is integrated inside the pressure vessel 12, 14, 16. In the illustrative example, the pressure vessel 12, 14, 16 is generally cylindrical and defines a cylinder axis A (labeled only in FIG. 2). The steam generator 30 is a straight-tube once-through steam generator (OTSG) 30 disposed in the upper vessel 14 above the CRDM 24, as seen in FIG. 1. The OTSG 30 includes straight tubes 32 arranged vertically in parallel with the cylinder axis A in an annular “downcomer” volume 34 defined between: (i) a hollow central riser 36 disposed coaxially in the upper portion 14 of the generally cylindrical pressure vessel, and (ii) the upper portion 14 of the generally cylindrical pressure vessel. The hollow central riser 36 defines a central riser flow path 38 inside the central riser 36. The OTSG 30 also includes an outer shroud 40 surrounding the tubes 32 disposed in the downcomer volume 34, and an inner shroud 42 disposed between the central riser 36 and the tubes 32. (Note that in FIGS. 2 and 3, the OTSG shrouds 40, 42 are shown and labeled, but the tubes 32 are omitted so as to more clearly show the annular downcomer volume 34 in FIGS. 2 and 3). The primary coolant flow path in the illustrative reactor is as follows. The central riser 36 has a bottom end proximate to the nuclear reactor core 10 to receive primary coolant heated by the nuclear reactor core 10, and a top end distal from the nuclear reactor core 10. Primary coolant heated by the nuclear reactor core 10 flows upward through the central riser flow path 38 inside the central riser 36. At the top of the central riser 36 the primary coolant flow turns approximately 180° (that is, from flowing generally upward to flowing generally downward). The primary coolant enters the tubes 32 of the OTSG 30 and flows downward through the tubes 32. The primary coolant is discharged from the lower ends of the tubes 32 into a primary outlet plenum 44, which passes the primary coolant flow back to the reactor inlet annulus 18 and back to the reactor core 10. In the illustrative embodiment, the illustrated internal reactor coolant pumps 26 are located on the cold side below the steam generator 30. In some embodiments, the coolant pumps 26 drive the primary coolant flow. In some embodiments, the coolant pumps 26 provide assistance to natural circulation driving the primary coolant flow. In either case, the location of the coolant pumps 26 on the cold side below the steam generator 30 advantageously promotes flow of the cooled primary coolant through the complex approximately 180° turn as the primary coolant is redirected upward into the core 10. The placement of the primary coolant pumps 26 on the cold side below the steam generator 30 also reduces the temperature and thermal stress on the coolant pumps 26, thus further facilitating the use of internal reactor coolant pumps 26. With continuing reference to FIGS. 1-3 and with further reference to FIG. 4, the outer and inner shrouds 40, 42 of the OTSG 30 define a fluid flow volume of the OTSG 30 between the shrouds 40, 42. This fluid flow volume surrounds the tubes 32, and has a feedwater inlet 50 and a steam outlet 52. Note that although a single inlet 50 and single outlet 52 are illustrated, in other embodiments there may be multiple inlets and/or multiple outlets, to provide redundancy and/or improved radial symmetry in the plane transverse to the axis A. Fluid (e.g., feedwater) is injected into the fluid flow volume at the feedwater inlet 50 and is discharged from the fluid flow volume (e.g., as steam) at the steam outlet 52. While in the fluid flow volume, the fluid flows outside the tubes 32 of the OTSG 30 in a generally upward direction generally opposite flow of primary coolant inside the tubes 32. With continuing reference to FIGS. 1-3 and with further reference to FIG. 4, in the operating state of the illustrative PWR, feedwater injected into the fluid flow volume of the OTSG 30 at the feedwater inlet 50 is converted to steam by heat emanating from primary coolant flowing inside the tubes 32 of the OTSG 30, and the steam is discharged from the fluid flow volume at a steam outlet 52. This is diagrammatically shown in FIG. 4, which shows portions of three illustrative tubes 32 carrying downward primary coolant flow (Fprimary). The fluid flow volume of the OTSG 30 is diagrammatically shown in FIG. 4 by indication of portions of the outer and inner shrouds 40, 42 that define the fluid flow volume of the OTSG 30. To facilitate correlation with FIGS. 1-3, the axial direction corresponding to the axis A of the generally cylindrical pressure vessel is also indicated in FIG. 4. The fluid flowing in the fluid flow volume of the OTSG 30 is sometimes referred to herein as “secondary” coolant, and the generally upward “counter” flow of the secondary coolant in the fluid flow volume of the OTSG 30 is indicated as secondary coolant flow (Fsecondary) in diagrammatic FIG. 4. During the upward flow, heat emanating from the primary coolant flow Fprimary transfers to the secondary coolant flow Fsecondary, causing the secondary coolant to heat until it is converted to secondary coolant flow having the form of steam flow (Ssecondary). (The steam flow Ssecondary is also diagrammatically indicated in FIG. 4 by using dotted arrows). Although not illustrated, the steam flow Ssecondary exiting the steam outlet 52 suitably serves as working steam that flows to and operates a turbine or other steam-operated device. In the illustrative embodiment, the fluid flow volume of the OTSG 30 is defined by the outer and inner shrouds 40, 42 that are separate from the central riser 36 and the upper portion 14 of the pressure vessel. Advantageously, this enables the OTSG 30 to be constructed as a unit including the tubes 32 and surrounding shrouds 40, 42, and to then be installed as a unit in the upper portion 14 of the pressure vessel. However, it is also contemplated for the inner shroud to be embodied as the outer surface of the central riser 36, and/or for the outer shroud to be embodied as an inner surface of the upper portion 14 of the pressure vessel. In embodiments which include the outer shroud 40 which is separate from the upper pressure vessel portion 14 (as in the illustrative embodiment), an annular space between the outer shroud 40 and the pressure vessel 14 may optionally be employed for a useful purpose. In the illustrative example, the annular space between the outer shroud 40 and the pressure vessel 14 defines a feedwater annulus 60 between an outer shroud 40 of the OTSG 30 and the pressure vessel (upper portion 14) buffers feedwater injected into the fluid flow volume at the feedwater inlet 50. Similarly, a steam annulus 62 between the outer shroud 40 of the OTSG 30 and the pressure vessel (upper portion 14) buffers steam discharged from the fluid flow volume at the steam outlet 52. In some embodiments, the feedwater annulus and the steam annulus have equal inner diameters and equal outer diameters. In such embodiments the outer shroud and the relevant pressure vessel portion have constant diameters over the axial length of the annuluses. In the illustrative embodiment, however, the feedwater annulus 60 has a larger outer diameter than the steam annulus 62. This is obtained by increasing the diameter of the upper pressure vessel portion 14 surrounding the feedwater annulus 60 as compared with the diameter of the upper pressure vessel portion 14 surrounding the steam annulus 62. In the illustrative embodiment the diameter of the outer shroud 40 remains constant over the axial length of the annuluses 60, 62. This configuration allows a larger local inventory of water so that the time available for steam generator boil-off is relatively longer in the event of a loss-of-feedwater (LOFW) accident. With reference to FIGS. 1 and 2, as already mentioned the flow circuit for the primary coolant includes an approximately 180° flow reversal as the primary coolant discharges from the central riser flow path 38 inside the central riser 36 and flows into the top ends of the tubes 32 of the OTSG 30. Optionally, a flow diverter 70 is provided to facilitate this flow reversal. The illustrative flow diverter 70 is disposed in the generally cylindrical pressure vessel 14 and has a flow diverting surface 72 facing the top end of the central riser that is sloped or (as illustrated) curved to redirect primary coolant discharged from the top end of the central riser 36 toward inlets of the tubes 32 of the OTSG 30. The flow diverter 70 is spaced apart from the top of the central riser 36 by a primary inlet plenum 74. At the lower end, the internal reactor coolant pumps 26 drive or assist the primary coolant flow as it makes another approximately 180° flow reversal to enter in an upward direction into the bottom of the reactor core 10. With reference to FIG. 5, in some embodiments there are no fluid couplings between the coolant pumps 26 and individual tubes 32 of the steam generator 30. In these embodiments the pumps 26 drive primary coolant flow downward generally and act on a plurality of outlets of nearby steam generator tubes 32. In the illustrative example shown in FIG. 5, a diagrammatically indicate path PP shows the flow of the primary coolant in the lower portion of the pressure vessel. The primary coolant flow path PP discharges from the lower ends of the tubes 32 (not specifically shown in FIG. 5) of the steam generator 30 into the primary outlet plenum 44. The lower end of the feedwater annulus 60 terminates at in the mid flange region 16, which allows for the reactor inlet annulus 18 to reside at a larger radius than the steam generator 30. The primary outlet plenum 44 extends from the smaller radius of the steam generator 30 to the larger radius of the reactor inlet annulus 18 so as to provide a fluid connection from the primary coolant outlet of the steam generator 30 into the reactor inlet annulus 18. However, as seen in FIG. 5, the primary coolant flow path PP makes a relatively sharp dogleg in order to transition from the smaller radius of the steam generator 30 to the larger radius of the reactor inlet annulus 18. In the illustrative embodiment, this flow transition is facilitated by locating the internal reactor coolant pumps 26 at or near the top of the reactor inlet annulus 18 proximate to the primary outlet plenum 44. The fluid coupling for primary coolant from the outlet of the steam generator 30 to the inlets of the internal reactor coolant pumps 26 is therefore indirect via the primary outlet plenum 44. Omitting fluid couplings between the steam generator and the coolant pumps facilitates independent maintenance of the steam generator 30 and the pumps 26. For example, during an opening of the pressure vessel the steam generator 30 can be removed from the pressure vessel while leaving the coolant pumps 26 intact and in place in the pressure vessel. Conversely, if suitable manways are provided for accessing the pumps 26, an individual coolant pump 26 can be removed or replaced while leaving both the steam generator 30 and the other coolant pumps 26 in intact and in place in the pressure vessel. Removal or replacement of either the steam generator 30 or one (or more) of the pumps 26 is also advantageously simplified since there are no steam generator/pump fluid couplings to disengage or engage. With reference to FIG. 6, an illustrative embodiment of one of the internal primary coolant pumps 26 is diagrammatically shown in sectional perspective view. In general, the internal primary coolant pumps can be embodied by any self-contained pump that is robust against the environment inside the pressure vessel. Some suitable pumps include spool pumps. See, e.g., Kitch et al., U.S. Pat. No. 6,813,328. The illustrative pump 26 shown in FIG. 6 is a spool pump including a cylindrical pump stator 2602 containing a concentrically arranged cylindrical pump rotor 2604 defining a cylindrical central flow region 2606 through which primary coolant flow Fprimary is driven from an inlet 2608 to an outlet 2610. Impeller blades 2612 extend between the pump rotor 2604 and a central hub 2614 so that the assembly of the rotor 2604, impeller blades 2612, and hub 2614 rotate together inside the stator 2602. Alternatively, the impeller blades can be mounted on the hub alone with other couplings securing the hub with the rotor, or the impeller blades be mounted “single-ended” extending inwardly from the inner surface of the rotor into the flow region 2606. Annular groove/nub couplings between the rotor and stator or other couplings (not shown) prevent axial shifting of the rotor 2604 within the stator 2602 (that is, preventing shifting of the rotor 2604 along the direction of the primary coolant flow Fprimary). The stator 2602 and the rotor 2604 cooperatively define a motor imparting torque that turns the rotating assembly 2604, 2612, 2614. Substantially any motor design can be employed. In the illustrative example of FIG. 6, the stator 2602 is electrically driven via electrical power lines 2616 and includes suitable electrical conductor windings (not shown) of copper or another suitably electrically conductive and heat-resistant material, while the rotor 2604 can be suitably embodied as windings or other conductors that inductively interact with the stator 2602, or as a permanent magnet, or so forth. It is also contemplated to employ an electrically powered rotor. By way of illustrative example, the motor defined by the stator 2602 and the rotor 2604 may be a salient pole motor, a brushless DC motor, or so forth. The stator 2602 is preferably hermetically sealed so that the primary coolant does not contact windings. Alternatively, the stator can be unsealed and the primary coolant windings made of materials that are robust against exposure to the primary coolant. Similarly, the windings, conductors, or electromagnet of the rotor 2604 should be hermetically sealed and/or should be made of material or materials that are robust against exposure to the primary coolant. Because the internal primary coolant pumps 26 are not coupled with the steam 30, the inlet 2608 and outlet 2610 of the pump 26 are not constrained to have any particular coupling connector or other particular configuration. In the illustrative pump 26 diagrammatically shown in FIG. 6, the inlet 2608 and outlet 2610 are maximally large, with radii constrained only by the inner diameter of the rotor 2604. This advantageously minimizes the fluid flow impedance. The lack of fluid coupling of the primary coolant pumps 26 with any particular component also facilitates maintenance or removal of the pump, as already noted. Yet another advantage of omitting fluid coupling with any particular component is that it provides flexibility in the number and size of internal primary coolant pumps included in the system. In spite of these advantages, if desired the inlet and/or outlet of the pumps can optionally have a particular configuration or coupling to facilitate and/or steer fluid flow. As yet another contemplated variation, a constriction of the outlet is optionally included to produce flow acceleration, or the cylindrical stator and rotor can be conical in order to provide primary coolant flow acceleration (if the cone is narrowing from inlet to outlet) or flow deceleration (if the cone is widening from inlet to outlet). The placement of the illustrative internal primary coolant pumps 26 is advantageous for the illustrative embodiment, since it facilitates fluid flow through the dogleg at the primary outlet plenum 44. In other embodiments, a different placement of the internal primary coolant pumps may be advantageous. In general, because the internal coolant pumps 26 are not coupled with the steam generator or any other component, they can be placed anywhere in the primary coolant flow circuit. Placement in the cold leg, as illustrated, advantageously reduces operating temperature and thermal stress on the pumps 26; however, placement in the hot leg is also contemplated. Although not illustrated, placement higher up in the pressure vessel, for example in the central riser flow path 38 inside the central riser 36 (which would place the pumps in the hot leg of the primary coolant flow circuit), is also contemplated so as to improve accessibility of the pumps for maintenance or replacement. Still further, it is contemplated to place internal primary coolant pumps at various places in the primary coolant flow circuit—for example, the embodiment illustrated in FIG. 5 can be modified by retaining the illustrated pumps 26 located at or near the top of the reactor inlet annulus 18 proximate to the primary outlet plenum 44, and additionally include additional pumps located in the central riser 36. As previously mentioned, the illustrative nuclear reactor is a pressurized water reactor (PWR) in which the primary coolant is sub-cooled and maintained under positive pressure. In some embodiments, the pressurization of the primary coolant is provided by an external pressurizer. However, in the illustrative embodiment the pressurization of the primary coolant is provided by an internal pressurizer. In this configuration, the flow diverter 72 also serves as a part of the divider plate 75 spaced apart from the top end of the central riser 36 by the aforementioned primary inlet plenum 74. The generally cylindrical pressure vessel 12, 14, 16 (and, more precisely, the upper pressure vessel portion 14) includes a sealing top portion 78 cooperating with the divider plate 75 to define an integral pressurizer volume 80 that is separated by the divider plate 75 from the remaining interior volume of the generally cylindrical pressure vessel 12, 14, 16. In the operating state of the PWR, the integral pressurizer volume 80 contains fluid (saturated primary coolant, liquid and steam) at a temperature that is greater than the temperature of the primary coolant disposed in the remaining interior volume of the generally cylindrical pressure vessel 12, 14, 16. In this embodiment, the saturation temperature is maintained by pressurizer heaters 82 (shown only in FIG. 1), while pressurizer spray nozzles 84 provide a mechanism for reducing the pressure by condensing some of the steam vapor in volume 80. The pressurizer heaters 82 may, for example, be electrical heaters. The divider plate 75 suitably includes openings (not shown) providing hydraulic fluid communication between the integral pressurizer volume 80 and the remaining interior volume of the generally cylindrical pressure vessel 12, 14, 16. This hydraulic fluid communication establishes the pressure level in the remaining interior volume of the generally cylindrical pressure vessel 12, 14, 16. Since there is a temperature difference across divider plate 75 between the pressurizer volume 80 and primary inlet plenum 74, the remaining primary fluid in the interior volume of the generally cylindrical pressure vessel 12, 14, 16 is maintained at sub-cooled liquid conditions at a temperature approximately 11° C. (20° F.) less than the saturation temperature in pressurizer volume 80. This level of sub-cooled liquid prevents the primary fluid in reactor core 10 from experiencing saturated bulk boiling which has a significantly higher vapor volume fraction than sub-cooled nucleate boiling typically present in pressurized water nuclear reactor cores. This prevention of bulk boiling in a PWR core is made possible by the pressurizer (80, 82, 84, 78, and 75) and is beneficial for the integrity of the nuclear reactor fuel rods by minimizing the probability of departure from nucleate boiling (DNB) which increases the fuel pellet and fuel cladding temperatures. Having set forth an illustrative integral PWR as an illustrative example in FIGS. 1-6, some further additional aspects and variants are set forth next. In one illustrative quantitative example, the reactor core 10 in the operating state operates at 425 MW thermal. The hot reactor coolant water flows in a circuit, called the hot leg, which includes the space above the core flowing around the CRDM's 24. The hot leg extends up the central riser 36 to the inlet plenum 74, wherein the reactor coolant subsequently enters into the tubes 32 of the straight-tube OTSG 30 via the central riser flow path 38. The straight-tube OTSG 30 encircles the central riser 36 and includes the annular array of steam generator tubes 32 disposed in the annulus between the central riser 36 and the outer shroud 40 of the OTSG 30. An advantage of this configuration is that the central riser 36 is a high pressure component separating the high pressure reactor primary coolant at 1900 psia (in this illustrative quantitative embodiment) from the lower pressure secondary coolant which in this example is at 825 psia. The use of an internal pressure part via the central riser 36 yields a compact and efficient design since the primary pressure boundary is internal to the steam generator 30 and serves the dual use as a riser defining the flow path 38 for the hot leg. One design consideration is that there is differential thermal expansion between central riser 36, the tubes 32, and the upper vessel 14. The differential expansion is further complicated by the feedwater annulus 60 containing feedwater at a substantially lower temperature than the steam in the steam annulus 62, which results in a range of temperatures in the upper vessel 14, causing additional thermal stress. One approach for mitigating the effect of these differential stresses is to balance the stresses over the operational and non-operational range of conditions of the steam generator. In one illustrative example, the tubes 32 are made of an austenitic nickel-chromium-based alloy, such as Inconel™ 690, and the tubes 32 are secured in a support made of steel. The support includes an upper tubesheet 90 and a lower tubesheet 92 (diagrammatically indicated in FIG. 2). In general, the austenitic nickel-chromium-based alloy will have a higher coefficient of thermal expansion than the steel. The balancing of the stresses over the operational and non-operational range of conditions is suitably accomplished by pre-stressing the Inconel™ 690 steam generator tubes 32 by expanding the tubes 32 into mating holes of the upper and lower tubesheets 90, 92. This expansion draws the tubes into tension via the Poisson effect. In general, the concept is that in the operating state of the nuclear reactor the primary coolant flowing in the tubes 32 of the OTSG 30 is at a relatively high temperature, for example a temperature of at least 500° C., and the tubes 32 of the OTSG 30 are designed to be under axial compression in this operating state at high temperature. On the other hand, the tubes 32 of the OTSG 30 are designed to be under axial tension in a non-operating state of the nuclear reactor in which the tubes 32 are at a substantially lower temperature such as room temperature, for example suitably quantified as a temperature of less than 100° C. The balancing of the stresses over the operational and non-operational range is achieved by pre-stressing the tubes 32 to be in axial tension at room temperature (e.g., at less than 100° C. in some embodiments), so that the differential thermal expansion between the Inconel™ 690 steam generator tubes 32 and the steel of the central riser 36 and vessel 14 causes the tubes to transition from axial tension to axial compression as the temperature is raised to the operating temperature, e.g. at least 500° C. in some embodiments. These differential thermal stresses among components 14, 32, and 36 are set up by common connection of the components at the tubesheet supports 90, 92 is also optionally reduced by having the feedwater nozzle 50 positioned low in the pressure vessel leaving a longer steam outlet annulus 62 to blanket the vessel with high temperature outlet steam, and by reducing axial length of the feedwater annulus 60 by employing a larger radius for the feedwater annulus 60. With brief reference to FIG. 7, a manufacturing sequence to prestress the tubes 32 to place them into axial tension is further described. In an operation 100, the tubes 32 are mounted in the tubesheets 90, 92 of the OTSG frame or support by expanding the tube ends to secure them to the tubesheets 90, 92. A consequential operation 102 is that this imparts axial tension to the tubes 32. In an operation 104, the OTSG 30 including the prestressed tubes 32 is installed in the pressure vessel 12, 14, 16 to construct the integral PWR of FIGS. 1-5. In an operation 106, the integral PWR is started up and brought to its operating state which has the effect of raising the temperature the primary coolant flowing in the tubes 32 of the OTSG 30 to an operating temperature of (in the illustrative example) at least 500° C. A consequential operation 108 is that this imparts axial compression to the tubes 32 due to the relatively higher coefficient of thermal expansion of the austenitic nickel-chromium-based alloy of the tubes 32 as compared with the steel of the central riser 36 and vessel 14 connected via tubesheets 90, 92. In some embodiments, in the operating state the OTSG 30 defines an integral economizer that heats feedwater injected into the fluid flow volume at the feedwater inlet 50 to a temperature at or below a boiling point of the feedwater. In such embodiments, the straight-tube OTSG 30 is an integral economizer (IEOTSG) design since the feedwater is heated by flow outside of the tubes 32. Feedwater enters through the feedwater nozzles 50, distributes throughout the feedwater annulus 60, and enters the tubes 32 via a gap or other passage (not shown) between the bundle shroud 40 and the lower tubesheet 92. In the operating mode, feedwater flows outside of the tubes 32 and there is forced convection heat transfer from the primary coolant flow to the feedwater flow followed by subcooled and saturated boiling to form the steam flow. Once the critical heat flux is reached at approximately 95% steam quality, the steam goes through a transition to stable film boiling followed by dryout at 100% steam quality. Thereafter in the tube bundle, the forced convection to steam raises the temperature to superheated conditions at which the steam exits the steam generator via the steam outlet annulus 62 and the steam outlet nozzle 52. The superheated steam does not require moisture separators before the steam is delivered to the steam turbine (although it is contemplated to include moisture separators in some embodiments). Some further aspects of the integral pressurizer are next described. The pressurizer controls the pressure of the primary coolant via the pressurizer heaters 82 and the pressurizer spray nozzles 84. To increase system pressure, the heaters 82 are turned on by a reactor control system (not shown). To decrease pressure, the spray nozzles 84 inject cold leg water from the top of the reactor inlet annulus 18 on the discharge side of the reactor coolant pumps 26 via a small external line (not shown). The pressurizer volume 80 is formed by a divider plate 70 which separates the space between the primary inlet plenum 74 and the pressurizer volume 80. The divider plate 70 optionally also serves as a flow diverter by including a perforated cylinder 124 (FIG. 8, top of divider plate not shown) or a cone shaped flow diverter surface 72 (FIG. 2) or other curved or slanted surface which aids in the turning of the flow in the primary coolant in the inlet plenum 74 before it enters the upper ends of the tubes 32 of the OTSG 30 setting up downward flow inside the tubes 30. The illustrative pressurizer including the pressurizer volume 80 and pressure control structures 82, 84 advantageously is a fully integral pressurizer (that is, is part of the pressure vessel 12, 14, 16) and advantageously has no pass-throughs for external CRDM's or other components. The central riser 36 forms a path 38 for the primary coolant flow leaving the reactor core 10 to reach the primary inlet plenum 74 of the steam generator 30. In this embodiment there is no horizontal run of piping for this purpose. As a result, if the reactor is operated in a natural circulation mode with the reactor coolant pumps 26 turned off (as may occur during a malfunction or loss of electrical power causing the pumps 26 to stop operating), the hot rising primary coolant is only impeded by the upper internals (e.g., the CRDM's 24). This flow resistance is not large compared to the flow resistance of the care 10 and the steam generator tubes 32 because the flow area is relatively large. The flow resistance of the central riser 36 is also a relatively small percentage of the total because of the large diameter of the path 38. The large inlet 2608 and outlet 2610 of the coolant pumps 26 further reduces flow resistance. In some existing nuclear steam supply systems, after a loss of coolant accident (LOCA) steam and non-condensable gases can collect at the high points of the reactor coolant pipes, and can inhibit the natural circulation loop between the reactor core and the steam generators. Advantageously, the straight-tube OTSG 30 with integral pressurizer volume 80 disclosed herein automatically removes non-condensable gases from the primary coolant circulation loop since there is only one high point at the top of the pressurizer volume 80. Buoyancy causes the non-condensable gases and vapor to go to the top of the pressurizer volume 80, where these gases and vapor do not interfere with the natural circulation loop. Another advantage of the disclosed straight-tube OTSG 30 is that it can optionally operate in multiple modes to remove decay heat from the reactor core 10. Starting with the normal operating state, if the reactor coolant pumps 26 stop operating, then the primary coolant water continues to circulate, albeit now via natural circulation, through the core 10 and through the steam generator tubes 32. As long as there is feedwater supplied to the inlet 50 of the steam generator 30, there is a large tube surface area to remove radioactive fission product heat from the core 10. If the primary coolant level falls below the level of the primary inlet plenum 74 during a LOCA, then the straight-tube OTSG 30 can operate as a condenser. In this mode, steam from boiling water in the reactor core 10 rises to fill the primary inlet plenum 74 and the pressurizer volume 80. The lower temperature water and steam on the secondary side (that is, in the fluid flow volume defined outside the tubes 32 by the shrouds 40, 42) causes condensation inside the steam generator tubes 30. By gravity alone, the condensate flows out of the steam generator tubes 32 into the primary outlet plenum 44 where it is returned to the core 10. In the straight-tube OTSG 30, the primary coolant pressure is inside the tubes 32. The primary coolant is at a substantially higher pressure than the secondary coolant flowing through the fluid flow volume defined outside the tubes 32 by the shrouds 40, 42. In some embodiments, in the operating state of the nuclear reactor the primary coolant flowing inside the tubes 32 is at a pressure that is at least twice a pressure of the secondary fluid (feedwater or steam) in the fluid flow volume. This enables the use of a thinner tube wall in tension. In contrast, if the primary coolant flows outside the tubes then the tube is in compression and a thicker tube wall is generally required. Some analyses have indicated that the tube wall in the tension design of the present OTSG embodiments can be made about one-half as thick as the tube wall thickness required for tubes placed in compression (for comparable tube diameter). The use of thinner tube walls translates into the OTSG 30 being substantially lighter and including substantially less Inconel™ 690 or other nickel-chromium-based alloy material used for the tubes 32. The weight saving of the straight-tube OTSG 30 is advantageous for an integral nuclear reactor. For example, in the illustrative embodiment of FIGS. 1-3, during refueling the core 10 is accessed by removing the steam generator 30. This entails disconnecting the OTSG 30 from the lower pressure vessel portion 12 via the mid-flange 16. The lightweight straight-tube OTSG 30 advantageously reduces the requisite size of the containment structure crane used for lifting the steam generator 30 off to the side during refueling. Advantageously, the illustrative internal primary coolant pumps 26 are located peripherally in the reactor inlet annulus 18 surrounding the reactor core 10, and so the pumps 26 do not need to be removed to perform the refueling. The straight-tube OTSG 30 also has service and maintenance advantages. Manways are readily provided proximate to the pressurizer volume 80 and the primary inlet plenum 74 to provide service access. Inspection of the tubes 32 can be performed during a plant outage via the primary inlet plenum 74 without removing the steam generator 30 from the pressure vessel. Eddy current inspection thusly performed can reveal tube thinning and tube cracks. If tube plugging is indicated by such inspection, the steam generator 30 can be removed during the outage and tube plugs can be installed at the lower tubesheet 92 and the upper tubesheet 90. In another approach, both tube inspection and tube plugging can be done during refueling when the steam generator 30 is placed off to the side of the reactor. In this case, there is easy access from the bottom for tube inspection and plugging. With reference back to FIG. 4, in the illustrative embodiment the steam generator 30 is a OTSG with straight tubes 32, and primary coolant Fprimary flows inside the tubes 32 of the steam generator 30 while the secondary coolant (water Fsecondary or steam Ssecondary) flows in the opposite direction outside of the tubes 32. As discussed herein, there are substantial advantages to this configuration. However, it is also contemplated to employ a flow configuration in which the primary coolant flows outside the tubes and the secondary coolant flows inside the tubes. Since the internal primary coolant pumps 26 are not coupled with the steam generator 30, using such a variant flow configuration would not entail modification of the pumps 26 or their location in the pressure vessel. Moreover, it is additionally or alternatively contemplated to replace the vertical tubes 32 of the illustrative OTSG 30 with slanted tubes, or with one or more helical tubes forming a helix around the central riser 36, or so forth. With reference to FIG. 8, a variant embodiment is described. This variant embodiment includes the IEOTSG 30 with tubes 32 mounted in upper and lower tubesheets 90, 92. In this variant embodiment, a modified upper pressure vessel portion 14′ differs from the upper pressure vessel portion 14 in that it does not have a larger diameter to provide a feedwater annulus with larger outer diameter as compared with the steam annulus. Rather, a feedwater annulus 60′ connected with the feedwater inlet 50 in the variant embodiment of FIG. 8 is of the same outer diameter as the steam annulus 62 that is connected with the steam outlet 52. The modified upper pressure vessel portion 14′ also differs from the upper pressure vessel portion 14 in that it does not include the integral sealing top portion 78. Rather, a separate sealing top portion 78′ is provided which is secured to the modified upper pressure vessel portion 14′ by an upper flange 120. Still further, the variant embodiment also does not include an integral pressurizer volume or the diverter plate 70. Rather, the sealing top portion 78′ defines a modified primary inlet plenum 74′ (but does not define a pressurizer volume), and the sealing top portion 78′ includes a curved surface 122 that cooperates with cylinder openings 124 at the top of the central riser 36 to perform the primary coolant flow diversion functionality. As the pressurizer volume 80 of the embodiment of FIGS. 1-3 is omitted in the variant embodiment of FIG. 8, primary coolant pressurization for the embodiment of FIG. 8 is suitably provided by self-pressurization. In this approach, steam vapor from the reactor core collects at the top of the steam generator vessel, that is, in the primary inlet plenum 74′. The compressibility of the vapor filled dome volume 74′ regulates the primary coolant pressure. To increase power, the feedwater flow into the feedwater inlet 50 is increased which increases the boiling lengths in the tubes 32. The reactor core 10 follows the load demand by increasing power via a negative moderator coefficient of reactivity due to the reduction in core inlet temperature from the steam generator 30. The core outlet temperature is maintained at a near constant temperature regulated by the pressure and saturation temperature of the steam dome volume 74′. Accordingly, for an increase in power, the temperature rise across the reactor core 10 increases while the reactor flow rate remains constant as determined by the reactor coolant pumps 26. Decreasing power employs analogous processes. As another illustrative variation (not shown), the tubes of the OTSG can be placed in different locations within the pressure vessel. In the illustrative embodiments of FIGS. 1-3 and 8, the OTSG 30 including tubes 32 is disposed entirely in the downcomer volume 34. More generally, however, tubes may be disposed in the downcomer volume (as illustrated), or in the central riser flow path 38 inside the central riser 36, or in both volumes 34, 36. As other illustrative variations, it has already been noted that the separate inner shroud 42 may instead be embodied as an outer surface of the central riser 36, and/or for the separate outer shroud 40 may instead be embodied as an inner surface of the upper portion 14 of the pressure vessel. Additionally, it is contemplated to integrate the lower tubesheet 92 with the mid-flange 16. With reference back to FIG. 5, another disclosed aspect is the use of a vertically staggered arrangement of the individual CRDM units 24. In illustrative FIG. 5, CRDM units are variously located with tops at different vertical levels L1-L4, with no two immediately neighboring CRDM units at the same vertical level. Another illustrative example, in which the CRDM units are staggered between only two vertical levels, is shown in application Ser. No. 12/722,696 filed Mar. 12, 2010 and titled “Control Rod Drive Mechanism for Nuclear Reactor”. Another way of stating this is that the CRDM units are staggered at two or more different distances from the nuclear reactor core such that no two neighboring CRDM units are at the same distance from the nuclear reactor core. In general, by staggering the CRDM units so that any two adjacent or neighboring CRDM units are not at the same vertical level or height, a more compact CRDM unit array is achieved as compared with conventional arrangements in which all CRDM units are at the same vertical level or height. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	description	This application is a continuation of U.S. application Ser. No. 11/779,899, filed Jul. 19, 2007, now U.S. Pat. No. 7,605,364 the contents of which are incorporated herein by reference. The present application claims priority from Japanese application JP 2006-253718 filed on Sep. 20, 2006, the content of which is hereby incorporated by reference into this application. The present invention relates to an electron scanning microscope for carrying out dimensional measurement of a micro-pattern formed on a semiconductor substrate, a method of evaluating a resolution of a scanning electron microscope, and a sample for valuating a resolution of a scanning electron microscope, and in particular to a scanning electron microscope incorporating a function of evaluating a resolution of a scanning electron microscope from a picked-up image. In a semiconductor manufacturing process, there have been demanded apparatuses for measuring dimensions with a higher degree of accuracy as the micro-patterns have been more and more fine. There has been known a scanning electro microscope for measuring a pattern width (a length measuring SEM (scanning electron microscope), or a CD (critical dimension) SEM), which are capable of picking up an image thereof with a magnification of one to five hundreds of thousands (100,000-500,000) as a dimension measuring tool for measurement of a micro-pattern having a size in the order of several ten nanometers. The demands for measuring accuracy of these apparatuses include not only enhancing the measuring accuracy of the individual apparatus but also reducing differences among measured dimensions of several apparatuses installed on a production line and as well as reducing variations in measured dimensions which are caused by aging (or deteriorating with age) of the apparatus. Of many factors for causing differences among measured dimensions of several apparatuses and for occurrence of variations in measured dimensions due to aging of the apparatuses, there may be exemplified differences and variation in resolution caused by differences among beam sizes and/or variation in the beam size due to aging. However, it is difficult to directly measure a size of an electron beam. Thus, in a scanning electron microscope, there has been used such a process that index values of resolution are measured from SEM images picked up by respective apparatuses, and differences among the beam sizes are evaluated by comparative evaluation of the index values. As a specific example of a technique for measuring a resolution, U.S. Pat. No. 6,545,275 (Patent Document 1) and Metrics of resolution and performance for CD-SEMs by D. C. Joy et al, Metrology, Inspection, and Process Control for Microlithography XIV, page 108 (Nonpatent Document 1) propose, as examples thereof, a method in which an image is picked up from a sample prepared by depositing gold particles on a silicon substrate, and frequencies are analyzed through Fourier transformation of the picked-up image in order to calculate an index value of resolution. Further, U.S. Pat. No. 5,969,273 (Patent Document 2) and Modeling and Experimental Aspects of Apparatus Beam Width as an Edge Resolution Measure, C. Archie et al, Metrology, Inspection, and Process Control for Microlithography XIII, page 669 (Nonpatent document 2) propose such a technique that an image is picked up from a pattern formed on a substrate so as to measure a width corresponding to a pattern edge part in order to calculate an index value of resolution. Furthermore, JP-A-2005-268231 (Patent Document 3) and Contrast-to-gradient method for the evaluation of image resolution taking account of random noise in scanning electron microscopy, T. Ishitani et. al, J. Electron Microscopy 53(3) page 245 (Nonpatent Document 3) propose such a technique that a plurality of partial resolutions is obtained from respective partial zones in a picked-up image, and an average of partial resolutions over the entire image is calculated in order to calculate an index value of resolution. In a scanning electron microscope apparatus for measuring dimensions of a pattern, a conventional resolution measuring process in which a picked-up image is used comprises the steps of (A-1) acquiring a picked-up image from a sample which is silicon substrate deposited thereon with gold or a porous silicon substrate, and (B-1) subjecting the picked-up image to Fast Fourier Transformation in order to analyze frequencies so as to calculate an index value of resolution. Further, another conventional resolution measuring process comprises the steps of (A-2) acquiring an image picked up from a pattern formed on a substrate, and (B-2) measuring a width corresponding to an edge part of the pattern from the picked-up image so as to calculate an index value of resolution. A secondary electron image obtained by the scanning electron microscope, is in general exhibited by a convolution integration of a f(x, y) and g(s,t), where f(x,y) is a signal determined by a material of a sample and a pattern shape, and g(s,t) is a shape of an electron beam irradiated onto the sample. That is, in order to measure a size of an electron beam from a secondary electron image, it is required to take into consideration an influence caused by the signal f (x, y) which is determined by a sample and which is included in the image. Estimating that the measurement of a resolution is carried out with the use of a dedicated sample, it is desirable for the sample to have one and the same pattern, one and the same pattern sectional shape and one and the same pattern distribution everywhere on the sample, even though it is not required to consider a variation in the signal due to a material quality. However, it is impossible to prepare such a sample, and the following problem will be caused. Since the sample used in (A-1) has such a feature that analogous patterns each having a size of several ten nanometers are distributed in random over the entire surface of the sample, if an image having many patterns can be obtained, it is expected to calculate an index value of resolution with respect to an averaged value of the signals (x, y). However, if the distribution densities, the averaged sizes or pattern sizes of the analogous patterns, are uneven, or if the sectional shapes vary thereamong, the averaged value of the signals (x,y) will be of course, changed, and accordingly, the sample should be prepared by controlling these items in order to decrease the dependency upon the individual sample characteristics. Similarly, even with a pattern formed on a substrate used in (A-2), the signals f(x,y) are different from one another, and accordingly, the index value of resolution depends upon the individual sample characteristics. Further, even with respect to a resolution evaluating algorithm, in the technique used in (B-1), if a pattern distribution on a sample becomes different, an index value of resolution obtained by calculation has a different characteristic and therefore dependency upon a pattern does not become small. In the technique used in (B-2), if a pattern roughness becomes different in an image zone used for calculation of a resolution, the index value of resolution will change, and accordingly, the pattern dependency is also not negligible. In view of the above-mentioned matters regarding the resolution monitor, a resolution problem inevitably has such a task that the preparation of a sample and the utilization of a measurement algorithm, which are capable of reducing the pattern dependency of an index value of resolution to be measured are required for precisely measuring a variation in size of an electron beam. An object of the present invention is to provide a sample for evaluation of a resolution of a scanning electron microscope, which is capable of stably evaluating a resolution with a high degree of sensitivity, a method of evaluating a resolution of a scanning electron microscope, and a scanning electron microscope. According to the present invention, there is provided a sample with a pattern having a sectional shape which is appropriate for a resolution monitor. The shape appropriate for the resolution monitor is specifically such that a pattern on the sample has a side wall which is inclined so as to prevent an irradiated primary electron beam from impinging on the side wall. Accordingly, there can be provided a pattern monitor which is independent from a sectional shape of a pattern. Further, according to the present invention, there is provided a resolution measuring method for evaluating a resolution with the use of an algorithm, by obtaining partial resolutions from partial zones in a picked-up image from the above-mentioned sample in order to calculate averaged partial resolution over the entire image so as to calculate an index value of resolution, as disclosed in the patent document 3. Thus, it is possible to provide a resolution monitor which can hardly be subjected to an affection caused by a pattern distribution, a pattern configuration and a sectional shape of a pattern. Further, with the use of the above-mentioned method, according to the present invention, there is provided a scanning electron microscope capable of managing a resolution. That is, according to the present invention, a sample for evaluating unevenness in resolution among several scanning electron microscopes or a sample for evaluating an aging in resolution of a designated scanning electron microscope is characterized in that a concave and convex pattern is formed on its outer surface, having a backward tapered sectional shape such as to have an upper part and a lower part which is narrower than the upper part. Further, according to the present invention, a sample for evaluating unevenness in resolution among several scanning electron microscopes or a sample for evaluating an aging in resolution of a designated scanning electron microscope is characterized in that a concave and convex pattern is formed on its outer surface, the concave and convex pattern being formed so that a side wall surface part on the concave and convex pattern falls in a shadow of the upper surface of the concave and convex pattern with respect to an electron beam perpendicularly incident upon the sample. Further, according to the present invention, there is provided a method for evaluating a resolution of a scanning electron microscope, characterized in that an image of the above-mentioned sample for evaluating a resolution of a scanning electron microscope is picked up successively by several scanning electron microscopes, and respective images picked up by these several scanning electron microscopes are processed in order to evaluate unevenness in resolution among the several scanning electron microscopes. Further, according to the present invention, there is provided a method for evaluating a scanning electron microscope, characterized in that image of the above-mentioned sample for evaluating a resolution of a scanning electron microscope is picked up, and image picked up is compared with data which has been stored in a storage means so as to evaluate an aging of resolution of the scanning electron microscope. Further, according to the present invention, there is provided a scanning electron microscope characterized in that the above-mentioned sample for evaluating a resolution is installed therein. Further, according to the present invention, there is provided a scanning electron microscope characterized by a function for correcting an image obtained by observing a sample to be observed or data obtained by processing the image, in accordance with a result of evaluation of a resolution with the use of the above-mentioned sample for evaluating a resolution. According to the present invention, the scanning electron microscope with the use of a sample for evaluating a resolution can measure a resolution with a higher degree of accuracy, and as a result, a variation in resolution of a scanning electron microscope and differences in resolution among scanning electron microscopes can be managed with a high degree of accuracy. Thus, it is possible to measures dimensions with a higher degree of reliability and a high degree of accuracy during, for example, a process of manufacturing a semiconductor pattern, resulting in an improvement in performance of a product and in an improvement in the yield thereof. These and other objects, features and advantages of the invention will be apparent from the following more particular description of preferred embodiments of the invention, as illustrated in the accompanying drawings. Explanation will be hereinbelow made of an embodiment of the present invention with reference to the accompanying drawings. In this embodiment, an image of a sample for evaluating a resolution, which has been obtained by a scanning electron microscope, is evaluated in order to evaluate and manage a resolution of the above microscope. (0) Sequence Referring to FIG. 1 which shows a sequence for evaluating a resolution, according to the present invention, the sequence comprises, at first, a step 0101 of setting a sample (B) for evaluating a resolution, onto a scanning electron microscope (A) whose resolution is to be evaluated, a step 0102 of picking up an image (C) of the sample (B) for evaluating a resolution, a step 0103 of calculating an index value (D) of resolution of the microscope from the picked-up image (C), a step 0104 of then storing the thus calculated index value (D) of resolution for each microscope and for each time series in order to monitor a condition of the scanning electron microscope, a step 0105 of removing the sample (B) for evaluating a resolution from the scanning electron microscope (A), and a step 0106 of correcting the scanning electron microscope (A) and the picked-up image (C) for a variation (F) in resolution in accordance with a result of the monitor of the condition of the scanning electron microscope which monitor was carried out at the step 0104 if the index value (D) of resolution becomes out of a preset range. The sample (B) for evaluating a resolution, which is set at the step 0101, may be beforehand set on the scanning electron microscope (A). Alternatively, the sample (B) for evaluating a resolution, may have been beforehand held at a predetermined position in or outside of the scanning electron microscope (A), and may be automatically set in the scanning electron microscope (A). Next, explanation will be hereinbelow made of the respective steps of the above-mentioned sequence in detail. (A) Scanning Electron Microscope Referring to FIG. 2 which shows a configuration of a scanning electron microscope (A) for measuring dimensions, according to the present invention, the scanning electron microscope is mainly composed of two portions, that is, an electron optical system 2000 for picking up or obtaining electron beam images, and a data processing system 2100 for processing the obtained images so as to measure an objective pattern. The electron optical system 2000 is mainly composed of a stage 0201 for carrying thereon a sample 0202, an electron source 0203 for emitting an electron beam 0208, a deflection lens 0204 for deflecting the electron beam 0208, an objective lens 0205 adapted to be controlled in order to pick up an image at a position of a focused point, a secondary electron detector 0206 having a function of converting secondary electrons produced from a sample, into an electric signal, an A/D converter 0207 for converting the detected electric signal into a digital signal, and a control portion 0211 for controlling the above-mentioned components. Meanwhile, the data processing system 2100 for measuring a picked-up pattern from image data is mainly composed of a process portion 0213 for processing an image and so forth, a storage portion 0212 for storing image data and various data adapted to be used for other processes, and an input/output portion 0214 having a function of allowing the user to input an image pick-up condition and parameters for the image process, the input/output portion 0214 also having a function of outputting an obtained result, among which portions data is delivered and received to and from one another through a data bus 0210. Further, the control portion 0211 shown in the figure, is adapted to carry out not only the control for the electron optical system but also control for measuring dimensions of a pattern from a picked-up image (the control portion 0211 is shown within the data processing system 2100 in the case of the configuration of the scanning electron microscope (A) according to the present invention, as shown in FIG. 2). This embodiment has a purpose of monitoring a variation in the shape of an electron beam, caused by an individual difference of the electron optical system, an aging and the like. Referring to FIG. 3 which shows a schematic view 0301 of a shape of an electron beam, the beam is diverged by a diverging angle α (0303) both forward and rearward of the incident direction of the electron beam, from a position where the beam is converged to the fullest, that is, a beam waist position 0302 as a mid point. It is assumed in this embodiment that the diverging angle α is not greater than about 1 degree. (B) Sample In this embodiment, a secondary electron signal image obtained by scanning the sample for evaluating a resolution by the above-mentioned electron beam over is analyzed in order to calculate an index value of resolution. The secondary electron signal image is the one which is an image of intensities of secondary electron signals emitted from the sample through the irradiation of a scanning electron beam over the sample. The secondary electron signals are in general represented by a convolution integral of two functions f and g, where f is a signal f(x,y) determined by a configuration and a material quality of a pattern on the sample, and g is an electron beam g(s,t). Referring to FIGS. 4A to 4D which schematically show the respective relationships between various sectional shapes of a pattern and a secondary electron signal, the intensity f of the secondary electron signal 0402 emitted from the pattern 0401 depends upon a relative angle β between the incident direction of the beam and the side wall of the pattern. The relative angle β 0403 is set to be positive 0404 in such a case that the electron beam is directly irradiated upon the side wall of the pattern, but to be negative 0405 in such a case that it is not directly irradiated upon the side wall. The relative angle β can be taken in an angle range from −90 to +90 deg., and it conceived that the following relationship (Formula 1) is satisfied in a range in which the relative angle β is greater than 0 deg. but not greater than 90 deg:f∝1/cos(β) Formula 1 Further, as understood from the above-mentioned relationship (Formula 1), a secondary electron signal having a relatively high intensity 0406 is emitted from the edge part of the pattern due to the so-called edge effect. In view of the above-mentioned principles, the intensities of the secondary electron signals are schematically shown in FIGS. 5A to 5E, with respect to various sectional shapes. The pattern on the sample has a top surface and a bottom surface. Assuming that both of the surfaces cross the incident direction of the electron beam, at a substantially right angle thereto, in such a case that the electron beam incident perpendicular to the sample is not made into contact with the side wall of the pattern, as denoted by 0501 and 0503 shown in FIGS. 5A and 5B, that is, the relative angle β between the incident direction (vertical direction) of the electron beam and the side wall of the pattern is not greater than 0 deg. (the width thereof is larger in the upper part (on the outer surface side) than in the lower part (on the base side) of the pattern in its sectional shape, that is, the sectional shape of the pattern is backward tapered), the signal f obtained by detecting the secondary electrons has a waveform which is enhanced only in its edge effect, as indicated by 0502 and 0504 shown in FIGS. 5A and 5B, that is, which does not depend upon the relative angle β. On the contrary, it can be understood that the secondary electron signal f is changed as indicated by 0508 in FIGS. 5D and 0510 in FIG. 5E, when the inclined angle of the side wall of the pattern varies so as to change the relative angle β between the side wall of the pattern and the incident direction of the electron beam, in view of the above-mentioned formula 0406 in such a case that the electron beam which is incident upon the sample, perpendicular thereto, is irradiated upon the side wall of the pattern, as indicated by 0507 in FIGS. 5D and 0509 in FIG. 5E, (the width is greater in the lower part than in the upper part of the pattern in its sectional shape, that is, the pattern has a forward tapered sectional shape), that is, the relative angle β between the incident direction (vertical direction) of the electron beam and the side wall of the pattern is positive. As stated above, since the image of the secondary electron signal is changed not only by the shape of the electron beam but also by the sectional shape of the pattern, there is raised such a task that variation in an index value of resolution in dependence upon a sectional shape of a pattern should be decreased whenever an index value of resolution is calculated from the picked-up image. By the way, it is extremely difficult to practically form a pattern having one and the same sectional shape everywhere on the sample. Accordingly, on the basis of the result shown in FIGS. 5A to 5D, there will be taken, for a sample producing a secondary electron image which can hardly be changed even though the sectional shape of the pattern on the sample is uneven, a sectional shape having such a side wall angle that the electron beam is prevented from irradiating upon the side wall of the pattern as possible as it can, that is, the relative angle β between the incident direction of the electron beam and the side wall of the pattern is not greater than 0 deg., as indicated by 0501 in FIGS. 5A and 0503 in FIG. 5B. Thus, with such a sectional shape that the top surface of the sample casts a shadow on the side wall surface thereof with respect to the electron beam which is incident upon the top surface of the sample, perpendicular thereto. By this sectional shape, the electron beam incident upon the sample will be prevented from being incident directly upon the side wall surface of the sample, thereby it is possible to detect a secondary electron signal which excludes any data concerning the side wall surface. Actually, the incident direction of the electron beam is inclined to the vertical direction or has a diverging angle as shown in FIG. 3, and accordingly, it is required to select a side wall angle of the pattern in view of this matter. For example, as shown in FIG. 6, if the beam waist 0302 of the electron beam 0301 which is incident upon the sample in the vertical direction is adjusted to a position in the vicinity of the top surface of the pattern 0601, a diverging angle α is produced in the part forward from the beam waist 0302 (said part is on the base side of the pattern 0601). Accordingly, in order to accurately evaluate a resolution with the use of the electron beam having such a diverging angle α, it had better to use a sample having a side wall which is inclined inward by an angle of not less than α with respect to the vertical direction. With the above-mentioned condition being satisfied, the unevenness of the secondary electron signals thus obtained can be reduced even though the sectional shapes of the pattern is uneven, thereby it is possible to calculate the index value of resolution with the reduced affection by the sectional shape of the pattern. As a sample which can satisfy the above-mentioned condition, there may be used, for example, a sample having an etching pattern which depicts lines, trenches, dots, holes or the like which has an arbitrary shape. FIG. 7 shows an example of a process for forming a sample having a side wall inclined inward of a pattern as shown in FIG. 6. That is, in a process before a manufacture of a semiconductor, a film 0701 made of a material having a high etching rate is formed on an Si substrate 0700 and a thin film 0702 made of a material having a low etching rate and having a thickness of about 10 nm is formed thereon. Next, a resist (which is not shown) is applied over the thus laminated film, and then, a pattern is exposed thereon, and developed. Then, the thus obtained resist pattern (which is not shown) is used as a mask for etching in order to form a pattern as shown in FIG. 7. Due to different etching rates, the lower film 0701 is greatly etched, and accordingly, a sample having a side wall inclined inward of the pattern can be provided. As an indicator of a pattern sectional shape, in addition to the inclined angle of the side wall, there may be exemplified a corner rounding (fitting) 1001 of a lower part of the pattern and a corner rounding (top rounding) of the upper part of the pattern, as shown in FIG. 10. If the fitting 1001 is present, the secondary electron signal is changed, depending upon a shape of the fitting, and accordingly, it is desirable that the fitting 1001 is as small as possible. Alternatively, as shown in FIG. 11, if the sample has such a sectional shape that the height of the pattern 1101 is sufficiently high, and accordingly, the secondary electrons 1103 emitted from the fitting part 1102 cannot reach a secondary electron detector 206, no affection by the fitting 1102 is exhibited in the secondary electron image, thereby it is possible to reduce the shape dependency of the sample. The secondary electron signal is also changed in dependence upon a shape of the top rounding 1002 shown in FIG. 10, and accordingly, it is desirable that the top rounding 1062 is small as possible and the top surface of the pattern is flat. However, as shown in FIG. 12, since the intensity of the secondary electron beam is strong in the part where the top rounding 1002 is produced, due to the above-mentioned edge effect, and can be hardly subjected to affection by a shape change in comparison with secondary electrons produced from the fitting 1001 or the fitting 1102. In particular, it may considered that a zone where a secondary electron signal having a strong intensity can be obtained due the edge effect, is substantially coincide with an electron scattering area 1201 in a pattern around the top of the edge as a center, and accordingly, even though the top rounding 1002 is present in this zone, it may be considered that the secondary electron signal cannot be easily changed. Thus, if the electron scattering zone in the pattern has a circular shape having a radius R, it is desirable that the top rounding is smaller than an arc having the radius R. The electron scattering zone in the pattern varies, depending upon a material quality of the sample, optical conditions (an acceleration voltage, a probe current and the like) of the electron beam and the like, within a range of radius from several nanometers to several ten nanometers. C. Image Pick-Up In order to evaluate a resolution with the use of the sample as mentioned above, with several repetitions, for one and the same zone, of such a process that a desired zone on the sample is scanned by the electron beam so as to obtain a secondary electron detection signal for one frame in order to obtain a second electron detection signals for several frames, the secondary electron signals for the several frames are added together (addition of frames) so as to obtain an image for evaluating a resolution. Since the detectability is high in the case of a condition that the size of one pixel being smaller than the size of the electron beam, it is desirable to pick up an image with a magnification which can satisfy the above-mentioned condition in order to detect a variation in the size of the electron beam. Although errors during measurement of a resolution caused by unevenness of sectional shapes of patterns, can be decreased to a certain degree with the use of the above-mentioned sample, there would be still remained affections by unevenness in pattern shape, pattern sectional shape and pattern distribution, and accordingly, in order to reduce such affections, index values of resolution are calculated N number of images, which have been obtained at N number of places on one and the same sample, and then, the evaluation of resolution is carried out by using an averaged value of the thus obtained index values of resolution. The required number N of picked-up images varies depending upon a required degree of accuracy for measuring a resolution. As an example, there may be used such a technique that the number N of picked-up images is selected so that a value which is obtained by dividing a degree V of unevenness in measurement for index values of resolution on a sample to be used, with a square root of N, is smaller than a required degree S. This technique has been known as a center limit theorem. The manner for determining a zone from which images at N places are picked up from a sample, may be random. Alternatively, as shown in FIG. 8, the N places at which images are picked up from a sample 0801 can be taken as denoted by reference numeral 0803 in which they are adjacent to one another for each evaluation of resolution, thereby it is possible to compare results of evaluation of resolution in adjacent patterns on the sample with each other. With the comparison between the adjacent patterns, even though a biased distribution of pattern shapes or sectional shapes is present within the sample, the measurements of resolution can be stably carried out without being affected by such biased distribution. D. Algorithm for Calculating an Index Value of Resolution Explanation will be hereinbelow made of an algorithm for calculating an index value of resolution from images for evaluating a resolution, which are picked up as stated above. As to the resolution calculating algorithm, there may be in general known several techniques, as exemplified in Nonpatent Documents 1 to 3, and any of these techniques may be used. However, in this embodiment, a CG (Contrast to Gradient) process is used. The CG process utilizes such a technique that index values of local resolution are calculated from local zones in an image, and a weighted average of index values of local resolution is obtained over the entire image as an index value of resolution. In this technique, since the index values are calculated from local zones at the first step, it is possible to decrease measurement errors for resolution caused by different pattern shapes and different pattern distributions on the sample. As stated above in the item C: Image Pick-Up, index values of resolution for N number of images are obtained, and an average thereof is then obtained as an index value of resolution for a scanning electron microscope to be evaluated, thereby it is possible to measure a resolution with less affection by a pattern. E: Monitor of Condition of Electron Microscope Referring to FIG. 9 which shows an example of a GUI of a system for monitoring a condition of the electron microscope with the use of calculated index values of resolution, the GUI includes a portion 0901 for comparing and displaying index values of resolution among electron microscopes, which have been measured on a day or during a period which is designated by a selection button 0910 on a screen, and a portion 0902 for displaying an aging of an index value of resolution for a single electron microscope designated by a designation button 0911 on the screen. Further, if each index value becomes out of a preset range 0903 or 0903′, or if a degree 0904 of unevenness in index value of resolution among electron microscopes or an aging 0905 becomes greater than a preset value, it is possible to issue an alarm for warning 0906. By monitoring a condition of the electron microscope with the use of the index values of resolution, the performance of the monitored electron microscope can be ensured, thereby it is possible to obtain a result of dimensional measurement with a high degree of reliability. F. Correction Finally, explanation will be made of a function of correcting electron microscopes, images or pattern dimensions measured from the images on the basis of the thus obtained index values of resolution. In order to correct the electron microscope, there may be used such a method that adjustment for microscope parameters and measurement of an index value of resolution are repeated until the index value of resolution becomes a desired value. As the microscope parameters, there may be exemplified an acceleration voltage, a parameter for adjusting aberration of an electron beam, a focus parameter, a parameter for adjusting an electromagnetic lens and the like. In addition, it is possible to adjust a parameter on the basis of data as to a relationship between a parameter adjusting value for adjusting a resolution and an index value of resolution which relationship has been previously checked. These adjustments are carried out by the control portion 0211 shown in FIG. 2. In order to correct the image, there may be exemplified a method for using an image filter in order to set the index value of resolution to a desired value. As the above-mentioned image filter, a Gaussian function type filter having a shape similar to that of the electron beam may be used. The index value of such a filter is adjusted based upon the index value of resolution, and with the use of the thus adjusted filter, images are processed for convolution, deconvolution or the like, thereby it is possible to reform the images picked up by plurality of electron microscopes having various index values of resolution, into images which are as those picked up the electron microscopes having the same index value of resolution. By measuring pattern dimensions from these images, unevenness of measured dimensions caused by different resolutions can be reduced. This correction is carried out by the processing portion 0213 shown in FIG. 2. In the case of direct correction for pattern dimensions measured from images, there may be exemplified a method for adding offsets to measured dimensions. That is, the relationship between an index value of resolution and a result of measurement of dimensions of a typical pattern have been beforehand checked, and differences between measured dimensions having a target index value of resolution and those having a practical index value of resolution are used as offsets, thereby it is possible to reduce differences in the measured dimensions caused by differences in resolution. These corrections is carried out by the processing portion 0213 in FIG. 2. Further, the correction may also be made in the combination of the correction for an electron microscope, the correction for the image, and the direct correction for pattern dimensions measured from images, each correction procedure is above-mentioned respectively. The evaluation of resolution of a scanning electron microscope and the correction therefor can be made through the sequence shown in FIG. 1, as stated above. Thus, dimensional data of a pattern, which is individually measured by a plurality of scanning electron microscopes, can be compared with one another with a relatively high degree of reliability. Further; it is possible to maintain a uniform degree of reliability for the resolution of a scanning electron microscope over a long time. The invention may be embodied in other specific forms without departing from the spirit or essential characteristics thereof. The present embodiment is therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being indicated by the appended claims rather than by the foregoing description and all changes which come within the meaning and range of equivalency of the claims are therefore intended to be embraced therein.
	description	The present invention relates generally to radiographic imaging and, more particularly, to an x-ray filter having dynamically displaceable x-ray attenuating fluid. Typically, in radiographic systems, an x-ray source emits x-rays toward a subject or object, such as a patient or a piece of luggage. Hereinafter, the terms “subject” and “object” may be interchangeably used to describe anything capable of being imaged. The x-ray beam, after being attenuated by the subject, impinges upon an array of radiation detectors. The intensity of the radiation beam received at the detector array is typically dependent upon the attenuation of the x-rays through the scanned object. Each detector element of the detector array produces a separate signal indicative of the attenuated beam received by each detector element. The signals are transmitted to a data processing system for analysis and further processing which ultimately produces an image. Generally, the x-ray source and the detector array are rotated about the gantry within an imaging plane and around the subject. X-ray sources typically include x-ray tubes, which emit the x-ray beam at a focal point. X-ray detectors typically include a collimator for collimating x-ray beams received at the detector, a scintillator for converting x-rays to light energy adjacent the collimator, and photodiodes for receiving the light energy from the adjacent scintillator and producing electrical signals therefrom. In a similar fashion, radiation detectors are employed in emission imaging systems such as used in nuclear medicine (NM) gamma cameras and Positron Emission Tomography (PET) systems. In these systems, the source of radiation is no longer an x-ray source, rather it is a radiopharmaceutical introduced into the body being examined. In these systems each detector of the array produces a signal in relation to the localized intensity of the radiopharmaceutical concentration in the object. Similar to conventional x-ray imaging, the strength of the emission signal is also attenuated by the inter-lying body parts. Each detector element of the detector array produces a separate signal indicative of the emitted beam received by each detector element. The signals are transmitted to a data processing system for analysis and further processing which ultimately produces an image. In most computed tomography (CT) imaging systems, the x-ray source and the detector array are rotated about a gantry encompassing an imaging volume around the subject. X-ray sources typically include x-ray tubes, which emit the x-rays as a fan or cone beam from the anode focal point. X-ray detector assemblies typically include a collimator for reducing scattered x-ray photons from reaching the detector, a scintillator adjacent to the collimator for converting x-rays to light energy, and a photodiode adjacent to the scintillator for receiving the light energy and producing electrical signals therefrom. Typically, each scintillator of a scintillator array converts x-rays to light energy. Each photodiode detects the light energy and generates a corresponding electrical signal. The outputs of the photodiodes are then transmitted to the data acquisition system and then to the processing system for image reconstruction. For radiographic imaging, such as x-ray imaging and computed tomography, x-ray exposure of a subject is always a concern. The amount of irradiation seen by a scan subject is generally referenced as “x-ray dose” and is factor that is paramount in prescribing a scan. That is, image quality is greatly influenced by the x-ray dose during data acquisition. In this regard, at higher dose levels, SNR is greater, which leads to better image quality. At higher dosage levels, however, the subject is exposed to greater amounts of irradiation. As there are strict guidelines to the amount of irradiation that a subject can experience, practitioners must limit x-ray dose and, as a result, sacrifice image quality. So, when prescribing a scan, practitioners choose a dosage level that will provide the best image quality without exceeding mandated irradiation levels. Adding to the difficulty in setting dosage levels is that subjects, such as medical patients, lack a uniform thickness. This is particularly problematic for CT systems where each view in a complete scan rotation presents a different angle of x-ray illumination to the subject. As such, it is difficult to optimize x-ray dose on a view-by-view basis. Moreover, subjects generally have variable attenuation characteristics across a given field-of-view. More specifically and in the context of medical patients, outside the skin line, there is no attenuation and the full flux of the x-ray beam is incident on the x-ray detector. Just within the skin line, the attenuation is much larger relative to outside the skin line and, as a result, fewer x-rays reach the detector. Since the x-ray source during a scan is operated so that the number of x-ray photons within the center of the field-of-view is sufficient to create an image, e.g., low noise, the excessive photons at the skin line interface are unnecessary for image quality. Typically, without a proper x-ray management device, this increased photon number at the skin interface imports additional dose to the subject and results in x-ray scatter into the imaged region. Therefore, it has been desirable to reduce the x-ray flux outside the imaging volume and to do so with each view. This is increasingly desirable for detectors that saturate at low x-ray flux levels. That is, energy discriminatory and photon counting detectors have a much lower saturation limit than conventional, energy integrating detectors. Despite the drawbacks associated with low flux saturation, photon counting detectors are desirable in order to ascertain energy information of the x-rays detected by the detectors which can be used for material discrimination. However, because direct conversion, photon counting detectors typically having a low flux rate limit, such as about 1 million cnts/sec/mm2), their use has been significantly limited. Additionally, the concerns of increased x-ray photons are not isolated to the skin line interface. If the field-of-view is relatively uniform for all views, then abrupt x-ray flux changes will only be experienced at the edge of the field-of-view. However, for medical imaging, such a case is rare. Typically, the asymmetry of the object being imaged results in very different flux profiles for different views. The impact of this variance can be mitigated if the center of the field-of-view has the thickest cross-section and the field-of-view boundary is marked by relatively high flux transition. However, if the object to be imaged is off-centered, then one side of the detector will see much higher flux rates than other side of the detector and this will change on a per view basis. It is also possible to have a great degree of variability within the field-of-view if the internal composition of the object causes large variations in signal level across the field-of-view. One skilled in the art will appreciate that numerous flux conditions other than those presented above can be encountered and lead to detector saturation and/or areas of unneeded x-ray flux. Heretofore, fixed shaped filters have been used to selectively attenuate x-rays at the edges of the field-of-view more than at the center of the field-of-view. A common fixed shape filter is generally referenced as a “bowtie” filter and is commonly used in CT systems to compensate for the thickest-in-the-middle characteristic of most medical patients. Known bowtie filters, however, are fixed in their shape and, thus, to accommodate the variations that could be encountered in a scan subject population, CT systems are generally equipped with several discrete bowtie filters. Not only does this lead to increased costs, but because the filters are static, the filters cannot be optimized for dynamic changes that occur as the source-detector rotates during a CT scan. In this regard, known bowties are ineffective in preventing detector saturation across the several views of a CT scan and can result in increased dosage levels to maintain image quality. It would therefore be desirable to design an x-ray flux management device that is effective in reducing detector saturation under high x-ray flux conditions while not compromising data acquisition under low x-ray flux conditions. It would also be desirable to have such an x-ray flux management device that provides further optimization of radiation dose during a scan. The present invention is a directed an x-ray flux management device that overcomes the aforementioned drawbacks. Conventional CT imaging scanners utilize detectors that convert x-ray photon energy into current signals that are integrated over a time period, then measured and ultimately digitized. A drawback of such detectors is their inability to provide independent data or feedback as to the energy and incident flux rate of photons detected. That is, conventional CT detectors have a scintillator component and photodiode component wherein the scintillator component illuminates upon reception of x-ray photons and the photodiode detects illumination of the scintillator component, and provides an integrated electrical current signal as a function of the intensity and energy of incident x-ray photons. While it is generally recognized that CT imaging would not be a viable diagnostic imaging tool without the advancements achieved with conventional CT detector design, a drawback of these integrating detectors is their inability to provide energy discriminatory data or otherwise count the number and/or measure the energy of photons actually received by a given detector element or pixel. Accordingly, the present invention is particularly applicable with CT systems having energy discriminating detectors that can provide photon counting and/or energy discriminating feedback. Energy discriminating detectors are capable of not only x-ray counting, but also providing a measurement of the energy level of each x-ray detected. While a number of materials may be used in the construction of an energy discriminating detector, including scintillators and photodiodes, direct conversion detectors having an x-ray photoconductor, such as amorphous selenium or cadmium zinc telluride, that directly convert x-ray photons into an electric charge have been shown to be among the preferred materials. A drawback of photon counting detectors, however, is that these types of detectors have limited count rates and have difficulty covering the broad dynamic ranges encompassing very high x-ray photon flux rates typically encountered with conventional CT systems. Generally, a CT detector dynamic range of 1,000,000 to one is required to adequately handle the possible variations in photon flux rates. In the very fast scanners now available, it is not uncommon to encounter x-ray flux rates of over 108 photons/mm2/sec when no object is in the scan field, with the same detection system needing to count only 10's of photons that manage to traverse the center of large objects. The very high x-ray photon flux rates ultimately lead to detector saturation. That is, these detectors typically saturate at relatively low x-ray flux levels. This saturation can occur at detector locations wherein small subject thickness is interposed between the detector and the radiographic energy source or x-ray tube. It has been shown that these saturated regions correspond to paths of low subject thickness near or outside the width of the subject projected onto the detector array. In many instances, the subject is more or less cylindrical in the effect on attenuation of the x-ray flux and subsequent incident intensity to the detector array. In this case, the saturated regions represent two disjointed regions at extremes of the detector array. In other less typical, but not rare instances, saturation occurs at other locations and in more than two disjointed regions of the detector. In the case of a cylindrical subject, the saturation at the edges of the array can be reduced by the imposition of a bowtie filter between the subject and the x-ray source. Typically, the filter is constructed to match the shape of the subject in such a way as to equalize total attenuation, filter and subject, across the detector array. The flux incident to the detector is then relatively uniform across the array and does not result in saturation. What can be problematic, however, is that the bowtie filter may not be optimum given that a subject population is significantly less than uniform and not exactly cylindrical in shape nor centrally located in the x-ray beam. In such cases, it is possible for one or more disjointed regions of saturation to occur or conversely to over-filter the x-ray flux and unnecessarily create regions of very low flux. Low x-ray flux in the projection results in a reduction in information content which will ultimately contribute to unwanted noise in the reconstructed image of the subject. Generally, high-sensitivity photon counting radiation detectors are constructed to have a relatively low dynamic range. This is generally considered acceptable for proton counting detector applications since high flux conditions typically do not occur. In CT detector designs, low flux detector readings through the subject are typically accompanied by areas of high irradiation in air, and/or within the contours of the scan subject requiring CT detectors to have very large dynamic range responses. Moreover, the exact measurement of photons in these high-flux regions is less critical than that for low-flux areas where each photon contributes an integral part to the total collected photon statistics. Notwithstanding that the higher flux areas may be of less clinical or diagnostic value, images reconstructed with over-ranging or saturated detector channel data can be prone to artifacts. As such, the handling of high-flux conditions is also important. The present invention includes an x-ray flux management device designed to prevent saturation of photon counting x-ray systems having detector channels characterized by low dynamic range. Dynamic range of a detector channel defines the range of x-ray flux levels that the detector channel can handle to provide meaningful data at the low-flux end and not experience over-ranging or saturating at the high flux end. Notwithstanding the need to prevent over-ranging, to provide diagnostically valuable data, the handling of low-flux conditions, which commonly occur during imaging through thicker cross-sections and other areas of limited x-ray transmission, is also critical in detector design. As such, the x-ray flux management device described herein is designed to satisfy both high flux and low flux conditions. A bowtie filter is presented that can present various x-ray attenuation profiles during a scan. The filter is constructed to have a fluidic envelope filled with attenuating fluid and a displacement insert or “zombie.” This zombie is designed to displace the attenuating fluid to achieve a desired attenuating or filtering profile. In this regard, the zombie can be rotated, twisted, moved, and otherwise contorted within the fluidic envelope as needed during the course of a scan. As the position and shape of the zombie is changed, the x-ray profile of the filter changes. In one preferred embodiment, the zombie is constructed from viscoelastic or elastomeric material that has a memory. In this regard, the zombie is constructed to have a complex shape when at rest, but can have its shape changed when external forces are placed thereon. As such, in this preferred embodiment, the zombie is constructed to have a shape matched to that of a typical medical patient when at rest. As the x-ray filtering needs change during the course of the scan, the filter can be rotated, compressed, stretched, and/or contorted to achieve additional filtering profiles. Therefore, in accordance with one aspect of the present invention, an x-ray attenuating filter is presented that includes an envelope containing x-ray accumulating fluid. The filter further has a fluid displacement device disposed within the envelope and configured to displace the x-ray attenuating fluid. According to another aspect, the present invention includes a CT system having an x-ray source, an x-ray detector, and an x-ray filter assembly disposed between the x-ray detector and the x-ray source along a path of irradiation. The x-ray filter assembly has a bowtie filter having a body with accumulating fluid disposed therein as well as an attenuating fluid displacement device sealingly enclosed within the body of the bowtie filter. The filter assembly further has a mechanized actuator connected to the attenuating fluid displacement device to dynamically position the attenuating fluid displacement device within the body of the bowtie filter to define a desired filtering profile for the bowtie filter. According to yet a further aspect of the present invention, an x-ray filter has a fluidic envelope having x-ray attenuating fluid disposed therein. The attenuating fluid is designed to filter x-rays projected from an x-ray source for an object to be scanned. The x-ray filter further has means for displacing the x-ray attenuating fluid within the fluidic envelope to achieve a desired x-ray filtering profile to prevent detector saturation during scanning of the object In accordance with yet a further aspect of the invention, an x-ray filter is presented that includes a fluidic envelope having x-ray attenuating fluid disposed therein. The x-ray attenuating fluid includes at least one of liquid metal, nanoparticles suspended in a non-settling solution, or a compound with pH control buffer dissolved in a liquid. Various other features and advantages of the present invention will be made apparent from the following detailed description and the drawings. The operating environment of the present invention is described with respect to a four-slice computed tomography (CT) system. However, it will be appreciated by those skilled in the art that the present invention is equally applicable for use with single-slice or other multi-slice configurations. Moreover, the present invention will be described with respect to the detection and conversion of x-rays. However, one skilled in the art will further appreciate that the present invention is equally applicable for the detection and conversion of other high frequency electromagnetic energy. While the present invention is applicable with a number of radiographic imaging systems, it is particularly well-suited for CT systems and, especially, those systems having detectors with relative small dynamic range, such as photon counting and energy discriminating detectors. In this regard, the present invention is believed to be a key enabler for the use of direct conversion and energy discriminating/photon counting detectors with conventional CT systems. Additionally, the invention is believed to be effective in limiting radiation exposure without sacrificing image quality. Referring to FIGS. 1 and 2, an exemplary computed tomography (CT) imaging system 10 is shown as including a gantry 12 representative of a “third generation” CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 toward a detector array 18 on the opposite side of the gantry 12. Detector array 18 is formed by a plurality of detectors 20 which together sense the projected x-rays that pass through a medical patient 22. Each detector 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuated beam as it passes through the patient 22. In one preferred embodiment, each detector is capable of providing energy level and photon count information. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to an x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detectors 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has a keyboard. An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28, gantry motor controller 30, and filter controller 52. In addition, computer 36 operates a table motor controller 44 which controls a motorized table 46 to position patient 22 and gantry 12. Particularly, table 46 moves portions of patient 22 through a gantry opening 48. As will be described in greater detail below, system 10 further has an x-ray filter 50 that is positioned between x-ray source 14 and detector array 18. The filter 50 is constructed to define various filtering profiles during the course of a scan. In this regard, the filter is operationally connected to a filter controller 52 that controls a motor and actuator to effectuate changes in the angle, position, shape, and otherwise orientation of attenuating fluid disposed in the filter. Referring now to FIG. 3, a cross-sectional view of filter 50 is shown. As illustrated, filter 50 preferably is constructed to have a fluidic envelope 54 defined by a rigid body enclosing x-ray attenuating fluid 56. The attenuating fluid 56 surrounds an x-ray fluid attenuating insert or “zombie” 58. As will be described, the zombie 58 may be rotated and/or translated by a rotor assembly 60. The zombie is also connected to an armature 62 that, as will be described, can effectuate changes in the shape thereof. The zombie is preferably constructed to have an elastomeric or viscoelastic shell 64 that is filled with foam or air. The attenuating fluid is preferably a dense fluid, such as liquid mercury or other liquid metals, with low-melting temperatures. However, it is contemplated that the fluid may be a combination of several liquid metals. Additionally, the attenuating fluid may take the form of high density powders or nanoparticles suspended in a non-settling colloidal suspension. In this regard, it is contemplated that tungsten or similar powders may be used, such as tungsten oxide mixed with a paint-binder system. It is further recognized that high density salts and other compounds with pH control buffers to provide stability dissolved in water, oil, or other liquids may also be used. For example, alkali halide salts in water, such as potassium iodide or cesium iodide are believed to be particularly applicable. In one preferred embodiment, the attenuating fluid comprises Na2WO4 in a solution of water, oil, organic, or non-organic liquid. Molecular liquids such as hexaiodobenzene are also believed to be practical as well as lubricous nanoparticles/liquid composites with low steric hindrance to zombie rotation. Ferro-fluids such as colloidal suspension of iron oxide particles may also be used. As described above, filter 50 is constructed to have a fluidic envelope containing high density x-ray attenuating fluid that is displaced by a zombie. The zombie, as illustrated in FIG. 4, has a plastic or elastomeric shell 64 that is preferably filled with foam or air. The shell can be rotated via rotor shaft 60 by actuator 66. The actuator can effectuate rotation and/or translation of the zombie within the fluidic envelope. The zombie further has an armature 62 that is connected to zombie face plate 68. The armature is connected to rotor shaft 60 and, when commanded, causes rotation of the shell 64. Additionally, the rotor shaft can also translate the zombie within the fluidic envelope. In the illustrated example, zombie 58 has one end connected to a motion controller 66 and another end that is free-floating. In this regard, the zombie may be translated or rotated freely within the fluidic envelope. However, the shape of the zombie cannot be adjusted. Accordingly, in another embodiment and as shown in FIG. 5, shell 64 is connected to pivot shaft 70 that sealingly extends into the fluid envelope. The pivot shaft preferably has a telescopic construction to support translation of the shell. Moreover, the pivot shaft provides a fixed point by which the shell can be rotated. As a result, the shape of the zombie can be contorted to provide a desired attenuating profile. Moreover, it is contemplated that when the telescoping pivot shaft is fully extended, the shell can be further translated away from the pivot shaft to achieve a stretching of the shell. Conversely, when the pivot shaft is fully retracted, the shell can be translated toward the shaft to compress the shell. Both of which provide additional flexibility in defining a desired x-ray attenuation profile. Additionally, not only is actuator 66 capable of translating and rotating the zombie shell, but it is further contemplated that the actuator may provide tilting of the shell when needed. Referring now to FIG. 6, another embodiment of the present invention is shown. In this embodiment, shell 64 is constructed to have a complex shape when at rest. In this regard, when external forces are not applied to the shell, the shell provides a unique, complex shape that is mirrored by the attenuating fluid in the fluid envelope. Moreover, the shell is constructed of material that will automatically retain this complex shape after external forces applied thereon are removed. Thus, if the shell is contorted to have a different shape, once those contorting forces are removed, the shell will return to its at-rest shape. In the exemplary zombie illustrated in FIG. 6, the shell has a generally frustoconical shape defined by a mid-level protuberance 64(a). As shown in the embodiment of FIG. 6, zombie 58 is constructed to have two rotor shafts connecting shell 64 to a rotary/linear motion control. In this regard, a solid rotor shaft 72 connects the motion controller 66 to the face plate 68. A hollow rotor shaft 74 is positioned circumferentially around the solid rotor shaft 72 and is connected to armature 62. The dual rotor system achieves rotation, translation, as well as, contortion of the shell. That is, rotor 72 is designed to rotate the entire zombie without causing a change in the shape of the zombie. On the other hand, rotor shaft 74 is designed to rotate independently of rotor shaft 72. In this regard, rotor shaft 74 rotates relative to shaft 72 which results in the actuator 62 rotating relative to the solid rotor shaft 72. As a result, the shape of the zombie contorts. Specifically, the shell is caused to twist around the linear axis defined by the solid rotor 72. One skilled in the art will appreciate that the dual rotor system includes a transmission or similar driving mechanism (not shown) to selectively rotate the rotor shafts independently or in tandem. Shown in FIG. 7 is an end view of the zombie of FIG. 6 illustrating one exemplary connection of the hollow rotor shaft and the solid rotor shaft. In this example, the hollow rotor shaft is connected to cam 76 and the zombie face plate 68 is connected to the solid rotor shaft 72. One skilled in the art will appreciate that rotation of the hollow rotor causes rotation of cam 76. Cam 76 is connected to an inner spring 78 such that rotation of the cam biases spring 78 relative to spring 80. Spring 80 is connected to the face plate 68. Therefore, rotation of the hollow rotor shaft effectuates a change in the shape of the zombie. It is understood that the motion controller 66, FIG. 6, may include a transmission (not shown) to selectively cause rotation of shafts 72 and 74. Further, it is contemplated that the transmission may include a rack and pinion or other gear arrangement to cause translation of the zombie. In this regard, the transmission may include a gear arrangement that causes independent rotation of rotor shaft 72 to effectuate a given orientation of the zombie in the fluidic envelope. The transmission may also include a gear arrangement that causes rotation of the hollow rotor shaft without rotation of the solid rotor shaft to effectuate shaping of the zombie and another gear arrangement, such as the aforementioned rack and pinion arrangement, to translate the zombie. In addition to geared arrangements, it is contemplated that rotation, translation, shaping, and tilting of the zombie can be achieved using hydraulic, pneumatic, and/or electrical circuits. Further, it is contemplated that a single or multiple motor/motion controllers may be used to effectuate rotation, translation, shaping, and/or tilting of the zombie. As illustrated in FIG. 6 and described herein, the zombie may have a complex shape, such as a shape that is generally matched to that of a medical patient. The zombie may also have a more simple shape, such as the frustoconical zombie illustrated in the embodiments of FIGS. 4-5. While a number of complex shapes are contemplated, in one preferred embodiment, the shape is preferably matched to that of a medical patient, as illustrated in FIG. 6. As shown, the zombie has a protuberance 64(a) that generally replicates a shoulder of a patient so that a general “neck” gradient 64(b) is also formed. Regardless of shape, it is preferred that the zombie be formed of deformable, pliable material so that the shape of the zombie can be changed. In a preferred embodiment, the zombie is constructed of viscoelastic or elastomeric material that retains its shape after external forces are removed therefrom. In this regard, when the zombie is not being contorted or after being contorted, the zombie will have its static or “natural” shape. Therefore, in a preferred embodiment, the zombie is constructed to have a shape suitable for whole body imaging of a medical patient. As such, the bowtie filter, by displacement of its attenuating fluid, is shaped for whole body imaging by default. However, as needed, the zombie's shape, position, and orientation can be adjusted, as described herein, to achieve differing bowtie filtering profiles as those profiles are needed for targeted anatomy imaging or whole body scans that require other than the default filtering profile. In particular, rotation of the zombie in concert with the rotation of the gantry system around the patient is envisioned. To achieve the variety in filtering profiles, it is contemplated that the zombie may be translated in the x, y, and z directions. In a preferred embodiment, the zombie is designed to be rotated relative to the z-direction (the axis defined lengthwise through the bore of the CT system), but could also be constructed to rotate relative to the x or y directions. As described above, it is contemplated that the zombie shell may be filled with foam, air, or other material. It is further contemplated that the zombie may be connected to a pump (not shown) to effectively add and remove air or other fluid from the zombie shell. In this regard, the inventors contemplate inflation as well as deflation of the zombie to achieve changes in zombie size. For this embodiment, the bowtie filter should be constructed of somewhat pliable material also to accommodate an increase in zombie size. However, it is also contemplated that attenuating fluid may be added or removed and its pressure changed as needed to account for variations in zombie size. Referring now to FIG. 8, package/baggage inspection system 82 includes a rotatable gantry 84 having an opening 86 therein through which packages or pieces of baggage may pass. The rotatable gantry 84 houses a high frequency electromagnetic energy source 88 as well as a detector assembly 90. A conveyor system 92 is also provided and includes a conveyor belt 94 supported by structure 96 to automatically and continuously pass packages or baggage pieces 98 through opening 86 to be scanned. Objects 98 are fed through opening 86 by conveyor belt 94, imaging data is then acquired, and the conveyor belt 94 removes the packages 98 from opening 86 in a controlled and continuous manner. As a result, postal inspectors, baggage handlers, and other security personnel may non-invasively inspect the contents of packages 98 for explosives, knives, guns, contraband, etc. Accordingly, the present invention includes an x-ray attenuating filter having an envelope containing x-ray attenuating fluid. The filter further has a fluid displacement device disposed within the envelope and configured to displace the x-ray attenuating fluid to define a desired attenuation profile for the filter. A CT system having an x-ray source, an x-ray detector, and an x-ray filter assembly is also presented. The x-ray filter assembly includes a bowtie filter having a body with attenuating fluid disposed therein as well as an attenuating fluid displacement device sealingly enclosed within the body of the bowtie filter. The filter assembly further has a mechanized actuator connected to the attenuating fluid displacement device to dynamically position the attenuating fluid displacement device within the body of the bowtie filter to define a desired filtering profile for the bowtie filter. An x-ray filter having a fluidic envelope is also disclosed. The fluidic envelope contains x-ray attenuating fluid that is designed to filter x-rays projected from the x-ray source toward an object to be scanned. The x-ray filter further has means for displacing the attenuating fluid within the fluidic envelope to achieve a desired x-ray filtering profile to prevent detector saturation during scanning of the object. An x-ray filter is also presented that includes a fluidic envelope having x-ray attenuating fluid disposed therein. The x-ray attenuating fluid includes at least one of liquid metal, nanoparticles suspended in a non-settling solution, or a compound with pH control buffer dissolved in a liquid. The present invention has been described in terms of the preferred embodiment, and it is recognized that equivalents, alternatives, and modifications, aside from those expressly stated, are possible and within the scope of the appending claims.
052727408	abstract	An agent for trapping the radioactivity of the fission products which appear in a nuclear fuel based on sintered uraniferous oxides in the course of irradiation characterized in that it comprises a stable oxygenated compound, a combination of at least two metallic oxides and at least one oxide of a non-radioactive isotope of the radioactive fission product or products whose radioactivity is to be trapped.
	summary
	summary
	abstract	A terminal end-piece for a fuel assembly of a pressurized water nuclear reactor, the assembly having fuel rods and a skeleton for supporting the fuel rods, the fuel rods extending in a longitudinal direction and being arranged at the nodes of a substantially regular network, the support skeleton comprising two terminal end-pieces and guide tubes that connect the terminal end-pieces, the fuel rods being arranged longitudinally between the terminal end-pieces, characterized in that the end-piece comprises noses for orientating the flow of a coolant fluid of the reactor along the adjacent longitudinal ends of the fuel rods, the noses being arranged in nodes of the substantially regular network in order to be positioned in a longitudinal continuation of at least some of the fuel rods and/or at least some of the guide tubes of the support skeleton.
	description	The present invention will be described in detail in conjunction with what are presently considered as preferred or typical embodiments thereof with reference to the drawings. In the following description, like reference characters designate like or corresponding parts throughout the drawings. Also in the following description, it is to be understood that such terms as xe2x80x9crightxe2x80x9d, xe2x80x9cleftxe2x80x9d, xe2x80x9cupperxe2x80x9d, xe2x80x9clowerxe2x80x9d and the like are words of convenience and are not to be construed as limiting terms. FIG. 1 is a sectional view showing generally and schematically a structure of a pressurized water reactor equipped with flow stabilizing members according to a first embodiment of the present invention. In this figure, some components are omitted from illustration. It should first be mentioned that in the pressurized water reactor now under consideration, the internal structure or internals of the reactor are, for the most part, essentially the same as the conventional nuclear reactor described hereinbefore. Accordingly, repetitive description thereof will be unnecessary. The pressurized water reactor according to the instant embodiment of the invention differs from conventional ones in that a short member 1 serving as a flow stabilizing member is mounted within the upper plenum 40 in an outer region extending outside of the fuel region along the inner wall of the core barrel 30 at a position lower than the outlet nozzle 12. In this conjunction, the phrase xe2x80x9couter region extending outside of the fuel regionxe2x80x9d means a region on the upper surface of the upper core plate 21 extending outside of the fuel assembly region 33 (indicated by a double-dotted line in FIG. 2) within the upper plenum 40 the outer periphery thereof corresponding to that of the core. Further, the phrase xe2x80x9cat a position lower than the outlet nozzle 12xe2x80x9d means that the top end of the flow stabilizing member 1 in the mounted state within upper plenum 40 is lower than a lowermost portion of the bore of the outlet nozzle 12. Next, referring to FIGS. 2 and 3, description will be made of a second embodiment of the present invention which is directed to a nuclear reactor for a four-loop plant in which the short members 1 according to the invention are employed as the flow stabilizing members. FIG. 2 is a schematic top plan view showing half of an essential portion of the nuclear reactor, and FIG. 3 is a perspective view showing a region where the short members are mounted in the vicinity of the outlet nozzles. As can be seen from the figures, the short members 1 serving as the flow stabilizing members, respectively, are disposed in the outer region extending outside of the fuel region in the proximity of the inner wall of the core barrel 30. More specifically, the short member 1 is mounted on the upper core plate 21 within the upper plenum 40 in a region extending or located outside of the fuel assembly region 33 (indicated by a double-dotted broken line) at a position close to the inner wall 30a of the core barrel 30. It should, however, be noted that in practical applications, the short member 1 is mounted at a certain distance from the inner wall 30a of the core barrel 30 in consideration of the thermal expansion of the short member 1. Furthermore, the short member 1 is mounted so that the top end thereof is positioned lower than the lowermost portion of a coolant outlet 12a of the outlet nozzle 12. This arrangement is effective for stabilizing the flow of the coolant within the upper plenum 40 in the regions located below the outlets 12a of outlet nozzles 12 as well as for reducing the hydrodynamic load acting on the short members 1. The length of the flow stabilizing member 1 is selected so that the top end thereof is positioned lower than the lowermost portion of the outlet 12a of the outlet nozzle 12 when the flow stabilizing member 1 is mounted on the upper core plate 21. More specifically, the flow stabilizing member 1 should preferably be dimensioned so that the length thereof falls within a range of from a position midway between the upper surface of the upper core plate 21 and the lowermost portion of the outlet 12a of the outlet nozzle 12 to a position lower than the lowermost portion of the outlet 12a. If the top end of the flow stabilizing member 1 is positioned higher than the lowermost portion of the outlet 12a of the outlet nozzle 12, the flow resistance of the coolant flowing within the upper plenum 40 toward the outlet nozzle 12 from a center portion of the reactor core will increase, and the hydrodynamic load acting on the short member 1 will increase as well to ultimately adversely effect the mechanical or structural strength of the internal structure of the reactor. On the other hand, if the length of the short member 1 is excessively short, it is difficult to stabilize the flow of the coolant along the inner wall 30a of the core barrel 30. For these reasons, the flow stabilizing member 1 should be so dimensioned that the length thereof falls within the range defined above. Furthermore, installation of the flow stabilizing member 1 can be realized by securing a bracket 2 which is fixed to the flow stabilizing member 1 onto the upper core plate 21 by means of bolts 3. The short length and ease of mounting of the flow stabilizing members on the upper core plate 21 can facilitate installation of the flow stabilizing members in existing equipment as well. Next, with reference to FIGS. 4 and 5, description will be made of the flow behavior of the coolant in the vicinity of the outlet nozzles 12 in the structure where the short members 1 are installed within the upper plenum in the manner described above. FIG. 4 is a schematic top plan view of half of a nuclear reactor illustrating disposition of the internal structural members within the upper plenum 40 together with the flow of the coolant in the vicinity of the outlet nozzles 12, and FIG. 5 is a schematic side view showing behavior of the coolant flowing in the vicinity of adjacent outlet nozzles 12. As can be seen from the figures, the coolant flows radially outward along the upper surface of the upper core plate 21 from a center region of the core to reach the inner wall 30a of the core barrel 30, whereupon the coolant flows toward the outlet nozzles 12 along the inner wall 30a of the core barrel 30 in the region outside of the outer periphery of the core. Here, a portion of the coolant or light water flowing along the inner wall 30a of the core barrel 30 flows below the outlet nozzle 12. Accordingly, when the short members 1 serving as the flow stabilizing members are not disposed, as in conventional reactors, streams of the coolant flowing in opposite directions collide with each other in a region between the adjacent outlet nozzles 12. In contrast, by disposing the short flow stabilizing members 1 according to the instant embodiment of the invention lower than the outlet nozzles 12, respectively, as mentioned previously, streams F of the coolant along the inner wall 30a can flow smoothly into the outlets 12a of the outlet nozzles 12 under the guiding action of the flow stabilizing members 1, as can be seen in FIG. 5. Further, a stagnation region S occurring between the flow stabilizing members 1 is decreased and the coolant in the stagnation region S is forced to smoothly flow upward into the outlet 12a of the outlet nozzle 12 under the constraining action exerted by the two flow stabilizing members 1. Accordingly, the coolant flow entering the outlet nozzle 12 can stabilized, as a result of which temperature fluctuation of the coolant flowing through the outlet pipe 42 connected to the outlet nozzle 12 can be effectively suppressed. Furthermore, measurement of variation or fluctuation of the temperature within the outlet pipe 42 connected to the outlet nozzle 12 in a demonstration test simulating flow behavior of the outlet nozzle 12 and upper plenum 40 show that temperature fluctuation in the outlet pipe 42 of the reactor equipped with the flow stabilizing members 1 can be suppressed or reduced by approximately half when compared with the structure in which no flow stabilizing members 1 are employed. Furthermore, the flow stabilizing member 1 according to the instant embodiment of the invention is implemented in a hollow cylindrical shape in view of the fact that the hollow cylindrical flow stabilizing member 1 can be manufactured with light weight and relatively low cost. However, the present invention is not restricted to a flow stabilizing member with such a shape. As long as the stream F of the coolant flowing along the inner wall 30a of the core barrel 30 can be stabilized, as described above, flow stabilizing members 1 with different structures such as a solid cylindrical or columnar structure, a plate-like structure, a prism structure or the like can be employed as well. A third embodiment of the present invention is directed to application of the flow stabilizing member to a nuclear reactor for a two- or three-loop plant. First, it is to be noted that the disposition of the flow stabilizing member(s) according to the instant embodiment of the invention can be equally adopted in the four-loop reactor plant described above. FIG. 6 is a schematic top plan view showing an essential portion of the nuclear reactor for a three-loop plant, wherein some components are omitted from illustration, and FIG. 7 is a perspective view showing a region of the upper plenum in the vicinity of the outlet nozzles 12. Here, it should be mentioned that the shape and the length of the flow stabilizing member 1 are essentially the same as those of the flow stabilizing member described hereinbefore. Accordingly, repeated description thereof will be unnecessary. The third embodiment of the invention differs from the preceding embodiments in that the outlet nozzles 12 are not disposed adjacent to each other. The flow stabilizing member 1 according to the third embodiment of the invention is disposed directly underneath a central portion of the outlet 12a of the outlet nozzle 12 in close proximity to the inner wall 30a of the core barrel 30. As mentioned hereinbefore, disposition of the flow stabilizing member in close proximity to the inner wall 30a of the core barrel 30 is very effective for reducing the hydrodynamic load applied to the flow stabilizing member 1 and also stabilizes flow of the coolant F in a space within the upper plenum 40 below the outlet nozzle 12. Now, description will turn to the flow behavior of the coolant in the vicinity of the outlet nozzle 12 in the reactor equipped with the flow stabilizing members 1 according to the third embodiment of the invention. Portions of the coolant flowing opposite to one another along the inner wall 30a of the core barrel 30 tend to collide with each other beneath the outlet nozzle 12. However, because the flow stabilizing member 1 is disposed underneath the center portion of the bore of the outlet nozzle 12, the streams of the coolant flowing along the inner wall 30a of the core barrel 30 are forced to flow upward uniformly at both sides of the flow stabilizing member owing to the flow guiding action thereof. Thus, the coolant can flow smoothly into the outlet 12a defined by the outlet nozzle 12. In other words, collision of the coolant streams flowing opposite to one another along the inner wall 30a beneath the outlet nozzle 12 can be effectively suppressed, whereby the stream of the coolant flowing into the outlet nozzle 12 can be stabilized. Consequently, temperature fluctuation within the outlet pipe 42 connected to the outlet nozzle 12 can be suppressed. Although it has been described that the flow stabilizing member 1 employed in the instant embodiment of the invention is installed underneath a center portion of the outlet of the outlet nozzle 12, it goes without saying that the flow stabilizing member 1 can be mounted at practically any position as long as the mounting position of the flow stabilizing member is covered by a region extending below and across the outlet nozzle 12. What is important is to avoid collision of the coolant streams in the vicinity of the region beneath the outlet nozzle 12, because then the flow of the coolant flowing into the outlet nozzle 12 can be stabilized. Accordingly, when it is impossible to mount the flow stabilizing member underneath the central portion of the outlet nozzle in view of structural limitations, then the flow stabilizing member 1 can be mounted at a position more or less deviated from a position underneath the center of the bore of the outlet nozzle 12 as long as the deviated position lies within the region extending beneath and across the outlet nozzle 12. Thus, the present invention can be applied to existing nuclear reactors without any appreciable difficulty. In the foregoing, exemplary embodiments of the present invention which are considered preferable at present and other alternative embodiments have been described in detail with reference to the drawings. It should, however, be noted that the present invention is not restricted to these embodiments and other variations and modifications can be easily conceived and realized by those skilled in the art without departing from the spirit and scope of the present invention.
	description	This application is a division of U.S. application Ser. No. 15/087,111, filed on Mar. 31, 2016, which is a continuation of U.S. application Ser. No. 13/399,082, filed on Feb. 17, 2012, now U.S. Pat. No. 9,312,037, which claims the benefit of U.S. provisional application Ser. No. 61/540,897, filed on Sep. 29, 2011, each of which is incorporated herein by reference in its entirety. This invention was made with government support under Contract No. DE-AC02-06CH11357 awarded by the United States Department of Energy to UChicago Argonne, LLC, operator of Argonne National Laboratory. The government has certain rights in the invention. This invention relates to methods and a novel device for producing radioisotopes for medical applications. More particularly, this invention relates to methods, as well as novel target units and sublimation devices for producing Cu67 radioisotope. In recent years medical researchers have indicated a desire to explore radioisotope therapy with beta-emitting sources that may simultaneously be monitored by imaging their photon emission. Beta particles with energies of a few hundred KeV have sufficient range in tissue (millimeters) that they can penetrate small tumor masses, without passing much further into the surrounding body and inadvertently destroying healthy tissue. Gamma rays of a few hundred KeV may be conveniently imaged with external cameras. An isotope that emits both particles must also have appropriate chemical properties in order to attach the isotope to a biologically active agent, such as a peptide or monoclonal antibody. Copper-67 (Cu67) has emerged as one of the most desired of these new radioisotopes; it emits beta particles with mean energy of 141 KeV and a gamma ray of 185 KeV. Its half-life of 2.6 days, however, demands rapid production, processing, and transfer to the medical clinic. Therapy of non-Hodgkin's lymphoma is perhaps the most recognized application for Cu67, but the dearth of supply has seriously inhibited the research effort in this area. Cu67 has been produced by two main processes, i.e., in nuclear reactors in small quantities, and by bombardment of zinc oxide (ZnO) with high energy protons. In the mid 1990s, Cu67 was produced by irradiation of ZnO in DOE-subsidized high-energy physics proton accelerators, e.g., BLIP at Brookhaven National Lab (BNL) and LAMPF at Los Alamos National Lab (LANL). By 2000, DOE changed its focus, with additional production being performed on the proton cyclotron at TRIUMF, in Canada, and import of the Cu67 to medical researchers in the United States. Reactor production of Cu67 is particularly difficult for several reasons. For example, neutron flux results in a number of harmful, unwanted other isotopes, which are difficult to remove from the desired Cu67. Human medical treatment applications require non-copper impurities to be reduced to parts-per-billion (ppb) levels, elimination of radioisotopes of copper other than Cu67, and a high specific activity (no more than a few hundred stable copper atoms for each Cu67 atom). In addition, the reactor method needs a sophisticated mechanical rabbit to retrieve the isotope from the core, and radioactive waste handling is costly (frequently requiring subsidization by national governments), which generally hinders economic production of radioisotopes. Linear accelerator (“linac”) production at BLIP and LAMPF was technically successful, but the two labs simply could not provide enough Cu67 to meet the demand. Production was limited to a total of about 1 Ci per year, due to scheduling demands on the accelerators for high-energy physics missions. Also, proton accelerator production requires irradiation of the target in a vacuum, and the machine must be opened to atmospheric pressure to recover the target, complicating the recovery. In the past, metal zinc target capsules have been used on electron accelerators to provide high yields of Cu67 via a photonuclear process (gamma rays from Bremsstrahlung convert Zn68 into Cu67). Zinc material was then irradiated, and Cu67 would be separated very quickly and efficiently using a sublimation process. Both the metal casting process into metal target capsules and subsequent sublimation attempts with metal apparatus have resulted in unacceptable levels of metal impurities, which were introduced by corrosive chemical reactions of zinc in the liquid and vapor phases. Accordingly, there is an ongoing need for improved methods for producing Cu67, particularly having a purity and specific activity suitable for medical applications. The present invention addresses this need. The present invention provides a photonuclear method for producing Cu67 radioisotope suitable for use in medical applications. The method comprises irradiating metallic zinc-68 (Zn68) contained within a closed ceramic capsule with a high energy gamma ray beam to convert at least a portion of the Zn68 to Cu67, and then isolating the Cu67 from the irradiated target. During irradiation, at least a portion of the Zn68 is converted to Cu67 by loss of a proton. Preferably, the irradiation is continued until the conversion of Zn68 to Cu67 yields a Cu67 activity of at least 5 milliCuries-per-gram of target (mCi/g). Our work has uncovered that composing the target capsule and sublimation apparatus out of ceramic materials that do not chemically react with molten zinc (e.g., alumina, aluminum nitride and boron nitride), and in particular alumina, offers a solution to avoiding the introduction of impurities during casting or sublimation known to take place in prior equipment. The present invention also provides an improved target unit for producing Cu67 radioisotopes. It also provides for easier handling and shipping of the target because of its use of low activation materials. The target unit includes a target body having a cage body coupled to a screw-on cap and a ceramic capsule containing the Zn68 target. The ceramic capsule is sealed within the target body between the cage body and the screw-on cap to form a substantially water-tight seal during irradiation. The ceramic capsule material must be selected to prevent chemical reaction with zinc; nevertheless, it must promote a solid physical contact between the capsule and solid Zn68 target ingot within the capsule. Even a small gap between the capsule and the Zn ingot would inhibit heat transport out of the zinc during high-power irradiation, resulting in melting and possible failure of the target. For this reason certain non-metals, such as graphite and boron nitride, are not appropriate for the target capsule. Alumina is an example of one satisfactory material of construction for the capsule. The initial stock of Zn68, and any additions of fresh stock to replace losses, must be substantially free of residual traces of oxygen. Substantially oxygen-free zinc promotes good physical contact between the cast ingot and the ceramic capsule. Substantially oxygen-free zinc can be prepared by subliming the Zn68 at least once prior to forming the target ingot. As used herein, the term “substantially oxygen-free zinc” and grammatical variations thereof, refer to trace oxygen levels within the target ingot that are low enough to prevent loss of adhesion between the capsule and the zinc target ingot during irradiation. The present invention also provides for an improved apparatus for subliming the irradiated metallic zinc target material from the Cu67 radioisotope. The sublimation apparatus comprises a ceramic sublimation body, which is a vacuum sealable tube with one open end. A ceramic capsule containing the irradiated metallic zinc target is placed within the sublimation body. The sublimation body is coupled to a vacuum source, which forms a leak-tight vacuum seal at temperatures between approximately 500 to about 700° C. The ceramic sublimation body material must be chosen to prevent chemical reaction with zinc liquid and vapor, but the zinc vapor from sublimation must deposit and physically adhere to the interior of cooler regions of the tube that are not directly heated. Additionally, upon later heating, the deposited zinc must melt and flow freely for subsequent recovery of the expensive Zn68 to refill and cast a new target ingot within a new capsule. For this reason certain non-metals, such as quartz/glass, are not appropriate for the sublimation body. Alumina is one example of a satisfactory material of construction for the sublimation tube. In addition, the present invention provides an improved method for recovering the sublimed Zn68 from the sublimation tube. In particular, the open ended ceramic sublimation tube is inverted over a hopper in order to fill a new ceramic capsule. The inverted sublimation tube and hopper are placed within a hermetic surround and heated in an inert atmosphere. The hopper funnels the molten zinc into the new ceramic capsule. For this process the hopper must be constructed from a non-metallic material which has no chemical reaction with molten Zn; graphite or glassy carbon are satisfactory materials, which may be easily fabricated into the desired hopper dimensions to properly align with the opening of the tube. Further details regarding sublimation and irradiation of zinc for producing Cu67 radioisotope are described in U.S. patent application Ser. No. 12/462,099, filed Jul. 29, 2009, the disclosure of which is incorporated herein by reference in its entirety. The present invention provides a method for producing Cu67 radioisotope comprising irradiating a metallic Zn68 target with a high energy gamma ray beam to convert Zn68 atoms to Cu67, and then isolating the Cu67 from the irradiated target. Preferably, the target to be irradiated comprises at least about 90% Zn68, more preferably at least about 95% Zn68, and even more preferably at least about 99% Zn68. It is particularly preferred that the Zn68 target include as low a level of copper contaminant as is practical, in order to minimize the amount of cold copper recovered after irradiation to produce radioactive Cu67. Zn68 containing low levels of copper can be obtained, for example, by repeated sublimation or by zone refining of the Zn68. At each sublimation stage less than 10% of the small amount of copper in the target material is transferred with the sublimed material, thereby affording a higher ratio of radioactive copper to cold copper after each cycle until substantially all of the cold copper is depleted from the zinc. The quantity, Q1, of initial copper in the bulk zinc target can be measured, as can the amount of copper, Q2, left in the sublimed zinc deposit. The metric r=(Q2/Q1)×100% (i.e., the percentage of copper left in the sublimed zinc) is a figure of merit, which provides an assessment of the efficiency of the sublimation process for removing trace amounts of copper from the bulk zinc. In six different sublimation runs, the percentage of copper removed from the zinc during sublimation was in the range of 85 to 99.5% (i.e., values of r=0.5%, r<1.4%, r=2.5%, r=3.6%, and r≤15% were observed). Based on these observations, recycling of the target zinc material will likely reduce trace amounts of cold copper by orders of magnitude after a few sublimation cycles. Thus, utilizing Zn68 that has been repeatedly sublimed (e.g., Zn68 sublimate recovered from repeated runs of the present methods) will lower the level of cold copper present in the Cu67 obtained after irradiation, and thus increase the specific activity of the Cu67 in the copper isolated from the process. The sublimation processing procedure can thus provide an extremely high specific activity of Cu67. For example, the radioisotope Cu67 product supplied to customers can have fewer than ten cold (nonradioactive, stable) copper atoms for each Cu67 atom. This is equivalent to a specific activity of ≥75 kCi/gram of copper. The Zn68 target present in the ceramic capsule can be configured in any suitable and convenient manner. For example, the target can be configured in the form of a frustum, a straight cylinder, or any other suitable shaped solid mass, and the like. The target and capsule can also be housed in a unit as desired, which preferably provides a water-tight seal for the capsule. The Zn68 within the capsule can be any solid monolithic ingot in tight contact with the capsule, such as a solid plate, a solid cylinder, or any other suitable configuration. Good physical contact between the solid ingot and the capsule can be achieved by pre-sublimation of the zinc to guarantee removal of oxygen from the metal. The target preferably has a mass in the range of about 100 to about 200 grams, although smaller and larger targets are suitable, as well. The Zn68 target is irradiated with a gamma ray beam having an intensity of at least about 1.3 kW/cm2, and comprising gamma rays having an energy of at least about 30 MeV. In a preferred embodiment, the gamma rays are produced by irradiating a tantalum target (Ta converter) with a high energy electron beam (e.g., 40-50 MeV, 6-10 kW) from a linear accelerator. The irradiation produces gamma rays of suitable energy for converting Zn68 to Cu67. Preferably, the tantalum is irradiated with a high power electron beam having a beam energy in the range of about 40 MeV to about 100 MeV and a beam current in the range of about 100 to about 200 microAmperes. Irradiation of the tantalum results in production of gamma rays having an energy in the range of about 40 to about 100 MeV, which is well suited for conversion of Zn68 to Cu67. Preferably, the irradiation is continued until the conversion of Zn68 to Cu67 yields a Cu67 activity in the target of at least about 5 milliCuries-per-gram of target (mCi/g), more preferably at least about 10 mCi/g, even more preferably at least about 20 mCi/g. Typical irradiation times are in the range of about 24 to 72 hours. The tantalum converter preferably has a thickness in the range of about 1 to about 4 mm and can comprise a single plate of tantalum or multiple stacked plates. Alternative converter materials include tungsten (preferably coated with a thin layer of Ta for chemical stability), or heavier metals such as lead (e.g., encased in a sealed jacket). The tantalum converter and the Zn68 target can be configured in any suitable manner within the electron beam of the linear accelerator. Due to the inevitable heating of the converter and target, cooling may be required during irradiation to avoid mechanical failure of the target (e.g., melting). Preferably, the converter and target are cooled by a recirculating cooling system (e.g., immersed in a forced-flow cooling water bath) while in the beam path of the linear accelerator. The target ceramic capsule is mounted in a holder or target unit that is water tight and may include cooling fins in a suitable number and size to aid in dissipating the heat generated during the irradiation, if desired. The target unit or holder with its included target preferably is immersed within cooling water during irradiation. After irradiation, the linear accelerator is shut down, the cooling water flow is stopped, and the target unit is removed for processing to recover the Cu67 therefrom. FIG. 1 illustrates a partial cross-sectional view of an exemplary embodiment of target unit 10, which houses the target and capsule during irradiation. Target unit 10 includes threaded cage body 20 and screw-cap 36, which can be screwed together to house capsule 40. Cage body 20 is substantially cylindrical having a top 22 defining an aperture 24 and an open male-threaded bottom portion 26, which defines opening 28 sized and configured to receive capsule 40. Cage body 20 also defines circumferential oblong apertures 30. A portion 32 of cage body 20 between male-threaded bottom portion 26 and apertures 30 defines a groove 35. Capsule 40 includes a closed end 42 and open end 44, together defining target cavity 45. Metal lid 46 includes closed end 48 and open end 50, which is sized and configured to receive open end 44 of capsule 40. Gasket 51 is disposed within lid 46 to seal against open end 44 of capsule 40. When assembled, closed end 42 of capsule 40 is received within open end 28 of cage body 20, while lid 46 covers open end 44 of capsule 40, with gasket 51 therebetween. Female threaded portion 38 of screw-cap 36 is engaged with male threaded portion 26 of cage body 20 such that screw-cap 36 and cage body 20 together exert sufficient force on cap 36 to provide a water-tight seal over open end 44 of capsule 40. Preferably, washer 53 is included between screw-cap 36 and closed end 48 of lid 46. In a preferred embodiment, gasket 51 is composed of graphite because it is highly resistant to radiation. Gasket 51 may be composed of other materials, excluding those containing copper. FIG. 2 provides an isometric view of assembled target unit 10. As illustrated in FIG. 2, screw-cap 36 includes flattened regions 37 to provide surfaces suitable to facilitate tightening of screw-cap 36 and cage body 20, e.g., by hand or with a wrench. FIG. 3 provides a top plan view of target unit 10, while FIG. 4 shows a bottom plan view, and illustrates the positioning of four flattened regions 37 symmetrically spaced along the circumference of screw-cap 36. FIG. 1A illustrates an alternative embodiment of target unit 10, in which cage body 20a defines a larger number of apertures 30a than cage body 20 of FIG. 1. Apertures 30 and 30a can be configured in any form or manner desired. The purpose of including apertures 30 or 30a in target unit 10 or 10a is to allow cooling water to contact capsule 40 during irradiation to prevent melting or partial melting of the zinc target ingot during irradiation. Capsule 40 is a ceramic crucible, and can be constructed of alumina or aluminum nitride, for example, because these materials do not chemically combine with zinc. Alumina is preferred because it is inexpensive and is a well-characterized material. Test results have shown that use of capsules composed of alumina by the disclosed methods and equipment do not introduce undesirable metal and other impurities into the resulting Cu67 in significant amounts. Tests also have shown that the initial zinc target (or any fresh zinc to make up for losses) should be substantially free from traces of oxygen, e.g., by pre-purifying the zinc by sublimation to eliminate traces of oxygen; this beneficially promotes good physical contact, after casting, between the cooled solid zinc ingot and the ceramic capsule. If oxygen is present in the zinc, a gap between the capsule and the zinc ingot may form upon cooling of the molten zinc after filling of the capsule. Such gaps can lead to inefficient cooling, and failure of the target. When assembled, a small expansion gap, between about 2 and about 3 mm, preferably is provided between the zinc ingot and metal lid 46. This gap is sufficient to provide the zinc with adequate thermal creep to avoid cracking the capsule as it expands under high-power heating. In other embodiments, a small zinc foil may be fitted within the gap to allow for current leakage during electron beam irradiation, from the zinc metal to the metal lid. Tests have shown there is no galvanic corrosion inside capsule 40 during beam operations. Cage bodies 20 and 20a provide physical protection to ceramic capsule 40, as well as an interface-connection to the target chamber at the electron linac. In a preferred embodiment, cage bodies 20 or 20a and screw-cap 36 are composed of different alloys of aluminum to minimize the possibility of thread galling. For example, cage bodies 20 or 20a can be composed of 6061 Al and screw-cap 36 can be composed of 2024 Al. The size and configuration of the target unit (e.g., 10 or 10a) is dictated by the size and configuration of the target chamber and amount of zinc to be irradiated. Thus, the configuration of the target unit may be varied without departing from the spirit of the invention. While the preferred embodiment utilizes a cage body, lid having a gasket, washer and screw-cap to secure the capsule within the target unit, fewer components may be utilized, provided that a water-tight seal is created for the target capsule. After the Zn68 has been irradiated for a sufficient period of time, the Cu67 produced in the target is isolated from the Zn68 by any suitable method. For example, the metallic target can be reacted with an acid to dissolve the metals and produce a mixture of metal ions (e.g., zinc and copper ions). The metal ions can then be separated from one another by chemical techniques that are well known in the art, including ion extraction, ion exchange, precipitation of insoluble metal salts, and the like. Preferably, the zinc is separated from copper by physical means, e.g., sublimation of zinc. Zinc can be readily sublimed away from copper at an elevated temperature under vacuum. In a preferred embodiment, the Cu67 is isolated by sublimation of the zinc at a temperature in the range of about 500 to about 700° C. under vacuum, preferably at a pressure in the range of about 10−3 to about 10−5 Torr, to remove a substantial portion of the zinc and afford a residue containing Cu67. Preferably, at least about 90%, 95% or 99% of the zinc is removed by sublimation, more preferably at least about 99.9%, even more preferably at least about 99.99%, on a weight basis. The Cu67-containing residue preferably is further purified by chemical means, such as reaction with an aqueous acid to form a solution of metal ions, followed by ion extraction, ion exchange, or a combination thereof to recover Cu67 ions. An example of sublimation apparatus 60 for use in the methods of the present invention is shown in FIG. 5 and FIG. 5A, in cross-section. Sublimation apparatus 60 comprises sublimation tube 62, capsule 40, coupler unit 66 and vacuum dome 64, which includes port 65 for attachment to a vacuum source. Sublimation tube 62 includes open end 61, which is sized and configured to have similar dimension to open end 63 of vacuum dome 64. Coupler unit 66 seals open end 61 of tube 62 to open end 63 of vacuum dome 64, by means of O-rings 86 and 88. FIG. 5A provides a detailed cross-sectional view of coupler unit 66, which comprises a tubular sheath 68, which is threaded at each end by male-threaded regions 70 and 72. Rings 74 and 78 include female-threaded regions 76 and 80, which are sized and configured to engage male-threaded regions 70 and 72 of sheath 68. Washers 82 and 84 are fitted within rings 74 and 78, respectively. O-rings 86 and 88 are disposed between the ends of sheath 68 and washers 82 and 84 when unit 66 is assembled. When rings 74 and 78 are screwed onto sheath 68, O-rings 88 and 86 are compressed between sheath 68 and washers 82 and 84. Ring 74 defines an aperture 71 which is sized and configured to receive open end 61 of sublimation tube 62, while ring 78 defines aperture 79, which is sized and configured to receive open end 63 of vacuum dome 64. O-rings 86 and 88 are sized to fit tightly against the exterior circumferences of sublimation tube 62 and vacuum dome 64, respectively. When rings 74 and 78 are tightened over sheath 68 with tube 62 and vacuum dome 64 received in apertures 71 and 79, O-rings 86 and 88 become compressed against tube 62 and vacuum dome 64 to form a vacuum-tight seal between tube 62 and vacuum dome 64. In use, sublimation apparatus 60 is assembled with capsule 40, which contains a solid ingot 90 of irradiated Zn68, and is situated within sublimation tube 62. Coupler unit 66 is tightened to provide a vacuum-tight seal, and the lower portion of tube 62 is heated to a temperature in the range of about 500 to about 700° C., while applying a vacuum in the range of about 10−3 to about 10−5 Torr via port 65. Zinc from ingot 90 sublimes and collects along the inner surface of tube 62 in areas that are not heated, leaving behind a residue of Cu67 in capsule 40 at the end of the sublimation process. The heating and sublimation cycle should be sufficiently slow to avoid thermal cracking of sublimation tube 62 as known by those of ordinary skill in the art. After sublimation is complete, heating is ceased, and the apparatus is allowed to cool at a relatively slow rate. Sublimation tube 62 preferably is composed of a ceramic material, such as alumina or boron nitride, because there is no chemical reaction between the ceramic and the zinc metal during sublimation. As there is no chemical reaction, no impurities are introduced to the Cu67. The material of construction of sublimation tube 62 may vary, provided that the selected material does not result in a corrosive chemical reaction with the Zn68 metal and Cu67 residue. Use of sublimation apparatus 60 is not limited to sublimation separation of Zn68 metal from Cu67 residue. If the sublimation body is used to sublime other types of materials, the sublimation body may be composed of a different material as known by those of ordinary skill in the art. Vacuum dome 64 can be composed of any suitable material, such as glass or metal. In the preferred embodiment, coupler unit 66 is composed primarily of stainless steel, with the exception of the O-rings, which can be any suitable chemically resistant polymeric material, such as, e.g., copolymers of hexafluoropropylene (HFP) and vinylidene fluoride (VDF or VF2), terpolymers of tetrafluoroethylene (TFE), vinylidene fluoride (VDF) and hexafluoropropylene (HFP), and the like, manufactured under the tradename VITON® by DuPont Performance Elastomers LLC. Other materials of construction may be utilized without departing from the spirit of the invention provided the chosen material does not result in unwanted contamination of the sublimed Zn68 and still provides for a leak-tight pressure seal. FIG. 5B and FIG. 5C provide isometric views of an alternative configuration for the vacuum dome and coupler. FIG. 5B shows the parts partially disassembled, while FIG. 5C shows the dome and coupler attached to each other. Coupler unit 66b includes sheath 68b, which is threaded at one end for engagement with threaded ring 74b, with an O-ring, not shown, as described above with respect to FIG. 5. Sheath 68b also includes flange 69b at its other end. Vacuum dome 64b includes gasket 67b which is sized and configured to seal against flange 69b, when open end 63b of dome 64b is received within sheath 68b. Clamp 75 is sized and configured to compress gasket 67b against flange 69b, forming a vacuum-tight seal. Dome 64b also includes flanged vacuum port 65b for connection to a vacuum source. In the embodiments shown in FIGS. 5B and 5C, the components (other than the gasket and O-ring) preferably are composed of a metal such as stainless steel. Test results have shown that the zinc-copper separation created through use of the disclosed sublimation apparatus and method is extremely efficient. Very little Cu67 transports with the sublimed-deposited zinc and extremely small amounts of zinc remain behind with the Cu67 in the capsule. The remaining Cu67 residue, however, can be further purified by dissolution in an acid (e.g., a mineral acid such as sulfuric acid, hydrochloric acid, phosphoric acid, nitric acid, or a combination of mineral acids). Tests have shown that ceramics, and in particular alumina, have negligible solubility in acids, so substantially no additional impurities are introduced through the further purification of the sublimed zinc by the acid solution. The sublimed zinc can be further processed to efficiently separate the remaining traces of zinc from the copper using ion exchange with a copper and/or zinc selective ion exchange resin (e.g., a quaternized amine resin), anion exchange (BioRad AG 1-X8 columns), or a chelating or solvating extractant, preferably immobilized on an ion exchange resin or silica substrate, to afford a Cu67 salt of suitable purity and specific activity for use in human medical applications. In one embodiment, the copper residue is dissolved in hydrochloric acid and the resulting Cu67 ions are purified on a quaternary amine ion exchange resin, as is well known in the art (see e.g., Mushtaq, A., Karim, H., Khan, M., 1990. Production of no-carrier-added 64Cu and 67Cu in a reactor. J. Radioanal. Nucl. Chem. 141, 261-269). Suitable metal chelating and solvating extractants are well known in the art and include, e.g., the CYANEX® brand extractants available from Cytec Industries, Inc., West Patterson, N.J., which comprise organophosphorous materials such as organophosphine oxides, organophosphinic acids, and organothiophosphinic acids. Such extractant can be immobilized on resin or silica beads, as is known in the art. See, e.g., U.S. Pat. No. 5,279,745; Kim et al., Korean Journal of Chemical Engineering, 2000; 17(1): 118-121; Naik et al., Journals of Radioanalytical and Nuclear Chemistry, 2003; 257(2): 327-332; Chah et al, Separation Science and Technology, 2002; 37(3): 701-716; and Jal et al., Talanta, 2004; 62(5): 1005-1028. The Cu67 recovered after ion exchange typically can be obtained in specific activity of up to 100 kCi/g at a purity suitable for human medical use. The Zn68 sublimate is preferably recycled for use as another target, so as to reduce the level of cold copper contaminant in the Zn68 target with each successive recycle, thus affording a radioactive copper residue containing a higher ratio of Cu67 to non-radioactive copper after each recycle stage, as described above. FIG. 6 shows exemplary recycling apparatus 100 to recycle Zn68 sublimate 105 for use as another target. Recycling apparatus 100 includes sublimation tube 62, hopper 102 and capsule 40 (e.g., as described in FIGS. 1-5). Sublimation tube 62 including Zn68 sublimate 105 on the interior wall of the tube is inverted and placed over hopper 102. Hopper 102 has a substantially cylindrical exterior and includes an internal funnel 104 configured to deposit molten liquid Zn68 into capsule 40 when sublimation tube 62 is heated to melt the zinc deposited on the interior of the tube. In the preferred embodiment, hopper 102 is composed of a high density, high purity graphite such as POCO; optionally the graphite can be coated with glassy carbon. Hopper 102, however, may be composed of a variety of different materials provided the material does not chemically react with the liquid zinc. During use recycling apparatus 100 is placed within a hermetic surround (not shown) as known by those of ordinary skill in the art to create an inert gas structure substantially free of oxygen around apparatus 100. The hermetic surround is then inserted into a furnace or other heating apparatus so that sublimed zinc 105 melts from sublimation tube 62. The hermetic surround may be composed of quartz, steel, or any other suitable material. Hopper 102 directs the molten liquid Zn68 into capsule 40. In the preferred embodiment, this process is done with an inert gas fill at atmospheric pressure, with temperatures in the range of about 450 to about 550° C. Experiments have shown that it is possible to process and recycle the zinc in the manner described into new target ingots contained within new capsules with negligible loss of the zinc material. The melt and fill cycle must be sufficiently slow (about 2 to about 3° C. per minute heating rate) to avoid thermal cracking of the sublimation tube (e.g., an alumina tube). Measurements have shown that the target unit disclosed herein results in very low radiation dose rate from the structural materials because alumina and aluminum are low-activation materials. After linac operations, the principal radiation hazard is provided by the zinc target material itself. Operations with enriched Zn68 (>99%) are characterized by even lower activation, since Cu67 will be the predominant isotope, and it has a very soft gamma emission which is easy to shield. The following example is provided to further illustrate certain aspects of the present invention, and is not to be construed as limiting the invention in any way. Sublimation separation of the irradiated metallic zinc from the Cu67 radioisotope was achieved on a zinc target ingot. The solid zinc target ingot within an alumina capsule was placed within a vacuum-tight alumina sublimation tube. The bottom of the sublimation tube was placed into a tube furnace and heated under an internal vacuum, to around 700° C. The sublimed zinc deposited on the cooler top of the sublimation tube, which was outside the furnace. Sublimation occurred very rapidly, at about greater than 40 g/h under a modest vacuum of about 1 mTorr. The heating and sublimation cycle was sufficiently slow, about less than 3° C. per minute, to avoid thermal cracking of the alumina. Once the sublimation process was complete, the furnace was shut down and the system was allowed to cool at a slow rate. All references, including publications, patent applications, and patents, cited herein are hereby incorporated by reference to the same extent as if each reference were individually and specifically indicated to be incorporated by reference and were set forth in its entirety herein. The use of the terms “a” and “an” and “the” and similar referents in the context of describing the invention (especially in the context of the following claims) are to be construed to cover both the singular and the plural, unless otherwise indicated herein or clearly contradicted by context. The terms “comprising,” “having,” “including,” and “containing” are to be construed as open-ended terms (i.e., meaning “including, but not limited to,”) unless otherwise noted. Recitation of ranges of values herein are merely intended to serve as a shorthand method of referring individually to each separate value falling within the range, unless otherwise indicated herein, and each separate value is incorporated into the specification as if it were individually recited herein. All numerical values obtained by measurement (e.g., weight, concentration, physical dimensions, removal rates, flow rates, and the like) are not to be construed as absolutely precise numbers, and should be considered to encompass values within the known limits of the measurement techniques commonly used in the art, regardless of whether or not the term “about” is explicitly stated. All methods described herein can be performed in any suitable order unless otherwise indicated herein or otherwise clearly contradicted by context. The use of any and all examples, or exemplary language (e.g., “such as”) provided herein, is intended merely to better illuminate certain aspects of the invention and does not pose a limitation on the scope of the invention unless otherwise claimed. No language in the specification should be construed as indicating any non-claimed element as essential to the practice of the invention. Preferred embodiments of this invention are described herein, including the best mode known to the inventors for carrying out the invention. Variations of those preferred embodiments may become apparent to those of ordinary skill in the art upon reading the foregoing description. The inventors expect skilled artisans to employ such variations as appropriate, and the inventors intend for the invention to be practiced otherwise than as specifically described herein. Accordingly, this invention includes all modifications and equivalents of the subject matter recited in the claims appended hereto as permitted by applicable law. Moreover, any combination of the above-described elements in all possible variations thereof is encompassed by the invention unless otherwise indicated herein or otherwise clearly contradicted by context.
039649660	description	DESCRIPTION OF THE INVENTION The accompanying drawing shows only so much of a liquid-metalcooled fast breeder reactor as is necessary to illustrate the setting of the present invention, the complete reactor being disclosed in U.S. Patent application Ser. No. 503,149, filed Sept. 4, 1974. The reactor includes a generally cylindrical pressure vessel 10 closed at the bottom by a bell 11 which bounds an inlet plenum 12. The vessel 10 has a plurality of inlet nozzles 13 above the bell 11 through which a heat-exchange fluid such as liquid sodium is supplied to the plenum 12 under pressure. The reactor core (not shown) is surrounded by a core barrel 14 and rests on support plate 15 which is supported from vessel 10 by conical skirt 16. A core catcher 17 according to the present invention of diameter greater than the reactor core is disposed in the pressure vessel 10 within inlet plenum 12 below the reactor core. Core catcher 17 includes a horizontal, circular baffle plate 18 having a plurality of openings 19 distributed in concentric circles about a central opening. Vertical heat transfer tubes 20 (4-inch, Type 304 stainless steel, schedule 5 pipe) are mounted in each of said openings by roll forming and welding. It will be noted that the flared end 21 of the tubes extends a short distance below the bottom of the baffle plate and that roll 21A is above the top of the baffle plate. The double attachment gives added assurance that a tube will never break loose during operation. Heat transfer tubes 20 are each drilled with three 2-inch diameter flow holes 22 near the end thereof and butt welded to an end cap 23 to close off the end. Baffle plate 18 is welded at its periphery to an imperforate, cylindrical baffle 24 to provide edge support for the plate. Core catcher 17 is supported from core support plate 15 by six uniformly spaced hanger rods 25, the lower end of each of which is welded to one of six radial beams 26 joined at the center by full-penetration welding to a hub 27. Radial beams 26 are intermittently welded to the bottom surface of baffle plate 18, channels 28 in the top of the beams accommodating the flared ends 21 of the vertical tubes. The end of radial beams 26 and the hub 27 are also depressed sufficiently to accommodate the flared end 21 of a central heat transfer tube 20. Radial beams 26 provide support for baffle plate 18 in bending. Hanger rods 25 are attached to core support plate 15 by means of retainers 29 such that up and down loads can be transmitted. Hanger rods 25 also provide lateral support for the core catcher by transmitting lateral loads by shear into the core support plate. Radial motions that may develop in the core catcher during transient phases of operation are allowed to occur by bearings 30 having slots 31 therein which are welded to outwardly extending lip 32 on cylindrical baffle 24 within notches 33 therein. By allowing the relative motion to occur between the core catcher and the hanger rod, thermal stresses are minimized. Important features of this invention follow: a. Containment volume can handle a complete core, even if the debris falls by a single stream through the core support plate. PA1 b. Adequate cooling of the heat-generating debris provided by natural insulation with no forced flow required. PA1 c. No device external to the reactor vessel is required. PA1 d. No attachments or modifications to the reactor vessel are required. PA1 c. Fabrication can be accomplished using conventional techniques. PA1 f. The core catcher provides for the accommodation of thermal expansion during transients. PA1 g. Criticality of debris is prevented by spatial separation. PA1 h. A screening effect is provided by the baffle assembly to eliminate possible flow blockage near the core inlet. In the extremely remote chance that an accident occurs which causes all or a part of the fuel to become molten, the fuel will cascade down through the core support plate and onto the core catcher between the heat-exchange tubes of the present invention wherein sodium continuing to flow through heat-exchange tubes by natural convection will cool the molten fuel and the heat-exchange tubes interspersed in the mass of molten fuel will eliminate any possibility that the fuel can attain a critical mass.
	summary
055240410	abstract	A collimator assembly for removing selected radiation output from a specimen. The assembly includes collimator elements with each element having walls comprised of a first material covered by an inner layer of a second material which preferentially absorbs inelastic scattered radiation created in the first material.
	summary
	claims	1. A bolus that is disposed in a particle beam therapy system and modulates energy distribution of a particle beam in accordance with a to-be-irradiated portion,wherein the shape of the bolus is set in such a way that, when a first reference point and a second reference point, which is at the downstream side of the first reference point, are given on a beam axis of a particle beam that enters the bolus at the upstream side thereof, and when an irradiation orbit of a particle beam that penetrates the bolus and reaches the to-be-irradiated portion is defined by a first slant from the beam axis with respect to a first axis that starts from the first reference point, that is perpendicular to the beam axis, and that includes the first reference point and by a second slant from the beam axis with respect to a second axis that is perpendicular to the beam axis and the first axis and includes the second reference point, a path length within the bolus of a particle beam in each of the irradiation orbits defined for combinations within a predetermined range of combinations of the first slant and the second slant, compensates a path length from a body surface situated at the upstream side of the to-be-irradiated portion to the to-be-irradiated portion. 2. The bolus according to claim 1, wherein the shape of the bolus is set in such a way that the following relationship is satisfied:LB(α, β)+LK(α, β)=R whereα denotes the first slant,β denotes the second slant,LB(α, β) denotes the path length, of a particle beam, within the bolus in an irradiation orbit defined by a combination of the first slant and the second slant,LK(α, β) denotes the path length from the body surface to the to-be-irradiated portion in the irradiation orbit defined by the combination of the first slant and the second slant, andR denotes the attainable depth corresponding to the energy of a particle beam that enters the bolus, respectively. 3. A bolus manufacturing method for manufacturing the bolus according to claim 2, comprising the steps of:acquiring inner-body depth data, which is the path length from the body surface to the to-be-irradiated portion, for each of combinations of the first slant and the second slant;setting the shape of a bolus in such a way that the path length is obtained by compensating the acquired inner-body depth data;creating bolus machining data, based on the set bolus shape; andmachining a bolus, based on the created machining data. 4. A particle beam therapy system comprising:an irradiation nozzle that scans a particle beam supplied from an accelerator by means of two electromagnets that range in the traveling direction of the particle beam and whose scanning directions are different from each other, and that irradiates the particle beam in such a way as to enlarge the irradiation field thereof; andthe bolus according to claim 2, disposed in a particle beam irradiated from the irradiation nozzle, wherein the bolus is disposed in such a way that the first axis for setting the shape of the bolus coincides with the scanning axis of the upstream electromagnet out of the two electromagnets and the second axis coincides with the scanning axis of the other electromagnet. 5. The particle beam therapy system according to claim 4, wherein the irradiation nozzle enlarges the irradiation field utilizing a spiral wobbling method. 6. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 5, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 7. The particle beam therapy system according to claim 4, wherein the irradiation nozzle enlarges the irradiation field utilizing a scanning method. 8. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 7, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 9. The particle beam therapy system according to claim 4, wherein scanning for one direction out of the two directions is performed by a deflection electromagnet that deflects the direction of a beam axis, and by regarding that the scanning axis of the deflection electromagnet passes through a point on the beam axis of a particle beam that enters the bolus, the scanning axis of the deflection electromagnet is made to coincide with the first axis or the second axis. 10. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 9, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 11. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 4, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 12. A bolus manufacturing method for manufacturing the bolus according to claim 1, comprising the steps of:acquiring inner-body depth data, which is the path length from the body surface to the to-be-irradiated portion, for each of combinations of the first slant and the second slant;setting the shape of a bolus in such a way that the path length is obtained by compensating the acquired inner-body depth data;creating bolus machining data, based on the set bolus shape; andmachining a bolus, based on the created machining data. 13. A particle beam therapy system comprising:an irradiation nozzle that scans a particle beam supplied from an accelerator by means of two electromagnets that range in the traveling direction of the particle beam and whose scanning directions are different from each other, and that irradiates the particle beam in such a way as to enlarge the irradiation field thereof; andthe bolus according to claim 1, disposed in a particle beam irradiated from the irradiation nozzle, wherein the bolus is disposed in such a way that the first axis for setting the shape of the bolus coincides with the scanning axis of the upstream electromagnet out of the two electromagnets and the second axis coincides with the scanning axis of the other electromagnet. 14. The particle beam therapy system according to claim 13, wherein the irradiation nozzle enlarges the irradiation field utilizing a spiral wobbling method. 15. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 14, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 16. The particle beam therapy system according to claim 13, wherein the irradiation nozzle enlarges the irradiation field utilizing a scanning method. 17. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 16, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 18. The particle beam therapy system according to claim 13, wherein scanning for one direction out of the two directions is performed by a deflection electromagnet that deflects the direction of a beam axis, and by regarding that the scanning axis of the deflection electromagnet passes through a point on the beam axis of a particle beam that enters the bolus, the scanning axis of the deflection electromagnet is made to coincide with the first axis or the second axis. 19. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 18, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet. 20. A treatment planning apparatus comprising:a three-dimensional data creation unit for creating three-dimensional data from image data of a body including the to-be-irradiated portion;an irradiation condition setting unit for setting an irradiation condition, based on the created three-dimensional data; anda bolus data creation unit for creating shape data on a bolus in the particle beam therapy system according to claim 13, based on the set irradiation condition, wherein the three-dimensional data creation unit creates the three-dimensional data by utilizing at least the first slant corresponding to beam deflection angle by the upstream electromagnet and the second slant corresponding to beam deflection angle by said the other electromagnet.
039727726	abstract	In a nuclear power station having a pressurized water reactor, blowdown water which is produced in the steam generators due to an accumulation of existing impurities, is not discarded but is returned to the circulation system through an electromagnetic filter, in combination with a mixed bed filter.
047327294	abstract	In the fast breeder reactor according to this invention, a medium-pressure plenum and a high-pressure plenum are disposed, in the mentioned order from above in a reactor core. Sodium is filled in the medium- and high-pressure plenums. The pressure in the medium-pressure plenum is lower than that in the high-pressure plenum. The lower ends of the entrance nozzles of a multiplicity of fuel assemblies loaded in the reactor core are supported by a supporting plate which forms a boundary between the medium- and high-pressure plenums. The entrance nozzles of control rod guide pipes disposed among the multiplicity of fuel assemblies are also supported by said supporting plate. The sodium in the medium-pressure plenum is supplied into each control rod guide pipe through a first opening provided in the lateral surface of the entrance nozzle of each control rod guide pipe. The sodium in the high-pressure plenum is supplied into each fuel assembly from the lower end of the entrance nozzle of each fuel assembly through a second opening provided piercing through the supporting plate. The sodium in the high-pressure plenum is supplied to the medium-pressure plenum through a pressure-reducing orifice.
	description	This application is a continuation application of International Application No. PCT/CN2018/100787, filed on Aug. 16, 2018, which claims priority to Chinese Patent Application No. 201711347618.5, filed on Dec. 15, 2017, and Chinese Patent Application No. 201721763785.3, filed on Dec. 15, 2017, the disclosures of which are hereby incorporated by reference. The present disclosure relates to a radioactive irradiation system, and more particularly to a neutron capture therapy system. The background description provided herein is for the purpose of generally presenting the context of the disclosure. Work of the presently named inventors, to the extent it is described in this background section, as well as aspects of the description that may not otherwise qualify as prior art at the time of filing, are neither expressly nor impliedly admitted as prior art against the present disclosure. As atomics moves ahead, radiotherapy such as Cobalt-60 therapy, linear accelerator therapy and electron beam therapy has been one of the major approaches to cancer treatment. However, conventional photon or electron therapy has undergone physical restrictions of radioactive rays. For example, a large amount of normal tissue on a beam path is damaged as tumor cells are killed. Moreover, tumor cells have different radiosensitivities, and as a result conventional radiotherapy falls short of treatment effectiveness on radioresistant malignant tumors (such as glioblastoma multiforme and melanoma). To reduce radiation damage to the normal tissue surrounding a tumor, targeted therapy in chemotherapy has been applied to radiotherapy. For highly radioresistant tumor cells, proton therapy, heavy particle therapy, neutron capture therapy, and the like using a radiation source with high relative biological effectiveness (RBE) are being actively developed at present. Among them, the neutron capture therapy combines the target therapy with the RBE. For example, the boron neutron capture therapy (BNCT). The boron neutron capture therapy provide a better cancer treatment option than conventional radiotherapy by specific grouping of boron-containing drugs in tumor cells is combined with precise neutron beam regulation to. The effect of BNCT depends on the concentration of boron-containing drugs and the quantity of thermal neutrons in the tumor cells, so it is also referred to as binary cancer therapy. It may be seen that in addition to the development of boron-containing drugs, the improvement in the fluxes and quality of neutron sources plays a significant role in the research of BNCT. In addition, a variety of radioactive rays are generated during radiotherapy, for example, low-energy to high-energy neutrons and photons, these radioactive rays may cause different levels of damage to normal tissue of a human body. Therefore, in the field of radiotherapy, how to provide effective treatment while reducing radiation pollution to external environment, medical workers and normal tissue of a patient is a vital subject. Therefore, it is necessary to propose a new technical solution to resolve the foregoing problem. To resolve the foregoing problem, one aspect of the present disclosure provides a neutron capture therapy system. The neutron capture therapy system includes a beam shaping assembly and a vacuum tube disposed in the beam shaping assembly. the beam shaping assembly, comprising a beam entrance, an accommodating cavity for accommodating the vacuum tube, a moderator adjacent to an end of the accommodating cavity, a reflector surrounding the moderator, a radiation shield disposed in the beam shaping assembly, and a beam exit, wherein the moderator moderates neutrons generated from a target to an epithermal neutron energy range, the reflector guides deflecting neutrons back to the moderator to enhance an intensity of an epithermal neutron beam, the radiation shield is configured to shield against leaked neutrons and photons to reduce a dose to a normal tissue in a non-irradiation area. The target is disposed at an end of the vacuum tube, the target undergoes a nuclear reaction with a charged particle beam entering through the beam entrance to generate neutrons, the neutrons form a neutron beam, and the neutron beam is emitted from the beam exit and defines a neutron beam axis. The moderator at least comprises two cylindrical moderating members with different outer diameters respectively, the moderator has a first end close to the beam entrance and a second end close to the beam exit, and the target is accommodated between the first end and the second end. Compared with the related art, the technical solution disclosed in this embodiment has the following beneficial effects: The moderator at least includes two cylindrical moderating members with different outer diameters respectively, and the target is accommodated in the moderator, which can reduce the material costs, greatly reduce the intensity of fast neutrons, and improve the neutron beam quality. Preferably, the moderator comprises a first moderating unit close to the beam entrance and a second moderating unit closely attached to the first moderating unit and close to the beam exit, the first moderating unit at least comprises two cylindrical moderating members with different outer diameters respectively, wherein all of the beam entrance, the moderator and the beam exit are extended along the neutron beam axis, and wherein a distance from the target to the beam exit is less than a distance from the first end to the beam exit. Further, the first moderating unit comprises three cylindrical moderating members with different outer diameters respectively, the first moderating unit comprises a first moderating portion close to the beam entrance, a second moderating portion closely attached to the first moderating portion, and a third moderating portion closely attached to the second moderating portion, the first moderating portion, the second moderating portion and the third moderating portion are sequentially arranged along an axial direction of the neutron beam, the first moderating portion defines a first outer diameter, the second moderating portion defines a second outer diameter greater than the first outer diameter, the third moderating portion defines a third outer diameter greater than the second outer diameter, the second moderating unit defines a fourth outer diameter equal to the third outer diameter. Preferably, the first moderating portion comprises a first front end surface close to the beam entrance, a first rear end surface close to the beam exit and a first outer circumferential surface, the second moderating portion comprises a second front end surface closely attached to the first rear end surface, a second rear end surface close to the beam exit and a second outer circumferential surface, the third moderating portion comprises a third front end surface closely attached to the second rear end surface, a third rear end surface close to the beam exit and a third outer circumferential surface, the second moderating unit comprises a fourth front end surface closely attached to the third rear end surface, a fourth rear end surface close to the beam exit and a fourth outer circumferential surface, in the tangent surface passing through the neutron beam axis, the first front end surface intersects the first outer circumferential surface to obtain a first intersection point, the second front end surface intersects the second outer circumferential surface to obtain a second intersection point, the third front end surface intersects the third outer circumferential surface to obtain a third intersection point, and the first intersection point, the second intersection point and the third intersection point are located on a same straight line or one arc lin. Further, a reflection compensator is filled between the accommodating cavity and the vacuum tube, and the reflection compensator is lead or Al or Teflon or C. Further, the first end protrudes from the target along the neutron beam axis in a direction towards the beam entrance, and the second end protrudes from the target along the neutron beam axis in a direction towards the beam exit. Further, the reflector protrudes from the moderator on both sides of the neutron beam axis, the accommodating cavity comprises a reflector accommodating cavity surrounded by the reflector and a moderator accommodating cavity extending from the reflector accommodating cavity and surrounded by the moderator, the vacuum tube comprises an extending section accommodated in the reflector accommodating cavity and an insertion section extending from the extending section and accommodated in the moderator accommodating cavity, and the target is disposed at an end of the insertion section. Further, the neutron capture therapy system further comprises at least one cooling device, at least one accommodating pipe disposed in the beam shaping assembly for accommodating the cooling device and a lead alloy or an aluminum alloy filled between the cooling device and an inner wall of the accommodating pipe. Further, the neutron capture therapy system further comprises a shielding assembly disposed at the beam entrance and closely attached to the beam shaping assembly. Further, a cross section of the second moderating unit is conical or cylindrical or steped-shaped. Further, a depth of the target entering into the moderator is less than or equal to a length of the first moderating unit in an axial direction of the neutron beam. Further, the cooling device comprises a first cooling portion arranged in a vertical direction and located in front of the target for cooling the target and a second cooling portion and a third cooling portion extending in an axial direction of the neutron beam and respectively located on two sides of the vacuum tube, the first cooling portion is connected between the second cooling portion and the third cooling portion, the second cooling portion inputs a cooling medium into the first cooling portion, and the third cooling portion outputs the cooling medium in the first cooling portion. In another aspect of the present disclosure provides a neutron capture therapy system. The neutron capture therapy system includes a beam shaping assembly, a vacuum tube disposed in the beam shaping assembly and a target disposed at an end of the vacuum tube. The target undergoes a nuclear reactions with a charged particle beam entering through the beam entrance to generate neutrons. The beam shaping assembly includes a beam entrance, an accommodating cavity for accommodating the vacuum tube, a moderator adjacent to an end of the accommodating cavity, a reflector surrounding the moderator and a beam exit. The moderator at least comprises two hollow cylindrical moderating members with different outer diameters and same inner diameter respectively. Further, a reflection compensator is filled between the accommodating cavity and the vacuum tube, and the reflection compensator is lead or Al or Teflon or C. Further, the neutron capture therapy system further comprises at least one cooling device, at least one accommodating pipe disposed in the beam shaping assembly for accommodating the cooling device, and a lead alloy or an aluminum alloy is filled between the cooling device and an inner wall of the accommodating pipe. Further, the moderator comprises a first end close to the beam entrance and a second end close to the beam exit, and the target is accommodated between the first end and the second end. In yet another aspect of the present disclosure provides a neutron capture therapy system. The neutron capture therapy system includes a beam shaping assembly. The a beam shaping assembly includes a beam entrance, a moderator, a reflector surrounding the moderator, and a beam exit. The moderator at least comprises two hollow cylindrical moderating members with different outer diameters respectively. Further, the moderator comprises a first moderating unit close to the beam entrance and a second moderating unit closely attached to the first moderating unit and close to the beam exit, the first moderating unit at least comprises two hollow cylindrical moderating members with different outer diameters respectively. Further, the neutron capture therapy system further comprises at least one cooling device, at least one accommodating pipe disposed in the beam shaping assembly for accommodating the cooling device and a lead alloy or an aluminum alloy filled between the cooling device and an inner wall of the accommodating pipe. Further, the neutron capture therapy system further comprises a vacuum tube disposed in the beam shaping assembly, the beam shaping assembly further comprises an accommodating cavity for accommodating the vacuum tube, a reflection compensator is filled between the accommodating cavity and the vacuum tube, and the reflection compensator is lead or Al or Teflon or C. The “cone” or “conical body” in the embodiments of the present disclosure is a structure with an overall outer contour gradually becoming smaller from one side to the other in a direction in the drawings. The entire surface of the outer contour may have a smooth transition or may have nonsmooth transition. For example, many protrusions and grooves are provided in the surface of the conical body. The embodiments of the present disclosure are further described in detail below with reference to the accompanying drawings, so that those skilled in the art can implement the technical solutions according to the description. Neutron capture therapy (NCT) has been increasingly practiced as an effective cancer curing means in recent years, and BNCT is the most common. Neutrons for NCT may be supplied by nuclear reactors or accelerators. Take AB-BNCT for example, its principal components comprise, in general, an accelerator for accelerating charged particles (such as protons and deuterons), a target, a heat removal system and a beam shaping assembly. The accelerated charged particles interact with the metal target to produce the neutrons, and suitable nuclear reactions are always determined according to such characteristics as desired neutron yield and energy, available accelerated charged particle energy and current and materialization of the metal target, among which the most discussed two are 7Li (p, n)7Be and 9Be (p, n)9B and both are endothermic reaction. Their energy thresholds are 1.881 MeV and 2.055 MeV respectively. Epithermal neutrons at a keV energy level are considered ideal neutron sources for BNCT. Theoretically, bombardment with lithium target using protons with energy slightly higher than the thresholds may produce neutrons relatively low in energy, so the neutrons may be used clinically without many moderations. However, Li (lithium) and Be (beryllium) and protons of threshold energy exhibit not high action cross section. In order to produce sufficient neutron fluxes, high-energy protons are usually selected to trigger the nuclear reactions. The target, considered perfect, is supposed to have the advantages of high neutron yield, a produced neutron energy distribution near the epithermal neutron energy range (see details thereinafter), little strong-penetration radiation, safety, low cost, easy accessibility, high temperature resistance etc. But in reality, no nuclear reactions may satisfy all requests. The target in these embodiments of the present disclosure is made of lithium. However, well known by those skilled in the art, the target materials may be made of other metals besides the above-mentioned. Requirements for the heat removal system differ as the selected nuclear reactions. 7Li (p, n)7Be asks for more than 9Be (p, n)9B does because of low melting point and poor thermal conductivity coefficient of the metal (lithium) target. In these embodiments of the present disclosure is 7Li (p, n)7Be. It may be seen that the temperature of the target that is irradiated by an accelerated charged particle beam at a high energy level inevitably rises significantly, and as a result the service life of the target is affected. No matter BNCT neutron sources are from the nuclear reactor or the nuclear reactions between the accelerator charged particles and the target, only mixed radiation fields are produced, that is, beams comprise neutrons and photons having energies from low to high. As for BNCT in the depth of tumors, except the epithermal neutrons, the more the residual quantity of radiation ray is, the higher the proportion of nonselective dose deposition in the normal tissue is. Therefore, radiation causing unnecessary dose should be lowered down as much as possible. Besides air beam quality factors, dose is calculated using a human head tissue prosthesis in order to understand dose distribution of the neutrons in the human body. The prosthesis beam quality factors are later used as design reference to the neutron beams, which is elaborated hereinafter. The International Atomic Energy Agency (IAEA) has given five suggestions on the air beam quality factors for the clinical BNCT neutron sources. The suggestions may be used for differentiating the neutron sources and as reference for selecting neutron production pathways and designing the beam shaping assembly, and are shown as follows: Epithermal neutron flux >1×109 n/cm2 s Fast neutron contamination <2×10−13 Gy-cm2/n Photon contamination <2×10−13 Gy-cm2/n Thermal to epithermal neutron flux ratio <0.05 Epithermal neutron current to flux ratio >0.7 Note: the epithermal neutron energy range is between 0.5 eV and 40 keV, the thermal neutron energy range is lower than 0.5 eV, and the fast neutron energy range is higher than 40 keV. 1. Epithermal Neutron Flux The epithermal neutron flux and the concentration of the boronated pharmaceuticals at the tumor site codetermine clinical therapy time. If the boronated pharmaceuticals at the tumor site are high enough in concentration, the epithermal neutron flux may be reduced. On the contrary, if the concentration of the boronated pharmaceuticals in the tumors is at a low level, it is required that the epithermal neutrons in the high epithermal neutron flux should provide enough doses to the tumors. The given standard on the epithermal neutron flux from IAEA is more than 109 epithermal neutrons per square centimeter per second. In this flux of neutron beams, therapy time may be approximately controlled shorter than an hour with the boronated pharmaceuticals. Thus, except that patients are well positioned and feel more comfortable in shorter therapy time, and limited residence time of the boronated pharmaceuticals in the tumors may be effectively utilized. 2. Fast Neutron Contamination Unnecessary dose on the normal tissue produced by fast neutrons are considered as contamination. The dose exhibit positive correlation to neutron energy, hence, the quantity of the fast neutrons in the neutron beams should be reduced to the greatest extent. Dose of the fast neutrons per unit epithermal neutron flux is defined as the fast neutron contamination, and according to IAEA, it is supposed to be less than 210−13Gy-cm2/n. 3. Photon Contamination (Gamma-Ray Contamination) Gamma-ray long-range penetration radiation will selectively result in dose deposit of all tissues in beam paths, so that lowering the quantity of gamma-ray is also the exclusive requirement in neutron beam design. Gamma-ray dose accompanied per unit epithermal neutron flux is defined as gamma-ray contamination which is suggested being less than 210−13Gy-cm2/n according to IAEA. 4. Thermal to Epithermal Neutron Flux Ratio The thermal neutrons are so fast in rate of decay and poor in penetration that they leave most of energy in skin tissue after entering the body. Except for skin tumors like melanocytoma, the thermal neutrons serve as neutron sources of BNCT, in other cases like brain tumors, the quantity of the thermal neutrons has to be lowered. The thermal to epithermal neutron flux ratio is recommended at lower than 0.05 in accordance with IAEA. 5. Epithermal Neutron Current to Flux Ratio The epithermal neutron current to flux ratio stands for beam direction, the higher the ratio is, the better the forward direction of the neutron beams is, and the neutron beams in the better forward direction may reduce dose surrounding the normal tissue resulted from neutron scattering. In addition, treatable depth as well as positioning posture is improved. The epithermal neutron current to flux ratio is better of larger than 0.7 according to IAEA. To reduce the manufacturing costs of a beam shaping assembly of a neutron capture therapy system and obtain relatively good neutron beam quality, referring to FIG. 1, a first embodiment of the present disclosure provides a neutron capture therapy system 1. The neutron capture therapy system 1 includes a beam shaping assembly 10, a cooling device 20 disposed in the beam shaping assembly 10, a vacuum tube 30, and a shielding assembly 40 disposed outside of the beam shaping assembly 10 and closely attached to the beam shaping assembly 10. As shown in FIG. 1 and FIG. 2, the beam shaping assembly 10 includes a beam entrance 11, an accommodating cavity 12 configured to accommodate the vacuum tube 30, an accommodating pipe 13 configured to accommodate the cooling device 20, a moderator 14 adjacent to an end of the accommodating cavity 12, a reflector 15 surrounding the moderator 14, a thermal neutron absorber 16 adjacent to the moderator 14, a radiation shield 17 disposed in the beam shaping assembly 10, and a beam exit 18. A target 31 is disposed at an end of the vacuum tube 30. A nuclear reaction occur between the target 31 and a charged particle beam that enters through the beam entrance 11 and passes through the vacuum tube 30 to generate neutrons, the neutrons form the neutron beam, and the neutron beam is emitted from the beam exit 18 and defines a neutron beam axis X1 that coincides with the central axis of the vacuum tube 30. The moderator 14 moderates the neutrons generated from the target 31 to an epithermal neutron energy range, and the reflector 15 leads neutrons deflected from the neutron beam axis X1 back to the moderator 14 to enhance the intensity of an epithermal neutron beam. The reflector 15 protrudes from the moderator 14 on both sides of the neutron beam axis X1. The thermal neutron absorber 16 is configured to absorb thermal neutrons to protect superficial normal tissue from an overdose during treatment. The radiation shield 17 is configured to shield against leaked neutrons and photons to reduce a dose to a normal tissue in a non-irradiation area. In an accelerator-based neutron capture therapy system, an accelerator accelerates the charged particle beam. In one embodiment, the target 31 is made of lithium metal. Specifically, the target 31 is made of lithium metal in which the content of 7Li is 98% and the content of 6Li is 2%. The charged particle beam is accelerated to an energy sufficient to overcome the energy of the coulomb repulsion of atomic nuclei of the target, and the 7Li (p, n)7Be nuclear reaction occurs between the charged particle beam and the target 31 to generate neutrons. The beam shaping assembly 10 can moderate the neutrons to the epithermal neutron energy range, and reduce the content of thermal neutrons and fast neutrons. The moderator 14 is made of a material with a large fast neutron reaction cross section and a small epithermal neutron reaction cross section, the reflector 15 is made of a material with high neutron reflectivity, the thermal neutron absorber 16 is made of a material with a large thermal neutron reaction cross section. In one embodiment, the moderator 14 is made of a mixture of MgF2 and LiF that is accounts for 4.6% of MgF2 by weight, the reflector 15 is made of Pb, and the thermal neutron absorber 16 is made of 6Li. The radiation shield 17 includes a photon shielding 171 and a neutron shielding 172. In one embodiment, the photon shielding 171 made of lead (Pb) and the neutron shielding 172 made of polyethylene (PE). As shown in FIG. 7, the target 31 includes a lithium target layer 311 and an antioxidation layer 312 that is located on a side of the lithium target layer 311 and is configured to prevent the lithium target layer 311 from oxidating. The antioxidation layer 312 of the target 31 is made of Al or stainless steel. As shown in FIG. 1 and FIG. 2, the moderator 14 includes a first moderating unit 140 close to the beam entrance 11 and a second moderating unit 144 that is closely attached to the first moderating unit 140 and is close to the beam exit 18. The moderator 14 has a first end 146 close to the beam entrance 11, a second end 148 close to the beam exit 18, and a third end 147 that is located between the first end 146 and the second end 148. The third end 147 is located between the first moderating unit 140 and the second moderating unit 144. The beam entrance 11, the moderator 14, and the beam exit 18 all extend along the neutron beam axis X1. A distance from the target 31 to the beam exit 18 is less than a distance from the first end 146 to the beam exit 18. In other words, the first end 146 protrudes from the target 31 along a neutron beam axis X1 in a direction towards the beam entrance 11, and the second end 148 protrudes from the target 31 along the neutron beam axis X1 in a direction towards the beam exit 18. The first moderating unit 140 includes at least two hollow cylindrical moderating members with different outer diameters and the same inner diameter, respectively. Referring to FIG. 1, FIG. 2, FIG. 6, and FIG. 8, in the first embodiment, a third embodiment, and a fourth embodiment, the first moderating unit 140 includes three hollow cylindrical moderating members with different outer diameters and the same inner diameter, respectively. The first moderating unit 140 and the second moderating unit 144 are formed by stacking and splicing several moderating members formed from molds with corresponding sizes and then subjected to processes such as polishing, grinding, Specifically, the first moderating unit 140 includes a first moderating portion 141 close to the beam entrance 11, a second moderating portion 142 that is located on the right side of the first moderating portion 141 and closely attached to the first moderating portion 141, and a third moderating portion 143 that is located on the right side of the second moderating portion 142 and closely attached to the second moderating portion 142. That is, the first moderating portion 141, the second moderating portion 142, and the third moderating portion 143 are sequentially arranged along a direction of the neutron beam axis X1. The first moderating portion 141 defines a first outer diameter, the second moderating portion 142 defines a second outer diameter greater than the first outer diameter, the third moderating portion 143 defines a third outer diameter greater than the second outer diameter, the second moderating unit 144 defines a fourth outer diameter equal to the third outer diameter, and inner diameters of the first moderating portion 141, the second moderating portion 142, and the third moderating portion 143 are equal. The central axes of the first moderating portion 141, the second moderating portion 142, and the third moderating portion 143 coincide with the centerline of the second moderating unit 144. The central axes also coincide with the neutron beam axis X1. The first moderating portion 141 has a first front end surface 1411 located on the left side, a first rear end surface 1412 located on the right side, a first outer circumferential surface 1413, and a first inner circumferential surface 1414. The second moderating portion 142 has a second front end surface 1421 located on the left side, a second rear end surface 1422 located on the right side, a second outer circumferential surface 1423, and a second inner circumferential surface 1424. The third moderating portion 143 has a third front end surface 1431 located on the left side, a third rear end surface 1432 located on the right side, a third outer circumferential surface 1433, and a third inner circumferential surface 1434. The second moderating unit 144 has a fourth front end surface 1441 located on the left side, a fourth rear end surface 1442 located on the right side, and a fourth outer circumferential surface 1443. All of the first front end surface 1411, the second front end surface 1421, the third front end surface 1431, the fourth front end surface 1441, the first rear end surface 1412, the second rear end surface 1422, the third rear end surface 1432, and the fourth rear end surface 1442 are parallel to each other and are perpendicular to the neutron beam axis X1. The first rear end surface 1412 of the first moderating portion 141 is closely attached to the second front end surface 1421 of the second moderating portion 142, the second rear end surface 1422 of the second moderating portion 142 is closely attached to the third front end surface 1431 of the third moderating portion 143, and the third rear end surface 1432 of the third moderating portion 143 is closely attached to the fourth front end surface 1441 of the second moderating unit 144. An intersection line of the tangent surface passing through the neutron beam axis X1 and the first outer circumferential surface 1413 is perpendicular to the second front end surface 1421, an intersection line of the tangent surface passing through the neutron beam axis X1 and the second outer circumferential surface 1423 is perpendicular to the third front end surface 1431, and there is smooth transition between the third outer circumferential surface 1433 of the third moderating portion 143 and the fourth outer circumferential surface 1443 of the second moderating unit 144. As shown in FIG. 2, in the tangent surface passing through the neutron beam axis X1, the first front end surface 1411 intersects the first outer circumferential surface 1413 of the first moderating portion 141 to obtain a first intersection point 1410, the second front end surface 1421 of the second moderating portion 142 intersects the second outer circumferential surface 1423 to obtain a second intersection point 1420, the third front end surface 1431 of the third moderating portion 143 intersects the third outer circumferential surface 1433 to obtain a third intersection point 1430, the first intersection point 1410, the second intersection point 1420, and the third intersection point 1430 are located on a same straight line X2, and an angle between the straight line X2 and the neutron beam axis X1 is less than 90 degrees. The reflector 15 has an inner surface 150 surrounding the moderator 14, and the inner surface 150 is closely attached to the first front end surface 1411, the first outer circumferential surface 1413, the second front end surface 1421, the second outer circumferential surface 1423, the third front end surface 1431, the third outer circumferential surface 1433, the fourth rear end surface 1442, and the fourth outer circumferential surface 1443 of the moderator 14. As shown in FIG. 1, FIG. 2, FIG. 6, and FIG. 8, in the first embodiment, the third embodiment, and the fourth embodiment, the first moderating unit 140 includes three concentric hollow cylindrical moderating members with different outer diameters and the same inner diameter, respectively. As observed along a direction perpendicular to the neutron beam axis X1, outer contours of the first moderating portion 141, the second moderating portion 142, and the third moderating portion 143 are combined to form a step shape, therefore, the first moderating unit 140 is named a three-step moderator. As shown in FIG. 10 to FIG. 12, from sixth embodiment to eighth embodiment, the first moderating unit 140 includes two, four, ten hollow cylindrical moderating members with different outer diameters and the same inner diameter, respectively. That is, the first moderating unit 140 may be a two-step moderator, four-step moderator, ten-step moderator. In another embodiment, the first moderating unit 140 may further includes another quantity of hollow cylindrical moderating members with different outer diameters and the same inner diameter, respectively. For example, twelve cylindrical moderating members, fifteen cylindrical moderating members, and the like. In another embodiment, the second moderating unit 144 may also be disposed to be a steped-shaped moderator. Alternatively, a polygonal prism may be used in place of a cylinder to form the moderator. In addition, the first intersection point 1410, the second intersection point 1420, and the third intersection point 1430 may be located on one arc line instead of the straight line X2. In addition, according to an actual requirement, the moderating portion that form the first moderating unit 140 may be disposed to have a partially non-hollow structure. The central axes of the moderating portion of the first moderating unit 140 may not coincide with the central axis of the second moderating unit 144. Generally, the moderator is formed by stacking and splicing several moderating members that are formed from molds with corresponding sizes and then subjected to processes such as polishing, grounding. The moderator formed from the mold is disc-shaped. When the moderator is designed as an entire cylinder or cone, the volume of the moderator material consumed is a product of the size of the moderator in the direction of the neutron beam axis X1 and the bottom area of the disk. It should be noted that the conical moderator is obtained by grinding the cylindrical moderator. That is, the volume of material needed for the design of the moderator to be cylindrical or conical is the same. In the present disclosure, the first moderating unit 140 is designed to be a step-shaped moderator. On the premise that the size of the moderator in the direction of the neutron beam axis X1 and the maximum diameter of the moderator are remain unchanged, because the bottom areas of the disc-shaped moderator forming the steped-shaped moderator is gradually increasing, in the present disclosure, the moderator material needed when the moderator is designed to be the steped-shaped moderator is less than the material needed when the moderator is designed to be the entire cylindrical moderator or conical moderator. As can be learned, the steped-shaped moderator in the present disclosure can greatly reduce the material for manufacturing the moderator, thereby reducing the manufacturing cost. Referring to FIG. 2, the accommodating cavity 12 is a cylindrical cavity that is surrounded by the reflector 15 and the first moderating unit 140 of the moderator 14. The accommodating cavity 12 includes a reflector accommodating cavity 121 surrounded by the reflector 15 and a moderator accommodating cavity 122 extending from the reflector accommodating cavity 121 and surrounded by the first moderating unit 140 of the moderator 14. That is, the moderator accommodating cavity 122 is surrounded by the first inner circumferential surface 1414 of the first moderating portion 141, the second inner circumferential surface 1424 of the second moderating portion 142, and the third inner circumferential surface 1434 of the third moderating portion 143. The vacuum tube 30 includes an extending section 32 surrounded by the reflector 15 and an insertion section 34 extending from the extending section 32 and inserted into the moderator 14, the extending section 32 is accommodated in the reflector accommodating cavity 121, and the insertion section 34 is accommodated in the moderator accommodating cavity 122. The target 31 is disposed at an end of the insertion section 34 of the vacuum tube 30, and the end is flush with the third rear end surface 1432 of the first moderating unit 140. In the first to third embodiment and fifth to eighth embodiments, the vacuum tube 30 is partially inserted into the moderator 14, that is, the target 31 is disposed in the moderator 14. Mark the depth of the target 31 enters into the moderator 14 as X. The value of X is equal to the size of the moderator accommodating cavity 122 in the direction of the neutron beam axis X1, that is, the size of the first moderating unit 140 in the direction of the neutron beam axis X1. In another embodiment, the depth X of the target 31 entering into the moderator 14 may be less than or greater than the length of the first moderating unit 140 in the direction along the neutron beam axis X1. That is, along the direction of the neutron beam axis X1, the target 31 may be disposed to extend within the first moderating unit 140 or extend beyond the first moderating unit 140 and into the second moderating unit 144. Correspondingly, when the target 31 is disposed to extend within the first moderating unit 140 along the direction of the neutron beam axis X1, the first moderating unit 140 is disposed to have a partially non-hollow structure. When the target 31 is disposed to extend beyond the first moderating unit 140 and into the second moderating unit 144 along the direction of the neutron beam axis X1, the first moderating unit 140 is disposed to have a hollow structure, and the second moderating unit 144 is disposed to have a partially hollow structure. Referring to FIG. 1, FIG. 2, and FIG. 3, a gap exists between the accommodating cavity 12 and the vacuum tube 30, a reflection compensator 50 is filled in the gap, and the reflection compensator 50 is Pb or Al or Teflon or carbon that can absorb or reflect neutrons. The reflection compensator 50 can reflect neutrons reflected or scattered into the gap into the moderator 14 or the reflector 15, thereby increasing the intensity of epithermal neutrons and reducing the time that an irradiated body needs to be irradiated. In another aspect, it avoids leakage of neutrons to the outside of the beam shaping assembly 10 to cause adversely affect to the instruments of the neutron capture therapy system, and improves radiation safe. As shown in FIG. 1 and FIG. 2, the accommodating pipe 13 includes a second accommodating pipe 132 and a third accommodating pipe 133 that extend along the direction of the neutron beam axis X1 and are respectively located on two sides of the accommodating cavity 12 at 180° intervals and a first accommodating pipe 131 that is disposed in a plane perpendicular to the neutron beam axis X1 and is located between the target 31 and the moderator 14. The second accommodating pipe 132 and the third accommodating pipe 133 extend beyond the accommodating cavity 12 in the direction of the neutron beam axis X1 and communicate with the first accommodating pipe 131 respectively. That is, the first accommodating pipe 131 is located at an end of the accommodating cavity 12 and between the target 31 and the moderator 14, and the second accommodating pipe 132 and the third accommodating pipe 133 are respectively located on two sides of the accommodating cavity 12 and are respectively communicated with the first accommodating pipe 131, so that the accommodating pipe 30 is arranged in a “[”-shaped structure. Referring to FIG. 2, the second accommodating pipe 132 and the third accommodating pipe 133 respectively include a second reflector accommodating pipe 1321 and a third reflector accommodating pipe 1331 located on an outer side of the reflector accommodating cavity 121 and a second moderating unit accommodating pipe 1322 and a third moderator accommodating pipe 1332 extending from the second reflector accommodating pipe 1321 and the third reflector accommodating pipe 1331 and located on the outer side of the moderator accommodating cavity 122, respectively. In this embodiment, the second accommodating pipe 132 and the third accommodating pipe 133 extend in the direction along the neutron beam axis X1 and are parallel to the neutron beam axis X1. That is, an angle between the second accommodating pipe 132 and the third accommodating pipe 133 and the neutron beam axis X1 is 0°. In the first embodiment and the second embodiment, the second accommodating pipe 132 and the third accommodating pipe 133 are not in communication with the accommodating cavity 12, that is, the second accommodating pipe 132 and the third accommodating pipe 133 are separated from the accommodating cavity 12 by the reflector 15 and the moderator 14. In another embodiment, the second accommodating pipe 132 and the third accommodating pipe 133 may be in communication with the accommodating cavity 12, that is, an outer surface the vacuum tube 30 accommodated in the accommodating cavity 12 is partially exposed in the second accommodating pipe 132 and the third accommodating pipe 133. In conclusion, the second accommodating pipe 132 and the third accommodating pipe 133 are located outside an inner wall of the accommodating cavity 12. In this embodiment of the present disclosure, the second accommodating pipe 132 and the third accommodating pipe 133 are disposed to be arc-shaped pipes extending along an axial direction of the vacuum tube 30, in another embodiments, the second accommodating pipe 132 and the third accommodating pipe 133 may be disposed to be rectangular pipes, triangular pipes or another polygonal pipes. In this embodiment of the present disclosure, the second accommodating pipe 132 and the third accommodating pipe 133 are two independent accommodating pipes that are separated in a circumferential direction of the accommodating cavity 12. In another embodiments, the second accommodating pipe 132 and the third accommodating pipe 133 are in communication with each other in the circumferential direction of the accommodating cavity 12, that is, the second accommodating pipe 132 and the third accommodating pipe 133 are replaced with one accommodating pipe surrounding the accommodating cavity 12. As shown in FIG. 5, the cooling device 20 includes a first cooling portion 21 arranged in a vertical direction and located in front of the target 31 for cooling the target 31 and a second cooling portion 22 and a third cooling portion 23 extending in the direction of the neutron beam axis X1 and respectively located on two sides of the vacuum tube 30 and parallel to the neutron beam axis X1. The first cooling portion 21 is connected between the second cooling portion 22 and the third cooling portion 23. The first cooling portion 21 is accommodated in the first accommodating pipe 131 arranged in the direction perpendicular to the neutron beam axis X1, and the second cooling portion 22 and the third cooling portion 23 are respectively accommodated in the second accommodating pipe 132 and the third accommodating pipe 133 arranged in the direction of the neutron beam axis X1. The second cooling portion 22 inputs a cooling medium into the first cooling portion 21, and the third cooling portion 23 outputs the cooling medium in the first cooling portion 21. The first cooling portion 21 is located between the target 31 and the moderator 14. One side of the first cooling portion 21 is in direct contact with the target 31, and the other side of the first cooling portion 21 is in contact with the moderator 14. The second cooling portion 22 and the third cooling portion 23 respectively include a first cooling section 221 and a second cooling section 231 located on the outer side of the reflector accommodating cavity 121 and a third cooling section 222 and a fourth cooling section 232 respectively extending from the first cooling section 221 and the second cooling section 231 and located on the outer side of the moderator accommodating cavity 122. The third cooling section 222 and the fourth cooling section 232 are respectively in communication with the first cooling portion 21. That is, the first cooling portion 21 is located at an end of the insertion section 34 of the vacuum tube 30, and is located on a side of the target 31 and is in direct contact with the target 31, the second cooling portion 22 and the third cooling portion 23 are respectively located on an upper side and a lower side of the vacuum tube 30 accommodated in the accommodating cavity 12 and communicate with the first cooling portion 21, respectively, so that the entire cooling device 20 is disposed to a “[”-shaped structure. In this embodiment, the first cooling portion 21 is in plane contact with the target 31, the second cooling portion 22 and the third cooling portion 23 are both tubular structures made of copper, and the second cooling portion 22 and the third cooling portion 23 extend along the direction of the neutron beam axis X1 and are parallel to the neutron beam axis X1, that is, an angle between the neutron beam axis X1 and each of the second cooling portion 22 and the third cooling portion 23 is 0°. The first cooling portion 21 includes a first contact portion 211, a second contact portion 212, and a cooling groove 213 located between the first contact portion 211 and the second contact portion 212 for passing the cooling medium. The first contact portion 211 is in direct contact with the target 31, and the second contact portion 212 may be in direct contact or may be in indirect contact with the moderator 14 through air. The cooling groove 213 has a input groove 214 communicating with the second cooling portion 22 and a output groove 215 communicating with the third cooling portion 23. The first contact portion 211 is made of a thermally conductive material. An upper edge of the input groove 214 is located above an upper edge of the second cooling portion 22, and a lower edge of the output groove 215 is located below a lower edge of the third cooling portion 23. The benefit of this arrangement is that the cooling device 20 can feed cooling water into the cooling groove 213 more smoothly and cool the target 31 in time, the heated cooling water can also be output from the cooling groove 213 more smoothly, and moreover, the water pressure of cooling water in the cooling groove 213 can further be reduced to a particular degree. The first contact portion 211 is made of a thermally conductive material (a material such as Cu, Fe, and Al with high thermal conductivity) or a material that can both conduct heat and suppress foaming, the second contact portion 212 is made of a material that suppresses foaming. The material that suppresses foaming or the material that can both conduct heat and suppress foaming is made of any one of Fe, Ta or V. The target 31 is irradiated by accelerated particles at a high energy level, which causes a temperature rise and generate heat, the first contact portion 211 guides out the heat, and the cooling medium that flows in the cooling groove 213 takes away the heat to cool the target 31. In this embodiment, the cooling medium is water. Referring to FIG. 2, the shielding assembly 40 covers a left end surface of the beam shaping assembly 10 and is closely attached to the end surface to prevent a neutron beam and a γ ray formed at the target 31 from overflowing from the left end surface of the beam shaping assembly 10. The shielding assembly 40 includes Pb and PE. Specifically, the shielding assembly 40 includes at least two layers of Pb and at least one layer of PE. In this embodiment, the shielding assembly 40 includes a first Pb layer 41 closely attached to the left end surface of the beam shaping assembly 10, a PE layer 42 closely attached to the first Pb layer 41, and a second Pb layer 43 covering the PE layer 42 and closely attached to the PE layer 42. Pb can absorb the γ ray overflowing from the beam shaping assembly 10 and reflects neutrons overflowing from the beam shaping assembly 10 back to the moderator 14 to increase the intensity of the epithermal neutron beam. Referring to FIG. 1, FIG. 2, FIG. 6, FIG. 8, and FIG. 10 to FIG. 12, in the first embodiment, the third embodiment, the forth embodiment and the sixth embodiment to the eighth embodiment, the moderator 14 part is composed of multi-step moderator. In the fifth embodiment, as shown in FIG. 9, the moderator 14 is composed of an entire cylindrical moderator. In another embodiments, the moderator 14 may be composed of one conical moderator and one cylindrical moderator, or two conical moderator in second embodiment shown in FIG. 4. In second embodiment, a moderator 14′ is composed of two opposite conical moderators, and in the present disclosure, the moderator 14′ in second embodiment 2 is referred to as a double-conical moderator. Referring to FIG. 4, the moderator 14′ has a first end 141′, a second end 142′, and a third end 143′ located between the first end 141′ and the second end 142′, The cross sections of the first end 141′, the second 142′, and the third end 143′ are circular, and diameters of the first end 141′ and the second end 142′ are less than the diameter of the third end 143′. A first conical body 146′ is formed between the first end 141′ and the third end 143′, and a second conical body 148′ is formed between the third end 143′ and the second end 142′. The target 31 is accommodated in the first conical body 146′. In the second embodiment, an angle between the neutron beam axis X1 and each of the second accommodating pipe 132, the third accommodating pipe 133, the second cooling portion 22, and the third cooling portion 23 is 0°. In another embodiments, the angle between the neutron beam axis X1 and each of the second accommodating pipe 132, the third accommodating pipe 133, the second cooling portion 22, and the third cooling portion 23 may be alternatively any other angle greater than 0° and less than or equal to 180°. For example, as shown in FIG. 6, an angle between the neutron beam axis X1 and each of a second accommodating pipe 132′, a third accommodating pipe 133′, the second cooling portion 22′, and a third cooling portion 23′ is 90°. As shown in FIG. 6, it shows a neutron capture therapy system 1″ according to the third embodiment of the present disclosure. The second cooling portion 22′ and the third cooling portion 23′ of a cooling device 20′ are perpendicular to the neutron beam axis X1. That is, the cooling device 20′ is disposed to an “I”-shaped structure to cool the target 31 in the inserted vacuum tube 30. A first cooling portion 21′ in the “I”-shaped cooling device 20′ is same as the first cooling portion 21 in the “[”-shaped cooling device 20. The difference is that the second cooling portion 22′, the third cooling portion 23′ and the first cooling portion 21′ of the “I”-shaped cooling device 20′ are located in the same plane perpendicular to the neutron beam axis X1′, and the second cooling portion 22′ and the third cooling portion 23′ respectively pass through the moderator 14′ along the direction perpendicular to the neutron beam axis X1. That is, an angle between the neutron beam axis X1 and each of the second cooling portion 22′ and the third cooling portion 23′ is 90°, so that the entire cooling device is disposed into a rectangle, that is, the foregoing “I”-shaped structure. Referring to FIG. 6 again, correspondingly, the accommodating pipe 30′ is also set to an “I”-shaped structure, the first accommodating pipe 131′ of the “I”-shaped accommodating pipe 30′ is same as the first accommodating pipe 131 of a “[”-shaped cooling pipe 30. Difference is that the second accommodating pipe 132′, the third accommodating pipe 133′ and the first accommodating pipe 131′ of the “I”-shaped accommodating pipe 30′ are located in the same plane perpendicular to the neutron beam axis X1, and the second accommodating pipe 132′ and the third accommodating pipe 133′ respectively pass through the moderator 14′ along the direction perpendicular to the neutron beam axis X1. That is, an angle between the neutron beam axis X1 and each of the second accommodating pipe 132′ and the third accommodating pipe 133′ is 90°, so that the entire accommodating pipe is disposed to be a rectangle, that is, the foregoing “I”-shaped structure. It is readily conceivable that in the structures shown in FIG. 4 and FIG. 9, the cooling device 20 and the accommodating pipe 30 may also be disposed to “I”-shaped structures. FIG. 8 is a schematic diagram of the neutron capture therapy system shown 1 in FIG. 1 or the neutron capture therapy system 1′ shown in FIG. 6 in which the cooling device 20, 20′ is removed and the target 31 does not insert into the moderator 14. Compared with the neutron capture therapy system 1 disclosed in FIG. 1 or the neutron capture therapy system 1″ disclosed in FIG. 6, the neutron capture therapy system 1 disclosed in FIG. 8 only set the target 31 outside the moderator 14. That is, the accommodating cavity 12 for accommodating the vacuum tube 30 does not extend into the moderator 14 but is only surrounded by the reflector 15. The structures of the moderator 14, the reflector 15, the shielding assembly 40, the cooling devices 20, 20′, the thermal neutron absorber 16, the radiation shield 17, and the like are the same as the structures disclosed in FIG. 1 or FIG. 6. For related description, please refer to the foregoing description of related structures, details are not described herein again. FIG. 9 is a schematic diagram of the neutron capture therapy system 1 in which the cooling device 20, 20′ is removed and the first moderating unit is a stepless moderator according to the present disclosure. Compared with the neutron capture therapy system 1 disclosed in FIG. 1 or the neutron capture therapy system 1″ disclosed in FIG. 6, the neutron capture therapy system 1 disclosed in FIG. 9 only replaces the first moderating unit 140 from a three-step moderator to a stepless moderator. That is, the first moderating unit 140 is composed of a hollow cylindrical second moderating unit with an outer diameter equal to an outer diameter of the cylindrical moderator 144. The structures of the reflector 15, the shielding assembly 40, the cooling devices 20, 20′, the thermal neutron absorber 16, the radiation shield 17, and the like are the same as the structures disclosed in FIG. 1 or FIG. 6. For related description, please refer to the foregoing description of related structures, details are not described herein again. FIG. 10 is a schematic diagram of the neutron capture therapy system 1 in which the cooling device 20, 20′ is removed and the first moderating unit is a stepless moderator according to the present disclosure. Compared with the neutron capture therapy system 1 disclosed in FIG. 1 or the neutron capture therapy system 1″ disclosed in FIG. 6, the neutron capture therapy system 1 disclosed in FIG. 10 only replaces the first moderating unit 140 from a three-step moderator to a two-step moderator. The structures of the reflector 15, the shielding assembly 40, the cooling devices 20, 20′, the thermal neutron absorber 16, the radiation shield 17, and the like are the same as the structures disclosed in FIG. 1 or FIG. 6. For related description, please refer to the foregoing description of related structures, details are not described herein again. FIG. 11 is a schematic diagram of the neutron capture therapy system 1 in which the cooling device 20, 20′ is removed and the first moderating unit is a stepless moderator according to the present disclosure. Compared with the neutron capture therapy system 1 disclosed in FIG. 1 or the neutron capture therapy system 1″ disclosed in FIG. 6, the neutron capture therapy system 1 disclosed in FIG. 11 only replaces the first moderating unit 140 from a three-step moderator to a four-step moderator. The structures of the reflector 15, the shielding assembly 40, the cooling devices 20, 20′, the thermal neutron absorber 16, the radiation shield 17, and the like are the same as the structures disclosed in FIG. 1 or FIG. 6. For related description, please refer to the foregoing description of related structures, details are not described herein again. FIG. 12 is a schematic diagram of the neutron capture therapy system 1 in which the cooling device 20, 20′ is removed and the first moderating unit is a stepless moderator according to the present disclosure. Compared with the neutron capture therapy system 1 disclosed in FIG. 1 or the neutron capture therapy system 1″ disclosed in FIG. 6, the neutron capture therapy system 1 disclosed in FIG. 12 only replaces the first moderating unit 140 from a three-step moderator to a ten-step moderator. The structures of the reflector 15, the shielding assembly 40, the cooling devices 20, 20′, the thermal neutron absorber 16, the radiation shield 17, and the like are the same as the structures disclosed in FIG. 1 or FIG. 6. For related description, please refer to the foregoing description of related structures, details are not described herein again. Referring to FIG. 1, FIG. 2, FIG. 4, and FIG. 6, there is a gap between the second cooling portions 22, 22′ and the third cooling portions 23, 23′ and inner walls of the second accommodating pipes 132, 132′ and the third accommodating pipes 133, 133′, respectively. A reflection compensator 80, 80′ are filled in the gaps, respectively. The reflection compensator 80, 80′ are substances such as a lead alloy or an aluminum alloy that can absorb or reflect neutrons. The reflection compensator 80, 80′ can reflect neutrons reflected or scattered into the gap into the moderator 14 or the reflector 15, thereby increasing the yield of epithermal neutrons and reducing the time that the irradiated body needs to be irradiated. In another aspect, it avoids leakage of neutrons to the outside the beam shaping assembly 10 to cause adversely affect to the instruments of the neutron capture therapy system, and improves radiation safety. In this embodiment of the present disclosure, the content of lead in the lead alloy is greater than or equal to 85%, and the content of aluminum in the aluminum alloy is greater than or equal to 85%. Simulated experiments are performed below to statistics and analyze the epithermal neutron fluxes, fast neutron fluxes, and epithermal neutron forwardness reference values and the intensity of γ rays in the related structures of the present disclosure. In all the simulated experiments of the present disclosure, the energy of the charged particle source is 2.5 MeV and 10 mA, the count surfaces of epithermal neutron fluxes and fast neutron fluxes are located at the beam exit 18 of the beam shaping assembly 10, the diameter of the beam exit 18 is 14 CM, and the count surface of the intensity of the γ ray is the left end surface of the beam shaping assembly 10. Referring to FIG. 1 and FIG. 2, the target 31 in the first embodiment is accommodated in the moderator 14. Referring to FIG. 8, the target 31 in the fourth embodiment is disposed outside the moderator 14. To compare the impact of the arrangement positions of the target 31 in the first embodiment and the fourth embodiment on the epithermal neutron fluxes, the fast neutron fluxes, and the neutron forwardness, simulated experiments are performed to obtain the data in Table 1 for comparison and analysis. In the present disclosure, the thickness of the moderator 14 is the size of the moderator 14 in the direction of the neutron beam axis X1. TABLE 1Epithermal neutron fluxes, fast neutron fluxes, and epithermalneutron forwardness reference values when the target is accommodatedin the moderator and is disposed outside of the moderatorEpithermalEpithermalFastneutron forward-neutronneutronness referenceflux (n/flux (n/Modelvaluecm2/sec)cm2/sec)The target is outside0.6791.28E+091.38E+08of the moderator.The thickness of themoderator is 25 cm.The target is inside0.6821.26E+091.21E+08the moderator.The thickness of themoderator is 25 cm. It can be learned from Table 1 that compared with the target 31 set outside of the moderator 14, when the target 31 is accommodated in the moderator 14, the neutron forwardness does not change significantly, the intensity of fast neutrons is reduced by 12.52%, and the intensity of an epithermal neutron beam is only reduced by 1.83%. It can be learned that the arrangement manner of the target 31 is accommodated in the moderator 14 is better than the arrangement manner of the target 31 is disposed outside the moderator 14. It should be noted that the closer the epithermal neutron forwardness reference value is to 1, the better the epithermal neutron forwardness. Referring to FIG. 1 and FIG. 2, in the first embodiment, the first moderating unit 140 is a three-step moderator. Referring to FIG. 9 to FIG. 12, in the fifth to the eighth embodiments, the first moderating unit assemblies 140 is set as a stepless moderator, a two-step moderator, a three-step moderator, a four-step moderator, and a ten-step moderator respectively. To compare the impact of the first moderating unit assemblies 140 with different quantities of steps on epithermal neutron fluxes, fast neutron fluxes, and neutron forwardness, in the present disclosure, on the premise of keeping the angle θ and the depth X of the target 31 entering the moderator 14, the first moderating unit 140 is set as a stepless moderator, a two-step moderator, a three-step moderator, a four-step moderator, and a ten-step moderator respectively. The simulated experiments are performed to obtain the data in Table 2 for comparison and analysis. TABLE 2Epithermal neutron fluxes, fast neutron fluxes, and epithermalneutron forwardness reference values when the first moderatingunit is set as a stepless moderator, a one-step moderator,a two-step moderator, a three-step moderator, a four-stepmoderator, and a ten-step moderator respectivelyEpithermalEpithermalFastneutron forward-neutronneutronness referenceflux (n/flux (n/Modelvaluecm2/sec)cm2/sec)Stepless moderator0.6821.26E+091.21E+08Two-step moderator0.6821.27E+091.22E+08Three-step moderator0.6821.28E+091.22E+08Four-step moderator0.6811.28E+091.23E+08Ten-step moderator0.6821.28E+091.24E+08 It can be learned from the data in Table 2 that setting the first moderating unit 140 as a stepless (cylindrical moderator) or multi-step moderator has a little affect to the intensity of epithermal neutrons, the intensity of fast neutrons, and neutron forwardness. However, a smaller amount of moderator material is needed to manufacture a multi-step moderator than a stepless moderator. In consideration of both material costs and manufacturing process costs, preferably, the first moderating unit assembly 140 is set as a three-step or a four-step moderator. Referring to FIG. 1 to FIG. 4, FIG. 8, and FIG. 10 to FIG. 12, a gap exists between the accommodating cavity 12 and the vacuum tube 30, and the reflection compensator 50 is filled in the gap. To compare the impact of filled or unfilled the reflection compensator 50 in the gap on the intensity of epithermal neutrons, the intensity of fast neutrons, and epithermal neutron forwardness, Table 3 is provided for detailed comparison and analysis. TABLE 3Epithermal neutron fluxes, fast neutron fluxes, and epithermal neutron forwardnessreference values when the reflection compensator is filled and unfilledEpithermalneutronEpithermalFast neutronforwardnessneutron fluxIncreasefluxIncreaseModelreference value(n/cm2/sec)ratio(n/cm2/sec)ratioTwo-stepWithout reflection0.6821.27E+091.22E+08compensationWith reflection0.6831.36E+097.37%1.26E+083.72%compensationThree-stepWithout reflection0.6821.28E+091.22E+08compensationWith reflection0.6831.37E+097.33%1.27E+083.68%compensationFour-stepWithout reflection0.6811.28E+091.23E+08compensationWith reflection0.6831.37E+09 735%1.27E+083.56%compensationTen-stepWithout reflection0.6821.28E+091.24E+08compensationWith reflection0.6831.38E+097.46%1.28E+083.40%compensation It can be learned from Table 3 that compared with the reflection compensator 50 is not filled in the gap between the accommodating cavity 12 and the vacuum tube 30, when the reflection compensator 50 is filled in the gap between the accommodating cavity 12 and the vacuum tube 30, the intensity of an epithermal neutron beam is increased by 7.33% to 7.46%, the neutron forwardness is significantly change. The present disclosure only lists the data obtained through simulated experiments of the moderator 140 is set as a multi-step moderator. However, research indicates that when the moderator 14 is set as the entire cylindrical moderator shown in FIG. 9 or is set as the double-conical moderator shown in FIG. 4 or is set as a moderator that includes one conical moderator and one cylindrical moderator or is set as a moderator that includes a multi-step moderator and one conical moderator, the intensity of epithermal neutrons can be increased in different degrees by filling the reflection compensator 50 in the gap between the accommodating cavity 12 and the vacuum tube 30, and the neutron forwardness is not significantly affected. Referring to FIG. 1, FIG. 2, and FIG. 8 to FIG. 12, the shielding assembly 40 is arranged at a left end of the beam shaping assembly 10 in the present disclosure, that is, a charged particle beam entrance end, to prevent the neutron beam and the γ ray formed at the target 31 from overflowing from the left end surface of the beam shaping assembly 10. The following list the data of the intensity of neutrons and the intensity of the γ ray at the left end of the beam shaping assembly 10 and the intensity of epithermal neutrons, the intensity of fast neutrons, and epithermal neutron forwardness reference values at the beam exit 18 of the beam shaping assembly 10 when the first moderating unit assemblies 140 is set as a stepless moderator, a two-step moderator, a three-step moderator, a four-step moderator, and a ten-step moderator respectively and when the shielding assembly 40 and/or the reflection compensator 50 is disposed or the shielding assembly 40 and/or the reflection compensator 50 is not disposed, to analyze the impact of the shielding assembly 40 and the reflection compensator 50 on the intensity of neutrons and the intensity of the γ ray at the left end of the beam shaping assembly 10 and the intensity of epithermal neutrons, the intensity of fast neutrons, and epithermal neutron forwardness at the beam exit 18 of the beam shaping assembly 10. The same unit “n/cm2/sec” is used for the intensity of neutrons, the γ ray, epithermal neutrons, and fast neutrons. TABLE 4The intensity of neutrons and the intensity of the γ ray at the left end of the beam shaping assembly and the intensity of epithermalneutrons, the intensity of fast neutrons, and epithermal neutron forwardness reference values at the beam exit of the beam shaping assemblyBeam exitModelEpithermalEpithermalShieldingReflectionLeft end of a beam shaping assemblyneutron forwardnessEpithermalFastneutron/FastassemblycompensatorNeutronRatioγ rayRatioreference valueneutronneutronneutronSteplessNoNo2.45E+08100.00%6.63E+06100.00%0.6841.26E+091.21E+0810.46NoYes2.22E+0890.68%5.26E+0679.27%0.6841.36E+091.25E+0810.81YesNo9.67E+0739.43%3.61E+0654.38%0.6841.31E+091.22E+0810.76YesYes9.31E+0737.95%3.45E+0652.01%0.6841.38E+091.26E+0811.00Two-stepNoNo2.46E+08100.00%6.43E+06100.00%0.6841.28E+091.22E+0810.48NoYes2.23E+0890.71%5.11E+0679.44%0.6831.37E+091.26E+0810.84YesNo9.76E+0739.69%3.54E+0655.08%0.6831.33E+091.23E+0810.78YesYes9.40E+0738.19%3.40E+0652.81%0.6831.40E+091.27E+0811.02Three-stepNoNo2.46E+08100.00%6.36E+06100.00%0.6831.28E+091.23E+0810.46NoYes2.23E+0890.70%5.06E+0679.55%0.6831.37E+091.27E+0810.82YesNo9.81E+0739.83%3.51E+0655.21%0.6821.33E+091.24E+0810.76YesYes9.44E+0738.31%3.37E+0653.00%0.6821.40E+091.28E+0811.00Four-stepNoNo2.47E+08100.00%6.31E+06100.00%0.6821.29E+091.23E+0810.47NoYes2.24E+0890.73%5.02E+0679.65%0.6821.38E+091.27E+0810.84YesNo9.84E+0739.90%3.50E+0655.46%0.6821.33E+091.24E+0810.77YesYes9.46E+0738.38%3.36E+0653.26%0.6831.41E+091.2SE+0811.02Ten-stepNoNo2.47E+08100.00%6.23E+06100.00%0.6831.29E+091.24E+0810.40NoYes2.24E+0890.74%4.96E+0679.68%0.6831.38E+091.2SE+0810.79YesNo9.89E+0740.05%3.47E+0655.80%0.6821.34E+091.25E+0810.71YesYes9.51E+0738.52%3.34E+0653.62%0.6821.41E+091.29E+0810.98 It can be learned from Table 4 that adding the shielding assembly 40 can significantly reduce the intensity of the γ ray and the intensity of the neutron beam behind the beam shaping assembly 10, the shielding assembly 40 does not significantly affect the intensity of epithermal neutrons and the intensity of fast neutrons at the beam exit 18, and adding the reflection compensator 50 can significantly increase the intensity of epithermal neutrons at the beam exit 18. The present disclosure only lists the data obtained through simulated experiments of the moderator 140 is set as a stepless moderator or multi-step moderator. However, research indicates that when the moderator 14 is set as the double-conical moderator shown in FIG. 4 or is set as a moderator that includes one conical moderator and one cylindrical moderator or is set as a moderator that includes a multi-step moderator and a conical moderator, the intensity of epithermal neutrons can be increased in different degrees, the intensity of the γ ray and the intensity of the neutron beam behind the beam shaping assembly 10 can be reduced in different degrees, and the neutron forwardness is not significantly affected by filling the reflection compensator 50 in the gap between the accommodating cavity 12 and the vacuum tube 30 and by disposing the shielding assembly 40 at the left end of the beam shaping assembly 10. In the following, the effect of changing the depth X of the target 31 entering into the moderator 14, that is, changing the size of the first moderating unit 140 in the direction of the neutron beam axis X1, on epithermal neutron fluxes, fast neutron fluxes, and neutron forwardness is analyzed through simulated experiment date under the premised of keeping the angle θ unchanged. TABLE 5Epithermal neutron fluxes, fast neutron fluxes, and neutron forwardness reference values when thedepth X of the target entering into the moderator are respectively 5 CM, 10 CM, 15 CM, and 20 CMEpithermalEpithermalFast neutronEpithermalneutron forwardnessneutron fluxfluxneutron/FastModelreference value(n/cm2/sec)Ratio(n/cm2/sec)RationeutronSteplessX = 5 cm0.6811.27E+09100.00%1.25E+08100.00%10.16X = 10 cm0.6821.26E+0999.21%1.21E+0896.80%10.43X = 15 cm0.6831.25E+0998.43%1.19E+0895.20%10.46X = 20 cm0.6821.24E+0997.64%1.19E+0895.20%10.45Two-stepX = 5 cm0.6811.28E+09100.00%1.26E+08100.00%10.10X = 10 cm0.6821.27E+0999.22%1.22E+0896.83%10.44X = 15 cm0.6821.26E+0998.44%1.20E+0895.24%10.54X = 20 cm0.6821.25E+0997.66%1.19E+0894.44%10.54Three-stepX = 5 cm0.6811.28E+09100.00%1.27E+08100.00%10.04X = 10 cm0.6821.28E+09100.00%1.22E+0896.06%10.42X = 15 cm0.6821.27E+0999.22%1.20E+0894.49%10.55X = 20 cm0.6831.26E+0998.44%1.19E+0893.70%10.56X = 5 cm0.6811.28E+09100.00%1.27E+08100.00%10.05Four-stepX = 10 cm0.6811.28E+09100.00%1.23E+0896.85%10.43X = 15 cm0.6821.27E+0999.22%1.20E+0894.49%10.59X = 20 cm0.6821.26E+0998.44%1.19E+0893.70%10.56Ten-stepX = 5 cm0.6811.28E+09100.00%1.28E+08100.00%10.00X = 10 cm0.6821.28E+09100.00%1.24E+0896.88%10.36X = 15 cm0.6821.28E+09100.00%1.21E+0894.53%10.55X = 20 cm0.6821.27E+0999.22%1.20E+0893.75%10.57 It can be learned from Table 5 that as the depth of the target 31 extending into the moderator 14 increases, the intensity of the epithermal neutron beam slightly decreases (about 2%), the intensity of fast neutrons decreases by about 6%, the epithermal neutron forwardness shows no significant change, and the ratio of the epithermal neutron flux to the fast neutron flux is increased. The present disclosure only lists the data obtained through simulated experiments of the moderator 140 is set as a stepless moderator or a multi-step moderator. However, research indicates that when the moderator 14 is set as the double-conical moderator shown in FIG. 4 or is set as a moderator that includes one conical moderator and one cylindrical moderator or is set as a moderator that includes a multi-step moderator and a conical moderator, as the depth of the target 3 extending into the moderator 14 increases, the intensity of the epithermal neutron beam slightly decreases, the intensity of fast neutrons decreases, the epithermal neutron forwardness shows no significant change, and the ratio of the epithermal neutron flux to the fast neutron flux is increased. To compare the impact on yield of epithermal neutrons, a contamination amount of fast neutrons, and an irradiation time when the reflection compensator 80 are respectively a lead alloy and an aluminum alloy and there is no reflection compensator 80 (that is, air is filled) in the gaps between the cooling devices 20, 20′ and the accommodating pipes 13, 13′, Table 6 to Table 8 are listed for detailed comparison. Table 6 shows the yield of epithermal neutrons (n/cm2 mA) when filling air, aluminum alloy, and lead alloy respectively under different accommodating cavity hole diameters: TABLE 6Yield of epithermal neutrons (n/cm2mA)Accommodating cavity hole diameter (CM)16 CM18 CM20 CM22 CM24 CM26 CMAir8.20E+077.82E+077.38E+076.97E+076.56E+076.22E+07Aluminum alloy8.74E+078.58E+078.40E+078.23E+078.07E+077.88E+07Lead alloy8.94E+078.88E+078.79E+078.69E+078.63E+078.53E+07 Table 7 shows contamination amounts of fast neutrons (Gy-cm2/n) when filling air, aluminum alloy, and lead alloy respectively under different accommodating cavity hole diameters: TABLE 7Contamination amount of fast neutrons (Gy-cm2/n)Accommodating cavity hole diameter (CM)16 CM18 CM20 CM22 CM24 CM26 CMAir7.01E−137.51E−138.23E−138.95E−139.80E−131.06E−12Aluminum alloy6.54E−136.83E−137.17E−137.54E−137.90E−138.37E−13Lead alloy6.56E−136.83E−137.18E−137.52E−137.87E−138.29E−13 Table 8 shows an irradiation time (minute) that an irradiated body requires when filling air, aluminum alloy, and lead alloy respectively under different accommodating cavity hole diameters: TABLE 8Irradiation time (Min) that an irradiated body requiresAccommodating cavity hole diameter (CM)16 CM18 CM20 CM22 CM24 CM26 CMAir30.8631.1632.2932.6633.4234.12Aluminum alloy29.6529.0730.4629.4229.2229.39Lead alloy28.9428.0028.3727.7627.9128.04 It can be learned from Table 6 to Table 8 that when the accommodating cavity hole diameter is the same, compared with air filling, the yield of epithermal neutrons is higher when filled with lead alloy or aluminum alloy, and the contamination amount of fast neutrons and the required irradiation time is less. The neutron capture therapy system disclosed in the present disclosure is not limited to the content in the foregoing embodiments and the structures represented in the accompanying drawings. For example, the moderator may be disposed to be a cone or a polygonal prism, several cooling devices may be disposed, and several accommodating pipes are correspondingly provided. All obvious changes, replacements or modifications made to the materials, shapes, and positions of the members based on the present disclosure fall within the protection scope of the present disclosure. Although the illustrative embodiments of the present invention have been described above in order to enable those skilled in the art to understand the present invention, it should be understood that the present invention is not to be limited the scope of the embodiments. For those skilled in the art, as long as various changes are within the spirit and scope as defined in the present invention and the appended claims, these changes are obvious and within the scope of protection claimed by the present invention.
052290662	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to a method and apparatus for determining the degree of insertion or the axial position of any and all control rods in a nuclear reactor core by analyzing the output signals of a number of strings of fixed location, neutron or gamma ray sensitive incore detectors. 2. Description of the Related Art Knowing the position of control rods in a nuclear core of a nuclear reactor is essential to safe operation of a nuclear power plant as well as a legal requirement for continuation of an operating license. Currently control rod position, where control rods include safety rods, grey rods, shim rods and displacer rods, as well as specifically "control" rods, can be determined in two different ways. The first uses a coil stack that sits on top of the reactor vessel in which the control rod drive shaft moves up and down. A magnetic impedance produced voltage generated by the coil stack is proportional to the length of the rod drive shaft residing in the coil stack, thereby allowing rod position to be inferred through the voltage output of each coil stack. Occasionally the rod position indicated by this system can be in question. Conventionally, the rod position at startup, as indicated by the coil stack indicators, is checked against the step demand counters to verify that the indicators are valid. If a rod position cannot be verified, because, for example, a coil stack is inoperable due to a mechanical or electrical failure, flux mapping can be performed to indicate rod position. If the position of the target rod cannot be verified, the rod position indicator must be considered inoperable. A reactor shutdown is usually necessary if more than one rod is considered inoperable. The second method of determining rod position determines enthalpy rise deviations in core power distribution using core exit thermocouples and an inlet temperature detector to determine control rod position change relative to a rod reference position as described in U.S. Pat. No. 4,927,594. Because enthalpy deviation changes with respect to rod position, the change in rod position can be determined from the magnitude of the deviation. By adding the change to the reference rod position, the actual rod position can thereby be determined. The accuracy of the rod position determined by a system in accordance with this second method, needs to be enhanced and supplemented to provide the operator with the most reliable indication of rod position possible, so that unnecessary reactor shutdowns are avoided. SUMMARY OF THE INVENTION It is an object of the present invention to provide an enhanced method of determining control rod axial position in a nuclear reactor core. It is another object of the present invention to supplement conventional rod position indication systems to provide plant operators with a diverse confirmation of rod position. It is also an object of the present invention to provide power distribution signatures that allow rod position and distribution abnormalities to be detected. It is also an object of the present invention to determine rod position by performing signature analysis on changes (deviations) in fixed incore detector responses. It is a further object of the present invention to determine rod position during reactor changes when not at full power or in a steady state. It is another object of the present invention to provide a protection grade rod position indication system, as defined by IEEE or ANSI standards. The above objects can be attained by a system which determines rod position by first producing a fixed incore detector response signature database appropriate to current core conditions. When a core thermal neutron or gamma ray flux distribution perturbation is detected the approximate rod configuration is determined by scanning the database for a match. If a match is found the rod position is known. If a match is not found the closest rod configuration is used as a reference position from which a search is conducted to determine rod position. These together with other objects and advantages which will be subsequently apparent, reside in the details of construction and operation as more fully hereinafter described and claimed, reference being had to the accompanying drawings forming a part hereof, wherein like numerals refer to like parts throughout.
	summary
	claims	1. A system, comprising:a spent fuel container coupled to at least one inlet path and at least one outlet path, the at least one inlet path configured to transport a non-reactive gas into the spent fuel container, the at least one outlet path configured to transport potentially radioactive vapor and gas from the spent fuel container, the at least one inlet path and at least one outlet path further coupled to a circulation path;at least one circulation pump coupled to the circulation path, wherein the at least one circulation pump comprises a vacuum circulating pump when the system is operating in a vacuum drying mode, and a gas circulating pump when the system is operating in a cooling mode, the vacuum circulating pump configured to remove vapor and other gases from the container and transport the vapor and other gases through the circulation path, and the gas circulating pump configured to circulate the non-reactive cooling gas through the circulation path;a heat exchanger coupled to the circulation path;a non-reactive gas source coupled to the circulation path; anda waste vent coupled to the circulation path; whereinthe spent fuel container is configured to receive the non-reactive gas via the at least one inlet path;the waste vent is configured to receive vapor and gases from the spent fuel container via the circulation path; andthe heat exchanger is configured to cool a gas exiting the spent fuel container from the at least one outlet path. 2. The system of claim 1, wherein the circulation pump is configured to create a negative pressure in the spent fuel container relative to the atmospheric pressure when the system is operating in vacuum drying mode. 3. The system of claim 1, wherein the circulation pump is configured to lower a pressure within the spent fuel container when the system is operating in vacuum drying mode. 4. The system of claim 1, wherein the heat exchanger further comprises:a cooling medium inlet configured to receive a cooling medium entering the heat exchanger;a cooling medium outlet configured to transport the cooling medium exiting the heat exchanger; anda radioactive waste liquid disposal system configured to transport a radioactive waste liquid from the heat exchanger. 5. The system of claim 1, further comprising at least one heat exchanger shutoff valve system configured to isolate the heat exchanger from the circulation path. 6. The system of claim 1, further comprising a non-reactive gas shutoff valve system configured to isolate the non-reactive gas source from the circulation path. 7. The system of claim 1, further comprising a circulation path shutoff valve system configured to cease circulation of gasses through the circulation path. 8. The system of claim 1, wherein the spent fuel container further comprises a spent fuel canister disposed within a cask, and the at least one inlet path and at least one outlet path are communicatively coupled to the spent fuel canister. 9. The system of claim 1, further comprising a spent fuel container bypass valve system configured to isolate the spent fuel container from the circulation path. 10. The system of claim 1, further comprising a radioactive waste gas disposal system coupled to the circulation path, the radioactive waste gas disposal system configured to remove radioactive waste gas from the circulation path. 11. The system of claim 1, further comprising a circulation path vent configured to vent excess system pressure to another location. 12. The system of claim 1, wherein a non-reactive gas is transported into the spent fuel container via the at least one inlet path and comprises at least one of helium, argon, carbon dioxide, and nitrogen. 13. The system of claim 12, wherein the non-reactive gas further comprises another inert gas.
	description	The United States Government has rights in this invention pursuant to Contract No. DE-AC03-76SF00098 between the United States Department of Energy and the University of California. This application claims benefit of priority to U.S. application Ser. No. 10/100,962 filed Mar. 18, 2002 entitled “CYLINDRICAL NEUTRON GENERATOR” (now U.S. Pat. No. 6,907,097), hereby incorporated by reference, which in turn claims priority to Provisional Applications Ser. Nos. 60/276,669 filed Mar. 16, 2001 and 60/316,792 filed Aug. 31, 2001, also incorporated by reference. The invention relates to neutron tubes or sources, and more particularly to neutron tubes or sources based on plasma ion generators, including compact neutron tubes or sources which generate a relatively high neutron flux using the D-D reaction. Conventional neutron tubes employ a Penning ion source and a single gap extractor. The target is a deuterium or tritium chemical embedded in a molybdenum or tungsten substrate. Neutron yield is limited by the ion source performance and beam size. The production of neutrons is limited by the beam current and power deposition on the target. In the conventional neutron tube, the extraction aperture and the target are limited to small areas, and so is the neutron output flux. Commercial neutron tubes have used the impact of deuterium on tritium (D-T) for neutron production. The deuterium-on-deuterium (D-D) reaction, with a cross section for production a hundred times lower, has not been able to provide the necessary neutron flux. It would be highly desirable and advantageous to make high flux D-D neutron sources feasible. This will greatly increase the lifetime of the neutron generator, which is unsatisfactory at present. For field applications, it would greatly reduce transport and operational safety concerns. For applications such as mine detection, where thermal neutrons are presently used, the use of the lower energy D-D neutrons (2.45 MeV rather than 14.1 MeV) also would decrease the size of the neutron moderator. The present invention has three potential competitors for field or small-laboratory use: (1) isotopic sources based on a sample of a radioactive substance, e.g. californium-252, that emits neutrons; (2) accelerator sources, usually based on an ion source feeding a radiofrequency quadrupole (RFQ) linac and thence a neutron production target; and (3) conventional neutron tubes. Of these, the most direct and significant competitors are commercially available neutron tubes. As for the others, RFQ-based sources have never become a major commercial presence due to cost and complexity, and the safety concerns and lack of time structure that are inherent to isotopic sources limit their applications. Neutronics can identify possible explosives and nuclear materials by their composition, not just by their shape or density the way x-ray machines do. Since the September 11 terrorist attacks, detection of explosives and fissionable materials has become an urgent national need. Detecting such materials hidden in baggage or cargo is challenging under real-world conditions. Thermal neutron analysis (TNA) has been tried for inspection of checked baggage and cargo at airports. Low-energy neutrons cause nitrogen in explosives to emit gamma rays and cause fissile materials to give off neutrons of their own. The first-generation TNA screeners were too large, complex, and expensive; FAA-approved screening devices presently on the market use x-rays to look at shapes and densities, rather than using neutronics to detect actual composition. Besides the obvious considerations of cost-effectiveness and acceptable footprint, systems for inspecting baggage and cargo must offer trustworthiness (reliability combined with freedom from both false positives and false negatives), plus high throughput so that spot checks can be replaced by comprehensive inspection without bottlenecking an already heavily burdened process. Systems are also needed for relatively nonintrusive inspection of larger objects, e.g. an intermodal cargo container, or a vehicle. Detection of land mines or unexploded ordnance is another related application of great worldwide importance. A compact neutron generator design with a high neutron flux and adapted for these uses would be highly advantageous. Neutron logging instruments consist of a neutron generator and gamma-ray detector packaged so as to fit into a small (e.g. 2-inch-diameter) borehole. Analyzing the gamma ray spectrum due to neutron capture and inelastic scattering in the subsurface allows elements in the medium to be identified. Applications include oil and mineral exploration, and basic geological studies. A neutron generator design with a high neutron flux and adapted for use in a borehole would be highly advantageous. The invention is a cylindrical neutron generator formed with a coaxial RF-driven plasma ion source. A deuterium plasma (or a deuterium and tritium plasma) is produced by RF excitation in a plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator and is separated therefrom by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360° and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. There are two alternate basic embodiments of the neutron generator, in which the position of the plasma generator and neutron target are reversed. In one embodiment the plasma generator is in the center and the neutron target is on the outside, and in the second embodiment, the plasma generator is on the outside and the target is on the inside. The plasma generator may be either cylindrical or annular shaped, and the target is a cylinder. The neutron target surrounds the cylindrical plasma ion generator or is positioned inside the annular shaped plasma ion generator. In both cases the plasma generator and target are coaxial or concentric. The embodiment with the target on the outside is preferred since the target area is larger. A more complex embodiment of the neutron generator, which combines the two basic embodiments, has a nested configuration that is formed by nesting concentric targets and plasma regions. The nested configuration places a coaxial target both inside and outside the plasma generating region, and nests several targets and plasma generating regions to increase the neutron flux. This invention enables the generator to operate with high current density, high atomic species and practically unlimited beam size in the axial or longitudinal direction. The structure is compact and rugged, e.g. the RF antenna can form part of the plasma electrode and chamber wall. Thus the source's lifetime should be greatly increased because no weak components exist. The geometry is ideal for borehole applications. The source is ideal for many neutronic applications. Because of the increased target area, the much safer D-D reaction can be used, eliminating any tritium from the source. FIGS. 1A, 2A, 3A show the neutron source geometry of a first embodiment 10 of the invention, which has a cylindrical neutron generating target outside a cylindrical plasma ion source. Neutron generator 10 has a cylindrical plasma ion source 12 at its center. The principles of plasma ion sources are well known in the art. Conventional multicusp ion sources are illustrated by U.S. Pat. Nos. 4,793,961; 4,447,732; 5,198,677; 6,094,012, which are herein incorporated by reference. Ion source 12 includes an RF antenna (induction coil) 14 for producing an ion plasma 20 from a gas which is introduced into ion source 12. Antenna 14 is typically made of titanium tubing, which may be water cooled. For neutron generation the plasma is preferably a deuterium ion plasma but may also be a deuterium and tritium plasma. A deuterium plasma with current density of about 100 mA/cm2 can be produced. Ion source 12 also includes a pair of spaced electrodes, plasma electrode 16 and extraction electrode 18, along its outer circumference. Electrodes 16, 18 electrostatically control the passage of ions from plasma 20 out of ion source 12. Electrodes 16, 18 contain many longitudinal slots 19 along their circumferences so that ions radiate out in a full 360° radial pattern. Coaxially or concentrically surrounding ion source 12 and spaced therefrom is cylindrical target 22. Target 22 is the neutron generating element. Ions from plasma source 12 pass through slots 19 in electrodes 16, 18 and impinge on target 22, typically with energy of 120 keV to 150 keV, producing neutrons as the result of ion induced reactions. The target 22 is loaded with D (or D/T) atoms by the beam. Titanium is not required, but is preferred for target 22 since it improves the absorption of these atoms. Target 22 may be a titanium shell or a titanium coating on another chamber wall 24, e.g. a quartz tube. Flange 26 extends from the ends of chamber wall 24. The extraction apertures in electrodes 16, 18 are in the form of slots 19 whose length can be extended to any desired value. The beam impinges on the target 22 in 360° and therefore the target area can be enhanced by at least 2 orders of magnitude over conventional neutron sources. Thus the same or greater neutron flux can be generated from D-D reactions in this neutron generator as can be generated by D-T reactions in a conventional neutron tube, eliminating the need for radioactive tritium. The neutrons produced, 2.45 MeV for D-D or 14.1 MeV for D-T, will also go out radially in 360°. By making the neutron generator as long as practical in the axial or longitudinal direction, a high neutron current can be obtained. FIG. 2A shows further details of neutron generator embodiment 10 from FIG. 1A. Induction coil (RF antenna) 14 is positioned inside concentric cylindrical electrodes 16, 18. Ions passing through the slots 19 in electrodes 16, 18 strike target (surface) 22. FIG. 3A shows some further details and minor variations of the design. The entire generator 10 is contained within a vacuum chamber 27 which is spaced apart from target chamber wall 24. Only a single extraction grid 18 is shown; plasma grid 16 is not needed since the ions can be extracted with a single grid. Chamber wall 24, on which target coating 22 is formed, is surrounded by target cooling coils 28. Permanent magnets 30 are arranged in a spaced apart relationship, running longitudinally along plasma ion generator 12, to from a magnetic cusp plasma ion source. The principles of magnetic cusp plasma ion sources are well known in the art, as cited above. FIGS. 1B, 2B, 3B, 3C show the neutron source geometry of a second embodiment 40 of the invention, which is similar to neutron generator 10, except that the cylindrical ion source and neutron generating target are in a reversed position, i.e. the cylindrical neutron generating target is inside the cylindrical plasma ion source. Neutron generator 40 has a cylindrical plasma ion source 42 at its outside. Ion source 42 includes an RF antenna (induction coil) 44 for producing an ion plasma 50. Ion source 42 also includes a pair of spaced electrodes, plasma electrode 46 and extraction electrode 48, along its inner circumference. Electrodes 46, 48 electrostatically control the passage of ions from plasma 50 out of ion source 42 into the interior of neutron generator 40. Electrodes 46, 48 contain many longitudinal slots 49 along their circumferences so that ions radiate in a full 360° radial pattern. Extraction electrode 48 is inside plasma electrode 46, the reverse of neutron generator 10, since the direction of plasma flow from the plasma ion source 42 is radially inward rather than radially outward, as in neutron generator 10. Ion source 42 coaxially or concentrically surrounds and is spaced from an inner cylindrical target 52. Target 52 is the neutron generating element. Ions from plasma source 42 pass through slots 49 in electrodes 46, 48 and impinge on target 52, typically with energy of 120 keV to 150 keV, producing neutrons as the result of ion induced reactions. The target 52 is loaded with D (or D/T) atoms by the beam. Titanium is not required, but is preferred for target 52 since it improves the absorption of these atoms. Neutron generator 40 is enclosed in an external chamber 54. The extraction apertures in electrodes 46, 48 are in the form of slots 49 whose length can be extended to any desired value. The beam impinges on the target 52 in 360° and therefore the target area can be enhanced. However, between neutron sources 10 and 40, for the same outside source diameter, the target in source 10 will be larger because of its larger diameter. The neutrons produced, 2.45 MeV for D-D or 14.1 MeV for D-T, will also go out radially in 360°. By making the neutron generator as long as practical in the axial or longitudinal direction, a high neutron current can be obtained. FIG. 2B shows the internal structure of neutron source 40 without chamber 54, and FIG. 3B shows the structure of FIG. 2B inside chamber 54, with flanges 56 extending from the ends of chamber 54. FIG. 3C shows a minor design change in which the RF antenna 44 is incorporated into the plasma electrode 46 and the chamber wall 54. This arrangement makes the source more compact and efficient. FIGS. 4A-B show the neutron source geometry of a third embodiment 60 of the invention, which has a nested configuration that is formed by nesting concentric neutron generating targets and plasma generating regions. The nested configuration of source 60 is a combination of sources 10, 40, placing a coaxial target both inside and outside a plasma generating region, and nesting several targets and plasma generating regions to increase the neutron flux. Except for the additional number of each component, each one is structured and functions essentially the same as in the basic embodiments. Inside a cylindrical chamber 62, a plurality of concentric or coaxial alternating targets 64 and plasma generating regions 66 are arranged. Each target 64 is a cylinder. Each plasma generating region 66 is annular and has an RF antenna (induction coil) 68 positioned therein. While four targets 64 alternating with three plasma generating regions 66 are shown, at least one plasma generating region 66 with two targets 64 are needed and any number of additional nested layers may be added depending on the desired neutron yield. Each plasma generating region 66 is bounded by extraction electrodes 70 on both its inner and outer surfaces. Electrodes 70 contain longitudinal slots, as previously shown, through which ions are extracted from plasma generating regions 66 and directed onto targets 64. A deuterium plasma with current density of about 100 mA/cm2 can be produced. Chamber 60 has a water inlet 72 and water outlet 74 for circulating water or other coolant to remove heat from the targets 64, as described further herein. The targets are typically made of copper with a thin coating of titanium on the surface. The power density generated by the beam is about 600 W/cm2 which can be removed by water cooling. As shown in FIGS. 4C-D, plasma generating regions 66 and targets 64 are nested coaxially. Extraction electrodes 70 with slots 76 bound the regions 66. The width of regions 66 is d1 and the gap from electrode 70 to target 64 is d2; d1 and d2 are typically 2.5 cm. Water inlet 72 and outlet 74 in chamber 62 connect to coolant channels in the chamber wall and targets 64 so that the targets can be cooled by flowing coolant during operation. A rod 78 may extend into the center of the source inside the inner target. Samples may be placed there for irradiation. A high voltage source 80 is connected between the extraction electrodes 70 and targets 64 to extract the ions from regions 66 where they are formed and accelerate them onto the targets 64 where they are collected and react. With a gap of about 2.5 cm, and adequate pumping in the region outside the ion sources, an extraction voltage of 80 kV or higher may be used. As shown in FIGS. 4E-F, RF antennas 68 are disposed within plasma generating regions 66. As shown in FIG. 4E, a coil of antenna 68 loops around the annular region 66 in a radial plane. As shown in FIG. 4F, a plurality, e.g. three, coils of antenna 68 are spaced apart axially in different radial planes along the length of the source. The separate coils of antenna 68 are all connected together in parallel by linear elements 80 which extend axially along the source. Thus the antenna generally comprises three or more antenna loops with a bi-filar arrangement, normally connected in parallel. The antenna is typically made of titanium tubing, which may be water cooled. FIG. 5 is a graph of the neutron production cross sections of the D-D and D-T reactions. Although the D-D cross section is much lower than the D-T cross section, the large target surface area provided by the cylindrical geometry, makes a source based on the D-D reaction practical since high neutron flux can be obtained. Thus the hazards of dealing with tritium in the source can be eliminated. FIG. 6 shows an application of the compact high flux neutron source for borehole instrumentation, e.g. oil well logging. A neutron generator 90 according to the invention, in combination with a gamma ray detector 92, is lowered into a borehole 94, with necessary electrical connections made through cable 96. Neutrons emitted by generator 90 interact with features in the ground, e.g. ground contamination region 98, and produce gamma rays that are detected by detector 92. The signals are analyzed by techniques known in the art to identify subterranean features or the presence of resources. Because of the cylindrical geometry of generator 90, it can be made of a diameter suitable for a typical borehole, while its length can be selected to give a sufficiently high neutron flux to improve detection capability. The combination of simplicity, compactness, and high flux offered by the compact neutron source of the invention is also advantageous for security applications such as thermal neutron analysis (TNA) to inspect baggage at airports. Compared to other technologies for performing TNA, it is much cheaper and simpler than RFQ-based systems, substantially smaller than previous neutron tubes capable of the same flux (which is important because high neutron flux is allows high throughput), and if D-D reactions are used, not only intrinsically safer than radioactive sources, but also, in contrast to them, capable of being gated rapidly on and off to allow finer discrimination by the detector and accompanying software. FIG. 7 illustrates a system 100 for detecting explosives 102 or fissile nuclear material 104 in a suitcase 106. System 100 includes a compact neutron source 108 of the type as generally described above but which may be slightly modified to also provide neutrons in an axial direction by extending the target to cover an end of the source and extracting ions axially as well as radially from the plasma generating region. Source 108 is connected by a power cable 110 to an external power supply 112. Source 108 is surrounded by a moderator 114 which slows the neutrons generated by source 108 to thermal energy to perform TNA. The moderator is further surrounded on the sides and back end by shielding 116 which prevents neutrons from escaping, so that all emitted neutrons are emitted from the front end and can be directed to the desired inspection point. Thermal neutrons may be reflected from the shield toward the open end. Also neutron source 108 could be placed transverse to the direction shown. External gamma ray detector 118 is used to detect gamma rays emitted by explosives and external neutron detector 120 is used to detect neutrons from fissile nuclear material. As described above, the neutron tube with coaxial geometry increases the available target area, thus giving up to 1000× the neutron output of existing tubes of comparable size. Such high output from so small a device opens up new horizons for explosives detection and also for R&D applications of neutronics. In a representative small version of this highly scalable design (length 26 cm, diameter 28 cm, and weight about 18 kg) with a single target, an expected output has been calculated of ˜1.2×1012 n/s for D-D neutrons and ˜3.5×1014 n/s for D-T. The neutron flux of course depends on several parameters, including voltage, current, and the size of the tube. The ultimate limiting factor in output is the power density on the target, which is conservatively limited at ˜650 W/cm2. Tubes with nested concentric targets and plasma regions can have as much as 10× higher average neutron flux than a tube of similar size (e.g. length 35 cm, diameter 48 cm) with a single target, e.g. ˜1.6×1013 n/s for D-D neutrons and ˜4.5×1015 n/s for D-T. As a broad general principle, “comparable” conventional neutron tubes (e.g., with diameters of several tens of cm and lengths of up to a few hundred cm) produce 106 to 108 neutrons per second in D-D operation. Many neutronics applications could be improved by a small high-flux neutron source according to the invention. These include: Condensed matter physics. Scattering of slow neutrons in condensed matter (solids or liquids) can determine structure on the atomic or molecular level. Neutrons penetrate deeply into matter, enabling study of new materials in realistic temperatures, pressures, and other ambient conditions. Material science. Point defects, dislocations, interphase boundaries, intrinsic junctions with microcracks, pores, etc. can be studied. Studies of molecular compounds. Small-angle neutron scattering (SANS) is a powerful method to investigate polymer systems and surface-active substances. Specular reflection provides information about the structure along the surface. Biology. Neutrons can “see” hydrogen better than photons can, so details of the structure and function of some biological systems can be better studied. Medical applications. Boron neutron-capture therapy trials; brachytherapy. Engineering analysis. Neutron diffraction probes internal stresses in multiphase materials. Both R&D and nondestructive evaluation could benefit. Earth sciences. Neutrons can probe the texture of rocks and minerals and the effects of external pressure on the structure of samples. Changes and modifications in the specifically described embodiments can be carried out without departing from the scope of the invention which is intended to be limited only by the scope of the appended claims.
043022944	abstract	A nuclear reactor fuel assembly comprises a bundle of parallel rods which are transversely spaced by cross-pieces, and is supported by support tubes which extend between, and are fixed at their ends to, two grids. The grids have cells which are aligned with the fuel rods and through which the fuel rods can be withdrawn. The grids are detachably connected to end plates by sockets which are removably engaged in the end plates and cells of the grids.
	abstract	The present invention relates to a data processing system and method for seamlessly integrating laboratory information management system (LIMS) functionalities into an enterprise resource planning (ERP) system. The invention enables to keep the additional complexity that is required for the provision of LIMS functionalities at a minimum as all laboratory instrumentation setup information is centrally stored in a server computer but not in the ERP system.
046801596	abstract	A storage container is disclosed for accommodating fuel rods of disassemb irradiated nuclear reactor fuel elements. The storage container includes a circular storage space into which a cage receiving the canned fuel rods is inserted. The individual fuel-rod cans are arranged in the insert cage in the form of a circle in order to achieve a good removal of heat. Hold-down springs bear on the end faces of the cans. The radial extension of the fuel-rod cans towards the longitudinal center of the container is limited to accommodate an empty square central shaft in the middle of the insert cage.
	abstract	A radiation protection device for providing protection of a body part that includes active bone marrow from ionizing radiation may include a radiation protection component configured to be placed adjacent to and externally cover the body part so as to reduce a radiation dose absorbed in that body part.
	abstract	The present invention realizes a nanotube probe with high durability that can be manufactured in short time with less impurities adhered to the holder sustaining the nanotube. The nanotube probe according to this invention is constructed by fastening a nanotube 8 on the protruded portion 4 of a cantilever by way of at least two partial coating films 12a and 12b. One or more additional partial coating films may be formed in the intermediate area between these two partial coating films. Each partial coating film is formed by irradiating electron beam 10 on the position where the nanotube 8 is in contact with the protruded portion 4 of the cantilever. The partial coating films are separated not to overlap each other. By minimizing the size of partial coating film as well as by narrowing down the beam diameter, coating time may be further shortened. With the beam diameter narrowed down, excessive deposit of impurities can be put under control.
	summary
	summary
	description	This application is a divisional application and claims the benefit of U.S. patent application Ser. No. 11/852,237, filed Sep. 7, 2007, which claims priority to U.S. patent application Ser. No. 10/821,658, filed Apr. 8, 2004, now U.S. Pat. No. 7,277,521 which claims priority to U.S. Provisional Patent Application No. 60/461,624 filed Apr. 8, 2003, the disclosures of which are hereby incorporated by reference herein. A part of this invention was made with United States Government support from Contract No. DE-AC03-76SF00098 between the U.S. Department of Energy (DOE) and the Lawrence Berkeley National Laboratory. The United States has certain rights in this invention. The present invention relates in general to the detection of special nuclear materials (“SNM”) in suspect containers. In particular, the present invention uses high-energy gamma rays emitted from fission products or fragments to identify SNM (i.e., 235U and 239Pu) in cargo containers and other potential sites. Special nuclear material (SNM) is defined by Title I of the Atomic Energy Act of 1954 as plutonium, uranium-233, or uranium enriched in the isotopes uranium-233 or uranium-235. The definition includes any other material which the Nuclear Regulatory Commission determines to be special nuclear material, but does not include source material. The Nuclear Regulatory Commission (“NRC”) has not declared any other material as SNM. SNM is only mildly radioactive, but it includes some fissile material, uranium-233, uranium-235, and plutonium-239, that, in concentrated form, can be the primary ingredients of nuclear explosives. These materials, in quantities greater than formula quantities, are defined as “strategic special nuclear material” (SSNM). The uranium-235 content of low-enriched uranium can be concentrated (i.e., enriched) to make highly enriched uranium, the primary ingredient of a nuclear weapon. Since Sep. 11, 2001, an increased urgency has been associated with the development of new and improved means for the detection and prevention of the clandestine transport of nuclear weapons materials and other materials for producing weapons of mass destruction. A particularly difficult problem is posed by highly-enriched uranium (e.g., 235U) and plutonium (e.g., 239Pu) that might be hidden in large sea-going cargo containers, which may be filled with masses approaching 27 MT and which might represent areal densities of more than 50 g cm−2 through which an identifying signal must penetrate to reach a detector. Passive detection methods (e.g., see “Passive Nondestructive Assay of Nuclear Materials,” edited by D. Reilly, N. Ensslin, and H. Smith, Jr., NUREG/CR-5550, LA-UR-90-732 (1991)) based on measurements of neutrons and/or photons are either inapplicable or impractical in many such cases. Traditional methods of radiography are unlikely to provide a unique signature of highly-enriched 235U and 239Pu. Active interrogation with neutrons or high-energy photons in a variety of forms (e.g., see “Ionizing Radiation Imaging Technologies for Homeland Security,” D. J. Strom and J. Callerame, Proceedings of the 36th Midyear Topical Meeting, Health Physics Society, Jan. 26-29, 2003, San Antonio, Tex., and “A Review of Neutron Based Non-Intrusive Inspection Technologies,” T. Gozani, Conference on Technology for Preventing Terrorism, Hoover Institution, Mar. 12-13, 2002, Stanford University, Stanford, Calif.) currently depends upon the observation of β-delayed neutrons following induced fission to provide a unique signature for 235U and 239Pu. However, the shielding provided by a thick hydrogenous cargo can be so large that this method will fail or will have very low detection sensitivity. In addition, considering that millions of cargo and other containers enter the United States each year, and considering that SNM might be hidden in some of these containers, in order to prevent the entry of any hidden SNM into the United States, a detection method needs to be effective without having to open and unload the containers. Furthermore, not only does a detection system need to be non-invasive, it must be able to perform its detection function in as short a time as possible, so as to not overly burden the flow of goods into the U.S. via these containers. There is therefore a need for a system and a method of detecting special nuclear materials (“SNM”) in suspect containers that does not suffer from the above described shortcomings. The present invention is directed to methods and systems that use either neutrons or high-energy photons (e.g., gamma-rays) to irradiate a fully loaded cargo or other container. Such neutrons or gamma-rays have a sufficient flux and energy level to induce fission in any SNM inside the container. After the neutron or photon irradiation is completed, a detector, or an array or arrays of detectors are used to interrogate the container for high energy (e.g., above 3 MeV) gamma rays that are produced by radioactive decays of fission products. In one embodiment, the present invention is directed to a method of detecting the presence of special nuclear materials in a suspect container. The method includes irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials; detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal; comparing the detector signal with a threshold value to form a comparison; and detecting the presence of the special nuclear materials using the comparison. In another embodiment, the present invention is directed to a system for detecting the presence of special nuclear materials in a suspect container. The system includes a neutron beam source configured for irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials; a detector configured for detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal; a comparator for comparing the detector signal with a threshold value to form a comparison; and a presence detector for detecting the presence of the special nuclear materials using the comparison. For a further understanding of the nature and advantages of the invention, reference should be made to the following description taken in conjunction with the accompanying drawings. The embodiments of the present invention are directed to methods and systems that use either neutrons or high-energy photons (e.g., gamma-rays) to irradiate a fully loaded cargo or other container. Such neutrons or gamma-rays have a sufficient flux and energy level to induce fission in any SNM inside the container. After the neutron or photon irradiation is completed, a detector, or an array or arrays of detectors are used to interrogate the container for high energy (e.g., above 3 MeV) gamma rays that are produced by radioactive decays of fission products. The inventors herein have shown that the yields of high-energy gamma rays following the thermal neutron-induced fission of 235U and 239Pu are large enough to permit the detection of kilogram-sized quantities of SNM hidden inside of cargo or other containers. The inventors herein have also shown that the energy spectrum of gamma rays emitted by fission products is qualitatively different from that produced by other material that would be commonly found in cargo or other containers. In addition, the inventors herein have determined the effective half-life of these gamma rays to be approximately 20 seconds. The energy spectrum and/or the time dependence (i.e., half-life) and/or the combination of the energy spectrum and the time dependence of the gamma-ray spectrum provides a unique signature for the SNM and its detection. The embodiments of the method and system of the present invention enable a fully loaded cargo container to be screened for SNM in a period on the order of one minute or less. Furthermore, the embodiments of the method and system of the present invention are easily scalable to enable the screening of smaller sized packages such as luggage items at airports for SNM in a period on the order of one minute or less. This ability of the embodiments of the present invention to detect the presence of SNM in suspect containers is quite remarkable considering that such containers come in an enormous range of sizes and loadings. For example, such containers are closed and randomly loaded where one is unaware whether the contents are apricots, bubblegum, bombs, fabrics, metals, plastics, steel, SNM, or wood. In particular, the embodiments of the present invention use the high-energy gamma rays emitted from short-lived fission fragments to identify SNM in cargo containers and other potential sites. As used herein high-energy gamma rays refer to gamma rays having an energy level higher than approximately 3-4 million electron volts (MeV). Also as used herein, short-lived fission fragments refer to fission fragments having a half that is less than approximately one minute. The active interrogation of a mass of highly enriched uranium (“HEU”), Pu, or SNM, embedded in a cargo container, with either 2.5 MeV deuterium-deuterium (“D-D”) neutrons or 14 MeV deuterium-tritium (“D-T”) neutrons, has been studied by some. A cargo container as used herein, refer to standard containers that are commonly made of steel that are typically available in the 20-foot or 40-foot lengths and which are approximately 8-foot wide by 8.5-foot high, that are used to transport goods on cargo ships. Some containers are larger and some are smaller. In those studies, a reasonable worst-case scenario assumes that the cargo container is otherwise filled with hydrogenous material at a water-equivalent density of about 0.4 gm cm−3. As an example of the effectiveness of the embodiments of the present invention, this worse-case scenario has been considered here with the further constraint that the SNM is located at the center of the container and that a distance of 1.5 m must be penetrated before radiations can reach a detector. It is known that some effort has been expended to investigate the possible use of delayed neutrons as the signal carrier for the presence of SNM. To demonstrate the advantages of the embodiments of the present invention, an evaluation of the relative merits of signals from delayed neutrons and the high-energy gamma rays from short-lived fission products is presented below. This evaluation shows the effectiveness of the embodiments of the present invention for the worst-case scenario, and clearly demonstrates that high-energy gamma rays from the decay of fission products offer a significant advantage in comparison to the signals from delayed neutrons. Delayed Neutrons The yields of delayed neutrons from thermal fission of 235U and 239Pu are about 0.017 and 0.0065 per fission, respectively. The half lives of the delayed neutron precursors lie in the range of about 0.1-56 s, and the ENDFB-IV nuclear data set energy spectra are shown in FIG. 1. The data on the yield of delayed neutrons shows that approximately half of the intensity has an energy less than 0.6 MeV and there are very few neutrons with energies above about 1.5 MeV. Because of thermalization and capture of the neutrons in hydrogen, there may be a very small probability for escape of delayed neutrons to an external detector. The results of calculations using nuclear engineering texts show that the root mean squared distance from birth of a 2 MeV neutron at the target until its absorption in hydrogen is about 15 cm in water at normal density, and thus the effective distance that must be traversed through normal water from the target to the detector is approximately 60 cm. The probability for escape of neutrons to a detector can be approximately estimated in two ways. Beyond about 40 cm from a point source of fission neutrons in water, the flux of neutrons with energies En>1 MeV is approximately G ⁡ ( r ) = 0.12 ⁢ ⁢ ⅇ - 0.103 ⁢ ⁢ r w 4 ⁢ π ⁢ ⁢ r 2 ⁢ ⁢ cm - 2 ⁡ ( source ⁢ ⁢ particle ) - 1 , where rw is the distance penetrated in water at normal density. The quantity 4 πr2 G(r), representing the probability of survival per source particle independent of the 1/r2 flux loss that will affect all radiations emitted from the source, is found to be about 2.5×10−4 (source particle)−1. Because the average thermal neutron will be captured in hydrogen within a few cm of where it is produced, this is a measure of the probability that any fission neutron will produce a thermal neutron that escapes to a detector. A second estimate is obtained from the Fermi-age approximation. This gives the spatial distribution of the neutron density of a given energy that has slowed down from some source energy as q ⁡ ( r , τ ) = ⅇ - r 2 / 4 ⁢ τ ( 4 ⁢ π ⁢ ⁢ τ ) 3 / 2 ⁢ cm - 3 ⁢ ⁢ ( source ⁢ ⁢ particle ) - 1 , where τ is the Fermi age in cm2. The approximate value of τ for thermal neutrons slowing down from a fission source is 31 cm2 in water. Estimating the velocity of a thermal neutron as 2200 m s−1, the quantity 4 πr2 q(r,τ) v is about 2.0×10−6 (source particle)−1. Although both estimates are rather rough approximations, they clearly indicate a very low probability of a fission neutron producing a thermal neutron that can escape to a detector. Because of their smaller average energies, the attenuation of delayed neutrons is expected to be significantly larger than for fission neutrons and thus the probability that they can produce a thermal neutron that can escape to a detector is expected to be smaller yet. The conclusion is that the direct observation of delayed neutrons under the assumed limiting conditions will likely afford a very low sensitivity for detecting SNM. On the other hand, indirect observation of the delayed neutrons is possible via their capture by hydrogen (“H”) to produce 2.2 MeV gamma rays (or by capture by other nuclides in more realistic situations). The attenuation coefficient for 2 MeV gamma rays in water is about 0.049 cm−1. Neglecting the size of the target and the 1/r2 flux loss, the probability for escape of such photons to a detector uncollided would be about 0.053. So, in effect, the direct or indirect observation of delayed neutrons under the assumed limiting conditions will likely afford a very low sensitivity for detecting SNM. Considering that delayed neutrons afford a very low sensitivity for detecting SNM in cargo or other containers, the inventors herein have focused their efforts on the detection of gamma rays from short-lived fission products. The inventors herein have demonstrated that gamma rays from short-lived fission products escape to a detector with significantly higher probability than the delayed neutrons or the capture gamma rays that result from them. Delayed Gamma Rays from Short-Lived Fission Products It is known that approximately 90% of the total yield of fission products from thermal fission of 235U is contained in 32 mass chains located at A=88-103 and A=131-146. For thermal fission of 239Pu, the light-massed peak increases in mass number by about 8-10 but the heavy massed peak remains fixed. Because the charge distribution is so narrow (FWHM ˜1.4 e), the majority of the chain yield will be found in one or two nuclides. A nuclide produced with Z=ZP, where ZP is the most probable atomic number for a given mass number, has a yield of about 0.5 of the chain yield. The values of ZP for 239Pu fission are 0.2-0.3 e greater than for 235U fission and thus essentially the same nuclides are considered for a fixed mass number in the two cases. For orientation purposes, only those nuclides with half lives less than a few minutes, and for which the probability for emission of a gamma ray with Eγ>4.0 MeV is at least 10−2 per decay, are directly considered. In Table 1 are the nuclides of interest and their relevant properties. TABLE 1Short-lived, high-yield fission products with probability >0.01 for emission of γ-rays with Eγ > 4.0 MeVHalf-LifeEγ—235U239Pu235U239PuNuclide(s)(keV)Iγ—(%)CY (%)CY (%)Iγ—f−1 (%)Iγ—f−1 (%)Br-8655.11.60.48954074.60.0007360.00022555192.80.0004480.00013762110.580.00009282.84E−05Br-8755.62.030.69418140.0008120.00027646452.20.00044660.00015247841.80.00036540.000124496220.0004060.00013851950.530.000107593.66E−0552010.550.00011165 3.8E−0554740.380.000077142.62E−05Br-8816.31.780.5140221.510.00026878 7.7E−05414840.0007120.00020444951.20.00021366.12E−0545633.20.00056960.00016347221.760.000313288.98E−0549861.950.00034719.95E−0550201.510.00026878 7.7E−0551970.950.00016914.85E−0552120.640.000113923.26E−0552960.720.000128163.67E−0554560.640.000113923.26E−05Br-894.351.090.3540861.80.00019620.00006341663.80.00041420.00013343541.20.00013080.00004245020.880.000095923.08E−05Rb-901584.51.2841366.70.0030150.000858436680.00360.00102446462.250.00101250.00028851871.170.00052650.00015 Rb-90m2581.240.74241931.140.000141368.46E−0544541.180.000146328.76E−05Rb-9158.45.582.140784.10.00228780.00086142651.40.00078120.000294Rb-924.54.821.9246382.20.00106040.00042248091.10.00053020.000211483610.0004820.00019249231.10.00053020.00021151882.50.0012050.00048 52151.10.00053020.00021152491.10.00053020.00021155841.70.00081940.000326563220.0009640.00038457390.70.00033740.00013458790.70.00033740.00013459010.90.00043380.00017360040.590.000284380.00011360300.790.000380780.00015261150.80.00038560.000154Sr-9523.95.273.0140751.220.0006430.000367Y-980.551.921.5244508.90.001710.001353I-136m46.91.261.6545601.410.000177660.00023348892.20.00027720.00036350910.540.000068048.91E−0551871.040.000131040.00017252550.580.000073089.57E−05Totals31.014.30.031060.0122 The fifth and sixth columns of Table 1 provide cumulative yields of the nuclides, and the seventh and eight columns provide the absolute intensity of gamma rays per fission. As is shown above, eleven nuclides are listed in the table with half lives in the range 0.55-158 s. All but one have half lives in essentially the same range as the delayed neutrons. For most of the nuclides, the cumulative yield is significantly larger than the independent yield and that implies that an additional ten or so nuclides with comparable or shorter half lives might have significant probabilities for emission of high-energy gamma rays. The total of the cumulative yields of the eleven nuclides is approximately twice as large for fission of 235U as it is for 239Pu. The total probability per fission for observing a gamma ray with Eγ>4.0 MeV from decay of these nuclides is about 0.031 and 0.012, respectively, for the two fission systems. These are about a factor of two larger than the delayed neutron yields and represent conservative estimates. The attenuation coefficient for 4 MeV gamma rays in H2O is 0.034 cm−1, and thus 13% of such gamma rays would escape from the container (e.g., 1.5 m distance) uncollided as compared to about 5.3% for 2 MeV gamma rays. If the 2.2 MeV photons from neutron capture on hydrogen were used as a surrogates for delayed neutrons, the high-energy gamma rays from the fission products offer, conservatively, a factor of about 5 larger probability for escape to a detector. While the capture photons are monoenergetic, the fission product gamma rays vary considerably in energy. Unless one used a high-resolution instrument, such as a germanium (“Ge”) detector, one will not be able to resolve these lines but one would also be unlikely to distinguish the capture photons either. Thus, what one is looking for is an elevated continuum that lasts for a few minutes following the neutron burst. The use of gamma ray detection for discovering illicit SNM may be limited by both the natural background and by the decay of activation products, especially those with half lives on the order of seconds or minutes. The natural background is dominated by a gamma ray at 1.461 MeV (40K) and the highest energy line of high intensity is that at 2.614 MeV (208Tl). Apart from very weak lines resulting from neutron capture of the terrestrial neutron background and rare high-energy interactions, no gamma ray lines with energies exceeding 4 MeV are found. The characteristics of short-lived activation products with lifetimes comparable to the fission products listed in Table 1 are shown in Table 2. TABLE 2Neutron activation products with short half livesEγ—Half LifeI— > 2.0ThreshAct. Prod.Reaction(MeV)(sec)Iγ—(abs)MeV (abs)(MeV)% abundC-1518O(n, a)5.32.40.635.290.2N-1616O(n, p)6.17.10.6710.2599.87.10.049Na-2626Mg(n, p)2.521.1.070 > 2.08.86112.54Al-3030Si(n, p)2.233.61.05 > 2.08.043.12.63.5K-4444Ca(n, p)2.151326>0.4 (>2.0)4.992.092.523.66S-3736Ar(n, γ)3.13000.9400.3440Ar(n, a)2.5699.6 With the exception of the (n,γ) and (n,α) reactions on the Argon (Ar) isotopes, all of the other reactions have thresholds greater than about 5.0 MeV, and, if D-D neutrons are used as the interrogation source, these reactions will not take place. Ar comprises about 0.93% of air. The (n,α) excitation function on 40Ar shows a maximum of 0.02 b at an energy of about 8.7 MeV and drops to less than about 0.001 b at 5.0 MeV. Thus, with D-D neutrons, the source produced by this reaction is expected to be very weak. Therefore, in the zeroth order, the (n,γ) cross section on 36Ar may be neglected because of its low atomic abundance. If D-T neutrons are used, the production of these interfering nuclides will take place only in that volume where the neutrons have not been moderated enough to drop their energies below about 5 MeV. While attractive from the point of view of minimizing interference from activation products, the use of D-D neutrons comes with the handicap of a production cross section of about a factor of 100 less than that possible with D-T sources, a deficit that may be too large to incur. However, by using a partially moderated D-T source the fraction of incident neutrons that lies above 5 MeV is substantially reduced. As an example, it may be possible to surround the D-T source with Be of an optimum thickness determined by detailed Monte Carlo calculations. Therefore, the intensity enhancement from a D-T target may be maintained without undue production of neutron activation products. Regardless of which neutron source is chosen, the average neutron that can penetrate to the target will be thermal or very nearly so. In order to detect the presence of SNM using high-energy gammas emitted from fission products, the characteristic energy spectrum and time dependence of these high-energy gammas was measured. In order to do so, it was advantageous to have a switchable high-intensity source of neutrons of variable energy that can be used to irradiate targets of 235U and/or 239Pu. The 88″ Cyclotron at Lawrence Berkeley National Laboratory (“LBNL”) provides such a beam, measurement and shielding facility. At the 88″, neutrons were produced in large numbers by deuteron fragmentation. The 88″ provided deuteron beams up to 60 MeV with currents up to 10 μA. The neutrons were produced on average with half the deuteron energy and their angular distribution was forward peaked. In this manner, a large number of neutrons were directed onto a suitable moderator and then onto the target of interest. A target delivery and transfer system (e.g., rabbit system) was also used at the LBNL facility that enabled the irradiation of the target inside an existing cave and then the transfer of the target to a remote shielded counting station where Ge, sodium iodide (“NaI”), or plastic scintillator detectors were located. In addition, appropriate electronics and data acquisition systems necessary for such measurements, were used to make the measurements. Using the system described above, the feasibility of the methodology and system of the embodiments of the present invention was demonstrated by conducting the following exemplary experiment. A deuteron source (e.g., 1 μA of 16-MeV such as the LBNL's 88″ Cyclotron) was used to bombard a beryllium (“Be”) source to produce source of neutrons. The neutrons were then moderated (i.e., slowed down) using a combination of steel and polyethylene. Highly enriched 235U, depleted U, and 239Pu targets were irradiated with the neutrons and then transported to a shielded counting station using a pneumatic transfer system, as is known to those skilled in the art of nuclide detection. Gamma ray counting was performed with large germanium (“Ge”) scintillator detectors. Time-based data was acquired using an ORTEC NOMAD system running GAMMAVISION. Using this bombardment and detection setup, many gamma particles above 4 MeV were detected and decay curves as a function of energy were determined. These as well as other aspects of the embodiments of the present invention and how it is generalized for containers in general and cargo containers in particular are described below in further detail. The measurement methodology disclosed above describes a method that provides unequivocal signatures of 235U and 239Pu that provides high sensitivity in the presence of thick hydrogenous and other cargos. The system and method in accordance with the embodiments of the present invention is based in part on the relatively high intensity of γ rays with Eγ≧3.0 MeV that are emitted from short-lived fission fragments (e.g., see Chu, S. Y. F., Ekstrom, L. P., and Firestone, R. B., WWW Table of Radioactive Isotopes, http:ie.lbl.gov/toi (1999), and England, T. R. and Rider, B. F., ENDF-349, LA-UR-94-3106 (1994)). These β-delayed γ rays have yields in fission that are approximately an order of magnitude larger than the corresponding β-delayed neutron intensities from the thermal fission of 235U and 239Pu. They are likely to be transmitted through thick hydrogenous material with 102-103 times the probability likely for β-delayed neutrons. Their energies lie above interferences from normal environmental radioactivity. In addition, the energy spectra and time dependencies for emission of the β-delayed γ rays provide unique signatures for 235U and 239Pu. In order to capture the main properties of the high-energy delayed γ rays, the γ-ray spectra following thermal neutron induced fission of 235U and 239Pu was measured. Using the setup and facility described above, neutrons were produced by bombarding a 1-inch thick water-cooled Be target with 16-MeV deuterons from the Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. Neutrons were then moderated using a 15 cm cube of steel surrounded by up to 45 cm of polyethylene. The steel cube was located immediately downstream of the Be target. A pneumatic transfer system shuttled targets between an irradiation location inside the polyethylene and a remote shielded counting station with a transit time of 2-3 s. The thermal neutron flux at the irradiation site was approximately 1.5×106 cm−2 s−1. 235 U (93% isotopic content), 239Pu (95% isotopic content) and, as representative of the characteristics of some cargo loadings, wood, polyethylene, aluminum, sandstone, and steel were irradiated. In each case, targets were repeatedly subjected to cycles of 30-s irradiations followed by 30-s counting periods, during which 10 sequential 3.0-s γ-ray spectra were acquired. Counting began 3 s after the end of irradiation. γ-rays were detected with an 80% relative efficiency coaxial high purity Ge (“HPGe”) detector and with a 30-cm×30-cm×10-cm plastic scintillator. Data were acquired and sorted using ORTEC PC-based electronics and software. FIG. 2 is a γ-ray spectra observed in the HPGe detector in 30 seconds of live time following the neutron irradiation of 0.568 grams of 239Pu and of 115 grams of steel. In order to display these two spectra on the same plot, offsets of 30 and 10 counts per channel were added to the data obtained from the 239Pu and steel targets, respectively. FIG. 2A (inset) is a graph of background-corrected decay curves for gamma rays in the energy intervals 3000-4000 keV and 4000-8000 keV observed from the 239Pu target. Similar results were obtained from a 235U target. FIG. 2 shows γ-ray spectra for E≧1.0 MeV acquired with the HPGe detector from irradiation of 0.568 grams of 239Pu and 115 grams of steel. The temporal behavior of detected high-energy events is shown in the inset, FIG. 2A. Both the energy and temporal distributions of the high-energy γ rays from thermal fission of 235U are quite similar to those shown for 239Pu but their intensity per fission is about a factor of 3 larger. Also, results similar to those shown for steel were found from the irradiation of wood, polyethylene, aluminum and sandstone in the most important characteristic, i.e., no spectrum indicated the presence γ rays with energies exceeding 3.0 MeV. From the steel target, a small number of lower-energy γ rays produced by the decays of long-lived isotopes such as 56Mn (t1/2=2.58 hours) were observed. To the contrary, the spectrum from 239Pu is indicative of fairly intense γ-ray emission at E≧3.0 MeV that extends to at least 5.5-6.0 MeV. It is also clear, as expected, that the high-energy intensity is spread over a relatively large number of lines rather than concentrated in only a few. Thus, a simple and sensitive method to identify fissile material may integrate the total number of events in a wide energy interval, regardless of whether the events represent full- or partial-energy depositions. The results from this type of analysis for the energy intervals 3-4 MeV and 4-8 MeV are shown in FIG. 2A (inset). The integrated numbers of events from irradiated 235U and 239Pu showed decays with a short effective half-life of approximately 25 seconds, whereas those from all other materials tested showed much longer decay times. The two features—large numbers of γ rays with energies above 3.0 MeV and a short effective half-life—are unique signatures of 235U and 239Pu. Because of the high-density of γ-ray lines produced by the decay of fission fragments, a practical system for interrogating large objects does not necessarily require high-resolution detectors, such as the above HPGe detector. For example, the energy spectrum shown in FIG. 2 was generated using a high-resolution detector and thus various sharp energy counts are displayed. However, had a low-resolution detector been used, then the overall triangular shape of the spectrum of FIG. 2, without the sharp lines would have been produced. In fact, essentially the same results shown in FIG. 2 were obtained with the low-resolution plastic scintillator described above. This is particularly significant because such scintillators are sufficiently low in cost that allow one to form a large array of such devices surrounding a cargo container to provide a large solid angle for detecting photons. To demonstrate that a system and method as described above is easily scalable, even all the way up to a large container, and thus yields practical results in reasonable times, the response of an array of detectors following a 30-s irradiation of a cargo container with a source producing 101114 MeV neutrons s−1 is estimated as follows. As a worse-case scenario, in the full-scale system, the cargo is assumed to be wood with a 5-cm (diameter) sphere of 239Pu located at its center. An embodiment of such a full-scale system 300 is shown in FIG. 3. FIG. 3 shows neutron beam source 302 directing neutrons at a cargo container 304 that is suspected of containing SNM 306. Any beam generating system that is capable of providing such a flux may be used with the system of the present invention. For example, a compact linear accelerator, such as a LINAC and an appropriate target (e.g., Be) may be configured to provide the necessary flux. Preferably the beam emits neutrons isotropically so as to adequately scan the container. Alternately, the beam may be an anisotropic beam that is scanned across the container using a scanning system. The neutron beam may be a D-D or a D-T produced beam. The cargo container 304 is surrounded by an array of detectors 308 that are used to detect high-energy gammas that are emitted from fission products produced by the thermal fission of the SNM nuclei by the neutrons that have been moderated on their way to the SNM target 306. The detector or detectors, or array of detectors may be Ge or HPGe detectors or liquid or plastic scintillators, or other suitable gamma ray detectors. In one embodiment, the cargo container 304 is moved on a rail car in position relative to the beam source 302 and the detectors 308. The cargo is then irradiated for a time period (e.g., 30 sec.) and then after the irradiation, counting is conducted for another time period (e.g., 30 sec.). Alternately, the cargo container 304 is placed on a moving conveyor and it is irradiated and counted in a continually moving configuration. The counting period is not limited to a 30-second period, so long as the period is capable of adequately capturing gamma rays having a half-life on the order of 20 to 30 seconds. For a worse-case determination, the 101114 MeV neutrons s−1 beam is considered to be approximately 15 feet away from the container 304. With no attenuation, the neutron flux at a distance of approximately 15 feet will be approximately 3.84 neutrons/cm2-sec (e.g. 1/r2 attenuation). Based on a very conservative estimate that 90% of all neutrons are absorbed by other cargo, then the resulting flux at the SNM target will be approximately 3.83 neutrons/cm2-sec. Integrated over a 30-second irradiation window, the resulting neutron fluence is approximately 1.1×105 n/cm2. Referring to FIG. 4, and considering that a thermal neutron has an attenuation length in 235U or 239Pu of on the order of 1 mm, then the available target mass for a 5 cm diameter target is approximately 500 grams. Using conservative text book calculations, the resulting gamma yield of gamma particles above 3 MeV that are emitted in a 30 second window is approximately 1.0×105 gamma particles. Again using a very conservative estimate and estimating that approximately 10% of the high-energy gammas escape the container, then it is estimated that approximately 1000 high-energy gamma events are expected to be detected in a 30-second counting window following the thermal fission of 239Pu, and approximately 350 detected γ-ray events above 3 MeV for 235U. These very conservative scaling calculations show that with currently available technology, an entire cargo container may be scanned for 235U and 239Pu and other SNM in approximately less than one minute. Possible interferences from activities induced in other materials are few and can be negated substantially by appropriate choice of the interrogating source, as is known to those of skill in the detection of radio nuclides. Furthermore, the system in accordance with the embodiments of the present invention, when combined with a radiographic imaging system, is even more attractive for rapid identification of 235U and 239Pu and other fissile materials in a wide range of applications. As will be understood by those skilled in the art, the present invention may be embodied in other specific forms without departing from the essential characteristics thereof. For example, the source of neutrons may be any source including a D-D or a D-T source that gets moderated on its way to the SNM target to induce a thermal fission in a portion of the SNM. Or that the detectors and their signal processing software and devices may be any setup that is capable of obtaining a time-dependant energy spectrum for the high-energy gamma rays that have been emitted from the fission products of the thermal fission of a portion of the SNM. These other embodiments are intended to be included within the scope of the present invention, which is set forth in the following claims.
043303714	description	Turning now to the drawings, wherein like components are designated by like reference numerals throughout the various figures, a nuclear reactor generally indicated by the reference numeral 10 is illustrated in FIG. 1. In a preferred embodiment, this reactor is a pool type liquid metal fast breeder reactor but may be of any other type compatible with the present invention to be described hereinafter. As seen in FIG. 1, reactor 10 includes a main vessel structure 12 which extends below a reactor cover assembly 14 and which contains a number of reactor components within a pool of hot sodium generally indicated at 16 and a pool of cold sodium generally indicated at 18. The reactor components within the vessel structure include an overall core assembly 20 comprised of a central core 22 containing both fuel and blanket assemblies (not shown) and a coolant plena 24 supported in a fixed position within vessel structure 12 on and by an overall structural arrangement 26 which is constructed in accordance with the present invention and which will be discussed in detail hereinafter. Other reactor components include one or more primary circulation pump assemblies 28 and associated discharge piping interconnecting the pump between the cold pool 18 and plena 24 for passing coolant from the cold pool to and through the core assembly for cooling the fuel and blanket assemblies and thereafter to hot pool 16. At least one intermediate heat exchanger 38, is provided to receive coolant from the hot pool for passage back to the cold pool after heat from the coolant has been transferred to an independent medium. An instrument tree 40 is mounted over the core. With the exception of structural support arrangement 26, all of these reactor components including ones which may or may not have been illustrated but which are not pertinent to the present invention are known and may be readily provided by those with ordinary skill in the art. Therefore, these latter components will not be discussed herein except where necessary for a better understanding of the present invention. In this latter regard, it should be noted that vessel structure 12 includes reactor vessel 42 having a bottommost, closed section 44 as illustrated best in FIG. 3. Referring specifically to FIGS. 2 to 6, attention is now directed to a detailed description of structural arrangement 26 which, as stated previously, is constructed in accordance with the present invention. As will be seen hereinafter, this arrangement includes a grillage of I-beams generally indicated at 46 in FIG. 2, a circumferential box girder 48, best illustrated in FIG. 3, and a circumferential support skirt 50, also best shown in FIG. 3, all of which are interconnected together in the manner to be described for achieving the objectives discussed previously. As best seen in FIG. 2, the grillage 46 of I-beams includes a first group of horizontally extending, spaced apart and substantially parallel I-beams 52 and a second group of horizontally, spaced apart and substantially parallel I-beams 54 extending transverse (preferably normal) to and across each of the I-beams 52 at specific crossing points or junctures 56. Each of the I-beams 52 includes a longitudinally extending intermediate section comprised of a vertically extending center web 58, an integrally formed, continuous lower horizontal flange 60 at the bottom of the web and integrally formed, upper horizontal flange sections 62 at the top of the web. The flange sections 62 define spaces 64 between one another and include upwardly chamfered welding surfaces 66 on opposite sides of each space. Finally, a fabrication access hole 68 is provided through web 58 a predetermined distance between each space 64 along with a vertical slot 70 extending upward from each access hole to the top edge of the web so as to open into the space. Each of the I-beams 54 is similar to I-beams 52 to the extent that it includes an intermediate section comprised of a center web 72 and longitudinally spaced fabrication access holes 74. However, the intermediate section of each I-beam 54 includes an integrally formed continuous flange 76 at the top edge of its web (rather than at its bottom) and discontinuous flange sections 78 at the bottom of its web (rather than at its top). Sections 78 define spaces 80 therebetween and include downwardly directed chamfered surfaces 82 in vertical alignment with associated access holes 74. Finally, each access hole and space 80 include a vertically extending slot 84 therebetween. Only the intermediate sections of the I-beams 52 and 54 have thus far been described. For reasons to be discussed hereinafter, each of these I-beams also includes opposite end sections 86 (in the case of I-beams 52) and 88 (in the case of I-beams 54) which are best shown in FIG. 2. As seen there and in FIG. 6, each of these end sections is flangeless, that is, it includes only web 58 or 72 without an associate flange and flange sections and, while it may include an access hole as shown in FIG. 3, it does not necessarily include an associated slot. As stated previously, the I-beams 52 and 54 cross one another at crossing points or junctures 56. Each I-beam 54 is welded to and mechanically interlocked with each I-beam 52 at each crossing point or juncture 56 in a particular way which is best illustrated in FIG. 4. One of the crossing points is shown therein in an exploded view such that a given access hole 74, its associated space 80 and slot 84 of I-beam 54 are vertically aligned above a given access hole 68, its associated spaces 64, and slot 70 of I-beam 52. The two I-beams illustrated in FIG. 4 are interconnected by lowering the top I-beam 54 into and past the slot 70 and access hole 68 of lower I-beam 52 such that this latter hole and the hole 74 are perpendicular to but otherwise aligned with one another with slot 70 above these holes and slot 84 below them. The space 64 between flange sections 62 is just wide enough to receive an aligned transverse section of flange 76 such that welding surfaces 66 engage the underside of this transverse flange section. In a similar manner, the space 80 is just wide enough to receive an aligned transverse section of flange 60 so that welding surfaces 82 engage the top side of this latter transverse flange section. Each welding surface is welded to its engaging flange surface to provide a welded joint which is maintained in a state of compression. At the same time, the I-beams are mechanically interlocked such that the lower I-beam 52 supports the upper I-beam 54 even if the welded joint should fail. The foregoing has been a description of the way in which a given pair of I-beams 52 and 54 are interconnected at a particular crossing point or juncture 56. It is to be understood that the I-beams are interconnected in this way at each and every crossing point in a preferred embodiment. Referring specifically to FIGS. 2 and 3 in conjunction with FIGS. 5 and 6, attention is now directed to circumferential box girder 48. While not shown, this girder extends entirely around the outer periphery of I-beam grillage 46 in a circular fashion. As seen best in FIG. 3, the box girder includes an inner circumferential, vertical support plate 88 and an outer circumferential support plate 90 which also extend vertically except for a circumferentially extending, outward and downwardly facing step or shoulder 92. The two support plates are interconnected to one another to form a boxed cross section by top and bottom interconnecting plates 94 and 96, respectively (illustrated only in FIG. 2). All of these plates are interconnected together in a suitable manner, preferably by welded joints, to provide the cross-sectional configuration illustrated in FIG. 3. Reinforcement webs 98 may be suitably connected within and made part of the overall box girder at spaced-apart radial positions within the box girder, as seen in FIG. 2. In addition, for reasons to be discussed below, the inner circumferential support plate 88 includes a plurality of circumferentially spaced apart, vertical through slots 100, one of which is best seen in FIG. 6. There is one through slot 100 for each I-beam end section 86 and 88 and each slot is positioned to receive therethrough an associated I-beam end section as illustrated in FIG. 6. As shown there, an end section 86 extends through the slot 100 and engages the inner surface of outer circumferential plate 90. In this regard, in a preferred embodiment, the end of each I-beam end section is preferably stepped in a way which conforms with the stepped configuration of the outer circumferential support plate so as to engage the inner surface of the latter along its entire vertical extent. As seen in FIG. 6, end section 86 is not only mechanically interlocked with the box girder as a result of passing through slot 100 and sitting on step 92, but it is also welded to both plates as indicated by the various weld joints 102 in FIGS. 5 and 6. From the foregoing, it should be apparent that the grillage 46 of I-beams 52 and 54 and the circumferential box girder 48 are welded and mechanically interlocked to form an integral unit. This overall unit is supported in a horizontally extending position on top of the top circumferential edge of previously mentioned support skirt 50. As best seen in FIG. 3, the bottommost circumferential edge of the support skirt rests against and is connected to the inner surface of vessel 42 across the bottom section 44 thereof, preferably by means of a welded joint. The top circumferential edge of the skirt engages circumferential shoulder 92 and is welded thereat by a weld joint maintained in a state of compression. However, it should be apparent that even if the welded joint should fail, the mechanical interlocking configuration between the stepped support plate 90 and the circumferential skirt is such that the latter will continue to support the entire box girder and its associated grillage of I-beams. With structural support arrangement 96 constructed in the foregoing manner, should all the weld joints fail, the various components making up the arrangement are nevertheless mechanically interlocked to provide sufficient structural integrity to support core assembly 20 in a limited downwardly deflected position. The amount of deflection which actually results from a total failure is detectable so that appropriate action may be immediately taken. In an actual working embodiment where the overall structural arrangement includes the dimensions illustrated in the drawing and where the weight being supported is about 2000 tons, the amount of deflection or sag in the structure resulting from a total failure in its welded joints is approximately 2 inches while the deflection in individual joints due to individual failures there have been found to be much less than 0.25 inches (a limit set by the reactivity insertion due to the relative control rod withdrawl of that amount). The advantage of a relatively large total deflection, for example, one as large as 2 inches, before the entire core assembly collapes is that such a deflection is detectable under normal operation of the reactor, specifically by means of the control rod position indicators which would show an increased insertion from the same radioactivity/power level. This means that additional deflection monitors are not necessary. Another advantage of the present arrangement resides in the ability of the grillage of I-beams to be readily assembled in the field. The mechanical interlocks not only serve as a mechanical backup to the welded joints should the latter fail as discussed previously, but also simplifies and improves field fabrication by simplifying alignment of the I-beams and minimizing distortion during welding. The box girder stiffens the overall structure against deflections during seismic events and reduces stresses in the support skirt. The latter is also sufficiently flexible to absorb differential expansion and rotation of the box girder. In a preferred embodiment, the I-beams are constructed of type 304 stainless steel, the plates making up the box girder are constructed of type 304 stainless steel as is the support skirt. However, it is to be understood that the type of material selected as well as the size, shape and strength requirements generally of the individual components making up arrangement 26 will be dictated and readily provided by the overall reactor design.
052767183	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 illustrates the shape of a first embodiment of a control blade for a nuclear reactor according to the present invention, the overall body of the control blade for a nuclear reactor being represented by reference numeral 10. The control blade 10 for use in a nuclear reactor is arranged to be inserted into or withdrawn form the nuclear reactor core by a control-blade drive mechanism (control rod drive) (omitted from illustration). Each of fuel assemblies is a set composed of four fuel assemblies, the fuel assemblies being disposed in the reactor core portion of a boiling water nuclear reactor serving as a light water reactor. As a result of the insertion/withdrawal of the control blade to and from the reactor core portion, the operation of the nuclear reactor can be shut down and as well as the power of the reactor can be adjusted and controlled. The control blade 10 for a nuclear reactor comprises an upper structure member 11 serving as an upper structure means, a lower structure member 12 serving as a lower structure means, a central tie member 13 disposed between the upper structure member 11 and the lower structure member 12 and serving as a central tie means and four rectangular wings 14 connected to the central tie member 13 to form a cross-shaped lateral cross section and serving as a wing means. A handle 15 for handling the control blade 10 for a nuclear reactor is integrally provided for the upper structure member 11. On the other hand, a speed limiter 16 is fastened to the lower portion of the lower structure member 12. A coupling socket 17 is disposed below the speed limiter 16, the coupling socket 17 being detachably fastened to the control blade drive mechanism (omitted from the illustration). Reference numerals 18 and 19 respectively represent guide rollers for smoothly guiding the control blade 10 for a nuclear reactor. The central tie member 13 for connecting the four wings 14 of the control blade 10 may be continuously formed from the upper structure member 11 toward the lower structure member 12. As an alternative to this, a plurality of central tie members 13 are disposed at intervals of 20 to 30 cm. The top end portion 14a of the wing 14 integrally fastened to the central tie member 13 by welding or the like is secured to the upper structure member 11, while the lower portion of the same is secured to the lower structure member 12. Each of the wings 14 is, as shown in FIGS. 2 and 3, formed by a rectangular hafnium metal plate, a metal plate containing hafnium mainly, a plate made of an alloy of hafnium and zirconium (Zr) or an alloy of hafnium and titanium (Ti) or a plate made of an alloy the main component of which is zirconium (Zr) and titanium (Ti). The wing 14 has a multiplicity of accommodating holes 20 disposed in the widthwise direction of the wing 14 for length L which is equivalent to the overall axial length of the core of the nuclear reactor the length L being equivalent to the height of the effective core portion of the nuclear reactor, that is, for example, 3.6 m (144 inches) or 3.7 m (146 inches). Each of the accommodating holes 20 formed in the wing 14 are sectioned into regions (A) to (E) or more regions when viewed from the front end in the direction of the insertion. Hollow tubular members (sleeves) 21 each of which is made of Zr or pure Zr are inserted into the accommodating holes 20 in the region (A), the Zr or pure Zr hollow tubular members 21 each of which is filled with Zr or pure Zr particles (grain) 22 are inserted into those in the region (B) and Zr or pure Zr tubular members 2, into each of which an Hf metal rod (a hafnium alloy rod or Ag-In-Cd alloy rod) 23 which is a long-lived absorbing material member is inserted, are inserted into those in the region (C). Furthermore, B.sub.4 C particles (powder) 27a and Zr or pure Zr particles 27b are, after being mixed, inserted into the accommodating holes 20 in the region (D) and B.sub.4 C particles are inserted into the accommodating holes 20 in the region (E). Since hafnium (Hf) serving as the long-lived type neutron absorber is resonance neutron absorbing material, it is necessary to form into a rod shape having a large surface area with respect to the volume. Hafnium is a metallic element which is extremely chemically stable. A multiple kinds of isotopes Hf-176, Hf-177, Hf-178, Hf-179, Hf-180 and the like exist each of which is able to satisfactorily absorb neutrons and which relatively considerably absorb neutrons of resonance energy. In particular, Hf-177 and Hf-178 exhibit an excellent neutron-absorption effect. For example, Hf-176 absorbs a neutron to change into Hf-177 ,And sequentially absorbs neutrons to change into Hf-180 through Hf-179 via Hf-178. As a result, one Hf atomic nucleus is able to absorb a plurality of neutrons and continues to absorb neutrons for a long time. Therefore, it can be said that hafnium is a long-lived type neutron absorber. In the above-described regions (A) to (D), a Hf metal rod (or an Ag-In-Cd alloy rod) 24 is disposed in the outer side end portion of the wing 14. Furthermore, a sheet (strip) 25 made of Zr or pure Zr is disposed between the Hf metal rod 24 and the end surface of the above-described accommodating holes 20. The Hf metal rod (or the Ag-In-Cd alloy rod) 24 is, as shown in FIGS. 4 to 7, surrounded by a sheet 25 made of Zr or pure Zr. In the above-described region (E), an Hf-Zr alloy rod (or an Hf-Ti alloy rod) 26 is disposed in the outer side end portion (peripheral) of the wing 14. The above-described regions (A) to (D) are exposed to a large amount of neutrons. In particular, the regions (A) to (C) are considerably exposed to the same. Therefore, a boron compound (exemplified by B.sub.4 C and EuB.sub.6), the life of which is relatively short, is not enclosed, but the Hf metal rod (or Ag-In-Cd alloy rod) 23 exhibiting a long neutron-absorbing life is enclosed in the form of the long-lived type neutron absorber. The region (D) is exposed to a relatively large amount of neutrons and is a region the neutron absorption characteristic of which must be improved. Therefore, the boron compound must be used in the region (D) in such a manner that a mixture 27 formed by mixing the Zr or pure Zr particles 27b, which are hydrogen absorbers, and B.sub.4 C 27a is enclosed in order to absorb tritium (.sup.3 T) which is produced as a result of a reaction between boron and the neutrons and hydrogen (H) which can be produced due to the radiolysis of water. The above-described region (A) possesses a function as a gas plenum serving as a portion for absorbing (.sup.3 T) and (H) . It is preferable that the axial length of the region (A) be restricted to a length (about 1 to 3 cm) equivalent to the length of one to three accommodating holes 20. A typical example of the designed control blade according to this embodiment is arranged in such a manner that the thickness of the wing 14 is about 8 mm, the diameter of the accommodating hole 20 is about 6 mm and the distance between the central axes of the two accommodating holes 20 is about 8 mm. Therefore, according to this design example, the length for one hole is about 8 mm (to 1 cm) in a lengthwise direction of the wing 14 and that for three holes is 28 mm (=3.times.8+2.times.2) to 3 cm. In the above-described region (B), absorptions of .sup.3 T and H are mainly performed. In a case where the Zr or pure Zr particles 22 are enclosed, it is preferable that the length (the number) of the holes in a lengthwise direction of the wing in the portion which is the sum of the regions (A) and (B) be restricted to be shorter than 2 to 3 cm (about 2 to 3 holes) for the purpose of preventing the deterioration in the reactivity worth of the control blade. Assuming that the region (A) includes one hole and the region (B) includes two holes, three holes are disposed in the region (A)+(B), causing the length to be about 3 cm similarly to the above-described structure. In any case, since the regions (A) and (B) are not filled with the neutron absorber, the reactivity worth will be deteriorated if the regions (A) and (B) are arranged to be long. The region (C) is a portion in which means to prevent swelling of the long-lived type neutron absorber in such a manner that the tubular member 21 made of Zr or pure Zr absorbs .sup.3 T ahd H and as well as absorbs the generation of stress due to the swelling of the long-lived type neutron absorber. The Zr or pure Zr serving as the hydrogen absorber is a soft material exhibiting excellent hydrogen absorbing performance. A gap is necessarily formed between the accommodating hole 20 and the tubular member 21 made of Zr or pure Zr, the gap as well as serving as the swelling absorbing space. Furthermore, the structure is arranged in such a manner that the Zr particles 27b mixed with B.sub.4 C 27a absorb .sup.3 T and H in the region (D) so as to prevent the diffusion to the other portions. Since the above-described region (E) is exposed to neutrons by a small amount, the ratio of production of .sup.3 T is low and as well as the radiolysis speed is low. Therefore, there is no special means in this region. The length of the regions (A) to (D) of the control blade of a type which is inserted during the operation is usually arranged to be 1/4 to 3/4 of the overall length L of the reactor core and about 1/2.multidot.L of the same. However, it is varied depending upon how the control blade is used. For example, in a case where it is completely removed during the operation, it might be considered feasible to make the total length of (A) to (D) to be shorter than 1/4.multidot.L of the same, for example, about 30 cm or shorter. Then, the reason why the above-described range (1/4 to 3/4 of the overall length L of the reactor core) is employed will now be described. That is, evaluating the actual and average neutron exposure amount distribution in the control blade of a type which is inserted during the operation, the amount of exposure is plane and high in a region of a length of about 1/2 of the overall length L from the front end portion of the insertion of the control blade 10. The amount of neutron exposure increases in a region of about 15 cm (in a region of about 35 cm or shorter) from the leading end portion of the insertion. In particular, an extremely high amount of the exposure is shown in a region of about 5 cm in the front end portion. On the other hand, the amount of the neutron exposure is rapidly reduced at a position of about 1/2.multidot.L from the front end portion of the insertion of the control blade 10 toward the end portion of the insertion. The amount of the neutron exposure is considerably reduced in a region of 1/4.multidot.L from the end portion of the insertion toward the front portion of the insertion. The control blade is changed depending upon how it is used and the above-described value 1/2.multidot.L is changed in a range between 1/4 and 3/4.multidot.L. In the above-described range from (A) to (D), the sheet (strip) 25 made of Zr or pure Zr and disposed on the outer surface of the accommodating hole 20 and the Zr or pure Zr material placed to surround the outer Hf metal rod (or the Ag-In-Cd alloy rod) 24 act similarly to the tubular member 21 disposed in the above-described region (C) and made of pure Zr. The total length of the region (A) to (C) is arranged to be about 15 to 35 cm. If it is longer than the above-described length, it is not preferable because the weight of the control blade will be increased and the reactivity worth can be deteriorated. A second embodiment of a control blade 10A for a nuclear reactor according to the present invention will now be described with reference to FIGS. 9 and 10. Referring to FIGS. 9 and 10, the same or equivalent elements to those according to the above-described first embodiment are given the same reference numerals at the time of making the descriptions about them. Referring to FIG. 9, a range of the front end portion of the insertion designated by symbol (a) is constituted by integrally forming, by welding, metal the main component of which is Hf (for example, Hf metal containing Zr by 2 to 3 wt. %) with an Hf-Zr alloy or an Hf-Ti alloy member accommodating the accommodating holes 20. The length of this range is arranged to be about 3 to 35 cm, usually about 10 to 15 cm. Since this range is exposed to a large amount of neutrons, the Hf material, which is the long-lived absorber, is used as it is (approximating 100%, usually 97%). A range (b) is constituted similarly to that shown in a cross sectional view taken along line II--II of FIG. 2. It is preferable that the number of the accommodating holes be about two or less. In a range (c), the tubular member 21 made of Zr or pure Zr is inserted into the accommodating hole 20. Furthermore, the Hf metal rod (or the Ag-In-Cd alloy rod) 23 the diameter of which is reduced and the length of the same is shortened and the Zr or pure Zr particles 22 which are hydrogen absorbers are alternately inserted into it. The tubular member 21 made of Zr or pure Zr forms a small gap in association with the accommodating hole 20, the small gap serving as a space into which swelling can be received. In addition, the tubular member 21 acts to absorb swelling of the Hf metal rod (or the Ag-In-Cd alloy rod) 23, serves as a hydrogen getter for getting .sup.3 T and H and forms a space for absorbing the lengthwise swelling of the Hf metal rod (or the Ag-In-Cd alloy rod) 23. The total length of (b) and (c) is usually arranged to be about 15 cm. It is preferable that the range (c) has about two or three accommodating holes 20 bored therein. However, since the material for the wing forming the accommodating hole 20 contains the neutron absorber of Hf, it can be eliminated. In the region (d), the tubular memer 21 made of Zr or pure Zr is inserted into the accommodating hole 20. Furthermore, a mixture of the boron compound and the hydrogen absorber made of Zr or pure Zr particles are inserted into the above-described tubular member 21 made of Zr or pure Zr. The tubular member 21 made of Zr or pure Zr forms the swelling space from the accommodating hole 20. Furthermore, the tubular body serves as a hydrogen getter for getting .sup.3 T and H and forms a region for absorbing (relieving) the swelling of boron. The Zr or pure Zr particles mainly serve as the hydrogen getter for getting .sup.3 T and H. A range (e) is constituted similarly to that shown in a cross sectional view taken along line V--V of FIG. 2. Although the structures shown in FIGS. 2 and 9 are described to show the embodiments, the control blade for a nuclear reactor is in actual constituted in a considerably complicated manner because the above-described embodiments are combined. They can, of course, be simplified to meet a desire. For example, the portions shown in the cross sectional views respectively taken along lines II--II and III--III of FIG. 2 can be replaced by the portion shown in the cross sectional view taken along lines IV--IV. Furthermore, the regions (c) and (d) of FIG. 9 may be replaced by structures to be described later referring to FIGS. 11(B) to 11(E) and 11(G). FIGS. 11(A) to 11(I) respectively illustrate structures of the accommodating hole portions of the control blade for a nuclear reactor according to the present invention. Referring to FIG. 11(A), the neutral absorber made of the boron compound (exemplified by B.sub.4 C and EuB.sub.6) 27 is enclosed in the tubular member (sleeve) 21 made of Zr or pure Zr. The Hf metal rod (or the Ag-In-Cd alloy rod) 23 is disposed at the outer side end portion of the wing. Furthermore, the sheet (strip) 25 made of Zr is disposed in the portion from the accommodating hole 20. Referring to FIG. 11(B), the diameter of the Hf metal rod (or the Ag-In-Cd alloy rod) 23 to be inserted is reduced for the purpose of maintaining the swelling space. The Hf metal rod 23 disposed at the outer end portion of the wing is surrounded by the tubular member 21 made of Zr or pure Zr and the sheet 25 made of Zr or pure Zr. Referring to FIG. 11(C), the diameter of the Hf metal rod (or the Ag-In-Cd alloy rod) 23 to be inserted is reduced. Furthermore, a multiplicity of projecting portions 23a which can be easily deformed at the time of swelling are formed in the circumferential direction of the Hf metal rod 23. Referring to FIG. 11(D), the Hf metal rod (or the Ag-In-Cd metal rod) 23 to be inserted is constituted by forming projection 23b around a bolt so that portions to be crushed are formed. Referring to FIG. 11(E), the length of the Hf metal rod (or the Ag-In-Cd metal rod) 23 to be inserted is made to be shorter than the depth of the accommodating hole 20 so that the pure Zr particles hydrogen absorber 22 are inserted in the space created as a result of shortening the above-described length. Referring to FIG. 11(F), the Hf metal rod (or the Ag-In-Cd metal rod) 23 to be inserted are sectioned into a plurality of short pieces (elements) so that the Zr or pure Zr particles 22 are inserted between the short pieces. Referring to FIG. 11(G), the Hf metal rod (or the Ag-In-Cd metal rod) 23 to be inserted is vertically divided in the lengthwise direction. The portions between the vertically divided sections have projections which can be crushed by the swelling or the strips made of pure Zr are placed between the vertically divided sections. Therefore, the former structure is employed to relief the swelling, while the latter structure is able to relief the swelling and as well as is able to serve as a hydrogen getter because the low hardness (soft) pure Zr strip 25 is interposed. Referring to FIG. 11(H), the boron compound 27, for example, B.sub.4 C particles 27a mixed with the Zr or pure Zr particles 27b are enclosed in the accommodating hole 20. Furthermore, Hf particles (grain) 28 are enclosed in the outer side end portion of the wing 14. The Hf particles 28 placed in the outer portion of the wing 14 possesses a function as the long-lived t e neutron absorber and as a hydrogen getter. Since Hf displays restricted hydrogen absorbing performance in comparison to that possessed by the Zr or pure Zr, the particle size of it is made to be smaller, that is powder (small particles) in comparison to the Zr or pure Zr particles 22 to enlarge the surface area. Referring to FIG. 11(I), the Hf particles 28 are enclosed in the portion adjacent to the end portion of the accommodating hole 20. Furthermore, the boron compound 27 are enclosed in the accommodating hole 20. In addition, a Zr string is disposed between the end surface of the accommodating hole 20 and the Hf metal rod 24 surrounded by Zr or pure Zr on the outer side end portion of the wing 14. In this case, the Hf particles 28 possesses the function of the hydrogen getter, which is the hydrogen absorber, and the function of the long-lived type neutron absorber. If an oxide is formed on the surface of the Hf metal rod 23 to be inserted into the accommodating hole 20, the hydrogen absorption of the Hf metal rod 23 can be restricted. Therefore, further swelling can be restricted. The description "further swelling can be restricted" means that additional swelling due to the hydrogen absorption is substantially prevented because the swelling has been taken place in such a manner that the Hf metal rod 23, in which swelling has taken place due to the oxidation, is inserted into the accommodating hole 23 while keeping a certain gap. In an oxide film is formed on the inner surface of the accommodating hole 20, the oxide film thus-formed will prevent the generation of the swelling from the inner surface of the accommodating hole 20 made of the Hf, Hf-Zr, or Hf-Ti alloy and constituting the wing 14 can be prevented or reduced. FIGS. 12 and 13 respectively illustrates a state when a selected neutron absorber or a hydrogen getter is enclosed in the accommodating hole 20 so as to hermetically weld there. The hollow tubular member 21 made of Zr or pure Zr is inserted into the accommodating hole 20, the hollow tubular member 21 being filled with the neutron absorber 27 composed of the B.sub.4 C powder. Furthermore, the Hf metal rod 24 surrounded by the sheet 25 made of Zr or pure Zr is placed in the vicinity of the outer side end portion of the wing. Since the Hf alloy member forming the wing 14, that is, the Hf-Zr member (or the Hf-Ti member), is formed into a shape having an end opening, this portion is turned inside so as to close the outer portion by welding as shown in FIG. 13. Since the Hf alloy which constitutes the wing and the Zr or pure Zr are changed to an alloy in association with each other, a portion of the sheet 25 made of the Zr or pure Zr which surrounds the Hf at the time of the hermetical welding is melted and is moved to the welded portion. However, no problem arises because the above-described alloy is an extremely safety alloy. At the time of welding, an Hf alloy member composed substantially similarly to the wing structural view may be placed on the reverse (inner) side of the welding portion similarly to the conventional structure shown in FIGS. 35 and 36. FIGS. 14 and 15 illustrates a third embodiment of the control blade for a nuclear reactor according to the present invention. The same or equivalent elements to those according to the above-described first embodiment are given the same reference numerals. Although this embodiment is basically arranged in the same manner as the above-described first and second embodiments, the Hf metal rod 23 in which a portion of the above-described embodiments is employed is inserted into the front portion 1.sub.3 of the insertion of the control blade. In a region of length 1.sub.4 (about 1/4 to 1/2 of the overall length L of the effective portion) from the above-described front portion toward the end portion of the insertion of the control blade, the accommodating hole 20 is formed into an elongated hole so that a larger quantity of the neutron absorber composed of the boron compound 27a such as B.sub.4 C is inserted. Furthermore, the pure Zr particle 27b are mixed so as to serve as hydrogen getter (the hydrogen absorber). In this case, the reactivity worth can be improved since a larger quantity of boron is enclosed. Since the ratio of .sup.3 T generation is low in the portion 1.sub.2 and as well as there is no necessity of particularly improving the reactivity worth, usual circular holes are simply arranged in which the neutron absorber such as B.sub.4 C is enclosed. FIG. 16 illustrates a fourth embodiment of a control blade 10C for a nuclear reactor according to the present invention, where the same or equivalent elements to those according to the above-described first embodiment are given the same reference numerals. The control blade according to this embodiment is arranged in such a manner that the front portion 1.sub.1 (about 1/4 to 1/2.multidot.L) of the insertion thereof is structured such that the accommodating hole is formed in the Hf-Zr or Hf-Ti alloy plate and the end portion 1.sub.2 of the insertion is structured such that two Hf plates are disposed to confront each other while holding a gap therebetween. Water is enclosed in the portion between the two Hf plates so as to serve as the neutron moderator and to perform the cooling operation. The prevention of the absorption of .sup.3 T and H from the inside of the accommodating hole and the like are arranged similarly to the above-described first to third embodiments. FIG. 17 illustrates a fifth embodiment of a control blade 10D for a nuclear reactor according to the present invention, where the same or equivalent elements to those according to the above-described first embodiment are given the same reference numerals. The control blade 10(D) according to this embodiment is arranged in such a manner that the wing 2 is constituted by placing the neutron absorber in a U-shaped sheath. In the region 1.sub.1 (about 1/4 to 1/2.multidot.L) of the front portion of the insertion of the control blade, the accommodating hole is formed in the Hf metal plate, Hf-Zr or Hf-Ti alloy of an ordinary composition containing 2 to 3 wt. % Zr. Furthermore, the neutron absorber is enclosed. The end portion 1.sub.2 of the insertion is arranged in such a manner that conventional neutron absorbing rods structured such that the B.sub.4 C powder is enclosed in a stainless steel pipe are arranged. If the U-shape sheath is made of the Hf-Zr or Hf-Ti alloy, the reactivity can be improved and the life can be lengthened. The sheath member may be made of stainless steel in a case where the above-described requirements are not made. The neutron absorbing element to be inserted into the portion 1.sub.1 is sectioned into a portion 1.sub.11 and a portion 1.sub.12. If the density of Hf is made to be high in the portion 1.sub.11, it is effective to improve the reactivity and to lengthen the life. However, since the weight of the control blade is undesirably increased, the density of Hf is lowered in the portion 1.sub.12 so as to reduce the weight and the overall cost. The portion 1.sub.3 acts to restrict a local peaking of the neutron flux due to the reduction in the quantity of the neutron absorber in the boundary portion. Accordingly, the Hf metal is placed. Since the control blade for a nuclear reactor according to this embodiment is arranged in such a manner that a mixture of a material containing boron and at least either the Zr or pure Zr particles or hafnium particles is enclosed in the accommodating hole, hydrogen and tritium can be absorbed and the hydrogen absorption on the inner surface of the accommodating hole can be prevented. Furthermore, since the boron compound surrounded by the zirconium or pure zirconium sheet is enclosed in the accommodating hole formed in the wing, swelling taken place due to the neutron reaction of the boron compound can be absorbed and as well as hydrogen and tritium can be absorbed. As a result, generation of stress in the accommodating hole and hydrogen absorption can be prevented. In addition, since hafnium metal member or Ag-In-Cd alloy member is longitudinally sectioned so as to hold the Zr or pure Zr strip between two confront sides, the zirconium strip acts to a relief portion for stress which acts on the accommodating hole due to swelling because the pure zirconium strip has low hardness. Furthermore, since it exhibits excellent hydrogen absorbing performance, it is able to absorb hydrogen if hydrogen exists prior to other materials. As a result, the hydrogen absorption into the inner surface of the accommodating hole, the Hf metal and Ag-In-Cd alloy can be prevented and as well as swelling taken place caused from the hydrogen absorption can be prevented. Therefore, the generation of stress acting on the accommodating hole can be prevented. Furthermore, the control blade for a nuclear reactor according to this embodiment is arranged in such a manner that a tubular member made of zirconium or pure zirconium exhibiting satisfactory hydrogen absorbing performance is inserted into at least a portion of the accommodating holes disposed in a region of at least 3 cm and as well as 35 cm or shorter from the front portion of the insertion. Furthermore, the portion inside and outside of the tubular member is formed into a non-hermeical shape so as to serve as a gas plenum. As a result, the hydrogen absorption of the hafnium and Ag-In-Cd alloy to be inserted can be prevented. In addition, the generation of stress due to the swelling can be prevented and to restrict the gas pressure rise in the control blade. Consequently, the soundness of the control blade can be satisfactorily maintained. Then, a sixth embodiment of a control blade for a nuclear reactor will now be described. A control blade 30 for a nuclear reactor is a conventional typed control blade as shown in FIG. 18, the control blade 30 having four wings 31 which accommodating an improved neutron absorbing rod 32. Each of the wings 31 comprises sheath plate 33 having a deep U-shape cross section and serving as a sheath plate means in such a manner that the opening sides of the sheath plate 33 is secured to a central tie rod 34 serving as a central tie means in such a manner that the opening sides of the wings 31 are disposed to form a cross-shape lateral cross section. A multiplicity of neutron absorbing rods 32 are disposed in line in the sheath plate 33, the neutron absorbing rod 32 serving as a neutron absorber rod means. Reference numeral 35 represents an upper structure member, 36 represents a lower structure member, 37 represents a handle, 38 represents a speed limiter and 39 represents a coupling socket. As shown in FIG. 19, the neutron absorbing rod 32 has an elongated cladding or covering pipe 40 serving as a poison tube, the covering pipe 40 accommodating neutron absorbing members and material 41 and 42. The two end portions of the covering pipe 40 are closed by plugs 43 serving as plug means. Although the covering pipe 40 is usually made of stainless steel, it may be made of Hf metal, an Hf alloy the main component of which is Hf and Zr or another Hf alloy the main component of which is Hf and Ti. The neutron absorbing rod 32 is mainly sectioned into three portions X, Y and Z. The portion X is a portion exposed to neutron by a large amount when it is accommodated in the control blade 30. Therefore, a long-lived type neutron absorbing member 41 made of Hf metal, an Hf alloy composed of Hf and Zr or Ti, or an Ag-In-Cd alloy is disposed in the portion X. According to this embodiment, the diameter of the long-lived type neutron absorbing member 41 is arranged to be smaller, by a certain quantity, than the inner diameter of the covering pipe 40 as unsealed type inner pipe. Furthermore, a thin sleeve 44 made of pure Zr, Hf, Ti or stainless steel is formed around the long-lived neutron absorbing member 41 is inserted into the covering pipe 40. In the portions Y and Z, the boron compound 42 formed into powder and serving as the neutron absorbing member is disposed. Since the portion Y is exposed to neutron by a relatively large amount, pure Zr particles and/or Hf powder 45 serving as the hydrogen getter (the hydrogen absorber) is mixed with the boron compound 42. Since the portion Z is exposed to neutron by a reduced quantity and thereby the ratio ofgeneration of tritium (.sup.3 T) is low, the hydrogen getter is omitted from the structure. Although the boron compound 42 is exemplified by B.sub.4 C, EuB.sub.6 or BN may be employed. Since the boron compound 42 encounters swelling because He gas is produced as a result of a reaction with neutron, the charging density of the boron compound 42 must be determined after the quantity of swelling which can be generated due to the neutron irradiation has been estimated. Although the length of the portion X can be made to be 3/4 of the overall length L if necessary, it can be set to about 3 to 5 cm if a short length is required. The length must be determined depending upon how to use the control blade 30 which accommodates the neutron absorbing rod 32. In a case where the control blade 30 is considerably inserted into the nuclear core during the operation of the nuclear reactor, it is arranged to be about 1/2 of the overall length L. In case where the control blade 30 is withdrawn from the nuclear core during the operation of the nuclear reactor, it is usually arranged to be about 15 cm. Although the length of the portion Z is usually arranged to be about 1/4 to 3/4 of the overall length L, the length of portions (Y+Z) may be arranged to be a length obtained by subtracting 15 cm from the overall length in a case where the control blade 30 is fully withdrawn from the operation. Furthermore, the length of the portion Y may be made to be zero. A metal wool 46 made of Hf, Zr, stainless steel or iron is enclosed in a portion between the plug 43 and the long-lived type neutron absorbing member 41. Also a metal wool 47 is interposed between the long-lived type neutron absorbing member 41 and the boron compound mixture layer Y for the purpose of preventing the mixture of the powder boron compound with the portion around the long-lived type neutron absorbing member 41. The above-described metal wool 47 may be basically the same as the metal wool 46 placed between the plug 43 and the long life type neutron absorbing member 41. However, it is preferable that wool made of the neutron absorbing material, for example, Hf wool, be employed in a case where the length into which the metal wool 47 is enclosed is longer than about 5 mm. The reason for this lies in that, if non-absorbing material is employed, the neutron flux peak will take place in the above-described portion, causing the adjacent boron compound to be locally exposed to neutron by a large amount. As a result, the soundness for the neutron absorbing element will be deteriorated. On the other hand, the length of the layer in which the metal wool 46 is enclosed is usually made to be about 5 to 10 mm. With the neutron absorbing rod thus-constituted, tritium produced by the boron compound can be absorbed by the Zr particles or the Hf powder mixed with the boron compound so as to prevent the diffusion of it into the long-lived type neutron absorbing member. Furthermore, since the long-lived type neutron absorbing member is surrounded by the sleeve made of pure Zr, Hf or Ti serving as a hydrogen getter or a stress absorber or a sleeve made of stainless steel serving as a stress relaxer, the stress generation in the covering pipe due to swelling can be satisfactorily prevented. FIGS. 24 to 29 respectively illustrates modifications of the portion X shown in FIG. 20, where the equivalent elements to those shown in FIG. 20 are given the same reference numerals. According to a first modification shown in FIG. 24, the diameter of the long-lived type neutron absorbing member 41 is, by a certain quantity, made smaller than the inner diameter of the covering pipe 40 so as to create a gap 50 between the long-lived type neutron absorbing member 41 and the covering pipe 40. Furthermore, the undesirable looseness generated due to the reduction of the diameter of the long-lived type neutron absorbing member 41 is prevented by forming a plurality of small projecting portions 51 in the portions of the long-lived type neutron absorbing member 41. As a result of the structure thus-constituted, the projecting portions 51 can be easily crushed and the generation of large stress in the covering pipe 40 can be prevented even if the long-lived type neutron absorbing member 41 generates swelling. According to a second modification shown in FIG. 25, the local projecting portions 51 according to the first modification are replaced by thread type projecting portions 52 formed on the surface of the long-lived type neutron absorbing member 41. According to this modification, an effect similar to that obtainable according to the first modification can be obtained. FIG. 26 illustrates a third modification which is arranged in such a manner that dimplings 53 are formed for the purpose of making the covering pipe 40 locally projecting and coming contact with the long-lived type neutron absorbing member 41. In a case where the long-lived type neutron absorbing member 41 encounters swelling, the dimplings 53 can easily restore their original shapes. Therefore, the generation of large stress cannot be generated in the covering pipe 40. FIG. 27 illustrates a fourth modification in which the long-lived type neutron absorbing member 41 is sectioned into a multiplicity of short pieces. Furthermore, at least either the Zr or pure Zr particles or the Hf powder 45 is disposed between the pieces thus-formed. According to this modification, the Zr or pure Zr particles or the Hf powder 6 absorbs hydrogen and tritium. Therefore, the prevention of swelling in the long-lived , type neutron absorbing member 41 can be substantially prevented. Although the Zr or pure zr particles and the Hf powder 45 generates swelling, it can be present as it is in the same gap even if the swelling takes place because it is in a low density state. Therefore, the generation of large stress in the covering pipe 40 can be prevented. FIG. 28 illustrates a fifth modification which is arranged in such a manner that the long-lived type neutron absorbing member 41 is longitudinally divided into elongated pieces. Furthermore, small projecting portions 54 which can be easily crushed by swelling are locally formed between the elongated pieces. If swelling takes place, the small projecting portions 54 are sequentially crushed so that the generation of large stress in the covering pipe 40 can be prevented during the above-described effect. Therefore, the time at which stress is generated can be delayed considerably. FIG. 29 illustrates a sixth modification in which the long-lived type neutron absorbing member 41 is, similarly to the fifth modification, longitudinally sectioned. Furthermore, strips 55 made of Zr or pure Zr are interposed in at least a portion of gaps between the elongated pieces. Since strip 56 made of Zr or pure Zr absorbs hydrogen, the generation of swelling in the long-lived type neutron absorbing member 41 can be substantially prevented. Although the strip 55 made of Zr or pure Zr encounters swelling, the generation of large stress in the covering pipe 40 can be satisfactorily prevented because the hardness is small and the gap will absorb the swelling. Although omitted from illustration, a seventh modification is arranged in such a manner that an oxide film is formed on the surface of the long-lived type neutron absorbing member 41. In this case, since the oxide film prevents the hydrogen absorption, the generation of swelling in the long-lived type neutron absorbing member 41 can be prevented. Eighth and ninth modifications are modofications about the portion Y shown in FIG. 20. FIGS. 30 and 31 respectively are vertical cross sectional views of the same. The eighth modification shown in FIG. 30 is arranged in such a manner that the boron compound 42 enclosed in a non-sealed type inner pipe 56 made of pure Zr, Hf or stainless is accommodated in the covering pipe 40. In a case where the inner pipe 56 is made of pure Zr, the inner pipe 56 serves as a hydrogen getter and a stress relaxer, and as well as serves as a neutron absorber. The inner pipe 56 made of stainless steel simply serves as a stress relaxer by forming gaps to prevent the swelling generated due to the He gas of the boron compound. Therefore, it is preferable in this case that a mixture composed by mixing pure Zr particles (grain) or Hf powder (grain) serving as a hydrogen getter with the boron compound 42 be inserted into the inner pipe 56. As a result, diffusion of tritium generated by the boron compound 42 to the long-lived type neutron absorbing member 41 can be prevented. A control blade shown in FIG. 32 has been developed recently and disclosed in Japanese Patent Laid-Open No. 2-254895. FIG. 33 is an enlarged view of a portion L shown in FIG. 32. A neutron absorbing member charge portion 58 is formed into a circular cross section, while the outer portion of the same is formed into a substantially square shape by padding material in units of 90.degree.. By alternately welding the pad portions 59 (to form welded portions 60), the control blade shown in FIG. 32 and having an outer shape which is substantially the same as the shape shown in FIG. 14 can be obtained. However, an apparent difference between the structure shown in FIG. 32 and that shown in FIG. 19 lies in that the sheath 33 is omitted from the structure shown in FIG. 32. Since the thickness of the wing 31 must be the same so as to be mounted on the same nuclear reactor, the diameter of the neutron absorbing member charge portion 58 can be enlarged by a quantity which corresponds to the sheath 33 omitted from the structure. Therefore, a larger amount of neutron absorber absorbers can be enclosed. As a result, the reactivity worth of the control blade for a nuclear reactor can be improved and as well as the nuclear life can be lengthened. As shown in FIGS. 19 and 20, the structure of the neutron absorbing rod 32 can be similarly adapted to a neutron absorbing rod 61 shown in FIGS. 32 and 33 and formed into a substantially square shape. The inventor of the present invention has disclosed the structure of an inner pipe for use as a control blade in Japanese Patent Laid-Open No. 2-2983. Although a structure for preventing the deterioration in the reactivity worth by using a neutron absorber in the plug for the inner pipe has been disclosed in Japanese Patent Laid-Open No. 2-2983, the present invention is arranged in such a manner that the pure Zr sleeve (inner pipe) serving as a hydrogen getter is used. Although a similar structure of an inner pipe has been disclosed in Japanese Patent Laid-Open No. 2-13888, the concept of the hydrogen getter is not included. As can be understood from the above-made description, the neutron absorbing rod is constituted in such a manner that the covering pipe can be protected from excessively large stress. Therefore, a long-lived type neutron absorbing element exhibiting an extremely improved soundness can be provided. Although the embodiments of the neutron absorbing rod according to the present invention is adapted to a control blade for use in a water boiling nuclear reactor, the present invention is not limited to this. The present invention can be adapted to a control blade for use in a pressurized water reactor. Furthermore, the structure of the neutron absorbing rod can be adapted to a control blade for use in a light water reactor, a heavy water reactor, a converter reactor or a fast breeder reactor. Although the invention has been described in its preferred form with a certain degree of particularly, it is understood that the present disclosure of the preferred form has been changed in the details of construction and the combination and arrangement of parts may be resorted to without departing from the spirit and the scope of the invention as hereinafter claimed.
060027347	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The preferred embodiment of the invention will be disclosed in multiple sections which cover the apparatus preferred mode of operation, system calibration and results. FIG. 1a is a schematic side view of an ore assay apparatus 10 according to the present invention. The system is intended for detecting the presence in a sample 11 of certain assay elements whose nuclei have relatively long-lived isomeric states. The apparatus includes an irradiation system 12 for irradiating the sample, a detector system 15, preferably removed from the vicinity of irradiation system 12, for detecting and quantifying the intensity of characteristic decay products, and a sample transport system 18 for moving the sample. The sample 11 in the irradiation position is drawn in solid lines, and in broken lines as a phantom sample in the detection position. The sample 11 preferably consists of a cylindrical holder filled with ore of a known weight. Essentially no preparation of the sample, such as drying and crushing, is required. The sample holders or "pans" are preferably filled with ore from a hopper (not shown) and given an identifying mark such as a bar code so that they can be tracked through the assay system. It is preferred that the sample containers be vibrated during filling so that the maximum amount of ore can be effectively "packed" into the sample holder in order to obtain maximum assay sensitivity for given irradiation and count times. Again referring to FIG. 1a, the irradiation system 12 operates to produce a relatively intense beam of gamma rays in the 6-9 MeV energy range. This range of gamma ray energy, while not critical, provides gamma rays of high enough energy to produce significant isomeric excitation, but not so high as to cause photoneutron and photofission processes. The irradiation system 12 preferably includes an electron linear accelerator 20, which is often referred to as a "linac" for brevity. The output beam of the linac 20 impinges upon a target 22 comprising material with a large atomic number Z such as tungsten. This will be referred to as a "high-Z" target. The resultant gamma rays are preferably collimated to a 20-30 degree cone by a conic collimator 23 and directed to the sample 11 and to a beam stop 25. An electron linear accelerator is preferred over an isotopic source of gamma because it produces copious photons, on the order of 10.sup.15 to 10.sup.16 photons/second, which is equivalent to a megaCurie isotopic source, although .sup.60 Co can be used if other operating conditions are suitable. A suitable linear accelerator is the Linatron Model 3000 manufactured and sold by Varian, Inc. The sample pan is preferably 12 to 14 inches (30 to 36 centimeters) in diameter, 1 to 3 in. (2.5 to 8 cm) thick, and contains approximately 2-3 liters of ore material. The sample 11 is mounted on a sample positioning apparatus 13 and is only partially exposed to the flux of gamma radiation. Referring to both FIG. 1a and to the top view of the sample shown in FIG. 1b, a portion or sector of sample, which is preferably defined by a 60 degree arc 14, is oscillated by the sample positioning apparatus 13 about the sample axis 16 during irradiation of this portion of the sample in order to obtain uniform exposure of the portion to the gamma ray flux. When a portion of the sample 11 has been irradiated for a sufficient length of time, the irradiation system 12 is turned off. As shown in FIG. 1a, the sample is then rapidly moved by means of a transportation system 18 or "trolley" to a detector system 15 for analysis. The trolley is operated under the control of a trolley motor and control system 40. Additional details of the irradiation, isomeric excitation, and decay process are provided in a subsequent section of this disclosure entitled "OPERATION". Detector system 15 includes an array of detectors 30 and an array of detectors 31 positioned above and below the sample 11, respectively, so that the irradiated portion defined by the arc 14 is exposed to the top detector array 30 and the lower detector array 31. The relative positions of sample and detector are best shown in FIG. 1c and in FIG. 3 which is a top view of the sample/detector geometry. After counting for a sufficient time, the sample is returned to the irradiation position and rotated by the sample positioning apparatus 13, under the control of a sample position control system 42, so that the next sequential portion of the sample 11 is exposed to the gamma ray flux. After irradiation, the sample is again transported by the trolley 18 to the detector system 15 for counting of that irradiated segment. This process is continued until all sample segments have been irradiated and counted. Using segment portions defined by the arc 14 of 60 degrees, the irradiation-count sequence is repeated six times. All counts are then combined to obtain a representative assay of the entire sample 11, as will be detailed in a subsequent section. The segmented irradiation and counting of the sample, and the oscillation of the sample during radiation, not only reduces the adverse effects of sample non homogeneity but also reduces the adverse effects caused by any variation of sample geometry. Again referring to FIG. 1a, each detector array 30 and 31 preferably includes clusters of high resolution germanium (Ge) detectors which are cooled by liquid nitrogen cryostat systems 33 and 35 or another type of electrical or mechanical cooling apparatus, respectfully, with associated electronics 46. Details of the detector system 15 will be presented in subsequent sections of this disclosure. Although high resolution type detectors are preferred, the gamma ray detectors 30 and 31 may be of any type suitable for detecting gamma rays in the range of about 0.05-1.0 MeV which is the range of typical isomeric transitions). An example of an alternate detector would be a sodium iodide scintillation crystal optically coupled to a photomultiplier tube. As an additional alternate, only one gamma ray detector can be used resulting in a loss of detection efficiency. Arrays of liquid nitrogen cooled germanium diode detectors are desirable since they have sufficient energy resolution to resolve gamma radiations with nearly the same energies, such as the gold isomeric photon emission "line" at 279.5 KeV from neighboring thorium line at 278 KeV. The gamma ray energies and isomeric half-lives for selected elements are set in the table below. ______________________________________ Element Energy (KeV) Half-Life (sec.) ______________________________________ Gold 279.5 7.2 Silver 88 & 92 42 Barium 662 156 Iridium 130 4.9 Hafnium 220 19 ______________________________________ Shielding structure is required to prevent neutrons and gamma rays produced at the target 22 from reaching the detectors 30 and 31. This is considered as "background" in the assay process. To this end, a body 35 of low-Z material (such as paraffin) is disposed near the accelerator assembly 12 to thermalize the neutrons, and high Z shielding 37 (such as lead) is disposed between the accelerator assembly 12 and the detector assembly 15. Although the accelerator assembly is preferably constructed from materials such that neutrons are not produced by the photoneutron or photofission processes, there is no assurance that the irradiation of the ore material will not generate neutrons by these processes. The low Z shielding material 35 thermalizes these neutrons and preferably contains materials such as boron to capture the thermal neutrons before they can reach the detector assembly 15. Gamma radiation is obviously generated in the vicinity of the accelerator assembly 12 from electrons impinging upon the target 22, from various photon reactions within the sample and surrounding material, from neutron capture reactions, and from other processes. Most of this gamma radiation is absorbed by the high Z shielding material 37. The detector assembly 15 is shielded by additional high Z gamma ray shielding material 39 to isolate the detectors from naturally occurring gamma ray emitters such as thorium, uranium and potassium isotopes, and from extraneous gamma radiation from the accelerator assembly that might penetrate the shield 37. As an additional background reduction means, cadmium jackets about 1/8-inch thick are placed around each detector cluster to absorb any thermal neutrons not absorbed in the shield 35. FIG. 2 is a functional block diagram of the system. Simply stated, it illustrates means for performing and controlling the sample position, irradiate, count, and count analysis operations previously described. It should be understood, however, that the system can be configured in other ways with other equipment to perform the same basic steps of the invention. As shown in FIG. 2, a linac controller 40 under the control of a clock 80 initiates and terminates the operation of the linac based accelerator system according to predetermined irradiation and quiescent (or count) times. The clock 80 and the linac controller 50 cooperate with the trolley motor and control system 40 (see FIG. 1a) to move the sample 11 to and from the irradiation position and the count position. The trolley motor and control system 40 includes a transport controller 52 which generates a trolley motor signal 54 which, in turn, initiates and terminates a trolley motor 56 thereby conveying the sample to and from the irradiation and the counting position. The transport controller 52 also operates the sample position control 42 thereby positing the sample 11 such that the desired portions defined by the arc 14 are irradiated and counted. Still referring to FIG. 2, the transport controller 52 and the clock 80 also inhibit and enable the detector electronics and controls 46 so that the irradiated sample is counted at predetermined counting intervals and in a predetermined sequence. Since the detectors produce electronic impulses or "pulses" whose amplitudes are generally proportional to impinging photon energy, a spectrum of measured gamma radiation can be obtained by sorting pulses as a function of amplitude or "height". A peak in the resulting histogram usually indicates monoenergetic gamma radiation of an energy corresponding to that amplitude, and can therefore be used to measure both energy and intensity of impinging gamma radiation. Pulses representative of the energy of gamma radiation impinging upon the detector clusters in the detector system 15 are input into the electronics 46 which comprises an amplifier 60, an analog-to-digital converter 62, and a histogram memory 64. This forms a detected "spectrum" representing a plot of detected gamma ray intensity as a function of gamma ray energy. Functions of the elements 62 and 64 can be performed by commercially available pulse height analyzers. Measured spectra and sample weights are then preferably input into a personal computer (PC) 70 for analysis in which measured intensities of gamma radiation of specific energy are converted into assay concentrations. These assay results can be stored in a storage device or transferred to another computer 74 for additional analysis, combination with assay results from a plurality of other assay systems, and the like. FIG. 3 illustrates in more detail the arrangement of the gamma ray detector cluster 30 and 31. The top cluster preferably comprises six germanium diode (Ge) detectors. The detectors are preferably the planar type manufactured by Canberra Industries, Inc. Clusters of three Ge detectors 92 are mounted preferably on a common cooling element of a "Trident" cryostat. As shown in FIG. 3 and FIG. 1a, six Ge detectors are configured about the arc 14 of sample 11, above the sample, forming the cluster 30. Six Ge detectors configured in the identical geometry are positioned below the sample as cluster 31. This arrangement optimizes the sensitivity of the detector system 15 to activity induced within the sample by the accelerator irradiation system 12. FIG. 4 is a measured spectrum of counts as a function of photon energy obtained with the Ge detector clusters shown in FIG. 3, showing a representative detector output in a photon range that spans the gold peak identified by the numeral 100 at the characteristic energy of 279.5 KeV. The spectrum is typically a sum of spectra recorded in each of the twelve individual gamma ray detectors and for each of the six sample sectors. The area under the peak 100 is proportional to the intensity of 279.5 KeV gamma radiation impinging upon the detector. This area is determined by subtracting an appropriate "background" level 102 of counts from the total counts recorded in the energy "window" which encompassed the peak 100 from a low energy identified by the numeral 104 to a high energy identified by the numeral 106. Other approaches, such as spectrum fitting, can be used to determine the contribution to the spectrum from the decay of the gold and other isomer. This process is well known in the art and computer software is commercially available to perform such calculations. Energy resolution is of prime importance in obtaining accurate assay results. As given in a previous example, the resolution of the detector system 15 must be sufficient to resolve the desired gold peak 100 from an emission at 277 KeV from naturally occurring thorium which is commonly found in ore. Additional improvements in the accuracy of the assay measurement can be achieved by removing spectral interference from the detector output by such techniques as assaying a sample devoid of the assay element, but otherwise like the unknown samples, and subtracting the interference from the detector output measured from the unknown sample. Additionally, known spectral response functions of the gamma detector to monoenergetic gamma rays can be used to subtract the background underlying the assay peak using the previously mention methodology of spectrum fitting. A consistent theme in the design of the assay system 10 is the minimization of background radiation in order to allow detection of the sometimes low intensity of the gamma rays resulting from the isomeric transitions. To this end, those components in the accelerator irradiation system 12 that are exposed to the gamma ray beam are fabricated of materials having an energy threshold for photoneutron production that is greater than the maximum gamma ray energy. This serves to prevent the production of neutrons which can cause neutron activation of the sample and the gamma ray detector. Unfortunately, no such control can be employed over the materials contained within the ore sample and photoneutrons can be produced. The shielding structure serves to prevent nuclear activation of the gamma detectors by those neutrons unavoidably produced in the sample during gamma irradiation. Gamma ray detectors 30 and 31 may also be shielded from any fission-produced delayed neutrons emerging from the sample by surrounding them with a layer of low-z material such as water, plastic, or paraffin. The steps taken to reduce unwanted background radiation ensure that the signals produced by gamma ray detector assembly 15 result primarily from the gamma rays from the assay elements in the sample. It is also important that the detection and measurement process enhance the desired signal in order to give the maximum sensitivity and accuracy. In this regard, the timing of the overall process is important, and preferably entails irradiation of the sample for the time approximately equal to the half life of the element being assayed. This process may be repeated until the measurement precision is sufficiently high. In such a case, sufficient time must be allowed between successive irradiations so that the activities of other species half lives greater than that of the element being assayed decay to a negligible level. OPERATION The operation of the process can be described mathematically using the following relationships between sample activity A and the irradiation time, delay time, and count time. For irradiation that is uniform (or rapidly and uniformly pulsed), the activity builds up during irradiation according to the formula: EQU A=kI.sigma.c(1-exp(-0.693T/.tau.))/.tau. where A is the activity; PA1 k is the constant depending on the irradiation and target geometry; PA1 I is the current of electrons of energy E striking the target to form the gamma ray beam (the intensity of which is proportional to I); PA1 .sigma. is the average activation cross section for an X-ray spectrum with maximum energy E; PA1 c is the concentration of the assay element; PA1 T is the total irradiation time; and PA1 .tau. is the half-life of the activation product. After a uniform irradiation, the activity decays exponentially as follows: EQU A=kI.sigma.c(1-exp(-0.693T/.tau.))/.tau.exp(-0.693t/.tau.) where t is the time elapsed since the end of irradiation. The gamma rays emitted by the isomers may be uniquely associated with the assay elements by measuring either or both of the gamma ray energy and the decay time of the detected radiation. Irradiation and count times are selected to yield optimum statistical accuracy of the measurement, while meeting reasonable assay throughput required in commercial applications of the system. Operation of the assay system 10 is summarized with reference to FIGS. 1a, 1b, 1c and 2. (1) Sample material is loaded into a sample holder pan 11, and the weight of the sample material is determined. (2) The sample is placed on the sample position apparatus 13 and oscillated about its axis 16 for 5 seconds while irradiating the first of six sample segments with the accelerator irradiation system 12. (3) The sample transport trolley 18 is then activated by the trolley motor and controller 40 to move the sample 11 to the detector assembly 15 while maintaining the proper sample orientation. (4) After a five second count period, the sample position apparatus 13 indexes the sample 11 to the next segmental position, the trolley 18 returns the sample to the irradiation position, and the irradiation-count cycle is repeated until all six sample segments have been irradiated and counted. (5) Counts and sample weight are transferred to the PC 70 where computer software converts the net counts pertinent to the assay material (e.g. gold) and any significant moisture content measured in the sample into an assay, and also computes the statistical error associated with the assay. If the statistical error is above a predetermined level, the sample may be passed through the irradiation-count cycle again in order to reduce this statistical error. (6) After assaying, the sample is placed on an output elevating conveyor (not shown) for return to the sample loading area. (7) Assay results from multiple assays and optionally from multiple assay systems are input into the computer 74 for tabulation and for additional analysis. It should be understood that the parameters used in the above example are typical and preferred for gold ore assay, but the parameters such as irradiation and count times must be varied with the physical properties of time constants for other elements, as can the volume and dimensions of the sample, without destroying the integrity of the assay system. CALIBRATION The system is calibrated by using samples representative of materials such as ores and containing known amounts of assay materials, using gold as an example. It is preferred that these known amounts are of the same orders of magnitudes that are expected from actual assays. It is also preferred that the calibration sample matrix is similar in weight, density, elemental concentration and moisture content to the ore material. The calibration samples are irradiated and counted using the same timing parameters as those used for ore assays. FIG. 5 is a plot of measured counts, C, in the gold peak as a function of know concentration of gold, M.sub.Au, in the calibration samples. FIG. 5 is used to graphically illustrate how a calibration relationship or calibration "curve" 150 is obtained by fitting the analysis of four calibration samples, and how this calibration relationship is subsequently used to obtain a quantitative ore assay from measured counts C in the gold peak. The system can be calibrated in any units related to the gold content of the ore, such as counts per ounce/ton, troy ounce per short ton or kilo-rad per oz./ton. The sample containing the lowest concentration 140 of gold yields a count 141 and is plotted as point 130. Results for calibration samples containing progressively higher known concentrations of gold are analyzed and plotted as points 131, 132, and 133, respectively. A curve 150 is then fitted through the four calibration points thereby yielding the desired calibration relationship for the assay apparatus. RESULTS The calibration relationship, graphically represented by the curve 150 in FIG. 5, is subsequently used to convert counts from ore samples into assay results. As an example, assume that the ore sample yields a gold count C represented by the point 134. A horizontal line projected from this point intersects the calibration curve 150 at a point 137, and a vertical line projected from the point 137 intersects the abscissa representing M.sub.Au at a point 136, thereby yielding the gold content of the ore. It should be understood that the above discussion and FIG. 5 are presented graphically for purposes of illustration, and the actual calibration relationship and assay determination are performed arithmetically in the computer 70. If the ore and sample calibrations differ significantly in geometry, the resulting assay results must be corrected for changes in irradiation geometry, counting geometry, and sample self absorption of the gold gamma ray emission using techniques known in the art. The ore samples must also be normalized to a weight defined by the calibration relationship as is well known in the art. It is again emphasized, however, that the segmented irradiation and counting of the sample and the oscillation of the sample during irradiation greatly decreases the dependence of the assay system upon sample geometry. Furthermore, if the moisture content or matrix of the calibration and ore samples differ significantly, assay results must be corrected for these factors using techniques known in the art. ALTERNATE EMBODIMENT The invention can be embodied to provide a non destructively analysis system for any element which is susceptible to photon excitation and which produces an isotope or an isomer which decay by the emission of radiation which can be identified and quantified. A functional diagram of such an embodiment is shown in FIG. 6. The irradiation and detector systems are again denoted by the numerals 15 and 12, respectfully. In this embodiment, the analyzed sample can be conveyed from the irradiation system 12 to the detector system 15 for counting. An example of such a sample is a piece of metal which is being analyzed to determine its gold content, or silver content, or barium content, or the content of any element susceptible to a photon excitation which yields a decay radiation which can be quantified and identified with the assay element of interest. Alternately, the sample can be left in place, and a representative portion of the sample can be irradiated with the irradiation system, removed after irradiation, and replaced with the detector system 15. An example of such an analysis is an airplane wing, where the irradiation system is placed at a specified spot for irradiation, subsequently removed after irradiation, and replaced at that same spot with the detector assembly 15. The element of interest might be iridium or any other element susceptible to photon activation. Since either the sample 11' can be moved from irradiation assembly to detector, or the sample can remain stationary and the irradiation and detector systems can be interchanged for sample analysis, the functional relationship between these three elements is indicated by the broken lines 97 and 98. The invention is not limited to photon activation reactions which result in the emission of isomeric gamma radiation. The invention can use any photon activation which results in measurable and identifiable decay radiation from the assay elements of interest. The detectors of the detection system must be selected to optimally detect the decay radiation from the activated assay elements. Count data from the detector assembly 15 are transferred to the computer 70 where assay concentrations are computed using previously discussed methods. The computer 70 outputs results 99 of the analysis. SUMMARY It can be seen that the present invention provides a technique that rapidly and accurately provides assays of large and non homogeneous ore samples with minimal sample preparation compared with methodology of the prior art. The technique has a similar sensitivity to gold as fire assay, less than 0.005 troy ounce per short ton, but can handle large sample weights (typically 10 kg) to give better average assay numbers. The technique makes it possible to process large volumes in a reasonable time (typically 600 samples within 24 hours). Other stated objects of the present invention are also met. While the above is a complete description of the preferred embodiment of the invention, alternative constructions, modifications, and equivalents may be used. For example, it is possible to avoid or reduce the need for neutron shielding around the gamma ray detector if the detector system is located a considerably greater distance from the irradiation system. Some gamma ray absorber is likely to be desirable to reduce naturally occurring background radiation. Furthermore, the irradiation and count scheme can be modified such that, as an example, segments of the sample are irradiated and oscillated while irradiating, but the entire sample is counted rather than counting irradiated segments. In addition, the invention can be configured to analyze any type of sample material for elements susceptible to photon activation analysis. Therefore, the above description and illustrations should not be seen as limiting the script of the invention which is defined by the appended claims.
	summary
055315453	claims	1. A mine roof cable bolt structure including, in combination: an elongated metal tubular member having a hollow interior, reaction structure provided said tubular member, a cable bolt provided a cylindrical member whereby to create a peripheral sectional enlargement, as to a portion of said cable bolt, which has a nominal transverse peripheral dimension greater than the corresponding transverse cross-section of said hollow interior of said tubular member, said sectional enlargement thereby cooperating with said tubular member within an interference fit, whereby to expand radially outwardly said tubular member in said tubular member's elastic range, for creating a pressure bubble, whereby to offer elastic resistance to movement of said metal tubular member, over said sectional enlargement. 2. The mine roof cable bolt structure of claim 1 wherein said cable bolt is provided with an interior axial king wire and multiple cable strands helically wound over said king wire whereby to provide said peripheral sectional enlargement, said cylindrical member being mounted on said king wire for expanding said strands in the region of said cylindrical member. 3. Mine roof cable bolt structure including, in combination: an elongated exteriorly threaded metal tubular member having remote and proximate ends, said tubular member having an internal bore provided with and tapering into a bore enlargement at said proximate end thereof, a cable bolt comprising a cable length having a king wire and multiple strands wrapped about said king wire, said cable bolt having a transverse enlargement, said enlargement cooperating in an installed press-fit within said bore enlargement, a reaction member disposed upon said tubular member, a torquing nut threaded upon said tubular member and backing said reaction member, said enlargement of said cable bolt being constructed to enter into said bore in an interference fit, whereby to expand elastically said tubular member as said enlargement proceeds in relative longitudinal movement within said bore, whereby to create a pressure bubble interference fit for resistively controlling relative movement between said enlargement and said tubular member, said enlargement being at least in part formed by an elongated cylindrical member of surface hardness of the order of that of said strands of said cable length. 4. Cable bolt apparatus for securement over an external, apertured mine roof bearing plate and within a strata borehole aligned with said bearing plate at its aperture, said cable bolt apparatus including, in combination: an elongated tubular member having an internal bore, external threads, and a peripherally enlarged, proximate bore area contiguous with said bore; a cable length comprising a king wire and helical strands wrapped thereabout, said cable length having a remote end constructed for securement within said borehole and a proximate, peripherally enlarged end nominally mounted in friction-fit relationship within said enlarged, proximate bore area; nut means threaded upon said tubular member and constructed for abutting against said external bearing plate and thereby preloading said cable length, said tubular member and said peripherally enlarged end of said cable length being mutually constructed whereby to provide, through elastic expansion of said tubular member and consequent radial compression thereof against said peripherally enlarged end, a pressure bubble offering controlled resistance to relative movement between said cable length and said tubular member in response to mine roof strata settling. 5. The cable bolt apparatus of claim 4 wherein said king wire is provided a cylindrical member, resistant to plastic deformation, disposed upon said king wire and forming with said strands said peripherally enlarged end. 6. The cable bolt apparatus of claim 4 wherein said king wire is provided with a series of end-to-end disposed cylindrical members, respectively resistant to plastic deformation, disposed upon said king wire and forming with said strands said peripherally enlarged end. 7. The cable bolt apparatus of claim 4 wherein a hardened cylindrical, tubular member having a slit side wall is provided said cable length, whereby to form said peripherally enlarged end. 8. A method of controlling the dilation of a mine roof, as produced through settling of strata thereabove, comprising the steps of: (1) providing a bore hole; (2) anchoring a cable bolt at its remote end within said borehole; (3) providing an elongated, cylindrical enlargement of said cable bolt at its proximate end; (4) providing an elongated, exteriorly threaded metal tubular member of radially elastic expansion characteristics, said metal tubular member having a hollow interior nominally less in transverse cross-section than that of said cylindrical enlargement, said metal tubular member thereby receiving said cable bolt at said cylindrical enlargement in a tubular-member-elastic-expansion interference fit; (5) providing for said tubular member a reaction plate and also a torquing nut, threaded upon said tubular member and backing said reaction plate;, (6) preloading said cable bolt through tightening said torquing nut against said reaction plate, and (7) creating a controlled, travel resistant pressure bubble as between said cable bolt and said tubular member, whereby to retard in a controlled resistive manner the descent of said tubular member relative to said cable bolt in response to dilation of said mine roof as occasioned through strata settling.
058754075	claims	1. A method for immobilizing waste chloride salts containing cesium for long term storage, comprising: mixing substantially dry, non-aqueous cesium chloride with chabazite; and heating the mixture to a temperature sufficient to form a pollucite. mixing non-aqueous cesium chloride with zeolite A; and heating the cesium chloride and zeolite to a temperature sufficient to form a product comprised of pollucite and sodalite. blending and heating zeolite with a waste chloride salt to trap the waste chloride salt ions within the zeolite structure to form a salt-loaded zeolite; adding chabazite; combining the salt-loaded zeolite and chabazite with glass; and heating and pressurizing the combination to form a ceramic composite. blending and heating zeolite with a waste chloride salt to trap the waste chloride salt ions within the zeolite structure to form a salt-loaded zeolite; heating the salt-loaded zeolite to a temperature sufficient to form sodalite; adding chabazite; combining the sodalite zeolite and chabazite with glass; and heating and pressurizing the combination to form a ceramic composite. 2. The method according to claim 1 wherein the step of heating the mixture includes heating the mixture to a temperature above the melting temperature of cesium chloride. 3. The method according to claim 1 wherein the step of heating the mixture includes heating the mixture to a temperature above about 700.degree. C. 4. The method according to claim 1 further comprising the steps of cooling the pollucite, contacting the pollucite with glass, and heating the glass and pollucite to a temperature sufficient to form a pollucite and glass product. 5. The method according to claim 4 wherein the step of heating the glass and pollucite includes heating the glass and pollucite to a temperature above about 700.degree. C. 6. The method according to claim 4 wherein the glass is borosilicate glass. 7. The method according to claim 4 wherein the glass is initially present as glass frit. 8. The method according to claim 4 wherein the step of heating the glass and pollucite includes hot isostatic pressing. 9. A method for immobilizing waste chloride salts containing cesium for long term storage, comprising: 10. The method according to claim 9 wherein the cesium chloride and zeolite are heated to a temperature of above about 700.degree. C. 11. A method for improving the retention of cesium in waste products, comprising: 12. The method according to claim 11, wherein the step of adding the chabazite includes adding 10% by weight chabazite. 13. The method according to claim 11, wherein the chabazite is added during the blending of the waste chloride salt and the zeolite. 14. The method according to claim 11, wherein the chabazite is added after the blending of the waste chloride salt and the zeolite, and prior to heating the zeolite and waste chloride salt. 15. The method according to claim 11, wherein the zeolite is zeolite A. 16. A method for improving the retention of cesium in waste products, comprising: 17. The method according to claim 16 wherein the step of adding the chabazite includes adding 10% by weight chabazite. 18. The method according to claim 16 wherein the zeolite is zeolite A.
	description	This application claims priority under 35 U.S.C. 119(e) of U.S. provisional application Ser. No. 62/686,748, filed on Jun. 19, 2018, the contents of which is hereby incorporated by reference in its entirety and for all purposes. The disclosure pertains to improvements in a gamma radiation source (including radiological and radiographic sources), typically containing low-density alloys or compounds or composites of iridium in mechanically deformable and compressible configurations, for use within a sealed encapsulation, and methods of manufacture thereof. Improvements in iridium sources have been described in PCT/US2017/033508 entitled “Low Density Spherical Iridium”; PCT/US2017/050425 entitled “Low Density Porous Iridium”; and PCT/US2015/029806 entitled “Device and Method for Enhanced Iridium Gamma Irradiation Sources.” The disclosures of these applications are well-suited to their intended purposes. However, further improvements and refinements are sought. It is therefore an object of this application to provide improvements and refinement with respect to the above-identified prior art. Objects of this disclosure include: 1. developing a deformable and/or compressible low density iridium alloy containing 30-85% (volume percentage) Iridium, preferably in the range of 30-70%, more preferably in the range of 40-60%. 2. the alloying constituents ideally or typically should not irradiate to produce other radionuclides that generate interfering gamma rays. 3. the alloying constituents ideally or typically should not have excessively high density or high neutron activation cross-section, which could decrease the activation yield or decrease the source-output yield of Iridium-192. 4. the alloying constituents ideally or typically should produce an alloy that is workable in that the alloy needs to be sufficiently ductile/deformable/compressible whereas pure iridium and most of its alloys are brittle and unworkable; the alloy ideally or typically should preferably have a lower melting point than pure iridium (a melting point less than 2000 degrees Centigrade would be desirable to lower processing costs and simplify thermal technologies); and the alloy ideally or typically should be substantially physicochemically inert (i.e., it does not oxidize/corrode/decompose under conditions of manufacture or use). The general class of compounds that are predicted to have suitable mechanical and density properties are called L21 Heusler structures. Specifically, these comprise Ir2M1N1, where M and N represent two different metals. Ir2MnAl is described above. Ir2CrAl Is a potential alternative. There may be others, e.g., Ir2Al and Ir2Al11B. With regard to the L21 Heusler compounds and structures, a range of compounds and structures should be taken into account. It is known that after irradiation of a L21 Heusler compound like Ir2MnAl, it would transmute to Ir2−(x+y)PtxOsyMnAl where “x+y” is the proportion of iridium that transmutes to platinum and osmium. There is typically approximately 10-20% conversion, depending on neutron flux, enrichment, irradiation time and decay lime (burn-up/transmutation) in an irradiation. Iridium-191 (37.3% in natural iridium, approximately 80% in enriched iridium) activates to Iridium-192 of which approximately 95% decays to Platinum-192 and 5% decays to Osmium-192 over the life of the source. Iridium-193 (62.7% in natural iridium, ˜20% in enriched iridium) activates to Iridium-194, which all decays to Platinum-194 in the reactor. In summary, an irradiated disk may contain roughly 5-20% platinum and 0.25-1% osmium after activation, depending on the flux, time and enrichment. It is the post-irradiated alloy that is desired to be ductile, deformable or compressible. The addition of platinum to iridium is likely to increase ductility. Even if pre-irradiated alloy disks do not have optimum mechanical properties for source manufacture, post-irradiated disks may. Quaternary alloys that contain small amounts of other ingredients, such as, but not limited to, platinum or osmium, or other purposeful additions included before irradiation (such as, but not limited to, chromium) may improve the physicochemical and mechanical properties without activating adversely. Ternary and quaternary alloys are synthesized to account for the conversion of 10-20 atom % of the Iridium to its daughters platinum and osmium in the nuclear reactor. Representative alloys in this regard include Ir1.8 Pt0.2MnAl and Ir1.6 Pt0.4MnAl, also including a very small percentage of osmium. A further representative alloy is Ir3Zr0.25 V0.75. Similarly, yttrium alloyed with iridium has increased ductility. Stable, natural 89Yttrium activates with very low cross section to form a very small amount of radioactive 90Yttrium, a pure beta emitter with a 64 hour half-life. It is therefore an acceptable metal to co-irradiate with Iridium. It does not produce long term interfering gamma rays. Moreover, 90Y decays to stable zirconium. Yttrium is therefore one of the preferred alloying additives. The most likely composition is IrY (i.e. 50/50-atomic percent alloy), but other ratios of IrxYy may also have increased ductility. Further representative alloys include IrY, Ir0.9 Pt0.1Y, and Ir0.8 Pt0.2Y. The density of Ir2MnAl is reported or calculated to be 13.89 g/cc vs. 22.56 g/cc for pure iridium (i.e., 61.5%). Further studies may confirm or refine this number. This is slightly higher than optimum for many applications, therefore this alloy may be used for porous or 3-D printed shapes that contain empty spaces, so that the net density may be reduced to the optimum range of 30-85% (preferably in the range of 30-70%, more preferably 40-60%), as illustrated in the various figures of this application. It is also expected that these compounds may have anti-ferromagnetic properties. These alloys may be formed by mixing powdered elements in molar proportions, e.g. Ir2+Mn+Al and heating—e.g. arc melting or using a high temperature vacuum furnace. As a variant of this basic method, it is expected, under some circumstances, to advantageously first pre-alloy Mn+Al and then mix/process this with pure iridium. MnAl melts at approximately 1500 degrees Centigrade. Other approaches may include pre-alloying iridium and aluminum and then adding Mn or Mn+Al later. The alloy composition Al7Ir3 (i.e. 30 mol % Iridium) is reported to have a eutectic at approximately 1930 degrees Kelvin (1657 degrees Centigrade). Reference is made to the article “Antiferromagnetism in γ-Phase Mn—Ir Alloys,” as reported in the Journal of the Physical Society of Japan in 1974, pages 445-450 (Online ISSN: 1347-4073, Print ISSN 0031-9015). This article indicates that antiferromagnetic disordered γ-phase Mn(1−x)Irx (0.05<x<0.35) alloys exists. Mixing an Ir+Mn alloy in this composition range, e.g. Mn7Ir11 powder or granules with Al7Ir3 powder or granules in equimolecular proportions followed by thermal processing (arc melting or furnace) is expected to produce an alloy with a composition of Ir14Mn7Al7 (═Ir2MnAl). In accordance with the above, the alloy Ir2MnAl forms an embodiment of the present disclosure of a gamma radiation source (including radiological and radiographic sources). It is believed to have ductile properties similar to steel. Additionally, manganese and aluminum are not expected to generate interfering gamma rays after activation by neutron irradiation. Platinum, osmium, chromium or mixtures thereof may likewise be found within the alloy. This alloy or similar alloys (such as with ternary additions of other non-activating elements) is expected have suitable mechanical properties to make deformable/compressible radiation sources. Although the addition of manganese slightly increases the density with respect to iridium plus aluminum or iridium plus aluminum plus Boron-11, it is expected that the metallurgical properties of Ir2MnAl may offer significant processing advantages. Iridium manganese copper alloys are also applicable to the present disclosure. These alloys are expected to be ductile and have a melting point significantly below 2000 degrees Centigrade and potentially as low as 1300 degrees Centigrade, depending upon the alloy composition after irradiation. These alloys are disclosed in U.S. Pat. No. 4,406,693 entitled “Method for Refining Contaminated Iridium,” issued on Sep. 27, 1983. However, it is expected that aluminum will be preferable over copper as a tertiary alloying element in most applications. Yttrium alloyed with indium is likewise a relevant material for the present disclosure. Furthermore, reduced density may be achieved in some embodiments by the use of porous, microporous or macroporous (i.e., metal foam) forms of the alloy of choice. All radiation sources are typically designed and expected to be inserted into a sealed encapsulation. Referring now to FIGS. 1 through 4, one sees illustrations of an embodiment of a deformable/compressible non-solid shape for a gamma radiation source 100 (which may be a radiological or radiographic source) that may be made using a deformable/compressible iridium alloy. Gamma radiation source 100 may be manufactured by 3-D printing but is not limited thereto. Further, gamma radiation source 100, as well as all embodiments disclosed herein, are implemented within a sealed encapsulation. Gamma radiation source 100 of FIGS. 1-4 includes a central ring or disk 102, along with an upper ring or disk 104 and a lower ring or disk 106 of somewhat reduced diameter. Rings 102, 104, 106 generally share a common rotational axis 108, as shown in FIGS. 2-4, and are generally parallel to each other in an uncompressed or uniformly compressed configuration. Upper ring 104 is positioned above the central ring 102 by arms 110, 112, 114 spiraling rotationally outwardly from an exterior circumferential surface of upper ring 104 to an interior circumferential surface of central ring 102. Similarly, lower ring 106 is positioned below the central ring 102 by arms 120, 122, 124 spiraling rotationally outwardly from an exterior circumferential surface of lower ring 106 to an interior circumferential surface of central ring 102. The elasticity and flexibility of spiraling arms 110, 112, 114, 120, 122, 124 allows for forces generally parallel with the common rotational axis 108 to compress the gamma radiation source 100 from the configuration shown in FIGS. 1 and 2 to the configuration shown in FIG. 4. Furthermore, in the compressed configuration of FIG. 4, gamma radiation source 100 is sealed within an encapsulation 117. Those skilled in the art will recognize that different shapes and configurations of encapsulation may be used for different applications, and that shapes different from that of the illustrated encapsulation may be used. FIGS. 5A and 5B illustrate an embodiment of gamma radiation source 100 wherein concentric co-planar rings 125, 126, 127, 128, 129 of deformable/compressible iridium alloy area positioned around a center 123, with radial structural spoke segments 131 extending from center 123 to innermost ring 125, and then between successively or sequentially concentrically adjacent rings, 125, 126; 126, 127; 127, 128; and 128, 129. FIG. 5B illustrates the elongated shape of the side view of gamma radiation source 100. The resulting configuration can be folded and/or compressed into different shapes to achieve an increased average density. This gamma radiation source 100 is made from a deformable/compressible iridium alloy, may be made by 3-D printing, and is sealed within an encapsulation (see FIG. 4, element 117). FIG. 6A-6E illustrate embodiments of the gamma radiation source 100 which include a central cylindrical shaft-type hub area 130 with a rotational axis 134 at the center and with propeller-type radial extensions 132 extending therefrom. Additionally. FIG. 6F includes an outer circular ring 136 joining the distal ends of the propeller-type radial extensions 132. These propeller type radial extensions 132, in the illustrated uncompressed slates, are oriented at an angle analogous to the pitch or blade angle of a conventional propeller. While different applications may use different angles, a typical pitch or propeller angle may be in the range of 30 to 60 degrees. However, as a result of forces of compression generally parallel to the rotational axis 134, the propeller angle of the propeller-type radial extensions 132 reduces so that the angle between the planar surface of the central cylindrical shaft-type hub area 130 and the propeller-type radial extensions 132 reduces so that the propeller-type radial extensions 132 approach a planar configuration with the central cylindrical shaft-type hub area 130. This decreases the volume which generally envelopes the gamma radiation source 100, thereby increasing the average density within the volume. These gamma radiation sources 100 are made from a deformable/compressible iridium alloy, may be made by 3-D printing, and are sealed within an encapsulation (see FIG. 4, element 117). FIGS. 7A-7F illustrate spiral configurations of the gamma radiation source 100 comprising a rod, tube or other extended configuration 200 of deformable iridium alloy, or similar material. Rod, tube or other extended configuration 200 includes a first end 202 and a second end 204. The spiral configuration places first end 202 at an interior location in the spiral and the second end 204 at an exterior location in the spiral. The spiral configuration, along with the deformable, and possibly elastic, property of the rod, tube or other extended configuration 200 allows the spiral to be tightened so as to occupy less volume, and therefore have a higher average density. In many applications, these shapes are adaptable to 3-D printing. FIG. 7G illustrates an embodiment of gamma radiation source 100 wherein a rod, tube or other extended configuration 200 of deformable iridium alloy or similar material is successively looped and placed at increasing radial locations, within each of four quadrants 230, 232, 234, 236. As shown in FIG. 7G, alternate loops may extend between two adjacent quadrants. The resulting structure can be stretched or compressed within the plane of gamma radiation source 100 or folded upon itself to alter the average density of the gamma radiation source 100. In many applications, this shape is adaptable to 3-D printing. FIG. 7H illustrates an embodiment of gamma radiation source 100 wherein a rod, tube or other extended configuration 200 of deformable iridium alloy or similar material is wrapped in a three-dimensional spiral shape so as to form a quasi-spherical shape in that the rod, tube or other extended configuration 200 covers a first portion of a quasi-spherical shape and a second portion of a quasi-spherical shape is left open, with ends 202, 204 generally at opposite poles of the quasi-spherical shape. The resulting three-dimensional spiral shape of the gamma radiation source 100 can be twisted or otherwise compressed into a configuration of increased average density. In many applications, this shape is adaptable to 3-D printing. FIGS. 8A-8G illustrate further embodiments of the gamma radiation source 100 of the present disclosure. FIG. 8A illustrates how a rod, tube or other extended configuration 200 of deformable iridium alloy or similar material may be wrapped or looped within a single plane. This gamma radiation source 100 may be twisted or compressed into a configuration of increased average density. In many applications, this shape is adapted to 3-D printing. FIG. 8B illustrates an embodiment of a gamma radiation source 100 similar to that of FIG. 7H. A rod, tube or other extended configuration 200 of deformable iridium alloy or similar material is wrapped in a three-dimensional spiral shape so as to form a quasi-ellipsoidal shape in that the rod, tube or other extended configuration 200 covers a first portion of a quasi-ellipsoidal shape and a second portion of a quasi-ellipsoidal shape is left open, with ends 202, 204 generally at opposite poles of the quasi-ellipsoidal shape. The resulting three-dimensional spiral shape of the gamma radiation source 100 can be twisted or otherwise compressed into a configuration of increased average density. In many applications, this shape is adaptable to 3-D printing. FIG. 8C illustrates an embodiment of the gamma radiation source 100 wherein a ribbon-like configuration 300 of deformable iridium alloy or similar material is wrapped in a three-dimensional projectile or nosecone-type shape. This shape may be pushed downward to form a tightly wrapped spiral configuration of increased average density. In many applications, this shape is adaptable to 3-D printing. FIG. 8D illustrates an embodiment of gamma radiation source 100 similar to that of FIGS. 1-4. In FIG. 8D, a relatively larger ring 102 is provided with, along with a relatively smaller ring 104 in an upward position. Rings 102, 104 generally share a common rotational axis 108. Ring 104 is positioned above the ring 102 by arms 110, 112, 114 spiraling outwardly from an exterior circumferential surface of ring 104 to an interior circumferential surface of ring 102. The elasticity and flexibility of arms 110, 112, 114 allows for forces generally parallel with the rotational axis to compress the gamma radiation source 100. In many applications, this shape is adaptable to 3-D printing. FIG. 8E illustrates an embodiment of gamma radiation source 100 which includes a series of interlocking sleeves 401-409 which are slidably engaged with inwardly or outwardly adjacent interlock sleeves. Interlocking sleeves 401-409, which are formed of a deformable iridium alloy or similar material may also be implemented as a spiral configuration of a single sheet of material. A spiral wire configuration 410 of similar material is engaged within an inner diameter of interlocking sleeve 409. This gamma radiation source 100 can be compressed to a reduced volume, thereby resulting in higher average density. In many applications, this shape is adaptable to 3-D printing. FIG. 8F illustrates an embodiment of gamma radiation source 100 which is somewhat similar to that of FIGS. 7H and 8B in that a rod, tube or other extended configuration 200 of deformable iridium alloy or similar material is wrapped in a three-dimensional spiral shape so as to form a quasi-conical shape (with an open circular base) in that the rod, tube or other extended configuration 200 covers a first portion of the walls of a quasi-conical shape and a second portion of the walls of the quasi-conical shape is left open. The resulting three-dimensional spiral quasi-conical shape of the gamma radiation source 100 can be twisted or otherwise compressed into a configuration of increased average density. In many applications, this shape is adaptable to 3-D printing. FIG. 8G illustrates an embodiment of gamma radiation source 100 wherein two adjacent disks 420, 422 each include first and second rods, tubes or other extended configurations 201, 203 of deformable iridium alloy or similar material are wrapped in a concentric spiral pattern. In the illustrated configuration, the first and second rods 201, 203 are wrapped in a clockwise configuration in first disk 420 and counterclockwise in second disk 422. The disks 420, 422 may be varied in relationship to each other, folded or otherwise compressed to vary the average density thereof. In many applications, this shape is adaptable to 3-D printing. Thus the several aforementioned objects and advantages are most effectively attained. Although preferred embodiments of the invention have been disclosed and described in detail herein, it should be understood that this invention is in no sense limited thereby.
	claims	1. A method for measuring the drop time of at least one control or shutdown rod in a nuclear reactor, said method comprising the steps of: (a) releasing a rod from a resting position above a nuclear reactor core; (b) guiding the rod through a plurality of energized coils; (c) recording a signal present at the plurality of energized coils, said signal being recorded from said step of releasing a rod from a resting position above a nuclear reactor core until the rod comes to a resting position within the reactor core, said signal including both a coil power measurement and a rod drop trace, said coil power measurement including information about the instantaneous power of the plurality of energized coils, said rod drop trace including information about the electrical signal induced in the plurality of energized coils by the rod; and (d) filtering said signal to isolate the rod drop trace from coil power measurement to identify the drop time. 2. The method of claim 1 further comprising the step of calculating a rod drop time from the rod drop trace. claim 1 3. The method of claim 1 further comprising the step of applying a filter to isolate the rod drop trace from the coil power measurement to identify the drop time. claim 1 4. The method of claim 1 wherein the step of releasing the rod from a resting position above a nuclear reactor core further includes the step of releasing a plurality of rods simultaneously from a resting position above a nuclear reactor core. claim 1 5. The method of claim 4 further comprising the step of overlaying one rod drop trace obtained from one of the plurality of rods with another rod drop trace obtained from another of the plurality of rods. claim 4 6. The method of claim 1 further comprising the step of storing at least the rod drop trace obtained during said step of filtering said signal to isolate the rod drop trace from the coil the power measurement to identify the drop time. claim 1 7. The method of claim 6 wherein said step of storing at least the rod drop trace obtained during said step of filtering said signal to isolate the rod drop trace from coil power measurement to identify the drop time includes the step of storing said signal obtained during said step of filtering said signal to isolate the rod drop trace from the coil power measurement to identify the drop time. claim 6 8. The method of claim 6 wherein said step of storing at least the rod drop trace obtained during said step of filtering said signal to isolate the rod drop trace from coil power measurement to identify the drop time includes the step of storing the coil power measurement obtained during said step of filtering said signal to isolate the rod drop trace from the coil power measurement to identify the drop time. claim 6 9. The method of claim 6 further comprising the step of overlaying one rod drop trace with another rod drop trace from said step of storing at least the rod drop trace obtained during said step of filtering said signal to isolate the rod drop trace from the coil power measurement to identify the drop time. claim 6 10. A method for measuring a drop time of control and shutdown rods in a nuclear reactor having a coil-based rod position indication system, said method comprising the steps of: (a) releasing a rod from a resting position above a nuclear reactor core; (b) sensing a coil current from a coil-based rod position indication system; (c) separating an induced current representing a rod drop position trace from said coil current; and (d) calculating a rod drop time from said induced current. 11. A method for measuring a rod drop time of a rod in a nuclear reactor core having a coil-based rod position indication system, said method comprising the steps of: (a) sensing a coil current from a coil-based rod position indication system; (b) separating an induced current from said coil current; (c) determining a start time of a rod drop event from said induced current; (d) determining a first stop time of the rod drop event from said induced current, said first stop time representing a time at which the control rod initially reaches a fully inserted position within a reactor core; and (e) determining a second stop time of the rod drop event from said induced current, said first stop time representing a time at which the control rod comes to a resting position within the reactor core. 12. A method for measuring a drop time of a rod in a nuclear reactor core having a coil-based rod position indication system, said method comprising the steps of: (a) sensing a coil current from a coil-based rod position indication system; (b) separating an induced current representing a rod drop position trace from said coil current; (c) determining from said induced current a first time period representing a period of time for the rod to travel from a top position in a reactor core to a bottom position in the reactor core; and (d) determining from said induced current a second time period representing a period of time for the rod to come to rest after reaching the bottom position in the reactor core.
	claims	1. A method of observing a specimen in a field of view of an electron microscope comprising the acts of:setting the magnification of said electron microscope;setting conditions for moving said field of view;setting a starting position for said field of view;moving said field of view based upon said condition;illuminating said specimen with an electron beam having a first angle and forming a first transmission image of said specimen in said field of view;adjusting said electron beam to a second angle and forming a second transmission image of said specimen in said field of view; andcalculating a degree of coincidence between said first and second transmission images.
	claims	1. An apparatus including multiple ion guns and multiple associated ion optical columns for focused ion beam processing of materials or imaging, comprising: a sealable ion gun chamber having positioned therein multiple ion guns, each ion gun capable of generating an ion beam; multiple ion optical columns, each ion optical column being associated with one of the multiple ion guns for focusing and directing the corresponding ion beam toward a target; a primary vacuum chamber for containing a target for processing or imagine; and a vacuum valve associated with each of the ion guns, the vacuum valves selectively opening to allow the corresponding ion beam to pass from the associated ion gun to the target or selectively closing to seal the corresponding ion gun chamber. 2. An apparatus including multiple ion guns and multiple associated ion optical columns for focused ion beam processing of materials or imaging, comprising multiple sealable ion gun chambers each including one or more ion guns, each ion gun capable of generating an ion beam; multiple ion optical columns, each ion optical column being associated with one of the multiple ion guns for focusing and directing the corresponding ion beam toward a target; a primary vacuum chamber for containing a target for processing or imaging a vacuum valve associated with each of the ion guns, the vacuum valves selectively opening to allow the corresponding ion beam to pass from the associated ion gun to the target or selectively closing to seal the corresponding ion gun chamber. 3. The apparatus of claims 2 in which each sealable gun chamber includes a vacuum pump. 4. The apparatus of claim 2 in which the vacuum valves associated with the ion guns in each of multiple gun chambers are connected so that the vacuum valves in each gun chamber open and close using a single control. claim 2 5. The apparatus of claim 1 in which each of the ion optical columns includes a deceleration lens element maintained near ground potential. claim 1 6. The apparatus of claim 1 , in which each of the ion optical columns includes optical elements and in which corresponding ones of at least one of the optical elements in different ones of the ion optical columns within the sealable gun chamber comprise an optical element bar to provide a common voltage to corresponding optical elements within the gun chamber. claim 1 7. The apparatus of claim 6 , in which electrically isolated lens elements are placed in the optical element bar to allow independent control of some of the optical elements comprising the optical element bar. claim 6 8. The apparatus of claim 1 in which the ion optical columns includes multiple lens elements and further comprising means for collecting secondary particles through at least one of the lens elements for imaging or characterizing the target. claim 1 9. The apparatus of claim 8 , in which at least one of the ion optical column further includes a deflector for deflecting secondary particles out of the path of the ion beam and in which the ion optical column includes at least one lens element between the deflector and the target, the lens element electrically biased relative to the target to create an electrical field to accelerate the charged secondary particles up through and past the lens element for detection. claim 8 10. The apparatus of claim 8 , in which the target is biased to the same polarity as that of the charge on the secondary charged particles to accelerate the charged particles up through and past the lens and electrostatic deflector for detection. claim 8 11. The apparatus of claim 8 , further comprising means for detecting charged secondary particles where the detector of charged particles is a channel plate multiplier or scintillator detector placed substantially perpendicular to the primary beam with a center hole for the primary beam to pass through. claim 8 12. The apparatus of claim 8 , further comprising a magnetic deflector, a Wien filter or an electrostatic deflection device for deflecting the secondary particles away from the ion beam path for collection. claim 8 13. The apparatus in claim 8 , further comprising a mass spectrometer for Secondary Ion Mass Spectrometry for detecting and characterizing the secondary charged particles. claim 8 14. The apparatus of claim 1 in which at least some of the ion guns and ion optical columns are tilted at an angle of about three degrees to a normal to the sample surface. claim 1 15. The apparatus of claim 14 in which the ion guns in a first one of the multiple ion beam gun chamber are tilted at an angle of about three degrees in a first direction from a normal to the sample surface and in which the ion guns in a second one of the multiple ion beam gun chamber are tilted at an angle of about three degrees from a normal to the sample surface in a direction opposite to the first direction. claim 14 16. The apparatus of claim 1 in which each of the ion optical columns includes beam offset, scanning, steering and stigmation controls and in which the beam offset, scanning, steering and stigmation can be controlled independently for each column. claim 1 17. The apparatus of claim 1 further comprising a high voltage supply for providing a high voltage to corresponding optical elements in multiple ones of the ion optical columns. claim 1 18. The apparatus of claim 17 further comprising means for adjusting the voltage in one of the ion optical columns to deviate from the high voltage provided by the high voltage power supply. claim 17 19. A multiple column focused ion beam system comprising: multiple ion guns for forming multiple ion beams; a bar having holes for forming therein multiple ion optical lenses, each ion optical lens corresponding to one of the multiple ion beam sources, each ion beam source and ion optical lens forming pan of an ion beam optical column; and a power supply for applying a voltage to lenses corresponding to the bar, thereby applying a common voltage to ion optical lenses in different optical columns. 20. The apparatus of claim 19 in which the bar comprises a flat conductive bar and in which the power supply provides a voltage directly to the conductive bar, the holes in the conductive bar functioning as ion optical lenses. claim 19 21. The apparatus of claim 19 in which the bar comprises a flat conductive bar having electrically isolated lenses formed therein and in which the power supply provides a common voltage to all lenses in the bar and selectively provides a second voltage to individual lenses in the bar. claim 19 22. The apparatus of claim 19 in which the bar comprises a flat non-conductive bar having electrically isolated lenses formed therein and in which the power supply provides a common voltage to all lenses in the bar and selectively provides second voltages to individual lenses in the bar. claim 19 23. The apparatus of claim 19 further comprising means for collecting through the lenses secondary particles emitted from the target, the secondary particles being used to image or to characterize the target. claim 19 24. The apparatus in claim 22 in which individual emitters are restarted by biasing either the extractor with respect to the emitter/suppressor elements about xe2x88x922000 V, or by biasing the emitter/suppressor elements with respect to the extractor element about 2000 V, in the individual guns as needed. claim 22 25. The apparatus of claim 1 wherein at least one said ion beam is generated employing a liquid metal ion source. claim 1 26. The apparatus of claim 8 , in which said multiple lens elements includes an electrostatic final lens for focusing the ion beam onto the target and further comprising a means for collecting secondary particles emitted from the target and traveling through said electrostatic lens for imaging or characterizing the target. claim 8 27. The apparatus of claim 2 , in which the ion optical column includes multiple lens elements and further comprising means for collecting secondary particles through at least one of the lens elements for imaging or characterizing the target. claim 2 28. The apparatus of claim 2 in which each of the ion optical columns includes beam offset, scanning, steering and stigmation controls and in which the beam offset, scanning, steering and stigmation can be controlled independently for each column. claim 2 29. The apparatus of claim 2 further comprising a high voltage supply for providing a high voltage to corresponding optical elements in multiple ones of the ion optical columns. claim 2 30. The apparatus of claim 2 wherein at least one said ion beam is generated employing a liquid metal ion source. claim 2
043476210	abstract	The invention pertains to the method and apparatus for the confining of a stream of fusible positive ions at values of density and high average kinetic energy, primarily of tightly looping motions, to produce nuclear fusion at a useful rate; more or less intimately mixed with the fusible ions will be lower-energy electrons at about equal density, introduced solely for the purpose of neutralizing the positive space charge of the ions. Ions under high kinetic energy are introduced into an annular reaction chamber having a primarily axial strong magnetic field and an essentially radial electric field and assume in the chamber a quasi-trochoidal motion in which the kinetic energies in their small diameter looping components of motion are greater by at least an order of magnitude, than the kinetic energies in the relatively slow crossed field advance motions with which the ions circulate circumferentially around the axis of the annular reaction chamber.
	description	The field of the application relates to treatment systems, and more particularly, to systems and methods for energy modulated radiation therapy. Radiation therapy involves medical procedures that selectively deliver high doses of radiation to certain areas inside a human body. A radiation machine for providing radiation therapy includes an electron source that provides electrons, and an accelerator that accelerates the electrons to form an electron beam. The electron beam is delivered downstream where it strikes a target to generate radiation. The radiation is then collimated to provide a radiation beam having a certain desired characteristic for treatment purpose. In other cases, instead of radiation, a particle beam (e.g., electron beam, proton beam, etc.) may be used as treatment energy to treat the patient. Systems and methods for energy modulated radiation therapy is described herein. A medical apparatus includes: a beam deflector having an electromagnet configured to provide a first magnetic field for deflecting a particle beam; and a current control configured to adjust a current of the electromagnet in correspondence with an energy level associated with an accelerator. Optionally, the apparatus further includes a permanent magnet configured to provide a second magnetic field. Optionally, the permanent magnet comprises a rare earth magnet. Optionally, the electromagnet comprises a coil surrounding the permanent magnet. Optionally, the apparatus further includes an accelerator for providing the particle beam, the accelerator having an energy switch configured to change an energy level of the particle beam. Optionally, the energy switch is configured to change the energy level of the particle beam within a duration that is less than one second. Optionally, the accelerator comprises a fixed-field alternating gradient (FFAG) accelerator, or a non-scaling fixed-field alternating gradient (NS-FFAG) accelerator. Optionally, the apparatus further includes an ion chamber, wherein the ion chamber comprises dosimetry circuit. Optionally, the control is also configured to adjust a parameter in the dosimetry circuit in correspondence with the energy level associated with an accelerator. Optionally, the electromagnet comprises a laminated steel. Optionally, the current control is configured to increase the current of the electromagnet in correspondence with an increase in the energy level associated with the accelerator. Optionally, the current control is configured to decrease the current of the electromagnet in correspondence with a decrease in the energy level associated with the accelerator. Optionally, the electromagnet comprises a pretzel magnet. Optionally, the apparatus further includes a beam output coupled to the beam deflector, wherein the beam output is moveable to deliver treatment energy from a plurality of gantry angles that includes at least a first gantry angle and a second gantry angle. Optionally, the apparatus further includes an energy adjuster configured to adjust the treatment energy so that the treatment energy has a first energy level when the beam output is at the first gantry angle, and a second energy level when the beam output is at the second gantry angle. Optionally, the energy adjuster is configured to adjust the treatment energy in a continuous manner. Optionally, the energy adjuster is configured to adjust the treatment energy in a discrete manner. Optionally, the apparatus further includes an energy adjuster configured to adjust the treatment energy so that the treatment energy has a first energy level when the beam output is at the first gantry angle, and a second energy level when the beam output is at the first gantry angle. Optionally, the beam output is configured to deliver the treatment energy without using any flattening filter. A medical apparatus includes: an accelerator configured to provide a particle beam; an energy switch configured to change an energy level of the particle beam within a duration that is less than one second; and a target configured to receive the particle beam; wherein the accelerator is configured to provide the particle beam directly onto the target without using a beam deflector at an end of the accelerator. Optionally, the apparatus further includes a beam output coupled to the accelerator, wherein the beam output is moveable to deliver treatment energy from a plurality of gantry angles that includes at least a first gantry angle and a second gantry angle. Optionally, the apparatus further includes an energy adjuster configured to adjust the treatment energy so that the treatment energy has a first energy level when the beam output is at the first gantry angle, and a second energy level when the beam output is at the second gantry angle. Optionally, the energy adjuster is configured to adjust the treatment energy in a continuous manner. Optionally, the energy adjuster is configured to adjust the treatment energy in a discrete manner. Optionally, the beam output is configured to deliver the treatment energy without using any flattening filter. Optionally, the accelerator comprises a fixed-field alternating gradient (FFAG) accelerator, or a non-scaling fixed-field alternating gradient (NS-FFAG) accelerator. Optionally, the apparatus further includes an ion chamber, wherein the ion chamber comprises dosimetry circuit. Optionally, the apparatus further includes a control configured to adjust a parameter in the dosimetry circuit in correspondence with the energy level of the particle beam. A treatment method includes: configuring a medical system for delivering a first treatment beam having a first energy level; delivering the first treatment beam by the medical system towards a patient that is on a patient support; configuring the medical system for delivering a second treatment beam having a second energy level; and delivering the second treatment beam by the medical system towards the patient; wherein the act of configuring the medical system for delivering the second treatment beam comprises changing an energy that is associated with an accelerator by an energy switch, and adjusting a current of an electromagnet by a current control in correspondence with the energy associated with the accelerator. Optionally, the first treatment beam is delivered towards the patient from a first gantry angle, and the second treatment beam is delivered towards the patient from a second gantry angle. Optionally, the first treatment beam is delivered towards the patient from a first gantry angle, and the second treatment beam is delivered towards the patient from the first gantry angle. Optionally, the medical system comprises an ion chamber, wherein the ion chamber comprises dosimetry circuit, and wherein the method further comprises adjusting a parameter in the dosimetry circuit in correspondence with the energy associated with the accelerator. Optionally, the acts of configuring the medical system, and the acts of delivering are performed so that the first treatment beam transitions to the second treatment beam in a continuous manner. Optionally, the acts of configuring the medical system, and the acts of delivering are performed so that the first treatment beam and the second treatment beam are delivered in discrete manner. A treatment planning method includes: defining control points in a treatment plan; setting up energy switching in one or more of the control points; and performing treatment optimization on the treatment plan based at least in part on the energy switching that is set up in the one or more of the control points. Optionally, the act of setting up energy switching comprises prescribing the energy switching between two gantry positions. Optionally, the act of setting up energy switching comprises prescribing the energy switching at a same gantry position. Optionally, the energy switching is set up so that the treatment plan will include a delivery of a first energy beam for treating a patient, and a delivery of a second energy beam for treating the patient. Optionally, the first energy beam comprises photons, electrons, protons, or a combination of two or more of the foregoing. A treatment planning method includes: defining control points for a treatment plan; receiving an input indicating that energy switching is desirable; and performing treatment optimization on the treatment plan based at least in part on the control points and the received input, wherein the treatment optimization is performed to determine whether to provide the energy switching for one or more of the control points. Other and further aspects and features will be evident from reading the following detailed description. Various embodiments are described hereinafter with reference to the figures. It should be noted that the figures are not drawn to scale and that elements of similar structures or functions are represented by like reference numerals throughout the figures. It should also be noted that the figures are only intended to facilitate the description of the embodiments. They are not intended as an exhaustive description of the invention or as a limitation on the scope of the invention. In addition, an illustrated embodiment needs not have all the aspects or advantages shown. An aspect or an advantage described in conjunction with a particular embodiment is not necessarily limited to that embodiment and can be practiced in any other embodiments even if not so illustrated, or if not so explicitly described. FIG. 1 illustrates a radiation treatment system 10. The system 10 includes an arm gantry 12, a patient support 14 for supporting a patient 20, and a control system 18 for controlling an operation of the gantry 12 and delivery of radiation. The system 10 also includes a radiation source 22 that projects a beam 26 of radiation towards the patient 20 while the patient 20 is supported on support 14, and a collimator system 24 for changing a cross sectional shape of the radiation beam 26. The radiation source 22 may be configured to generate a cone beam, a fan beam, or other types of radiation beams in different embodiments. Also, in other embodiments, the source 22 may be configured to generate a proton beam, electron beam, or neutron beam, as a form of radiation for treatment purpose. Also, in other embodiments, the system 10 may have other form and/or configuration. For example, in other embodiments, instead of an arm gantry 12, the system 10 may have a ring gantry 12. In the illustrated embodiments, the radiation source 22 is a treatment radiation source for providing treatment energy. In other embodiments, in addition to being a treatment radiation source, the radiation source 22 can also be a diagnostic radiation source for providing diagnostic energy for imaging purpose. In such cases, the system 10 will include an imager, such as the imager 80, located at an operative position relative to the source 22 (e.g., under the support 14). In further embodiments, the radiation source 22 may be a treatment radiation source for providing treatment energy, wherein the treatment energy may be used to obtain images. In such cases, in order to obtain imaging using treatment energies, the imager 80 is configured to generate images in response to radiation having treatment energies (e.g., MV imager). In some embodiments, the treatment energy is generally those energies of 160 kilo-electron-volts (keV) or greater, and more typically 1 mega-electron-volts (MeV) or greater, and diagnostic energy is generally those energies below the high energy range, and more typically below 160 keV. In other embodiments, the treatment energy and the diagnostic energy can have other energy levels, and refer to energies that are used for treatment and diagnostic purposes, respectively. In some embodiments, the radiation source 22 is able to generate X-ray radiation at a plurality of photon energy levels within a range anywhere between approximately 10 keV and approximately 20 MeV. Alternatively, the source 22 may not be a radiation source, and may instead be a particle source configured to provide a particle beam (e.g., electron beam, proton beam, etc.) as treatment beam. In other embodiments, the source 22 may be configured to provide a combination of photon beam and particle beam (e.g., electron beam, proton beam, etc.) for treatment. In further embodiments, the radiation source 22 can be a diagnostic radiation source. In such cases, the system 10 may be a diagnostic system with one or more moving parts. In the illustrated embodiments, the radiation source 22 is carried by the arm gantry 12. Alternatively, the radiation source 22 may be located within a bore (e.g., coupled to a ring gantry). In the illustrated embodiments, the control system 18 includes a processing unit 54, such as a processor, coupled to a control 40. The control system 18 may also include a monitor 56 for displaying data and an input device 58, such as a keyboard or a mouse, for inputting data. The operation of the radiation source 22 and the gantry 12 are controlled by the control 40, which provides power and timing signals to the radiation source 22, and controls a rotational speed and position of the gantry 12, based on signals received from the processing unit 54. Although the control 40 is shown as a separate component from the gantry 12 and the processing unit 54, in alternative embodiments, the control 40 can be a part of the gantry 12 or the processing unit 54. In some embodiments, the system 10 may be a treatment system configured to deliver treatment radiation beam towards the patient 20 at different gantry angles. During a treatment procedure, the source 22 rotates around the patient 20 and delivers treatment radiation beam from different gantry angles towards the patient 20. While the source 22 is at different gantry angles, the collimator 24 is operated to change the shape of the beam to correspond with a shape of the target tissue structure. For example, the collimator 24 may be operated so that the shape of the beam is similar to a cross sectional shape of the target tissue structure. In another example, the collimator 24 may be operated so that different portions of the target tissue structure receive different amount of radiation (as in an IMRT procedure). FIG. 2A illustrates some of the components of the radiation treatment system 10. As shown in the figure, the radiation treatment system 10 includes an electron source 200 for generating and providing electrons. The radiation treatment system 10 also includes a particle accelerator 202 having a first end 204 and a second end 206. The first end 204 of the accelerator 202 is coupled to the electron source 200 for receiving the electrons. The accelerator 202 is configured to accelerate the electrons to form an electron beam 208. The radiation treatment system 10 further includes a beam deflector 210 configured to turn the direction of the electron beam 208. As shown in the figure, the beam deflector 210 includes an electromagnet 212 and a permanent magnet 214. In other embodiments, the beam deflector 210 may include only the electromagnet 212 without the permanent magnet 214. In the illustrated example, the radiation treatment system 10 also includes a target 220 for receiving the electron beam 208. The electrons in the electron beam 208 strike the target 220 to generate radiation. The radiation goes through an ion chamber 230 configured to measure a dose associated with the radiation. The radiation is then collimated by a collimator 240 into a desired radiation beam 242. In the illustrated embodiments, the radiation treatment system 10 is configured to provide variable treatment energies for treating a patient. The radiation treatment system 10 also includes an energy switch 256 coupled to the accelerator 202 for changing an energy (e.g., an energy level) associated with the accelerator 202. The radiation treatment system 10 also includes a current control 258 for adjusting a current for the electromagnet 212, and a control 260 for adjusting a parameter in a dosimetry circuit 232 of the ion chamber 230 in correspondence with the energy associated with the accelerator 202. In the illustrated figure, the current control 258 and the control 260 are shown as separate components from the control 40. In other embodiments, the current control 258 and/or the control 260 may be integrated with the control 40 of the radiation treatment system 10. During use, the electron source 200 generates electrons and feeds the electrons to the accelerator 202. The accelerator 202 accelerates the electrons to form an electron beam 208 that travels down the longitudinal axis of the accelerator 202. The accelerator 202 is excited by a power, e.g., microwave power, delivered by a Magnetron (not shown) at a frequency, for example, between 1000 MHz and 20 GHz, and more typically, between 2800 and 3000 MHz. In other embodiments, the Magnetron can have other configurations and/or may be configured to provide power at other frequencies. The power delivered by the Magnetron may be in a form of electromagnetic waves. In other embodiments, instead of Magnetron, a klystron may be used. The electrons generated by the electron source 200 are accelerated through the accelerator 202 by oscillations of the electromagnetic fields within cavities of the accelerator 202, thereby resulting in the electron beam 208 having certain level of energy. The electron beam 208 exits the second end 206 of the accelerator 202 and is deflected by the beam deflector 210 at the second end 206 of the accelerator 202. In the illustrated embodiments, the control 40 controls the energy switch 256 so that the electron beam 208 exiting the accelerator 202 will have a first energy level. The current control 258 is configured to adjust a current for the electromagnet 212 in the beam deflector 210 in correspondence with an energy (e.g., energy level) associated with the accelerator 202. In some embodiments, the energy associated with the accelerator 202 may be an energy level of an electric field in the accelerator 202 caused or to be caused by an operation of the energy switch 256. In other embodiments, the energy associated with the accelerator 202 may be the energy level of the electron beam 208 generated or to be generated by the accelerator 202. As the energy of the electron beam 208 is increased, more magnetic field energy is provided by the electromagnet 212 in order to change the direction of the electron beam 208 by a certain pre-determined angle. Similarly, as the energy of the electron beam 208 is decreased, less magnetic field energy is provided by the electromagnet 212 to change the direction of the electron beam 208 by the same pre-determined angle. In the illustrated example, the electron beam 208 is deflected by 270° using the beam deflector 210. In other embodiments, the beam deflector 210 may be configured to deflect the electron beam 208 by other range of angles. In the illustrated example, the electromagnet 212 is surrounding the permanent magnet 214. The electromagnet 212 is configured to provide a first magnetic field MF1, and the permanent magnet 214 is configured to provide a second magnetic field MF2. The combined magnetic field MFT will thus be MF1+MF2. In some cases, the magnetic field MF2 provided by the permanent magnet 214 corresponds with the lowest energy level of the radiation beam that is to be generated. For example, if the system 10 is configured to provide a radiation beam having an energy level that ranges from 2 MeV to 20 MeV, then the permanent magnet 214 may be configured to provide the magnetic field MF2 that corresponds with the energy level of 2 MeV for the radiation beam. In such cases, the electromagnet 212 is configured to provide the magnetic field that ranges from 0 to MF1 (depending on the amount of current provided for the coils of the electromagnet 212), wherein the combined magnetic field of MF2+0 corresponds with the beam energy level of 2 MeV, and the combined magnetic field of MF2+MF1 corresponds with the beam energy level of 20 MeV. Accordingly, the combination of the electromagnet 212 and the permanent magnet 214 provides an unique combined magnetic field for each of the available beam energy levels. In other embodiments, the magnetic field MF2 provided by the permanent magnet 214 corresponds with a beam energy level that is lower than the lowest energy level of the radiation beam to be generated. For example, if the system 10 is configured to provide a radiation beam having an energy level that ranges from 2 MeV to 20 MeV, then the permanent magnet 214 may be configured to provide the magnetic field MF2 that corresponds with a beam energy level of 1 MeV. In such cases, the electromagnet 212 is configured to provide the magnetic field that ranges from 0 to MF1 (depending on the amount of current provided for the coils of the electromagnet 212), wherein the combined magnetic field of MF2+0 corresponds with the beam energy level of 1 MeV, and the combined magnetic field of MF2+MF1 corresponds with the beam energy level of 20 MeV. Also, in other embodiments, the electromagnet 212 may not surround the permanent magnet 214. Instead, the electromagnet 212 may be placed next to the permanent magnet 214 in a side-by-side configuration. In some embodiments, the permanent magnet 214 may be implemented using a rare earth permanent magnet. By using a rare earth magnet that is precisely machined and positioned, the amount of electromagnet needed to generate a desired magnetic field may be reduced. Furthermore, in other embodiments, instead of one electromagnet 212 and/or one permanent magnet 214, the beam deflector 210 may include multiple electromagnets and/or multiple permanent magnets. It should be noted that the combination of the electromagnet 212 and the permanent magnet 214 is advantageous because it allows a range of magnetic fields to be generated, while reducing the size of the beam deflector 210 (compared to the scenario in which just electromagnet is used). This in turn, provides a compact head (i.e., the structure upstream from the collimator 240) allowing more clearance to the patient and/or room walls. The hybrid bend magnet is also cheaper to construct and maintain due to smaller electromagnet with fewer copper windings (compared to the scenario in which just electromagnet is used), and due to lower amount of electrical current and cooling involved in the operation of the hybrid bend magnet. Furthermore, the hybrid bend magnet has the benefit of reducing magnetic hysteresis, which allows faster switching of energy of the electron beam. In some embodiments, the bend magnet's energy hysteresis cycle may be speed up by using laminated steel. Other materials may be used for the bend magnet in other embodiments. After the electron beam 208 is deflected by the beam deflector 210, the electron beam 208 strikes the target 220 to generate radiation. Alternatively, the electron beam 208 may not strike any target, and instead be used directly as treatment energy. Also, in other embodiments, instead of electron beam, the accelerator may provide other particle beams, such as proton beam, for treatment. The radiation or beam goes through the ion chamber 230 that measures a dose associated with the radiation or the beam. The dosimetry circuit 232 of the ion chamber 230 is configured to calculate a dose based on the radiation or energy received and one or more parameters (e.g., a dose calculation parameter). In the illustrated embodiments, the control 260 is configured to adjust a parameter (e.g., a dose calculation parameter) at the dosimetry circuit 232 in correspondence with the energy associated with the accelerator 202. As similarly discussed, the energy associated with the accelerator 202 may be an energy level of an electric field in the accelerator 202 caused or to be caused by an operation of the energy switch 256, or an energy level of the electron beam 208 that is generated or to be generated. The control 260 is advantageous because as the energy level of the electron beam 208 changes, the relation between the radiation energy and dose may not be constant. The control 260 allows the dosimetry circuit 232 of the ion chamber 230 to be updated so that ionization from changing energies is handled correctly. In particular, in order to correctly calculate a dose value associated with the generated radiation, the dose calculation parameter in the dosimetry circuit 232 is adjusted by the control 260. In one implementation, the relation between the different energy levels associated with the accelerator 202 and dose values may be determined empirically, and the resulting relation may then be programmed into the control 260 so that the control 260 can determine how much to adjust the dose calculation parameter based on the energy level associated with the accelerator 202. In other embodiments, the relation may be determined mathematically. After the radiation goes through the ion chamber 230, the radiation is then collimated by the collimator 240 into a desired radiation beam 242. In some embodiments, the collimator 240 may be a multi-leaves collimator having a plurality of moveable fingers configured to shape a cross section of the radiation beam 242. In other embodiments, the collimator 240 may be a block collimator, or any other types of collimator. In some embodiments, the control 40 is configured to operate the energy switch 256, the current control 258, and the control 260 in correspondence with each other so that a radiation beam 242 with a desired characteristic is provided at a beam output 270 as defined by the collimator 240. Also, in some embodiments, the operation of the energy switch 256, the current control 258, and the control 260 may be performed in synchronization with each other. In other embodiments, instead of the control 40 controlling the energy switch 256, the current control 258, and the control 260, the operation of the energy switch 256 and/or the control 260 may be based on a configuration of the energy switch 256. In one implementation, the energy switch 256 may include a probe that is moveably mounted to a cavity. In such cases, by varying an amount of insertion of the probe into the cavity, the energy associated with the accelerator 202 may be adjusted. In some embodiments, the operation of the current control 258 and/or the control 260 may be based on an amount of insertion of the probe into the cavity at the energy switch 256. The energy switch 256 may include a sensor for measuring an amount of insertion by the probe into the cavity. Based on the sensed value (representing the amount of insertion of the probe, and also representing the energy associated with the accelerator 202), the current control 258 may then adjust the current for the electromagnet 212. Similarly, based on the sensed value (representing the amount of insertion of the probe, and also representing the energy associated with the accelerator 202), the control 260 may then adjust the parameter in the dosimetry circuit 232 at the ion chamber 230. In further embodiments, the operation of the current control 258 and/or the control 260 may occur before the operation of the energy switch 256. For example, in one implementation, the current control 158 may be configured to first adjust the current for the electromagnet 212 so that the magnetic field for deflecting the electron beam 208 corresponds with an energy level of the electron beam 208 to be achieved. Then in response to the adjustment of the current, the energy switch 256 changes the energy level associated with the accelerator 202 so that the desired energy level of the electron beam 208 is achieved. The control 260 may also change the parameter in the dosimetry circuit 232 in correspondence with the adjusted current for the electromagnet 212. In another example, the control 260 may be configured to first adjust the parameter in the dosimetry circuit 232 so that the operation of the dosimetry circuit 232 will correspond with an energy level of the electron beam 208 to be achieved. Then in response to the adjustment in the dosimetry circuit 232, the energy switch 256 changes the energy level associated with the accelerator 202 so that the desired energy level of the electron beam 208 is achieved. Also, in response to the adjustment in the dosimetry circuit 232, the current control 158 then adjusts the current for the electromagnet 212 so that the magnetic field for deflecting the electron beam 208 corresponds with the energy level of the electron beam 208 to be achieved. In another example, the operation of the energy switch 256, the current control 258, and the control 260 may occur simultaneously or substantially simultaneously (e.g., within 1 second, or preferably within 0.5 second, and more preferably within 0.1 second). The configuration of the system 10 is advantageous because it allows an energy of the electron beam 208 (and also the energy of the resulting radiation beam 242) to be changed during treatment. The change of the energy can occur quickly, such as within 3 seconds, and preferably within 1 second, and more preferably within 0.5 second, and even more preferably within 0.1 second. In some cases, the system 10 allows intra field switching of energy. The system 10 may also allow changes of energy while the gantry is rotating (e.g., as in RapidArc treatments). In some embodiments, the system 10 may allow a first treatment beam with a first energy to be delivered to the patient from a first gantry angle, and a second treatment beam with a second energy to be delivered to the patient from a second gantry angle. In one implementation, the first treatment beam with the first energy is delivered to the patient while the gantry 12 is at the first gantry angle. The system 10 then stops the delivery of treatment beam, and moves the gantry 12 to the second gantry angle. The system 10 then delivers the second treatment beam with the second energy to the patient while the gantry 12 is at the second gantry angle. In other embodiments, instead of discretely changing the energy in a step-wise manner, the system 10 may continuously change the energy level of the treatment beam while the treatment beam is being delivered towards the patient. For example as the gantry 12 rotates from the first gantry angle to the second gantry angle, the treatment beam may remain “on” and the energy of the treatment beam may be changed continuously from a first level to a second level. As another example, the continuous change of the energy level may be performed by the system 10 while the gantry 12 is at a certain gantry position. Furthermore, in some cases, while the gantry 12 is at a certain gantry position, the system 10 may perform intensity-modulate radiotherapy (IMRT). The IMRT may be performed using the collimator 24, or using both the collimator 24 and the switching of the energy. In some embodiments, the change of the energy may be discrete change in IMRT fields. In other embodiments, the change of the energy may be continuous change in the IMRT fields. The IMRT fields may utilize a step-and-shoot scheme, or a sliding window technique. Also, in some embodiments, the change of the energy may be performed on a field-by-field basis. In addition, allowing changes in energy in the radiation treatment is advantageous. This is because different radiation beam energies have different penetrating powers. For example, a beam energy of 6 MV may achieve a maximum dose at a depth of 1.6 cm below a skin of the patient, while a beam energy of 15 MV may achieve a maximum dose at a depth of 2.9 cm. Thus, higher energy beam may be utilized to reach deeper tumors. In some embodiments, fast and automated energy change may be performed during treatment to maximize dose conformity through difference in depth doses. Also, low energy radiation (e.g., below 6 MV) may have better dose falloff, leading to lower dose to normal tissue. Accordingly, the ability for the system 10 to change energy for different treatment fields will increase conformity to the tumor while reducing integral dose to normal tissue. In one treatment technique, if a beam direction from a certain gantry angle traverses a tumor that is relatively close to the patient's skin, then lower beam energy may be used. Similarly, if a beam direction from a certain gantry angle traverses the tumor that is relatively far from the patient's skin, then higher beam energy may be used. For example, if a target is 2 cm under the patient's skin when viewed from a gantry position of 0°, and the same target is 3 cm under the patient's skin when viewed from a gantry position of 90°, then the system 10 may deliver beam with energy of 6 MV when the gantry 12 is at gantry angle of 0°, and may deliver beam with energy of 15 MV when the gantry 12 is at gantry angle of 90°. In some embodiments, the system 10 is configured to provide radiation beam for treating the patient without using any flattening filter. In other embodiments, the system 10 may be configured to flatten an energy profile across a cross-section of the beam before delivering the beam to treat the patient. The flattening may be achieved using one or more flattening filters. In one implementation, different filters for different respective beam energies may be provided. A mechanical positioner may be configured to place a selected one of the flattening filters (i.e., selected based on the beam energy level) at a beam path to thereby flatten the treatment beam. It should be noted that the configuration of the system 10 is not limited to the example described, and that the system 10 may have other configurations in other embodiments. For example, in other embodiments, the radiation treatment system 10 may not include the beam deflector 210. Instead, the accelerator 202 may be configured to provide the electron beam 208 directly to the target 220 without going through any beam deflector at the second end 206 of the accelerator 202 (FIG. 2B). Removing the beam deflector is advantageous because it may allow intra field switching of energy to be more easily implemented. In some embodiments, the switching of energy may be implemented while the gantry is continuously moving (e.g., as in RapidArc treatments). Also, in other embodiments, the accelerator 202 may be a fixed-field alternating gradient (FFAG) accelerator, or a non-scaling fixed-field alternating gradient (NS-FFAG) accelerator. As discussed, in some embodiments, the beam deflector 210 may include both the electromagnet 212 and the permanent magnet 214. In some embodiments, one or each of the electromagnet 212 and the permanent magnet 214 may be implemented using an Enge mirror or pretzel magnet (FIG. 3). As shown in the figure, the principle of the Enge pretzel magnet is that a charged particle entering a field increasing with x will exit at the same angle as it entered, if the angle is a specific function of the field gradient. Also, it should be noted that in addition to the beam deflector 210 and the dosimetry circuit 232, the system 10 may be configured to adjust other components based on a desired energy level. For example, in some embodiments, the system 10 may be configured to adjust beam steering servos based on an energy level of the beam being provided or to be provided. FIG. 4 is a schematic axial sectional view of an example of the accelerator 202 of FIG. 1. The accelerator 202 comprises a main body 470 having a first end 472, a second end 474, and a chain of electromagnetically coupled resonant cavities (electromagnetic cavities) 416 between the first and second ends 472, 474. The accelerator 202 also includes a plurality of coupling bodies 421, each of which having a coupling cavity 420 that couples to two adjacent cavities 416. The accelerator 202 also has an energy switch 480 (which may be an example of the energy switch 256 of FIG. 1). The accelerator 202 is excited by microwave power delivered by a microwave source at a frequency near its resonant frequency, for example, between 1000 MHz and 20 GHz, and more preferably, between 2800 and 3000 MHz. The accelerator 202 may be excited by power having other levels in other embodiments by an energy source. The microwave source can be a Magnetron or a Klystron, both of which are known in the art. The power enters one cavity 416, preferably one of the cavities along the chain, through an opening 415. In some embodiments, the accelerator 202 is configured to be operated with an automatic frequency control, such as that described in U.S. Pat. No. 3,820,035, for controlling an operation of a microwave source. The automatic frequency control helps the microwave source (or the RF driver) determine the accelerator 202 resonance by developing an error voltage that tracks a frequency error. The U.S. Pat. No. 3,820,035 is expressly incorporated by reference herein. Alternatively, or additionally, a control, such as that disclosed in U.S. Pat. No. 3,714,592, can be provided to provide feedback to the microwave source (e.g., a Magnetron) by deflecting some of the reflected signal generated by the accelerator 202, and sending it back into the microwave source. U.S. Pat. No. 3,714,592 is expressly incorporated by reference herein. In some embodiments, the wall 444 of the main body 470 adjacent to a gun source 414 (which may be an example of the electron source 200 of FIG. 1) can include one or more pump out holes (not shown) for improving molecular flow conductance. In such cases, the accelerator 202 can further include a tuning ring (not shown) secured to an interior surface of the wall 444 for compensating the detuning from the pump out holes. The tuning ring can be manufactured with the wall 444 as a single unit. Alternatively, the tuning ring and the wall 444 can be manufactured separately, and then assembled together. Also, in some embodiments, the accelerator 202 can further include a copper plate, such as that described in U.S. Pat. No. 3,546,524, disposed at the interior face of the wall 444. The copper plate functions to terminate and shape the electric field. During use, a linear beam 412 of electrons is injected into the accelerator 202 by a electron gun source 414 at the first end 472. The beam 412 may be either continuous or pulsed. The beam 412 passes through a first section 476 of the accelerator 202 in which electrons are captured and accelerated, and enters a second section 478 of the accelerator 202 where the captured electrons are further accelerated. Amplitude of the electric field in the second section 478 (i.e., downstream) can be adjusted by operation of the energy switch 480. Since the formation of electron bunches from an initial continuous beam takes place in the first section 476 of the accelerator 202, the bunching can be accomplished and/or optimized there and not degraded by the varying accelerating field in the output cavities 416 of the second section 478. The spread of energies in the output beam is thus made independent of the varying mean output electron energy. By controlling the RF power input (which changes the relative electric field between the first and second sections 476, 478), and the energy switch 480 (which changes the electric field in the second section 478), one can optimize spectrum of energies and maintain stable charging (or filling) of the accelerator 202. In some embodiments, the accelerator 202 may optionally include a field step control to provide a change in the electric field (e.g., a stepped field) to decrease the range of field variation associated with operation of the energy switch 480. This use of field step has an effect of decreasing separations of resonant modes of the accelerator 202, so that an optimum range of beam energies can be generated. This in turn results in a relatively stable bandwidth, allowing the accelerator 202 to generate x-ray beam with a wider range of energy levels and minimum energy spread. In some embodiments, the field step control enables the accelerator 202 to provide x-ray beam or a particle beam (e.g., electron beam, proton beam, etc.) having an energy level that ranges from approximately 4 to 20 MeV. In other embodiments, the x-ray beam may have an energy level that is below 4 MeV (e.g., 2 MeV, or in the kV range). In further embodiments, the x-ray beam may have an energy level that is higher than 20 MeV. In some embodiments, the field step control is located further away from the beam source 414 than the energy switch 480, and is positioned adjacent to the energy switch 480. Alternatively, the field step control may be located at other positions, such as between the beam source 414 and the energy switch 480, or further downstream from the energy switch 480. The field step control may be implemented using different sizes and/or shapes of openings that are coupling between a side cavity and two adjacent main cavities, using a ring that is secured to a dividing wall separating adjacent cavities, using a nose (secured to a dividing wall that separates adjacent cavities) with a different size and/or shape compared to that of an adjacent nose, and/or using a beam aperture with a different size and/or shape compared to that of an adjacent beam aperture. Field step controls have been described in U.S. Pat. No. 7,339,320, the entire disclosure of which is expressly incorporated by reference herein. In the illustrated embodiments, the electromagnetic cavities 416 are doughnut shaped with aligned central beam apertures 417 which permit passage of the beam 412. The cavities 416 may have projecting noses 419 of optimized configuration in order to improve efficiency of interaction of the microwave power and electron beam. The cavities 416 are electromagnetically coupled together through the coupling cavities 420, each of which is coupled to each of the adjacent pair of cavities 416 by an opening 422. The coupling cavities 420 are resonant at the same frequency as the accelerating cavities 416 and do not interact with the beam 412. The frequency of excitation is such that the chain is excited in standing wave resonance with a π/2 radian phase shift between each coupling cavity 420 and the adjacent accelerating cavity 416. Thus, there is a π radian shift between adjacent accelerating cavities 416. The π/2 mode has several advantages. It has the greatest separation of resonant frequency from adjacent modes which might be accidentally excited. In other embodiments, the shift between adjacent cavities 416 may have other values. Also, when the chain is properly terminated, there are very small electromagnetic fields in coupling cavities 420 so the power losses in these non-interacting cavities are small. The first and last accelerating cavities 426 and 428 are shown as having one-half of an interior cavity 416. It is of course understood that, in alternative embodiments, the terminal cavities 426, 428 may each be a full cavity or any portion of a cavity. The spacing between accelerating cavities 416 is about one-half of a free-space wavelength, so that electrons accelerated in one cavity 416 will arrive at the next accelerating cavity in right phase relative to the microwave field for additional acceleration. Alternatively, the accelerating cavities 416 can have other spacing. In some embodiments, most of the accelerating cavities 416 and most of the coupling cavities 420 are similar such that the fields in most of the accelerating cavities 416 are substantially the same. Alternatively, the accelerating cavities 416 and/or the coupling cavities 420 can have other configurations such that the fields in some of the cavities 416 are different. In the illustrated embodiments, the first section 476 (i.e., the “buncher”. section) has 3-½ cavities 416, and the second section 478 (i.e., the “accelerating” section) of the accelerator 202 has 2-½ cavities 416. Alternatively, each of the sections 476, 478 of the accelerator 202 can have other number of cavities 416. The energy switch 480 of the accelerator 202 is mounted to a cylindrical cup-shaped body 450 having a cavity 434 and an opening 451, and includes a probe 456 inserted through the opening 451, and a choke 458 coaxially surrounding at least part of the probe 456. The choke 458 is a double quarter-wave choke configured to facilitate transmission of high current around the opening 451 by functioning as an impedance transformation of short circuit to opened circuit. The body 450 is attached to the main body 470 of the accelerator 202 such that the cavity 434 is coupled to adjacent cavities 416 through respective openings 438, 440. The energy switch 480 also includes a pair of axially projecting conductive capacitively coupled noses 454 having opposed end faces that extend axially into the cavity 434. The body 450 and the noses 454 are similar to the body 421 and noses 424 discussed previously. In some embodiments, the cavity 434 (the switched side-cavity) is tuned to the same frequency as are the other coupling cavities 420. Such can be accomplished, for example, by varying a diameter or cross sectional dimension of the probe 456 when the probe 456 is at least partially inserted into the cavity 434. Alternatively, the tuning can be accomplished by varying separation between the noses 454 when the probe is not inserted into the cavity 434. The probe 456 is positioned such that it is offset from a center line 459 of the body 450. In the illustrated embodiments, the probe 456 is located upstream of the center line 459 of the body 450. Alternatively, the probe 456 can be located downstream of the center line 459. The probe 456 is preferably circular cylinder although it could have other cross sectional shapes. In the illustrated embodiments, the probe 456 is made from stainless steel, but can also be made from other materials. The probe 456 has a lumen 457 extending along its length. During use, cooling fluid can be delivered into the lumen 457 (e.g., via another tube inserted coaxially into the lumen 457) for cooling of the probe 456. In alternative embodiments, the probe 456 has a solid cross section and does not have a lumen. The use of a single probe provides physical room for the mechanisms which engage the end of the probe 456 to advance and retract the probe 456 without mechanical interference. The mechanism (not shown) can comprise electrically actuated solenoid(s) or pneumatically operated cylinder(s). Movement of the probe 456 is through the vacuum wall via bellows 461, which provides a vacuum seal. During use, the pair of noses 454 function as coupling resonators, and the probe 456 functions as a third resonator. By varying a degree of insertion of the probe 456 into the cavity 434, distances between the probe 456 and the noses 454 change correspondingly, thereby altering the magnetic fields which couple to the openings 438, 440. This in turn alters the energy level of the beam downstream from the switch 480. It should be noted that the type of switch that can be employed with the accelerator 202 is not necessarily limited to the example discussed previously, and that other types of switches known in the art can also be used. By means of non-limiting examples, accelerator switches such as those described in U.S. Pat. Nos. 4,382,208, 4,286,192, US 2007/0215813 can be used. U.S. Pat. No. 6,366,021 teaches switching electric fields in a coupling cavity by inserting two probes of selected diameter to provide different upstream and down stream electric field coupling to adjacent accelerating cavities. U.S. Pat. Nos. 6,366,021, 4,382,208, and 4,286,192, and US 2007/0215813 are expressly incorporated by reference herein. Also, in alternative embodiments, the energy switch 256 can be located at other position along the length of the accelerator 202, instead of that shown in the illustrated embodiments. Furthermore, although only one energy switch is shown in the previously described embodiments, alternatively, the accelerator 202 can have a plurality of energy switches. FIG. 5 illustrates a treatment method 500 that may be performed by the radiation treatment system 10. The treatment method 500 includes configuring a radiation system for delivering a first radiation beam having a first energy level (item 502); delivering the first radiation beam by the radiation system towards a patient that is on a patient support (item 504); configuring the radiation system for delivering a second radiation beam having a second energy level (item 506); and delivering the second radiation beam by the radiation system towards the patient (item 508). In some embodiments, the act of configuring the radiation beam for delivering the second radiation beam in item 506 comprises changing an energy that is associated with the accelerator 202 by the energy switch 256, and adjusting a current of the electromagnet 212 by the current control 258 in correspondence with the energy associated with the accelerator 202. The adjustment of the current may occur before, after, or simultaneously with, the adjustment of the energy that is associated with the accelerator 202 by the energy switch 256. In some cases, the switching of energy may be performed between treatment delivery fields using a fast and automated method. Alternatively or additionally, the switching of energy may be performed during an treatment delivery field. In some embodiments, the first radiation beam in item 504 is delivered towards the patient from a first gantry angle, and the second radiation beam in item 508 is delivered towards the patient from a second gantry angle. In other embodiments, the first radiation beam in item 504 is delivered towards the patient from a first gantry angle, and the second radiation beam in item 508 is delivered towards the patient from the first gantry angle. Also, in some embodiments, the treatment system 10 comprises the ion chamber 230 with the dosimetry circuit 232. In such cases, the method 500 further comprises adjusting a parameter in the dosimetry circuit 232 in correspondence with the energy associated with the accelerator 202. The adjusting of the parameter in the dosimetry circuit 232 may occur before, after, or simultaneously with, the adjustment of the current for the electromagnet 210. In some embodiments, the acts of configuring the radiation system 10 (items 502, 506), and the acts of delivering (items 504, 508) are performed so that the first radiation beam transitions to the second radiation beam in a continuous manner. In other embodiments, the acts of configuring the radiation system 10 (items 502, 506), and the acts of delivering (items 504, 508) are performed so that the first radiation beam and the second radiation beam are delivered in discrete manner. FIG. 6 illustrates a treatment planning method 600. The method 600 includes defining control points in a treatment plan; setting up energy switching in one or more of the control points (item 602); and performing treatment optimization on the treatment plan based at least in part on the energy switching that is set up in the one or more of the control points (item 604). In some embodiments, a user interface may be provided that allows a user to define the control points in the treatment plan, and to set up energy switching in one or more of the control points. Also, in some embodiments, the features (e.g., control points, and energy switching parameters) of the treatment plan, and the treatment plan itself, may be stored in a non-transitory medium for later use. In some embodiments, the act of setting up energy switching in item 604 comprises prescribing the energy switching between two gantry positions. In other embodiments, the act of setting up energy switching comprises prescribing the energy switching at a same gantry position. Also, in some embodiments, the energy switching is set up in item 604 so that the treatment plan will include a delivery of a first energy beam for treating a patient, and a delivery of a second energy beam for treating the patient. FIG. 7 illustrates an example of a user interface 1100 that may allow a user to determine a treatment plan in accordance with some embodiments. As used in this specification, the term “user” may refer to a single person, or a plurality of persons. In some cases, the user interface 1100 may be used in the method 600 to determine a treatment plan. The user interface 1100 includes a screen 1102 displaying an input interface 1104. The input interface 1104 may be generated by a processor that executes a set of instruction programmed to provide the image of the input interface 1104. In the illustrated embodiments, the input interface 1104 includes a table 1106 having fields that allow the user to input parameters and/or values. In the illustrated example, the user has defined in table 1106 control points 1108, parameter 1110 for the allowable starting point of support 14, parameter 1112 for allowable ending point of support 14, parameter 1114 for allowable gantry starting angle, and parameter 1116 for allowable gantry ending angle. As shown in the example, a control point may represent treatment parameter(s) at a given time, dose, gantry angle, etc., or an interval between two points of varying treatment parameters (e.g., “0-1 interval,” “1-2 interval”). The table 1106 includes various input fields for allowing the user to input values for the parameters at different control points. As shown in the example, a value may be a numerical value, or an instruction (e.g., “interpolate”—which specifies that values for the corresponding control point are to be calculated in accordance with a prescribed scheme). In some embodiments, the user needs not enter all or any value for the input fields. In such cases, the processor/software for determining the treatment plan is configured to determine the values for the various fields in the table 1106. In the illustrated example, a trajectory is defined by the control points 1108, wherein each control point 108 defines a region in parameter space. The optimized trajectory has to pass through the defined region before proceeding to the next control point 108. The allowed region (range of parameters) between the control points is also defined. Thus, the control points define the region where the machine control points must be placed. In some cases, the processor/software for determining the treatment plan is configured to generate machine control points based on these rules and the defined parameters. For example, the processor/software may perform optimization based on geometric properties of target region(s) and healthy region(s). The processor/software may then continue with the optimization using dose based method(s), e.g., direct aperture method, fluence based method, etc. In some cases, the processor/software may perform Pareto optimization, or any of other types of multi-criteria optimization. In some embodiments, the user interface 1100 allows the user to input initial values for some or all of the parameters. During the optimization process, the processor/software optimizes the values based on certain user-defined constraints (e.g., size, shape, and location of target, path of source, etc.). In other embodiments, the processor/software may be configured to determine the values for the parameters without any initial input values from the user. In the illustrated example of FIG. 7, the trajectory would move the support 14 from 0 cm to 40 cm in the Z direction (from control point 0 to control point 1), and back to 0 cm (from control point 1 to control point 2). The gantry 12 would rotate from somewhere between 0° and 45° to between 315° and 360°, and back to between 0° and 45°. The Z-positions of the support 14 would be interpolated (e.g., linearly, or using some other interpolation scheme) between the control points. The trajectory then would move the gantry to 45° where treatment beam is to be delivered at 10 MeV (control point 3). The trajectory then would move the patient support in the Z direction to position 10 cm, and would move the gantry to 320° where treatment beam is to be delivered at 15 MeV (control point 4). The trajectory then would move the gantry from 320° to 270°, where through this gantry range, the treatment beam is to be delivered at energy level that ranges from 15 MeV to 6 MeV. In some embodiments, the optimizer of the processor/software that is used to perform method 600 is configured to determine the route between the control points for gantry angles in the 0° to 360° interval. In some cases, the user interface 1100 also allows the user to perform simple operations on defined trajectory. For example, in some embodiments, the trajectory of FIG. 7 may be stretched in the Z-direction by applying a multiplication of 2 in the Z-direction of the support 14. In other embodiments, at least part of the trajectory may be shifted. In some embodiments, the user interface 1100 allows the user to save the designed trajectory in a medium. The trajectory may be saved as a part of a treatment plan, which will be used later in a treatment procedure. Alternatively, or additionally, the trajectory may be saved as a trajectory class. In some cases, the trajectory classes may be organized based on specific machines (e.g., different machines may have different classes of trajectories), patient anatomy, location of target regions, sizes of target regions, shapes of target regions, and/or other disease specific factors. In such cases, a user may retrieve a trajectory from one of the available trajectory classes, based on the specific machine, target region's shape, size, and location, and type of disease. The user may then revise the retrieved trajectory to fine-tune it so that is can be better used for a specific treatment for a specific patient. For example, the user may perform a multiplication and/or an adding procedure for any part (e.g., a parameter type) of the trajectory, to thereby fit the dimensions and/or positions of a target in a specific patient. It should be noted that the type of parameters that may be defined using the user interface 1100 is not limited to the example discussed, and that the user interface 1100 may allow the user to define other parameters, such as gantry angle, positions (e.g., x, y, z) of support 14, orientations (Ox, Oy, Oz) of support 14, dose (e.g., user may specify whether dose is to be delivered for a control point), dose rate, leaves' positions, and speed limits (e.g., of gantry rotation, leaves movements, support 14 movements, etc.). Also, the user interface 1100 may allow the user to prescribe whether to provide a constant energy at a certain gantry angle, or to provide a variable energy at a certain gantry angle. The user interface 1100 may further allow the user to prescribe whether to provide a constant energy while rotating the gantry, or whether to provide a variable energy while rotating the gantry. As illustrated in the above embodiments, the user interface 1100 provides a flexible method for a planner to communicate to the optimizer which class of trajectories is considered for a specific case. The trajectory is defined as a set of control points, in which some parameters are to be optimized, and other parameters are to be interpolated. Energy switching may be implemented as one or more parameters through the user interface 1100. In some embodiments, parameters that are not optimized are interpolated using an interpolation scheme. The user interface 1100 also allows ranges to be defined, and provides tools for a user to manipulate the trajectory class. In some cases, the parameters to be optimized may be different for different intervals of the treatment. Thus, the user interface 1100 provides a tool for allowing a user to define a trajectory that is flexible enough for different applications, and is easy to converge to a good solution (because not all of the parameters need to be optimized—some of the parameters may be interpolated). Specialized Processing System FIG. 8 is a block diagram illustrating an embodiment of a specialized processing system 1600 that can be used to implement various embodiments described herein. For example, the processing system 1600 may be configured to perform the method 500 of FIG. 5, and/or the method 600 of FIG. 6. The processing system 1600 may also be an example of the processing unit 54 (or a component thereof), an example of the current control 258 (or a component thereof), and/or an example of the control 260 (or a component thereof). The processing system 1600 may also be any processor described herein. Referring to FIG. 8, the processing system 1600 includes a bus 1602 or other communication mechanism for communicating information, and a processor 1604 coupled with the bus 1602 for processing information. The processor system 1600 also includes a main memory 1606, such as a random access memory (RAM) or other dynamic storage device, coupled to the bus 1602 for storing information and instructions to be executed by the processor 1604. The main memory 1606 also may be used for storing temporary variables or other intermediate information during execution of instructions to be executed by the processor 1604. The processor system 1600 further includes a read only memory (ROM) 1608 or other static storage device coupled to the bus 1602 for storing static information and instructions for the processor 1604. A data storage device 1610, such as a magnetic disk or optical disk, is provided and coupled to the bus 1602 for storing information and instructions. The processor system 1600 may be coupled via the bus 1602 to a display 167, such as a cathode ray tube (CRT), for displaying information to a user. An input device 1614, including alphanumeric and other keys, is coupled to the bus 1602 for communicating information and command selections to processor 1604. Another type of user input device is cursor control 1616, such as a mouse, a trackball, or cursor direction keys for communicating direction information and command selections to processor 1604 and for controlling cursor movement on display 167. This input device typically has two degrees of freedom in two axes, a first axis (e.g., x) and a second axis (e.g., y), that allows the device to specify positions in a plane. In some embodiments, the processor system 1600 can be used to perform various functions described herein. According to some embodiments, such use is provided by processor system 1600 in response to processor 1604 executing one or more sequences of one or more instructions contained in the main memory 1606. Those skilled in the art will know how to prepare such instructions based on the functions and methods described herein. Such instructions may be read into the main memory 1606 from another processor-readable medium, such as storage device 1610. Execution of the sequences of instructions contained in the main memory 1606 causes the processor 1604 to perform the process steps described herein. One or more processors in a multi-processing arrangement may also be employed to execute the sequences of instructions contained in the main memory 1606. In alternative embodiments, hard-wired circuitry may be used in place of or in combination with software instructions to implement the various embodiments described herein. Thus, embodiments are not limited to any specific combination of hardware circuitry and software. The term “processor-readable medium” as used herein refers to any medium that participates in providing instructions to the processor 1604 for execution. Such a medium may take many forms, including but not limited to, non-volatile media, volatile media, and transmission media. Non-volatile media includes, for example, optical or magnetic disks, such as the storage device 1610. A non-volatile medium may be considered an example of non-transitory medium. Volatile media includes dynamic memory, such as the main memory 1606. A volatile medium may be considered an example of non-transitory medium. Transmission media includes coaxial cables, copper wire and fiber optics, including the wires that comprise the bus 1602. Transmission media can also take the form of acoustic or light waves, such as those generated during radio wave and infrared data communications. Common forms of processor-readable media include, for example, a floppy disk, a flexible disk, hard disk, magnetic tape, or any other magnetic medium, a CD-ROM, any other optical medium, punch cards, paper tape, any other physical medium with patterns of holes, a RAM, a PROM, and EPROM, a FLASH-EPROM, any other memory chip or cartridge, a carrier wave as described hereinafter, or any other medium from which a processor can read. Various forms of processor-readable media may be involved in carrying one or more sequences of one or more instructions to the processor 1604 for execution. For example, the instructions may initially be carried on a magnetic disk of a remote computer. The remote computer can load the instructions into its dynamic memory and send the instructions over a telephone line using a modem. A modem local to the processing system 1600 can receive the data on the telephone line and use an infrared transmitter to convert the data to an infrared signal. An infrared detector coupled to the bus 1602 can receive the data carried in the infrared signal and place the data on the bus 1602. The bus 1602 carries the data to the main memory 1606, from which the processor 1604 retrieves and executes the instructions. The instructions received by the main memory 1606 may optionally be stored on the storage device 1610 either before or after execution by the processor 1604. The processing system 1600 also includes a communication interface 1618 coupled to the bus 1602. The communication interface 1618 provides a two-way data communication coupling to a network link 1620 that is connected to a local network 1622. For example, the communication interface 1618 may be an integrated services digital network (ISDN) card or a modem to provide a data communication connection to a corresponding type of telephone line. As another example, the communication interface 1618 may be a local area network (LAN) card to provide a data communication connection to a compatible LAN. Wireless links may also be implemented. In any such implementation, the communication interface 1618 sends and receives electrical, electromagnetic or optical signals that carry data streams representing various types of information. The network link 1620 typically provides data communication through one or more networks to other devices. For example, the network link 1620 may provide a connection through local network 1622 to a host computer 1624 or to equipment 1626 such as a radiation beam source or a switch operatively coupled to a radiation beam source. The data streams transported over the network link 1620 can comprise electrical, electromagnetic or optical signals. The signals through the various networks and the signals on the network link 1620 and through the communication interface 1618, which carry data to and from the processing system 1600, are exemplary forms of carrier waves transporting the information. The processing system 1600 can send messages and receive data, including program code, through the network(s), the network link 1620, and the communication interface 1618. It should be noted that as used in this specification, the term “beam”, or any of other similar terms, may refer to one or more beam(s). For example, the system 10 may deliver a treatment beam towards the patient, and may vary the energy of the treatment beam from a first energy level to a second energy level. In such cases, the treatment beam may considered as a single treatment beam with variable energies, or alternatively, as a first beam having the first energy level and a second beam having the second energy level. The above definition applies even in the scenario in which different energy beams are delivered discretely. For example, the system 10 may deliver a first treatment beam having a first energy level, and then stop the treatment beam delivery. The system 10 may then deliver a second treatment beam having a second energy level. In this scenario, the system 10 may be considered as delivering multiple energy beams at different respective energies, or alternatively, as delivering a treatment beam having different instances of deliveries and different energy levels. Also, in some embodiments, a treatment beam may be a photon beam (e.g., x-ray beam), a particle beam (e.g., electron beam, proton beam, etc.), or a combination of a photon beam and a particle beam. Also, the various features described with reference to the beam deflector 210 may be incorporated into other radiation systems in other embodiments, and they are not limited for application to radiation systems with energy switching. For example, the rare earth magnet, the hybrid magnet, etc. may be utilized in any radiation system, with or without energy switching. Although particular embodiments have been shown and described, it will be understood that it is not intended to limit the claimed inventions to the preferred embodiments, and it will be obvious to those skilled in the art that various changes and modifications may be made without department from the spirit and scope of the claimed inventions. The specification and drawings are, accordingly, to be regarded in an illustrative rather than restrictive sense. The claimed inventions are intended to cover alternatives, modifications, and equivalents.
	description	This application claims priority to U.S. Provisional Application No. 62/777,317 filed Dec. 10, 2018, the entirety of which is incorporated by reference. The technical field relates to wearable protective garments. Specifically, the technical field relates to a wearable exoskeleton frame that supports heavy apparel worn by a user. In various fields and professions, workers often wear heavy apparel or safety apparatuses while they perform their duties. Military, police, and other law enforcement personnel often need to wear protective body armor that may weigh more than protective gear in other professions. Firefighters often wear Kevlar coats, water-resistant layers, and heat insulation which combined can weigh over fifty pounds, and physicians and other health care staff who work near x-rays or CT scanners often wear heavy lead aprons to protect them from radiation. The weight of protective apparel is often applied directly on a user's shoulders and spine causing fatigue, poor posture, injury, overheating, skin irritation, and decreased productivity. An exoskeleton frame that decreases the force and weight applied to a user's shoulders and spine by heavy apparel would be well received in the art. According to one aspect, an exoskeleton frame includes a frame body including a front torso member positioned such that when the exoskeleton frame is worn by a user the front torso member is located along a front of a torso of the user; a back torso member positioned such that when the exoskeleton frame is worn by a user the back torso member is located along a back of the torso of the user; a first shoulder band extending from the front torso member to the back torso member; and a second shoulder band extending from the front torso member to the back torso member; and an adjustable belt removably attached to the front torso and back torso member, wherein the adjustable belt is configured to direct a weight of apparel that is worn by the user over the exoskeleton frame to a weight-bearing area of the user located between knees and abdomen of the user, wherein the first shoulder band and second shoulder band are configured to support the weight of the apparel such that the weight of the apparel is not applied to a first shoulder or a second shoulder of the user. According to another aspect, a method for making an exoskeleton frame comprising providing a piece of material; forming a preformed frame body out of the piece of material, such that the preformed frame body includes a first shoulder band, a second shoulder band, an opening between the first shoulder band and second shoulder band, a front torso member, and a back torso member; and manipulating the preformed frame body into a frame body such that when the frame body is worn by a user, the front torso member is located in front of a torso of the user and the back torso member is located along a back of the torso of the user, and the first shoulder band and second shoulder band are configured to support the weight of apparel that is worn by the user over the frame body such that the weight of the apparel is held off of a first shoulder and a second shoulder of the user. According to another aspect, an exoskeleton frame comprises a frame body including a front torso member positioned such that when the exoskeleton frame is worn by a user the front torso member is located along a front of a torso of the user; a back torso member positioned such that when the exoskeleton frame is worn by a user the back torso member is located along a back of the torso of the user; a first shoulder band extending from the front torso member to the back torso member; a second shoulder band extending from the front torso member to the back torso member; and an opening positioned between the first shoulder band and second shoulder band such that when the exoskeleton frame is worn by a user, a neck of the user is located within the opening, wherein the front torso member and back torso member are configured to receive an adjustable belt that is configured to direct a weight of apparel that is worn by the user over the exoskeleton frame to a weight-bearing area of the user located between knees and abdomen of the user, wherein the first shoulder band and second shoulder band are configured to support the weight of the apparel such that the weight of the apparel is held off of a first shoulder and a second shoulder of the user. A detailed description of the hereinafter-described embodiments of the disclosed apparatus and method are presented herein by way of exemplification and not limitation with reference made to the Figures. Although certain embodiments are shown and described in detail, it should be understood that various changes and modifications might be made without departing from the scope of the appended claims. The scope of the present disclosure will in no way be limited to the number of constituting components, the materials thereof, the shapes thereof, colors thereof, the relative arrangement thereof, etc., and are disclosed simply as an example of embodiments of the present disclosure. A more complete understanding of the present embodiments and advantages thereof may be acquired by referring to the following description taken in conjunction with the accompanying drawings, in which like reference numbers indicate like features. With reference to FIG. 1, a perspective view of an exoskeleton frame 10 is shown according to one embodiment. The exoskeleton frame 10 is configured to be worn by a user (as hereinafter shown and described with respect to FIG. 4). The exoskeleton frame 10 includes a frame body 11 and a belt 20. The frame body 11 is made of a curved piece of material. The frame body 11 includes a front torso member 17 and a back torso member 18. The exoskeleton frame includes a first shoulder band 12 and a second shoulder band 13 that each extend between the front torso member 17 and the back torso member 18. The first shoulder band 12 has a first top edge 12A, and the second shoulder band 13 has a second top edge 13A. The frame body 11 has an inner surface 1 and an outer surface 2. The frame body 11 may be rigid. The frame body may be semi-rigid and have flexibility. In this embodiment, the back torso member 18 includes a back plate 18A and a first vertical element 14 that extends from the back plate 18A to a first end 14A of the first vertical element 14. The first and second shoulder bands 12, 13 extend from the back plate 18A opposite the first vertical element 14. The front torso member 17 has a chest plate 17A, which is located in front of a chest area of the user when the exoskeleton frame 10 is worn by a user. The front torso member 17 includes a second vertical element 15 extending from the chest plate 17A opposite the second shoulder band 13 to a second end 15A of the second vertical element 15. The front torso member 17 further includes a third vertical element 16 extending from the chest plate 17A opposite the first shoulder band 12 to a third end 16A of the third vertical element 16. The first, second, and third ends 14A, 15A, 16A may be flat, rounded, or have another shape. In this embodiment, the first end 14A is flat, and the second and third ends 15A, 16A are rounded. The chest plate 17A and back plate 18A may each be a panel, a plate, a plaque, a surface, and the like. The exoskeleton frame 10 is not limited to having a back torso member 18 with a first vertical element 14. For example, the back torso member 18 may have one or more vertical elements. Further, the front torso member 17 is not limited to having a second and a third vertical element 15, 16, and may have less than two or more than two vertical elements. The exoskeleton frame 10 has a first edge 22 that extends around the frame body 11 and abuts the inner surface 1 and outer surface 2 of the frame body 11. The frame body 11 further includes an opening 350 having a second edge 23 defined by the first and second shoulder bands 12, 13, the front torso member 17, and the back torso member 18. When the exoskeleton frame 10 is worn by the user, the user's neck and head are located within the opening 350. In an embodiment, the first and second edges 22, 23 may be enclosed in a soft material such as rubber, fabric, and the like. For example, the first and second edges 22, 23 may have a rubber lip, a fabric hem, a soft border, and the like. As another example, the first and second edges 22, 23 may be rounded. A “U” shaped space 300 is provided between the second vertical element 15, the third vertical element 16, and chest plate 17A. In this embodiment, the space 300 is “U” shaped. The space 300 may have another shape, such as a “V” shape or other shape. In an embodiment, the second and third vertical elements 15, 16 may not be straight, and may have shape that provides a different shape space 300, such as a circular space, rectangular space, and the like. The space 300 is configured to accommodate differently sized users wearing the exoskeleton frame 10 such that the exoskeleton frame 10 does not place pressure onto a stomach area of the user, or obstruct a user's stomach area from extending to its natural extent. With continuing reference to FIG. 1, in this embodiment, the front torso member 17 has a curved shape such that when the exoskeleton frame 10 is worn by a user, the chest plate 17A curves about the front of the user's chest area. The curve of the front torso member 17 may be configured such that when the exoskeleton frame 10 is worn by the user, the inner surface 1 at the second and third vertical elements 15, 16 lays flat along or against the front of the user's torso. The curve of the front torso member 17 may be configured to mimic the curve and shape of a user's torso such that when apparel is worn over the exoskeleton frame 10, the apparel falls or lays naturally as if worn on a user's body without an exoskeleton frame such as exoskeleton frame 10. The exoskeleton frame 10 may have flexibility such that when the exoskeleton frame 10 is worn by the user, the second and third vertical elements 15, 16 are flat along or against the front of the user's torso. As another example, the second and third vertical elements 15, 16 may have flexibility such that when the exoskeleton frame 10 is worn by the user, the inner surface 1 of the second and third vertical elements 15, 16 is flat along or against the front of the user's torso on either side of the space 300. The first vertical element 14 has a first plurality of slits 50 located proximate to the first end 14A. The second vertical element 15 has a second plurality of slits 30 located proximate to the second end 15A. The third vertical element 16 has a third plurality of slits 40 located proximate to the third end 16A. In this embodiment, the first, second, and third pluralities 50, 30, 40 of slits include 3 pairs of parallel slits arranged on top of one another along the first, second, and third vertical elements 14, 15, 16. Each pair of slits is configured to receive the belt 20 such that the belt is attached to the frame body 11 and the exoskeleton frame 10 is securable to a user. The arrangement of pairs of slits along each of the first, second, and third vertical members 14, 15, 16 provides for adjustable fit of the exoskeleton frame 10 on a user such that a user of any height can wear the exoskeleton frame 10. For example, a user with a shorter torso may insert the belt 20 through the top-most pair of slits of the first, second, and third pluralities of slits 50, 30, 40, and a user having a longer torso may insert the belt through the pair of slits closest to the first, second, and third ends 14A, 15A, 16A. The pairs of slits closest to the first, second, and third ends 14A, 15A, 16A may be one (1) inch away from the first, second, and third ends 14A, 15A, 16A. As another example the pairs of slits closest to the first, second, and third ends 14A, 15A, 16A may be more than one inch from the first, second, and third ends 14A, 15A, 16A, for example, two (2) to six (6) inches, and the like. Each of the first, second, and third vertical elements 14, 15, 16 may include any number of slits or other belt attachment structures. The exoskeleton frame 10 may include other belt attachment structures, and is not limited to receiving the belt 20 by a first, second, and third pluralities of slits 50, 30, 40. For example, the first, second, and third vertical members 14, 15, 16 may include one or more holes, slots, bores, cuts, spaces, openings, apertures, eyelets, buttons, hook and loop surfaces, fasteners, ties, hooks, tabs, and the like that are configured to removably receive or attach the belt 20 to the exoskeleton frame 10. As another example, the first, second, and third vertical elements 14, 15, 16 may each include a vertically adjustable track. For example, the first, second, and third vertical elements 14, 15, 16 may each include a plurality of notches arranged vertically on a track, and the belt 20 may be affixed to each of the vertically adjustable tracks such that the arrangement of the belt 20 along each of the first, second, and third vertical elements 14, 15, 16 is adjustable along each track by the notches. As shown in FIG. 1, the belt 20 is removably attached to the first, second, and third vertical elements 14, 15, 16 by being inserted through a pair of slits on each of the first, second, and third vertical elements 14, 15, 16. The belt 20 may also be removably attached to the exoskeleton frame 10 by being tied to one or more slits of the first, second, and third pluralities of slits 50, 30, 40, laced through one or more slits of the first, second, and third pluralities of slits 50, 30, 40, looped around one or more slits of the first, second, and third pluralities of slits 50, 30, 40, and the like. In this embodiment, each pair of slits of the first, second, and third pluralities of slits 50, 30, 40 are aligned. For example, each middle pair of slits on the first, second, and third vertical elements 14, 15, 16 are aligned such that when the belt 20 is inserted through each middle pair of slits, the belt 20 is configured to extend around the first, second, and third vertical element 14, 15, 16 and horizontally around a weight-bearing area of the user. A weight-bearing area of the user may be an area between the user's knees and abdomen, such as the user's hips, waist, and thighs. The belt 20 is made out of a flexible material. For example, the belt 20 may be made out of polypropylene webbing, leather, elastic, cord, string, rope, fabric, canvas, and the like. In this embodiment, the belt 20 has a buckle 21 configured to secure the belt 20 to a user at a desired weight-bearing area between the user's knees and abdomen. The belt 20 is not limited to having a buckle 21 and could be securable by hook-and-loop fasteners, fasteners, buttons, magnets, hooks, and the like. The belt 20 is adjustable in length to accommodate differently sized users. For example, an operable length of the belt 20 may be adjusted by pulling the belt through the buckle 21 or other fastener, or by the belt 20 being made out of an elastic material that can stretch and retract based on a user's dimensions. The buckle 21 is oriented such that a user can access the buckle 21 when the exoskeleton frame 10 is worn by the user such that the user can secure and unsecure the exoskeleton frame 10 to the user. In another embodiment, instead of a flexible belt 20, the exoskeleton frame 10 may include a rigid support structure including a rigid belt configured to fit around the weight-bearing area of the user's body. A rigid belt may include one or more segments, such as a “C” shaped rigid segment extending from the first vertical element 14 to the second, and third vertical elements 15, 16 and one or more moveable segments between the second and third vertical elements 15, 16 configured to open to receive a user and configured to secure closed into a continuous rigid belt around the weight-bearing area of the user. For example, one or more moveable segments of a rigid belt may be hingedly or foldably attached to the second and third vertical elements 15, 16. With continuing reference to FIG. 1, the frame body 11 may be made of a rigid, or semi-rigid material, such as plastic. The frame body 11 may be made of low-density polyethylene. The frame body 11 may include a high-density polymer. In one embodiment, the frame body 11 may be made out of a material that attenuates x-rays such that a user, for example, a physician who needs to wear a protective garment such as a lead apron, may wear a lead apron that weighs less than a standard lead apron on top of the exoskeleton frame 10 because of the additional x-ray attenuation provided by the frame body 11 of the exoskeleton frame 10. In another embodiment, the frame body 11 may provide all x-ray attenuation needed for the user such that a lead apron does not need to be worn in the presence of x-rays or CT scanner. With continuing reference to FIG. 1, in this embodiment, the front torso member 17 and second torso member 18 are flexibly movable apart from one another to accommodate users having differently sized torsos. To accommodate a wider user, the belt 20 may be loosened, for example, at buckle 21, such that the user fits between the back torso member 17 and front torso member 18. To accommodate a narrower user, the belt 20 may be tightened, for example, at buckle 21, such that the space between the front torso member 17 and back torso member 18 is narrower. With reference to FIG. 2, a front view of the exoskeleton frame 10 is shown. The exoskeleton frame 10 is symmetrical down a center line 500. The first, second, and third ends 14A, 15A, 16A are aligned in this embodiment. In another embodiment, the first, second, and third ends 14A, 15A, 16A may not be aligned. For example, in one embodiment, the second and third vertical elements 15, 16 may extend longer than the first vertical member 14 such that the second and third ends 15A and 16A are below the first end 14A. In another embodiment, the first vertical element 14 may extend longer than the second and third elements 15, 16 such that the first end 14A is lower than the second and third ends 15A, 16A. With continuing reference to FIG. 2, in this embodiment, the first, second, and third vertical 14, 15, 16 have a straight shape. In another embodiment, the first, second, and third element 14, 15, 16 may have a different shape, such as a wavy or curved shape. The first vertical element 14 has a first width 14B, the second vertical element 15 has a second width 15B, and the third vertical element 16 has a third width 16B. The second width 15B of the second vertical element 15 and third width 16B of the third vertical element 16 are equal in this embodiment. The first width 14B of the first vertical element 14 is wider than the second and third widths 15B, 16B in this embodiment. In another embodiment, the first, second, and third widths 14B, 15B, 16B may be equal. In another embodiment, the second and third vertical elements 15, 16 may be wider than the first vertical element 14. With reference to FIG. 3, a back view of the exoskeleton frame 10 is shown. In this embodiment, the back torso member 18 of the exoskeleton frame 10 has a “Y” shape that extends from the first end 14A to the first and second top edges 12A, 13A of the first and second shoulder bands 12, 13. In another embodiment, the frame body 11 may be separable into two pieces. For example, the frame body may be separable into pieces that separate at the first shoulder band 12 and second shoulder band 13 such that each piece can be stacked on top of the other for storage of the exoskeleton frame 10 when not in use. A first piece may include the front torso portion 17 and a first half of each of the first and second shoulder bands 12, 13 extending from the front torso portion 17, and a second piece may include the back torso portion 18 and a second half of each of the first and second shoulder bands 12, 13. Each half of the shoulder bands 12, 13 may extend to the first and second top edges 12A, 13A of the first and second shoulder bands 12, 13. The separable pieces of the frame body 11 may be assembled into the frame body 11. For example, each of the first halves of the first and second shoulder bands 12, 13 may securably interlock with the second halves of the first and second shoulder bands 12, 13, respectively, for example, by a clamp. As another example, each of the first halves of the first and second shoulder bands 12, 13 may have a protrusion such as a peg, pin, nail, fastener, tab, nub, knob, and the like, configured to be securably inserted into a hole of each of the second halves of the first and second shoulder portions 12, 13, such as a hole, slit, slot, bore, opening, and the like. When not in use, the two pieces are separable and the belt 20 is removable from one of the first or second pieces. The first piece and second piece are then stackable such that the exoskeleton frame 10 can be stored and take up less space than when the exoskeleton frame 10 is fully assembled. Referring now to FIG. 4, a perspective view of a first user 100A and a second user 100B each wearing an exoskeleton frame 10 is shown. With respect to each user 100A, 100B, the first vertical element 14 of the exoskeleton frame 10 extends downward along the first and second users' 100A, 100B backs or spines, and the second and third vertical elements 15, 16 are arranged along the fronts of the users' 100A, 100B torsos. The chest plate 17A of the exoskeleton frame 10 is positioned adjacent to a chest area of each of the first and second users 100A, 100B. The inner surface 1 of the frame body 11 is adjacent to the first and second users' 100A, 100B bodies, and the outer surface 2 of the frame body 11 faces away from the first and second users' 100A, 100B bodies. The belt 20 secures the exoskeleton frame 10 to a weight-bearing area of the users 100A, 110B. A weight-bearing area may be an area between the users' 100A, 100B knees and abdomen. In this embodiment, the belt 20 is secured to each users' 100A, 100B waist 140 such that when a user puts on apparel such as a protective garment over the exoskeleton frame 10, a weight of the apparel is directed to that weight-bearing area of the user and not the user's shoulders or spine. The belt 20 secures the exoskeleton frame to the user 100A, 100B such that a gap 70 is provided between each of the user's 100A, 100B shoulders, and the first and second shoulder bands 12, 13. For example, as shown with respect to the first user 100A, a gap 70 is provided between each of the first user's 100A shoulders 110 and the first and second shoulder bands 12, 13. As shown with respect to the second user 100B, a gap 70 is provided between each of the second user's 100B shoulders 120 and the first and second shoulder bands 12, 13. When apparel such as a protective garment is worn over the exoskeleton frame 10 (hereinafter shown and described in FIG. 5), the first and second shoulder bands 12, 13 hold the apparel such that the apparel is lifted off and above a users' 100A, 100B shoulders 110, 120, and simultaneously, the belt 20 directs a weight of the apparel to a weight-bearing area of the respective user such as the waist 140 of the user 100A, 100B and off of the user's shoulders, neck, spine, and back. As shown in the embodiment in FIG. 4, the belt 20 is arranged to direct a weight of apparel worn by the first and second users 100A, 100B to the waist 140 of the first and second user 100A, 100B. The belt 20 is securable to the user 100A, 100B such that the weight of apparel worn over the exoskeleton frame 10 is kept on the weight-bearing area of the user such as the user's 110A, 100B waist 140 while the user 100A, 100B is in a standing position, when the user 100A, 100B bends over, or bends to the right or left, or makes another movement. In this embodiment, the exoskeleton frame 10 is one-size-fits-all. The same size exoskeleton frame 10 may provide gaps 70 above differently sized user's shoulders by the position of the belt 20 along the first, second, and third vertical elements 14, 15, 16. For example, by attaching the belt 20 to a different set of slits 50, 30, 40 on the first, second, and third vertical members 14, 15, 16. As an example shown in FIG. 4, the first user 100A is taller than the second user 100B. The first user 100A is wearing the exoskeleton frame 10 with the belt 20 inserted through the middle pair of slits 30B, 40B of the second and third vertical element 14, 15 and the middle pair of slits of the first vertical element (not shown) such that the belt 20 is located at the waist line 140 of the first user 100A, and such that a gap 70 is provided between each of the shoulders 110 of the first user 100A and the first and second shoulder bands 12, 13. The second user 100B is wearing the exoskeleton frame 10 with the belt 20 inserted through the top pair of slits 30A, 40A of the second and third vertical element 15, 16 and top pair of slits of the first vertical element 14 (not shown) such that the belt 20 is located at the waist 140 of the second user 100B, and such that a gap 70 is provided between each of the shoulders 120 of the second user 100B and the first and second shoulder bands 12, 13. A user that is taller than the first user 100A may need to attach the belt 20 to the bottom sets of slits of the first, second, and third vertical members 14, 15, 16 in order to secure the exoskeleton frame 10 to the user such that a gap 70 is provided above each of the user's shoulders. With continuing reference to FIG. 4, in this embodiment, flexibility of the frame body 11 permits the first and second shoulders 12, 13 of the exoskeleton frame 10 to compress downwards when apparel such as a protective garment is worn over the exoskeleton frame 10. The gaps 70 provided underneath each of the first and second shoulder bands 12, 13 are configured such that any compression of the first and second shoulder bands 12, 13 does not cause any force or weight of the apparel to be applied to the user's 100A, 100B shoulders, neck, spine, back, and the like. For example, the first and second shoulder bands 12, 13 are configured such that any compression by apparel worn over the exoskeleton frame 10 leaves a gap 70 between each of the user's shoulders and the first and second shoulder bands 12, 13. When apparel worn by the user 100A, 100B over the exoskeleton frame 10 is taken off, the first and second shoulder bands 12, 13 may decompress and extend upwards back into a default decompressed position. Flexibility of the frame body 11 may further permit the exoskeleton frame to be more lightweight than a rigid frame body 11 embodiment in which the frame body 11 has no flexibility. In an embodiment in which the frame body 11 is not flexible, depending on the material of the frame body 11, the frame body 11 may need to be thicker in order to be rigid such that any weight of apparel such as a protective garment is kept off of a user's 100A, 100B shoulders and spine by the first and second shoulder bands 12, 13 and gaps 70 provided thereunder. The first and second shoulder bands 12, 13 may be configured to support apparel with differently sized shoulders. With continuing reference to FIG. 4, the first shoulder band 12 of the exoskeleton frame 10 has a width 12C, and the second shoulder band 13 of the exoskeleton frame 13 has a width 13C. The widths 12C, 13C of the first and second shoulder bands 12, 13 may be configured such that shoulders of a protective garment or other apparel worn over the exoskeleton frame 10 do not hang over the edges of the first and second shoulder bands 12, 13, but are fully supported by the widths 12C, 13C of the first and second shoulder bands 12, 13. In another embodiment, the first and second shoulder bands 12, 13 may include an adjustable width element, such as a tab that is extendable from the first and second shoulder bands 12, 13 parallel to the user's shoulders 110, 120 such that shoulders of a protective garment that are wider than the first and second shoulder bands 12, 13 are supported by the shoulder bands 12, 13 and the extended tabs. In another embodiment, the first and second shoulders 12, 13 of the exoskeleton frame may be configured to receive a shoulder extending attachment such as a shoulder pad, a panel, a tab, and the like, that can releasably affixed to each of the first and second shoulder bands 12, 13 such that the first and second shoulder bands 12, 13 are customizable in width to receive different apparel. As an example, an exoskeleton frame 10 may be sold or otherwise provided with one or more pairs of shoulder-extending attachments that may be affixed by fasteners, hook-and-loop fasteners, ridges configured to snap onto the first and second shoulder bands 12, 13, ties, buttons, hooks, and the like. In another embodiment, the exoskeleton frame 10 may include one or more straps to further secure the exoskeleton frame 10 to a user. For example, the chest plate 17 may include one or more straps that are attachable to the back torso member 18 when the exoskeleton frame 10 is worn by a user. One or more straps on the chest plate 17 may further keep the exoskeleton frame 10 centered such that the back torso member 18 is located along a user's spine, and such that the front torso member 17 is centered in front of the user's torso when the exoskeleton frame 10 is worn. When a user such as first and second users 100A, 100B, is wearing apparel such as a protective garment over the exoskeleton frame 10, the frame body 11 provides a barrier between the user's body and the apparel such that the apparel is not in direct contact with the user's body. This permits heat generated by the user's body to move freely without being trapped within the apparel against the user's body, which helps prevent the user from over-heating and sweating. In an embodiment, the exoskeleton frame 10 may include further temperature control features. For example, the frame body 11 may include one or more pads affixed to the inner surface 1 that are configured to rest against a user's body. As an example, an area of the inner surface 1 at the chest plate 17A may include a pad, and an area of the inner surface 1 at each of the first, second, and third vertical elements 14, 15, 16 may include a pad. In another embodiment, the frame body 11 may include one or more pockets configured to contain thermoregulating liquid or receive a pouch or enclosure of thermoregulating liquid such as chilled water or a coolant. In yet another embodiment, the frame body 11 may have a fabric covering such as cloth, polyester, cotton or the like to make the exoskeleton frame 10 soft. In one embodiment the exoskeleton frame 10 may be permanently enclosed in fabric. In another embodiment, the first, second, and third vertical elements 14, 15, 16 may be permanently or removably enclosed in fabric. Referring now to FIG. 5, a front view is shown of the first user 100A of FIG. 4 wearing a protective garment 200 over the exoskeleton frame 10. The protective garment 200 is a lead apron of the type that a physician or health care staff may wear to protect themselves from x-rays or CT scanners. Different apparel such as body armor or other protective gear worn by military, police, and other law enforcement personnel; coats; suits; firefighter gear; and the like used by personnel in any profession involving wearable gear or protection may be worn over an exoskeleton frame 10. In this embodiment, the protective garment 200 has a front 201, a first shoulder 212 and a second shoulder 213. The protective garment 200 also has a back (not shown) and two hook-and-loop panels 210 that extend around the user 100A from the back of the protective garment 200 and attach to the front 201 thereby securing the protective garment 200 to the user 100A. The protective garment 200 is worn by the user 100A over the exoskeleton frame 10 such that the first shoulder 212 of the protective garment 200 is supported by the first shoulder band 12 of the exoskeleton frame 10 and such that the second shoulder 213 of the protective garment 200 is supported by the second shoulder 13 of the exoskeleton frame 200. The first and second shoulders 12, 13 of the exoskeleton frame 10 is supporting the protective garment 200 such that force of the weight of the protective garment 200 is not applied to the user's shoulders 110, spine, neck, or back. A gap 70 between each of the user's shoulders 110 and the first and second shoulders 212, 213 of the protective garment 200 is provided by the first and second shoulder bands 12, 13 of the exoskeleton frame 10. In this embodiment the belt 20 (not shown) of the exoskeleton frame 10 is arranged through a pair of slits on each of the first, second, and third vertical elements 14, 15,16 such that the weight of the protective garment 200 is directed to a weight-bearing area 140 of the user 100A such as the user's 100A waist. The weight-bearing area 140 may be the user's 100A waist, hips, thighs, abdomen, and the like. The gaps 70 provided between each of the first user's 100A shoulders 210 and the first and second shoulders 212, 213 of the protective garment increases the mobility and flexibility of the user when wearing the protective garment 200. For example, the range of motion of the user's 100A arms and shoulders 110 is increased because the user's arms and shoulders 110 are not bearing the weight of the protective garment 200. This may increase the productivity and precision of the user's activity, for example, surgery preparation, surgery, and patient manipulation. These benefits are provided by the exoskeleton frame 10 when worn underneath other gear and garments as well. As another example, in the case of military and law enforcement personnel wearing body armor, gaps 70 between the first and second shoulder bands 12, 13 of the exoskeleton frame 10 and shoulders 110 of the users may increase the physical capabilities of the user by reducing muscle fatigue from the weight of the body armor, and increasing flexibility by keeping weight off of the shoulders, spine, and back of the user, thereby increasing physical abilities to move, run, lift, drag, carry, and the like. As yet another example, in the case of a firefighter wearing protective garments such as Kevlar, gaps 70 provided between the first and second shoulder bands 12, 13 of the exoskeleton frame 10 and shoulders 110 of the users may increase the physical capabilities of the user by reducing muscle fatigue from the weight of the protective garments and increasing muscle power for firefighting activities such as lifting, rescue, and the like. By keeping the weight of the protective garment 200 and other protective garments and apparel off of the user's shoulders and spine and directing the weight to a weight-bearing area of the user such as the user's abdomen, waist, thighs, or hips, deleterious effects of wearing a heavy apparel such as a protective garment are prevented, for example, neck pain, back pain, orthopedic injury, musculoskeletal damage, Interventionalist's Disc Disease, and the like. Further, apparel such as lead aprons often applies a force on a user's neck, shoulders, and back that pulls the user's neck, shoulders, and back forward due to the weight of a protective garment. Wearing an exoskeleton frame 10 underneath such a protective garment prevents this pulling force from being applied to the user and injuring a user's neck, shoulders, and back. Accordingly, early retirement, missed work days, and compromised work performance may thereby be decreased. In one embodiment, the exoskeleton frame 10 may include one or more sensors that monitor one or more health metrics of a user. As an example, one or more sensors may be affixed to the inner surface 1 or outer surface 2 of the exoskeleton frame 10. One or more sensors may be affixed to the exoskeleton frame by adhesive, a pocket, a button, a fastener, hook-and-loop enclosures, and the like. One or more sensors on the exoskeleton frame 10 may measure heart rate, EKG, breathing rate, internal body temperature, external body temperature, temperature of a space in between the user and a protective garment worn over the exoskeleton frame 10, and the like. The one or more sensors may be configured to transmit measured data. For example, one or more sensors may transmit by Bluetooth, Wi-Fi, and the like. With reference to FIG. 6, a front view is shown of a preformed frame body 400 of an exoskeleton frame 10 in a step of a method of making the exoskeleton frame 10. The preformed frame body 400 may be formed, for example cut, out of a sheet of material such as plastic, for example, low-density polyethylene, such that the preformed frame body 400 includes back torso member 16 and front torso member 17. The preformed frame body 400 has been cut such that the front torso member 17 includes the second vertical element 15 and third vertical element 16. The preformed frame body 400 has been cut such that the second vertical element 15 and third vertical element 16 have a space 300 between the second vertical element 15 and third vertical element 16. The preformed frame body 400 has further been cut to include an opening 350 which, when the preformed frame body 400 is shaped into the frame body 11, will define a space within which a user's neck and head will be located when the exoskeleton frame 10 is worn by the user. The preformed frame body 400 has a length 401 that extends between the first end 16A of the back torso member 18 to the second and third ends 15A and 16A. The preformed frame body 400 has a center point 402 at the center of the length 401. The center point 402 defines a bend location where the preformed frame body 400 may be bent or curved in half such that a first top edge 12A of the first shoulder band 12 and a second top edge 13A of the second shoulder band 13 are formed at the center point 402. A method of making an exoskeleton frame 10 may include providing a piece of material; forming a preformed frame body, such as preformed frame body 400, out of the piece of material, such that the preformed frame body includes a first shoulder band, such as first shoulder band 12, a second shoulder band, such as second shoulder band 13, an opening between the first shoulder band and second shoulder band, such as opening 350, a front torso member, such as front torso member 17, and a back torso member, such as back torso member 18; manipulating the preformed frame body into a frame body, such as frame body 11, such that when the frame body is worn by a user, the first shoulder band and second shoulder band are located above a first shoulder and a second shoulder of the user, the front torso member is located in front of a torso of the user and the back torso member is located along a back of the torso of the user; and attaching an adjustable belt such as belt 20 to the front torso member and back torso member. A method of making an exoskeleton frame may include forming the preformed body out of the piece of material such that the back torso member includes a first vertical element, such as first vertical element 14, having a first end, such as first end 14A, and such that the front torso member includes a chest plate, such as chest plate 17A, a second vertical element, such as second vertical element 15, having a second end, such as second end 15A, and a third vertical element, such as third vertical element 16, having a third end, such as third end 16A, wherein the first shoulder band and second shoulder band extend from the chest plate to the back torso member, wherein the second vertical element and third vertical element extend from the chest plate opposite the first shoulder band and second shoulder band. A method of making an exoskeleton frame may include forming the preformed body out of the piece of material such that a space, such as space 300, is located between the second vertical element and the third vertical element. A method of making an exoskeleton frame may include forming at least one belt attachment structure in each of the front torso member and back torso member. In an embodiment of a method of making an exoskeleton frame, manipulating the piece of material into a frame body may include bending the preformed frame body such that a first end of the back torso member is aligned with a second end of the front torso member. The method may further include applying heat to the preformed frame body while the preformed frame body is bent. In an embodiment of the method, the piece of material may be a low-density polyethylene. In another embodiment, the piece of material may be a flat sheet of material. In a method of making an exoskeleton frame, forming a preformed frame body out of the piece of material may include cutting the piece of material. In another embodiment, forming a preformed frame body out of the piece of material may include stamping the piece of material. In one embodiment, a method of making an exoskeleton frame may include forming at least one belt attachment structure, such as one or more pairs of slits 50, 30, 40 on or in each of the back torso member, first vertical element, and second vertical element. The method may further include removably attaching a belt to each of the at least one belt attachment structure on or in each of the first, second, and third vertical elements. In one embodiment a method of making an exoskeleton frame may include loading an electronic file containing computer instructions for forming the preformed frame body into a cutter or stamper, for example, a computer numeral controlled router machine, and the cutter may cut the preformed frame body out of a piece of rigid material, for example, a thick sheet of low-density polyethylene. The method may include bending the preformed frame body in half at a mid-point such as mid-point 402 along a length, such as length 401 of the preformed frame body, into a frame body, and placing the preformed frame body in a shape guiding tool such as a jig, and holding the frame body, for example, by a clamp, strap, and the like. The method may include applying heat to the frame body, for example, by a hot air gun, cooling the frame body, and removing the frame body from the shape guiding tool. In one embodiment, the preformed frame body may be cut from a sheet of material. In another embodiment, the preformed frame body may be stamped out of a sheet of material. In another embodiment, a plurality of preformed frame bodies may be formed from a plurality of sheets of material at the same time, for example, by stamping the preformed frame bodies out of the plurality of sheets of material. In an embodiment, a method of making an exoskeleton frame may include affixing one or more pads, temperature control devices, or sensors on the frame body, such as an inner surface, for example, inner surface 1, of the frame body. In another embodiment, a method of making an exoskeleton frame may include bending the front torso portion along a center line, such as center line 500, extending vertically along the center of the front torso portion such that the front torso portion is curved, and such that when a user is wearing the frame body, the front torso portion curves about the front of the user's torso. The method may further include heating the frame body 11 while the front torso portion is bent along the center line to produce the curve of the front torso portion. The descriptions of the various embodiments of the present invention have been presented for purposes of illustration, but are not intended to be exhaustive or limited to the embodiments disclosed. Many modifications and variations will be apparent to those of ordinary skill in the art without departing from the scope and spirit of the described embodiments. The terminology used herein was chosen to best explain the principles of the embodiments, the practical application or technical improvement over technologies found in the marketplace, or to enable others of ordinary skill in the art to understand the embodiments disclosed herein.
	description	The present invention claims the benefit of Provisional Application No. 60/514,711 filed Oct. 27, 2003. The present invention relates generally to diagnostic imaging and, more particularly, to a method and apparatus to optimize dose efficiency by dynamically filtering radiation emitted toward the subject during radiographic imaging in a manner tailored to the position and/or shape of the subject to be imaged. Typically, in computed tomography (CT) imaging systems, an x-ray source emits a fan-shaped beam toward a subject or object, such as a patient or a piece of luggage. Hereinafter, the terms “subject” and “object” shall include anything capable of being imaged. The beam, after being attenuated by the subject, impinges upon an array of radiation detectors. The intensity of the attenuated beam radiation received at the detector array is typically dependent upon the attenuation of the x-ray beam by the subject. Each detector element of the detector array produces a separate electrical signal indicative of the attenuated beam received by each detector element. The electrical signals are transmitted to a data processing system for analysis and subsequent image reconstruction. There is an increasing desire to reduce radiation expose to a patient during radiographic data acquisition. It is generally well known that significant radiation or “dose” reduction may be achieved by using an attenuation filter to shape the intensity profile of an x-ray beam. Surface dose reductions may be as much as 50% using an attenuation filter. It is also generally known that radiation exposure for data acquisition from different anatomical regions of a patient may be optimized by using specifically shaped attenuation filters tailored to the anatomical region-of-interest (ROI). For example, scanning of the head or a small region of a patient may be optimized using a filter shape that is significantly different than a filter used during data acquisition from the heart. Therefore, it is desirable to have an imaging system with a large number of attenuation filter shapes available to best fit each patient and/or various anatomical ROIs. However, fashioning an imaging system with a sufficient number of attenuation filters to accommodate the numerous patient sizes and shapes that may be encountered can be impractical given the variances in a possible population. Additionally, manufacturing an imaging system with a multitude of attenuation filters would increase the overall manufacturing cost of the imaging system. Further, for optimum dose efficiency, i.e. best image quality at the lowest possible dose, the attenuation profile created by the attenuation filter should be particular to the patient. That is, it is desirable and preferred that when selecting a pre-patient or attenuation filter that it be adjusted according to the particulars of the patient, such as the patient's size, shape, and relative position in the bore of the scanner, be taken into account. By taking these and other particulars into consideration, radiation exposure can be optimized for the patient and the scan session. Known CT scanners use both an attenuation filter and dynamic current modulation to shape the intensity of the x-ray beam incident to the patient. To reduce radiation exposure, the attenuation filer is typically configured to minimize x-ray exposure to edges of the patient where path lengths are shorter and noise in the projection data has a less degrading impact on overall image quality. Accordingly, one such implementation of the attenuation filter is the bowtie filter, which, as a function of form, increases attenuation of x-ray intensity incident upon of the peripheral of the imaging subject. However, improper patient centering and/or bowtie filter selection can significantly degrade image quality and dose efficiency because x-ray attenuation is misapplied to the particulars of the subject. The bowtie filter is aligned with a point of maximum radiation dose or isocenter. The bowtie filter minimizes attenuation of x-ray intensity to isocenter and attenuates radiation significantly with radial distance beyond the center region of the bowtie, because, ideally, the isocenter corresponds to an imaging center of the subject. However, this is not always the case, e.g. when the subject is mis-centered in the scanner. FIG. 1 illustrates a bowtie filter ideally matched to a patient. Specifically, bowtie filter 10 is aligned within an imaging beam 12 such that an x-ray profile 14 is generated by the incidence of the imaging beam 12 upon the patient 16. However, if the patient 16 is not centered with respect to the bowtie center and the corresponding isocenter, significant image degradation can occur. The degradation is dependent upon a multitude of factors, such as the size of the central region of the bowtie filter, size and shape of the patient, and the amount and direction of patient mis-centering. FIG. 2 illustrates one such example of a bowtie filter opening that is improperly matched to the patient. That is, the bowtie filter 10 is aligned within the imaging beam 12 such that an improper x-ray profile 18 is generated by the incidence of the imaging beam 12 upon the patient 16. Specifically, photon incidence or flux at the edges of the patent may increase image noise to a level that may be prohibitively high for diagnostically valuable images. Recent improvements in imaging devices include a continuously adjustable bowtie filter having a pair of filtering elements to compensate for factors that may lead to non-ideal imaging. Such a filter is described in U.S. Ser. No. 10/605,789, the disclosure of which is incorporated herein and is assigned to GE Medical Systems Global Technology Co., LLC, which is also the Assignee of this application. Each filter element has a long low attenuating tail section that varies in attenuation power across its length such that as the elements are moved relative to one another, the attenuation of the beam is controlled. Each filter element is dynamically positioned with a dedicated motor assembly. The filter elements may be positioned in the x-ray beam so as to shape the profile of the x-ray beam to match a desired ROI or anatomical point-of-interest. The filter portions are positionable and adjustable using precision positioners to control the radiation pattern for the patient or the anatomy currently being imaged. However, image degradation may occur if the bowtie opening created is too small for a large patient since useful x-ray needed for imaging is attenuated by the bowtie thereby causing high image noise. As a result, the operator must manually determine the appropriate beam width and position according to size, shape, and positioning of the subject within the scanner bore. A properly sized bowtie configuration, however, does not ensure acceptable image quality. If the subject is mis-centered, image degradation may still persist. This degradation is typically a result of two factors. First, if subject mis-centering is caused by mis-elevation of the subject with respect to the bowtie filter then the calculation of tube current will result in an underestimate of the subject size. Referring to FIG. 3, a patient 16 is shown mis-centered in an x-ray beam 12. Specifically, the patient 16 is positioned at an improper centering elevation 20 by a centering error 22 below a proper centering elevation or y-position 24. As a result, a portion of the imaging beam 26 is not incident upon the patient 16 and a projection area 28 is understated by an error margin 30 because the patient 16 intercepts fewer rays in the imaging beam 12. As such, when determining tube current with the imaging tube at top-dead-center, as is convention, a lower tube current than actually required for proper imaging will be determined. As a result of the lower tube current, excessive image noise will be present relative to the user's selection. For example, a calculated milliamp (mA) that is 30% lower than actually required for proper imaging occurs for a typical 30 cm×20 cm body mis-centered in elevation by three cm. In such a case, noise introduction is increased by approximately 15% from a properly centered, properly imaged, patient. Secondly, patient mis-centering with respect to elevation may also position the thickest part of the patient such that x-rays for lateral projections pass through the thickest part of the bowtie which results in over-attenuation of the imaging beam. Referring to FIG. 4, the patient 16 is shown mis-centered within the imaging beam 12 by a centering error 22 below the proper centering elevation 24. As a result, the imaging beam 12 passes through the thickest parts of the bowtie filter 10 and patient 16, as exemplified by projection route 32. Such mis-centering can result in an additional image noise increase by as much as 70%. These errors can cause images of such high noise that the diagnostic value is compromised. Moreover, since traditional CT imaging methods rely on operator input to perfect patient centering, including elevation, elevational patient mis-centering can be common. Furthermore, traditional edge detection methods rely on identifying the center of the patient indirectly by detecting the edges of the patient, which can be particularly susceptible to error. Additionally, recent advancements in detector technology has increased the desire to control x-ray flux management to within very accurate constraints. For example, photon counting (PC) and energy discriminating (ED) detector CT systems have the potential to greatly increase the medical benefits of CT by differentiating materials such as a contrast agent in the blood and calcifications that may otherwise be indistinguishable in traditional CT systems. Additionally, PC and ED CT systems produce less image noise for the same dose than photon energy integrating detectors and hence can be more dose efficient than conventional CT systems. However, while PC and ED CT systems have the potential to realize numerous advantages over traditional CT detectors, the systems may be impractical for some scan protocols. Therefore, it would be desirable to design an apparatus and method to automatically control flux by dynamically filtering radiation emitted toward the subject during radiographic imaging in a manner tailored to the position and/or shape of the subject to be imaged so as to optimize radiation exposure during data acquisition. It would be further desirable to have a system that tailors the radiation emitted toward the subject during data acquisition based on a scout scan of the subject. Furthermore, it would be advantageous to have a system and method of controlling x-ray flux management to avoid photon pileup. Additionally, it would be desirable to have a system and method of dynamically adjusting radiation filtering to follow a user defined-region-of interest. It would also be desirable to have an apparatus to automatically collect patient centering and surface elevation information include a direct method of detecting patient centering. Furthermore, it would be desirable to have a method of accurately determining patient mis-centering within an imaging volume and adjusting the patient position to compensate for the determined mis-centering. The present invention is a directed method and apparatus to optimize radiation exposure that overcomes the aforementioned drawbacks. The present invention includes a method and apparatus for user-selection of a region-of-interest (ROI) and automatically controlling an x-ray dose received by the subject during imaging to be tailored to the position and/or shape of the ROI. In accordance with one aspect of the invention, a method of imaging is disclosed that includes positioning a subject in an imaging device, performing at least one scout scan, and marking a user-defined ROI. An attenuation characteristic of an attenuation filter is then automatically adjusted based on the user-defined ROI. In accordance with another aspect of the invention, a tomographic system is disclosed that includes a rotatable gantry having a bore centrally disposed therein, a table movable within the bore and configured to position a subject for tomographic data acquisition, and a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject. A detector array is disposed within the rotatable gantry and is configured to detect high frequency electromagnetic energy projected by the projection source and impinged by the subject. An attenuation filter is positioned between the high frequency electromagnetic energy projection source and the subject. A computer is included that is programmed to display a user interface including an illustration of a position of the subject and allow selection of a ROI and determine an attenuation profile of the attenuation filter based on the user-selected ROI. In accordance with another aspect of the invention, a computer readable storage medium having stored thereon a computer program representing a set of instructions is disclosed. The instructions, when executed by at least one processor, cause the at least one processor to perform at least one scout scan, display an interface including a reconstructed image from the at least one scout scan and receive user-selection identifying a ROI. The instructions then cause the at least one processor to adjust at least one of an attenuation filter configuration and a subject position based on the ROI. Various other features, objects and advantages of the present invention will be made apparent from the following detailed description and the drawings. The present invention is directed to a method and system that automatically determines the patient's size, shape, and centering within an imaging volume and dynamically controls x-ray flux accordingly. Preferably, one or two scout scans together with a plurality of sensors integrally formed with the CT scanner provide patient particulars. The present invention uses the information to provide centering information to the user, allow user input, automatically re-center the patient elevation, correct projection area measurements for dynamic tube current control and select the correct bowtie filter for the optimum dose efficiency. The operating environment of the present invention is described with respect to a four-slice computed tomography (CT) system. However, it will be appreciated by those skilled in the art that the present invention is equally applicable for use with single-slice or other multi-slice configurations. Moreover, the present invention will be described with respect to the detection and conversion of x-rays. However, one skilled in the art will further appreciate that the present invention is equally applicable for the detection and conversion of other high frequency electromagnetic energy. The present invention will be described with respect to “third generation” CT systems but is equally applicable with a wide variety of CT systems. That is, it is contemplated that the present invention may be utilized with energy integrating, photon counting (PC), and/or photon energy discriminating (ED) CT detector systems. Specifically, it should be recognized that the present invention provides a technique that controls and effectively limits detector saturation. The technique is adaptable such that it may be tailored to specific the requirements and constraints of a particular detector type and/or detector arrangement. For example, energy integrating detectors, which integrate the amount of x-ray flux recorded during an exposure time, have an inherent flux level tolerance that is relatively high. On the other hand, direct conversion detectors such as photon counting detectors, which actually count each photon as it passes, have a very different, typically lower, flux level tolerance. The present invention provides a dynamically adaptable technique whereby specific flux level tolerances may be observed to avoid detector saturation. Referring to FIGS. 5 and 6, a computed tomography (CT) imaging system 100 is shown as including a gantry 102 representative of a “third generation” CT scanner. Gantry 102 has an x-ray source 104 that projects a beam of x-rays 106 through a filter assembly 105 toward a detector array 108 on the opposite side of the gantry 102. Detector array 108 is formed by a plurality of detectors 110 which together sense the projected x-rays that pass through a medical patient 112. Each detector 110 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuated beam as it passes through the patient 112. Moreover, the detectors may be photon energy integrating detectors, photon counting, and photon energy discriminating detectors. During a scan to acquire x-ray projection data, gantry 102 and the components mounted thereon rotate about a center of rotation 114. Rotation of gantry 102 and the operation of x-ray source 104 are governed by a control mechanism 116 of CT system 100. Control mechanism 116 includes an x-ray controller 118 that provides power and timing signals to an x-ray source 104, a gantry motor controller 120 that controls the rotational speed and position of gantry 102, and filter assembly controller 123 that controls filter assembly 105. A data acquisition system (DAS) 122 in control mechanism 116 samples analog data from detectors 110 and converts the data to digital signals for subsequent processing. An image reconstructor 124 receives sampled and digitized x-ray data from DAS 122 and performs high speed reconstruction. The reconstructed image is applied as an input to a computer 126 which stores the image in a mass storage device 128. Computer 126 also receives commands and scanning parameters from an operator via console 130 that has a user interface device. An associated cathode ray tube display 132 allows the operator to observe the reconstructed image and other data from computer 126. The operator supplied commands and parameters are used by computer 126 to provide control signals and information to DAS 122, x-ray controller 118 and gantry motor controller 120. In addition, computer 126 operates a table motor/table centering controller 134 which controls a motorized table 136 to position patient 112 and gantry 102. Particularly, table motor/table centering controller 134 adjusts table 136 to move portions of patient 112 through and center patient 112 in a gantry opening 138. Sensors 140 are positioned within gantry opening 138 to collect patient position and contour data. Sensors 140 are connected to a sensor controller 142 that controls the operation of sensors 140 and provides the acquired data to computer 126 to be processed. As shown in FIGS. 7 and 8, an x-ray system 150 incorporating the present invention is shown. The x-ray system 150 includes an oil pump 152, an anode end 154, and a cathode end 156. A central enclosure 158 is provided and positioned between the anode end 154 and the cathode end 156. Housed within the central enclosure 158 is an x-ray generating device or x-ray tube 160. A fluid chamber 162 is provided and housed within a lead lined casing 164. Fluid chamber 162 is typically filled with coolant 166 that will be used to dissipate heat within the x-ray generating device 160. Coolant 166 is typically a dielectric oil, but other coolants including air may be implemented. Oil pump 152 circulates the coolant through the x-ray system 150 to cool the x-ray generating device 160 and to insulate casing 164 from high electrical charges found within vacuum vessel 168. To cool the coolant to proper temperatures, a radiator 170 is provided and positioned at one side of the central enclosure 158. Additionally, fans 172, 174 may be mounted near the radiator 170 to provide cooling air flow over the radiator 170 as the dielectric oil circulates therethrough. Electrical connections are provided in anode receptacle 176 and cathode receptacle 178 that allow electrons 179 to flow through the x-ray system 150. Casing 164 is typically formed of an aluminum-based material and lined with lead to prevent stray x-ray emissions. A stator 170 is also provided adjacent to vacuum vessel 168 and within the casing 164. A window 182 is provided that allows for x-ray emissions created within the system 150 to exit the system and be projected toward an object, such as, a medical patient for diagnostic imaging. Typically, window 182 is formed in casing 164. Casing 164 is designed such that most generated x-rays 184 are blocked from emission except through window 182. X-ray system 150 includes an attenuation filter assembly 186 designed to control an attenuation profile of x-rays 184. As stated, the present invention provides a means to determine patient particulars such as patient size, shape, and centering from one or two scout scans. The information is used to provide centering information to the user, allow user selection of a ROI, automatically center the patient elevation, correct projection area measurements for dynamic tube current control, and select the correct bowtie filter configuration for the optimum dose efficiency. The methods include automatic selection of the proper bowtie filter opening to control the impact of the bowtie filter and patient mis-centering on tube current or x-ray flux modulation. Referring now to FIG. 9, a flow chart setting forth the steps of an imaging technique in accordance with the present invention is shown. The technique is particularly tailored for dynamic bowtie and tube current control. The technique begins at 200 with the performance of at least one scout scan and/or sensing a patient elevational profile 202 to determine a required tube current modulation 204 in the x, y and z-directions for a desired image noise assuming a properly centered patient. As will be described in detail, the scout scan(s) may be a lateral scout scan or an anterior-posterior (AP) or posterior-anterior (PA) scout scan. Depending on the orientation of the available scout scan(s), a starting CT scan angle, a location in the z-direction, and positions for the left and right filter segments of the continuously variable bowtie filter, such as that described in commonly assigned patent application U.S. Ser. No. 10/605,789, are selected 206. The starting bowtie (attenuation) filter positions are determined 206 independently for each side, as will be described with respect to FIG. 10. Once the starting bowtie filter positions have been set 206, scanning begins 208. Bowtie position information is collected and included for each projection during the scan to allow the bowtie attenuation profile to be properly normalized during image reconstruction. Bowtie positioning repeatability is preferably maintained within ten micrometers to allow dynamic calibration and correction of the moving bowtie during patient scanning. That is, during the scan 208 the information from the scout scan(s) and/or sensed patient elevational profile 202 is/are used to adjust operating parameters. Specifically, a maximum edge x-ray flux is sensed 210 and a closed loop feedback system is utilized to determine whether such is within a select range 212. If it is determined that the maximum edge x-ray flux is outside the selected range 214, the bowtie filter is adjusted to maintain the maximum edge x-ray flux 216. That is, as the maximum flux at the edge of the imaging object increases, an associated filter segment of the bowtie filter is moved toward isocenter. Conversely, if the flux at the edge relative to the center flux is below the selected range, an associated filter segment of the bowtie filter is moved away from isocenter. However, if it is determined that the maximum edge x-ray flux is inside the selected range 218, the bowtie filter is not adjusted and sensing of the maximum edge x-ray flux continues. At the same time, a mean x-ray flux rate at the central portion of the imaging subject is sensed 220 and a determination of whether the mean x-ray flux rate at the central portion of the imaging subject is outside a selected range is made at 222. If the mean x-ray flux rate at the central portion of the imaging subject is outside the selected range 224, the tube current (mA) is adjusted to maintain the desired mean x-ray flux rate in the central projection region of the imaging subject 226. On the other hand, if the mean flux rate at the central portion of the imaging subject is within the selected range 228, no change to tube current is made and sensing 220 continues. However, since mA modulation influences the edge flux, it is contemplated that the control of edge flux levels may be done relative to the average central flux level. As such, in accordance with one embodiment of the present invention, the adjustment of bowtie filter toward isocenter 216 and the adjustment of tube current 226 are based on an interdependent consideration of both the sensed maximum flux at the edge of the imaging object 210 and the sensed mean flux rate at the central portion of the imaging object 210. Furthermore, as will be described, it is contemplated to use a priori positioning for dynamic bowtie positioning with the feedback loop and to use filter positioning moves to prevent photon pileup only when absolute flux limitations are at risk thereby also compensating for the fact that mA modulation influences the edge flux. In this way, the bowtie filter can be positioned for optimum dose efficiency based on imaging subject size, shape, and centering as a first priority whereby positioning for prevention of photon pileup during scanning has precedence. However, it is contemplated that for situations where the central projection region may have the highest x-ray flux, such as for AP projections when scanning legs, for example, adjusting the mA to avoid photon pileup in the center of the projections may be given priority over the tube current modulation objectives. As will be shown, reliable patient size and centering determinations can be made from projections using two orthogonal scout scans or a single scout scan. However, as will be shown, the present invention includes systems to compensate for the absence of a second scout scan by accurately sensing patient elevational contours. The present invention also includes a method of improved calculations of subject center whereby the centroid (center of mass) is determined from two orthogonal scout projections or estimated from a single scout scan. The method of FIG. 9 may be utilized to maintain the x-ray flux rates below an absolute maximum limit of a detector. That is, x-ray rates for some detector configurations may be significantly lower than other detectors. Hence, flux rates must be carefully managed to avoid count rate saturation (photon pileup). Since patient attenuation, projection centering error, and desired flux rate levels are known; x-ray flux rates can be controlled by appropriate filter positioning and tube current adjustments using expressions representing the fundamental x-ray physics attenuation and absorption equations. Referring now to FIG. 10, an example of a patient mis-centering to the left is illustrated 230. That is, the patient 232 is mis-centered with respect to the isocenter 234 of the x-ray flux passing through a bowtie filter configuration 236. The bowtie filter configuration 236 includes a left bowtie portion 238 and a right bowtie portion 240 that are dynamically adjustable by left control motor 242 and right control motor 244. The bowtie configuration and the tube current are controlled such that a central patient region 246 is within a desired object flux 248 according to a desired image noise. Furthermore, the left bowtie portion 238 is positioned to maintain a left patient edge region flux profile 250 below a max flux limit 252. Similarly, the right bowtie portion 240 is positioned to maintain a right patient edge region flux profile 254 below the max flux limit 252. Accordingly, a filter and patient x-ray flux profile 256 has a relative flux level 258 below the max flux limit 252. As such, flux rates are maintained to remain under the x-ray rates required for specific detector configurations, thereby avoiding photon pileup. Specifically, the system may be used to overcome limitations, such as photon pileup, which is commonly encountered with the use of photon counting (PC) and photon energy discriminating detectors (ED) CT as opposed to traditional photon energy integrating CT detectors. Photon counting CT systems include detector systems that are capable of distinguishing between photons such that a photon is differentiated from another photon and counted as it is received by the detector. Energy discriminating CT systems are capable of tagging each photon count with its associated energy level. As will be described in detail below, the present invention provides a means to determine an imaging subject's size, shape, and centering and to use this information to provide centering information for automatically re-center patient elevation. Accordingly, as shown in FIG. 10, x-ray flux management may be controlled to maintain a flux profile 250, 254 that is below a max flux limit 252 of a specific detector and its respective flux limits. For example, the flux profile 250, 254 may be specifically controlled to satisfy the requirements of ED or PC CT detectors so as to avoid photon pileup. Referring now to FIG. 11, FIG. 11 provides a detailed method for adjusting pre-imaging and imaging parameters. Specifically, two scout scans are performed 300 that include an AP scout scan and a lateral scout scan. From the two scout scans a centroid projection 302 is made. Specifically, the distance of the centroid from a point of reference is made. In a preferred embodiment, the point of reference is isocenter of the x-ray fan beam and the distance of the centroid from isocenter is determined. However, it is also contemplated that the point of reference may be the center of the medical imaging device or the center of the bore of the medical imaging device, or any other stationary point that is readily identifiable. Additionally, it is contemplated that the point of reference may be a map of an ideally positioned imaging subject with similar physical features. In any case, the distance of the centroid from the point of reference is used to geometrically calculate an x and y centering error for the patient relative to a reference position 304. In accordance with a preferred embodiment of the present invention, the reference position is at a center of the scanning bay located in the y-direction. However, it is also contemplated that the reference position may be arbitrarily selected as long as the reference position is fixed with respect to patient position within the CT bore. Having calculated the y-axis patient centering error over the extent of the prescribed CT scan, the system determines the mean center with respect to the reference position, to provide the optimum fixed table height for the duration of the CT scan. The x and y mis-centering is then compared to a threshold 305. Accordingly, a direct determination of the center of the imaging subject is made. That is, by utilizing centroid calculations the center of maximum attenuation that should be positioned in the maximum x-ray field is determined rather than the physical center relative to the edges of the object. If the mis-centering is less than the threshold 306, indicating that the current position of the patient is within the imaging tolerance of the system, no adjustment is necessary and the system is ready for scanning 307. However, if the mis-centering is greater than the threshold 308, the operator is notified of the centering error 309 and presented with an auto-correction prompt whereby the operator is prompted to accept or reject 310 the table elevation change. However, it is contemplated that operator approval may be bypassed whereby auto-correction is completed without operator approval 310. As such, a fully automated correction system may be implemented. As will be described with respect to FIGS. 11 and 12, should the operator reject the auto-correction 311, the operator may use a graphical indication or other means to enter a user-selected centering correction 312 according to which scanning is performed 313. Should the operator accept the auto-correction 314, the patient elevation is automatically corrected 316. Additionally, the x and y centering errors are used to correct the projection area (PA) 318. The PA is the sum of the attenuation values of the x-rays that intercept the patient. Therefore, PA is dependent on the distance of the patient from the fan beam x-ray source. By utilizing the accuracy of the centroid calculated x and y centering errors, a corrected PA is directly calculated using geometric equations 318. The PA from both the AP and lateral scouts can be corrected using the centering error determined from the orthogonal scout scan and the average AP scout scan. That is, lateral PA can be used to improve the accuracy of a tube current modulation noise prediction algorithm 322. Additionally, the oval ratio (OR) is directly computed 320 using the projection measure (PM) ratio from the two orthogonal scouts to further improve the accuracy of the tube current modulation noise prediction 322. That is, the tube current is then boosted to compensate for the centroid calculated mis-centering 324. Once adjustments according to the centroid calculated mis-centering are complete 316–324, the proper bowtie filter configuration is selected 326. Specifically, for a given bowtie filter shape and a given patient size and shape there exists an optimum opening, measured in flat width (FW), that provides the best image quality at the lowest dose. The optimum value is the value of FW that maximizes a quality factor Q as calculated as follows: Q = KC ⁡ ( a , b , FW ) N ⁡ ( a , b , FW ) ⁢ D ⁡ ( a , b , FW ) ;where; N is the overall noise in the image or scan data (standard deviation); D is the dose to the object; C is the contrast between two materials such as iodine and water (dependent on the spectral characteristics of the system); K interpolates linearly between Q = 1 N 2 ⁢ D ⁢ ⁢ ⁢ and ⁢ ⁢ ⁢ Q = C N 2 ⁢ D ; a and b are the axes parameters for an ellipse; and FW is one half of the flat width (i.e. ½ the length of the uniform low attenuation region of the bowtie filter in mm). In accordance with an alternative embodiment, quality factor may be determined using a single diameter parameter d, where d is the average of a and b. In either case, once the proper bowtie filter configuration is selected 226, the system is ready for scanning 328. As such, the patient table is raised or lowered dynamically during the execution of a helical CT scan to accommodate the changing optimum elevations depending on patient anatomy and centering/mis-centering. Elevation data is included in the scan data header to properly position the views during image reconstruction. If a continuous bowtie is present, the bowtie is positioned dynamically to follow the sineogram of the patient. That is, an attenuation pattern may be utilized that maps a dynamic configuration of the attenuation of the bowtie so as to achieve desired attenuation over time, i.e. during data acquisition. Referring now to FIG. 12, the optimum bowtie filter opening can be determined experimentally by constructing various phantom sizes and shapes and then scanning the phantoms with various bowtie filters having different FW values, reconstructing images, measuring the noise, dose, and contrast for each case, and fitting a curve to the Q values vs. FW as shown in FIG. 13. The optimum FW value for a given patient size can then be determined by reviewing FW value against the Q value. Specifically, the FW value where Q is at a maximum is the optimum FW value, as illustrated in FIG. 13. The Q values can also be determined by computer modeling using fundamental x-ray physics attenuation and absorption equations to estimate the noise, contrast, and dose in the image for each case. The contrast weighting value K can be chosen between 0 and 1. In the given example, the value of K is zero in order to exclude any benefits of improved object contrast. From experimental data or simulations the set of optimum bowtie opening values can be determined versus patient or object size as shown in FIG. 12. The relationship is approximately linear and can be represented by the equation FW=0.45 (d−10) for the K=0 assumption where d is the patient diameter in centimeters. Patients with diameters less than 10 cm would use a bowtie opening FW value of zero. As such, optimum bowtie opening FW can be accurately selected given the patient diameter d. The patient diameter can be determined from the PM (amplitude of projection) and a patient density assumption μ. The average PM can be obtained from the orthogonal scout scan pair since d=avg(PM/μ). For the human body, the density assumption μ can be assumed to be 0.2, which is the attenuation coefficient of water, except for the chest and head. For the chest and head, μ can be approximated as 0.14 and 0.24, respectively, due to the density decrease of the lungs and the density increase of the skull. For CT systems with a continuously variable addressable bowtie, the FW value can be determined directly by the equation, d=avg(PM/μ). On CT systems without an addressable continuous bowtie, the equation can be used to select the nearest optimum bowtie from the selection of available discrete bowtie filters. For example a set of discrete bowtie filters that covers the patient range from infants to large obese adults would typically include bowtie filters with openings having FW values of 1, 5, 9, and a flat filter. From the graph on FIG. 12, the following lookup table can be constructed to automatically select the most optimum discrete bowtie for the patient as follows: DIAMETER:<15 cm15 to 25 cm>25 to 35 cm>35 cmBOWTIE FILTER:FW 1FW 5FW 9Flat The optimum filter opening, however, is dependent on how well the patient is centered in addition to the patient's diameter. The effect of patient mis-centering is comparable to a patient radius increase for the projections perpendicular to the mis-centering axis. Hence the proper filter selection is a function of the patient diameter plus the mis-centering and can be determined using the equation, FW=0.45 (d−10+2ew), where e is the patient mis-centering error in centimeters and w is a weighting factor or function. The weighting factor is typically 1.0 but could be less than 1.0 to constrain the dose increase that would otherwise result when the bowtie is opened to fully account for the worst case effect of mis-centering. The value of w could also be a function of the object size, shape, and mis-centering to more closely match the behavior of image noise with mis-centering of various size objects. A discrete bowtie selection can also be obtained by adding the centering error factor (2ew) to the phantom diameter for the lookup table index. For example, from the table, a 24 cm patient with a 3 cm error would be considered a 30 cm diameter and hence, filter FW 9 should be selected instead of FW 5 for the centered case. Furthermore, in the event that tube current modulation is used and the patient is mis-centered in a smaller than optimum bowtie, the mA can be boosted to avoid an unacceptable noise increase in the image. Referring now to FIG. 14, an example of the user interface through which manual entry of a user-selected centering correction 312, FIG. 11, may be entered is shown. In the given example, the user is performing a spine study. In this case, a spine study is optimally centered on the spine 410 instead of the overall attenuation centroid for the patient 412. However, the automatically calculated adjustments will be based upon the mean center of the patient over the scan length and yield the overall attenuation centroid of the patient as the center point 412. Accordingly, the automatically calculated adjustments based on the centroid calculations to compensate for mis-centering are not optimal for the spine study and the operator will choose to manually enter a user-selected centering correction such that recentering is along the mean center of the spine over the scan length 414. Referring to FIG. 15, another view of the example of the user interface through which manual entry of a user-selected centering correction may be entered on a pair of scout scans is shown. Through the interface, the user marks the location of the spine or other area of interest on scout scans using cursor markers 416. Via the user-defined cursor markers 416, a diameter of interest 418 is defined that includes a center of interest 420 independent of the centroid calculated isocenter 422. Accordingly, the patient table may be raised or lowered dynamically during the execution of a CT scan to accommodate the changing optimum elevations depending on patient anatomy to track the user-defined cursor markers. Elevation data is included in the scan data header to properly position the views during image reconstruction. If a continuous bowtie filter is present, the bowtie filter may be controlled dynamically to follow the sineogram of the patient. That is, if the location of a ROI is designated 418 via markers 416, the bowtie filter is dynamically positioned to follow the sineogram of the ROI. This positioning obtains improved image quality for the ROI and reduces dose elsewhere. FIG. 16 illustrates an implementation of the method illustrated in FIG. 11 when only a lateral patient scout scan 510 is available. From the lateral scout scan 510, a centroid projection is made at 512 and y mis-centering is determined relative to a reference position 514 according to the methods previously described. However, since no AP scout scan data is available, x mis-centering is assumed to be zero. The assumption that x mis-centering is 0 provides a reasonable estimation as long as the operator utilizes the edges of the patient table as a guide when positioning the patient in x. Then, having determined the y axis patient centering error over the extent of the prescribed CT scan, the system determines the mean center to provide the optimum fixed table height for the duration of the CT scan. The y mis-centering is then compared to a threshold 516. If the mis-centering is less than the threshold 518, indicating that the current position of the patient is within the imaging tolerance of the system, no adjustment is necessary and the system is ready for scanning 520. However, if the mis-centering is greater than the threshold 522, the operator is notified of the centering error 524 and presented with an auto-correction prompt whereby the operator is prompted to accept or reject 526 the table elevation change. However, it is contemplated that operator approval may be bypassed whereby auto-correction is completed without operator approval 526. As was described with respect to FIGS. 11 and 12, should the operator reject the auto-correction 528, the operator may use a graphical indication or other means to enter a user-selected centering correction 530 according to which scanning is performed 532. Should the operator accept the auto-correction 534, the patient elevation is automatically corrected 536. The PA, PM, and OR are calculated 538–540 from the single scout using known methods utilizing y mis-centering calculations. However, since only one PM is available from the single scout scan, the diameter for bowtie selection is determined by the equation, d=(PM/μ)(OR+1)/2 because the OR, by definition, is the ratio of the axis parameters of the elliptical patient model. As such, mA modulation is calculated 542, the mA boost factor is implemented 544, and the appropriate bowtie is selected 546 based on the y-axis centering information using the methods previously described herein. Accordingly, scanning is performed at 548. FIG. 17 illustrates an implementation of the method illustrated in FIG. 11 when only an AP patient scout is available. Fundamentally, the method shown in FIG. 17 is substantially similar to that of FIG. 16; however, the y-axis centering error can not be directly determined since it is in the same orientation as the scout projections. Nevertheless, an estimate of the y-axis error relative to a reference position can be made if elevation information relative to the surface of the patient is available. Once the AP scout scan is complete 610, the system determines whether the surface elevation of the patient is known 612. If the surface elevation of the patient is unknown 614, the operator is prompted to manually select a bowtie filter configuration and calculate tube current per traditional manual methods 616 and a scan is performed 617. However, if the surface elevation of the patient is known or derived 618, as will be described with respect to FIGS. 18–20, an estimation of y mis-centering is performed 620. The estimation of y mis-centering is then compared to a threshold 622. If the mis-centering is less than the threshold 624, indicating that the current position of the patient is within the imaging tolerance of the system, no adjustment is necessary and the system is ready for scanning 626. However, if the mis-centering is greater than the threshold 628, the operator is notified of the centering error 630 and presented with an auto-correction prompt whereby the operator is prompted to accept or reject 632 the table elevation change. However, it is contemplated that operator approval may be bypassed whereby auto-correction is completed without prior operator approval 632. As was described with respect to FIGS. 11 and 12, should the operator reject the auto-correction 634, the operator uses a graphical indication or other means to enter a user-selected centering correction 636 according to which scanning is performed 638. Should the operator accept the auto-correction 640, the patient elevation is automatically corrected 642. The PA is then corrected 644 for the estimated y-axis centering error. This is done by direct geometric calculations or as a fitted function of elevation, PA, and OR, as will be described with respect to FIG. 18–20. As such, mA modulation is determined 646, the mA boost factor is implemented 648, and the appropriate bowtie is selected 650 based on the y-axis centering information using the methods previously described herein. Accordingly, the system is ready for scanning 652. However, it is also contemplated that estimations for PA, PM and mis-centering may be generated from the surface contour of the patient. As such, it is possible to determine mA modulation, boost mA to compensate for patient mis-centering, and select a desired bowtie configuration without the benefit of scout scans. That is, for the selection of the bowtie filter configuration, it is assumed that the patient is centered and the bowtie configuration is selected based on patient size estimated from the PM and density assumption of μ as previously described herein. Referring to FIGS. 18, 19, and 20, surface elevation information about the patient can be obtained by various methods. If the patient 706 is resting directly on the patient table, as in FIG. 18, the table elevation can be used to determine y-axis centering error. Specifically, with respect to FIG. 18, the table height 708 is known and, as such, the upper horizontal axis 710 of the patient 706 is known or reasonably estimated. Therefore, once the vertical axis 712 is determined, as described above, the upper center 714 of the patient 706 can be determined from the intersection of the upper horizontal axis 710 of the patient 706 and the vertical axis 712. Accordingly, the center 716 of the patient 706 is disposed halfway between the upper center 714 and the table height 708. Given the determination of these values, table elevation relative to isocenter (E) can be calculated by solving for the equation E=R+H−C, wherein H is the height of the table 714, R is the difference between the center 716 of the patient 706 and the table height 708, and C is the height of the upper center 714 of the patient 706. Specifically, mis-centering is determined by measuring the offset of the contour projections from isocenter. However, in cases where the patient 706 is propped up, as in FIG. 19, with pillows or other positioning devices 718, the centering can be determined from a laser or sonic displacement measuring device positioned on the gantry or otherwise disposed on the scanner to locate the top surface of the patient 706. As such, a vector of position information is collected and associated with each scout projection to allow the centering error to be calculated as a function of the z-direction. Specifically, since the center 716 of the patient 706 cannot readily be readily discerned because it is not disposed halfway between the upper center 714 and the table height 708 due to the offset created by the positioning device 718, as shown in FIG. 18, a laser or sonic displacement sensor 720 may be utilized to determine a distance L to the upper horizontal axis 710 of the patient 706. As such, E can be calculated in this case according to: E=C−R−L. However, referring to FIG. 20, it is also contemplated that a plurality of lasers and/or sonic displacement 720 sensors may be utilized to measure the distance from an array of points to obtain the specific contour of the patient 706. As such, an improved accuracy determination of overall patient contour is achieved. In any case, the PA can be determined from the external patient contour and the μ for the associated anatomy as described previously herein. The OR is determined directly from the distance measurements or from the PM which can be determined from the μ and patient surface distances. Mis-centering is determined by measuring offset of the contour projections from isocenter. Once the contour of the patient is known, it is possible to calculate the projection error ratio and fit it to a cubic or other function of elevation, PA, and OR to determine equation coefficients in order to calculate the PA corrected for y-axis centering error according to the following: PA = P / C1 + ( C2 * E ) + ( C3 * P ) + ( C4 * 0 ) + ( C5 * E * P ) + ( C6 * E * O ) + ( C7 * P * O ) + ( C8 * E 2 ) + ( C9 * P 2 ) + ( C10 * O 2 ) + ) ⁢ C11 * E * P * O ) + ( C12 * E 2 * P ) + ( C13 * E 2 * O ) + ( C14 * P 2 * E ) + ( C15 * P 2 * O ) + ( C16 * O 2 * P ) + ( C17 * O 2 * P ) + ( C18 * E 2 ) ;wherein: Eq coeffVariableC1constantC2elevationC3PAC4OVRC5elevation * PAC6elevation * OVRC7PA * OVRC8elevation2C7PA2C10OVR2C11elevation * PA * OVRC12elevation2 * PAC13elevation2 * OVRC14PA2 * elevationC15PA2 * OVRC16OVR2 * elevationC17OVR2 * PAC18OVR3andE is the table elevation relative to isocenter;P is the measured projection area;PA is the projection area corrected for table/patient elevation; andO is the oval ratio. It is contemplated that the above-described invention be utilized with “third generation” CT systems as well as a wide variety of other CT-type systems. That is, it is contemplated that the present invention may be utilized with energy integrating, PC, and ED CT detector systems. Furthermore, it is contemplate that the above-described invention may be utilized with non-traditional and non-medical CT applications. For example, it is contemplated that the above-described invention may be utilized with a non-invasive package/baggage inspection system, such as the system shown in FIG. 21. Referring now to FIG. 21, package/baggage inspection system 800 includes a rotatable gantry 810 having an opening 812 therein through which packages or pieces of baggage may pass. The rotatable gantry 810 houses a high frequency electromagnetic energy source 814 aligned with an attenuation filter 815 as well as a detector assembly 816. A conveyor system 818 is also provided and includes a conveyor belt 820 supported by structure 822 to automatically and continuously pass packages or baggage pieces 824 through opening 812 to be scanned. Objects 824 are fed through opening 812 by conveyor belt 820, imaging data is then acquired, and the conveyor belt 820 removes the packages 824 from opening 812 in a controlled and continuous manner. As a result, postal inspectors, baggage handlers, and other security personnel may non-invasively inspect the contents of packages 824 for explosives, knives, guns, contraband, and the like. Therefore, in accordance with one embodiment of the current invention, a method of diagnostic imaging is disclosed that includes determining a position of a subject in a scanning bay relative to a reference position, automatically adjusting an attenuation characteristic of an attenuation filter based on the determined position of the subject and imaging the subject. In accordance with another embodiment of the invention, a computer readable storage medium is disclosed that has stored thereon a computer program representing a set of instructions. When the instructions are executed by at least one processor, the at least one processor is caused to receive feedback regarding mis-centering of a subject to be scanned, determine a value of mis-centering of the subject to be scanned, and adjust at least one of an attenuation filter configuration and a subject position based on the value of mis-centering. The processor is then caused to acquire radiographic diagnostic data from the subject. In accordance with still another embodiment of the invention, a tomographic system is disclosed. The tomographic system includes a rotatable gantry having a bore centrally disposed therein, a table movable within the bore and configured to position a subject for tomographic data acquisition within the bore, and a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject. A detector array is disposed within the rotatable gantry and configured to detect high frequency electromagnetic energy projected by the projection source and impinged by the subject and an attenuation filter positioned between the high frequency electromagnetic energy projection source and the subject. A computer is programmed to adjust at least one of an attenuation characteristic of the attenuation filter and a table position based on a specific position of the subject in the bore. In accordance with yet another embodiment of the invention, a method of centering a subject in a medical imaging device is disclosed that includes positioning a subject in a scanning bay, comparing a center of mass of the subject to a reference point, and repositioning the subject in the scanning bay to reduce a difference in position between the center of mass of the subject and the reference point. In accordance with another embodiment of the invention, a computer readable storage medium having stored thereon a computer program representing a set of instructions is disclosed. The instructions, when executed by at least one processor, causes the at least one processor to determine a centroid of a subject, determine a value of mis-centering of the centroid of the subject within a medical imaging device, and adjust a position of the subject within the imaging device to compensate for the value of mis-centering. In accordance with yet another embodiment of the invention, a method of medical imaging is disclosed that includes positioning a subject in a medical imaging device, determining a value of mis-elevation of the subject, and adjusting an elevation of the subject device to reduce the value of mis-elevation. In accordance with still another embodiment of the invention, a tomographic system is disclosed that includes a rotatable gantry having a bore centrally disposed therein, a table movable within the bore and configured to position a subject for tomographic data acquisition within the bore, and a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject. The tomographic system also includes a detector array disposed within the rotatable gantry and configured to detect high frequency electromagnetic energy projected by the projection source and impinged by the subject and computer. The computer is programmed to determine a centroid of the subject and adjust an elevation of the subject to align the centroid with a reference position. In accordance with one embodiment of the invention, a method of imaging is disclosed that includes positioning a subject in an imaging device, performing at least one scout scan, and marking a user-defined region-of-interest (ROI). An attenuation characteristic of an attenuation filter is then automatically adjusted based on the user-defined ROI. In accordance with another embodiment of the invention, a tomographic system is disclosed that includes a rotatable gantry having a bore centrally disposed therein, a table movable within the bore and configured to position a subject for tomographic data acquisition, and a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject. A detector array is disposed within the rotatable gantry and is configured to detect high frequency electromagnetic energy projected by the projection source and impinged by the subject. An attenuation filter is positioned between the high frequency electromagnetic energy projection source and the subject. A computer is included that is programmed to display a user interface including an illustration of a position of the subject and allow selection of a ROI and determine an attenuation profile of the attenuation filter based on the user-selected ROI. In accordance with another embodiment of the invention, a computer readable storage medium having stored thereon a computer program representing a set of instructions is disclosed. The instructions, when executed by at least one processor, cause the at least one processor to perform at least one scout scan, display an interface including a reconstructed image from the at least one scout scan and receive user-selection identifying a ROI. The instructions then cause the at least one processor to adjust at least one of an attenuation filter configuration and a subject position based on the ROI. In accordance with yet another embodiment of the invention, a tomographic system is disclosed that includes a rotatable gantry having a bore centrally disposed therein, a table movable within the bore and configured to position a subject for tomographic data acquisition within the bore, and a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject. A detector array is disposed within the rotatable gantry and configured to detect high frequency electromagnetic energy projected by the projection source and impinged by the subject and at least one sensor is included to provide subject position feedback. In accordance with another embodiment of the invention, a computer readable storage medium is disclosed having stored thereon a computer program representing a set of instructions. When the instructions are executed by at least one processor, the at least one processor is caused to receive feedback regarding a subject position from at least one sensor of an imaging device and determine a centering error from the feedback. In accordance with one more embodiment of the invention, a method of imaging is disclosed that includes positioning a subject in an imaging device, collecting positioning information of the subject from at least one sensor disposed in proximity of the imaging device, and determining a relative position of the subject within the imaging device from at least the position information. The present invention has been described in terms of the preferred embodiment, and it is recognized that equivalents, alternatives, and modifications, aside from those expressly stated, are possible and within the scope of the appending claims.
052805107	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a schematic representation of a system 10 for implementing the present invention. A vacuum chamber 12 is connected to a vacuum pump 14 by which the ambient air can be removed. The tubular component 16 to be coated, is supported within the vacuum chamber. A source material rod 20 is positioned, preferably axially, within the component 16, in spaced relation from the confronting inside surface 18. The rod 20 includes a distribution of field emitter structure having a generally nodular, and preferably pointed shape, which will hereafter be referred to as emitter tips 22. The rod 20 must, of necessity, have a rather small effective outside diameter, especially when used for coating the inside of nuclear fuel rod tubing, which is generally on the order of 0.5 inch. The emitter can be formed as a wire on which the desired source material for the coating, has been flame sprayed. Techniques for fabricating the emitter rod per se, do not form a part of the present invention, but rather can be adapted for use in the present invention, by reference to available literature on conventional field emitter technology. In some instances, the rod 20 can be of a type having gated field emitters (the gates 24 are shown in FIG. 1 as a dashed line). FIG. 2 shows a schematic of a gated field emitter rod 100, wherein the rod core 102 has nodular structure such as 108 formed thereon, preferably including a point 110. An insulating material 104 is layered on the core material 102, and a top layer of conductive material 106 covers the insulating material 104. A basin, or bowl is formed by the absence of insulating and conductor material 104, 106 in a region immediately surrounding the emitter structure 108. As an example, the distance from the point 110 to the insulating layer 106 is on the order of the height or projection of the point 110 from the core 102, such distance typically being in the range of about 1-10 .mu.m. In the conventional utilization of a gated field emitter such as shown in FIG. 2, the emitter tip or nodule 108 is biased negatively with respect to its gate 106. This structure will emit electrons 112 if the electrical field at the emitter point 110 is large enough to induce field emission from the tip material. Above a critical current, a gated field emitter structure 108 will exhibit a failure which will cause the point 110 and then a portion of the rest of the tip 108 to be evaporated. This process has been investigated by J. Browning, N. E. McGruer, S. Meassick, C. Chan, W. Bintz and M. Gilmor, in an article entitled "Gated Field Emitter Failures: Experiment and Theory", to be published in IEEE Transactions, Plasma Science, October, 1992. As shown in FIG. 1, the tips 22 are negatively biased via tip power supply 28 and conductor 26, whereas the gates 24 are positively biased via power supply 34 and conductors 30,36. The component substrate can be negatively biased via conductor 32. With non-gated field emitters, which are preferred for the present invention, the tips 22 are biased negatively via conductor 26 and tip power supply 28, whereas the workpiece, i.e., the tubular component 16, would be positively biased via conductors 30,36 and workpiece bias power supply 34. Above a critical current failure of the emitter tip 22 will occur, causing a portion of the tip to be evaporated as microparticles. Because of the small size of the field emitter structures, i.e., less than 10 .mu.m feature size, arrays of field emitters can be used to deposit material on large work pieces while single emitters can be utilized for deposition of material on small workpieces. The size of the emitter tips (volume of material available in each tip), surface density of emitter tips, and the energy available for each tip failure can be used to adjust the amount of material evaporated and, therefore, the thickness of a deposited coating on a workpiece. Arrays of field emitters can be fabricated on sources of arbitrarily complex shape and used for deposition on complexly shaped work pieces. In particular, emitters can be fabricated on a small wire appropriately sized for use on the inside of fuel rods, instrument thimble tubes, burnable poison rods, control rods, or control rod guide tubes. The rate of failures, energy available for each failure, and the surface density of emitters can be used to adjust the rate of deposition, coating thickness and coating parameters. A large fraction of the material evaporated from the field emitter is ionized, allowing a potential gradient between the emitter tip and work piece to accelerate the ions thereby changing the properties of the deposited coating, increasing adherance, decreasing crystal size. The material of the deposited coating is determined by the composition of the emitter tips. Thus, the composition of the deposited coating can be changed by adjusting the composition of individual emitter tips or by making individual emitter tips of different material. The tips 22 thus need not be of the same material as the carrying core of rod 20 and all tips need not be of the same material. As represented at 38 in FIG. 10, the atmosphere within the vacuum chamber 12 can be backfilled with a reactive gas such as nitrogen or oxygen, or carbon vapor plasmas can be introduced, whereby the source material evaporated from the tips 22 chemically reacts with the nitrogen, oxygen and/or carbon, before adhering to the component surface 18. In any event, the tip material deposits on the component substrate surface 18, either in the form as originated on the emitter, or in a form as modified by reaction with the reactive gas in the chamber. The ability to achieve different coatings along different longitudinal regions of the component, with the ability to layer one component upon another, as described in said copending applications, is another advantage achievable with the present invention. Thus, where desirable, neutron absorber poison material can be coated with and in combination with wear resistant or hydrogen getter material. The following tables list particularly desirable coatings that are achievable with the present invention. Table 1 lists a variety of burnable poison metals that can be deposited in accordance with the present invention (BN is a burnable poison coating that can be formed with the use of nitrogen as the reaction gas introduced via reaction gas source 38): TABLE 1 ______________________________________ Burnable Poison Metals and Metallic Compounds ______________________________________ Gadolinium Erbium Boron ZrB.sub.2 BN TiB.sub.2 ______________________________________ Table 2 is a representative list of burnable poison ceramic materials including glasses that are usable with the present invention: TABLE 2 ______________________________________ Burnable Poison Ceramics and Glasses ______________________________________ 20 Li.sub.2 080B.sub.3 15 Na.sub.2 085B 20 B.sub.4 C ______________________________________ Table 3 is a representative list of getter material that can be deposited in accordance with the present invention: TABLE 3 ______________________________________ Getter Material ______________________________________ Yttrium Zirconium-Nickel alloys Zirconium-Titanium-Nickel alloys ______________________________________ Table 4 is a representative list of wear and/or corrosion resistant metal compounds. TABLE 4 ______________________________________ Metals and Metallic Compounds ______________________________________ ZrN TiN CrN HfN TaAlVN TaN ______________________________________ Table 5 is a representative list of wear and/or corrosion resistant ceramic materials including glasses that are usable with the present invention: TABLE 5 ______________________________________ Ceramics and Glasses ______________________________________ Zr.sub.2 O.sub.3 Al.sub.2 O.sub.3 TiCN TiC CrC ZrC WC Calcium Magnesium aluminosilicate Sodium Borosilicate Calcium Zinc borate ______________________________________
	abstract	A scanning charged particle microscope which facilitates adjustment, has a deep focal depth, and is provided with an aberration correction means. The state of aberration correction is judged from a SEM image by using a stop having plural openings and the judgment result is fed back to the adjustment of the aberration correction means. A stop of a nearly orbicular zone shape is used in combination with the aberration correction means.
	claims	1. A system for bringing a nuclear power plant into a safe state after extreme effect, the system comprising:a steam generator;a separation tank;a storage tank;a pump;a heat exchanger;a control unit configured to control the system;a first water valve;a second water valve;a first air valve;an inlet pipeline from the steam generator to the separation tank;two pipelines comprising a first pipeline and a second pipeline, the first pipeline configured to provide a path for water in the separation tank to be transferred to the storage tank, the second pipeline configured to provide a path for gas in the separation tank to be transferred to the storage tank;a discharge pipeline from the storage tank to the pump; andan outlet pipeline from the storage tank to the pump and from the pump to the steam generator;the separation tank located above the steam generator and connected by the two pipelines to the storage tank; andwherein the heat exchanger is installed in the outlet pipeline downstream of the pump, the first water valve is installed in the inlet pipeline, and the separation tank is connected with the storage tank by the first pipeline with the second water valve installed therein and the second pipeline with the first air valve installed therein. 2. The system according to claim 1, wherein a deaerator configured to remove steam from the system is used as the storage tank. 3. The system according to claim 1, wherein the system comprises a vertical steam discharge pipeline with a second air valve; and the steam generator is equipped with the vertical steam discharge pipeline with the second air valve. 4. The system according to claim 1, wherein the system comprises a plurality of steam generators connected in parallel to each other to the inlet and outlet pipelines. 5. The system according to claim 1, wherein at least a part of the inlet pipeline is configured to have an upward slope towards the separation tank.
	description	This application claims the benefit of U.S. Provisional Application No. 61/588,550, filed Jan. 19, 2012, the disclosure of which is hereby expressly incorporated by reference in its entirety herein. Part of the operation of a nuclear power plant is the removal and disposal of irradiated nuclear fuel assemblies. Most early reactors were originally built to store from three to five years' capacity of irradiated fuel assemblies in a storage pool. From the storage pool, the irradiated fuel assemblies could be reprocessed or sent to long-term storage. However, as a result of uncertainties in the federal policies relating to reprocessing of irradiated fuel and also in the establishment of permanent irradiated fuel dumps, on-site irradiated fuel storage facilities have been stressed to their capacity for storing these irradiated fuel assemblies. To prevent the forced shutdown of nuclear power plants as a result of the overcrowding of storage pools, a number of near-term irradiated fuel storage concepts have been developed and/or utilized. One such near-term concept in use is the dry storage of irradiated fuel. Nonetheless, early developments in irradiated fuel dry storage in the United States anticipated that this would be a short-term measure, with removal of irradiated fuel to more permanent geologic storage required by Federal Law starting in 1998. As it became apparent that this would not happen, and that interim dry storage would be a larger scale and longer term effort, the following change occurred in the demands placed on dry storage systems. As the initial inventory of low burnup, long cooled irradiated fuel residing in pools was transferred to dry storage, and as power plants increased the enrichment and burnup of their fuel, the need to store fuel with ever greater residual decay heat has grown. The fuel gives off heat from the decay of the radioactive elements, and so the storage system must be able to keep the fuel cladding cool enough that it does not deteriorate during the dry storage period without the use of active coolers such as fans. Early systems were developed with a capability for about 24 kW of decay heat per system; current needs are in excess of 40 kW. Various structures have been developed to transport and store the irradiated fuel in secure canisters. One type of canister uses a lattice structure to form compartments to locate the fuel within the transport and storage canisters. The lattice structure is of “egg crate” design composed of interlocking transverse plates. However, existing baskets of egg crate design have used very expensive materials. Such materials include, for example, borated stainless steel, extruded profiles of enriched boron aluminum and metal matrix composites. Thus, a need exists of constructing transport and storage canisters from lower cost and more common materials. A system was developed for horizontal modular dry irradiated fuel storage, as described in U.S. Pat. No. 4,780,269, the disclosure of which is hereby expressly incorporated by reference. However, there exists a need for improvements to that system. Embodiments of the present disclosure described herein are directed to fulfilling this and other needs. This summary introduces a selection of concepts in a simplified form that are further described below in the Detailed Description. This summary is not intended to identify key features of the claimed subject matter, nor is it intended to be used as an aid in determining the scope of the claimed subject matter. A system for transportation and storage of spent nuclear fuel disclosed herein enhances the internal heat transfer during storage by the efficient use of high strength steel to construct the storage canisters, allowing more room for conductive material, aluminum or copper. The rejection of heat external to the fuel storage canister is enhanced by the mechanical application of fins to the canister outside cylindrical surface, or by the use of efficient and effective heat shields between the canister and the concrete storage module, including designs that increase the surface area for convective heat rejection from the heat shields, in comparison to conventional heat shields. The present disclosure employs lower cost materials employed in a novel way to construct an “egg crate” type transport and storage canisters for the irradiated fuel. The plates of the egg crate structure employ low alloy, high strength steel plates that encase aluminum and a thin sheet of a metallic base neutron absorbing material forming a functionally graded structure where the steel provides structural stability, the aluminum conducts heat, and the neutron absorber prevents a neutron chain reaction. In accordance with further aspects of the present disclosure, a canister for the transportation and storage of nuclear fuel assemblies includes a basket assembly receivable into a canister shell. The basket assembly includes a plurality of interlocking structural plates that are disposed in spaced parallel relationship to each other in a first direction, as well as a plurality of structural plates disposed in a second direction transverse to the first direction. The structural plates including transverse slots formed along the plates so that the slots of the structural plates disposed in a first direction engage with the slots of the structural plates disposed in the second, transverse direction. The structural plates are composed of a plurality of separate layers, including outer layers composed of a structural material, at least one inner layer composed of heat conducting material, and at least one inner layer composed of neutron-absorbing material. In accordance with further aspects of the present disclosure, the outer layers of the structural plates are formed to encase the inner layers of the structural plates. In this regard, the margins of the outer layers extend over the edges of the inner layers and are joined to each other. In a further aspect of the present disclosure, elongated locking keys extend along and engage with adjacent edge portions of adjacent structural plates to lock the adjacent edge portions together and to align the adjacent edge portions together. In this regard, grooves are formed along the edge portions of the structural plates. These grooves are sized to closely receive the locking key therein. Also, holes are formed in the structural plates, whereby the locking key passes through the holes of the structural plates that extend transversely to the length of the locking keys. In accordance with a further aspect of the present invention, transition rails extend lengthwise of the canister at the outer perimeter of the basket assembly to interconnect the structural plates. The transition rails have an outer curvature in the direction transverse to the length of transition rails that correspond to the circumference of the canister. In addition, the transition rails are at least partially hollow to receive a stiffening structure therein to enhance the structural integrity and rigidity of the transition rails. In accordance with a further aspect of the present disclosure, storage modules for containing nuclear fuel assemblies in storage canisters include concrete bottom, side, and top walls. The modules are configured for air flow therethrough by natural convection to dissipate the heat emitted from the nuclear fuel assemblies. At least one heat transfer structure is disposed within the module and positioned to transfer heat from the canister to the air flowing through the module. In addition, at least one heat shield is disposed in the module to shield the interior of the module from heat emitted from the nuclear fuel assemblies. In accordance with another aspect of the present disclosure, the concrete top, bottom and side walls of the module are composed of a mixture of concrete and metallic fibers serving to reinforce the concrete. In accordance with a further aspect of the present disclosure, the heat transfer structure includes fins that are disposed within the module for transferring heat from the canister to the air flowing through the module. The fins are placed into contact with the canister once the canister is positioned within the module. In accordance with a further aspect of the present disclosure, a heat shield structure and/or heat transfer structure includes barriers that extend along one or more of the side walls, top wall and bottom wall of the module. Such barrier is spaced from the module walls to provide an air flow interface between the barrier and the module walls. The heat barrier structure is selected from the group consisting of plate structures, corrugated wall structures, and tubular wall structures. In accordance with another aspect of the present disclosure, a single port tool is provided for the canister for fluid flow communication with the interior of the canister for draining water from the canister and replacing the draining water with make-up gas. The single port tool includes a single opening formed in the canister and a shield plug insertable within the opening. The single sport tool is in fluid flow communication with the interior of the canister and has a first passageway therethrough for receiving a drain tube for draining the water from the canister, and a second opening therethrough for receiving make-up gas and directing such make-up gas to the interior of the canister. In a further aspect of the present disclosure, the canister has a housing and a cover, and a single port is formed in the cover of the canister for reception of the single port tool. A horizontal modular dry irradiated (e.g., spent) fuel storage system will first be described. With reference to FIGS. 1-4, a horizontal modular dry irradiated fuel storage system 10 in accordance with embodiments of the present disclosure is presented. In addition, processes for storing irradiated fuel are described. As set forth in greater detail below, the systems and processes described herein are improvements to previous systems and processes described in U.S. Pat. No. 4,780,269 (as seen in FIGS. 1-4), the disclosure of which is hereby expressly incorporated by reference. Referring to FIG. 1, the system 10 uses a specially designed dry shielded canister assembly 12, which is shown in greater detail in FIGS. 5-10, as described in greater detail below. The canister assembly 11 is inserted into a transfer cask 14. The transfer cask 14 and canister assembly 11 can be placed by a crane 16 into an irradiated fuel storage pool 18 filled with water (see FIGS. 1 and 2). Irradiated fuel contained in fuel assemblies (see, e.g., fuel assembly 20) can be stored in the pool 18. To remove the irradiated fuel from the pool 18, the fuel is placed in the canister assembly 11, and appropriate seals and covers (as described below) are affixed to the canister assembly 11 before the transfer cask 14 is removed from the pool 18. Referring to FIG. 2, upon removal from the pool 18, water is forced out of both the canister assembly 11 and the transfer cask 14 with a pressurized gas being applied through selected ports of the canister assembly and cask. The canister assembly 11 is further dried by using a vacuum pump to evacuate the residual water from the canister assembly 11. After evacuation of the canister assembly 11, helium gas is pumped into the canister assembly 11. As the transfer cask 14 (containing the canister assembly 11 and irradiation fuel assemblies 20) is removed from the pool 18, appropriate radiation shielding is provided for the contained irradiated fuel assemblies by the shielded end plugs of the canister assembly 11 and the transfer cask 14. Referring now to FIG. 3, the transfer cask 14 can be loaded into a horizontal position onto a transfer trailer 22 having a specially designed skid 24. The skid 24 allows the transfer cask 14 to be moved in three dimensions to permit alignment of the cask 14 with a horizontal storage module 25, which can be seen in FIG. 4, for dry storage of the canister assembly 11. Referring to FIG. 4, the cask 14 is aligned with a port 28 in the dry storage module 25 to extract the canister assembly 11 from the transfer cask 14 for storage in the horizontal storage module 25. In the illustrated embodiment, a hydraulic ram 30 is at least partially insertable through a second port 32 at the opposite end of the dry storage module 26 to extract the canister assembly 11 from the transfer cask 14 for storage in the horizontal storage module 25. Alternatively, a winch (not shown) or another extraction device could be used in place of ram 30 to extract the canister assembly 11 from the transfer cask 14. It should further be appreciated that the reverse operation of pushing the canister assembly 11 into the dry storage module 25 can also be accomplished. Referring to FIGS. 5-10, detailed views of an improved horizontal dry storage module 26 are provided. The horizontal dry storage module 26 includes a housing 40 having a top section 41. The housing 40 is in block or rectilinear form and is preferably constructed from reinforced concrete, which may be positioned on a load-bearing foundation 42 (see, e.g., FIG. 4). In a previous design, the housing 40 was formed from concrete reinforced with rebar. However, in the improved design, the housing 40 is reinforced with metal fiber, for example, steel fiber, to increase blast and earthquake resistance and provide long-term crack resistance. The metal fiber also reduces shrinkage and cracking of the concrete in the short term, thereby decreasing water incursion and also increasing spalling resistance in the long term. In sum, the use of steel or other comparable fibers to reinforce the concrete increases the toughness, tensile strength, density, and dynamic strength of the concrete. It should be appreciated that vertical storage modules or other storage modules (not shown), having housings that are reinforced with metal fiber, for example, steel fiber, are also within the scope of the present disclosure. Also, it is to be appreciated that the use of metal fiber to reinforce the concrete can be used in lieu of or in addition to primary and secondary rebars used in standard concrete construction. Also, it is to be appreciated that other high-strength fibers can be used in place of or in addition to metal fibers, such as fiberglass fibers, glass fibers, or carbon fibers. The housing 40 includes an inlet 44 at one end and an interior volume 46 designed for receiving and containing a canister assembly 12. Embedded in housing 40 is an underlying support assembly 48 to support the canister assembly 12 when it is fully inserted into housing 40. The support assembly 48 may also be configured to allow the canister assembly 12 to slide easily in and out of the housing 40. As shown in FIGS. 5-7, the support assembly includes parallel slide rail assemblies 49 extending along the housing in the lower portion of the interior volume 46. The slide rails may include slide strips composed of material that is galvonically compatible with the canister 12 and durable under the radiation level and temperature within interior volume 46. The slide rails themselves may be composed of such material or a coating or surface treatment may be applied to the slide rails. The housing 40 includes a closure device 50 to cover the inlet 44. The closure device 50 may be constructed from steel and/or concrete and/or other appropriate radiation protection media. The closure device includes an inner, round-shaped cover plug 54 and an outer hat plate 52 that is sized to overlap the front wall of the housing surrounding the inlet 44. The wet plug 54 closely fits within inlet 44. As can be seen in FIGS. 5 and 6, the closure device 50 can be appropriately positioned in place when a canister assembly 12 is disposed in the module 26. Referring to FIG. 7, the housing 40 may be designed and configured to allow similar housings 40 to be placed adjacent other housings, which may be interlocked therewith. Therefore, several housings can be stacked together in series to provide additional shielding to minimize radiation leakage. Referring to FIGS. 6 and 7, the horizontal dry storage module 26 may include a heat dissipation assembly 60. In the illustrated embodiment, the heat dissipation assembly 60 includes a plurality of curved, relatively thin fins 62 spaced along the module 26. The fins 62 are either lowered onto, or clamped like a clamshell onto, the outer surface of the canister assembly 12 after the assembly is installed in the module 26. The fins 62 enhance convective heat transfer from the canister surface to the air flowing through the module 26. As can be seen in FIG. 6, in one embodiment of the present disclosure, the heat dissipation assembly 60 consists of a series of curved fins that are mounted to the underside of a longitudinal bar 64. The bar in turn depends from a series of rods 66 that extend through the top section 41 of the housing 40 to terminate at threaded upper end portions that engage threaded fasteners 66. The heat dissipation assembly is initially retracted, positioned at the top or roof of the module 26 by rotation of fasteners 66 on rods 64. Once the canister assembly 12 has been inserted within module 26, the fasteners 46 are used to lower the bar 64 and associated fins 62 down onto the upper surface of the canister. Although in FIG. 6, the upper threaded ends of bars 64 are shown as protruding above the upper surface of housing top section 41, instead the upper end of the rods 64, as well as threaded fastener 66, may be disposed below the top surface of top section 41. In this regard, wells or sockets may be formed in the upper surface of the top section 41, so that once the assembly 60 has been deployed downwardly against canister assembly 12, the wells or sockets can be plugged or otherwise securely closed off. Rather than being constructed as shown in FIGS. 6 and 7, the heat dissipation assembly 60 may be constructed in two separate sections, with each section hinged to the interior of housing 40, for example, along the lower side portions of the housing. Once canister assembly 12 has been installed in the module 26, such hinged fin sections could be rotated to bear against the exterior of canister assembly 12 in a clamshell-like arrangement. Rather than constructing the heat dissipation assembly 60 as a movable unit, the assembly can be formed from stationary fins, for example, fins 62′ or 62″ shown in FIGS. 8A and 8B. Flexible, heat transmitting interfaces 68 or 69 are provided along the edges of fins 62′ and 62″ that face the canister 12. In FIG. 8A, the interface 64 is in the form of a hollow, bulbous shape that can be deformed when the canister 12 is slid into module 26. As shown in FIG. 8A, the interface forms an oval or elliptical shape when pressed against the exterior of canister assembly 12. In FIG. 8B, the interface is in the form of a flexible lip assembly 69 that flexes and presses against the exterior of canister assembly 12 when the canister is slid into place within module 26. As noted above, both of these interfaces are highly heat conductive. The fins 62 may be constructed from aluminum or any other suitable metal or non-metal material designed for heat conduction and collection. Referring to FIG. 9, one or more access ports 92 may be provided in the front wall of the module 26 and on plate 50 for inspection of the module 26 interior space 46 and the surface of the canister assembly 12 during long term service, off-normal events, etc. As seen in FIG. 9, the ports 92 maybe closed off by suitable shielding plugs 94. The ports 92 may be various configurations and at various locations on the front wall and plate 52. Referring to FIG. 10, tubular heat shields 96 are positioned in the interior space 46 of the module to increase the surface area for transferring radiant heat from the canister 12 the air flowing through the module 26 relative to if the shield was composed of a flat plate, while at the same time protecting the housing 40 (which is made from reinforced concrete) from excessive heat. The heat shields 96 can be composed of standard square or rectangular cross-sectional metallic tubes, for example, steel or aluminum, or other heat conducting material. The individual tubes can be secured adjacent to each other by welding, mechanical fastening, or other expedient means. A mechanical fastener could include rods that extend transversely through the tubes. Alternatively, transverse tie rods could extend transversely over the exterior of the tubes, with the tie rods welded or otherwise fastened to the tubes. Also, the surface of the shields 96 that face canister 12 can be treated to increase their radiant emissivity, and thus increase their ability to absorb or otherwise capture infrared heat from canister assembly 12. The tubular shields 96 are mounted to the interior side walls and top walls of the housing 40 by suitable brackets thereby to space the shields from the adjacent walls of the housing 40. This provides a relatively cool layer of air between the shield and the concrete wall of the canister assembly 12 thereby to protect the concrete from excessive heat, which of course can weaken the structural integrity of the concrete. Rather than using heat shields 96 of tubular construction, the heat shields can be of other constructions, including one or more substantially flat plates or a plate of corrugated construction, with the corrugation taking many cross-sectional shapes, such as semi-circular, rectilinear, triangular, etc. Also of course these alternative constructions for heat shields 96 can be composed of various materials having different levels of radiant heat absorption and heat conduction. As noted above, the heat shields 96 enhance the overall heat rejection capability of canister assembly 12 by increasing the surface area for heat rejection. The heat shield is heated both by radiation and by air flowing from the canister to the near surface of the shield by natural convention. The tubular or corrugated heat shields increase the surface area compared to a flat heat shield, thus making the heat shield more effective for transferring heat to the cooler air which flows inside the tubes composing the shields 96, as well as the air that flows between the tubes and concrete wall of the housing 40. This directly increases the surface area available for transferring heat away from the canister 12. Also, the tubes that compose shield 96 provide two separate shielding surfaces, one facing the canister and one facing the concrete wall, thereby increasing the ability of the shield 96 to serve as a heat barrier and protecting the concrete walls of the housing 40 from being overheated. Referring to FIG. 11, a basket assembly 70 for being disposed in the canister 12 to hold fuel assemblies 20 will now be described in greater detail. The basket assembly 70 is in the form of a rack positioned internal to the canister assembly 12 for locating and supporting the fuel assemblies during storage and transportation. Referring to FIGS. 11, 12 and 12A, the basket assembly 70 has a structure composed of functionally graded plates 72 that interlock in a criss-cross or “egg crate” matrix to define a plurality of tubes 74 (square or rectilinear in cross-section) for receiving individual fuel assemblies. The plates 72 are formed in a plurality of layers for structure, heat transfer, and neutron absorption as described more fully below. Referring specifically to FIGS. 13-17, the plates 72 may include a multi-layer structure. As a non-limiting example, the plates 72 may have a four-layer structure including first and second steel outer layers 80 and 82, a heat conductor interior layer 84, and a neutron absorber layer 86. As a non-limiting example, the steel outer layers 80 and 82 may be a high strength, low alloy steel, a high-strength steel, a carbon steel, stainless steel or other comparable materials. As a non-limiting example, the heat conductor layer 84 may be manufactured from aluminum or copper or other highly heat conductive metal or material. As a non-limiting example, the neutron absorber layer 86 may be manufactured from a material whether metallic, ceramic or a composite, that contains an element that absorbs thermal neutrons. Such materials include, but are not limited to, boron, cadmium, and gadolinium. As such, the layer 86 may be composed of a metal matrix composition, such as a composite of fine boron carbide particles in an aluminum or aluminum alloy matrix. The aluminum matrix may consist of 99% pure aluminum. Also, it is to be understood that the heat conducting function and neutron absorption function can be combined into a single layer of material that can both conduct heat and absorb neutrons. Such materials can include but cannot be limited to, aluminum or copper with embedded particles of boron carbide. The plates 72 may include flush fasteners 76 for securing the layers of the plate to each other in face-to-face relationship, see FIG. 18. Suitable fasteners 76 may include, for example, threaded fasteners, rivets or welded joins. In the illustrated embodiments of FIGS. 14 and 15, holes 88 for receiving fasteners 76 may be formed by punching, drilling or other methods in the plates 72. Referring to FIG. 18, an exemplary threaded torque limiting fastener 76, which is flush with the exterior surfaces of the plate 72, on both sides, is shown. The fastener has a bolt section 76A composed of a beveled head 76B and a shank 76C. The threaded section 76D engages with the interior of a threaded nut 76E, which also has a beveled head 76F. The beveled heads 76B and 76F bottom against beveled counter bores formed in the layers 80 and 82. Then the fastener 76 is fully engaged the heads 76B and 76F of the fastener are flush with or beneath the outer surfaces of plate layers 80 and 82. In one embodiment of the present disclosure, the layers of the plates 72 are furnace-brazed together. Exemplary constructions for the furnace-blazed plates 72 are shown in FIGS. 16 and 17. Referring to FIG. 16, the edges of layers 80 and 82 are bent around and over each other in overlapping fashion at 89A for adding buckling resistance. Referring to FIG. 17, the bent edges of layers 80 and 82 are welded to each other along the butt seam 89B to form a rigid tubular structure with the other components (layers 84 and 86) encased inside the tube. In one embodiment of the present disclosure, the plates 72 may include a black oxide coating on one or both steel layers 80 and 82 to provide improved radiation heat transfer from the fuel assemblies (not shown) to the basket assembly 70. In addition, the outer surfaces of the plates 72 may further include a hydrophobic silicon dioxide coating to improve water shedding and thereby reduce drying time. The plates 72 can be constructed in different thicknesses and widths. The thicknesses of the plates can depend on various factors, including the weight of the fuel being transported and stored, the amount of heat conduction desired by layer 84 as well as the level of neutron absorption desired for a layer 86. The widths of the plate 72 can depend on the overall length of the basket assembly 70, since such length is composed of plates 72 stacked lengthwise upon each other. As a non-limiting example, the plate 70 can range in width from about 10 inches to about 16 inches or even wider. The basket assembly 70 shown in FIG. 11 is composed of plates 72 that are fitted together in criss-cross or “egg crate” manner. Also referring to FIGS. 11, 12, and 12A, the plates 72 have transverse slots 73 that extend a quarter of the way across the width of the plate. As a consequence, when the plates 72 are fitted together so that the slots 73 of the criss-crossing plates engage each other, adjacent plates in the vertical direction mate edgewise against each other. In this manner, a plurality of vertical cells 74 are formed for the full height of the basket assembly 70. Ideally, each of the cells 74 are only slightly larger in cross-section than the nuclear fuel assemblies that are contained or stored in the basket assembly 70. As will be appreciated at the very top and bottom of the basket assembly 70, the plates 72 are only half as wide as throughout the remaining height of the basket assembly. Moreover, the slots 73 in the upper most and lower most plates 72 extend half-way through the width of such plates. As a consequence, the bottom edges of all the lower most criss-crossing plates are on the same plane. Likewise, at the top of basket assembly 70, the upper edges of the upper most criss-crossing plates 72 are also of the same elevation. Referring specifically to FIG. 12A, as an optional construction of basket assembly 70, the longitudinal edges of the plates 72 are formed with a groove 74 extending along the upper and lower edges of each of the plates 72. The groove is sized to receive a close fitting bar or key 75 that is sized to be very closely receivable within the opposing grooves 74 of adjacent plates 72. The rod or bar 74 passes through openings 75A formed in the plates 72 in alignment with the two opposing slots 73 of a plate 72 and half way between such opposing slots 73. As will be appreciated by the foregoing construction, the bars 72 lock the adjacent edge portions of adjacent plates 72 together to form a very rigid construction for the basket assembly 70. The width of the groove 74 can be the thickness of the plate inner layers 84 and 86. As such, the groove 74 is formed by extending the outer layers 80 and 82 beyond the edges of the inner layers 84 and 86. Referring to FIGS. 19A and 19B, transition rails 90 and 92 may be designed for placement along the outer perimeter of the basket assembly 70 to help form the cylindrical outer structural shape of the basket assembly 90 when received in a canister assembly 12, see FIG. 11. In that regard, the rails 90 and 92 may be configured as cast or extruded aluminum alloy rails to provide strength and creep resistance to the basket assembly under long term exposure to the fuel assemblies at high temperature. The transition rail 90, shown in FIG. 19A, is generally triangular in cross section and having an outer curved side or surface 91 of a transverse curvature corresponding to the overall outer curvature of basket assembly 70, shown in FIG. 11. To provide structural integrity to the rail 90, an interior bracket or brace 91A may be utilized. As illustrated in FIG. 19A, the brace 91A is shown in the form of a rectangular tubular member. Through-holes 91B are formed in brace 91A in alignment with corresponding holes formed in the adjacent wall 91C of transition rail 90 through which appropriate fasteners may be engaged. Such fasteners, shown in FIG. 11, also extend through the adjacent plates 72 of the basket assembly. It will be appreciated that this construction aids in creating the basket 70 as a very rigid structure. Other than the cross sectional area taken out by the walls of brace 91A, the interior of transition rail 90 is hollow to minimize the weight of the rail and also to allow air to pass therethrough to aid in heat dissipation. As shown in FIG. 11, two sets of rails 90 are used in each quadrant of the basket 70. Two sets of transition rails 92 are also used in each quadrant of basket 70. The transition rails 92 are thinner in cross section than rails 90, but do include a curved outer surface 93 of a transverse curvature corresponding to the outer diameter of basket 70. The rails 92 include a longitudinal opening 93A for reception of a reinforcing tube 93B extending lengthwise through the rail. The reinforcing tube 93B is provided to help stiffen the rail 92. Of course, reinforcing members of other shapes can be used in place of tube 93B. Also, through cavities 93C and D extend lengthwise through the rail 92. These cavities help reduce the weight of the transition rail without significantly reducing the structural integrity of the rail. Moreover, air is able to flow through the cavities 93C and 93D, extending the length of the transition rail 92, thereby to help dissipate the heat generated from the fuel assemblies 20 disposed within the basket 70. The transition rail 92 is secured to adjacent plates 82 by fasteners 93E that extend through aligned openings formed in the plate 72 and in rails 92, see FIG. 11. Also, when in place, groove 93F formed in the interior wall section of transition rail 92 mate with the end portions of plates 72 that protrude beyond the furthest outward cross plate 72, for instance, as shown in FIG. 11. This interlocking relationship with the ends of the plate 72 also add to the rigidity of the construction of the basket 70. Returning to FIG. 7, a canister assembly 12 is shown in a module 26. Referring now to the cross-sectional view of FIG. 20, the canister assembly 12 is a substantially cylindrical container having an outer shell 96 and a distal end 98 and is designed for containing the basket assembly 70 for storage and transportation of fuel. The canister assembly 12 further includes a closure assembly 100 at its proximal end, as described in greater detail below. Most lightweight reactor fuel is in the range of about 146 to 201 inches in length. As such, the canister assembly 12 is constructed at a length corresponding to the length of the reactor fuel. As discussed above with reference to FIG. 2, the canister assembly 12 must be dried after it has been removed from the pool 18. In that regard, water must drained from both the canister assembly 12 and the transfer cask 14 that surrounds the canister assembly 12. See, for example, FIG. 2. Referring to FIGS. 21-23, in accordance with one embodiment of the present disclosure, a canister assembly 12 has been designed with an end closure assembly 100 that includes a shield plug 102 and an inner top cover plate 104 outward of the shield plug and a single integrated vent and drain port tool 106. The shield plug and inner top cover plate 104 close off the proximal end of outer shell 96. The shield plug is relatively thin and can be composed of material to contain the nuclear fuel assemblies within the canister assembly. Such materials may include, for example, steel, lead, tungsten and depleted uranium. The integrated port tool 106 has the capability to drain water and also provide an inert gas (e.g., helium) cover for fuel assemblies 20. Therefore, the canister assembly 12 includes means for controlling the gas that enters the interior of the canister assembly 12 while the water is being pumped out. The port tool 106 of FIGS. 21-23 may be configured as an adapter to replace separate drain and vent ports that are conventionally used in existing canister assemblies. In the illustrated embodiment of FIGS. 21-23, the port tool 106 generally includes an adaptor body 108 extending through inner top cover plate 104 and into the shield plug 102. The port assembly 106 also includes a vent 110 in communication with the interior of canister assembly 12 for gas supply into the canister assembly and a water removal tube 112, extending through a central passageway formed in body 108, for water to exit from the canister assembly 12. The vent 110 is composed of an outer nipple 111 that is connected to a vent passageway 114 formed in the adapter body 108. In FIG. 22, the vent 110 is shown as consisting of a tube 111C that extends through a vent passageway extending through the adapter body 108. In use, gas is supplied to the vent 110. Water may be pumped via tube 112 or forced out at tube 112 by gas pressure applied at vent 110. The adaptor body 108 of the port tool 106 is attachable to the inner top cover plate 104 by any suitable means, including threading, a bayonet lock, screw flange, or quick thread from the top. An exemplary threaded attachment 114 is shown in the illustrated embodiment of FIG. 22. In addition, a plurality of elastomeric x-rings 115A and o-rings 115B ensure a tight seal between the port assembly 106 and the inner top cover plate 104. O-rings 115C also are disposed between water removal tube 112 and the passageway extending through adaptor body 108. A port is formed by a cup 116 welded under the inner top cover plate 104. The cup 116 has a center hole 118 for receiving the water removal tube 112. The hole 118 has a diameter that is sized slightly larger than the outer diameter of the water removal tube 112 to provide an annular flow path for the backfill gas entering from vent 110. The water removal tube 112 may be a removable drain tube, that extends the length of the canister assembly 112. A top view of the port tool 106 in the canister assembly 112 can be seen in FIG. 23. The port is located at the perimeter of the basket, as can be seen in hidden view through outer cover 104, as viewed from the top of FIG. 23. The port and port tool 106 provide advantages over existing drain ports. These advantages include reduced manufacturing costs by the ability to provide a deep port in a relatively thin lower plate rather than the thick cover plates or vent and drain block commonly used. Moreover, the port assembly of the present disclosure reduces operation time and dose by reducing the number of ports that need to be closed (from two to one), and by the use of the thick adaptor body 108 that acts as a radiation shield. By sliding the tube 112 in the adaptor body 108, the gap between the bottom of the tube and the bottom end of the canister can be adjusted to optimize the removal of aspirated droplets, thus optimizing the removal of all water from the canister assembly 12. In addition, because the tube 112 is entirely removed during vacuum drying as shown in FIG. 24, the large opening improves the conductance for vacuum drying, which also reduces drying time, optimizing the removal of all water from the canister assembly 12. In addition, the port assembly 106 improves the conductance for vacuum drying, which also reduces drying time. Fuel loading operations will now be described. After fuel has been loaded into the canister assembly 12 (see, for example, FIG. 1), the shield plug 102 (shown in FIGS. 21 and 22) is installed while the canister assembly 12 and surrounding cask assembly 14 still remains under water. Rotational orientation of the canister assembly within the cast assembly is controlled by a key on the side wall of the canister assembly 12. The shield plug 102 does not engage a drain tube. A short hose is inserted into the canister assembly 112 to drain the water as needed from the canister assembly 12 and the inner top 104 cover is installed after the cask assembly 14 has been set down. The inner top cover 104 is then welded, and the drain tube 112 and port tool 106 are then installed. After the drain tube 112 and port tool 106 have been installed, the drain tube 112 is pushed to the bottom of the canister 112, then raised up about ⅜ inch (10 mm) and secured with a locking collar, not shown. The inert gas (e.g., helium) supply is attached to the vent tube 110, and the water pump is attached to the water drain tube 112. Gas flow and water pumping is initiated. In that regard, the gas pressure under the port assembly 106 should be slightly positive. At the first sign of cavitation (air in the water pump), the drain tube 112 is lowered and pumping is continued until water is no longer being pumped out. The water pump is them disconnected from the drain tube 112 and a vacuum pump with a water trap is attached to the drain tube 112. Gas continues to be supplied through the vent tube 110 while gas and water are removed from the canister assembly 12 by the vacuum pump. The drain tube 112 can be raised and lowered slightly during vacuuming to find the ideal gap between the drain tube 112 and the bottom of the canister assembly 12. Referring now to FIG. 26, another embodiment of a port assembly 306 for a canister assembly 12 is shown. In the embodiment of FIG. 26, the port assembly 306 includes a permanent tube 322 in the canister assembly 12 with a cup 340. A short removable tube 312 is connectable to the permanent tube 322 for drain operation, and removed for vacuum drying. The tube section 312 is comparable to the upper end of tube 112, discussed above. The permanent tube 322 in the canister assembly 12 connects with the threaded cup 340. Alternatively, the cup can be permanently affixed to tube section 312. The cup 340 can move enough to self align. The permanent tube 322 can also move up and down, but does not rotate and it does not engage the shield plug or any other lid component other than the tube 312 that is part of the port tool 106. The system described herein can be used to provide a solution to the problem of storage of irradiated fuel assemblies. The system is particularly appropriate for use as an interim solution to the irradiated fuel storage problem until provided by governmental authorities. Accordingly, the present disclosure provides for a relatively inexpensive temporary storage facility for irradiated fuel assemblies. The system uses and reuses existing casks to transfer canisters with the irradiated fuel assemblies to modules 26 for near-term storage. Further there is no requirement for a lifting crane at the storage site, because horizontal loading and unloading is enabled. In addition, the fuel canisters 12 can be comprised of a thin-walled material, because the canister is always protected either by the module 26 or by the transfer cask 14. In view of the use of existing technology and equipment, investment in horizontal dry storage module 26 can be spread over a number of years, because the modules 26 need only be fabricated and positioned as they are required. Also, when appropriate long-term solutions for the storage of irradiated fuel assemblies have been reached, the modules 26 can be easily deactivated and the assemblies still inside the canisters can be transported to the permanent storage facility. While illustrative embodiments have been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the disclosure.
043107656	abstract	A neutron accelerator tube having a target section, an ionization section, and a replenisher section for supplying accelerator gas to the ionization section. The ionization section is located between the target and the replenisher section and includes an ionization chamber adapted to receive accelerator gas from the replenisher section. The ionization section further includes spaced cathodes having opposed active surfaces exposed to the interior of the ionization chamber. An anode is located intermediate the cathodes whereby in response to an applied positive voltage, electrons created by field emission are transmitted between the opposed active surfaces of the cathodes and produce the emission of secondary electrons. The active surface of at least one of the cathodes is formulated of a material having a secondary electron emission factor of at least 2. One cathode member located in the tube adjacent to the replenisher section may have a protuberant portion extending axially into the ionization chamber. The other cathode spaced from the first cathode member in the direction of the target has an aperture therein along the axis of the protuberant portion. An annular magnet extends around the exterior of the ionization chamber and envelops the anode member. Means are provided to establish a high permeability magnetic flux path extending outwardly from the opposed poles from the magnet to the active surfaces of the cathode members.
	summary
	summary
	abstract	An x-ray reflector may include: a substrate; a first layer formed on the substrate, the first layer including a relatively higher-Z material, where Z represents the atomic number; and a second layer formed on the first layer, the second layer including a relatively lower-Z material; at least one of the first layer and the second layer exhibiting a taper in an axial direction extending between a first end of the substrate and a second end of the substrate.
048881503	abstract	A control rod for a nuclear reactor comprising a number of absorber plates (13-16) which are connected to each other along a center line on the control rod and which are each provided with a plurality of bored channels (18b), which extend at least substantially perpendicularly to the center line of the rod, contain boron carbide or other absorber material which swells upon irradiation and are sealed off from communication with the surroundings of the control rod. Within at least one region of an absorber plate, each channel is arranged at a smaller distance to an adjacent channel than to the surface of the absorber plate. Preferably, each channel within the stated region of the absorber plate is arranged at a smaller distance to the adjacent channel on one of its sides than to the adjacent channel on its other side.
048790907	summary	BACKGROUND OF THE INVENNTION The invention described herein relates to nuclear reactor fuel assemblies and more particularly to a zircaloy fuel assembly grid designed to improve strength, and reactor performance, and to be manufactured at a cost less than conventional grids. It is well known that the fuel or fissionable material for heterogeneous nuclear reactors is conventionally in the form of fuel elements or rods which are grouped together. These groupings or fuel assemblies also include rods comprising burnable poisons and hollow tubes through which control element assemblies are arranged to pass. The liquid moderator-coolant, normally water, flows upwardly through the reactor core in channels or longitudinal passageways formed between the members that comprise the core. One of the operating limitations on current reactors is established by the onset of film boiling on the surfaces of the fuel elements. The phenomenon is commonly referred to as departure from nucleate boiling (DNB) and is affected by the fuel element spacing, system pressure, heat flux, coolant enthalpy and coolant velocity. When DNB occurs, there is a rapid rise in the temperature of the fuel element due to the reduced heat transfer which can ultimately result in failure of the element. Therefore, in order to maintain a factor of safety, nuclear reactors must be operated at a heat flux level somewhat lower than that at which DNB occurs. This margin is commonly referred to as the "thermal margin". Nuclear reactors normally have some regions in the core which have a higher neutron flux and power density than other regions. This situation may be caused by a number of factors, one of which is the presence of control rod channels in the core. When the control rods are withdrawn, these channels are filled with moderator which increases the local moderating capacity and thereby increases the power generated in the fuel. In these regions of high power density known as "hot channels", there is a higher rate of enthalpy rise than in other channels. It is such hot channels that set the maximum operating conditions for the reactor and limit the amount of power that can be generated, since it is in these channels that the critical thermal margin is first reached. Attempts have been made in the past to solve these problems and increase DNB performance by providing the support grid structures employed to contain the members of the fuel assembly with integral flow deflector vanes. These vanes can improve performance by increasing coolant mixing and rod heat transfer ability downstream of the vanes. These attempts to improve performance have met with varying success depending on the vane design and the design of other grid components which can impact the effectiveness of vanes. To maximize the benefit of the vanes, the size, shape, bend angle, and location of the vanes must be optimized. The vanes are especially beneficial adjacent to the aforementioned hot channels. The remaining components of the grid which include the strips, rod support features and welds must be streamlined to reduce the turbulence generated in the vicinity of the vanes. Further constraints on designing the grids include minimizing grid pressure drop and maximizing grid load carrying strength. Grids are generally of a first and second plurality of half-slotted strips in "egg-crate" configuration and are spaced along the fuel assembly to provide support for the fuel rods, maintain fuel rod spacing, promote mixing of coolant, provide lateral support and positioning for control assembly guide tubes, and provide lateral support and positioning for an instrumentation tube. The grid assembly usually consists of individual strips that interlock to form a lattice. The resulting square cells provide support for the fuel rods in two perpendicular planes; in general, each plane has three support points: two support arches and one spring. The springs and arches are stamped and formed in the grid strip and thus are integral parts of the grid assembly. The springs exert a controlled force, preset so as to optimally maintain the spring force on the fuel rod over the operating life of the fuel assembly. Fuel assemblies employing spacer grids with flow deflector vanes of the prior art have usually been fabricated substantially or entirely of Inconel or a zirconium-tin alloy, i.e., zircaloy. An Inconel grid has the advantage of greater strength because of better material characteristics and because the brazing process bonds the intersection of the strips along its entire length. Brazing also has the advantage of providing little or no obstruction to flow. Due to the increased strength, the strip thickness of an Inconel grid can be reduced relative to the zircaloy grid to reduce pressure drop and turbulence in the vicinity of the vanes. The use of annealed zircaloy has been directed by its desirable combination of mechanical strength, workability, and low neutron capture cross-section. The most important of these characteristics is its low neutron capture cross-section which makes the nuclear fission more efficient, thus making the nuclear reactor operate more economically. However, to achieve a strength equivalent to that of an Inconel grid, the strip thickness for a zircaloy grid must be increased, thus creating more turbulence and higher pressure drop. Also, the joining of the interlocking zircaloy strips has always been by welding which requires the melting of some grid material to form a weld nugget. The increases strip thickness and weld nuggets for zircaloy grids of the prior art increase turbulence and grid pressure drop and reduce the effectiveness of the vanes. Therefore, the DNB performance of a zircaloy grid containing flow vanes of the prior art will be degraded relative to an Inconel grid design. In U.S. Pat. No. 4,089,741, a split vaned grid is disclosed in which first and second welding tabs are disposed in intersecting relation. Fusing of the protruding tabs at the intersection points down into the intersection joints occurs such that the protruding tabs are consumed whereupon there is formed in said vanes an opening at the base thereof, but within the bent and flow exposed vanes and not the vertical sections supporting them. The openings have a shape of the same general configuration as that of said first protruding tab, whereby flow is through the opening and in that patent, it is alleged the flow mixing capability of said spacer is improved. FIG. 1 is a prior art view showing what happens to create flow separation when a vane such as that of U.S. Pat. No. 4,089,741 has a nugget weld "unshielded" from the flow and an opening in the vane itself. U.S. Pat. Application Ser. No. 856,888 of Donald W. Krawiec, now U.S. Pat. No. 4,725,402 assigned to the assignee of the instant invention teaches "shielding" the weld nugget from the flow path within the confines of the strips in openings along their lines of intersection, to minimize pressure drop. This application does not specifically disclose integral vanes of the type in U.S. Pat. No. 4,089,741, however, which have openings which increase pressure drop by flow separation during flow therethrough. Water table tests were performed to visualize how the weld nugget and the welding hole cutout in the prior art vane for a nugget affects the flow passing by and through the vane. FIG. 1 illustrates the prior art flow pattern with a nugget and its weld hole in the vane. It can be seen that the weld nugget/weld access hole generates a very large wake, which, in turn, promotes decay of the vane effectiveness downstream of the grid. Velocity measurements downstream of the grid, both in water table tests and in an air model, using Laser-Doppler Anemometry, support the claim that the vane of the invention is more effective in directing the flow into the fuel rod gap because the weld nugget/weld access hole is not present in the vane. SUMMARY OF THE INVENTION The present invention overcomes the above briefly discussed deficiencies and disadvantages of the prior art by enhancing the strength and mixing ability of the grid. Grids employing the present invention have a measurable beneficial effect on reactor performance, operating cost and efficiency when compared to the prior art. In accordance with the invention, the crush strength of a zircaloy reactor fuel assembly spacer grid is increased. This increase is principally attributable to a novel and improved perimeter and interior strip. The novel perimeter strip is characterized by small stiffening ribs and round dimple stiffening features, both of which have been located differently in the prior art, as seen, for example, in U.S. Pat. Nos. 4,224,107 and 3,751,335. The ribs of the invention extend around the perimeter strips at two elevations and are ridged inwardly. The dimples also extend inwardly into some or all of the fuel rod receiving grid sectors serving to rigidize the perimeter strip and functioning as either arches for fuel rod support or backup arches for the integral fuel rod positioning springs which extend inwardly from the perimeter strip. The junction of the internal strip to the perimeter strip, in accordance with the present invention, is characterized by a weld seam of substantially greater length than has previously been employed. The increased weld seam length also enhances the strength of the grid. Interior orthogonal strips are designed to limit cutouts in the unslotted section of the strip. This is accomplished by the use of small cantilevered springs, designed to laterally impress a controlled resistive force on each fuel rod. The spring's size allows it to be located in the slotted section of the interior grid strip. The design maintains a load path through the unslotted interior strip which is much larger than in the prior art and thus leads to a much higher strength grid as compared to grids of equal size of the prior art. In addition, grid support features, i.e. support arches and springs, have been positioned in a staggered manner so that turbulence produced is minimized, thus obtaining even better performance from the integral flow deflector vanes. The staggered positioning of grid support features also reduces grid pressure drop and promotes coolant mixing by staggering the flow through a grid cell. DNB tests performed for the zircaloy grid design indicate these unique features significantly increase coolant mixing and DNB performance by comparing the test results to an Inconel grid design with similar flow vanes. These comparisons verified that the zircaloy grid design produced improvements in performance relative to the Inconel design. As described previously, grids are formed of "egg-crate" construction by zircaloy strips which form multiple cells or sectors, each sector having springs on two adjacent walls and a pair of projections or arches on each of the other two walls forming a sector. The springs laterally impress controlled resistive forces on each fuel rod in the assembly. Although this fuel assembly design performs exceptionally well in a nuclear reactor, one disadvantage inherent in the design is that the inwardly projecting springs and arches occasionally mark or score the surface of fuel rods during the time they are being pulled or pushed into the fuel assembly grids. In carrying out this fuel rod loading operation, the grids are held immovably in position while a longitudinal steel rod attached to the end of a fuel rod push or pull it axially through the aligned openings or sectors in the grids. As the rod engages the springs and arches in the grid sectors, their edges engage the exposed surface of the moving fuel rod and, in some cases, score its surface sufficiently deep so as to cause the rod to fall outside established fuel rod surface specifications. To eliminate this problem, arches and springs have been designed with a crown. The crown's size has again been optimized with respect to flow blockage to minimize turbulence and pressure drop and the scorability of the rod. The number of spacer grids employed in a single fuel assembly will be minimized, to an extent commensurate with structural requirements, in the interest of enhancing reactor operating efficiency. While possessing adequate resistance to buckling under normal operating conditions, laboratory tests have shown that prior art zircaloy spacer grids may not have the mechanical strength required to absorb severe lateral stresses as might be encountered as a result of high seismic loading. Higher strength grids are required in plants whose locations are in areas of high seismic activity. While the strength of reactor fuel assembly spacer grids could be increased by the use therein of metals having a greater stiffness than annealed zircaloy, most of such higher strength materials are also characterized by higher neutron capture cross-section when compared to zircaloy and a principal objective in the design of a fuel assembly for a nuclear reactor is to maximize operating efficiency by minimizing neutron capture. To maintain a zircaloy grid strip material while obtaining the required strength, the unslotted section of the grid strip was provided with minimum cutouts. This was accomplished by the use of a small cantilevered spring, designed to laterally impress a controlled resistive force on each fuel rod. The spring's size allowed it to be located in the slotted section of the grid strip. The design maintains a continuous load path of unslotted material which is much larger than previous art and thus has a much higher strength as compared to grids of equal thickness and height. Tests have been performed which support this claim and have shown an increase in strength of 15-20% over grids of the prior art. Intermediate welds of the type taught in U.S. Application Ser. No. 856,888, now U.S. Pat. No. 4,725,402 were also provided to improve the strength of the grid. The slots were tapered at their ends to facilitate welding at intermediate locations, thus improving grid strength. Tests have also been performed which support this claim and have also shown an increase in strength of approximately 15% over grids of the prior art. This increase would be additive to that described in the paragraph above. The outer strips of the grid have also been optimized with respect to strength, handling, turbulence generation, and pressure drop. To obtain additional strength, small ribbed and round dimple stiffeners were employed along the strip's entire length. These stiffeners did not only increase the buckling resistance of the grid but improve the strip's resistance to interact, i.e catch or hang-up with adjacent fuel assemblies which reduces the potential for handling damage. In addition, the optimized outer strip design more effectively spreads accident loadings throughout the grid interior strips thus increasing strength. The outer strip also diverts just enough flow to the interior of the fuel assembly to match the thermal power distribution of the fuel array and eliminate any corrosion concerns on peripheral fuel rods. The novel features described herein summarize the improvements made to a straight strip zircaloy grid containing flow vanes to improve upon reactor performance, load carrying strength, pressure drop, and handling performance relative to zircaloy grid designs of the prior art. To improve reactor performance, the integral flow deflector vanes were optimized with respect to size, shape, and bend angle in order to maximize coolant mixing and fuel rod heat transfer downstream of the vanes. The weld nuggets were optimized with respect to size, strength, corrosion resistance, and location, i.e., the nugget is positioned upstream in the grid strip and is shielded by the integral flow deflector vanes. By recessing the nugget into the grid with no cutout in the vane, the turbulent wake produced by the nugget and cutout has less impact on vane performance. The location of the nuggets, upstream and shielded by the integral flow deflector vanes, leads to less turbulent wakes and better vane mixing and fuel rod heat transfer performance than the prior art. The novel staggering of grid support features also increased vane performance and lowered grid pressure drop as compared to prior art. It will occur to those skilled in the art that the grid strip thickness can be reduced by virtue of the increased strength provided by the staggered support configuration and the stronger outer strip design. This, in turn, reduces the pressure loss experienced by the coolant in flowing through the reactor core. Having in mind the above and other objects that will be evident from an understanding of this disclosure, the present invention comprises the combinations and arrangements and method as illustrated and disclosed in the presently preferred embodiment of the invention which is hereinafter set forth.
039502201	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The drawing illustrates a portion of the pressure vessel 6 in a boiling water reactor. The vessel 6 (e.g., the bottom wall of the vessel) has an opening 1 which is provided therein for the purpose of installing a novel internal primary recirculating pump P. In addition, the opening 1 simultaneously serves for admission of feed water by way of an annular chamber 2 in the body 9 of the pump P. The chamber 2 is closely adjacent to the rotor 5 of the pump P. Prior to entering the pressure vessel 6, the stream of feed water flows from the annular chamber 2 between the blades of the wheel 4 of a single stage turbine which is mounted on the pump shaft 3, preferably at the suction side of the rotor 5. The shaft 3 is driven by an electric motor 7 which is mounted in the body 9 of the pump P. In the illustrated embodiment, the turbine wheel 4 serves as a carrier for the pump rotor 5. When the supply of current to the motor 7 is interrupted due to an unforeseen failure, the turbine wheel 4 prevents rapid deceleration of the shaft 3 to zero speed to thus insure that the cooling of the boiling water reactor proceeds for a sufficient interval of time. A bypass channel 8 between the pressure chambers of the turbine and pump rotor is preferably formed as a slightly enlarged sealing gap and serves to prevent an overheating of the turbine under certain operating conditions. If desired, the illustrated single-stage turbine can be replaced with a multi-stage turbine. The turbine replaces costly, bulky and complex motor generator sets which are used in conventional boiling water reactors to insure the supply of additional energy in the event of current failure. The turbine insures that the circulation of feed water in the pressure vessel 6 continues even if the motor 7 is disconnected from the energy source. Moreover, and as long as the quantity of feed water which is being admitted to the chamber 2 remains substantially unchanged, the turbine including the wheel 4 (or a multi-stage turbine) is capable of insuring that the pump P remains in operation at a reduced RPM. The cold feed water which is admitted by way of the chamber 2 is caused to mix with hot reactor water upstream of the pump rotor 5. This produces a desirable supercooling affect and a substantial NPSH gain. The motor 7 may be a normal electric motor or a wet or canned motor. Without further analysis, the foregoing will so fully reveal the gist of the present invention that others can, by applying current knowledge, readily adapt it for various applications without omitting features which fairly constitute essential characteristics of the generic and specific aspects of my contribution to the art and, therefore, such adaptations should and are intended to be comprehended within the meaning and range of equivalence of the claims.
	description	This invention pertains to an apparatus for detecting and analyzing equipment operational parameters. More particularly the invention pertains to apparatuses for imaging and dynamic signal analysis for monitoring the status of equipment health. Imaging devices include focal plane array devices that sense infrared or visible light. Dynamic signal analysis devices include vibration or ultrasonic detectors. Infrared imagers are commonly used for thermographic inspections of equipment. State of the art for infrared inspection process involves use of an uncooled, radiometric, focal plane array, infrared camera plus visible camera built into a lightweight, hand-held package with onboard digital memory, an LCD display, and interactive user interface. Visible light imaging systems are also used for inspection of equipment. Examples of such applications are borescopes, fiberscopes, and even conventional video cameras. Various dynamic measurement systems have also been developed to monitor the operational health of equipment. Examples of such systems are vibration analysis devices, sonic or ultrasonic measurement devices, and electromagnetic spectrum analyzers In addition, various devices have been developed for measurement of conditions that are often more static in nature, such as temperature, pressure, and lubrication properties. Typically, vibration analysis and infrared analysis have been handled as distinct and separate condition monitoring techniques with regard to walk-around inspections, routes, or surveys. The maintenance departments of industrial plants have employed totally separate devices for each different condition monitoring method. For example a typical industrial plant often uses an infrared camera for infrared inspection, a multi-frequency sonic and ultrasonic inspection system for acoustic monitoring, a videoscope for video inspection, a minilab oil analyzer for on-site oil analysis, and a fast Fourier transform (FFT) equipment analyzer for vibration, flux, and current analysis. Existing technology does not adequately address all of the needs for integrating the collection of imaging information with other sensor measurements. What is needed is a system that provides portable imaging capability with portable dynamic sensor measurement capability plus optionally portable static measurement capability. With regard to the above, in one of its embodiments the invention provides a portable instrument for inspecting equipment. The portable instrument includes a first sensor interface for a focal plane array imaging sensor, where the first sensor interface includes electro-mechanics configured to receive imaging sensor data. The portable instrument also includes a second sensor interface for a dynamic sensor, where the second sensor interface includes electro-mechanics configured to receive dynamic sensor data. Further, the portable instrument incorporates a digital memory that stores an operating system and application instructions and a dataset. A processor is proved that runs the operating system and is operatively connected to the digital memory and operatively connected to the first sensor interface and is operatively connected to the second sensor interface. The processor is configured to use at least a portion of the application instructions for recording in the dataset at least a portion of imaging sensor data and for recording in the dataset at least a portion of dynamic sensor data. The processor is further configured to use at least a portion of the application instructions to operate on the imaging sensor data and to derive at least one dynamic indication of equipment health. Additionally, the portable instrument has a display that presents information; and a user interface that in cooperation with the processor controls what information is presented on the display. Alternate embodiment provides a portable apparatus for inspecting equipment that includes a first sensor interface for a focal plane array imaging sensor, where the first sensor interface includes electro-mechanics configured to receive imaging sensor data, and a second sensor interface for a dynamic sensor, where the second sensor includes electro-mechanics configured to receive dynamic sensor data. The portable apparatus also incorporates a processor operatively connected to the first sensor interface and operatively connected to the second sensor interface and configured with an application instruction for analyzing the dynamic sensor data and for deriving at least one dynamic indication of equipment health, and a display that presents information, and a user interface that in cooperation with the processor controls what information is presented on the display. The portable apparatus also incorporates a focal plane array imaging sensor that is operatively connected to the first sensor interface for sending imaging sensor data to the first sensor interface, and a dynamic sensor that is operatively connected to the second sensor interface for sending dynamic data to the second sensor interface. A further alternate embodiment is an apparatus for inspecting equipment that includes a portable instrument and a base station. The portable instrument incorporates a first sensor interface for a focal plane array imaging sensor, where the first sensor interface includes electro-mechanics configured to receive imaging sensor data, and a second sensor interface for a dynamic sensor, where the second sensor includes electro-mechanics configured to receive dynamic sensor data. The portable instrument also incorporates a processor operatively connected to the first sensor interface and operatively connected to the second sensor interface, plus a display that presents information, a user interface that in cooperation with the processor controls what information is presented on the display, and a wireless transmitter that is configured cooperatively with the processor to transmit at least a portion of the imaging sensor data and at least a portion of the dynamic sensor data. The a base station has a wireless receiver configured to receive at least a portion of imaging sensor data and at least a portion of dynamic sensor data transmitted by the transmitter in the portable instrument. The base station also includes a central processor that is operatively connected to the receiver, a station display that presents information, and a station user interface that in cooperation with the central processor controls what information is presented on the station display. A different embodiment presents a method for inspecting equipment that involves storing in an instrument (a) application instructions for receiving, storing and analyzing focal plane array imaging sensor data to derive at least one imagery indication of equipment health, and (b) application instructions for receiving, storing, and analyzing dynamic sensor data to derive at least one dynamic indication of equipment health, and (c) application instructions for correlating at least one imagery indication of equipment health with at least one dynamic indication of equipment health. The method continues with receiving and storing focal plane array imaging sensor data and dynamic sensor data in the instrument using at least a portion of the application instructions, deriving at least one imagery indication of equipment health using at least a portion of the application instructions, and deriving at least one dynamic indication of equipment health using at least a portion of the application instructions. The method concludes with the step of correlating at least one imagery indication of equipment health with at least one dynamic indication of equipment health. A further alternate method embodiment is a method for inspecting equipment that includes storing in an instrument application instructions for capturing and transmitting imaging sensor data from a focal plane array imaging sensor and application instructions for capturing and transmitting waveforms from a dynamic sensor, and storing in a base station application software for (a) receiving, storing and analyzing imaging sensor data to derive at least one imagery indication of equipment health, and (b) application software for receiving, storing, and analyzing waveforms to derive at least one dynamic indication of equipment health, and (c) application software for correlating at least one imagery indication of equipment health with at least one dynamic indication of equipment health. The method includes a step of capturing imaging sensor data with a focal plane array imaging sensor and transmitting at least a portion of the imaging sensor data from the instrument to the base station using at least a portion of the application software. A further step is receiving and storing in the base station at least a portion of the imaging sensor data transmitted by the instrument using at least a portion of the application software. The method includes capturing dynamic sensor data with a dynamic sensor and transmitting at least a portion of the dynamic sensor data from the instrument to a base station using at least a portion of the application software, with the further step of receiving and storing in the base station at least a portion of the dynamic sensor data transmitted by the instrument using at least a portion of the application software. The method concludes with deriving at least one imagery indication of equipment health using at least a portion of the application software, and deriving at least one dynamic indication of equipment health using at least a portion of the application software. An alternative embodiment provides a system for inspecting equipment. the system incorporates a portable instrument that includes a focal plane array imaging sensor selected from selected from the group consisting of (a) an infrared focal plane array imaging sensor and (b) a visible focal plane array imaging, where the focal plane array imaging sensor is configured to generate imaging sensor data and a dynamic sensor selected from the group consisting of (a) a vibration sensor and (b) a sonic sensor and (c) an ultrasonic sensor and (d) a flux sensor and (e) a current sensor, where the dynamic sensor configured to generate dynamic sensor data. The portable instrument further includes digital memory that stores an operating system and application instructions and a dataset. The portable instrument has a processor that runs the operating system and is operatively connected to the digital memory and operatively connected to the imaging sensor and is operatively connected to dynamic sensor. The processor is configured to use at least a portion of the application instructions for recording in the dataset at least a portion of the imaging sensor data and for recording in the dataset at least a portion of the dynamic sensor data. The processor is further configured to use at least a portion of the application instructions to operate on the imaging sensor data and the dynamic sensor data stored in the dataset and to derive at least one dynamic indication of equipment health. the portable instrument also includes a display that presents at least a portion of the imaging sensor data and presents at least a portion of the dynamic sensor data. The portable instrument also has a user interface that in cooperation with the processor controls what imaging sensor data and what dynamic sensor data is presented on the display. A further alternative embodiment provides a method for automating inspection of an equipment item using both imaging and dynamic signal analysis. The method begins with providing a battery-operated inspection device having a processor, memory, a display having at least one window data input, and a user interface. The method proceeds with providing imaging data and dynamic signal data for the equipment item to the processor. The method also includes steps of using the processor to derive a dynamic indication of equipment health based upon a least a portion of the inputted dynamic signal data and using the processor to establish an association data element. the method also includes a step of providing a user interface selection to allow a user to view at least a portion of the imaging data and at least a portion of the dynamic indication of equipment health on at least one window on the display while the user performs the inspection. One advantage of these and other embodiments is the improved ability to analyze the health of equipment. Incorporation of means to gather data in the field is also important in some embodiments. Other advantages of various embodiments include integrating the functions of a portable instrument with a base station. Also, as will be seen in the detailed description of various embodiments, provisions for analyzing imaging sensor data and dynamic sensor data are incorporated to meet previously identified needs. Finally, embodiments are provided that incorporate combined analysis of imaging sensor data and dynamic sensor data thereby enhancing the overall versatility and utility of various embodiments for maintenance and preventive maintenance operations. The present invention provides an apparatus for efficiently identifying and analyzing concerns possibly requiring maintenance for various types of equipment and machinery such as power circuits, transformers, switchgear, motor control centers, motors, pumps, fans, presses, drive trains, gear boxes, etc. The term “equipment” will be used and understood herein to include machinery and to cover devices with moving part as well as devices without moving parts. Many embodiments described herein allow for complex analysis, including summation, of multiple signals representing equipment characteristics through a plurality of sensors, and provides the opportunity for economy, time savings and safety through operation of a portable platform connected by contact or wireless means to a both dynamic signal analysis and focal plane array imaging sensors. A portable platform is a form of an instrument. The preferred embodiments employ a multiple-technology, highly automated, portable inspection system that combines infrared inspection with other portable condition monitoring technologies. One aspect of the most preferred embodiments is a portable platform that the technician carries to the field. The portable platform typically includes a processor with software constructed in a housing. The portable platform generally also includes portable display, portable power system, and data input and data output capability. It has mouse or touch-screen or button or other user interface capabilities for use by the field technician. Another aspect of the most preferred embodiments of the portable platform is the incorporation of at least one sensor interface in the portable platform. A sensor interface typically comprises electro-mechanics (hardware, firmware, or both) that are capable of receiving data from a sensor and conveying the data to central processing unit in portable platform so that the data may be stored in electronic memory. In some embodiments, the sensor interface also includes electro-mechanics for transmitting data from the portable platform to the sensor. A sensor interface may also incorporate electronics and firmware tools for translating signals from a sensor into useful data. For example, a sensor interface may include an analog to digital converter, a sampling circuit or sampling software, a frame grabber, or a format conversion tool such as hardware or software for converting NTSC or PAL video signals to VGA or SVGA format for presentation on a display, or for converting such signals to .jpg (or similar) files for storage in an electronic memory. A sensor interface may also include data authentication tools such as time stamping, encryption, and file locking software, although such data authentication tools may alternately be provided by application instructions that reside in the electronic memory and are executed by the portable platform. Some examples of a sensor interface are a video capture card, an RS-232 serial port module, a parallel port, a universal serial bus (USB) card, an analog interface adapter, an input/output card, and a data acquisition board. In the most preferred embodiments, the sensor interface is operatively connected to the housing of the portable platform, meaning that is mechanically mounted and electronically integrated with the other electronics. Generally, sensor interfaces are designed to accommodate dynamic sensor data. Dynamic sensor data represents information having a time domain, meaning that the measurements detected by the sensor vary over time and that variation is recorded from a start time to an end time. However, sensor interfaces are often also designed to accommodate static sensor data. Static sensor data represents measurements taken at a single point in time. Imaging techniques can include either infrared or visible detection sensors having either analog or digital output. Infrared cameras are often used as infrared imaging sensors and digital cameras are often used as visible imaging sensors. In both cases the field of view for the image describes an area of interest. Infrared imaging includes both radiometric and non-radiometric type detector arrays. Imaging may be individual frame or multiple frames. Imaging may include enhancements by magnification, zoom, light amplification, or optical wave guides, or other techniques. These images produced by such imaging sensors are examples of imaging sensor data. Typically image data analysis produces an array of values such as emissivity or temperature or another imagery indication of equipment health. Other imagery indications of equipment health may include the following. Point value representation normally associated with either the center pixel or a cursor. Maximum scalar value determination. Minimum scalar value determination. Average scalar value determination. Median scalar value determination. Absolute or standard deviation scalar value representing a range of values. Delta scalar value or differential determination. Contour of scalar values or connection of pixels having similar values. Alarm limit scalar values methods for distinguishing values inside or outside of alarm conditions. Histogram showing statistical profile representation of scalar values within a selected area. Line profile showing scalar values corresponding to a linear path on the image Other individual, differential, or statistical analysis of scalar values with or without considering pixel position. Distance in pixels or other dimensional units between features on the image. Number(s) of item(s) with particular characteristics on at least a portion of the image. Classification of characteristics of object(s) in the image based on particular visual characteristics. Comparison of image being analyzed with one or more reference imaging sensor data. Results of parametric analysis of the image using a digital image analysis software tool. Results of parametric analysis using other graphical image analysis tools. Dynamic sensors typically employ devices such as accelerometers, piezoelectric components, electrical current or voltage probes, thermocouples, pitot tubes, and sonic or ultrasonic detectors. Dynamic analysis or dynamic signal analysis techniques include, but are not limited to Fast Fourier Transform (FFT) vibration analysis, waveform vibration analysis, spectral vibration analysis, stress wave analysis, transient analysis, sonic analysis, ultrasonic analysis, FFT flux analysis, and FFT current analysis. Such analysis generally produces one or more dynamic indications of equipment health. Examples of dynamic indications of equipment health are: Speed Overall value Less than one times turning speed value One times turning speed value Two times turning speed value Three to eight times turning speed value Nine to thirty-five times turning speed value More than thirty-five times turning speed value One times line frequency value Two times line frequency value 4 kHz peak value 4 kHz average value 4 kHz peak hold value 30 kHz peak value 30 kHz average value 30 kHz peak hold value 40 kHz peak value 40 kHz average value 40 kHz peak hold value 150 kHz peak value 150 kHz average value 150 kHz peak hold value The dynamic indications of equipment health are often associated with particular locations or orientations. Here are some examples of such particular locations or orientations: Motor outboard horizontal Motor outboard vertical Motor outboard axial Motor inboard horizontal Motor inboard vertical Pump inboard horizontal Pump outboard horizontal Pump outboard axial Inlet Outlet Suction Discharge X, Y, Z or other coordinate locator Angle or other relative orientation Many preferred embodiments provide for the correlating of imaging analysis data and dynamic or static data analysis data. In a very basic form, the correlating is accomplished by simply making both imaging sensor data and dynamic or static data available substantially simultaneously to a technician so that results can be reviewed comparatively. Correlating data may also involve such actions as adjusting scales to common units, identifying data sets that pertain to the same equipment or measurements, matching imaging sensor data files with dynamic sensor data files, and performing multivariate analysis. In some instances this correlating includes the calculation of one or more equipment health combined statistics that are derived from a joint analysis of imaging sensor data and dynamic/static data. Examples of equipment health combined statistics are: Temperature of an excessively vibrating bearing. Overlay of a thermal trend and a vibration level trend. Viscosity of oil at the highest temperature point in a machine. A plot of peak vibration versus temperature. Amperage at hottest spot in a power line. Comparison of a thermal image and an ultrasonic image. Dimensional location of hottest point in a furnace. Thermographic image data and corresponding ultrasonic dB values for inlet and outlet positions on a steam trap or other valve. Infrared image showing fluid level, compared to level sensor output showing same fluid level. Infrared image showing relatively hot coupling verifying vibration analysis results indicating misalignment. Visual image showing adhesive wear indications from mixed mode or boundary lubrication compared to elevated ultrasonic dB levels. Delta-temperature correlated with heterodyned ultrasonic sounds from electrical discharge or corona on power line insulator connections. Visual strobe imaging synchronized with a vibration fault frequency. Bore scope image of gear or bearing components identified by characteristic vibration spectrum. Animation of an otherwise static visual image using data from modal vibration analysis. Correlation of image and vibration due to misalignment before and after thermal growth. Comparison of ultrasonic dB levels with oil level. Verification that heat showing up on a thermographic image is caused by increased friction due to adhesive wear (e.g., boundary lubrication regime) by measuring airborne ultrasonic signature in the vicinity of the relatively hot location on the thermogram. Comparing data from wear debris image analysis with PeakVue® vibration data. Correlation of ultrasonic leak detection with thermographic image for a system containing compressed or heated gas Validation of stator faults by comparing thermographic image with motor flux analysis. In integrating imaging and dynamic sensor data it is beneficial to store to at least one association data element that identifies what imaging sensor data is associated with what sensor data. In most cases this association is the result of taking and recording both the imaging sensor data and the sensor data related to a particular piece of equipment at approximately the same time. However in some cases the association may relate to changes that occur over time, comparative information taken from multiple machines, or other considerations. The association data elements may be established by creating a data field in an independent database that links the identity one or more image files with the identity of one or more sensor data files. In other cases association data elements may be established by creating matching data records in separate file fields in both the image file(s) and the sensor data file(s) that are associated with each other. One image file may be associated with only one sensor data file, or one image file may be associated with multiple sensor data files, or multiple image files may be associated with one sensor data file. It is even possible that multiple image files and multiple sensor data files are all associated. Examples of association data elements are an electronic date and time stamp, a job code identifier, an operator identifier, a location identifier, a subject identifier, or even a random number that ties both the imaging sensor data and the dynamic sensor data together so that the two data sets are identified as being associated with each other. An example of combining imaging and associated analysis with dynamic analysis in a portable system is the integration of infrared thermography with portable vibration analysis. The infrared focal plane array and vibration transducer are two of the sensors from which the technician collects data while in the field using the portable platform. This combination provides a view of equipment health enabling fault isolation. Certain faults trigger temperature changes, some trigger both temperature and vibration signatures, and some trigger only vibration indications. Examples are motor stator shorts, coupling faults, and imbalance respectively. This method of using multiple techniques to view equipment faults is called fault isolation. The use of thermographic imaging allows the technician to survey a large area on the machine in a rapid sweep, quickly locating hot or cold sites. Temperature anomalies are easily identified with the focal plane array technique that would likely have been missed using point temperature measurements. Temperature excursions are commonly associated with equipment faults. Embodiments that include portable vibration analysis provide the technician with a further beneficial tool to assess such equipment faults. Another example of imaging and dynamic analysis is the integration of infrared thermography with sonic or ultrasonic analysis. In this case the infrared focal plane array and sonic or ultrasonic transducers are two of the sensors from which the technician collects data while in the field using the portable platform. The term sonic sensor refers to a sensor that detects transmission media waves at frequencies up to the top of the human audible range, whereas the term ultrasonic sensor refers to a sensor that detects transmission media waves above the top frequency of the human audible range. Both ultrasonic and infrared focal plane array technologies are well suited for area surveys. The combination provides greater insight than use of either one independently. For example, steam traps are best surveyed using both infrared and sonic/ultrasonic measurements. The focal plane array is able to identify steam flow and blow-by. Sonic and ultrasonic sensors are able to identify performance of many mechanisms inside the steam trap. The combination isolates faults and provides intuitive insight as to normal versus abnormal operation. Certain traps are faulty when steam is blowing by. For others this is normal. The visible indication and ultrasonic signature allows the user to fully understand the operation and interpret the normal or fault condition. In the same way, ultrasonic and infrared are excellent combined field technologies for electrical applications in which corona, arcing, or discharge may occur. By understanding that ultrasonic signature often indicates friction and sustained high friction generates temperature excursions, the operator equipped with a single platform including both measurements has the advantage. Still another aspect of some embodiments is the integration of infrared thermography with oil analysis information collected while on a route or survey. In this case the infrared focal plane array and an oil sensor are two of the sensors from which the technician collects data using the portable platform. Four of the common equipment fault conditions revealed through lubricant analysis include fatigue, abrasion, adhesion, and corrosion. Fault isolation can be enhanced using the combination of infrared focal plane array inspection with lubricant analysis. For example, adhesion often results from inadequate lubrication which may be caused by low viscosity. A viscosity sensor reporting low oil viscosity combined with the presence of heat in load bearing regions provides an indication of cause and effect. The viscosity sensor can be installed in the machine with communication via cable or wireless method, or the viscosity sensor can be transported to the equipment by the operator. Typically an oil sensor is a static sensor and the measured data are static sensor data. Other examples of sensors that are normally used as static sensors are pressure gages, temperature probes and linear displacement gages. The combination of three or four data analyses (e.g., oil analysis, sonic/ultrasonic analysis, vibration analysis, and focal plane array infrared analysis) may provide the operator with immediate and accurate indication of equipment health that would not have been derived by one or even two of the separate technologies. For example, in the event that the oil level falls below critical level causing the oil sensor to trigger “low oil” the resulting condition can be a lubricant starved bearing condition with high ultrasonic signature, high vibration as well as a temperature excursion. The operator who observes this situation fills the oil level, noting the return to normal oil level, normal sonic/ultrasonic signature, normal temperature, and vibration signature with distinctly quantified vibration faults and fault frequencies. The combination allows a field operator to make an accurate assessment of the equipment condition and effect of corrective actions. Yet another aspect of the invention is the integration of infrared thermography with bore-scope inspection while on a route or survey. In this case the infrared focal plane array and the bore-scope are two of the sensors from which the technician collects data using the portable platform. In this case the infrared looks at the outside surface temperature while the bore-scope interrogates the inside aspects of a difficult to access volume containing critical mechanisms. A bore-scope is one example of a visible image sensor. A visible image sensor is a sensor that detects light in the human visible spectrum. Other examples of visible image sensors are camcorders, optical microscope imagers, and digital cameras. Typically, the technician will determine which measurements will need to be made and carry only the items needed for a particular survey. Sensors in the sensor suite may be carried by the technician or may already be installed in the field application. Sensors may be connected by cable to the portable platform. An alternative configuration is for the sensor to be in wireless communication. Another alternative is for the sensor to collect measurements in one operation and then transfer the data to the portable platform in a second operation using electrical contacts or wireless communication. Also a microphone recording device, typically built into the portable platform, may be used to provide voice annotation. A microphone is a specialized form audible detector that is excluded from the general category of sonic sensors. Many embodiments incorporate a base computer that in preferred embodiments consists of a personal computer with database and application software, data input/output, and a printer. Either the portable platform or the base computer or both can be part of a network or server or internet application. An aspect of preferred embodiments is a portable platform which may include a processor (such as a central processing unit), display, power supply, transmitter, user interface, database, application instructions (firmware or software), and at least one sensor interface. Often a personal computer is adapted to become the computing portion of a portable platform. In some of these embodiments the portable platform uses cables and electrical contact to transmit data. In an alternate embodiment, communications may be accomplished using wireless means such as infrared or radio frequency or microwave. Such wireless communications may be used between the portable platform and one or more of the sensors. Wireless communications may also be used between the portable platform and the host system. Generally the application instructions loaded in the portable platform direct the functioning of the unit. However in some cases some of the software, database, and other functions may be equivalently performed in the host or base platform instead of on the portable platform by, for example, using networks, file servers, and so forth. An overarching theme of many embodiments is the ability to operate on imaging sensor data and dynamic or static sensor data. Operating on the data may include such actions as editing the data, reformatting the data, adding annotations to the data, simultaneously viewing image data and sensing data, correlating data from the same sensor, correlating the data from different sensors, or analyzing the data. Referring now to FIGS. 1, 2, and 3, a sensor suite (122) is shown to be in communication with a portable platform (123 and 207) to perform condition monitoring analyses. Analysis is typically done by a technician who walks a route or performs an area survey carrying as a minimum the portable platform (123 and 207) and a dynamic signal analysis sensor such as either a sonic sensor (118), an ultrasonic sensor (117), a vibration sensor (119), a flux sensor also called flux coil (115) or current sensor (114). In addition the technician carries an imaging sensor such as the infrared camera (121) or visible camera (120). Other devices such as an oil sensor (116) may be used as well. The infrared camera (121) is typically an uncooled focal plane array type imager provides either a formatted digital signal or analog video signal to the portable platform (123). In one embodiment the ergonomically designed infrared imager (121) houses the imaging optics, detector, drive electronics, optical modulator, laser-pointer and four standard or rechargeable AA size batteries. The infrared camera (121) is an example of an image sensor and the output of the infrared camera is referred to as imaging sensor data. In alternate embodiments a visible spectrum digital camera may be the image sensor that produces the imaging sensor data. Application instructions (109) (typically stored as software or firmware) run on a processor (124) (such as a central processing unit) under an operating system (141) (such as Windows CE®) in the portable platform (123 and 207). The application instructions (109) are used to interpret the analog or digital signal, present this information on the portable display (125), and use the processor (124) to save imaging sensor data (147A, 147B) in the digital memory (140) of the portable platform (123 and 207). The digital memory (140) may include read only memory, random access memory, and media memory such as compact disc data storage. A visible camera (120) may be integral with the infrared camera (121) or may be independent. The visible camera may include optional lighting accessories such as flash, it may include special optics and wave guide accessories. An optical wave guide is valuable for bore-scope inspection in hard to reach locations of equipment. One embodiment for the visible camera (121) is a point-and-shoot liquid crystal display camera. Another embodiment uses a digital camera suitable for either live or still frame photography including zoom-coupled smart auto-focus system, automatic light guide zoom flash, and low-light features. The one-way or two-way data transfer sensor communication link (112) between the camera(s) (120, 121) and the portable platform (123 and 207) may be rigid and attached; may be temporarily attached; may be flexible, allowing reorientation different from that of the portable platform; may be connected via cable; or may be in communication with the portable platform via wireless means. In a preferred embodiment, the cameras (120 and 121) are temporarily stored in the portable platform (123 and 207) in one of the storage spaces (220A or 220B or 220C or 220D or 220E-224). In the preferred embodiment a sensor communication link (112) with cameras (120, 121) and power for cameras (120, 121) is provided through one of the connection ports such as general purpose ports (206A, 206B) or alternate connector port (208C). An alternative communication and power connection for cameras (121, 120) is the PCMCIA card interface (211). Such connections between cameras (120, 121) and the portable platform (123 and 207) are examples of sensor interfaces. A preferred embodiment for a dynamic sensor (119) is a piezoelectric transducer such as a single- or tri-axial accelerometer. Many other dynamic signal analysis sensors can be used instead. These normally supply 4 to 20 mA signal output transmitted through sensor communication link (112). Alternative dynamic signal measurement devices to the vibration sensor (119) include the sonic sensor (118), ultrasonic sensor (117), flux sensor (115), or current sensor (114). One example includes both sonic and ultrasonic measurements. This device measures decibel values including peak value, peak hold, and average in selected frequency sonic and ultrasonic frequency ranges which are typically 4 kHz, 30 kHz, and 40 kHz. Another example of a sonic sensor is a microphone. An example of a flux sensor is a current frequency clamp. Power may be supplied from the portable platform (123 and 207) to the sensor(s) (114, 115, 116, 117, 118, 119, 120, 121) through the sensor communication link (112). This power may be temporarily supplied for purpose of recharging battery or may be supplied for the entire time the sensor is in use. For wireless applications using sensor communication link (112) the sensor (114, 115, 116, 117, 118, 119, 120, 121) normally uses it's own battery power which may or may not be recharged by the portable platform (123 and 207). Sensor ports (208A, 208B) on the portable platform (123 and 207) may include features that provide one-way or two way data transfer between the portable platform (123) and one or more sensors (114, 115, 116, 117, 118, 119, 120, 121), power management and sensor signal interpretation for sensor(s) (114, 115, 116, 117, 118, 119, 120, 121). Sensor ports (208A, 208B) are examples of sensor interfaces. Optional accessories used in conjunction with dynamic signal analysis include a speed sensor, a tachometer, or a strobe. In preferred embodiments, a dynamic sensor (119, 118, 117, 115, 114) is connected through sensor communication link (112) to the portable platform (123 and 207) via a sensor port (206A or 206B), which preferably is ruggedized, or via alternate channels such as general purpose ports (206A, 206B) or communication port (208C), or to a PCMCIA card in the PCMCIA slot (211). Such connections serve as sensor interfaces. When it is not in use, the sensor(s) (114, 115, 116, 117, 118, 119, 120, 121) is(are) stored on the back of the portable platform (123) in a space provided for that purpose. An alternate embodiment uses wireless communication (112) between the sensor(s) (114, 115, 116, 117, 118, 119, 120, 121) and the portable platform (123 and 207). In the preferred embodiment the oil sensor (116) is fixed rigidly to the lubricating oil system. The oil sensor (116) measures some aspect of the lubrication system such as oil quality, oil level, oil contamination, or mechanical wear debris. For example a capacitive oil sensor can be used to measure the dielectric permittivity of oil and trigger low oil when the level falls to the level of the oil sensor (116). An alternative to mounting the oil sensor (116) in the oil system is to dip the sensor into the oil. The oil sensor (116) transmits measurements to the portable platform (123 and 207) via electrical or wireless connection (112). The portable platform (123 and 207) may include optional safety rating for use in hazardous environments including potentially explosive atmospheres. In a preferred embodiment, the portable platform (123 and 207) includes digital signal processing (DSP), to enable fast measurement time for greater productivity. Productivity is reduced by reducing data collection time and simplifying analysis using high real-time rate, fast auto-ranging, and an extended dynamic range. In the most preferred embodiments the portable platform (123 and 207) is small and lightweight so that it can easily be carried up ladders and into tight areas, even on the longest routes. The rugged housing will resist damage, and it stands up to harsh operating conditions. The backlit display and special electroluminescent keypad eliminate operational problems in dimly lit areas. For operation in harsh environments, speed enables the user to obtain quality data with minimal personal exposure. The portable platform (123 and 207) typically includes a processor (124), a portable display (125 and 204), a portable power source (126), one or more data input and data output ports such as transmitter (127) and general purpose ports (206A, 206B), a field user interface (111, 200, 201, 202, 203, 205, 212, 213), a portable database (108), application instructions (109), one or more sensor ports (208A, 208B), and an alternate connector port (208C). General-purpose ports (206A, 206B) and alternate connector port (208C) may be serial ports, USB ports, or custom ports designed specifically for use in the portable platform (123 and 207). When connected to the sensor suite (122), the portable platform (123 and 207) is typically in one-way or two-way communication (112) with the sensor suite (122) through a sensor ports (208A, 208B) using wired or wireless mechanisms. In alternate embodiments general purpose ports (206A, 206B) or alternate connector port (208C) may be used in combination with, or in place of, sensor ports (208A, 208B) for connecting portable platform (123 and 207) to the sensor suite (122). In such embodiments general purpose ports (206A, 206B) and alternate connector port (208C) serve as a sensor interface. In some embodiments the portable platform (123 and 207) may receive data from sources other than a sensor suite. For example, the portable platform may be equipped for receiving image data and sensor data from a network server, from another portable platform, from a base station, or from another similar source. In the most preferred embodiments, a primary function of the portable platform (123 and 207) is capturing imaging sensor data (147A, 147B) and dynamic sensor data (146A, 146B). In some embodiments the process of capturing the data includes storing the data in the portable database (108). The portable platform (123 and 207) may also communicate with a base station (101) via one-way or two-way data transfer via electrical contact or wireless mechanisms such as an appropriate PCMCIA card interface (211). In preferred embodiments the portable platform (123 and 207) includes at least a transmitter for data transfer and the base station (101) includes at least a receiver for data transfer. In some embodiments the process of capturing data does not include storing the data in the portable platform (123 and 207); instead the data are transmitted directly to the base station (101) without storage in the portable platform (123 and 207). The base station (101) typically also includes a central processor, which is typically a conventional central processing unit but in alternate embodiments may be thin client processor, an application specific integrated circuit, or similar electronics. The base station (101) generally incorporates station digital memory, such as read only memory, random access memory, and media memory such as compact disc data storage. The base station (101) typically runs application software (106) which controls data input/output through sensor communication link (112) and (when used) base communication link (107). The application software (106) also accepts user input through user interface One embodiment of the portable platform employs a modified version of a equipment analyzer which already works in conjunction with a plurality of dynamic signal sensors including vibration sensor (119), sonic sensor (118), ultrasonic sensor (117), flux sensor (115), and current sensor (114). Major modifications to the equipment analyzer accommodate the preferred embodiment include more than one sensor ports (208A, 208B), application instructions (109), and portable database (108), to accommodate the addition of camera(s) (120, 121) and oil sensor(s) (116). The preferred embodiment for the portable platform employs carrying straps attached to slots (209A, 209B, 209C, 209D) on the portable platform. In the preferred embodiment a portable display (125, 204) is a transflective, color, liquid crystal display with selectable backlighting. One optional configuration is to use a touch-screen display. Another optional configuration is a microphone (213) for sound recording at least voice information when using the portable platform (123 and 207). Still another optional configuration includes a speaker (214) for audio output. In some embodiments it is envisioned that the portable platform (123 and 207) or base station (101) might include voice recognition and/or text-to-voice features to facilitate expanded user interface. In the preferred embodiment the portable display (125) functions as part of the user interface (111) providing visual communication to the user in the field. An optional configuration includes microphone and speaker so that the video interface is supplemented with audio output and sound recording for voice annotation. The preferred embodiment uses rechargeable batteries to provide portable power (126) to operate the portable platform (123) and may also be used to power sensor(s) (114, 115, 116, 117, 118, 119, 120, 121). Transmitter (127) is used to provide data output through electrical, wireless, visual, and possibly audio means. Sensor(s) (114, 115, 116, 117, 118, 119, 120, 121) connect to the portable platform by electrical or wireless mechanisms. In the preferred embodiment the portable database(s) (108) is(are) derived from that(those) used in a equipment analyzer with additional support as needed to accommodate visible and infrared cameras (120, 121) and oil sensor(s) (116). An alternate embodiment is to employ all of the database elements into the portable database. This database already supports imaging and dynamic signal analysis data sources, but generally requires modification to operate on the processor (124) rather than the central processor (102) as part of the optional base station (101). In this alternate embodiment, replication or a similar concept may be used to synchronize the portable database (108) with the stationary database (105) which may also be a derivative of the stationary database (105). In the preferred embodiment the application instructions (109) operate on the portable platform (123,207) enabling embedded intelligence, fast data collection, advanced bearing analysis using stress wave analysis methodology, reliable slow speed measurements, single or dual channel or multi-channel analysis, balancing, laser alignment, cascade, transient analysis, motor monitoring, imaging, and thermography. The user is directly notified about the nature of a developing fault at the time of measurement. This enables the user to focus attention on critical machine issues as soon as they are identified and collect additional diagnostic information while still at the machine site. In the preferred application instructions (109) embodiment the user can choose from a menu of special tests and it automatically configures itself for collecting additional data to focus in on the problem. Optional features for the application instructions (109) include demodulation, used for early detection tool due to its ability to isolate specific fault frequencies associated with the developing bearing or gear fault; and stress wave analysis, which goes beyond demodulation's ability to identify the fault by providing an objective, trend suitable measure of the fault severity. In one embodiment the application instructions (109) includes advanced digital technology to detect the stress waves generated by faults such as fatigue cracking, cracked gear teeth, abrasive wear, scuffing, or impacting in their earliest stages. The early and accurate detection provided by stress wave analysis results in improved maintenance planning, enabling the user to lower costs, decrease downtime, and reduce spare parts inventory. In one embodiment the application instructions (109) includes slow speed technology to take reliable readings with single or even dual integration as low as 10 RPM. In one embodiment the application instructions (109) includes cascade analysis which quickly captures a series of FFT spectra during startup or coast-down, which can be displayed as a waterfall plot. It can also be used for short-term continuous monitoring of critical machine problems. The application instructions (109) and portable platform (123 and 207) may support dual channel capability which can reduce data collection time by as much as 50%. The productivity gains alone typically justify the investment. Far beyond productivity improvements, the dual channel analyzer opens up new analysis possibilities to confirm faults such as misalignment, looseness, cracks, and structural resonance. The dual channel analyzer also provides filtered orbit analysis. The dual channel dynamic signal analyzer includes the advanced cross-channel program as a standard module to determine the root cause of a failure. Embedded intelligence makes cross channel analysis easy to use with minimal training. The companion software can be used to analyze and archive the results, plus provides a custom data export link to operating deflection shape and modal analysis software. Optional application instructions (109) programs are available for the dynamic signal analyzer may be used for transient, balancing and laser alignment. The optional advanced transient program turns the dynamic signal analyzer into a single or dual channel digital tape recorder with full analysis capabilities. Another optional application instructions (109) program supports the analyzer being used for shop or field balancing. In a particular embodiment the graphical interface makes operation simple and helps avoid the typical errors made in balancing setup. The program systematically removes background vibration while the balancing watchdog expert alerts the user to other conditions that could complicate the balance job. The application instructions (109) supports imaging and image analysis. In the preferred embodiment this includes visual imaging, thermal imaging. Optional thermographic image analysis includes display of radiometric images in selectable modes including grayscale, ironbow, rainbow, and/or other pallets. Typically the display field also includes a legend depicting temperatures to which colors and shades correspond. The user normally selects one or more points for which actual temperature is displayed in text as well as pixel color/shade. Stored images and video clips can be recalled for display, annotation, and editing in the field. Such images and video clips are examples of imaging sensor data. The application instructions (109) used in the preferred embodiment supports watch variable data collection including notes and observations representing visual, audible, and otherwise perceived observations logged by the technician using the portable platform. Each sensor port (208A, 208B) included in the portable platform (123 and 207) is configured to one or more of the array of sensors in the sensor suite (122). In most embodiments this includes signal interpretation. It typically also includes mechanisms for attachment and transport of the one or more sensors when it or they is or are not in use, and sensor communication (112) with each sensor during use. It may also include power supplied to the sensors for the purposes of either measurement or recharging sensor batteries. Connection between the portable platform (123 and 207) and the sensor suite (122) includes physical connection and data communication. In one embodiment physical connection and communication to the sensors makes use of one or more ports such as PCMCIA card interface (211). In another embodiment communication is wireless using either radio frequency or infrared data transfer mechanisms. Sensors may be attached to the portable platform (123 and 207) in a convenient location such as on the back of the analyzer package (220A or 220B or 220C or 220D or 220E). One optional embodiment includes a feature enabling printing to be accomplished directly from the portable platform (123 and 207). Typically this is done using application instructions to format the signal with printer instructions. Communication to the printer may be made through a printer port (210) or via wireless means or directly to an integral printing device. Optional connection between the portable platform and the base station (101) is provided for the broadest application of this invention. In the preferred embodiment, the base station (101) may include these elements: computer operating system (143), receiver (104), stationary database (105), computer display (142), central processor (102), application software (105), and connection and base communication link (107) to the portable platform. The computer operating system (143) is typically Microsoft Windows® or Microsoft Windows CE®, although other operating systems may be used. The central processor (102) is typically a desktop central processing unit. The base communication link (107) may be wired or wireless. The base station is an optional embodiment because all of these functions could be performed in the portable platform if one desires. One may prefer to maintain a stationary database (105) and all that goes with it to provide greater integration with other asset management database and software applications, to provide multi-user network, to provide memory backup, to provide extended memory and analysis tools, to allow the user to perform final analysis and reporting functions in a comfortable location, to allow others ready access to the information, for fuller integration with systems, and for other individual reasons. A standard desktop CPU is preferred as the central processor (102) in the base station (101), although other devices such as an application specific integrated circuit (ASIC) may be used. In the preferred embodiment, the printer (103) is in communication with the central processor (102) as part of the base station (101) system. Typically the printer drivers are provided with the base software (106) although one could easily connect or incorporate or simply communicate the printer, when used, directly with the portable platform (123 and 207). The printer (103) is one way to report results. Other, equally acceptable methods include electronic reporting via file transfer using data files or PDF style reports or other style reports. An alternate reporting method is email messaging or voice mail messaging or display messaging or other ways to let the intended recipient know about data and information derived from sensor inputs and personal observations. There are many embodiments for the receiver (104) on the base platform (101) including devices using file transfer protocol, replication, serial communication, manual data entry, and others. The base communication link (107) can include physical connection or wireless connection between the base station (101) and portable platform (123 and 207). This base communication link (107) may include linkage through an Ethernet or internet or intranet or other network. The preferred embodiment for base station software and stationary database are extended versions of commercially available reliability software (106) and stationary database (105). Logical upgrades to these commercially available tools are required to support the invention. An alternate embodiment uses similar software and database structure on both the portable platform (123 and 207) and base station (101), including replication or another method for synchronizing database information between the two systems. Another alternate embodiment uses wireless Ethernet or similar communication through the base communication link (107) such that the stationary database (105) receives data directly or shortly after it is collected in the field. In this case the portable database(s) (108) can be very small or nonexistent since the data is being directly stored onto the base station (101). One can envision that by using Ethernet or internet or other on-line communication as at least part of base communication link (107) between the portable platform (123 and 207) and base station (101) that the distinction between base software (106) and application instructions (109) can shift such that the base software can supplement or replace or update the functions envisioned in this invention for the application instructions (109). In the same way, the database functions may be performed in the stationary database (105) or the portable database (108) or some combination of both. By using base communication link (107), some or all of the software functions (106 and 109), the database functions (105, 108), and optional printer function (103), and even functions of the sensor ports (208A, 208B) may be performed using either the portable platform (123 and 207) or the base station (101) or some combination of both. A physical embodiment of a portable platform is depicted in FIGS. 2 and 3 as portable platform (207). FIG. 2 portrays the front (or top) of the portable platform (207), and FIG. 3 portrays the back (or bottom) of the portable platform (207). A display (204) performs a plurality of functions including display of header information (205) for location, equipment information, summary, or other information used to orient the user to the type of information on the remaining portions of the display. Another function of the display (204) in the preferred embodiment is the allocation of one or more windows (203A, 203B, 203C, 203D) of the display (204) to reporting measurements, images and graphics to facilitate translating measured data into useful information. In this application, a window refers to any pop-up or overlay or highlighted area or sector or portion or otherwise set-aside vicinity of the display in which a particular function is performed. The windows (203A, 203B, 203C, 203D) of the display (204) may depict vibration spectrum, vibration waveforms, bar graph, visual images, trend plots, tabular data, setup information, etc. These display windows (203A, 203B, 203C, 203D) are easily adapted for dynamic signal analysis and for imaging analysis functions. Functions historically performed on the display of an infrared camera may be performed on the display windows (203A, 203B, 203C, 203D) including but not limited to the following list: live thermal image, live visual image, frozen visual image, frozen thermal image, text annotation, graphic annotation, temperature at cursor points, temperature histogram, temperature profile, alarms, parameters, user instructions, etc. Functions historically performed on the display of a dynamic signal analyzer may be performed on the display windows (203A, 203B, 203C, 203D). These include but are not limited to active waveform, active spectrum, individual waveform, individual spectrum, trend, data table, alarms, parameters, graphic representation, dynamic or frozen modal analysis, demodulated spectrum or waveform, cascade, transient, etc. Another display function in the preferred embodiment it to provide dynamic or changeable function selections (212) corresponding to function keys (200). In the case of a touch-screen display (204) the function keys (200) might be combined with function key descriptions (202). Preferred embodiments use sensor ports (208A, 208B) for data input and output from sensors. Alternately other ports such as the PCMCIA card interface (211) or even printer port (210) may also be used. The preferred embodiment provides for user supplied data input through field user interface (111 in FIG. 1) and output through the portable display (204). In preferred embodiments, field user interface (111) includes dynamic function keys (200), function descriptions (202), keypad interface (201), changeable function selections (212), display windows (203A, 203B, 203C, 203D), header information (205), and microphone (213). The user may also provide input via one or more of the sensor (114, 115, 116, 117, 118, 119, 120, or 121 in FIG. 1) in communication with the portable platform through the portion of the sensor communication link (112 in FIG. 1) serving such sensor(s). The portable platform (123 and 207) is substantially contained in a housing (215). One or more strap slots (209A, 209B, 209C, 209D) may be provided in the housing (215) to facilitate attaching one or more carrying straps to the platform, or to facilitate tie-downs for retaining the portable platform in a fixture or carrying case. FIG. 3 also illustrates 5 storage spaces (220A or 220B or 220C or 220D or 220E, 221, 222, 223, and 224). These spaces are recesses in the back/bottom of portable platform (123 and 207) which are specifically dimensioned to hold sensors or other accessory devices used with the portable platform (123 and 207). An example of an application of embodiments is depicted in the switchyard diagram of FIG. 4. Various components of the switchyard are depicted. These are (from right to left) high voltage transmission lines (255), disconnect (254), transformer (253), potential and current transformers (252), circuit breaker (251), and low voltage transmission lines (250). Table 1 shows how a portable platform with multiple sensor technologies may be used to inspect a switchyard. TABLE 1Inspection of switchgear with an integrated system.Sensor ApplicabilityOilComponentIRVisibleUltrasonicVibrationAnalysisPower lineYesYesYesNoNoConnectionYesYesYesNoNoInsulatorYesYesYesNoNoBushingYesYesYesNoYesJunctionYesYesYesNoNoCouplingYesYesYesNoNoDisconnectYesYesYesNoNoCurrent transformerYesYesYesNoYesDisconnectYesYesYesNoNoMain transformerYesYesYesNoYesLoad tap changerYesNoYesNoNoBreakerYesYesYesNoYesOperating statusYesYesNoNoNoCooling systemYesYesNoNoNoMotorYesYesNoYesNoPumpYesYesNoYesNoGas compressorYesYesYesYesYes Table 1 presents a matrix of different combinations of sensors that may beneficially be employed to measure the health of particular components in an electrical switchyard. Various embodiments may be used to check each of the components listed in each row of the Table 1 by incorporating sensor capability identified in the columns where “yes” is listed in that row. Matrix elements labeled “No” indicate that the sensor in that column is generally not applicable for inspecting the component in that row. However, under special circumstances such use may be appropriate. FIG. 5 illustrates a method embodiment of the invention. The method begins with a storing application instructions in an instrument step (302). Then two additional processes are conducted. One of these processes involves a receiving and storing imaging sensor data in the instrument step (304) followed by a deriving imagery indication of equipment health step (306). The other of these two additional processes includes a receiving and storing dynamic sensor data in the instrument step (308) followed by a deriving dynamic indication of equipment health step (310). It is cautioned that while FIG. 5 might incorrectly be interpreted to show that the two additional processes are conducted in concurrently in parallel, the two additional processes do not necessarily have to be conducted concurrently in parallel (although they may be). For example, the receiving imaging sensor data in the instrument step (304) and the deriving imagery indication of equipment health step (306) may be completed before beginning the receiving dynamic sensor data in the instrument step (308) and the deriving dynamic indication of equipment health step (310), or steps (308) and (310) may be completed before beginning steps (304) and (306). Alternately, the receiving imaging sensor data in the instrument step (304) may be conducted first, followed by the receiving dynamic sensor data in the instrument step (308). Using the reverse sequence of those two steps is also possible. However, the deriving imagery indication of equipment health step (306) must be preceded (although not immediately preceded) by the receiving imaging sensor data in the instrument step (304), and the deriving dynamic indication of equipment health step (310) must be preceded (although not immediately preceded) by the receiving dynamic sensor data in the instrument step (308). After the deriving imagery indication of equipment health step (306) and the deriving dynamic indication of equipment health step (310) are completed, the step (312) of correlating imagery indication of machine health with dynamic indication of machine health may be undertaken and completed. FIG. 6 illustrates a different method embodiment of the invention. The method begins in a manner similar to the method of FIG. 5 with a storing application instructions in an instrument step (322). Then two further processes are conducted. One of these processes involves an acquiring imaging sensor data with an imaging sensor step (324), followed by a receiving and storing imaging sensor data in the instrument step (326) followed by a deriving imagery indication of equipment health step (326). The other of these two further processes includes acquiring dynamic sensor data with a dynamic sensor step (330), a receiving and storing dynamic sensor data in the instrument step (332), followed by a deriving dynamic indication of equipment health step (334). As with FIG. 5, the two further processes do not have to be conducted concurrently in parallel. Steps (324), (326) and (328) may be completed before beginning steps (330), (332) and (334), or vice versa. Alternately, the acquiring image data with an imaging sensor step (324) may be conducted first, followed by the acquiring dynamic sensor data with a dynamic sensor step (330). Using the reverse sequence of those two steps is also possible. However, sequence of steps following the acquiring image data with an imaging sensor step (324) have to proceed sequentially (but not immediately) in the order shown, and the sequence of steps following the acquiring dynamic sensor data with a dynamic sensor step (330) have to proceed sequentially (but not immediately) in the order shown. After the deriving imagery indication of equipment health step (328) and the deriving dynamic indication of equipment health step (334) are completed, the correlating imagery indication of machine health with dynamic indication of machine health step (336) may be undertaken and completed. FIG. 7 illustrates a further alternate method, one that incorporates the use of a base station. The method begins with a storing application instructions in an instrument step (342) and a storing application software in a base station step (342). These steps may be completed in any order. Then two subsequent processes are conducted. One of these processes involves the step (346) of capturing imaging sensor data with an imaging sensor operatively connected to the instrument and using the instrument to transmit the imaging sensor data to the base station, followed by the step (348) receiving and storing the imaging sensor data in the base station, followed by the step (350) of deriving imagery indication of equipment health in the base station. The other of these two subsequent processes includes the step (352) of capturing dynamic sensor data with a dynamic sensor operatively connected to the instrument and using the instrument to transmit the dynamic sensor data to the base station, followed by the step (354) of a receiving and storing dynamic sensor data in the base station, followed by a deriving dynamic indication of equipment health step (356). As with FIGS. 5 and 6, the two subsequent processes do not necessarily have to be conducted concurrently in parallel. However, sequence of steps following the step (346) of capturing imaging sensor data with an imaging sensor and transmitting the imaging sensor data to the base station have to proceed sequentially (but not immediately) in the order shown, and the sequence of steps following step (352) of capturing dynamic sensor data with a dynamic sensor and transmitting the dynamic sensor data to the base station have to proceed sequentially (but not immediately) in the order shown. After the step (350) of deriving imagery indication of equipment health in the base station and step (356) of the deriving dynamic indication of equipment health in the base station are completed, the correlating imagery indication of machine health with dynamic indication of machine health step (358) may be undertaken and completed. The foregoing description of preferred embodiments for this invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise form disclosed. Obvious modifications or variations are possible in light of the above teachings. The embodiments are chosen and described in an effort to provide the best illustrations of the principles of the invention and its practical application, and to thereby enable one of ordinary skill in the art to utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. All such modifications and variations are within the scope of the invention as determined by the appended claims when interpreted in accordance with the breadth to which they are fairly, legally, and equitably entitled.
	summary
040295451	claims	1. A nuclear fuel element comprising (a) a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof and (b) an elongated composite cladding container comprising an outer layer of a material selected from the group consisting of zirconium or a zirconium alloy having two coatings bonded on the inside surface with the first coating being an undeformed diffusion barrier of constant thickness selected from the group consisting of chromium and chromium alloys and the second coating being an undeformed metal layer of constant thickness selected from the group consisting of copper, nickel, iron and alloys thereof, said diffusion barrier being about 0.00005 to about 0.001 inch in thickness, said metal layer being about 0.0001 to about 0.002 inch in thicknes, each of said coatings having an impurity content of less than about 1 weight percent, and said cladding container enclosing said core so as to leave a gap between said core and said container during use in a nuclear reactor. 2. A nuclear fuel element of claim 1 which has in addition a cavity and a nuclear fuel material retaining means in the form of a helical member positioned in the cavity. 3. A nuclear fuel element of claim 1 in which the metal layer is copper. 4. A nuclear fuel element of claim 1 in which the metal layer is nickel, 5. A nuclear element of claim 1 in which the metal layer is iron. 6. A nuclear fuel element of claim 1 in which the outer layer is a zirconium alloy. 7. A nuclear element of claim 1 in which the diffusion barrier is chromium. 8. A nuclear fuel element of claim 1 in which the diffusion barrier is a chromium alloy. 9. A nuclear fuel element of claim 1 in which each of said coatings has an impurity content of less than about 1000 parts per million. 10. A nuclear fuel element of calim 1 in which the nuclear fuel material is comprised of uranium dioxide. 11. A nuclear fuel element of claim 1 in which the nuclear fuel material is a mixture comprising uranium dioxide and plutonium dioxide. 12. A nuclear fuel element which comprises an elongate composite cladding container of an outer layer of a material selected from the group consisting of zirconium or a zirconium alloy having two coatings bonded on the inside surface with the first coating on the outer layer being an undeformed diffusion barrier of constant thickness selected from the group consisting of chromium and chromium alloys and the second coating on the first coating being an undeformed metal layer of constant thickness selected from the group consisting of copper, nickel, iron and alloys thereof, said diffusion barrier being about 0.0005 to about 0.001 inch in thickness, said metal layer being about 0.0001 to about 0.002 inch in thickness each of said coatings having an impurity content of less than about one weight percent, a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof disposed in and partially filling said container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container, a nuclear fuel material retaining means positioned in the cavity, and said cladding container enclosing said core so as to leave a gap between said core and said container during use in a nuclear reactor.
	description	The present invention relates to a technology for monitoring nuclear reactor power during operation. In the boiling water reactor (BWR), output power can be controlled by changing a core flow and thereby changing a steam ratio (void fraction) inside a boiling reactor core. However, it is known that depending on the core flow and other operating conditions, neutron flux distribution and liquidity in the reactor core are destabilized by delayed transportation of voids and a negative feedback effect caused by negative void reactivity coefficients in the reactor core. There is concern that occurrence of such a nuclear thermal hydraulic destabilization phenomenon may result in considerable oscillation of output power and flow rate, which may deteriorate cooling characteristics in terms of fuel rod surface temperature and may damage the soundness of fuel rod cladding tubes. Accordingly, in designing fuels and reactor cores for the boiling water reactor, the nuclear thermal hydraulic stability is analyzed to produce a design that gives sufficient margin to stability so as to prevent such an oscillation phenomenon from occurring in any of the expected operating ranges. In such a range where deterioration in nuclear thermal hydraulic stability is expected, limited operation is preset for safety. Nuclear reactors of some types are provided with a safety setting so that in the unlikely event where the nuclear reactor reaches the operation limited range, output power is lowered by insertion of control rods and the like so that the nuclear reactor can get out of the operation limited range. There are a large number of nuclear power plants which allow, from a viewpoint of Detect and Suppress, power oscillation phenomena while accurately detecting the power oscillation phenomena attributed to nuclear thermal hydraulic destabilization and suppressing the oscillations before the fuel soundness is damaged. Accordingly, a power oscillation detection algorithm with use of dedicated detection signals for detecting the power oscillation phenomenon, which is referred to as OPRM (Oscillation Power Range Monitor), has been proposed (e.g., Patent Literature 1). There is also known a technique to analyze principal components in an oscillation based on a plurality of nuclear instrumentation signals, extract independent components different in oscillation modes, and to evaluate core-wide stability and regional stability which are stability different in oscillation modes (e.g., Patent Literature 2). There is also known a technique to evaluate nuclear thermal hydraulic stability in consideration of parameters such as reactor core average neutron flux measurement values (APRM), delay corresponding to a heat-transfer time constant in fuel rods, and main steam flow rate measurement values (e.g., Patent Literature 3). Further, there is known an apparatus adapted to perform prediction analysis of stability based on decay ratios and to issue an alarm when sequentially detected stability of the reactor core exceeds a predicted value (e.g., Patent Literature 4). There is also known a technology to successively calculate deterioration indexes of nuclear thermal hydraulic stability in a boiling water reactor based on plant information, such as reactor core power distributions, reactor core flows, reactor core pressure and feed water temperature, and to issue an alarm when these indexes exceed preset values (e.g., Patent Literature 5). In addition, there is known a method for monitoring stability based on neutron flux space mode distributions obtained by calculation of reactor core characteristic values using a large number of LPRM signals (e.g., Patent Literature 6). Patent Literature 1: U.S. Pat. No. 5,555,279A Patent Literature 2: JP2002-221590A Patent Literature 3: JP2002-181984A Patent Literature 4: JP2000-314793A Patent Literature 5: JP2000-121778A Patent Literature 6: JP1999-231089A As the boiling water reactors are made to have a larger size, a higher power density and a higher burn-up, their nuclear thermal hydraulic stability is generally declined. However, in Patent Literature 1, a measure is not taken for such problem of the boiling water reactors. While an allowance of nuclear thermal hydraulic stability is inevitably declined in connection with increase in reactor core power and power density, enhancement of monitoring accuracy in nuclear thermal hydraulic stability is now demanded more than before. However, in the cases of Patent Literature 2 to Patent Literature 6, it is not possible to enhance the monitoring accuracy in the nuclear thermal hydraulic stability more than before, while an allowance of nuclear thermal hydraulic stability is inevitably declined in connection with increase in reactor core power and power density. The present invention has been made in order to solve the above-mentioned problems and it is an object of the present invention to provide a technology for enhancing the monitoring accuracy and reliability in nuclear thermal hydraulic stability of a nuclear reactor. First Embodiment The embodiments of the present invention will be described hereinbelow with reference to the accompanying drawings. A nuclear power generation system shown in FIG. 1 includes: a nuclear reactor 10 which heats furnace water by the heat generated through nuclear fission of nuclear fuel and thereby generates steam; a main line 21 which guides the generated steam to a turbine 22; a generator 23 coaxially connected with the turbine 22 which is rotationally driven by the steam to convert rotational kinetic energy to electric energy; a condenser 24 which cools and condenses the steam, which was expanded in the process of doing its work in the turbine 22, into condensate water; and a water supply line 26 which sends the condensate water to the nuclear reactor 10 with a pump 25. Feed water returned to the nuclear reactor 10 is reheated as furnace water, and the above-stated process is repeated to perform continuous power generation. To sustain the power generation in a stable manner, a nuclear reactor power monitor 30(50) is provided. The nuclear reactor 10 includes: a pressure vessel 11 filled with furnace water and provided with a shroud 15 fixed to the inside thereof; a core support plate 17 fixed to the shroud 15; a reactor core 16 enclosed by the shroud 15 which is supported by the core support plate 17; and a steam separator 13 which performs gas-liquid separation of the furnace water which has been changed into a gas-liquid two-phase flow by passing through the reactor core 16. The steam as the one product obtained by steam separation in the steam separator 13 is guided to the main line 21 as described above so as to contribute to power generation, while the other product obtained as separated water joins the feed water returned through the water supply line 26. The thus-joined furnace water is made to flow down an area (downcomer D) between the shroud 15 and the pressure vessel 11 with a plurality of recirculation pumps 18 (only one pump is described in the drawing) provided in a circumferential direction, and is guided to a lower plenum area L. The furnace water guided to the lower plenum area L again passes the reactor core 16, where the water is heated into a gas-liquid two-phase flow before reaching an upper plenum area U. The gas-liquid two-phase flow that reached the upper plenum area U is again guided to the steam separator 13, where the aforementioned process is repeated. As shown in a horizontal sectional view of FIG. 2, the reactor core 16 includes: a square cylinder-shaped fuel assembly 33 containing a large number of fuel rods (omitted in the drawing); a control rod 32 which absorbs neutrons generated by a nuclear fission reaction to control output power; and an instrumentation pipe 34 whose upper and lower ends are respectively fixed to an upper grid plate 14 and the core support plate 17 and which supports nuclear instrumentation detectors 31 (31A, 31B, 31C, 31D) for detecting the neutrons as shown in FIG. 1. A large number of these component members are arrayed to structure the reactor core 16. One instrumentation pipe 34 is generally provided for 16 fuel assemblies 33. For example, an advanced boiling water reactor including 872 fuel assemblies is equipped with 52 instrumentation pipes 34. The nuclear instrumentation detectors 31A, 31B, 31C, 31D provided at four positions in a perpendicular direction of the instrumentation pipe 34 are each referred to as a level A, a level B, a level C, and a level D in accordance with a height position from the lower side. The furnace water which circulates inside the reactor core 16 flows into the furnace from the level A, where the furnace water is heated with fuel and starts to boil. The furnace water reaches the level B, the level C, and the level D in sequence while its water/steam two-phase state is being changed. The nuclear thermal hydraulic stability is greatly influenced by pressure propagation in the water/steam two-phase state. More specifically, due to a delay in pressure propagation of the furnace water which flows from the lower side to the upper side in the reactor core 16 as shown in FIG. 1, the two-phase state (water and steam ratio) is changed. This causes a response delay of the nuclear instrumentation detectors 31A, 31B, 31C, 31D, which in turn causes phase difference between the respective nuclear instrumentation signals S (SA, SB, SC, SD) detected at the level A, the level B, the level C, and the level D. Such phase difference in power oscillations in a flow direction of furnace water has a mechanism of causing mutual cancellation of the responses of the nuclear instrumentation signals S. Therefore, from the viewpoint of accuracy and reliability in monitoring the nuclear thermal hydraulic stability, it is preferable that a plurality of the nuclear instrumentation signals S at the same height level are grouped and evaluation is performed for each group. The necessity of performing stability monitoring on all the levels from the level A to level D is low. Accordingly, in each of the embodiments, evaluation of the nuclear thermal hydraulic stability is performed by targeting a level B group, which is generally said to have the highest sensibility for stability monitoring. The power oscillations relating to the nuclear thermal hydraulic stability are a macroscopic phenomenon which occurs in the entire reactor core due to destabilization of flow conditions inside a fuel channel which encloses the fuel assembly 33, the destabilization being caused by reactivity feedback to dynamic responses of neutron fluxes. It is considered that the reactivity feedback excites a neutron flux space mode, which results in occurrence of power oscillations. When the excited space mode is a basic mode, the power oscillations caused thereby are called core-wide oscillations. The core-wide oscillations basically have the same phase in each of the reactor core cross section at the same height level. In this case, a plurality of nuclear instrumentation signals S measured in the same cross section have almost no phase difference from each other. They are not cancelled by addition, and therefore oscillations can sufficiently be detected with use of average power range monitor (APRM) signals. In contrast, when the excited space mode is a higher order mode, the oscillations thereby are called regional oscillations. According to the higher order space mode distribution, the nuclear instrumentation signals S in the reactor core cross section at the same height have phase difference from each other. With a node of the higher order space mode distribution as a center line of oscillations, 180-degree phase difference appears across the center line, and oscillations are reversed at this center line. FIG. 3A shows a higher order space mode distribution in the regional oscillations. As shown in the horizontal cross sectional view of FIG. 3B, two areas a and b across an oscillation center line c, which corresponds to a node, are opposite in phase from each other, i.e., they oscillate with 180-degree phase difference from each other. In this case, if a plurality of the nuclear instrumentation signals S across these two areas a and b are averaged, oscillations are cancelled due to the phase difference. Accordingly, the amplitude of the averaged signals is flattened and this makes it difficult to detect oscillations. In short, it is not suitable for detection of such regional oscillations to use the APRM signal outputted as a signal formed by averaging all the reactor core signals. Although not shown in the drawings, use of the APRM signal is also unsuitable in the case of detecting local power oscillations which occur in a narrow area centering around a certain specific fuel assembly 33 (FIG. 2). As shown in FIG. 2, a power monitor 30 includes a first calculation unit 42 configured to calculate a first stability index based on time series data Xt (FIG. 6) indicating power oscillation in nuclear instrumentation signals S outputted from a plurality of nuclear instrumentation detectors 31 which detect neutrons in a reactor core 16; a first determination unit 44 configured to compare the first stability index and a first reference value and determine whether nuclear thermal hydraulic stability of the reactor core 16 is stable or deteriorated; a second calculation unit 46 configured to calculate a second stability index of the reactor core 16 based on the time series data Xt when the deteriorated state is determined in the first determination unit 44; and a second determination unit 47 configured to compare the second stability index and a second reference value and determine whether to perform suppressing operation of the power oscillation. The power monitor 30 is further configured to include a peak detection unit 45, a third determination unit 48, and a statistical processing unit 40. A grouping unit 41 is configured to divide the nuclear instrumentation detectors 31 into groups based on information (such as power distributions of the reactor core, higher order space mode distributions of neutron fluxes, or specified positions of fuel assembly 33) transmitted from a state estimation unit 60. It is to be noted that the nuclear instrumentation signals S processed in the later-described first determination unit 44 and the second determination unit 47 may be individual signals of the nuclear instrumentation detectors 31 selected from a group, or an average signal of the nuclear instrumentation detectors 31 in units of a group. The state estimation unit 60 includes a process calculator 61, a data base 62, and a nuclear thermal hydraulic simulator 63. The thus-configured state estimation unit 60 transmits information, such as reactor core power distributions estimated based on a physical model or a data base, higher order space mode distributions of neutron fluxes, and placement information on the nuclear instrumentation detectors 31, to the grouping unit 41. The statistical processing unit 40 is configured to apply a statistical method, such as an autoregression analysis method, an autocorrelation function method or a spectrum-analysis method, to time series data Xt (FIG. 6) indicating power oscillation in the nuclear instrumentation signals S outputted from the nuclear instrumentation detectors 31 which detect neutrons in the reactor core. The first calculation unit 42 is configured to calculate the first stability index based on the time series data in the nuclear instrumentation signals S. In a variation analysis unit 49 included in the first calculation unit 42, a variance or a standard deviation σT(t) is derived as the first stability index which indicates variations in oscillation period T in a plurality of the time series data Xt. It is to be noted that time series data Xt including an oscillation period varied beyond a fixed range is excluded as an outlier in calculation of the standard deviation σT(t). A reference value storage unit 43 is configured to store a first reference value, a second reference value and a third reference value which are used as the reference value in each of the first determination unit 44, the second determination unit 47 and the third determination unit 48. The first determination unit 44 is configured to compare the first stability index and the first reference value and determine whether nuclear thermal hydraulic stability of the reactor core is stable or deteriorated. If it is determined that the stability is stable, monitoring based on operation of the variation analysis unit 49 is continued, whereas if it is determined that the stability is deteriorated, the operation is then shifted to monitoring by peak detection. Even after the operation is shifted to the monitoring by peak detection, and the peak detection unit 45 is in operation, the monitoring by the variation analysis unit 49 is concurrently continued. The first determination unit 44 determines that the nuclear thermal hydraulic stability is deteriorated when the first stability index or the standard deviation σT(t) of oscillation periods exceeds the first reference value for a predetermined time in succession. The peak detection unit 45 is configured to execute peak detection of the time series data when it is determined in the first determination unit 44 that the stability is deteriorated. The peak detection unit 45 fits the time series data set on intervals by a polynomial and searches for a point where a derivative value of the polynomial is equal to zero as a peak. A spline function is applied as the polynomial. In fitting the time series data to the spline function, it is desirable to conduct spline interpolation at a sampling interval used when the nuclear instrumentation signals S are converted into digital data. The point where the derivative value of the polynomial is equal to zero is obtained as follows. First, the zero point is searched within a period of time, composed of half of the oscillation period and a margin, with a switchover point from the first determination unit 44 as an origin. Then, out of a plurality of intervals which are interposed in between data points that constitute the time series data, an interval in which a product of derivatives of the data points placed on both sides thereof is a minus-sign product is obtained. The obtained interval is further divided, and out of these dividing points, a point where an absolute value of a derivative is the minimum is searched as the zero point. Next, the above-stated search flow is repeated within a period of time, composed of half of the oscillation period and a margin, with the searched point as an origin so as to obtain the zero point. The second calculation unit 46 is configured to calculate the second stability index of the reactor core based on the result of peak detection of the time series data. Here, an amplitude or a decay ratio of the plurality of the time series data is used as the second stability index. In short, the searched peak position (time axis) is substituted into the spline function of that interval to have a peak value (vertical axis). A difference between the peak value and a next peak value obtained in the same manner is used as an amplitude, and a time difference between the adjacent peak positions is used as an oscillation period. The second determination unit 47 is configured to compare the second stability index and the second reference value and to determine whether to operate a power suppression device 35. When the second determination unit 47 does not make a determination to operate the power suppression device 35 in a predetermined time, operation of the second calculation unit is stopped. It is to be noted that the time taken for stopping the operation of the second calculation unit 46 is desirably set longer than the time taken for the first determination unit 44 to make a determination. While the second calculation unit 46 is operated, the first calculation unit 42 is also concurrently operated. Accordingly, when a determination to execute the reactor core power control is not made by the second determination unit in a predetermined time, operation of the second calculation unit 46 is stopped and only the operation of the first calculation unit 42 is continued. Even when the second determination unit 47 determines that operation of the power suppression device 35 is not necessary based on the second stability index, the third determination unit 48 still determines that operation of the power suppression device 35 is necessary if the first stability index satisfies the third reference value which is set more severely than the first reference value. In response to the determination by the first determination unit 44, the second determination unit 47 and the third determination unit 48, any of automatic activation signals of an alarm device (omitted in the drawing), an oscillation information providing device (omitted in the drawing) and the power suppression device 35 may be issued in stages as a power oscillation suppression operation. Now, with reference to FIG. 4, a decay ratio, an oscillation period, and amplitude will be defined by using an oscillatory impulse response at the time of applying disturbance to a system. Assuming that peaks of the impulse response are set in order as X1, X2, X3, X4, . . . , and their appearing time are each set as t1, t2, t3, t4, . . . , the decay ratio, the oscillation period, and the amplitude, which are generally used as indexes indicating the stability of the nuclear thermal hydraulic stability, are defined as follows:Decay ratio=(X3−X4)/(X1−X2)Oscillation period=(t3−t1) or (t4−t2)Amplitude=(X3−X4) or (X1−X2) As for the phase difference, a time difference in tn between a plurality of signals is defined as an angle with one period being 360 degrees. If the decay ratio is less than 1, the impulse response is attenuated and therefore the system is stable, whereas if the decay ratio is more than 1, oscillations grow and the system becomes unstable. When the decay ratio is 1, the oscillations continue with constant amplitude. With a shorter oscillation period, oscillations grow or attenuate more quickly. An inverse of the oscillation period is generally referred to as a resonance frequency or a natural frequency, which is expressed in the unit of Hz or cps. While a graph view of FIG. 4 shows an ideal impulse response, a response of an actual nuclear instrumentation signal S does not form an ideal impulse response as shown in FIG. 5. More specifically, in the actual nuclear instrumentation signals S, high-frequency noise and low-frequency trends may be superposed or may make iterated vibration as shown in FIG. 5, so that monotonous attenuation and growth are not demonstrated in some cases. Accordingly, even if peaks are directly detected from the response of the actual nuclear instrumentation signal S and the decay ratio is calculated based on the amplitude obtained, it is impossible to accurately estimate the decay ratio since the response used is not a genuine impulse response of the system. Therefore, in order to calculate the impulse response, it is necessary to estimate a transfer function of the system, and for the estimation, it is necessary to apply a statistical method using a certain data length (a function of the statistical processing unit 40 of FIGS. 2 and 6). The decay ratio derived from the response of the nuclear instrumentation signal S of FIG. 5 with use of the thus-estimated transfer function is about 0.7 to 0.8. It can be said that the system is stable according to the response though growth and attenuation is repeated therein. In this case, unless fuel soundness is threatened by power oscillations, it suffices to monitor the system to confirm the instability thereof does not progress, and additional operation to stabilize the system is not particularly needed. The decay ratio is appropriate in the point that it is an index capable of directly evaluating the nuclear thermal hydraulic stability. However, accurate estimation of the decay ratio requires a certain data length. When the oscillation state depends on the reactor core position as in the case of regional oscillations, the value of the decay ratio varies depending on the nuclear instrumentation signals S to be observed. Accordingly, it becomes difficult to judge whether to conduct additional operation for improving core stability by activating a device such as the power suppression device 35 (FIG. 2). In the present embodiment, the statistical processing unit 40 and the first calculation unit 42 are used, and an oscillation period of the respective nuclear instrumentation signals S is applied as a parameter for monitoring deterioration of the stability. A plurality of nuclear instrumentation signals S divided into groups by the grouping unit 41 are analyzed by the variation analysis unit 49. Examples of indexes indicating variations include a variance and a standard deviation. Herein, the standard deviation σT(t) is adopted since its determination criterion is easily to select. The nuclear instrumentation detectors 31 (LPRM: Local Power Range Monitor system) are divided into groups as described later. Processing by the statistical processing unit 40 is performed in units of a group to obtain an oscillation period of each signal and a standard deviation σT(t) within each group. The standard deviation σT(t) is compared, in the first determination unit 44, with the first reference value in the reference value storage unit 43, and if the determination criterion is satisfied, the peak detection unit 45 is activated with the time of the determination as a reference. After the peak detection unit 45 is activated, the analysis of oscillation period variations by the first determination unit 44 is performed concurrently. With reference to FIG. 6A, description is given of a method for obtaining the oscillation period by a statistical method. First, digital processing and processing for removing noise and trend components are applied to nuclear instrumentation signals S to prepare time series data Xt. The oscillation period is obtained by a method such as a method for obtaining an autocorrelation function directly from the time series data Xt and setting a delay time with which the correlation function has a maximum value as the oscillation period, a method for obtaining spectral density by a method such as FFT (Fast Fourier Transform) and the autoregressive method and setting an inverse of a frequency (resonance frequency, Hz) at which the spectral density has a maximum value as the oscillation period, a method for obtaining a transfer function by the autoregressive method and obtaining the oscillation period from a resonance frequency estimated from a transfer function pole, and a method for obtaining an impulse response by the autoregressive method and obtaining the oscillation period based on a relation shown in FIG. 4. When any one of these method is used, a certain data length (including several oscillation periods or more) is needed in order to achieve accurate estimation. The autocorrelation function is a covariance of the time series data Xt at a certain time t and its past value Xt-1. A delay time with which the covariance value has a maximum value is equivalent to the oscillation period. In the autoregressive method, the time series data Xt is subjected to linear fitting such as Formula (1) of FIG. 6 to estimate an autoregression coefficient ak (FIG. 6(2)). In Formula (1), et represents Gaussian noise excluded from the fitting. Some algorithms which efficiently estimate such an autoregressive process have been suggested and widely used. A time-series temporal response characteristic is reflected upon an autoregression coefficient. By using this coefficient, a decay ratio, an oscillation period, and a phase difference can be obtained. Examples of the methods for obtaining the decay ratio and the oscillation period with the autoregression coefficient include those involving estimation from each of an impulse response, a spectral density and a transfer function. First, the impulse response is obtained by Formula (3) of FIG. 6 using the autoregression coefficient. The impulse response is the response shown in FIG. 4, with which the decay ratio and the oscillation period can be obtained based on Formula (4) of FIG. 6. To suppress estimation variations, an average value is used as an estimated value with respect to the decay ratio and the oscillation period. An autoregressive process is expressed as shown in Formula (5) of FIG. 6 where a delay operator Z−1(Xt-1=Z−1Xt) at discrete (digital) time is used. An inverse of A(Z−1) shown in Formula (7) of FIG. 6 is a transfer function from Gaussian noise to time series, and stability information is included in this transfer function. If this transfer function is used, a spectral density function S(f) will be given by Formula (6) of FIG. 6. A frequency fmax at which this spectrum becomes the maximum is the resonance frequency, and an inverse 1/fmax thereof serves as the oscillation period. A zero point of A(Z−1) corresponds to a pole of the transfer function. Since stability is determined by the positional relation of the pole on a complex plane, the decay ratio and the oscillation frequency can be estimated through estimation of the transfer function pole. When transfer function pole=(PR, PI) is set as in FIG. 6(8), the decay ratio and the oscillation period are expressed as in FIG. 6(10) based on a relation shown in Formula (9) of FIG. 6. In this case, Δt represents a sampling period of time series signals. Now, a group including total N nuclear instrumentation signals S is considered. An oscillation period of the i-th signal in this group is defined as Ti(t). Each oscillation period Ti(t) is calculated by using the statistical method described above. FIG. 7(A) shows a response example of each oscillation period Ti(t) calculated in the first calculation unit 42 based on a plurality of nuclear instrumentation signals S indicating a stable state, while FIG. 7(B) is a response example in the process of shifting to an unstable state. Although FIG. 7(A) and FIG. 7(B) are separately presented as upper and lower drawings, they show a consecutive result. The drawings indicate that in the stable state, variations in respective oscillation periods over time and variations among signals are both notable, whereas in the unstable states, both of these variations are small. In the variation analysis unit 49 (FIGS. 2 and 6), a variation (standard deviation) σT(t) of the oscillation periods in a group is calculated. It is to be noted that those having a top bar within a reference sign 49 of FIG. 6 represent an average value of the oscillation periods at a target time interval (represented by t) in a group. FIG. 8 shows a response of a standard deviation σT(t) of the oscillation periods of FIG. 7. As shown in the drawing, the standard deviation value falls at the point past 450 sec and converges into a constant value at the point around 750 sec where oscillation has sufficiently grown. This is considered because the nuclear thermal hydraulic stability depends on instability of respective fuel channels and on dynamic instability represented by density wave oscillation. In short, pass time of a two-phase flow is different in every fuel channel due to difference in a two-phase flow dynamic state, and the resonance frequency of density wave oscillation is typically intricate and different in every fuel channel. However, it is considered that a nonlinear frequency locking occurs in the process of macroscopic growth of an instability phenomenon over the entire reactor core through dynamic characteristics of neutron, and thereby the oscillation period of every fuel channel is locked to the oscillation period peculiar to the macroscopic instability. Accordingly, in the process of the macroscopic growth of the instability phenomenon, variations in the oscillation period decrease. As a result, the effectiveness of using the oscillation period variation as a parameter for monitoring the nuclear thermal hydraulic instability can be recognized. Note that discontinuous values which appear at the points around 100 sec and 200 sec in graph a of FIG. 8 are attributed to the existence of outliers shown in FIG. 9. When a sufficient number of nuclear instrumentation signals S are included in a group, the standard deviation has robustness against such outliers. However, since an average value and a variance (standard deviation) are not robust statistic values in their nature, it is necessary to remove such outliers at the time of calculating monitoring parameters. Such outliers may be removed by, for example, a method for calculating a standard deviation first and then removing an oscillation period portion which is out of an average oscillation period value by a constant-fold value of the standard deviation. More specifically, a standard deviation is calculated anew by using C as a constant and using only the oscillation period data in the range of the left formula of reference sign 49 in FIG. 6. The resulting standard deviation is use as a monitoring parameter. Graph line b shown in FIG. 8 represents a response in the case where C=1 and discontinuous response portions caused by outliers are removed. The value of C is set after comparison with other stability parameters such as the decay ratio was conducted and considered. The outliers may be generated due to failure of measurement systems or insufficient adjustment of ranges. In that case, the nuclear instrumentation signals S can be subjected to range check in advance and thereby removed from a monitoring group. In that case, the nuclear instrumentation signals S would be in the removed state until the failure or insufficient adjustment are eliminated. Apart from such a case, there is another case in which an outlier is calculated in the process of a statistical processing operation. In this case, the nuclear instrumentation signal S corresponding to the outlier is automatically excluded from the monitor group only at the point when the outlier is calculated. Once the outlier is no longer calculated, the pertinent nuclear instrumentation signal S would be put in the monitoring group again. A sampling period for converting the nuclear instrumentation signal S into digital time series data Xt is 25 msec. The sampling period is defined as a first reference value c obtained when the oscillation periods of the nuclear instrumentation signals S are aligned due to stability deterioration. The first reference value c is shown with a dashed dotted line in FIG. 8. Referring to the response of graph line b which was subjected to correction of the standard deviation outliers in FIG. 8, graph line b once goes below the first reference value c at the point around 225 sec and immediately goes above the 25-msec line, and then again goes below at the point around 470 sec and again immediately goes above the line in the same manner. Graph line b then goes below the first reference value c at the point around 540 sec. After this point, graph line b goes above the line at the point around 625 sec and keeps the state for about 25 sec, but immediately falls at the point past 650 sec and ends up to be a small value of 10 msec or less. It is at the point around 640 sec that the decay ratio actually starts to show rapid increase, and it is also at that point that a notable oscillation component is observed in the nuclear instrumentation signals S. Therefore, it is considered to be reasonable that the first determination unit 44 determines the stability deterioration of the nuclear instrumentation signals S and shifts to the peak detection mode at the point past around 600 sec. However, in order to remove those outliers which accidentally reach the first reference value c, the determination is not made based on only one deviance. Rather, duration time of deviance is monitored by using the point at which the response falls below the first reference value c as a reference. The mode is changed to the peak detection mode at the moment when the duration time exceeds a specified duration length. While the duration length also depends on the oscillation period and on the data length used for calculation thereof, the duration length is set at a value which is several times larger than the oscillation period value and the data length value or more. That is, the peak detection mode is changed when graph line b successively goes below the first reference value c several times or more. For example, if the duration length is set at 5 times, or a time period of 15 sec in FIG. 8, the operation mode is changed to the peak detection mode at the point around 555 sec. On the other hand, it is also possible to use, as the first stability index, a standard deviation of oscillation periods associated with a decay ratio. This parameter is effective in detection when the power oscillation appears not in the entire reactor core but in some regions as in the case of the regional oscillation for example. At the point around 450 sec in FIG. 8, though the response of graph b has not yet reached the first reference value c, the variations in the oscillation period are clearly decreasing as compared with the variations before this point. FIG. 10 shows a distribution in a decay ratio of the respective nuclear instrumentation signals S at this 450 sec point. A plurality of nuclear instrumentation signals S with a decay ratio of 0.8 are present spatially adjacent to each other. They are considered in the stage prior to development into a regional oscillation. For example, a start point of instability is defined as the time when a plurality of nuclear instrumentation signals S with a maximum decay ratio or with a decay ratio of 0.8 or more are present. Accordingly, if 30 msec, which is 20% larger than the sampling period, is set as a reference value, then the point of 450 sec is determined as the start point of stability deterioration, and the operation mode can be changed to the peak monitoring mode based thereon. In the case where stability is recovered after the operation mode has changed to the peak detection mode according to the determination of the stability deterioration by the first determination unit 44, determination to return the mode to the previous oscillation period variation monitoring mode is also needed. FIG. 11 shows a response of a standard deviation of oscillation periods which shifts from the state of stability deterioration with almost no variation to the state of increase in variations with recovery of stability. In short, in this graph line, a section after 230 sec corresponds to the recovered state of stability. Since the sampling period is 80 msec and the standard deviation of oscillation periods before the point of 230 sec is one digit smaller than the standard deviation after that point, it is indicated that the developed power oscillation is present in the section before the point of 230 sec. After execution of the operation for suppressing the power oscillation, the standard deviation of oscillation periods rapidly increases, exceeds the sampling period of 80 msec at the point around 240 sec, and then shifts to values in the range of about 100 to 140 msec. FIG. 12 shows a distribution in a decay ratio of respective nuclear instrumentation signals S at the points around 250 sec and 300 sec of FIG. 11. According to the distribution, a plurality of signals with a decay ratio of 0.8 or more remain, indicating insufficient stability. More specifically, if the sampling period, which is on a level with the criterion for judging the stability deterioration, is set as a criterion for judging the recovery of stability, there is a high probability that the system becomes anti-conservative on the aspect of safety. Therefore, a larger (severe) value is to be used for the criterion for judging the recovery of stability as compared to the criterion for judging the stability deterioration. For example, if the criterion for determining the recovery of stability using the standard deviation is set at a value 1.5 times the sampling period, and a duration time is set at a value 10 times the period, then the safety recovery criterion would not be reached throughout the example of FIG. 11. A description is now given of the grouping unit 41. In the ABWR, there are total 208 nuclear instrumentation signals S (LPRM signals). For monitoring the oscillation period variations, it is desirable to group the nuclear instrumentation detectors 31b (FIG. 1), which are in the level B, the highest level in average value of signals among four levels. This is because phase difference appears in oscillation among detectors different in shaft direction level from each other and also because power oscillation tends to occur when a lower power distribution is high, for example. It is possible to prepare a group which includes all of these nuclear instrumentation signals S (LPRM signals), and it is also possible to use any one of the APRM signals which are grouped into eight groups. However, when the APRM signals are used, oscillations expected to be detected by the group are core-wide oscillations. In this case, it is considered that there is no substantial difference from the general APRM monitoring. In other cases, it may be necessary to employ operation methods such as selecting channels including no outliers and damaged detectors and switching channels in the middle of operation. Since occurrence of regional oscillations is greatly influenced by lower distortion in the power distribution, it is effective, for determining the oscillation mode based on the phase difference, to select detectors out of four detectors 31a, 31b, 31c, and 31d (FIG. 1) and to make a group of the same level, that is, a group of either level A or level B. Reactor core management is conducted by targeting ¼ pattern as shown in FIGS. 13(A), 13(B), 13(C) and 13(D). The higher order space mode distribution generally corresponds to any one of these four patterns. Therefore, eight groups divided by these four patterns are prepared for regional oscillation monitoring. Based on the higher order space mode distribution obtained in the state estimation unit 60, a necessary number of the nuclear instrumentation signals S for use are selected within respective groups in descending order of the higher order space mode distribution. When there is no higher order mode distribution, signals may be selected in descending order of the power distribution, or selected, for example, from predetermined fixed locations, such as two signals from an outermost periphery portion and three signals from an inner side thereof. FIG. 14 shows the case where ten signals allocated to respective groups. In this case, ten detectors (detectors encircled in the drawing) nearest to a node with a higher higher-order space mode value are selected in the respective group region. The pattern of FIG. 14 is equivalent to the pattern of FIG. 13(D). As for the pattern of FIG. 13(C) which is orthogonal to the pattern of FIG. 14, two groups each having ten signals allotted thereto are also prepared. If the patterns of FIGS. 13(C) and 13(D) are predicted in higher order mode distribution prediction, it is not necessary to take the trouble of preparing the patterns of FIGS. 13(A) and 13(B). In this case, four groups would be selected for monitoring regional oscillations. Next, a group for local oscillation monitoring is set. Local oscillations tend to occur in a fuel channel which is thermally severest. Accordingly, the detectors are selected and observed which are nearest to the fuels, which are obtained by the state estimation unit 60 as the fuels having the severest radial power distribution, and to the fuels which are severe in terms of a gross power distribution including effects of an axial power distribution as well as the effects of the radial power distribution. When the detectors selected based on these two power distributions are different, two different groups may be prepared, or either the fuels severe in fuel soundness or the fuels severer in the higher order space mode distribution are appropriately selected and grouped as one group. As shown in FIG. 15, when the position of a detector d is selected as the fuels for local oscillation monitoring, four detectors e which are most adjacent to the detector d are selected, and five detectors composed of these four detectors and the detector d are grouped. In this case, stability indexes (oscillation period, decay ratio, and amplitude) are independently calculated from the detector d, while standard deviations of these stability indexes of five signals are calculated and used as local oscillation monitoring signals. Further, there is also a method in which one detector in the regional oscillation monitor group of FIG. 14, which is nearest to the peak of the higher order mode distribution, is independently set in the group for local oscillation monitoring. The methods described so far are the methods in which the detectors adjacent to the fuel assembly that is a monitoring object are prefixed based on the predicted power distribution and higher order mode distribution, and local oscillations are monitored based on the signals from the detectors. Aside from the methods disclosed, a method for sequentially selecting monitoring signals from the group used for core-wide oscillation monitoring may also be considered. In this method, among respective signal responses coming from the detectors which constitute the group, a signal representing the maximum value of the decay ratios or the amplitude is selected as a local oscillation monitoring signal. FIGS. 16(A) and 16(B) shows transition of the maximum decay ratio of the original nuclear instrumentation signals S used in FIG. 7. In the time period of about 1000 sec, the detector which gains the maximum decay ratio is changed over four areas: a->b->c->a. Of these detectors, the detector having the maximum decay ratio in area a is a detector a located on the lower right side of the reactor core. In area b, the maximum detector shifts to a detector b located at a position symmetrical to the detector a on the reactor core, and then shifts to a detector c in area c, before returning to the detector a in area a. Thus, the signal having the maximum decay ratio is defined as the nuclear instrumentation signal S determined to be most unstable and is used as a signal for local oscillation monitoring. However, since it is inadequate to frequently change the monitoring signal, signal change is conducted when the decay ratio of the monitoring signal shows a rapid fall or when the signal has a larger decay ratio for a time period several times longer than the oscillation period. The monitoring signal is fixed once the monitoring mode is changed to the peak detection mode. As shown in FIG. 3(A), the regional oscillation is an oscillation caused by deterioration of the space higher order mode of neutron fluxes. As a consequence, the distribution of the oscillation mode in the reactor core, i.e., the distribution of the phase difference in oscillation is similar to the higher order space mode distribution. Accordingly, a prediction function for predicting the power distribution and the neutron flux higher-order space mode in the operating state where the reactor core tends to be destabilized is additionally provided for effective monitoring of the regional oscillations. FIG. 2 shows the estimation unit 60 configured to estimate the power distribution and the neutron flux higher-order space mode. The process calculator 61 estimates plant parameters based on a nuclear thermal hydraulic physical model, while loading plant data. With use of the nuclear thermal hydraulic simulator 63 in the process calculator 61, the power distribution or the higher order mode distribution under operating conditions of stability deterioration is calculated. The operating conditions of stability deterioration are low flow-rate and high power conditions after pump trip. Such conditions are obtained in advance by conducting analysis in every operating cycle. In the case where the nuclear thermal hydraulic simulator 63 does not have a function for calculating the higher order mode distribution, distributions calculated off-line in advance corresponding to control rod patterns, burn-up and the like are stored in the data base 62, and the distributions stored in the data base are interpolated/extrapolated to calculate distributions under operating conditions. The thus-obtained power distributions or higher order mode distributions are used by the grouping unit 41 to set signal groups for monitoring the regional oscillation or the local oscillation. A description is now given of determination of the regional stability by grouping. FIG. 17 shows a standard deviation of the oscillation periods calculated in a group, which is made to include top ten signals in the higher order mode distribution as shown in FIG. 14 with use of the pattern of FIG. 13(C). Contrary to this, in FIG. 8 shown before, an average group of the entire reactor core is used. At the point around 450 sec of FIG. 17, the standard deviation of the oscillation periods is lower than 25 msec or a sampling period of the time series data. This value corresponds to a value 1.2 times larger than the criterion for determining stability deterioration in FIG. 8. This group includes the signals having the highest decay ratio in area c in FIG. 16, and corresponds to the determination of stability deterioration on the criterion of a maximum decay ratio of 0.8. Now, a specific example will be shown with respect to the peak detection mode. After stability deterioration is determined by monitoring of the standard deviation of the oscillation periods, it is necessary to promptly change the operating mode to the peak detection mode, which is capable of detecting growth of power oscillations in order to also avoid degradation in fuel and plant soundness due to power oscillations. In the state where stability has been deteriorated to some extent, it is considered that oscillation components which affect nuclear thermal hydraulic stability are more prominent than noise components. Accordingly, by fitting the prominent oscillation component into a polynomial capable of analyzing the components, peaks can be detect in an analytical manner. In this case, the necessary data length is smaller than that in the statistical method, so that higher response speed can also be provided. However, if fitting is performed by using the entire data length, fitting does not succeed in the case of relatively low-order functions. If functions are too complicated, they cause excessive data dependency and disturb robust fitting. Accordingly, piecewise polynomial approximation is employed which divides and fits time series data. A representative example thereof is spline functions, among which the most commonly-used cubic spline interpolation is used to approximate subintervals with cubic polynomials. More specifically, time of the time series data is divided into intervals of tn≦t≦n+1, and these intervals are interpolated with the following cubic:Sn(t)=an+bn(t−tn)+cn(t−tn)2+dn(t−tn)3 Thus, by fitting the time series data into the polynomial, a derivative can easily be obtained in an analytical manner. If the time when the derivative is equal to zero is obtained, the time is equivalent to the time when a peak (top or bottom) of an oscillation appears.dSn(t)/dt=bn+2cn(t−tn)+3dn(t−tn)2 FIG. 18 shows comparison between spline function a and derivative b obtained by differentiating the spline function a. Herein, the sampling interval of the time series data is used for spline interpolation data intervals. More specifically, the sampling intervals are interpolated by tn+1=tn+Δt where Δt represents a sampling period. With respect to the method for searching the time when derivative b is equal to zero, it is often difficult to stably obtain the root of the quadratic since the differential of the cubic spline function is a quadratic function. Accordingly, the following procedures are used. First, an interval where a value obtained by multiplying adjacent derivatives is negative is searched. Since the derivative at endpoints of the interval [tn, tn+1] is equivalent to a coefficient in a primary term of spline function a, they are respectively equal to [bn, bn+1]. In order to remove minute variations in estimated error of the derivatives, it is necessary to filter the time series data with a low pass filter. Once the interval where the derivative is equal to zero is estimated, the interval is then further divided and a derivative is obtained therein. A point where an absolute value of the derivative is the minimum is defined as the zero point. By inputting the time corresponding to this zero point into the spline function, a peak value can be obtained. FIG. 19 is a graph view formed by connecting peaks extracted using this method. The above procedures will be shown in the form of formulas. An interval [tn, tn+1] which satisfies bnbn+1≦0 is obtained based on linear coefficients bn and bn+1 in spline functions of adjacent intervals. In other words, there is a point where the derivative is equal to zero, i.e., a peak of oscillation, in this interval. Therefore, the interval is further divided. That is, the interval [tn, tn+1] is divided at equal intervals into N subintervals. Points between these subintervals are defined as [tn, tn+Δt/N, tn+2Δt/N . . . tn+1]=[t0n, t1n, tmn, . . . tNn], and the divided time is each substituted therein to obtain a derivative. Then, a dividing point where the derivative is most approximate to 0, i.e., where an absolute value thereof is the minimum, is obtained. More specifically, when m is varied from 0 to N, m=mmin which gains a smallest value of \|dSn(tmn)/dt\| is obtained. As a result, tn(mmin)=tn+mminΔt/N makes the position of the peak, and therefore this time is substituted to have a peak value of Sn(tn(mmin)). The peak is either a maximum value (top) or a minimum value (bottom). If the peak is the top, for example, then the next peak searched in the same technique ends up the bottom, and further the next peak will be the top. If the time at which the peak appears is defined as tk(mmin), time difference tk(mmin)−tn(mmin) makes an oscillation period. Since the difference between the top and the bottom makes an amplitude, a ratio between the present amplitude and the amplitude in one period previous makes a decay ratio. Thus, peak detection is repeated until a peak is detected, and once the peak is detected, the detected peak is used as an origin to detect the next peak. The procedures are repeated afterward. It is to be noted that an initial origin is the point of time when the variation monitoring mode is changed to the peak detection mode. An operation flow of the peak detection method will be explained with reference to FIG. 20. When the peak detection method is activated by mode change (S51), the mode change point is set as an origin (S52), and oscillation period T derived by the variation monitoring method before the mode change is used (S53) to set a range T/2 which is half the oscillation period T as the range T/2 has a high probability of including the next peak (S54). The time series data is passed through a low pass filter and then sequentially loaded (S55, S56). As a result, minute fluctuation components with a frequency larger than the frequency of the target nuclear thermal hydraulic stability (inverse of oscillation period) are removed. During a half period (with a certain amount of margin e added thereto) from the origin of peak search, cubic spline interpolation is performed with a sampling period as an interval width (S57). A spline factor bn in a secondary term is obtained, and an interval where a product bnbn+1 between bn and adjacent coefficient bn+1 is equal to zero or less is searched (S58). Once the interval is found, then the interval is further divided into N subintervals. That is, the sampling period Δt is further divided by Δt/N to obtain divided time (S59). Then, a derivative at each of the divided time is obtained, and a dividing point where an absolute value of the derivative is the minimum is searched (S60). The point tnmin makes a peak position, and a peak value Sn(tnmin) derived by substituting the point into the original spline function is stored (S61, S62). Once the peak point is found in this way, an endpoint tn+1 of the interval including the peak point is used as a new origin to find the next peak in a similar way (S61->S52). Based on the stored position of adjacent peaks and peak values, an amplitude, a period, and a decay ratio relating to the second stability index are estimated (S63). Since peaks are searched as tops and bottoms in turns, the amplitude is obtained by subtracting a bottom value from a top value, whereas the period is obtained as a difference between peaks composed of the top and the next top or the bottom and the next bottom. The decay ratio is obtained by dividing the amplitude of a rear peak in a set of adjacent peaks by the amplitude of a front peak. The parameters monitored by the peak detection method are the amplitude or maximum values of peaks, and the decay ratio. That is, in this stage, the stability has already been deteriorated to some extent, and oscillation components other than noise components are dominant in the output response. It is considered therefore that it is highly probable that the peaks are responses based on the power oscillation, while it is unlikely that the peaks are accidental one. There are a plurality of peaks to be monitored, such as average peaks in the group for monitoring average responses, average peaks in the group for monitoring regional oscillations based on the higher order mode, and unaveraged peaks for monitoring local oscillations. Since the behavior of these peaks is different depending on the types of power oscillations, the judgmental criterion for activating oscillation suppression devices is also different. FIG. 21 shows an amplitude response, and FIG. 22 shows a peak (top) response. Both the responses are normalized with signal values at the time when the monitoring mode is changed. Graph a of FIG. 21 and graph a of FIG. 22 represent each a response extracted from signals for local oscillation monitoring, graph b of FIG. 21 and graph b of FIG. 22 represent each a response extracted from ten average signals grouped for local oscillation monitoring in area b of FIG. 14, and graph c of FIG. 21 and graph c of FIG. 22 represent each a response extracted from average signals in the entire reactor core. Both in FIGS. 21 and 22, the peak response of the signal for local oscillation monitoring is fastest, followed by the peak of the signal for regional oscillation monitoring with slight delay, and the peak of the average signal follows with large delay. This is a natural consequence of the targeted power oscillation being the regional oscillation. The judgmental criterion for outputting a trip signal, which activates power oscillation suppression operation, is different depending on such differences as difference in signal characteristics, difference in characteristics of the power oscillation mode, and difference in impact of these differences on the entire plant. In other words, the reference value for outputting the trip signal is set to be small for the average signals which have the lowest sensibility and large impact on the entire plant, and is set to be large for the signals for local oscillation monitoring. For example, if the reference value of amplitude is set as 30% of the amplitude of the local oscillation signal, 20% of the amplitude of the regional oscillation monitoring signal, and 10% of the amplitude of the average signal, then the trip signal is to be outputted at the point around 860 sec when the amplitude of the regional oscillation signal reaches 20% in FIG. 21. If the reference value of the peak value is set as half of the amplitude, then the average signal reaches the reference value fastest and outputs the trip signal at the point around 770 sec in FIG. 22. In this power oscillation example, average power slowly goes up (there is a trend of low-frequency components), so that the trip signal is outputted fastest on the basis of the average value signal in the case of determination according to the peak value. In order to avoid power delay in the trip signal based on spline interpolation processing (FIG. 20: S57), the trip signal is set to be outputted if the value of a signal (average signal of each group and individual signal for local oscillation monitoring) which passed the low pass filter exceeds a predetermined value in spite of the trip determination not made by the peak detection method. For example, the predetermined value is 50% larger than a trip set value based on the peak value, or 15% of the trip set value based on the signal for regional oscillation monitoring. Thus, when trip determination is made without using the peak detection method, the stability of power has already been considerably deteriorated (e.g., a decay ratio of 0.8 or more). Accordingly, a possibility of rapidly grown power oscillation is higher than a possibility of accidental peaks. Therefore, the trip determination in this case is effective as a backup function. FIG. 23 shows a response example of the decay ratio obtained from the peak amplitude of the signal for regional oscillation monitoring. In the state where the amplitude is small, increase and decrease in the amplitude alternately occur between adjacent peaks like a bead, so that the decay ratio has large momentary fluctuation. The signal in this state is not appropriate as a monitoring signal. However, the decay ratio is effective as an index to determine whether or not rapid growth of oscillation is occurring. In the case of the response including alternate increase and decrease in the amplitude, a method using an average value of two decay ratios is effective. In order to avoid abnormal decay ratios in a small amplitude, it is recommended to enable this decay ratio monitoring when a value smaller than the trip set value based on the amplitude, e.g., ¼ of the trip set value, is exceeded. In this case, the trip based on the decay ratio is enabled when 5% of the amplitude is exceeded in FIG. 21, that is, the trip is enabled at the point around 780 sec. This makes it possible to avoid the trip being enabled before this point by an unusually large decay ratio in FIG. 23. If the trip determination value based on the decay ratio in the case of being determined by one amplitude ratio is set at 1.3, and the trip determination value based on the decay ratio in the case of being determined by average of two amplitude ratios is set at 1.2, then the trip is enabled at the point around 800 sec. The determination criterion for enabling the decay ratio-based trip needs to be set at a value which is higher, with a sufficient margin, than a regular noise level amplitude or several %. Even when any one of the trip determination criteria is not reached in the above power oscillation monitoring by the peak detection, it can still be judged that there is a high probability of occurrence of the nuclear thermal hydraulic oscillation if the variations, monitored in the oscillation period variation monitoring executed concurrently with the peak detection method, converge into a small value. In this case, although the peak detection is given high priority, the trip signal is outputted based on the oscillation period variation monitoring. As a result, it becomes possible to suppress power oscillation with higher reliability without compromising fuel soundness and plant soundness. The standard deviation representing oscillation period variations also depends on the number of signals used for calculation. As the number of signals increases, the value of the standard deviation generally decreases. Accordingly, it is necessary to separately set each set point depending on the number of signals in each group. For example, FIG. 8 shows the case of the average value signal group, so that the number of signals used is large. Accordingly, around at the point past 700 sec, the standard deviation largely falls and converges to an almost stable value. Contrary to this case, in the regional oscillation monitoring group, the number of target nuclear instrumentation signals S is ⅕ of that in the average value group. According to the definition of the standard deviation, the value is proportional to the square root of the inverse of the number of signals. Therefore, the value is about √5 to 2.24 times larger. FIG. 24 is an explanatory view showing a method for judging whether to output a trip signal based on a standard deviation of oscillation periods. Line q which defines a reset condition is set at ½ of the criterion for determining mode change, and line p which defines a trip condition is set at ¼ of the criterion for determining the mode change. Thus, as long as the reset condition is set less severely than the trip condition, the values of the lines which define the reset condition and the trip condition are not particularly limited. The time when the signal goes below line p is used as an origin to start time measurement, and when predetermined duration time K has elapsed, the trip signal is outputted. Even when the signal goes below line p once, the time measurement would be reset if the signal exceeds line q before elapse of duration time u (e.g., 10 times of the oscillation period). This is because a smaller deviation value generally has larger variations. In FIG. 24, the signal reaches line p that is the determination criterion of ¼ at the point around 746 sec first, but the signal immediately goes above the line. However, the time measurement of duration time u is maintained, and the criterion is continuously satisfied for almost 30 sec or 10 times of the oscillation period. Therefore, the trip signal would eventually be outputted. In the reactor core shown in FIG. 25(A), averaged signals of area a, averaged signals of upper area b across diagonal line d, and averaged signals of lower area c across diagonal line d are observed. FIG. 25(B) is a graph view showing phase difference in three combinations of these three areas. Between the average signals divided by diagonal line d into monitoring groups, the phase difference in oscillation is calculated with use of a statistical method. In the statistical method, when a cross correlation function or a cross spectrum that is spectral representation of the cross correlation function is obtained, phase difference can be calculated based on delay time with which the function value is the maximum in the case of the former. The phase difference can be calculated can also be calculated based on a spectral phase in the case of the latter. With these parameters, precise information on the power oscillation is obtained, which makes it possible to execute optimal monitoring for the oscillation mode and to enhance the aforementioned determination accuracy. Graph b-c of FIG. 25(B) represents phase difference between two regional oscillation monitoring signals of area b and area c. The phase difference reaches 180 degrees at the point past 600 sec. It is indicated that a combination of these areas constitutes regional oscillation which oscillates in phases opposite to each other. Graph b-a represents phase difference between two regional oscillation monitoring signals of area b and area a. Graph c-a represents phase difference between two regional oscillation monitoring signals of area c and area a. Both graph b-a and graph c-a do not form clear opposite phases. However, since graph b-a has a larger phase difference, it is estimated that an oscillation center line which serves as a node of regional oscillation is out of alignment with the center line d for group division. It is to be noted that all of these phase differences hardly appear in the case of core-wide oscillation. In the case where signs of regional oscillation has already been observed at the point around 600 sec first, and then the operating mode shifts to the peak detection mode as shown in FIG. 25(B), it is possible to preferentially monitor the average signals which represent area b and area c of FIG. 25(A). For example, as for trip conditions of the peak detection method, if it is not desirable, in terms of reliability, to activate the oscillation suppression operation by the trip of only one signal out of a plurality of group average signals, then setting is so changed that the operation is activated by generation of two or more trip signals. In short, the average signals being tripped in both area b and area c are regarded as development of regional oscillation, so that the power oscillation suppression operation may be activated. Alternatively, if the trip conditions of area b and area c are set less severely than the trip conditions of other signals, it becomes possible to ensure that the regional oscillation is suppressed only after these two signals are both tripped. Thus, if the information on the oscillation mode is acquired by the statistical method in the stage before the activation of the peak detection method, malfunction and operation delay can be minimized based on the information, and thereby optimized trip conditions can be implemented with high reliability. Although a description has been given of the example in which the present invention is applied to optimization of control on power oscillation suppression operation after determination of stability deterioration, the present invention is also applicable to determination of stability deterioration in a similar manner. More specifically, if information on the oscillation mode can be acquired at the time of monitoring convergence of oscillation period variations, the average signals for regional oscillation monitoring are preferentially monitored based on the acquired information. If the standard deviation of oscillation periods in the groups of area b and area c decreases in a similar way, it is then judged that the regional stability has been deteriorated. Note that in this case, the judgment of the stability deterioration is concluded only when the criterion for determining the stability deterioration is set less severely and both the standard deviation values satisfy the less-severe determination criterion. The peak detection mode relating to the group signals in the pertinent area is activated. As a consequence, the trip conditions for the peak detection method in area b and area c in this stage are made slightly less severe than those for other signals. FIG. 26 shows an operation flow of the power monitor. Based on the plant information, an expected state of the stability deterioration is predicted (S71). Based on the power distribution and the higher order mode distribution derived from the result, the nuclear instrumentation signals are grouped (S73). As plant information, digitized nuclear instrumentation signals S are inputted in sequence and are subjected to appropriate filtering processing (S72) so as to remove fluctuation components whose frequency bands are different from those of the nuclear thermal hydraulic power oscillations. With use of the signals based on the plant information, grouping is performed (S73), statistical processing is performed on individual signals or averaged signals (S74), and a standard deviation of oscillation periods in each group is outputted (S75). Further, phase difference between group signals is outputted (S81), and the oscillation mode is determined based on the phase difference (S82). The oscillation mode is first referred in the case of determining instability from the standard deviation of oscillation periods (S76). More specifically, based on the oscillation mode characteristics (core-wide, regional, or local oscillation, the center line thereof in the case of regional oscillation, and the position thereof in the case of local oscillation), the group signals optimum for determining instability are selected. In the case where instability has already been determined and the peak detection method has been activated (S78), determination of stability is contrarily performed (S76). If the reactor core is determined to be stable, the procedures from the statistical processing (S74) to the processing for calculating the standard deviation of oscillation periods (S75) are repeated. If instability in the standard deviation of oscillation periods is judged, an alarm is first sounded (S77) and the peak detection method is activated (S78). However, even when the peak detection method is activated, the processing for calculating the standard deviation of oscillation periods is concurrently conducted. In contrast, if stability is determined while the peak detection method is in activation (S76), then processing of the peak detection method is temporally interrupted. After the mode is changed to the peak detection method, peak (top) values, amplitudes, and decay ratios obtained by difference between the amplitudes are each outputted as monitoring parameters (S79). Based on these monitoring parameters, determination of the trip signal is conducted (S80A, S80B, S80C). Further, filtered signal data without being subjected to the peak detection method are also included in the monitoring parameters, and the trip signal determination is performed (S80D). For these monitoring parameters, their trip determination criteria are adjusted with reference to group characteristics and oscillation modes (S82), and their trip determination is performed (S80). The decay ratio is added to the monitoring parameters only when an amplitude value exceeds the criterion. When trip determination is performed in accordance with the trip determination criterion for every parameter based on the peak detection method (S80A, S80B, S80C), trip signals are inputted into a logical gate (S83). It is to be noted that the logical gate is formed of an OR gate in the case of executing oscillation suppression operation according to any one of the trip determinations (S80A, S80B, S80C), or formed of an AND gate in the case where the oscillation suppression operation is not operated if a plurality of trip determinations are not executed. The trip signal for the trip determination performed directly based on the standard deviation of oscillation periods (S80D) is inputted into the OR gate (S84). The results of trip determination (S80A, S803, S80C) based on the peak detection method are also inputted into the OR gate. In this OR logic, when any one of the trip signals is issued, the oscillation suppression operation would be activated automatically (S85). The oscillation suppression operation herein refers to power reduction operation by control rod insertion. (Second Embodiment) As shown in FIG. 27, a power monitor 50 includes: a first calculation unit 52 configured to calculate a decay ratio γ as a first stability index based on time series data t indicating power oscillation in nuclear instrumentation signals S outputted from a plurality of nuclear instrumentation detectors 31 which detect neutrons in a reactor core 16; a first determination unit 53 configured to compare the first stability index (decay ratio γ) and a first reference value D and determine whether nuclear thermal hydraulic stability of the reactor core 16 is stable or deteriorated; a second calculation unit 54 configured to calculate a second stability index R of the reactor core 16 by counting the time series data determined to indicate deterioration in the first determination unit 53; and a second determination unit 55 configured to compare the second stability index R and a second reference value P and determine whether to perform suppressing operation of the power oscillation. A power suppression device 80 includes a plurality of devices different in suppression level, such as a warning unit 81, an insertion preparation unit 82, and a control rod insertion unit 83. Specific physical quantity to be handled is different between in the first embodiment and in the second embodiment and onward. In order to avoid confusion, in the second embodiment and onward, the first stability index is expressed as the decay ratio γ, the second stability index is stated with “R” suffixed thereto, the first reference value is stated with “D” suffixed thereto, and the second reference value is stated with “P” suffixed thereto for differentiation. A grouping unit 51 is configured to classify a plurality of nuclear instrumentation detectors 31 into groups. In the second embodiment, grouping is performed according to the aforementioned levels A to D. Nuclear instrumentation signals S(1) to S(M) outputted from any one of these groups (used herein is a group of the nuclear instrumentation detectors 31 placed at level B) are outputted to the first calculation unit 52, where decay ratios γ(1) to γ(M) are calculated. In the second calculation unit 54, the second stability index R is calculated with this group as a unit, and the second determination unit 55 judges whether or not the activation instruction unit 56 activates the power suppression device 80. The first calculation unit 52 applies digital processing and processing for removing noise and trend components to the received nuclear instrumentation signals S, and extracts time series data made up only of power oscillation components. A statistical method is applied to the time series data to obtain the decay ratio with high precision. Examples of the method for statistically obtaining the decay ratio include a method for directly obtaining an autocorrelation function and setting a delay time with which the correlation function has a maximum value as the oscillation period, a method for obtaining spectral density by a method such as FFT (Fast Fourier Transform) and an autoregressive method and setting an inverse of a frequency (resonance frequency, Hz) at which the spectral density is the maximum as the oscillation period, a method for obtaining a transfer function by the autoregressive method and obtaining the oscillation period from a resonance frequency estimated from a transfer function pole, and a method for obtaining an impulse response by the autoregressive method and obtaining the oscillation period based on a relation shown in FIG. 4. When any one of these method is used, a certain data length (including several oscillation periods or more) is needed in order to achieve accurate estimation. FIG. 28(A) shows overwritten graphs representing decay ratios γ of six signals, selected out of M decay ratios γ sequentially outputted every 5 sec from the first calculation unit 52, in the process of the nuclear thermal hydraulic stability shifting from a stable state to an unstable state. FIG. 29(A) shows a frequency distribution of the decay ratios γ of a plurality of nuclear instrumentation signals (43 signals) at each point of time (400, 500, 600, 700, 800 and 900 sec) during the same period as in FIG. 28. The distribution at 400 sec indicates an average decay ratio γ of 0.45 and a standard deviation of 0.12, the distribution at 500 sec indicates an average decay ratio γ of 0.57 and a standard deviation of 0.11, the distribution at 600 sec indicates an average decay ratio γ of 0.57 and a standard deviation of 0.09, the distribution at 700 sec indicates an average decay ratio γ of 0.76 and a standard deviation of 0.15, and the distribution at 800 sec indicates an average decay ratio γ of 0.89 and a standard deviation is 0.07. In the distribution at 900 sec, the average decay ratio γ is 0.96 and the standard deviation is as small as 0.01, indicating an extremely coherent distribution. It is indicated that at the point of 900 sec, the nuclear thermal hydraulic instability state uniformly spreads over the entire reactor core. Thus, as the stable state changes to the unstable state, the decay ratio γ increases and its variations (standard deviation) tend to become smaller. FIG. 29(B) shows a frequency distribution of oscillation period T under the same conditions as in FIG. 29(A). In the oscillation period T, there is also shown that the variations tend to become smaller in connection with the destabilization as in the case of the decay ratio γ. However, it can be said that the monotonicity of the change in the oscillation period T due to the change from the stable state to the unstable state is inferior to the decay ratio γ. Consequently, it can be said that using the value of the decay ratio γ itself as a monitoring object is suitable for evaluating the nuclear thermal hydraulic stability with high accuracy. FIG. 28(B) shows one signal representing the signals of the decay ratio γ shown in FIG. 28(A). A first reference value D shown herein is a value for identifying whether the reactor core is in a stable state or an unstable state in the corresponding decay ratio γ. As the first reference value D takes a larger value, the criterion for judging the unstable state becomes more lenient, whereas as the value D takes a smaller value, the criterion for judging the unstable state becomes severer. A storage unit 58 (FIG. 27) stores one first reference value D or N first reference values D(n) (n; 1−N) each having a different value. In the first determination unit 53 (FIG. 27), each of a plurality of decay ratios γ(m) (m; 1−M) included in the group are compared in size with the first reference value D (Formula (3)).γ(m)≧D (3) The second calculation unit 54 counts the number of the decay ratios γ(m) which exceed the first reference value D (which satisfies Formula (3)). Operation of the second calculation unit 54 will be explained with reference to FIG. 28(B). In this case, the first reference value D is set at 0.8. As the reactor core is gradually destabilized, the decay ratio γ slowly increases, then rapidly increases at the point around 640 sec, and reaches the first reference value D at around 680 sec. The decay ratio γ goes below the first reference value D only for a short period but immediately exceeds the value D. Accordingly, at the moment when the decay ratio γ exceeds the first reference value D for the first time, a peripheral region of the pertinent nuclear instrumentation detector 31 is judged to be destabilized, and this destabilization is counted up in the second calculation unit 54. More specifically, out of the decay ratios γ(m) (m; 1−M) in the group at a certain point, one that satisfies Formula (3) is added as a(m)=1, and one that does not satisfy Formula (3) is added as a(m)=0. Then, the total of all the added-up a(m) (m; 1−M) is calculated, and the calculated value is divided by sum total M. The value thus normalized is defined as a second stability index R (Formulas (4) to (6)). The second stability index R takes a value from 0 to 1 (Formula (7)).a(m)=1 (γ(m)≧D) (4)a(m)=0 (γ(m)<D) (5)R=SUM{a(1):a(m)}/M (6)0≦R≦1 (7) Then, in the second determination unit 55, the second stability index R is compared in size with a second reference value P stored in the storage unit 58 (Formula (8)). The second reference value P is a value for identifying whether the reactor core is in the stable state or the unstable state in the pertinent group. If Formula (8) is satisfied, it is judged that the nuclear thermal hydraulic stability of the nuclear reactor is deteriorated, and the activation instruction unit 56 instructs activation to the power suppression device 80 of the reactor core.R≧P (8) FIG. 30(A) is a graph view showing the second stability index R calculated over the same period as in FIG. 28 with the first reference value D being set at 0.8. The second reference value P is set at 0.5 which signifies that more than half of the nuclear instrumentation signals S are destabilized. In FIG. 30(A), the second stability index R begins to take a non-zero value at the point around 640 sec, then rises at high speed, and exceeds the second reference value P (=0.5) at the point around 700 sec. As shown in the graph of FIG. 30(A), a rise speed of the second stability index R is relatively high. Accordingly, it can be said that the timing for judging that the group is in the unstable state (Formula (8) is satisfied) is less dependent on the value taken by the second reference value P. For example, time difference between the second reference values P taking a value of 0.4 and taking a value of 0.6 is only 25 sec. FIG. 30(B) is a graph view showing the second stability index R in the case of setting the first reference value D(n) (n; 1−N) in the range of 0.5 to 0.95 in stages. It is observed that the index R tends to have a steeper rising edge and becomes more linear as a preset value of the first reference-value D(n) becomes larger. When the set value of the first reference value D(n) is small, a plateau region is observed as the index R, which once took a non-zero value, stays there for a while without rising up. Accordingly, the storage unit 58 shown in FIG. 27 stores N first reference values N D(n) (n; 1−N) each having a different value. The reference value updating unit 57 updates the first reference value D(n), which is applied to Formulas (4) and (5), once the corresponding second stability index R(n) exceeds the second reference value P (satisfies Formula (8)). Whenever the second stability index R(n) corresponding to the updated first reference value D(n) satisfies Formula (8), the activation instruction unit 56 instructs gradual activation to the power suppression devices 80 (warning unit 81, insertion preparation unit 82, control rod insertion unit 83) which are different in the suppression level. For example in FIG. 30(B), in an initial state where the index R shifts from stability to instability, the first reference value D(1) is applied, so that an alarm is issued at the point around 400 sec when the index R reaches the second reference value P (=0.5). In the final stage, the first reference value D(N) is applied, and so the nuclear reactor is tripped at the point around 820 sec. Thus, since entry into the unstable state is warned about 7 min before reaching the final stage, it becomes possible to secure time for an operator to examine a cause thereof and manually take appropriate measures. It is thus possible to inform signs of destabilization to the operator in advance without activating the oscillation suppression device (control rod insertion unit 83) which may possibly scram the nuclear reactor all at once. Moreover, by evaluating the second stability index R while increasing the first reference value D(n) in stages, it becomes possible to take prompt measures against changes of state and to execute more flexible and more reliable monitoring. A description will be given of a map 70 which is applied to each embodiment with reference to FIG. 31. The map 70 shows placement configuration of a plurality of nuclear instrumentation detectors 31 included in the same group. The map further displays the decay ratio γ of each nuclear instrumentation detector 31 so that the decay ratios γ that exceed the first reference value D (that satisfy Formula (3)) and the decay ratios γ that does not exceed the first reference value D (that do not satisfy Formula (3)) are displayed in a distinguishable manner. In addition to the case where display of the map 70 may be updated in real time when processing of the first determination unit 53 is completed, there is a case where the map 70 is automatically displayed in synchronization with instruction by the activation instruction unit 56 to activate the power suppression device 80. Now, an operation example will be described with reference to the graph view of FIG. 30(B). When the second stability index R with the first reference value being D(1)=0.5, which is a first graph line, exceeds the second reference value P at the point around 400 sec, an alarm is issued. With this alarm sounding, the map 70 is displayed on a monitor screen, and the operator can monitor the monitor screen to check what kind of signs of destabilization phenomena are occurring. In short, a numeric value of the decay ratio γ in those nuclear instrumentation detectors 31 which exceed the first reference value D(1)=0.5 is encircled with a double frame as shown in FIG. 31. A numeric value of the decay ratio γ which indicates the maximum is encircled with a thick wire frame. Accordingly, the operator can estimate what kind of unstable mode is occurring. According to the map 70, the nuclear instrumentation detectors 31 which exceed the first reference value D are distributed across a diagonal line shown with a dashed dotted line, indicating the signs of regional instability with the diagonal line as a node of oscillation. It becomes easy to visually understand such a diagonal line by, for example, encircling the decay ratio indicating the minimum with a dotted frame in each row in an array of the nuclear instrumentation detectors 31. A description will be given of the operation of the nuclear reactor power monitor according to the second embodiment with reference to a flowchart of FIG. 32 (with reference to FIG. 27 where appropriate). First, a first reference value D(n) (n; 1−N) and a second reference value P are acquired (S11). Then, the nuclear instrumentation signals S(m) (m; 1−M) outputted from M nuclear instrumentation detectors 31 within the group are received (S12), and respective decay ratios γ(m) (m; 1−M) are calculated (S13). Each of M decay ratios γ(m) are compared with the first reference value D(n) (n=1), and 1 is counted for every decay ratio γ that exceeds the value D (S14: Yes, S15), and 0 is counted for every decay ratio γ that does not exceed the value D (S14: No, S16). This processing is executed for the decay ratios γ(m) within the group (S17: No). Once the processing is executed for all of M decay ratios (S17: Yes), a second stability index R(n) (n=1) is calculated (S18). Next, the second stability index R(n) (n=1) is compared with the second reference value P. If the index R does not exceed the value P (S19: No), a routine Q from S12 to S18 is repeated with n=1 being fixed. If the second stability index R(n) (n=1) exceeds the second reference value P (S19: Yes), then a first suppression device (warning unit 81) is activated (S20). Further, n=2 is set (S21: No), and the first reference value D(n) is updated (S11). Then, the routine from S12 to S21 is repeated. Once a value n is updated and the final N-th suppression device (control rod insertion unit 83) is activated (S21: Yes), the operation of the power monitor 50 is completed. Thus, with the second stability index R(n) derived with gradual increase in the first reference value D(n), oscillation suppression operation can be advanced gradually as in the stages of light alarm, serious alarm, control rod insertion preparation start, and control rod insertion. Particularly, preliminary alarm issuance can prevent unexpected and automatic activation of control rod insertion as an oscillation suppression device, resulting in reduction in load on the operator. (Third Embodiment) With reference to FIG. 33, a nuclear reactor power monitor according to a third embodiment will be explained. In FIG. 33, component members identical or corresponding to those in FIG. 27 are designated by identical reference signs, and foregoing descriptions are used therefor to omit detailed explanation. A power monitor 50 of the third embodiment is different from the power monitor of the second embodiment in that the activation instruction unit 56 instructs activation of the power suppression device 80 when a second stability index Rk exceeds the second reference value P at least in two or more groups. The power monitor 50 of the third embodiment is further different from the power monitor of the second embodiment in that a reference value correction unit 71 is included. A graph of FIG. 34 represents decay ratios of nuclear instrumentation signals SA to SD outputted from the nuclear instrumentation detectors 31A to 31D (FIG. 1) placed at positions different in a vertical direction. The graph indicates that the decay ratios γ of the nuclear instrumentation signals SA and SB of levels A and B located on a lower side in a vertical direction are more sensitive than the decay ratios γ of the nuclear instrumentation signals SC and SD of levels C and D on an upper side. In grouping the nuclear instrumentation detectors 31, the detectors 31 classified as the same height levels are conveniently grouped together. However, when such grouping is performed, it is necessary to take into consideration the difference in detector sensibility between the levels in advance. The reference value correction unit 71 (FIG. 33) has a function of obtaining, in each of K groups classified, an average value of decay ratios γk(m) in a unit of the k-th group (k; 1−K) and correcting a first reference value Dk based on a deviation uk among the obtained average values. More specifically, an average value <γk> of the decay ratios in the k-th group is expressed by Formula (9), where M denotes the number of decay ratios γk(m) to be outputted:<γk>=SUM{γk(1):γk(M)}/M (9) When a largest value among the average decay ratios <γk> of the respective groups 1 to K is defined as max<γ>, the deviation uk is expressed by Formula (10) and the first reference value Dk applied to each group is expressed by Formula (11):uk=max<γ>−<γk> (10)Dk=D−uk (11) A description will be given of the operation of the nuclear reactor power monitor according to the third embodiment with reference to a flowchart of FIG. 35 (with reference to FIG. 33 where appropriate). A plurality of nuclear instrumentation detectors 31 placed in the reactor core 16 are classified into K groups (S31). One group is herein constituted of M nuclear instrumentation detectors 31. Then, an initial value of the first reference value D and the second reference value P are acquired (S32). An average decay ratio <γk> of the k-th group is obtained in order starting from k=1 (S33, S34: No) till the average decay ratios of all K groups are obtained (S34: Yes). Then, a largest value among the average decay ratios <γk> of the respective groups 1 to K is defined as max<γ>, and a deviation uk between max<γ> and the average decay ratio <γk> of the k-th group is obtained (S35). Further, the initial value of the first reference value D acquired in step S32 is corrected using the deviation uk to acquire a first reference value Dk which is applied to the k-th group (S36). Then, the first reference value Dk which is different per group is applied to the routine Q (FIG. 32) to calculate a second stability index Rk for the k-th group (S37). When the second stability index Rk does not exceed the second reference value P (S38: No), the routine from steps S33 to S37 is repeated. Even when any one of K second stability indexes Rk exceeds the second reference value P, the routine from step S33 to step S38 is repeated unless a prescribed number of the second stability indexes Rk exceed the second reference value P (S39: No). Once the second reference value P is exceeded by a prescribed number of the groups (S39: Yes), the suppression device 80 is activated (S40), and the operation of the power monitor 50 is completed. Also in the third embodiment, as in the case of the second embodiment, a plurality of first reference values D(n) may be registered into the storage unit 58 in advance, and a plurality of power suppression devices 80 different in the suppression level (warning unit 81, insertion preparation unit 82, control rod insertion unit 83) may also be activated in stages. Examples of the grouping method may include various classifying methods other than the classifying method by the height level as described before. In the aforementioned classifying method in which grouping is performed by the height level, it is necessary to take into consideration in advance the difference in detector sensibility between the levels. Accordingly, if the plurality of nuclear instrumentation detectors 31 are classified so that the average decay ratio <γk> in each group is evened, such consideration is no longer necessary. In this case, in the operation flow of FIG. 35, the flow of step S33 to step S36 can be omitted. Examples of such grouping methods include a method for allotting the nuclear instrumentation detectors 31 of four levels A to D to all the groups at the same rate. For example, in this method, the detectors 31 are classified in order toward the right-hand side from an upper left part of the reactor core 16 of FIG. 27, so that the first group includes A/B/C/D/A/B/C . . . , the second group includes B/C/D/A/B/C/D . . . , the third group includes C/D/A/B/C/D/A . . . , and the fourth group includes D/A/B/C/D/A/B . . . . In this case, occurrence of deviation in the average decay ratios between the groups becomes less likely, and also the information on placement configuration of the nuclear instrumentation detectors 31 in the reactor core can also be stored. However, in the case of comparing the decay ratios γ displayed on the map 70 as in FIG. 31, it is necessary to estimate and correct deviation in a vertical direction level. Moreover, in the case of simply handling the entire detectors as one group, it suffices to use a signal obtained by averaging four signals of each instrumentation pipe 34. In the case of preparing only two groups, they may use, for example, an average of (A+C) and an average of (B+C), or an average of (A+D) and an average of (B+D). In the case of three groups, using a method for grouping by three levels except the level D makes it possible to omit correction. (Fourth Embodiment) With reference to FIG. 36, a nuclear reactor power monitor according to a fourth embodiment will be explained. In FIG. 36, component members identical or corresponding to those in FIG. 27 are designated by identical reference signs, and foregoing descriptions are used therefor to omit detailed explanation. A power monitor 50 of the fourth embodiment is different from the power monitors of the second embodiment and the third embodiment in that a weighting factor setting unit 72 configured to set a weighting factor W(n) is included and that the reference value updating unit 57 updates a second reference value P(j) applied to Formula (8) in the next routine once an integrated second stability index R exceeds the second reference value P(j) (satisfies Formula (8)). The second calculation unit 54 calculates a plurality of second stability indexes R(n) corresponding to N first reference values D(n) (n; 1−N) which are different in value, based on Formulas (4) to (6). Then as shown in Formula (12), each of the second stability indexes R(n) (n; 1−N) is multiplied by a corresponding weighting factor W(n) (n; 1−N), and resulting values are added up and outputted as a second stability index R.R=SUM{W(1)×R(1):W(N)×R(N)} (12) In this case, as shown in Formula (13), the weighting factor W(n) is normalized so that a sum of values corresponding to each of the second stability indexes R(n) is equal to 1.SUM{W(1):W(N)}=1 (13) In the graph view of FIG. 37(A), the second stability index R calculated in the same period as in FIG. 28 is shown with a solid line, while terms of R(n) and (n; 1−N) are shown with a broken line. In the foregoing second embodiment, the second stability index R(n) is expressed separately at each of a plurality of set points n. However, in the fourth embodiment, one integrated second stability index R is expressed in this way. It can be said that the integrated second stability index R includes characteristics of individual second stability index R(n) responses, that is, the second stability index R slowly rises in a stable state with a small decay ratio γ, and suddenly rises with progress of an unstable state. As shown in Formula (14), if the weighting factor W(n) is set to be larger with increase in set point n, accuracy of monitoring the progressing instability can be enhanced. As shown in Formula (15), the weighting factor W(n) can be expressed as a geometrical progression using a common ratio r.W(n+1)>W(n) (14)W(n+1)=r×W(n) (15) FIG. 37(B) shows the second stability index R when the common ratio r of the weighting factor W(n), which constitutes a geometrical progression, is gradually varied to 1.0, 1.5, and 2.0. Thus, rising-edge sensibility can be changed by varying the weighting factor W(n). If the common ratio r is set to be larger, rise with a small decay ratio γ is slow and therefore a plateau region will appear. More specifically, if stability does not greatly change, the integrated second stability index R keeps generally a constant value. In contrast, as the decay ratio γ becomes larger, an increasing rate of the second stability index R rapidly becomes larger, resulting in a steep rise. In FIG. 37(B), the point around 650 sec when the index R rises from the plateau coincides with the point of definite occurrence of regional oscillation though small in amplitude. It is possible, therefore, to advantageously distinguish such a physical phenomenon in a visually-aided manner. Thus, as the weighting factor W(n), an optimal value can be set depending on the purpose of detection, such as improvement of sensibility for instability detection, and identification of specific phenomena. A description will be given of the operation of the nuclear reactor power monitor according to the fourth embodiment with reference to a flowchart of FIG. 38 (with reference to FIG. 36 where appropriate). First, a first reference value D(n) (n; 1−N) of a set point n is acquired (S51), and a weighting factor W(n) corresponding to the set point n is set (S52). Further, a second reference value P(j) (j=1) is acquired (S53). A routine Q (FIG. 32) is applied to the first reference value D(n) (n; 1−N) to calculate individual second stability indexes R(n) and to calculate an integrated second stability index R (S54). If the integrated second stability index R does not exceed the second reference value P(j) (j=1) (S55: No), the routine of step S54 is repeated. Once the integrated second stability index R exceeds the second reference value P(j) (j=1) (S55: Yes), the j-th power suppression device 80 (j=1) is activated (S56), and a light alarm is issued. Further, j=2 is set (S57: No), and the second reference value P(j) is updated (S53). Then, the routine from step S53 to S57 is repeated. Once a value j is updated and the final J-th suppression device (control rod insertion unit 83) is activated (S57: Yes), the operation of the power monitor 50 is completed. Thus, also in the fourth embodiment, setting a plurality of second reference values P(j) makes it possible to gradually activate a plurality of power suppression devices 80 (warning unit 81, insertion preparation unit 82, control rod insertion unit 83) different in the suppression level as in the second embodiment. It is also possible in the fourth embodiment, as in the third embodiment, to set a plurality of groups and activate a suppression device once a prescribed number of groups exceed the reference value. As shown in the foregoing, the present invention makes it possible to detect power oscillations caused by nuclear thermal hydraulic instability in the nuclear reactor with high reliability and to suppress power oscillations without the power oscillations exerting serious influence on the soundness of fuels and plants, thereby contributing to safe and efficient operation of the nuclear reactor. The present invention is not limited to the above described embodiments. The invention can be appropriately deformed and implemented within the scope of common technical concepts. For example, the present invention can implement respective devices as respective functional programs by computer. The respective functional programs may be combined to form a program for monitoring nuclear thermal hydraulic stability of the reactor core. 10 Nuclear reactor, 11 Pressure vessel, 13 Steam separator, 14 Upper grid plate, 15 Shroud, 16 Reactor core, 17 Core support plate, 18 Recirculation pump, 21 Main line, 22 Turbine, 23 Generator, 24 Condenser, 25 Pump, 26 Water supply line, 30 Power monitor, 31 (31A, 31B, 31C, 31D) Nuclear instrumentation detector, 32 control rod, 33 Fuel assembly, 34 Instrumentation pipe, 35 Power suppression device, 40 Statistical processing unit, 41 Grouping unit, 42, 52 First calculation unit, 43 Reference value storage unit, 44, 53 First determination unit, 45 Peak detection unit, 46, 54 Second calculation unit, 47, 55 Second determination unit, 48 Third determination unit, 49 Variation analysis unit, 50 Power monitor, 51 Grouping unit, 56 Activation instruction unit, 57 Reference value updating unit, 58 Storage unit, 60 State estimation unit, 61 Process calculator, 62 Data base, 63 Nuclear thermal hydraulic simulator, 70 Map, 71 Reference value correction unit, 72 Weighting factor setting unit, 80 Power suppression device, 81 Warning unit, 82 Insertion preparation unit, 83 Control rod insertion unit, γ (γk, γ(m) m:1−M) Decay ratio, D (D(n) n:1−N) First reference value, P Second reference value, R (Rk, R(n) n:1−N) Second stability index, S (SA, SB, SC, SD, S(1)-S(m)) Nuclear instrumentation signal, uk Deviation, W(n)(n; 1−N) Weighting factor, Xt Time series data.
	claims	1. A method for taking tomograms of a patient's beating heart with the aid of a computed tomography unit, the method comprising:moving at least one focus, with an oppositely situated detector, around the patient to scan the beating heart;recording detector data, output by the detector and representing an attenuation of the beams emanating from the at least one focus, together with indirect or direct spatial orientation data of the beams;recording ECG signals of the beating heart, the detector data and ECG signals being stored in a temporally correlated fashion; andusing, to reconstruct the tomograms, detector data that originates from a selected cycle area of a cardiac cycle of the heart, the cycle area being selected automatically and individually per cycle for at least one cardiac cycle by pattern recognition,wherein the using includesbefore the automatic selection of the cycle area is carried out, a typical signal profile of the current ECG in at least one of the area and an adjacent one is determined manually, and the typical profile is subsequently automatically detected again in at least one cardiac cycle to which the determination of the desired area is oriented for the reconstruction. 2. The method as claimed in claim 1, wherein the profile of a P wave is used as typical signal profile. 3. The method as claimed in claim 2, further comprising:manually marking a time sector in a cycle in a visualized ECG, the profile, located in the sector, of the EGG being adopted as typical signal profile of a P wave (template Sp);subsequently automatically comparing the typical signal profile Sp with the signal profile Ep of further cycles of the EGG via a convolution function Pl(t) that corresponds to a successively temporally offset convolution;determining a greatest maximum per cycle in the temporal profile of the convolution function Pl(t), the actual position of the P wave per cycle, and thus the temporal end of the rest phase of the atrium of the heart per cycle, being determined therefrom. 4. The method as claimed in claim 2, wherein the cycle area whose data are used for the reconstruction precedes the detected area of the P wave. 5. The method as claimed in claim 2, wherein a neural network is used to compare the contour of the profile of the EGG signal with a prescribed contour. 6. The method as claimed in claim 2, wherein, for a cycle in which no P wave position is to be determined, the cycle area considered is determined by at least one known method. 7. The method as claimed in claim 3, wherein the successively temporally offset convolution is performed by calculating P 1 ⁡ ( t ) = S p ⊗ E p ⁡ ( t ) = 1 N · ∫ - ( tt s + tt e ) / 2 ( tt s + tt e ) / 2 ⁢ ⅆ τ · h c · S p ⁡ ( τ ) · E p ⁡ ( t - τ ) ⁢ t ∈ p 1 where: N = ∫ - ( tt s + tt e ) / 2 ( tt s + tt e ) / 2 ⁢ ⅆ τ · S p ⁡ ( τ ) 2 and ttS is the start and tte is the end of a cycle interval in which the pattern to be determined is presumed, τ is the integration variable in the time interval considered, t is the time, and pl is the cycle considered. 8. The method as claimed in claim 7, wherein the convolution function Pl(t) is normalized. 9. The method as claimed in claim 8, wherein the cycle area whose data are used for the reconstruction precedes the detected area of the P wave. 10. The method as claimed in claim 7, wherein the cycle area whose data are used for the reconstruction precedes the detected area of the P wave. 11. The method as claimed in claim 3, wherein the convolution function Pl(t) is normalized. 12. The method as claimed in claim 11, wherein the cycle area whose data are used for the reconstruction precedes the detected area of the P wave. 13. The method as claimed in claim 3, wherein the cycle area whose data are used for the reconstruction precedes the detected area of the P wave. 14. The method as claimed in claim 3, wherein a neural network is used to compare the contour of the profile of the EGG signal with a prescribed contour. 15. The method as claimed in claim 3, wherein, for a cycle in which no P wave position is to be determined, the cycle area considered is determined by at least one known method. 16. The method as claimed in claim 1, wherein a neural network is used to compare the contour of the profile of the ECG signal with a prescribed contour. 17. The method as claimed in claim 1, wherein, for a cycle in which no P wave position is to be determined, the cycle area considered is determined by at least one known method. 18. The method as claimed in claim 1, wherein the detector is a multirow detector. 19. A computer readable medium, including executable instructions, which when executed by a computer, cause the computer to perform the method of claim 1.
	abstract
H00012629	summary	BACKGROUND OF THE INVENTION The present invention is directed to a device for performing a large number of exacting measurements on radioactive fuel rods. There has been a need for performing a large number of exacting measurements on radioactive fuel rods. Due to the radiation, these measurements have to be performed underwater with remote controlled machinery of high reliability. A large number of measurements are required to establish the effects of long term operation of the fuel rods. The quantity of data which needs to be collected makes it difficult to perform the necessary measurements with manually operated instruments. Furthermore, it is necessary that the equipment perform the measurements in the same precise sequence each time while minimizing the chance of data rejection due to human error. The measurements need to be reproducible in subsequent tests or repeated measurement sequences. Calibration of the measuring system must be maintained for the length of the measuring sequence. Any adjustment to the measuring instruments needs to be performed remotely from the control console. SUMMARY OF THE INVENTION The present invention satisfies the above requirements. An objective of the present invention is to provide for a remote, underwater nondestructive examination of radioactive fuel rods without any degradation of the quality of the measurement. Another objective of the present invention is to automatically scan the fuel rod and then to collect and store the data obtained. A further objective of the present invention is to provide a system which is easy and safe to operate.
	abstract	A process for removal of tritium from materials that are contaminated thereby envisages the use of a detritiation reactor RT, in which the reaction for the removal of tritium from the waste takes place, the waste being recovered by a flow of moist inert gas in which an extremely low percentage of humidity is used. The heated waste releases a current of tritiated gases, the current of gases being removed from the reactor via the moist inert gas, which conveys it into a membrane reactor RM for decontamination. The membrane reactor, in fact, is able to remove selectively the tritium present in the mixture of gases: there is thus the dual advantage of purifying the mixture of gases and of recovering the tritium contained therein.
	summary
	summary
	description	This application is a continuation claiming the benefit under 35 U.S.C. §120 of U.S. application Ser. No. 12/078,409, filed Mar. 31, 2008, now U.S. Pat. No. 7,708,961 which claims the benefit under 35 U.S.C. §119(e) of U.S. Provisional Application No. 60/909,431, filed on Mar. 31, 2007, the entire contents of each of which are hereby incorporated herein by reference. 1. Technical Field Example embodiments relate to the production and extraction of radioisotopes from a source compound. 2. Description of the Related Art Therapeutic radiopharmaceuticals may be radiolabeled molecules used for delivering therapeutic doses of ionizing radiation with relatively high specificity to certain disease sites (e.g., cancerous tumors) in a patient's body. Additionally, recent research has been directed to the radiolabeling of monoclonal antibodies to evaluate the efficacy of radioimmunotherapy. A number of different radioisotopes have been used for these purposes, including α, β, and auger electron emitters. For those applications including site-specific therapy, it may be beneficial to use radiopharmaceuticals exhibiting higher specific activities. However, the presence of “cold” labeled antibodies may decrease the number of “hot” labeled antibodies that occupy the binding sites on the target cells. Consequently, reduced numbers of “hot” labeled antibodies may result in lower doses of ionizing radiation to the target cells, thus decreasing or impeding the ability of the treatment to induce the desired cell kill. Accordingly, higher specific radioactivity (SA) compounds may be beneficial to reduce the impact of “cold” labeled antibodies. 186Re has been investigated as a candidate for radiotherapy, because 186Re decays by β− emissions and has a half-life of about 3.7 days. Additionally, 186Re exhibits a chemical similarity to 99mTc, a radioisotope that has already been extensively studied and used in a variety of medical applications. 186Re may be produced in reactors via an 186Re(n, γ)186Re reaction. Although radioimmunotherapy using 186Re has been successfully performed, higher SA 186Re compounds remain relatively difficult to obtain. A method of isolating a radioisotope according to example embodiments may include ionizing a source compound containing a first isotope and a second isotope of an element so as to form charged particles of the first and second isotopes. The second isotope may have at least one of therapeutic and diagnostic properties when used as a radiopharmaceutical. The charged particles may be separated so as to isolate the particles of the second isotope and obtain a material having a specific activity above 30 curies/mg. An isolated radioisotope material according to example embodiments may be prepared by a method including ionizing a source compound containing a first isotope and a second isotope of an element so as to form charged particles of the first and second isotopes, the second isotope having at least one of therapeutic and diagnostic properties when used as a radiopharmaceutical; and separating the charged particles so as to isolate the particles of the second isotope, the method resulting in the isolated radioisotope material having a specific activity above 30 curies/mg. It will be understood that when an element or layer is referred to as being “on”, “connected to”, “coupled to”, or “covering” another element or layer, it may be directly on, connected to, coupled to, or covering the other element or layer or intervening elements or layers may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to” or “directly coupled to” another element or layer, there are no intervening elements or layers present. Like numbers refer to like elements throughout the specification. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that, although the terms first, second, third, etc. may be used herein to describe various elements, components, regions, layers and/or sections, these elements, components, regions, layers and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer or section from another region, layer or section. Thus, a first element, component, region, layer or section discussed below could be termed a second element, component, region, layer or section without departing from the teachings of example embodiments. Spatially relative terms, e.g., “beneath,” “below,” “lower,” “above,” “upper” and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It will be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the term “below” may encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. The terminology used herein is for the purpose of describing various embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a,” “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises” and/or “comprising,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. Example embodiments are described herein with reference to cross-sectional illustrations that are schematic illustrations of idealized embodiments (and intermediate structures) of example embodiments. As such, variations from the shapes of the illustrations as a result, for example, of manufacturing techniques and/or tolerances, are to be expected. Thus, example embodiments should not be construed as limited to the shapes of regions illustrated herein but are to include deviations in shapes that result, for example, from manufacturing. For example, an implanted region illustrated as a rectangle will, typically, have rounded or curved features and/or a gradient of implant concentration at its edges rather than a binary change from implanted to non-implanted region. Likewise, a buried region formed by implantation may result in some implantation in the region between the buried region and the surface through which the implantation takes place. Thus, the regions illustrated in the figures are schematic in nature and their shapes are not intended to illustrate the actual shape of a region of a device and are not intended to limit the scope of example embodiments. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which example embodiments belong. It will be further understood that terms, including those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. Example embodiments relate to the production and isolation of anionic species from a source material. For instance, the methods according to example embodiments may be suitable for producing and isolating 186Re (rhenium-186) radioisotopes. As a result, higher specific radioactivity compounds containing the 186Re radioisotopes may be generated. The 186Re compounds may be utilized in a variety of medical applications. For example, an 186Re compound may be attached to one or more antibodies that are specific to the targeted receptors and utilized in radiation therapy and/or diagnostic procedures. The methods and apparatuses according to example embodiments may also be suitable for producing other higher specific radioactivity materials which may be utilized in a broader range of research, therapeutic, and/or diagnostic applications. Conventional methods of producing 186Re may utilize 185Re (rhenium-185), 187Re (rhenium-187), or 186W (tungsten-186) as the starting material. The conventional method utilizing 185Re as the starting material may be represented by expression (1) below:185Re(n,γ)186Re (1)wherein the 185Re is converted to 186Re through neutron capture in a reactor. Although this method may have relatively high yield, separating the desired 186Re isotope from the source material may be difficult (e.g., via chemical separation), thus resulting in products exhibiting relatively low specific radioactivity. Conversely, it should be understood that a (γ,n) reaction for producing 186Re would utilize 187Re as the starting material. The conventional method utilizing 186W as the starting material may be represented by expression (2) below:186W(p,n)186Re (2)wherein the 186W is converted to 186Re through a proton induced reaction within a particle accelerator. Although this method may have in a relatively low yield, separating the desired 186Re isotope from the source material may be easier (e.g., via chemical separation), thus resulting in products exhibiting improved levels of specific radioactivity. However, because the cross-section for the 186W(p,n)186Re reaction is relatively low, producing patient-dose quantities of 186Re in a cost effective manner using this method may not be feasible. Additionally, a relatively large number of curies of therapeutic and/or diagnostic radioisotopes may be required for clinical trials. Accordingly, an accelerator-based 186W production method may not even be able to produce the necessary quantities of therapeutic and/or diagnostic radioisotopes for a single patient per day (let alone thousands of patients yearly). The methods and apparatuses according to example embodiments may involve the ionization and mass separation of 186Re from the 186Re starting material so as to facilitate the production of increased specific radioactivity 186Re compounds. The methods and apparatuses according to example embodiments may be able to achieve radioisotope production in the range of curies per day of material exhibiting relatively high specific radioactivity values (e.g. above 30 curies/mg). As discussed above, conventional 186Re therapeutic and/or diagnostic compounds produced by neutron capture in a reactor may have relatively low specific radioactivity. Consequently, increases in the specific radioactivity of 186Re compounds according to example embodiments may be investigated to determine to the level of specific radioactivity required to improve therapeutic and/or diagnostic efficacy relative to that of conventional 186Re compounds. Once a target specific radioactivity has been established (e.g., the antibody-conjugated 186Re according to example embodiments exhibits improved efficacy over the conventional lower specific radioactivity 186Re while maintaining acceptable specificity so as to reduce or avoid impacting cells that do not express the target cell surface marker), methods and apparatuses according to example embodiments may be employed to produce usable quantities of the 186Re compound having the target specific radioactivity via ionization and mass separation of the 186Re radioisotope. The increased availability of 186Re compounds having higher specific radioactivity may facilitate further chemical developments and clinical studies directed to the use of 186Re-radiolabeled antibodies or small molecules. Labeling an antibody with 186Re produced and recovered according to example embodiments may involve utilizing an activated ester as a bifunctional chelating agent (e.g., mercaptoacetyltriglycine (MAG3)). An example of a reaction scheme for the synthesis of the activated ester may be shown below by scheme (3). Although 188Re may be available in no-carrier-added form via a 188W generator, 186Re may be the more suitable radioisotope, at least with regard to matching the physical decay properties of the radioisotope with the cell repair cycle. For example, the decay properties of 186Re may include a β−Emax of about 1 MeV and a t1/2 of about 90 h, while the decay properties of 188Re may include a β−Emax of about 2 MeV and a t1/2 of about 17 h. Thus, the decay properties of 186Re may be more suited for the radioisotope therapy of small tumors. Additionally, generation of the 188W precursor (for 188Re production) involves a double neutron capture reaction which can be achieved at only a few reactors worldwide, while facilities capable of the 185Re(n,γ)186Re reaction are much more widely available. According to example embodiments, higher specific radioactivity 186Re compounds may be generated with greater ease from the 185Re(n,γ)186Re reaction product. Furthermore, 186Re compounds according to example embodiments may exhibit improved physical properties with regard to energy and half-life relative to 90Y and 131I, respectively, wherein 90Y and 131I are commonly used radioisotopes. The methods and apparatuses according to example embodiments relate to the production of increased specific radioactivity 186Re compounds. Additionally, the specific radioactivity of the 186Re compounds may be adjusted via the inclusion of natural rhenium so as to achieve a level of specific radioactivity that exhibits the desired balance of therapeutic and/or diagnostic efficacy and value. As discussed above, 186Re may be a suitable candidate for radiotherapy, because its decay properties include β− emissions and a half-life of about 3.7 days. Furthermore, 186Re has a chemical similarity to 99mTc, which has already been extensively studied. However, although production facilities capable of producing 186Re via the 185Re(n,γ)186Re reaction may be readily available, the conventional 185Re(n,γ)186Re reaction method typically results in a 186Re product exhibiting relatively low specific radioactivity which limits its utility in therapeutic and/or diagnostic applications involving site-specific targets. To improve the production of higher specific radioactivity 186Re compounds, methods and apparatuses according to example embodiments may employ a cusp ion source to ionize and extract the 186Re radioisotopes from the starting material. Additional information regarding cusp ion source technology may be found, for example, in Dehnel, et al., NIM B, vol. 241, pp. 896-900, 2005, the entire contents of which are incorporated herein by reference. FIG. 1 is an electrical schematic diagram of a rhenium ion source according to example embodiments. Referring to FIG. 1, plasma may be generated by electron emission from the filament 100 at a current of about 130 Amps. The plasma may be maintained in a stable state by the addition of hydrogen (H2) gas. As a result, the majority of the ions implanted into the Faraday cup 102 may be H− ions. The extraction lens 104 (e.g., 2 kV) and Faraday cup 102 (e.g., 20 kV Bias) may be maintained at a positive voltage so as to extract negative ions from the source. It should be understood that the rhenium ion source according to example embodiments is not limited to the parameters set forth in FIG. 1. Rather, one of ordinary skill in the art will readily appreciate that, in view of the present disclosure, other variations are possible. Using the example discussed above, initial tests may be conducted to determine the temperature of the plasma as a function of the resistance of resistor R2. As the filament current is increased, the arc across the plasma may also increase. The resistor R2 may limit the feedback between these two power supplies, so it may be beneficial to determine the highest resistance of the resistor R2 that will allow the maintenance a temperature that is sufficiently high to keep a rhenium oxide species volatile. A graph of this plasma temperature change with resistance is shown in FIG. 2. In light of the results shown in FIG. 2, the R2 resistor value may be maintained at about three ohms to ensure adequate vaporization. However, in view of the present disclosure, those of ordinary skill in the art will readily appreciate that a variety of circuits and apparatuses may be used to achieve the target plasma heating and that such modifications would not detract from the fundamental operation of the disclosed device. In a method according to example embodiments, H188ReO4 was utilized as the radioisotope source compound. The H188ReO4 was collected on a quartz dish, dried, and placed in the ion source chamber. The pressure in the ion source chamber was reduced to below atmospheric pressure, and hydrogen plasma was produced within the ion source chamber. Consequently, the plasma heated the radioisotope source compound to a temperature sufficient to induce vaporization of the source compound. As the molecules of the source compound vaporized and interacted with the plasma (e.g., H− ions), negatively charged species were produced and accelerated toward the collector assembly. In this instance, the collector assembly was a Faraday cup, although example embodiments are not limited thereto. Without being bound by theory, it is believed that the H− plasma interacts with the radioisotope source compound to produce one or more negatively charged ions (e.g., ReOn−) which are accelerated toward and collected in the Faraday cup. As will be appreciated by those ordinarily skilled in the art, this technique may also be applicable to other radioisotope source compounds (e.g., oxides, nitrides, carbides) which can be vaporized under the appropriate temperature and pressure combination maintained within the ion source chamber. Similarly, those ordinarily skilled in the art will also appreciate that the proper temperature and pressure may be a function of the materials utilized, the power applied, and the configuration of the source chamber and the ancillary equipment (e.g., gas mass flow controllers, valving, control systems, vacuum pumps, cooling assemblies). The ion source chamber according to example embodiments may be constructed and operated so as to enable the creation and maintenance of the appropriate temperature and pressure conditions within the ion source chamber. As a result, the radioisotope source material may be vaporized at a suitable rate without damaging the ion source chamber or generating undesirable levels of byproducts that would interfere with the collection and enrichment of the targeted radioisotope. For example, the radioisotope source compound utilized in the ion source may exhibit satisfactory vaporization at temperatures below about 1300° C. Additionally, it may be beneficial for the radioisotope source compound to exhibit satisfactory vaporization at temperatures below about 900° C. so as to allow for the utilization of a wider range of materials in the construction of the ion source chamber. Furthermore, it may be beneficial for the radioisotope source compound utilized in the ion source to exhibit satisfactory vaporization at pressures below about 1 Torr. As discussed above, the use of an appropriately sized R2 resistor according to example embodiments may allow the production of plasma capable of heating the source compound and its vessel to temperatures in excess of about 500° C., thereby volatilizing the rhenium oxide. Consequently, the source compound may dissociate within the plasma, with the resulting fragments becoming negatively charged ions (e.g., ReOn−). The negatively charged ions may be extracted from the ion source chamber and implanted on the Faraday cup. After an implant cycle, the Faraday cup may be removed and evaluated using gamma spectroscopy to determine the amount of radioactivity implanted in the Faraday cup. Ion source performance analysis indicates that the apparatus illustrated in FIG. 1 may achieve implant beam currents of about 1.2 mA (with H− constituting a major portion of the beam and the radioisotope source compound species ReOn− constituting a minor portion of the beam). FIG. 3 is a photographic image of a Faraday cup after one hour of irradiation with the extracted Re beam according to example embodiments. Because the power of the accelerated beam exceeded the tolerance of the Faraday cup based on its initial configuration, the Faraday cup became discolored and deformed, as shown in FIG. 3. When the Faraday cup and the source compound container from the ion source chamber were analyzed with a high purity Germanium detector for radioactivity, the initial results indicated that approximately 20% of the radioactivity that was volatilized from the source was actually implanted in the Faraday cup. Additional efforts may be directed toward improving the extraction percentage, wherein the extraction percentage may be the portion of the desired rhenium radioisotopes released from the source compound vessel (e.g., quartz dish). For example, by providing a combination of both stable and radioactive rhenium atoms on the source compound vessel used in the ion source chamber, the majority of the radioisotope atoms may be successfully vaporized, ionized, and collected at the target assembly (e.g., a Faraday cup). As will be appreciated by those ordinarily skilled in the art, various combinations of stable and radioactive rhenium atoms and extraction voltages may provide for further improvements in the extraction percentage. FIG. 4 is a plan view, side view, and perspective view of a water-cooled Faraday cup for a rhenium ion source according to example embodiments. A modified apparatus incorporating a water-cooling arrangement 106 for the Faraday cup 102 may reduce the damage suffered by the Faraday cup 102 during implantation. For example, the water-cooled Faraday cup 102 may be beneficial during prolonged implants and may increase the removability of the radioactivity from the source. The methods and apparatuses according to example embodiments may facilitate the production of useful quantities of increased specific radioactivity 186Re and related compounds. For example, a 186Re source compound may be placed in an ion source chamber and exposed to a temperature and pressure combination that is sufficient to induce the vaporization of the source compound. Hydrogen plasma may be utilized to both heat the source compound and to ionize the resulting molecular fragments to produce Re-containing anions. However, it should be understood that a plasma of hydrogen, helium, nitrogen, argon, oxygen, xenon, and/or krypton may also be utilized. The Re-containing anions may be extracted from the ion source chamber and collected in a positively-charged target vessel. As will be appreciated by those ordinarily skilled in the art, alternative configurations may provide for supplemental heating sources. For example, resistance heating and/or microwave heating may be used in lieu of or in addition to the plasma for vaporizing the source compound. Similarly, alternative structures (e.g., higher voltage filaments) may be utilized for imparting a charge to the vaporized source compound fragments so that the desired species (e.g., radioactive species) may be extracted from the ion source chamber and accelerated toward a collection assembly. Furthermore, the source compound may be introduced into the ion source chamber as a vapor (e.g., perrhenic acid). It should also be understood that the source compound may be ionized with laser radiation, electron bombardment, thermally heated surfaces, or ion sputtering. The charged particles resulting from ionization of the source compound may be negatively- or positively-ionized atoms and/or molecules. Thus, when properly configured according to the present disclosure, various alternative example embodiments may be attained for purposes of producing higher specific radioactivity compounds. Depending on the separation assembly (e.g., magnetic separation assembly), specific radioactivity values in the range of 30 curies/mg to over 300 curies/mg may be achieved using the methods and apparatuses according to example embodiments. As discussed above, a CUSP ion source may be used to separate 186Re from neutron-irradiated 185Re by ionizing perrhenate molecules and implanting them on a water-cooled Faraday cup. The CUSP ion source may provide satisfactory results even when the perrhenate ion beam is not controlled and is contaminated with a relatively high current negative ion hydrogen beam. Alternatively, a negative ion surface thermal ionization (NIST) process may be utilized to ionize the perrhenate molecules. Depending on the circumstances, negative ion surface thermal ionization may be more efficient and effective than CUSP ionization. Methods and apparatuses according to example embodiments with regard to negative surface ionization are described below. Furthermore, additional information relating to surface ionization may be found in Brown, Ian G. (Ed.), “The Physics and Technology of Ion Sources,” 2nd edition, Wiley-VCH, Weinheim, 2004, the entire contents of which are incorporated herein by reference. When a neutral atom or molecule impinges upon and is temporarily adsorbed by a heated surface during a negative ion surface thermal ionization (NIST) process, the heated surface may be hot enough to prevent the atoms from remaining adsorbed. As a result, the atoms or molecules may be ionized when leaving the heated surface. A negative ion may be produced when the work function (Φ) of the heated surface is smaller than the electron affinity (EA) of the atom or molecule impacting the heated surface. For example, referring to FIG. 5, when approaching a relatively hot surface 500, an atom/molecule 502 may become polarized by the forces between its nucleus and the free electrons inside the relatively hot surface 500. The atom/molecule 502 may adhere to the relatively hot surface 500 under the action of these forces. If the work function (Φ) of the relatively hot surface 500 is smaller than the electron affinity (EA) of the absorbed atom/molecule 502, then an electron 504 at the Fermi level in the conduction band of the relatively hot surface 500 may shift by tunneling to the electron affinity level of the atom/molecule 502. Consequently, there may be a probability that the adsorbed atom/molecule 502 will transition from a neutral state to a negative ionic state. If the temperature of the relatively hot surface 500 is sufficiently high, then the adsorbed atom/molecule 502 may accumulate enough energy to overcome the binding forces so as to result in thermal desorption. During thermal desorption, the adsorbed atom/molecule 502 may be ejected as an ion 506 with relatively low energy from the relatively hot surface 500. The likelihood of ionization may be described as a function of the surface temperature, the work function of the surface material, and the electron affinity of the atom/molecule to be ionized. The probability that a negative ion will be emitted may be mathematically expressed by a set of equations. For example, the equilibrium ratio (α) of ion flux (N−) to neutral flux (Nn) leaving from the heated surface may be provided by the Saha-Langmuir (S-L) equation as shown by equation (4) below: α = N - N n = g - g n ⁢ exp ⁡ [ q ⁡ ( EA - Φ ) kT ] ( 4 ) wherein: N−=emission rate of negative ions Nn=emission rate of neutral species φ=work function of the surface [eV] EA=electron affinity of atom or molecule [eV] k=Boltzmann's constant (8.617×10−5 eV/K) T=absolute surface temperature [K] g−, gn=statistical weighting factors for the negative ion and neutral atom/molecule, respectively. They are related to the total spin S of the respective species given by g = 2 ⁢ ⁢ S + 1 = 2 ⁢ ∑ i ⁢ s i + 1 ,wherein si is the spin on the ith electron The ionization efficiency (β) may be in equilibrium when the total number of particles (N0) is equal the sum of N−+Nn. The ionization efficiency (β) may be expressed by equation (5) below: β = α 1 + α = N - N 0 = 1 1 + g n g - ⁢ exp ⁡ ( q ⁡ ( Φ - EA ) kT ) ( 5 ) wherein: N−=emission rate of negative ions Nn=emission rate of neutral species φ=work function of the surface [eV] EA=electron affinity of atom or molecule [eV] k=Boltzmann's constant (8.617×10−5 eV/K) T=absolute surface temperature [K] g−, gn=statistical weighting factors for the negative ion and neutral atom/molecule, respectively. They are related to the total spin S of the respective species given by g = 2 ⁢ ⁢ S + 1 = 2 ⁢ ∑ i ⁢ s i + 1 ,wherein si is the spin on the ith electron In view of the above equations, it may be appreciated that higher temperatures may have higher ionization potential. Additionally, it may be appreciated from equation (6) below that the residence time (τ) of the impinging particle may be reduced with higher temperature. τ = τ 0 ⁢ exp ⁡ ( E ads kT ) ( 6 ) wherein: Eads=ion adsorption energy [eV] τ0=vibrational period of the ion near the surface [s] k=Boltzmann's constant (8.617×10−5 eV/K) T=absolute surface temperature [K] The ion adsorption energy (Eads) may a few eV, and τ0 may be about 10−13 s. The ionization probability may be independent of the initial kinetic energy as long as the initial kinetic energy is smaller than or comparable to the adsorption energy, because the residence time (τ) on the heated surface may be sufficient to ensure thermal equilibrium with the heated surface. A negative surface ion source apparatus according to example embodiments may include an evaporation unit, a vacuum system, an ionization unit, and an extraction unit. The extraction unit may include magnets for removing excess electrons. Ionization and extraction according to example embodiments may include transferring a 185/186Re mixture into a crucible and inserting the crucible into the evaporation unit. A vacuum may be established in the evaporation unit. The perrhenate molecules of the 185/186Re mixture may be evaporated under a vacuum. The perrhenate molecules then may be ionized in the ionization unit. The resulting perrhenate ions may be extracted from the ionization unit as a beam, wherein the beam may be shaped for injection into a mass separator to separate the 185Re from the 186Re. A method of isolating 186Re according to example embodiments will be discussed in further detail below. An irradiated chemically-undefined 185/186Re mixture may be chemically converted into a perrhenate salt (different counter ions are suitable). The perrhenate salt may be dissolved in water and transferred to a vaporization crucible. The water may be completely evaporated from the crucible, such that the 185/186Re perrhenates may be adhered to the walls of the crucible. The crucible may be made of a refractory material with a relatively low work function. For example, the crucible may be formed of tungsten (W), molybdenum (Mo), tantalum (Ta), or Lanthanum-Hexaboride, although example embodiments are not limited thereto. The cavity of the crucible may be comprised of a hollow cylinder with one side closed and the opening directly attached to the vaporization unit. The inner diameter and depth of the cavity may be in the mm to cm range and may be adjusted as needed. The crucible may be disposed in a filament of the evaporation unit for ohmic heating. After the crucible with the perrhenate has been inserted into the filament of the evaporation unit, a vacuum may be established (e.g., about 10−5 to 10−7 Torr). The crucible may be heated to a temperature of about 1500° C. After evaporation, the volatile perrhenates may drift into the ionization unit. The temperature of the ionization unit may be controlled separately. The ionizer may be made of a refractory material with a relatively low work function. The ionizer may have a tubular shape. The ionizer may also be filled with a porous material or a screen so as to enhance the ionizing process by increasing the surface area. The ionizer may be ohmically heated by a filament up to temperatures of about 1500° C. It may be beneficial for the transition connection between the evaporator and the ionizer to be relatively tight so as to reduce or prevent the loss of the volatile perrhenates. The transition connection may also provide thermal insulation between the evaporator and the ionizer to allow independent control of the evaporation and ionizing processes. Upon operation of the ion source, a plasma including of an equilibrium of volatile ionized and neutral perrhenates may be generated in the ionizer volume. An excess of free electrons, formed during the ionization process, may also be present. To reduce or prevent further acceleration of the excess free electrons, a relatively weak magnetic field may be established at the “exit” of the ionizer to draw the excess free electrons towards the screening electrode. The negatively ionized species may be accelerated from the ionizer by an electric field produced by a series of extraction electrodes having different voltage levels. The perrhenate ions and the excess free electrons may be initially accelerated from the ionizer region by the extraction electrode. The perrhenate ions may then be further accelerated and shaped by the screening electrode, whereas the excess free electrons (which have smaller mass) will hit the screening electrode and so be removed from the perrhenate ion beam. The final extracted perrhenate ion beam may be additionally shaped by magnetic and/or electrostatic beam optics and then injected into a mass separator to separate the 185Re from the 186Re. Although the example embodiments detailed above are directed to the production of higher SA 186Re compounds, the present disclosure is not limited thereto. For instance, the methods and apparatuses described above may be applied to the extraction of other radioisotope species that can be vaporized and charged within an ion source chamber constructed and operated in accord with the detailed description provided above (e.g., extraction of 99Mo from 98Mo and/or 100Mo). Accordingly, the methods and apparatuses according to example embodiments may be utilized to produce an increased volume of a range of higher SA radioisotope materials having a longer shelf life and improved therapeutic and/or diagnostic effects compared to conventional production and purification techniques. While example embodiments have been disclosed herein, it should be understood that other variations may be possible. Such variations are not to be regarded as a departure from the spirit and scope of example embodiments of the present disclosure, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
047567686	abstract	Chemical decontamination of metallic parts of nuclear reactor installation in which an oxidative treatment with a permanganic acid solution is applied before dicarbonic acids are used for further treatment. Rinsing operations are eliminated and smaller amounts of dicarbonic acids needed. Also the primary system of the nuclear reactor no longer requires emptying before effecting decontamination treatment.
	abstract	Example embodiments and methods are directed to irradiation target retention devices that may be inserted into conventional nuclear fuel rods and assemblies. Example embodiment devices may hold several irradiation targets for irradiation during operation of a nuclear core containing the assemblies and fuel rods having example embodiment irradiation target retention devices. Irradiation targets may substantially convert to useful radioisotopes upon exposure to neutron flux in the operating nuclear core and be removed and harvested from fuel rods after operation.
047864621	description	DETAILED DESCRIPTION OF THE INVENTION The monolith nuclear reactor support structure may be used with any type of nuclear reactor including light water reactor vessels and liquid metal reactor concept vessels. Liquid metal reactors of the pool or loop type may be supported by the monolith support structure of the invention. The preferred type of liquid metal reactor for use in the monolithic support structure of the invention is described in an application filed as of even date herewith by George Garabedian. The monolith support structure is made of reinforced concrete and comprises a base member and a unitary core member. The base member is intended to serve as a foundation to support the unitary core and other structures which need to be designed for seismic disturbances. The foundation may be placed at, above, or below grade depending on soil conditions and other considerations such as hydrology. For the preferred embodiment, the top of the basemat is at grade and the top of the monolith unitary case above grade is 64 feet. The unitary core is integrally formed with the base member using standard practices in the reinforced concrete arts. Generally light rebar density requiring simple lap splicing may be used for the unitary core. Other reinforcing techniques including pre-stressing and post-stressing using conventional materials can be employed. The unitary core member has a central vertical void for the reactor vessel and satellite vertical voids for housing satellite tanks that are radially arranged around the central vertical void. The unitary core also provides horizontal voids to permit cross-connection between the reactor vessel and the satellite tanks. It is also contemplated that inspection access passageways will be present in the unitary core structure as well as conduit passageway for ventilation and control and monitoring cables and devices. The various voids are made large enough to accomodate the reactor components as well as an exterior containment structure including appropriate layers of insulation. The ratio of the entire gross area of the surface of the monolith to the void area is greater than 2.0. In addition, the ligaments between voids or between a void and the edge of the monolith are sufficiently large to assure that deformation of the monolith is shear controlled. The present invention is best seen in FIG. 3 in an embodiment which shows central vertical void 2A for a liquid metal reactor vessel. The satellite vessel voids 4A are for satellite tanks that are connected to the reactor vessel by conduits that are placed in horizontal voids 6A. The reinforced concrete base 5A is a common basemate for supporting the unitary reinforced concrete core 7A as well. A steel deck platform 9A is provided above the top of the unitary concrete core 8A and serves as an operating platform as well as a cover for cabling and cable trays which are routed above the top surface of unitary core. A vertical access tunnel (not shown) is provided between the satellite vessels in order to service and maintain the expansion joints on the containment conduits connecting each satellite containment vessel with the reactor vessel containment vessel. Compartments 12A for steam generation means are formed adjacent to unitary reinforced concrete core 7A. Accessory compartments 14A are also formed adjacent to the unitary concrete core 7A to contain accessory piping and monitoring equipment. Vertical fuel storage void 16A will contain a fuel storage vessel in a suitable protective environment in close proximity to the central vertical void. FIGS. 5 through 8 illustrate alternate embodiments of this invention. FIG. 5 provides an embodiment whereby the satellite tanks contain steam generation means and separate compartments 12A of FIG. 4 are not required. FIG. 6 illustrates an embodiment wherein two satellite tanks 4A and a reactor vessel void 2A are provided. Steam generator voids 17A are also provided in the monolith as well as void 18A for auxiliary components. FIG. 7 provides an embodiment that features a single void 2A for the reactor vessel and FIG. 8 illustrates a support concept with reactor vessel void 2A and a void for a satellite tank with steam generation means 4A. The monolith support structure may preferably be used for the support of a reactor that is based on a single reactor vessel that is provided with a reactor core and its associated support systems and one or more satellite tanks that each include a pump for circulation of liquid metal and a heat exchanger. Each of the satellite tanks is connected to the reactor vessel by two conduits. These conduits are preferably arranged so that liquid metal may freely circulate between the reactor vessel and the satellite tanks. This arrangement provides for hydraulic interconnection of all vessels in the system thus providing a similar system behavior as in a pool vessel. Operation of this system with one pump out of service is thus made possible as well as means of accomodating a pump failure without the use of fast acting check valves. The advantage provided by this type of an apparatus is that a large amount of liquid metal is in direct contact with the reactor core so that if the core temperature rises, there is a substantial thermal inertia that resists a sharp or sudden rise in temperature. The upper and lower conduit means may be arranged to cause the hot liquid metal to flow from the reactor vessel to the satellite tank by convection to permit a flow of liquid metal to be established that will exert a cooling effect on the reactor core in the absence of a pumped or forced circulation of liquid metal. Each satellite tank may be provided with an intermediate heat exchanger that is a part of a primary liquid metal loop. Alternatively, the heat exchanger may include means for the direct generation of steam or it may be connected to a secondary liquid metal heat exchanger for the purpose of steam generation by a secondary liquid metal loop. It is contemplated that the satellite tank may be fabricated as a standardized module having a defined heat exchange capacity and/or steam generation capacity. One, two, three or more modules may be connected to a single reactor vessel that may also be substantially fabricated off site. It is envisioned that the reactor vessel size would remain uniform and for small power generation needs the available space within the core region utilized for storage of spent fuel. Thus for small power generation applications, a separate vessel for storage of spent fuel would not be required and for intermediate power generation applications the size requirements for this vessel would be reduced. The satellite tank is preferably fabricated with a transverse support structure and baffle assembly that extends across the central portion of the tank. This support structure provides means that engage and provide lateral support for the pump and heat exchanger. Vertical support is provided by a transverse support structure at the top of a satellite vessel. Provision for flow of liquid metal through the satellite tank is provided by insertion of a pump into a conduit which connects the bottom of the satellite tank with the gaseous space at the top of the vessel. The satellite tank is intended to have a lower plenum that is utilized to contain a supply of cold liquid metal. The lower plenum should be large enough to hold an amount of liquid metal that will impart a sufficient degree of thermal inertia to aid in the prevention of rapid temperature transients to either the pump or the reactor core if abnormal operating conditions are encountered. The reactor vessel will also have a lower plenum that will hold a quantity of liquid metal that still provides a degree of thermal inertia and mixing of flows from each satellite vessel. The upper and lower conduit means will have a relatively short length which can be basically considered to be extension of the vessel nozzles, piping support and snubber requirements and the need for auxilliary systems and structures are eliminated. Suitable thermal expansion means such as sinusoidal bellows may be used in the conduit piping. The lower conduit means may be fitted with an internal coaxial pipe that connects the pump outlet with the inlet of the core plenum that is placed around the lower portion of the reactor core. The internal coaxial pipe carries cold liquid metal that passes from the pump to the core plenum. The liquid metal flows from the core plenum upwardly through the core where it is heated and rises to the upper part of the reactor vessel. The hot liquid metal passes through the upper liquid metal conduit to the heat exchanger in the satellite tank. This heat exchanger may be an intermediate heat exchanger similar to that which is used in a pool vessel or it may be a uniquely designed heat exchanger that takes advantage of the design flexibility of the satellite vessel approach. The reactor vessel and satellite tank are provided with a unitary containment vessel that will contain any liquid metal leakage from the reactor vessel, satellite tank or the connecting conduits. The containment vessel extends completely around the satellite tanks and the reactor vessel and it may be provided with an exterior insulated cooling shroud that defines a space surrounding the reactor vessel and satellite tanks. The cooling shroud may be provided with cooling fins and a cooling medium such as air may be circulated between the containment vessel and the cooling shroud for a diverse and independent shutdown heat removal system. Sodium is the preferred liquid metal but other liquid metals and fluids may be utilized. The present invention is illustrated, as best seen in FIG. 1, in an embodiment for which one reactor vessel 2 is shown with four satellite tanks 6. The reactor vessel 2 and satellite tanks 6 are each secured within an individual containment vessel 4. Upper and lower containment passages 9 and 11 are provided to house upper and lower liquid metal conduits 10 and 12 which are connected to the reactor vessel 2. Upper and lower liquid metal conduits 10 and 12 afford communication between the reactor vessel 2 and each satellite tank 6. The reactor vessel 2 is best seen in FIG. 2 and is comprised of a reactor core 22 supported in the reactor vessel 2 by a support bracket 24 that is affixed to the side wall 25 of the reactor vessel 2 and to the side wall 23 of the reactor core 22. A core inlet plenum 28 is affixed to the lower inlet structure of the reactor core 22. The core inlet plenum 28 communicates with a core support structure 30 through a plurality of openings 29. The core inlet plenum 28 is connected to the central portion 26 of the reactor core 22 at the lower end 27. Openings 32 and 29A allow liquid metal to flow upwardly through holes (not illustrated) into the housing 35 for the fuel support modules. A single fuel assembly 80 is shown in the reactor core. The upper area of the reactor vessel 2 and of the satellite tank 6 are provided with slosh baffles 70 that maintain a quiescent sodium/gas interface and prevent any liquid from contacting the insulated reactor vessel cover or the insulated satellite tank cover 74. Inert gas space above the baffles 70, may be filled with argon. Housing 78 contains conventional core control rods, and refueling apparatus, not illustrated. The reactor vessel 2 may also further comprise an internal jacket 66 that is disposed on the upper side walls 64. Pipe inlet 68 is tapped off of reactor core plenum 28 to provide a flow of cool liquid sodium. The reactor vessel 2 is surrounded by containment vessel 4 which is sealed and is filled with an inert gas such as argon to act as a container to collect any liquid sodium that leaks from the reactor vessel 2. Alternate containment closure designs such as shown in FIG. 5 can be provided to enclose the inert gas space of the satellite or reactor vessels up to the reactor vessel suspension bracket 90. A cooling shroud 14 which bears a layer of insulating material 16 is placed around the containment vessel 4. Between the containment vessel 4 and cooling shroud exists a duct space 18 which may be utilized as a channel for circulation of air or any other coolant by any suitable moving means that are not illustrated. The containment vessel may be further extended above the suspension bracket to enclose as much of the inert gas space above the suspension bracket as desired. A gas space is also provided between the insulation and the cavity wall surface to enable circulation of coolant to keep the operating temperature of the concrete at pernissable levels. Alternatively, water cooling may be embedded beneath the concrete cavity surface for this purpose. The duct space 18 may serve as a housing for cooling fins 20 which are preferably mounted on the exterior wall of the containment vessel 4. The cooling fins are metal struts that are attached to the wall of the containment vessel to radiate heat. These cooling fins may serve as an alternate and independent means for decay heat removal by circulation of coolant past the fin surfaces. The system will also operate in the event a satellite vessel or the reactor vessel leaks liquid metal to the containment vessel. Alternate methods of operation of this system could include prefilling the containment vessel with liquid metal to further facilitate heat removal from the vessels. This mode of operation would also serve to limit the amount of liquid metal which could leak from the vessels. Expansion joints 8 are provided in the containment passages 9 and 11 between the satellite tanks 6 and the reactor vessel 2. A system satellite tank 6 is best seen in FIG. 2 and is further comprised of upper liquid metal conduit 10 and lower liquid metal conduit 12. Upper liquid metal conduit 10 is provided with expansion 3oints 38 that are integrally formed with the side wall 39 of satellite tank 6. The lower wall 58 of satellite tank 6 is provided with expansion joints 60. The upper liquid conduit 10 carries hot liquid sodium to the intermediate heat exchanger 36 where the heat is transferred to a secondary sodium loop 50. The heat exchanger 36 discharges cooled sodium into the lower plenum space 42. A second heat exchanger (not shown) is also present in the satellite tank. Each heat exchanger is supported by transverse support 90A. The lower plenum space 42 is configured to insure that a residual amount of liquid sodium is provided that will dampen thermal gradients encountered during abnormal operating conditions. A vertical housing 62A is provided to interconnect the lower plenum space 42 with the gaseous area 72 above the upper plenu. A vertical pump 44 is inserted in this housing, supported by the support structure 90A and it takes suction from plenum 42 with flow entering the pump suction via space 61. The intake of pump 44 is at intake port 37. Pump drive means 45, which comprise an electric motor, not illustrated, are located on the upper exterior outside surface of the satellite tank. Alternative pump means including electromagnetic fluid flow couplers may be used. The satellite tank 6 is provided with a transverse support structure 62 that supports pump housing 62A and provides lateral restraint to the housing and its pump, as well as lateral restraint to heat exchanger 36. The support structure also serves as a baffle to the stagnant liquid metal within it to separate the hot and cold pools. Inspection and fabrication assembly access holes 82 are provided in both the satellite tank 6 and the reactor vessel 2 to allow for inspection of the various components of the reactors and for use during the fabrication phase. An inspection and manufacture hole (not shown) is also provided in the top of the satellite vessel. Satellite suspension bracket 90 is used to support the weight of satellite tank 6 on concrete support 92 to which it is secured with bolts 91. Reactor vessel 2 is similarly supported on reactor vessel suspension bracket 94. Thermal insulation 96 is provided at the upper end of reactor vessel and satellite tank 6. Radiation shielding is provided in the upper portions of the vessels' support structure (above thermal insulation 96). Seismic forces are accomodated by having the reactor vessel support 94 rigidly attached to a concrete support 92 and satellite vessel supports 90 rigidly attached to the same concrete structure. The containment vessel is attached to the top of the reactor vessel and satellite vessels and moves with the vessels under a seismic disturbance. Expansion bellows 8 accomodate relative motion between the containment areas surrounding the reactor vessel and satellite vessels. The pump flow circuit comprises intake port 37 and discharge passage 54. Slip 3oint 53 at the lower portion of the discharge passage 54 is connected to liquid metal duct 57 that is coaxially located inside lower liquid metal conduit 12. Liquid metal duct 56 is supported by support brackets 63 and is provided with slip joint 56 to permit thermal expansion. The liquid metal duct 57 is connected to the core inlet plenum 28. In operation, the liquid sodium passes out through the reactor top 34 of reactor core 22 at a temperature of about 950.degree. F. The liquid sodium then passes through the upper support structure 78, mixes in the upper plenum area of the reactor and exits to the satellite tank through the upper liquid metal conduit 10. Within the satellite vessel the liquid sodium mixes within the upper plenum area and enters the heat exchanger 36. In the heat exchanger 36 heat is transferred to a secondary sodium heat transfer loop 50 that is connected to the heat exchanger 36. The cooled liquid sodium exits the heat exchanger into satellite tank plenum 42 at a temperature of about 670.degree. F. The cool sodium passes via channel 61 up to pump intake opening 37. The liquid sodium is taken into intake port 37 of pump 44 and discharged through discharge passage 54 into the liquid metal duct 57. The slip joint 56 is arranged so that any expansion of liquid metal duct 57 will not place any transverse stress on pump 44 or reactor core 22. The liquid sodium flows into the core inlet plenum 20 and through the core support structure 30 and upward through openings 32 upwardly through holes 33 (not shown) into the housing 35 for the fuel support modules and passes fuel assemblies such as fuel assembly 80 where it is heated to a temperature of about 950.degree. F. The hot sodium flows out of the top of reactor core 34. The static level 84 of the liquid sodium in the reactor is reached only when the pump is not operating. The satellite tank operating level 86 is less than the reactor vessel operating level 88. This is due to the pressure losses in upper liquid conduit 10. FIG. 9 illustrates a further emobodiment of the invention wherein satellite tank 6 is provided with an alternate means for thermal expansion. The lower wall 58 of satellite tank 6 does not have any provisions for expansion where it is joined by lower liquid metal conduit 12. Expansion means are provided by rollers 102 which support the satellite tank 6 and permit transverse motion caused by expansion of lower liquid metal conduit 12. The rollers are guided to move in a linear direction and seismic snubbers 103 are provided to limit the satellite vessel movements in the event of an earthquake. The reactor vessel is rigidly supported by means of support 94 to the concrete support structure 92. The upper part of satellite tank 6 may be provided with an internal cooling jacket 104 which provides cooling by the circulation of cool liquid sodium which is taken from the output of pump 44 through line 106 which is connected to liquid metal duct 57 at outlet 108. FIG. 10 is a view of an embodiment of a satellite tank of the invention having a pump 110 that is coaxially located in the satellite tank 6 and supported principally by upper tranverse support 113. Support bracket 111 which primarily supports heat exchanger 115 may be placed at a level in the satellite tank 6 so that the vertical expansion of the satellite tank relative to the reactor vessel will not place undue stress on the horizontal conduits and pump discharge line 57. The pump housing 112 is covered on its upper exterior with a mixing baffle 114 that directs primary sodium into heat exchanger 115 that is provided with tubes 117, primary liquid sodium flows through the tubes 117, to plenum space 126. Secondary sodium enters from secondary sodium inlet duct 122 from an inlet (not shown) and is passed to the lower part 124 of the heat exchanger 115. The secondary sodium is directed around the periphery of the heat exchanger by a directing tube (not shown) which is connected to the end 128 of secondary sodium inlet duct 122. The secondary sodium outlet 130, is connected to a connecting tube (not shown) at inlet 131. The connecting tube extends partway around the periphery of the upper region of heat exchanger 115 to collect the hot secondary sodium that flows up past tube 117 which is representative of a plurality of tubes in the heat exchanger and heat exchanger baffles 123. Coils 116 are placed in the upper region of the satellite tank 6 and are connected to inlet conduit 118 and outlet conduit 120 to form a backup cooling system for use if the primary cooling system fails. Holes 119 may be of different dimensions or may have different spacing or may have a combination of different dimensions and spacing around the circumference of mixing baffle 114 to promote uniform flow of liquid metal to avoid any localized overheating. The pump bracket 132 supports liquid metal duct 57. Flow nozzle 56a as shown in FIG. 5 provides an alternate means for connection in pump discharge line 57 that may be utilized in any of the embodiments of this invention. A gap between the flow nozzle 56a and the end of flow conduit 57a is provided to accomodate thermal expansion effects. The nozzle size and pump discharge pressure is such that a small induced flow directly from the bottom plenum of the satellite tank is possible under normal operating conditions. At shutdown when natural circulation of the primary coolant is desirable, the gap between the nozzle 56a and duct 57a will assure an unimpeded and direct flow path for natural convective flow to the core. Line 57a may be provided with a diffuser to minimize pressure losses, also spacers 61 may also be provided within the annular space of the coaxial ducts. FIG. 11 illustrates a sectional elevation of a combined primary and secondary system satellite tank that includes a helical coil steam generation system. The containment vessel upper transverse support structure and insulation is now shown in the view for clarity. Central downcomer 146 acts as a support and housing for the pump 148. Support bracket 147 also provides support and stability. An annular mixing plenum 150 is provided above the upper end of heat exchanger 152. Hot liquid primary sodium is passed from upper liquid conduit 10 through holes 151 in the annular mixing plenum and then downward through vertical tubes 154 which run through annular heat exchanger 152. Heat exchanger 152 which may be welded in place or detachably affixed with suitable fasteners, is provided with baffles 156 and associated to rod 165 that insure slow upward passage of secondary sodium that is received from secondary sodium inlet duct 158. Hot secondary sodium is taken from the heat exchanger at header 160 into line 162 which directs the flow to the entrance mixing area 164 of the steam generator 161 which is provided with steam generator coils 166. The steam generator coils 166 are connected to water inlet 168 and steam outlet 170. They are formed into a helix which is partially shown. Secondary liquid sodium flows down around the steam generator coils 166 through annular heat exchanger chamber 172. At the lower end 171 of annular heat exchanger chamber 172, pump intake ports 174 direct the cool secondary liquid sodium to the pump chamber. FIG. 11A is a partial cutaway view of the pump chamber of the pump 148. The primary pump rotor 140 and secondary pump rotor 142 are mounted on common pump shaft 144. The pump intake ports 174 direct the liquid metal to the intake channel 176 which directs the liquid sodium to secondary pump rotor 142. The output is pushed through sodium inlet duct 158 to heat exchanger 152 which is shown in FIG. 6. Cool primary liquid sodium is passed through holes 180, shown in FIG. 6, to intake channels 182. The cool liquid sodium is passed from intake channels 182 to primary pump rotor 140 which pushes liquid sodium to the reactor core plenum, (not shown) through duct 57. Pump bearing 145 is lubricated by a feed of liquid sodium that is supplied by a tube (not shown) that is tapped off of the pump discharge circuit. Upper connecting joints 181 and lower connecting joints 183 are provided to facilitate removal of the pump internals. A perforated shield 178 that may have differently sized or differently spaced holes 180 is provided to promote uniform mixing of cool sodium in lower plenum space 42. The lower plenum space 42 is hydraulically connected with the reactor vessel (not shown) and other satellite vessels (not shown) through lower liquid metal conduit 12. Annular space 159 serves the function of providing a conduit for the lower plenum to communicate with the upper gaseous space. Spacers 163 are provided for lateral stabilization of the steam generator under seismic conditions. The normal operating level of liquid sodium 184 in the satellite tank 6 is at the upper level of the steam coils 166. The level of liquid sodium 185 in the pump chamber 186 is above the level of pump rotor 142 shown in FIG. 6A. During operation, the primary pump discharge is at a higher pressure than the secondary circuit pump discharge and a continuous linkage of primary flow enters the secondary pump via the leakage past the pump bearing. The primary coolant volume is maintained by an auxilliary make up and clean up system (not shown) which is housed outside the satellite tank. This system (not shown) removes a flow equal to or greater than the pump bearing leakage flow, purifies the secondary sodium, and returns sodium to the primary system to maintain its inventory as well as any excess flow back to the secondary system circuit. In the event a steam generator tube leak occurred, the secondary system could be pressurized as a result of sodium-water reactions. A rupture disk at the top of the satellite vessel wou1d serve to vent the pressure to a blow off system (not shown) thus maintaining the pressure integrity of the secondary sodium system circuit. Alternative pump means such as electromagnetic pumps or electromagnetic fluid flow couplers which do not require pump bearing seal leakages can be utilized for this concept. Alternatively, these steam outlet nozzles may be located at the bottom of the vessel by routing the tubing discharge from the top of the helical bundle down past the exterior of heat exchanger 214. FIG. 12 is a diagrammatic sectional elevation of a steam generator for use in a secondary sodium loop according to the invention. The generator comprises a tank 200, that has water inlets 202 and 204 and steam outlets 206 and 208. The intake pipe 210 is connected to a nozzle (not shown) as a part of a secondary liquid sodium loop that transfers heat from a primary sodium loop to the steam generator 200. Water inlets 202 and 204 are connected to a plurality of tubes that are formed into a helical coil that is configured as an annular heat exchanger 214. The tubing exits the helical coil and is directed to manifolds at the steams outlet nozzles 206 and 208. The annular heat exchanger 214 is provided at the upper end 216 with a manifold 218 that allows hot sodium to enter the annular heat steam generator. The lower plenum has a central downcomer 224 that is perforated with holes 225 that permit cool sodium to flow inwardly and upwardly to pump 226 which is operated by motor 227. The output pipe 228 of the pump 226 acts as a conduit to bring cool liquid sodium to outlet nozzle 230. Inert gas space 232 is filled with a suitable inert gas such as argon. Bracket 234 stabilizes the pump housing 235 and bumpers 236 provide lateral support while accomodating thermal expansion and seismic forces. Nozzle 230 sidewalls are envisioned to be extended coaxially around piping 228 and 210 up to the entrance and exit nozzles for this piping at the top of the satellite tank, using suitable thermal expansion means such as bellows. Appropriate insulation (not shown) may be used on the upper interior walls of the steam generator. Any alternate steam generator may be utilized with the novel reactor of the invention such as the steam generator disclosed in ASME paper 80-C2/NE-29 which is incorporated by reference. FIG. 13 is a partial cross-section of the outer walls of the reactor tank or satellite tanks of the invention. Vessel wall 300 is tht interior wall of either the reactor vessel or the satellite tank. The containment vessel 301 surrounds the vessel wall and on its exterior surface has a plurality of cooling fins 303 that are provided between containment vessel 301 and shroud 302. An insulating layer 304 is placed on the exterior of shroud 2 between the cavity wall of the liner 305, for concrete wall 306, to define an annular space 307 between insulating layer 304 and liner 305. The liquid metal reactor of the invention may be fabricated in dimensions that are selected according to the desired output. Generally reactor vessel and satellite tanks may be from about 5-15 meters. In diameter and from about 15-23 meters in height. The upper and lower liquid metal conduits may be about 50 cm-130 cm in diameter and the containment vessel may be spaced approximately 20-35 cm from the satellite tanks and reactor vessel. The cooling fins may be 0.1 cm-5 cm thick and may extend from the wall of the containment vessel for 2 cm-20 cm to the shroud. The reactor vessel and the satellite tanks may be spaced from 1 to 10 meters apart. While the invention has been disclosed with respect to the particular drawings and embodiments shown and described, the invention is not to be limited thereby but is only to be limited by the scope of the appended claims.
039754715	description	The invention is further explained in the following examples. EXAMPLE 1 As fuel particles there were employed spherical kernels of UO.sub.2 having a grain diameter of 220 .mu.m. These particles were provided with a three layer coating of a pyrolytically deposited coating, the total thickness of the coating being 180 .mu.m. An intermediate layer of SiC having a thickness of 23 .mu.m was present between the two outermost carbon layers. (The thickness of the SiC layer is included in the total thickness of the carbon layers). The particles having a diameter of 580 .mu.m and a density of 2.3 g/cm.sup.3 contained 21.85 weight % uranium. The fertile particles of ThO.sub.2 having a grain diameter of 617 .mu.m were provided with a double layer coating of carbon (also pyrolytically deposited) having a total thickness of 144 .mu.m. The particles having a diameter of 905 .mu.m and a density of 3.99 g/cm.sup.3 contained 63.23 weight % thorium. As the molding powder there was employed a mixture consisting of 64 weight % natural graphite, 16 % graphitized petroleum coke and 20 % of a thermoplastic phenol-formaldehyde novolak resin as a binder. The fuel and fertile particles were overcoated in separate processes with the molding powder with addition of methanol in a rotating drum. The amounts supplied were so selected that in the finished compacts there was present a graphite matrix portion of about 55 volume %. There was sprayed on the finished overcoated particles stearic acid as a 5 % solution in trichloroethylene as the lubricant together with 1 % hexamethylenetetramine as the hardener. Based on the binder portion in the matrix powder the hardener addition amounted to 3 weight % and the lubricant addition to 15 weight %. The particles were dried at room temperature. 5.2 grams of overcoated fuel particles and 49 grams of overcoated fertile particles after admixing were filled into a cylindrical compression mold having a diameter of 16 mm and a length of 180 mm and held constant at 150.degree.C. and compressed from both sides. The amounts given correspond at a matrix density of 1.70 g/cm.sup.3 to a cylinder length of 100 mm. This length was attained at a compression of less than 20 kp/cm.sup.2. After expulsion, the compacts underwent a closed heat treatment at 1800.degree.C. Subsequent electrolytic decomposition of over 100 of the compacts in dilute nitric acid and fluorometric uranium analysis gave values of 7 to 15 micrograms of free uranium and the chemical thorium analysis of 10 to 25 micrograms per compact which is less than half the heavy metal content of one single particle. Thus, it is proven that no particles are destroyed by this process of the invention. The found values only depend on a surface contamination with uranium or thorium. EXAMPLE 2 Using the process of published German application DAS 2 215 577 as described on column 4, lines 1-34 FIG. 1 and Example 1, there was employed a molding powder of the same composition as for the overcoating and the composition was compressed in cavities of an elastic rubber disc to form three-dimensional isotropic granulates having a 1.5 mm particle size. The entire disclosure of the German application DAS 2 215 577 is hereby incorporated by reference. A mixture of 10.4 grams of overcoated fuel particles and 24.5 grams of overcoated fertile particles according to Example 1 and 15 grams of the granulates produced in the elastic rubber disc were coated wih hardener and lubricant and compressed in a manner analogous to Example 1. The length referred to (100 mm) at a matrix density of 1.7 g/cm.sup.3 was already obtained at about 20 kp/cm.sup.2. The uranium analysis carried out analogous to Example 1 showed 15 micrograms of free uranium. Free thorium was ascertained as 20 micrograms, which likewise proves there was a compression free of particle damage. The composition applied to the overcoat outer surface can consist of or consist essentially of the hardener and lubricant.
053613775	claims	1. A method for producing electrical power from steam generated by a nuclear reactor comprising the steps of: (a) providing a nuclear reactor engaged to a steam generator for generating steam when heated aqueous product is passed therethrough; (b) passing heated aqueous product through the steam generator of step (a) to produce steam; (c) passing the produced steam of step (b) through a superheater to superheat the produced steam to a temperature where the produced steam has an enthalpy above about 1450 BTU per lb.; (d) passing the superheated produced steam of step (c) through a first turbine to expand the superheated produced steam and produce steam having an enthalpy above about 1250 BTU per lb; (e) reheating the produced steam of step (d) to obtain a reheated steam having an enthalpy above about 1470 BTU per lb.; (f) passing the obtained reheated steam of step (e) through a second turbine coupled to a generator in order to expand the obtained reheated steam and generate electrical power with the generator; (g) discontinuing the passing step (c) and subsequently bifurcating the produced steam of step (b) into a first steam stream and a second steam stream; passing the first steam stream through a heat exchanger; passing the second steam stream through a third turbine to expand the second steam stream and produce an expanded second steam stream; dividing the expanded second steam stream into a first expanded second steam stream, a second expanded second steam stream; and a third expanded second steam stream; passing the first expanded second steam stream through said heat exchanger; passing the second expanded second steam stream into said heat exchanger to heat the same through a heat exchange relationship with the first steam stream and the first expanded second steam stream and produce a heated second expanded second steam stream; passing the third expanded second steam stream through a third heater; passing the reheated second expanded second steam stream through a fourth turbine coupled to a second generator to expand the heated second expanded second steam stream, causing the second generator to generate electricity and produce an expanded heated second expanded second steam stream. (a) providing a nuclear reactor engaged to a steam generator for generating steam when heated aqueous product is passed therethrough; (b) passing heated aqueous product through the steam generator of step (a) to produce steam; (c) passing the produced steam of step (b) through a superheater to superheat the produced steam; (d) passing the superheated produced steam of step (c) through a first turbine to expand the superheated produced steam and produce steam; (e) reheating the produced steam of step (d) to obtain a reheated steam; (f) passing the obtained reheated steam of step (e) through a second turbine coupled to a generator in order to expand the obtained reheated steam and generate electrical power with the generator; (g) discontinuing the passing step (c) and subsequently bifurcating the produced steam of step (b) into a first steam stream and a second steam stream; passing the first steam stream through a heat exchanger; passing the second steam stream through a third turbine to expand the second steam stream and produce an expanded second steam stream; dividing the expanded second steam stream into a first expanded second steam stream, a second expanded second steam stream; and a third expanded second steam stream; passing the first expanded second steam stream through said heat exchanger; passing the second expanded second steam stream into said heat exchanger to heat the same through a heat exchange relationship with the first steam stream and the first expanded second steam stream and produce a heated second expanded second steam stream; passing the third expanded second steam stream through a third heater; passing the reheated second expanded second steam stream through a fourth turbine coupled to a second generator to expand the heated second expanded second steam stream, causing the second generator to generate electricity and produce an expanded heated second expanded second steam stream. (a) providing a nuclear reactor engaged to a steam generator for generating steam when heated aqueous product is passed therethrough; (b) passing heated aqueous product through the steam generator of step (a) to produce steam; (c) bifurcating the produced steam of step (b) into a first produced steam stream and a second produced steam stream; (d) passing the first produced steam stream of step (c) through a superheater to superheat the first produced steam stream; (e) passing the superheated first produced steam stream of step (d) through a first turbine to expand the superheated first produced steam stream and produce an expanded superheated first produced steam stream; (f) reheating the produced expanded superheated first produced steam stream of step (e) to obtain a reheated steam; (g) passing the obtained reheated steam of step (f) through a second turbine coupled to a generator in order to expand the obtained reheated steam and generate electrical power with the generator; (h) recovering expanded steam from the second turbine, said recovered expanded steam from the second turbine having an enthalpy greater than about 1050 BTU per lb.; condensing the recovered expanded steam into an aqueous product; passing the aqueous product through a first pump to pump the aqueous product to produce a pumped aqueous product; passing the pumped aqueous product through at least one first heater to produce an aqueous product having an elevated temperature, an elevated pressure and elevated enthalpy; passing the aqueous product having an elevated temperature, pressure and enthalpy through a second pump; passing subsequently the aqueous product from the second pump through at least one second heater to produce said heated aqueous product which is for being passed through the steam generator; (i) bifurcating the second produced steam stream of step (C) into a first steam stream and a second steam stream; passing the first steam stream through a heat exchanger; passing the second steam stream through a third turbine to expand the second steam stream and produce an expanded second steam stream; dividing the expanded second steam stream into a first expanded second steam stream, a second expanded second steam stream; and a third expanded second steam stream; passing the first expanded second steam stream through said heat exchanger; passing the second expanded second steam stream into said heat exchanger to heat the same through a heat exchange relationship with the first steam stream and the first expanded second steam stream and produce a heated second expanded second steam stream; passing the third expanded second steam stream through a third heater; passing the reheated second expanded second steam stream through a fourth turbine coupled to a second generator to expand the heated second expanded second steam stream, causing the second generator to generate electricity and produce an expanded heated second expanded second steam stream. 2. A method for producing electrical power from steam generated by a nuclear reactor comprising the steps of: 3. The method of claim 1 additionally comprising recovering expanded steam from the second turbine, said recovered expanded steam from the second turbine having an enthalpy greater than about 1050 BTU per lb.; condensing the recovered expanded steam into an aqueous product; passing the aqueous product through a first pump to pump the aqueous product to produce a pumped aqueous product; passing the pumped aqueous product through at least one first heater to produce an aqueous product having an elevated temperature, an elevated pressure and elevated enthalpy; passing the aqueous product having an elevated temperature, pressure and enthalpy through a second pump; passing subsequently the aqueous product from the second pump through at least one second heater to produce said heated aqueous product which is for being passed through the steam generator. 4. The method of claim 2 additionally comprising recovering expanded steam from the second turbine, said recovered expanded steam from the second turbine having an enthalpy greater than about 1050 BTU per lb.; condensing the recovered expanded steam into an aqueous product; passing the aqueous product through a first pump to pump the aqueous product to produce a pumped aqueous product; passing the pumped aqueous product through at least one first heater to produce an aqueous product having an elevated temperature, an elevated pressure and elevated enthalpy; passing the aqueous product having an elevated temperature, pressure and enthalpy through a second pump; passing subsequently the aqueous product from the second pump through at least one second heater to produce said heated aqueous product which is for being passed through the steam generator. 5. The method of claim 1 additionally comprising passing part of the produced steam of step (d) to a feed water pump to operate the same. 6. The method of claim 2 additionally comprising passing part of the produced steam of step (d) to a feed water pump to operate the same. 7. A method for producing electrical power from steam generated by a nuclear reactor comprising the steps of: 8. The method of claim 7 additionally comprising an apparatus produced to accomplish the method steps.
055307280	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to the measurement of objects using an optical measuring system. More particularly, the invention relates to a method of measuring the lengths of spent nuclear fuel rods so that the mass of the rods can be determined. 2. Discussion of Prior Art In the reprocessing of spent nuclear fuel it is important for inventory and accounting purposes that the amount of spent fuel to be reprocessed is calculated accurately. This requires a system for obtaining the mass of the spent fuel rods to be reprocessed. In one particular application, measurement of the mass of the fuel rods takes place in a decanning cave in which a metal cladding is stripped from fuel elements to provide fuel rods for subsequent reprocessing. Because of the hostile environment existing within the cave, systems for direct weighing of the fuel rods have many disadvantages. For example, the components of such a system are subjected to adverse conditions of radiation, heat, mechanical shock and water existing within the cave. Thus, these systems tend to have a short operational life and require frequent servicing and maintenance. Because of the inaccessibility and the danger of radiation exposure associated with handling the weighing system components in the cave, direct weighing systems are unsatisfactory. One known type of direct weighing system has a weighing device in which a weighing platform is supported on solid state load cells located in the decanning cave. Although such a system can produce accurate results, it has not proved satisfactory because the load cells are unable to withstand the mechanical shock loads and the high radiation fields existing within the cave and so tend to fail after a short period of time. SUMMARY OF THE INVENTION According to the present invention there is provided a method of measuring the linear dimension of an object, said method comprising the steps of obtaining a first optical image, said first optical image being of a support surface, digitising the first optical image and storing the first optical image in the form of grey level values, obtaining a second optical image, said second optical image being of an object to be measured when placed on said support surface, digitising the second optical image and storing the digitised second optical image in the form of grey level values, and processing the two stored digitised images to obtain a difference in grey level values between the first and second digitised images so as to determine the required linear dimension of said object. To overcome the above-mentioned disadvantages of the prior art we have devised a non-intrusive measuring system as defined above which optically measures the lengths of the objects such as fuel rods. Having obtained the length of a fuel rod the system is able to calculate the mass of the fuel rod using a known value of mass per meter for the fuel material. An advantage of the present invention is that it provides a non-intrusive, indirect weighing system for spent nuclear fuel rods and that none of the components of the system is located within the hostile environment of the decanning cave. As a result, the measuring system has a longer operational life than direct weighing systems and requires only a minimum amount of servicing and maintenance. A further advantage of the present invention is that once the system is set up and calibrated it does not require any further input from an operator. In a preferred embodiment the linear dimension is the length of a spent nuclear fuel rod. Preferably the support surface forms part of a tray for receiving spent nuclear fuel rods. The first and second digitised optical images are each stored as a plurality of columns, each column containing a plurality of pixels having a grey level value, the difference in grey level values of corresponding columns in the first and second optical images being determined as a root mean square value. Preferably a root mean square value is selected as a threshold value, said threshold value being selected so that root mean square values above the threshold value are indicative of the presence of a fuel rod, and root mean squares values below the threshold value are indicative of the support surface. In a preferred embodiment the two stored digitised images are processed using an algorithm which moves across the columns and processes each column in turn, the number of columns having a root means square value above said threshold value being a measure of the length of the fuel rod. Preferably the algorithm is adapted to perform an averaging test on the first of said columns which indicates a root mean square value above the threshold value and on a plurality of columns next succeeding said first column, the result of said test indicating whether or not said root mean square value of said first column represents one end of the fuel rod. The averaging test may be performed on the said first column and the next nine succeeding columns. Preferably the algorithm is also adapted after establishing said one end of the fuel rod to perform an averaging test on a subsequent column indicating a root mean square value below said threshold value and on a plurality of next succeeding columns, the result of said test indicating whether or not said root mean square value of the said subsequent column represents the other end of the fuel rod. The averaging test may be performed on the said subsequent column and the next nine succeeding columns.
	abstract	The invention relates to a multi-component cladding for a nuclear fuel rod that includes a combination of ceramic and metal components. More particularly, the invention is directed to a cladding that includes a ceramic composite having a zirconium composition deposited thereon to form a zirconium coated ceramic composite. The ceramic composite includes a ceramic matrix and a plurality of ceramic fibers. The cladding is effective to protect the contents of the cladding structure from exposure to high temperature environments during various load conditions of a nuclear reactor.
	summary
044407190	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows pertinent features of a direct cycle boiling water reactor (BWR) system. The invention is however applicable to other kinds of reactor systems as well. Various examples of such systems are shown in the book Nuclear Power Plants by R. L. Loftness, published in 1964 by Van Nostrand Company. Reactor systems typically include a containment structure 10 for structural protection. Within the containment 10, a pressure vessel 11 is suitably mounted on a pedestal 13. The pressure vessel 11 includes a head 14, which can be removed to refuel the reactor and give access to the reactor core 15. Reactor fuel destined for or removed from the core is temporarily stored underwater in a suitable upper fuel pool 16 containing demineralized water and located in the containment building. The containment 10 also holds a pressure suppression pool 17 for condensing steam in the unlikely event of certain kinds of postulated reactor accidents. The suppression pool 17 is suitably filled with water. One such accident involves the leakage of steam from the pressure vessel 11 within a drywell 18 of the reactor, which is bounded by suitable walls 20 defining underwater openings 22 in the suppression pool 17. A suitable pressure suppression system including a suppression pool is shown and described by C. P. Ashworth et al in an article entitled "Pressure Suppression," Nuclear Energy, August 1962, pages 313 through 321. The reactor generates steam which drives a turbine/generator combination 24 for producing electricity used by consumers and to some extent by nuclear installation itself. The steam generated by the reactor travels to the turbine/generator 24 along at least a single pipe or line 26, the flow through which is suitably controlled as for example by one or more valves, as in indicated at 27. A branch 30 of line 26 extends through suitable valves 32 and 33, the walls 20 of the drywell 18, and the containment 10, and carries steam to a Reactor Core Isolation Cooling (RCIC) system 35 to which the invention disclosed herein relates. The generation of steam in the reactor reduces the inventory of cooling water within the pressure vessel 11. The water within the pressure vessel can be said to "cool" the reactor fuel in the core 15, or the nuclear fuel can be considered to heat the water in the pressure vessel. The loss of water from the pressure vessel due to steam generation can be understood in terms of a "phase change" between the liquid and gaseous phases of materials. The water lost during steam production is replenished during reactor operation by a feedwater supply system 37 which includes a condensate tank 38. During normal operation this replacement water is transported to the pressure vessel by a suitable feedwater pump (not shown) operating through a pipe or line 40 extending to the pressure vessel 11 from a condensor (not shown). If the feedwater supply system 37 becomes unavailable, the RCIC substitutes for it long enough to permit complete reactor shutdown and depressurization of the vessel 11 from its normal operating pressure in excess of about 1,000 psi. The details of the normal operation of the feedwater supply system 37 and the features of its construction as well as its connection and interaction with the turbine/generator 24 by for example a pipe or line 43 are well known in the art. This is also true of the condensate tank 38 which holds water available to supply the condensor (not shown) of the feedwater supply system 37 in the course of normal operation. The water from the condensate tank 38, is also available (as is water from the upper fuel pool 16 and the suppression pool 17) to the RCIC by means of pipes or lines 45. As can be seen in FIG. 2, these lines 45 provide an inlet for water to a suitable valve 47. Other valves (not shown) may be employed to determine which of the various sources of water (condensate tank 38, supression pool 17, or fuel pool 16) or which combination of them will actually be employed to supply the water. In general, the various valves referred to above may be operated manually or may be governed by a suitable valve control unit or units (not shown), which are responsive to the liquid level of coolant in the reactor vessel 11 or an indication of the non-availability of the feedwater supply system 37. As will be shown, the steam generated by the reactor and passing through the branch 30 of the steam line 26 of the reactor system is employed to feed back the water to the reactor pressure vessel 11 for replenishment. This feedwater flows through a pipe or line 50. Some RCIC systems presently known and operating employ features such as those shown in U.S. Pat. No. 3,431,168 by J. E. Kjemtrup and issued in 1969. The patent indicates the provision of a turbine pump which supplies stored water to the reactor vessel to maintain the coolant level. The turbine shown is steam operated, and independent of the possible loss of electric power at plant site. However, the RCIC system of this invention includes jet pump and injector stages and cooling means or heat exchangers as will be explained hereinafter. A heat exchanger 53 is for example coupled to the output of the steam-driven injector 60, which drives a fluid such as water by applying steam from the reactor to it. Such an injector is commercially available from Pemberthy Corporation, and pertinent details of its operation will be provided hereinafter to permit a fuller understanding of the art involved in this invention. The heat exchanger 53 utilizes a suitable cooling medium for example low temperature water as a working fluid, which enters the heat exchangers 53 at an opening 54 and departs at an output 55. Convective forces, pump or other driving means (not shown) causes the cooling medium to flow through the heat exchanger 53. The heat exchanger 53 may be one of a variety of heat transfer devices available on the market to remove heat from a fluid traveling in a tube, pipe, or line. No specific model or type is recommended for application herewith. The details of such a selection fall well within the competence of one skilled in the art. The heat exchanger 53 taken in combination with the steam driven injector 60 can be considered to be a single stage of the injection system described. Valve 33 controls the application of steam from the reactor to respective branches of suitable pipes or lines 63 and 65. Pipe 63 leads steam to the steam input side of steam-driven injector 60. Pipe 65 provides steam to another or second steam-driven injector 70. The injectors 60, 70 exhibit similar features. In particular each includes input apertures for steam and water. Accordingly, inlets are provided for each injector: one for water, and one for steam. On the output side of injector 60, there is a drain or overflow pipe or line 71. A main output line 72 is utilized in normal operation to permit the discharge of water. During start-up and during operation, a suction force is created within each steam-driven injector by passing high pressure steam through an expansion nozzle. The expansion nozzle causes the high pressure inlet steam to be accelerated to supersonic velocity at subatmospheric pressure. The subatmospheric pressure causes the water to begin to pass through the injector in a suction chamber 74, and condensation of the high velocity steam in a condensation region 75 creates a high velocity fluid stream that produces a secondary suction effect in an after-region of the injector, which in turn shuts a check valve on the overflow or drain line and directs flow through the main output line 72. Considering only the first steam injector 60, steam is introduced at its steam inlet and flows through the suction chamber 74. The steam then passes through the condensation region 75 and initially departs through a drain check valve 73, into a suitable external region (not shown) adapted to accept overflow from the RCIC. The output of the injector 60 flows through a check valve 76 and respective branches of suitable pipes or lines 80 and 81 to convey output water to the appropriate ones of the components shown in FIG. 2. In particular, line 80 leads to heat exchanger 53 and line 81 leads to a heat exchanger 88, which has an inlet 89 for coolant and an outlet 55. Respective heat exchangers 53, 88 connect with outpit lines or pipes 91, 92 which are respectively coupled to a jet pump 100 and to steam injector 70. The output of the jet pump 100 provides water to the suction chamber 74 of the injector 60 through a line or pipe 101. The jet pump 100 has an inlet to receive water from any of lines 45 through valve 47. Steam injector 70 has a suction chamber 110, a condensation region 111, an overflow or drain check valve 113, and a main check valve 115. Its features are accordingly generally similar to those of steam injector 60. Suitable steam injectors and a jet pump for application to this invention may be obtained from the Pemberthy Division of Houdaille Industries in Illinois. For example, a pair of Pemberthy Automatic Injectors, and a jet pump of the LL, LM, or LH model series may be employed. The steam injector is considered to be a two-phase device, since it injects a liquid (i.e., water) by use of a gas (i.e., steam) and involves the conversion of the gaseous phase to a liquid phase upon condensation during its passage through the injector. The jet pump is of course a single-phase device, since water injected into other water involves only a single (the liquid) phase of H.sub.2 O. In the jet pump 100, water under high pressure from injector 60 through heat exchanger 53 is driven through a nozzle 117 of the pump, establishing a suction effect tending to draw water through valve 47. The nozzle 117 produces a high velocity stream of water to establish low pressure within the pump 100, which creates the desired suction. On the discharge or output side of the pump 100, there is a diffuser 119, shaped to reduce flow velocity and gradually boost pressure while minimizing energy losses. Further information regarding such pumps is available in Igor J. Karassik et al, Pump Handbook by McGraw Hill, published in 1976. Operation is initiated by turning on the steam at valve 33, whereby steam received from pressure vessel 11 (FIG. 1) flows through steam injectors 60 and 70 and exits through respective check valves 73, and 113. Introduction of water into both of these injectors 60, 70, by turning on valve 47, establishes condensation in the respective condensation regions 75 and 111, which produces a low pressure in the after-region, thereby closing the check valves 73, 113 and permitting an output through respective check valves 76, 115. A portion of the output of injector 60 is fed back through branch line 80 and jet pump 100. This provides higher pressure water to the suction chamber 74 and boosts the output pressure level of the first injector 60. The efficiency of the first injector 60 is maintained, by cooling the steam-heated water at the output of the injector 60 with heat exchanger 53 prior to introduction of the water into suction chamber 74 through jet pump 100. Without the heat exchanger 53, the overall output pressure level of the first injector 60 would be reduced. Water from the output of the first injector 60 is carried through pipe or line 81 to heat exchanger 88 and thence through a line 92 to second injector 70. Cooling of the water by heat exchanger 88 permits the second injector 70 to operate at an acceptably high efficiency level. Output water from second injector 70 is supplied to line 50 at an output pressure in excess of about 1,000 psi--enough to overcome the back pressure offered by the reactor and to permit the injection of replenishment water thereinto. The above description pertains to a single possible embodiment of the instant invention and is susceptible of reasonable modification by those skilled in the art. However, this invention is not meant to be limited to the preferred embodiment shown and described. Rather the claims set forth the invention and they are intended to cover all modifications reasonably within the spirit and scope of the invention.
050139452	claims	1. A linearly operating motor for stepwise advance of a driven member (11), such as a shaft, comprising an elongated body (10) with length variable properties, and means for changing the length of said body (10), said driven member (11) being provided with at least a first pivotally mounted locking means (12) and having a through-hole to drive said driven member (11), said first locking means (12), upon length increase of said body (10), being adapted to first swing so that a grip is obtained between said driven member (11) and hole and adapted to move said locking means (12) and the driven member (11) in the advance direction thereof. 2. A linearly operating motor for stepwise advance of a driven member (11), such as a shaft, comprising an elongated body (10) with length variable properties, and means for changing the length of said body (10), said driven member (11) being provided with at least a first locking means (12), said first locking means (12), upon length increase of said body (10), being adapted to be brought into engagement with said driven member (11) by means of said length increase and said body (10) being adapted to act upon the first locking means (12) in such a manner that said locking means and thereby also said driven member (11) are displaced at length increase of the body (10), said locking means comprising a ring (30) having a through-hole for the driven member (11) and a wedge (31) with a wedge surface adapted to rest against the ring (30), whereby at length increase of the elongated body, the wedge (31) is adapted to first press the ring (30) against the driven member (11) and then to displace the driven member (11) in the advance direction thereof. 3. A linearly operating motor for stepwise advance of a driven member (11), such as a shaft, comprising an elongated body (10) with length variable properties, and means for changing the length of said body (10), said driven member (11) being provided with at least a first locking means (12), said first locking means (12), upon length increase of said body (10), being adapted to be brought into engagement with said driven member (11) by means of said length increase and said body (10) being adapted to act upon the first locking means (12) in such a manner that said locking means and thereby also said driven member (11) are displaced at length increase of the body (10), said locking means comprising a first pivotally arranged element (24) and a friction element (25) connected thereto and intended for engagement with the driven member (11), whereby the elongated body (10) at its length increase is arranged to swing the first element (24) in a direction towards said driven member (11) by first forcing the friction element (25) to abutment against the driven member (11) and then advancing the driven member (11). 4. A motor as claimed in claim 1, wherein the locking means 12 is displaced against the action of a resilient element (22), which is adapted to apply a pretensioning force upon the elongated body (10).
047599014	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a nuclear reactor installation arranged in the cavity of a pressure vessel, with a nuclear reactor, the core, surrounded by a thermal side shield, which is being traversed from top to bottom by a cooling gas, with a plurality of main loops, each containing within the pressure vessel a heat exchanger and a blower, together with two gas conduits to connect the said components with the reactor core, and with a plurality of auxiliary loops for the removal of decay heat, wherein each auxiliary loop is connected by means of two gas conduits with the reactor core. In nuclear reactor installations, rapid shutdowns (insertion of absorber rods) are followed in the core by the development of decay heat, which initially amounts to approximately 4% of the nominal thermal capacity. This decay heat development declines rapidly at first and then slower and in its later variation approaches a value of zero asymptotically. In order to prevent an excessive rise in temperature in the reactor core, the decay heat must be removed from the core. 2. Description of the Prior Art It is part of the state of the art to equip nuclear reactor installations of higher capacities with auxiliary loops for the removal of decay heat, each of which contains auxiliary heat exchangers and auxiliary blowers. The auxiliary heat exchangers are connected by means of gas conduits with the hot gas and cold gas collector spaces of the reactor. During normal operation, the auxiliary heat exchangers are closed, for example by gravity actuated butterfly valves on the auxiliary blowers, which causes a slight backflow of cold gas over the auxiliary heat exchangers. No hot gas can therefore pass by free convection in the auxiliary heat exchangers. In high temperature reactors at least two auxiliary loops are provided for the removal of decay heat; they may be operated independently of each other and independently of the main loops. Because these systems are important from the standpoint of safety engineering, there are very high requirements concerning their operating availability and safety. In German Offenlegungsschrift No. 31 41 734 and 32 26 300, and in German P No. 33 44 527.3 high temperature reactors equipped with such auxiliary loops are described. Every installation for the removal of decay heat comprises: (1) an intermediate cooling water loop, wherein by means of a circulating pump, water is circulated through the auxiliary heat exchangers; (2) an intermediate cooler, in which the water transfers its heat to a service cooling water loop; and (3) the auxiliary loops with the auxiliary heat exchangers and the auxiliary blowers for the cooling gas. In the intermediate cooling water loop, a further circulating pump and a cooling tower are arranged. The intermediate cooler, the circulating pumps and the service cooling water loops are located outside the reactor pressure vessel. One disadvantage of the known decay heat removal installations is that it is composed of several active components, such as pumps. A second disadvantage is if a tube fractures in the auxiliary water exchangers, which must be anticipated, a large volume of water (up to several m.sup.3) may enter the reactor core. Furthermore, the auxiliary heat exchanger involved in no longer available for the removal of decay heat because it must be shut off as soon as possible. SUMMARY OF THE INVENTION In view of the above described state of the art, it is the object of the invention to develop a nuclear reactor installation of the above described structural type so that the disadvantages of the known decay heat removal facilities may be avoided. According to the invention, this object is attained in that each auxiliary loop comprises a bundle of independent, parallel heat pipes, together with a cooling gas blower, with the heat absorbing part of the heat pipes being arranged in an interruptable flow of cooling, and by an external cooling water system operated by means of a circulating pump which is provided as a heat sink for each bundle of heat pipes, with a cooling tower being located in a known manner in each loop. Compared to the known nuclear reactor installations, the plant according to the invention has the advantage that the installations for the removal of decay heat have a high degree of availability. First, if a tube fractures, which must be anticipated, only a limited, small amount of water may enter the reactor core. Thus, an additional expensive leakage monitoring system becomes superflous. Second, if there is a leak in a bundle of heat pipes, this would affect only one heat pipe and therefore, the removal of the decay heat is not affected because a bundle consisting of independent individual heat pipes provides a high degree of redundancy. In contrast, in case of a leak in a heat exchanger, the entire auxiliary loop is affected. The high availability of the decay heat removing installations is further enhanced by the fact that they are capable of operating with few active components, such as circulating pumps, since one cooling water loop per installation has been eliminated. The heat transmitting part of the heat pipes of each bundle terminates preferably in a reservoir filled with water, located above the pressure vessel, to which the cooling water system involved is connected. According to an advantageous further development of the invention, the reservoirs are equipped with at least one evaporator line for the water and are sufficiently large enough to be capable of removing the decay heat by evaporation for a certain period of time in case of a failure of the external active cooling water loop. In an emergency therefore the removal of the decay heat is possible without the use of active components. If in the nuclear reactor installation, the heat exchangers of the main loops are located in an annular space formed by the wall of the cavity and the thermal side shield, it is advantageous to arrange the heat absorbing part of the heat pipes also in this annular space. The part of the heat pipes, which transports the heat absorbed to the heat transfer part, is installed in bundles in a vertical passage located in the roof of the pressure vessel.
	claims	1. A method comprising:generating a gaseous fission product with a nuclear fission fuel element disposed in a reactor vessel;receiving the gaseous fission product into a plenum defined by a valve body associated with the nuclear fission fuel element disposed in a reactor vessel; andcontrollably venting the gaseous fission product from the plenum by operating a valve in communication with the plenum by displacing the valve by moving a flexible diaphragm coupled to the valve. 2. The method of claim 1, further comprising:mounting a cap on the valve; andextending a manipulator to the cap for manipulating the cap. 3. The method of claim 1, further comprising extending a manipulator to the valve for manipulating the valve. 4. The method of claim 1, further comprising:extending an articulated manipulator arm to the plenum; andcarrying a receptacle on the articulated manipulator arm, the receptacle engageable with the plenum for receiving the gaseous fission product controllably vented from the plenum. 5. The method of claim 1, wherein controllably venting the gaseous fission product from the plenum by operating venting means in communication with the plenum comprises controllably venting the gaseous fission product from the plenum by operating a valve responsive to a parameter chosen from pressure in the plenum and a type of gaseous fission product in the plenum. 6. The method of claim 1, further comprising sensing a parameter with a sensor in operative communication with the plenum. 7. The method of claim 6, wherein the parameter includes a parameter chosen from pressure in the plenum, a type of gaseous fission product in the plenum, and a radioactive fission product in the plenum. 8. The method of claim 6, further comprising transmitting a signal from the sensor. 9. The method of claim 8, wherein transmitting a signal from the sensor includes transmitting an identification signal identifying the valve body. 10. The method of claim 1, further comprising receiving the gaseous fission product into a reservoir coupled to the venting means. 11. The method of claim 10, wherein receiving the gaseous fission product into a reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter. 12. The method of claim 11, wherein separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a semi-permeable membrane. 13. The method of claim 11, wherein separating a condensed phase fission product solid from the gaseous fission product by passing the gaseous fission product through a filter comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through an electrostatic collector. 14. The method of claim 11, wherein separating a condensed phase fission product solid from the gaseous fission product by passing the gaseous fission product through a filter reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a cold trap. 15. The method of claim 10,wherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir coupled to a reactor vessel; andwherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir capable of being decoupled from the reactor vessel for removing the gaseous fission product from the reactor vessel. 16. The method of claim 10,wherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir coupled to a reactor vessel; andwherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir capable of remaining coupled to the reactor vessel for storing the gaseous fission product at the reactor vessel. 17. The method of claim 1, further comprising operating a coolant system in operative communication with the venting means for receiving the gaseous fission product controllably vented by the venting means. 18. The method of claim 1, further comprising removing the gaseous fission product from the coolant system to a removal system in operative communication with the coolant system. 19. The method of claim 1, wherein operating venting means associated with the nuclear fission fuel element comprises operating a reclosable valve. 20. The method of claim 1, wherein operating venting means associated with the nuclear fission fuel element comprises operating a sealably reclosable valve. 21. The method of claim 1, further comprising controlling operation of the venting means by operating a controller coupled to the venting means. 22. A method of operating a nuclear fission reactor, comprising:generating a gaseous fission product with a plurality of nuclear fission fuel element bundles disposed in a reactor vessel;receiving the gaseous fission product into a plenum defined by at least one of a plurality of valve bodies associated with respective ones of the plurality of nuclear fission fuel element bundles disposed in the reactor vessel;controllably venting the gaseous fission product from the plenum by operating a valve disposed in the at least one of the plurality of valve bodies, the valve being in communication with the plenum;displacing the valve by allowing movement of a flexible diaphragm coupled to the valve; andthreadably mounting a cap on the valve. 23. The method of claim 22, further comprising activating a plurality of nuclear fission fuel element bundles associated with respective ones of the plurality of valve bodies, at least one of the plurality of nuclear fission fuel element bundles being capable of generating the gaseous fission product. 24. The method of claim 22, wherein controllably venting the gaseous fission product comprises allowing movement of a flexible diaphragm capable of displacing the valve to a closed position. 25. The method of claim 22, further comprising extending an articulated manipulator arm to the cap for threadably dismounting the cap from the valve. 26. The method of claim 22, further comprising extending an articulated manipulator arm to the valve for operating the valve. 27. The method of claim 22, further comprising:extending an articulated manipulator arm to the plenum; andcarrying a receptacle on the articulated manipulator arm, the receptacle being engageable with the plenum for receiving the gaseous fission product controllably vented from the plenum. 28. The method of claim 22, wherein operating a valve in the at least one of the plurality of valve bodies comprises operating a valve responsive to a parameter chosen from pressure in the plenum and a type of gaseous fission product in the plenum. 29. The method of claim 22, further comprising sensing a parameter with a sensor in operative communication with the plenum. 30. The method of claim 29, wherein the parameter includes a parameter chosen from pressure in the plenum, a type of gaseous fission product in the plenum, and a radioactive fission product in the plenum. 31. The method of claim 22, further comprising transmitting a signal from the sensor. 32. The method of claim 31, transmitting a signal from the sensor includes transmitting an identification signal identifying the valve body. 33. The method of claim 22, further comprising receiving the gaseous fission product into a reservoir coupled to the valve, the gaseous fission product being vented by the valve. 34. The method of claim 33, wherein receiving the gaseous fission product into the reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter disposed in the reservoir. 35. The method of claim 34, wherein separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter disposed in the reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a semi-permeable membrane. 36. The method of claim 34, wherein separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter disposed in the reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through an electrostatic collector. 37. The method of claim 34, wherein separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a filter disposed in the reservoir comprises separating a condensed phase fission product from the gaseous fission product by passing the gaseous fission product through a cold trap. 38. The method of claim 33,wherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir coupled to a reactor vessel; andwherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir capable of being decoupled from the reactor vessel for removing the gaseous fission product from the reactor vessel. 39. The method of claim 33,wherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir coupled to a reactor vessel; andwherein receiving the gaseous fission product into a reservoir comprises receiving the gaseous fission product into a reservoir capable of remaining coupled to the reactor vessel for storing the gaseous fission product at the reactor vessel. 40. The method of claim 22, further comprising operating a coolant system in operative communication with the valve for receiving the gaseous fission product controllably vented by the valve. 41. The method of claim 40, further comprising removing the gaseous fission product from the coolant system to a removal system in operative communication with the coolant system. 42. The method of claim 22, wherein controllably venting a gaseous fission product comprises operating a reclosable valve. 43. The method of claim 22, wherein controllably venting the gaseous fission product comprises operating a sealably reclosable valve. 44. The method of claim 22, wherein controllably venting the gaseous fission product comprises operating the valve to controllably vent the gaseous fission product according to a predetermined release rate for minimizing size of an associated gaseous fission product clean-up system. 45. The method of claim 22, further comprising controllably operating the valve by operating a controller coupled to the valve.
	description	This U.S. application claims priority under 35 U.S.C 371 to, and is a U.S. National Phase application of, the International Patent Application No. PCT/CN2014/075763, filed Apr. 21, 2014, which claims the benefit of prior Chinese Application No. 201410043818.1 filed Jan. 29, 2014. The entire contents of the before-mentioned patent applications are incorporated by reference as part of the disclosure of this U.S. application. Embodiments of the present disclosure generally relate to a detecting technology field, and more particularly, to a fuel ball detecting method and system with a self-diagnosis function. Secure and efficient nuclear reactor is one of the key technologies to solve the energy resource shortage problems. Since the pebble-bed high temperature gas cooled reactor has a high power generation efficiency and a good inherent safety, and the fuel can be loaded and unloaded without turning off the reactor, it is very popular in many countries. Presently, the principle of a fuel component detecting device applied in the pebble-bed reactor is detecting the fuel component based on effects of a graphite conductor on the inductive reactance of the coil, and the detecting device can be divided into the following categories according to different structures. The first category is the inboard detecting device in which the detecting coil is predisposed in the stainless steel fuel ball pipeline and when the fuel ball passes through the detecting coil, the inductive reactance of the coil changes, and a fuel ball signal can be obtained by detecting the inductive reactance changes of the coil. The second category is the detecting device installed via side wall drillings in which a hole is formed in the side wall of the stainless steel fuel ball pipeline and the detecting coil is disposed in the hole. The second category device is similar to the electromagnetic proximity switch. Since the installation of the above two devices both may influence the gas tightness of the pipeline, the sensor structure and the installation have to ensure a gas tightness of the ball pipeline under a high pressure, resulting in the structure complication and mounting difficulty of the sensor. When a fault occurs in the sensor, in order to maintain and replace the sensor, the fuel ball pipeline is needed to be dismantled, which has a long construction time and easily causes radiation pollution, thus influencing utilizability of the reactor. Furthermore, the detecting coil, the frames and other attachments contact with the radioactivity fuel ball directly, which affects a working life of the sensor, so the anti-radiation capability of the sensor material is required to be good. In addition, the high temperature gas cooled reactor has various electromagnetic interferences and heavy nuclear radiation, and the working environment of the reactor is so complex that it's difficult to check and maintain the devices. The conventional detecting devices have the following defects: the fuel ball detecting device has a low electromagnetic compatibility and is easy to be interfered by an electromagnetic environment in the high temperature gas cooled reactor, thus resulting in a miscount; when a fault occurs in the fuel ball detecting device, the detecting device cannot generate an alarm initiatively, such that a counting loss and a miscount may occur during the failure period. Embodiments of the present disclosure seek to solve at least one of the problems existing in the prior art to at least some extent, such as damages on a fuel ball pipeline, a poor anti-interference capability and no initiative alarms during a failure period. Accordingly, a first objective of the present disclosure is to provide a fuel ball detecting method with a self-diagnosis function. A second objective of the present disclosure is to provide a fuel ball detecting system with a self-diagnosis function. In order to achieve above objectives, according to embodiments of a first broad aspect of the present disclosure, a fuel ball detecting method with a self-diagnosis function is provided, including: exciting a first detecting coil and a second detecting coil of a fuel ball sensor disposed outside a pipeline by a sinusoidal alternating current, wherein the first detecting coil and the second detecting coil wind around the pipeline and are set upstream and downstream the pipeline respectively; obtaining a first voltage signal U1 from the first detecting coil and a second voltage signal U2 from the second detecting coil; processing the first voltage signal U1 and the second voltage signal U2 by differential amplification, band pass filtering, phase sensitive detection and low pass filtering by a signal processor so as to obtain a fuel ball waveform signal U0; determining whether the fuel ball passes the pipeline according to U0 by a single chip microcomputer; determining whether the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally; and outputting a result showing whether the fuel ball passes the pipeline, when the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally. According to embodiments of the present disclosure, the fuel ball detecting method with a self-diagnosis function has at least the following advantages: a self-diagnosis function is added; when a fault occurs in at least one of the first and the second detecting coils, the signal processor and the single chip microcomputer, an alarm can be generated for the fault to be removed in time, which avoids the counting loss and improves reliability of a fuel ball counting result. According to embodiments of a second broad aspect of the present disclosure, a fuel ball detecting system with a self-diagnosis function is provided, including: a fuel ball sensor disposed outside a pipeline and comprising a first detecting coil and a second detecting coil winding around the pipeline and set upstream and downstream the pipeline respectively; an exciting module configured to generate a sinusoidal alternating current exciting signal to excite the first detecting coil and the second detecting coil; a resonance bridge detecting circuit configured to obtain a first voltage signal U1 from the first detecting coil and a second voltage signal U2 from the second detecting coil; a signal processor connected with the resonance bridge detecting circuit and configured to process the first voltage signal U1 and the second voltage signal U2 by differential amplification, band pass filtering, phase sensitive detection and low pass filtering, so as to obtain a fuel ball waveform signal U0; a single chip microcomputer connected with the signal processor and configured to determine whether the fuel ball passes the pipeline according to the U0; a self-diagnosis module connected with the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer and configured to detect whether the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally; and an outputting module connected with the single chip microcomputer and the self-diagnosis module respectively and configured to output a result showing whether the fuel ball passes the pipeline, when the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally. According to embodiments of the present disclosure, the fuel ball detecting system with a self-diagnosis function has at least the following advantages: a self-diagnosis function is added, and when a fault occurs in at least one of the first and the second detecting coils, the signal processor and the single chip microcomputer, an alarm can be generated for the fault to be removed in time, which avoids the counting loss and improves reliability of a fuel ball counting result. Additional aspects and advantages of embodiments of present disclosure will be given in part in the following descriptions, become apparent in part from the following descriptions, or be learned from the practice of the embodiments of the present disclosure. Reference will be made in detail to embodiments of the present disclosure. The same or similar elements and the elements having same or similar functions are denoted by like reference numerals throughout the descriptions. The embodiments described herein with reference to drawings are explanatory, illustrative, and used to generally understand the present disclosure. The embodiments shall not be construed to limit the present disclosure. In the specification, unless specified or limited otherwise, relative terms such as “central”, “longitudinal”, “lateral”, “length”, “width”, “thickness”, “front”, “rear”, “left”, “right”, “lower”, “upper”, “horizontal”, “vertical”, “above”, “below”, “up”, “down”, “top”, “bottom”, “inner”, “outer”, “clockwise”, “anticlockwise” as well as derivative thereof (e.g., “horizontally”, “downwardly”, “upwardly”, etc.) should be construed to refer to the orientation as then described or as shown in the drawings under discussion. These relative terms are for convenience of description and do not require that the present disclosure be constructed or operated in a particular orientation. In addition, terms such as “first” and “second” are used herein for purposes of description and are not intended to indicate or imply relative importance or significance. Thus, features limited by “first” and “second” are intended to indicate or imply including one or more than one these features. In the description of the present disclosure, “a plurality of” relates to two or more than two. In the description of the present disclosure, unless specified or limited otherwise, it should be noted that, terms “mounted,” “connected” “coupled” and “fastened” may be understood broadly, such as permanent connection or detachable connection, electronic connection or mechanical connection, direct connection or indirect connection via intermediary, inner communication or interaction between two elements. Those having ordinary skills in the art should understand the specific meanings in the present disclosure according to specific situations. In the description of the present disclosure, a structure in which a first feature is “on” a second feature may include an embodiment in which the first feature directly contacts the second feature, and may also include an embodiment in which an additional feature is formed between the first feature and the second feature so that the first feature does not directly contact the second feature, unless otherwise specified. Furthermore, a first feature “on,” “above,” or “on top of” a second feature may include an embodiment in which the first feature is right “on,” “above,” or “on top of” the second feature, and may also include an embodiment in which the first feature is not right “on,” “above,” or “on top of” the second feature, or just means that the first feature has a sea level elevation larger than the sea level elevation of the second feature. While first feature “beneath,” “below,” or “on bottom of” a second feature may include an embodiment in which the first feature is right “beneath,” “below,” or “on bottom of” the second feature, and may also include an embodiment in which the first feature is not right “beneath,” “below,” or “on bottom of” the second feature, or just means that the first feature has a sea level elevation smaller than the sea level elevation of the second feature. Reference throughout this specification to “an embodiment,” “some embodiments,” “one embodiment”, “another example,” “an example,” “a specific example,” or “some examples,” means that a particular feature, structure, material, or characteristic described in connection with the embodiment or example is included in at least one embodiment or example of the present disclosure. Thus, the appearances of the phrases such as “in some embodiments,” “in one embodiment”, “in an embodiment”, “in another example,” “in an example,” “in a specific example,” or “in some examples,” in various places throughout this specification are not necessarily referring to the same embodiment or example of the present disclosure. Furthermore, the particular features, structures, materials, or characteristics may be combined in any suitable manner in one or more embodiments or examples. Any procedure or method described in the flow charts or described in any other way herein may be understood to comprise one or more modules, portions or parts for storing executable codes that realize particular logic functions or procedures. Moreover, advantageous embodiments of the present disclosure comprises other implementations in which the order of execution is different from that which is depicted or discussed, including executing functions in a substantially simultaneous manner or in an opposite order according to the related functions. This should be understood by those skilled in the art which embodiments of the present disclosure belong to. The logic and/or step described in other manners herein or shown in the flow chart, for example, a particular sequence table of executable instructions for realizing the logical function, may be specifically achieved in any computer readable medium to be used by the instruction execution system, device or equipment (such as the system based on computers, the system comprising processors or other systems capable of obtaining the instruction from the instruction execution system, device and equipment and executing the instruction), or to be used in combination with the instruction execution system, device and equipment. It is understood that each part of the present disclosure may be realized by the hardware, software, firmware or their combination. In the above embodiments, a plurality of steps or methods may be realized by the software or firmware stored in the memory and executed by the appropriate instruction execution system. For example, if it is realized by the hardware, likewise in another embodiment, the steps or methods may be realized by one or a combination of the following techniques known in the art: a discrete logic circuit having a logic gate circuit for realizing a logic function of a data signal, an application-specific integrated circuit having an appropriate combination logic gate circuit, a programmable gate array (PGA), a field programmable gate array (FPGA), etc. Those skilled in the art shall understand that all or parts of the steps in the above exemplifying method of the present disclosure may be achieved by commanding the related hardware with programs. The programs may be stored in a computer readable storage medium, and the programs comprise one or a combination of the steps in the method embodiments of the present disclosure when run on a computer. The storage medium mentioned above may be read-only memories, magnetic disks or CD, etc. In addition, each function cell of the embodiments of the present disclosure may be integrated in a processing module, or these cells may be separate physical existence, or two or more cells are integrated in a processing module. The integrated module may be realized in a form of hardware or in a form of software function modules. When the integrated module is realized in a form of software function module and is sold or used as a standalone product, the integrated module may be stored in a computer readable storage medium. The present disclosure seeks to provide a fuel ball detecting method with a self-diagnosis function and a system for detecting and counting a graphite-based fuel ball in a loading pipeline, an unloading pipeline and a spent fuel pipeline, which has a great anti-interference capability and a high reliability. When a fault occurs in key devices in the detecting system, the fault can be detected and an alarm can be generated so as to avoid a miscount of the fuel ball. The detecting method and system can work reliably and detect the fuel ball accurately in atrocious electromagnetic environments, such that loading and unloading the fuel ball in various environments can be controlled effectively, thus ensuring a safe operation of a reactor. As shown in FIG. 1, the fuel ball detecting method with a self-diagnosis function according to embodiments of the present disclosure includes the following steps. In step A, a first detecting coil and a second detecting coil of a fuel ball sensor are excited by a sinusoidal alternating current, in which the fuel ball sensor is disposed outside a pipeline, and the first detecting coil and the second detecting coil wind around the pipeline and are set upstream and downstream the pipeline respectively. The first and second detecting coils are disposed at different locations of the pipeline, so as to ensure a time difference between time of the fuel ball passing through the first detecting coil and that of the second detecting coil. The sinusoidal alternating current is recorded as UE. The sinusoidal alternating current may be based on direct digital frequency synthesis technology and an AD9850 chip is used to full-digitally synthesize the sinusoidal alternating current having a stable amplitude and frequency, and then the sinusoidal alternating current is amplified by a LM1875 power chip and further input into the first detecting coil and the second detecting coil. An outer structure of the fuel ball sensor is shown in FIG. 2 (a) and an inner structure of the fuel ball sensor is shown in FIG. 2 (b). The fuel ball sensor includes: a housing 10; a pair of semi-ring frames 20A and 20B, the first detecting coil 30 and the second detecting coil 40, an aviation plug or a feed-through filter 50 and an electromagnetic sealing gasket 60. The pair of semi-ring frames 20A and 20B disposed over the pipeline fitted with each other. The first detecting coil 30 and the second detecting coil 40 respectively comprise a pair of semi-ring coils 30A, 30B and 40A, 40B connected with each other and winding around the pair of semi-ring frames, in which a wire rod, a winding method and a coil turn of each coil are the same with each other. The aviation plug or the feed-through filter 50 is connected with an output terminal of the first detecting coil 30 and the second detecting coil 40, so as to improve the anti-interference of the system. The electromagnetic sealing gasket 60 fills in a gap between the semi-ring coils and contacts with the housing 10, thus achieving a good shielding. The detecting coils in the above fuel ball sensor is designed as two semi-ring coils fitted with each other, which satisfies an external installation requirement and facilitates installation and maintenance. The coil has a narrow and flat shape, which can ensure a detecting sensitivity. In one embodiment, the coils are winded with two wires respectively so as to ensure a resistance deviation and an inductance deviation between the first detecting coil and the second detecting coil are less than 1%. The pair of semi-ring frames may be made of polysulfone plastics and the wire of the coils may be a polyimide varnished wire, so as to satisfy operating requirements for particular environments such as withstanding high temperatures and anti-radiation. The housing 10 itself has a good electromagnet isolation performance. However, radiation still can interfere and further be coupled to the coil in the fuel ball sensor through the gap in the housing 10 formed during the installation, and thus a shielding process is required. Therefore, the electromagnetic sealing gasket 60 is adapted to fill in the gap between the semi-ring coils and contact with the housing 10, so as to achieve a complete shielding structure. In step B, a first voltage signal U1 is obtained from the first detecting coil and a second voltage signal U2 is obtained from the second detecting coil by a resonance bridge detecting circuit. FIG. 3 is a circuit diagram showing a resonance bridge detecting circuit according to an embodiment of the present disclosure. As shown in FIG. 3, a first resistor R1 represents a first bridge arm and has a terminal connected with an output terminal of the exciting module to receive the exciting signal UE; a second resistor R2 represents a second bridge arm and has a terminal connected with the output terminal of the exciting module to receive the exciting signal UE; the first detecting coil L1, a first capacitor C1 and a potentiometer VR1 connected in parallel with each other represent a third bridge arm, in which a first terminal of the third bridge arm is grounded and a second terminal of the third bridge arm is configured to output the first voltage signal U1; the second detecting coil L2, a second capacitor C2 and a third resistor R3 connected in parallel with each other represent a fourth bridge arm, in which a first terminal of the fourth bridge is grounded and a second terminal of the fourth bridge arm is configured to output the second voltage signal U2. In step C, the first voltage signal U1 and the second voltage signal U2 are processed by differential amplification, band pass filtering, phase sensitive detection and low pass filtering by a signal processor so as to obtain a fuel ball waveform signal U0. Specifically, the processor obtains the fuel ball waveform signal U0 according to a lock-in amplification principle, as shown in FIG. 4. The first voltage signal U1 and the second voltage signal U2 are processed by a differential amplification circuit, a band pass filtering circuit, a phase sensitive detection circuit and a low pass filtering circuit sequentially. The initial sinusoidal signal is processed by a phase shifted circuit and a phase shifted initial sinusoidal signal is obtained. In the phase sensitive detection circuit, the first voltage signal U1 and the second voltage signal U2 are multiplied by the phase shifted initial sinusoidal signal. After being processed by the low pass filtering circuit the fuel ball waveform signal U0 is outputted, which can indicate changes of the first voltage signal U1 and the second voltage signal U2. A waveform of the fuel ball waveform signal U0 is shown in FIG. 5. When no ball passes the pipeline, excited by the sinusoidal alternating current, the first detecting coil generates the voltage signal U1 and the second detecting coil generates the voltage signal U2 due to the mutual inductance between the first detecting coil and the second detecting coil, in which the voltage signals U1 and U2 have a same frequency, a same phase position and an equal amplitude, and the fuel ball waveform signal U0 is constant. When the fuel ball passes the pipeline, an inductive reactance of the first detecting coil and that of the second detecting coils change due to an eddy current effect. The fuel ball passes through the first detecting coil and the second detecting coil at different time, so the amplitudes and phases of the voltage signals U1 and U2 change, excited by the sinusoidal alternating current. The fuel ball waveform signal U0 obtained by the resonance bridge detecting circuit is an approximate sinusoidal alternating signal. In step D, it is determined whether the fuel ball passes the pipeline according to U0 by a single chip microcomputer. (1) It is determined whether the fuel ball waveform signal U0 is complete. Specifically, the fuel ball waveform signal U0 is processed by a dual threshold comparison to generate adjacent pulses, and then it is determined whether the adjacent pulses comprise a peak pulse UP1 and a valley pulse UP2; if yes, it is determined the fuel ball waveform signal U0 is complete; and if no, it is determined the fuel ball waveform signal U0 is incomplete. (2) It is determined whether the fuel ball waveform signal U0 is continuous. Specifically, it is determined whether a time difference between a falling edge of the peak pulse UP1 and a rising edge of the valley pulse UP2 of the fuel ball waveform signal U0 adjacent to each other is less than a peak pulse width or a valley pulse width; if yes, it is determined the fuel ball waveform signal U0 is continuous; and if no, it is determined the fuel ball waveform signal U0 is discontinuous. The peak pulse width is determined by a peak amplitude threshold UTh1 of the fuel ball waveform signal U0 and the valley pulse width is determined by a valley amplitude threshold UTh2 of the fuel ball waveform signal U0.UTh1=(a base value of U0 without the fuel ball+a peak value of U0 with the fuel ball)/2UTh2=(a base value of U0 without the fuel ball+a valley value of U0 with the fuel ball)/2 The base value of U0 without the fuel ball is a preset value. (3) It is determined whether the fuel ball waveform signal U0 is symmetrical. Specifically, it is determined whether waveform widths of the peak pulse UP1 and the valley pulse UP2 of U0 are similar to each other; if yes, it is determined the fuel ball waveform signal U0 is symmetrical; and if no, it is determined the fuel ball waveform signal U0 is dissymmetrical. For example, when a difference between the peak pulse width and the valley pulse width is less than or equal to 25%, it is determined the fuel ball waveform signal U0 is symmetrical. (4) It is determined the fuel ball passes the pipeline when the fuel ball waveform signal U0 is complete, continuous and symmetrical. In step E, it is determined whether the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally. (1) It is determined whether the first detecting coil and the second detecting coil work normally by detecting voltages across the first detecting coil and the second detecting coil, and a first high level signal is output, when the first detecting coil and the second detecting coil work normally. Specifically, when the first and the second detecting coils work normally, the voltages across the first and the second detecting coils are high due to resonance phenomenon; when a fault (no matter a short circuit or a broken circuit) occurs in the first detecting coil or the second detecting coil, the voltage across the first detecting coil or the second detecting coil decreases, and thus the first high level signal output from a resonance detecting circuit is turned into a low level, of which a principle is shown in FIG. 6. (2) It is determined whether the signal processor works normally via a photocoupler disposed on a power source terminal of the signal processor, and a second high level signal is output when the signal processor works normally. Specifically, the photocoupler is disposed on the power source terminal of the signal processor to detect the signal processor. When a fault occurs, a signal output from the photocoupler turns into a low level instead of a high level, of which a principle is shown in FIG. 7. (3) It is determined whether the single chip microcomputer works normally by detecting a square wave output from the single chip microcomputer, and a third high level signal is output when the single chip microcomputer works normally. (4) The first high level signal, the second high level signal and the third high level signal are processed by AND operation to get a self-diagnosis output signal. (5) It is determined that the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally when the self-diagnosis output signal is a high level signal. In step F, a result showing whether the fuel ball passes the pipeline is output, when the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work normally. In addition, in an embodiment of the present disclosure, the fuel ball detecting method with a self-diagnosis function further includes step G in which an alarm is output when the self-diagnosis output signal is a low level signal, which means that at least one of the first detecting coil, the second detecting coil, the signal processor and the single chip microcomputer work abnormally. According to embodiments of the present disclosure, the fuel ball detecting method with a self-diagnosis function has at least following advantages: semi-ring coils fitted with each other are used as the outboard sensor, which makes the structure of the sensor simple; it's not necessary to destroy the pipeline during installation and maintenance, thus ensuring the completeness and gas tightness of the pipeline under a high pressure and reducing the radiation pollution; by providing a reasonable exciting signal and processing the fuel ball waveform signal U0, a good signal to noise ratio can be obtained and the system gain can be reduced; the first and the second detecting coils are designed with an electromagnetic compatibility, which improves the anti-interference capability; a self-diagnosis function is added and when a fault occurs in at least one of the first and the second detecting coils, the signal processor and the single chip microcomputer, an alarm can be generated for the fault to be removed in time, which avoids the counting loss and improves the reliability of a fuel ball counting result. As shown in FIG. 8, according to embodiments of the present disclosure, a fuel ball detecting system with a self-diagnosis function is provided, which includes: a fuel ball sensor 100, an exciting module 200, a resonance bridge detecting circuit 300, a signal processor 400, a single chip microcomputer 500, a self-diagnosis module 600 and an outputting module 700. The fuel ball sensor 100 is disposed outside a pipeline and includes a first detecting coil 30 and a second detecting coil 40 winding around the pipeline and set upstream and downstream the pipeline respectively. The exciting module 20 is configured to generate a sinusoidal alternating current exciting signal to excite the first detecting coil 30 and the second detecting coil 40. The resonance bridge detecting circuit is configured to obtain a first voltage signal U1 from the first detecting coil 30 and a second voltage signal U2 from the second detecting coil 40. The signal processor 400 is connected with the resonance bridge detecting circuit 300 and is configured to process the first voltage signal U1 and the second voltage signal U2 by differential amplification, band pass filtering, phase sensitive detection and low pass filtering, so as to obtain a fuel ball waveform signal U0. The single chip microcomputer 500 is connected with the signal processor 400 and is configured to determine whether the fuel ball passes the pipeline according to the U0. The self-diagnosis module 600 is connected with the first detecting coil 30, the second detecting coil 40, the signal processor 400 and the single chip microcomputer 500 and is configured to detect whether the first detecting coil 30, the second detecting coil 40, the signal processor 400 and the single chip microcomputer 500 work normally. The outputting module 700 is connected with the single chip microcomputer 500 and the self-diagnosis module 600 respectively and is configured to output a result showing whether the fuel ball passes the pipeline, when the first detecting coil 30, the second detecting coil 40, the signal processor 400 and the single chip microcomputer 500 work normally. In addition, the outputting module 700 is further configured to output an alarm when the self-diagnosis output signal is a low level signal, which means that at least one of the first detecting coil 30, the second detecting coil 40, the signal processor 400 and the single chip microcomputer 500 work abnormally. The fuel ball sensor 100, the resonance bridge detecting circuit 200, the exciting module 300 and the signal processor 400 in embodiments of the present disclosure have been described above, which are omitted here. In an embodiment of the present disclosure, the single chip microcomputer 500 includes: a first determining module, a second determining module, a third determining module. Specifically, the first determining module is configured to process the fuel ball waveform signal U0 by dual threshold comparison to generate adjacent pulses and to determine whether the adjacent pulses comprise a peak pulse UP1 and a valley pulse UP2; if yes, the first determining module determines the fuel ball waveform signal U0 is complete; and if no, the first determining module determines the fuel ball waveform signal U0 is incomplete. The second determining module is configured to determine whether a time difference between a falling edge of the peak pulse UP1 and a rising edge of the valley pulse UP2 of the fuel ball waveform signal U0 adjacent to each other is less than a peak pulse width or a valley pulse width; if yes, the second determining module determines the fuel ball waveform signal U0 is continuous; and if no, the second determining module determines the fuel ball waveform signal U0 is discontinuous. The peak pulse width is determined by a peak amplitude threshold UTh1 of the fuel ball waveform signal U0 and the valley pulse width is determined by a valley amplitude threshold UTh2 of the fuel ball waveform signal U0, in which UTh1=(a base value of U0 without the fuel ball+a peak value of U0 with the fuel ball)/2, UTh2=(a base value of U0 without the fuel ball+a valley value of U0 with the fuel ball)/2, in which the base value of U0 without the fuel ball is a preset value. The third determining module is configured to determine whether waveform widths of the peak pulse UP1 and the valley pulse UP2 of U0 are similar to each other; if yes, the third determining module determines the fuel ball waveform signal U0 is symmetrical; if no, the third determining module determines the fuel ball waveform signal U0 is dissymmetrical. Furthermore, the single chip microcomputer 500 determines the fuel ball passes the pipeline when the fuel ball waveform signal U0 is complete, continuous and symmetrical. In an embodiment of the present disclosure, the self-diagnosis module 600 includes: a first detecting unit, a second detecting unit, a third detecting unit and an AND gate circuit. Specifically, the first detecting unit is configured to determine whether the first detecting coil 30 and the second detecting coil 40 work normally by detecting voltages across the first detecting coil 30 and the second detecting coil 40, and output a first high level signal, when the first detecting coil 30 and the second detecting coil 40 work normally. The second detecting unit is configured to determine whether the signal processor 400 works normally via a photocoupler disposed on a power source terminal of the signal processor 400, and output a second high level signal when the signal processor 400 works normally. The third detecting unit is configured to determine whether the single chip microcomputer works normally by detecting a square wave output from the single chip microcomputer, and output a third high level signal when the single chip microcomputer works normally. Furthermore, the AND gate circuit is configured to process the first high level signal, the second high level signal and the third high level signal by AND operation to get a self-diagnosis output signal, in which when the self-diagnosis output signal is a high level signal, the self-diagnosis module 600 determines that the first detecting coil 30, the second detecting coil 40, the signal processor 400 and the single chip microcomputer 500 work normally. In order to achieve a better anti-interference capability, the detecting system has an isolated electromagnetic shielding structure to reduce radiated interference effects to the largest extent. Specifically, the signal processor 400 and the single chip microcomputer 500 may be disposed in an aluminium alloy melded and assembled case and inputs and outputs of signals and power source are realized via the aviation plug or the feed-through filter. Shielded cables are used to connect the first detecting coil 30, the second detecting coil 40, the resonance bridge detecting circuit 200 and the signal processor 300, and shielding layers of the shielded cables conduct well with the housing of the fuel ball sensor 100 and the aluminium alloy case of the signal processor 400. Power source wires adopt a varistor, a Ni—Zn magnet ring, a Mn—Zn common mode choke and a power source filter to remove conduction disturbance of a high voltage, a high frequency common mode, a low frequency common mode and a differential mode. Signal wires use the Ni—Zn magnet ring and the Mn—Zn common mode choke to remove conduction disturbance of the high frequency common mode and the low frequency common mode. According to embodiments of the present disclosure, the fuel ball detecting system with a self-diagnosis function has at least following advantages: semi-ring coils fitted with each other are used as the outboard sensor, which makes the structure of the sensor simple; it's not necessary to destroy the pipeline during installation and maintenance, thus ensuring the completeness and gas tightness of the pipeline under a high pressure and reducing the radiation pollution; by providing a reasonable exciting signal and processing the fuel ball waveform signal U0, a good signal to noise ratio can be obtained and the system gain can be reduced; the first and the second detecting coils are designed with an electromagnetic compatibility, which improves the anti-interference capability; a self-diagnosis function is added and when a fault occurs in at least one of the first and the second detecting coils, the signal processor and the single chip microcomputer, an alarm can be generated for the fault to be removed in time, which avoids the counting loss and improves the reliability of a fuel ball counting result. Although explanatory embodiments have been shown and described, it would be appreciated by those skilled in the art that the above embodiments cannot be construed to limit the present disclosure, and changes, alternatives, and modifications can be made in the embodiments without departing from spirit, principles and scope of the present disclosure.
053612841	summary	BACKGROUND OF THE INVENTION 1. Technical Field This invention relates to a class of devices for creating a range of corrosive conditions simulating those found in crevices of metal surfaces contained within a corrosive medium. It has particular application to the conditions found in heat exchangers, especially at the surfaces of heat exchange tubes within the secondary side of pressurized water reactor nuclear steam generators. This invention additionally relates to methods of monitoring and predicting tube crevice corrosion, also having applications to the operation of heat exchangers. 2. Description of the Prior Art The thin-walled metal tubes that are generally used in heat exchangers often suffer highly stressed conditions during extended periods of use. Stresses caused by elevated temperatures and pressures in combination with a corrosive chemical environment can lead to failure of these metals. This process occurs frequently at the crevices formed at interfaces with supporting structures. Corrosive debris suspended in a heat exchanger fluid can accumulate in the crevices as a corrosive sludge. Where a heat exchanger contains a hazardous material, or where the heat exchanger is within a system that is difficult or costly to shut down for maintenance, it is especially important to be able to monitor corrosive degradation, or even better, to predict catastrophic failure before it happens. Environmental concerns and cost considerations are both important in pressurized water reactor (PWR) nuclear steam generators. Because corrosion of the heat exchanger tubes of PWR steam generators present special concerns related to this invention, it is worthwhile to outline some of the details of these systems. Nuclear powered steam generators have three principle parts. A primary side contains radioactive hot water heated by the nuclear core. A secondary side contains non-radioactive water which, upon being converted into steam, powers the turbine generator. Heat is transferred from the primary side to the secondary side by a heat exchanger comprising a tube sheet in which the inlet and outlet ends of a plurality of U-shaped tubes are mounted. In PWR's, water from the primary side enters the U-shaped tubes through the inlets, flows through the U-shaped tubes and exits the outlets which are hydraulically isolated from the inlets by a divider sheet. A second hydraulic flowpath circulates non-radioactive water around the outside surfaces of the U-shaped tubes extending into the secondary side. Heat from the primary side is transferred across the metal boundaries of the U-shaped tubes to the secondary side. The U-shaped heat exchange tubes of nuclear powered steam generators are subjected to conditions which can lead to corrosion and failure of the tubes. Corrosion in the crevice regions of the heat exchanger is especially troublesome. Crevices are formed in the annular space between the heat exchange tubes and the tube sheet and also in the annular clearance between the tubes and the support plates in the secondary side. The support plates are used to uniformly space and align these tubes which otherwise would be buffeted about by the strong hydraulic flow around them. A sludge formed of particulates in suspension in the secondary side tends to collect in these crevices. These sludges are comprised predominantly of iron oxides and copper. The normal hydraulic circulation of the water in the secondary side is not sufficient to flush out the sludge from these crevices. In fact, the poor hydraulic circulation in these regions exacerbates the situation. The collection of sludge in the crevices impedes heat transfer from these regions, creating localized areas of elevated temperature (or "hot spots") in the tubes adjacent the sludge. The elevated temperatures in the "hot spots" allow higher local concentrations of corrosive impurities in solution, accelerating the corrosion process. Nuclear power plant operators periodically attempt to sweep the sludge out of the generator vessel by hydraulic means, however this is not always effective. The U-shaped heat exchange tubes of nuclear steam generators are typically formed from corrosion resistant nickel alloys, such as Inconel.RTM. 600, but corrosion in the crevices due to the elevated heat conditions and the high pressures found within the U-shaped tubes may ultimately penetrate the tube wall, resulting in leakage of radioactive water from the primary side into the non-radioactive water in the secondary side of the steam generator. Remedial action taken during maintenance shutdowns of the reactor can prevent such leakage, however it would be very useful to be able to know when corrosion is occurring before there is significant propagation. While early forms of steam generator corrosion (thinning and denting) could be readily related to a particular operating chemistry, more recent forms of corrosion (intergranular corrosion and cold leg thinning) usually cannot be related to a direct cause. As impurity levels in steam generators have been reduced, it has become progressively more difficult to correlate contaminant levels, as measured either in the steam generator blowdown during power operation or by monitoring the hideout return following a shutdown, with the severity of corrosion, as measured by the time period required to initiate corrosion, its subsequent propagation rate, and the number of affected tubes. Because of the inability of using the operating chemistry as a means of judging the possible future extent of corrosion, a more direct means of relating a particular operating chemistry to the future occurrence of corrosion is highly desirable. With such a capability, it would be possible to undertake a corrective action well in advance of the corrosion actually occurring. For this reason, model steam generators were developed to monitor the corrosion occurring in the heat exchange tubes of particular steam generators so that corrective action could be taken before failure of the tube walls. Model steam generators are bulky complicated apparatuses that attempt to mimic the actual steam generators. They contain all the hydraulic elements of an actual steam generator including circulating primary water, circulating secondary water and heat exchange tubes. They operate by subjecting an array of sample heat exchange tubes to a set of heat, pressure and chemical conditions approximating that which surrounds the heat exchange tubes in such nuclear steam generators. Their accuracy depends upon the accuracy of the simulation of the conditions within the operating steam generator. One type of model steam generator is described in U.S. Pat. Nos. 4,628,870, 4,635,589 and 4,640,233, all which were assigned to the Westinghouse Electric Corporation. A more compact model steam generator was disclosed in U.S. Pat. No. 4,637,346, also assigned to the Westinghouse Electric Corporation. Model steam generators, while being useful instruments for simulating conditions in an actual steam generator, do not subject the tubes in the model steam generators to exactly the same conditions as found in an operating generator. Although the same secondary side feedwater source as used in a PWR generator may be fed into the model generator's secondary side, this does not guarantee the same chemistry as that found within the power plant's secondary side. An additional defect they suffer is that it is difficult to change the primary tubes in the model steam generators, for example, to do inspection or destructive testing of the heat exchange tubes. SUMMARY OF THE INVENTION There exists a need for a simpler system which can monitor the actual corrosive degradation of heat exchanger tubes. There is also a need for a system that can accurately monitor corrosion. Accordingly, it is an object of the present invention to provide an apparatus for creating a range of corrosive conditions as found in fluid heat exchanger tubing. It is another object of this invention to provide an apparatus that is capable of monitoring tube corrosion within an operating heat exchanger. A further object of this invention is to provide an apparatus for predicting tube corrosion within an operating heat exchanger. It is an object of this invention to provide an apparatus for monitoring and predicting heat exchanger tube corrosion within operating PWR steam generators without interfering with the operation of the reactor. It is an object of this invention to provide a simplified, light-weight heat exchanger tube corrosion monitor and predictor. It is an object of this invention to provide a method for identifying and characterizing steam generator tube support plate crevice corrosion before it occurs in active heat transfer tubing without removing an active tube from the steam generator. These objects and others that will become more readily apparent are obtained in accordance with this invention. The present invention is a device for creating a corrosive condition on the exterior surface of an elongated sealed metal tube, such as those used in heat exchangers, and more particularly those used in PWR's. The device has means, such as a gas inlet port, for pressurizing the interior of the tube, and also has a heating means for providing heat to a predetermined position within the tube. Disposed exterior to the tube is a structure for retaining a corrosive medium in contact with the exterior of the tube proximate to the predetermined position. In one version of this device, the structure for retaining a corrosive medium in contact with the exterior surface of the tube is provided by a member simulating a tube support plate in a heat exchanger. The member, positioned approximately adjacent to the exterior surface of the tube, provides a crevice for trapping a corrosive sludge as found in the heat exchangers of PWR's. In another version, it is provided by a porous, consolidated material. A water-based solution having corrosive components can interpenetrate the porous material, simulating a sludge in a crevice. In either case, the purpose is to simulate the conditions in actual heat exchangers, wherein the heat exchange tubes are supported and held in place by structures that allow the collection of sludges in crevices at the surfaces of the tubes. Another aspect of this invention is that the device indicates corrosion taking place in an operating heat exchanger, thereby allowing an operator to take remedial measures before a leak occurs. This function is especially important for the heat exchangers of PWR's, where even a small leak of radioactive water (about 20 9 pm) from the primary side to the secondary side can cause the shutdown of the power plant. The device is adapted to be inserted into an active fluid heat exchanger, close to the heat exchange tubes or in an adjacent vessel. A measuring probe for providing a first signal indicative of a corrosion condition of the exterior surface of the tube proximate to the structure for retaining a corrosive medium allows in situ monitoring of the progress of tube corrosion conditions. The invention includes a method for creating corrosion of surfaces and monitoring that corrosion. The method allows for the creation of conditions similar to those found at the surfaces of U-tubes in heat exchangers having primary and secondary sides. The method uses an apparatus comprising an elongated sealed tube having an exterior surface, a heating source for providing heat to the tube at a first axial position within the tube, structure for retaining a corrosive medium, such as the sludge deposits found in the crevices between the U-tubes and support plates, in contact with the exterior surface proximate to the first axial position, means for pressurizing the interior of the tube, and a measuring probe for providing a first signal indicative of a corrosion condition of the exterior surface proximate to the first axial position. The method includes the steps of: first placing a portion of the sealed tube comprising the structure for retaining a corrosive medium in contact with a corrosive medium, such as the secondary side of a heat exchanger in an active PWR; pressurizing the interior of the tube to a first pressure; heating the exterior surface of the tube proximate to the first axial position to a first temperature with the heating source; and then providing the first signal indicative of a corrosion condition with the measuring probe after pressurizing and heating the tube. According to another aspect of this invention, a method is provided for creating corrosion of tube surfaces that is appropriate for monitoring heat exchangers. The method uses an apparatus comprising an elongated sealed tube, means within the tube for providing heat to the tube at a first axial position and for pressurizing the interior of the tube, and structure for retaining a corrosive medium in contact with the exterior surface of the tube proximate to the first axial position. First, a portion of the tube, comprising the structure for retaining a corrosive medium, is placed in contact with a corrosive medium. The interior of the tube at the first axial position is then heated and pressurized. Corrosion is enhanced by increasing the temperature and pressure. After a predetermined period of time, the temperature and pressure are reduced and the apparatus is removed from the corrosive medium. At this point, the tube can be examined for corrosion by nondestructive and/or destructive testing methods. Either method is appropriate for use in active heat exchangers, adjacent pressure vessels or for laboratory testing. By adjusting the internal pressure of the tube to at least about the pressure experienced within active heat exchanger tubing, corrosion is accelerated. Corrosion is also accelerated by bringing the local temperature of the exterior surface in contact with the corrosive medium to a temperature at least about the temperature of an active heat exchanger tube. The corrosive medium may be provided by a bench-top setup wherein the chemistry of the medium can be adjusted by an operator. It may also be provided by an active heat exchanger, such as the secondary side of an active PWR steam generator. Using the first method, an operator monitoring the signal indicative of a corrosion condition can take corrective action well in advance of the occurrence of significant corrosion of actual heat exchanger tubing. The second method involves simpler equipment and can be used in conjunction with regular servicing shutdowns of PWR's.
047160101	claims	1. Support apparatus for supporting and selectively locating a tool with respect to its work piece, and the work piece having first and second dimensions, said support apparatus comprising: (a) a tool carriage for receiving the tool; (b) a support structure for supporting and guiding said tool carriage with respect to the workpiece; (c) first drive means mounted on said support structure and coupled to said tool carriage for selectively driving said tool carriage along said first dimension; (d) rail means mounted on said tool carriage along said second dimension; (e) a support table mounted for movement along said rail means; (f) second drive means mounted on said tool carriage selectively driving the tool along said second dimension, said second drive means mounted on said support table and engaging said tool carriage to drive said support table along said rail means, whereby the tool may be accurately positioned with respect to the work piece; and (g) clamp means affixed to said support structure for engaging the work piece for suspending said support structure, whereby said support structure extends along said first dimension of said work piece. (a) a tool carriage for receiving the tool; (b) a strongback assembly for supporting and guiding said tool carriage with respect to the core barrel; (c) clamp means affixed to said strongback assembly for engaging the core barrel for suspending said strongback assembly along a first dimension substantially parallel to the axis of the core barrel; (d) first drive means mounted on said strongback assembly and coupled to said tool carriage for selectively driving said tool carriage along said first dimension; (e) a support table; (f) means mounted on said tool carriage for supporting said support table for movement along an arcuate path corresponding to said cylindrical configuration; and (g) second drive means on said tool carriage and coupled to said support table for selectively driving said support table along said arcuate path, whereby the tool is accurately positioned with respect to the core barrel. 2. Support apparatus as claimed in claim 1, wherein said support structure includes an indicator for aligning said support structure with respect to the workpiece. 3. Support apparatus as claimed in claim 1, wherein the workpiece includes a third dimension and there is further included third drive means mounted on said tool carriage for selectively driving the tool along aid third dimension with respect to the workpiece. 4. Support apparatus as claimed in claim 1, wherein the workpiece includes a third dimension and there is further included third drive means mounted on said support table for driving the tool along said third dimension with respect to the workpiece. 5. Support apparatus as claimed in claim 4, wherein there is included a tool mounting arm affixed to said third drive means to be driven thereby along said third dimension. 6. Support apparatus as claimed in claim 5, wherein said tool receiving arm includes means for releasably grasping the tool. 7. Support apparatus as claimed in claim 4, wherein the workpiece includes at least one work surface and there is further included sensing means for sensing the location of the work surface and fourth drive means mounted on said support table for disposing said sensing means along said third dimension with respect to the workpiece and its work surface. 8. Support apparatus as claimed in claim 7, wherein there is further included a sensing means arm coupled to said fourth drive means for carrying said sensing means. 9. Support apparatus as claimed in claim 1, wherein the workpiece includes at least one work surface and a pointer disposed thereon in a fixed relation with the work surface, and there is further included indicator means mounted upon said tool carriage to permit alignment of said support apparatus with respect to the work surface. 10. Support apparatus as claimed in claim 9, wherein said second drive means is mounted on said tool carriage in a fixed relationship with respect to said indicator means for moving the tool with respect to said indicator means as aligned with the pointer to align the tool with respect to the work surface. 11. Support apparatus as claimed in claim 10, wherein there is included encoder means mounted on said tool carriage and coupled to the tool for providing an output signal indicative of the relative movement of the tool with respect to said indicator means. 12. Support apparatus as claimed in claim 4, wherein there is included first encoder means mounted on said tool carriage and coupled to said support table for providing an output signal indicative of the relative movement between said support table and said tool carriage along said second dimension, and second encoder means mounted on said support table and coupled to the tool for providing an output signal indicative of the movement imparted to the tool by said third drive means along said third dimension. 13. In a nuclear reactor having a core barrel of a substantially cylindrical configuration with an axis and a plurality of flow holes disposed therethrough, a remotely controlled apparatus for supporting and selectively disposing a tool with respect to the core barrel, said remotely controlled apparatus comprising: 14. The remotely controlled apparatus as claimed in claim 13, wherein there is further included third drive means mounted on said support table and coupled to said tool for selectively disposing said tool along a second dimension aligned with respect to a radius of said cylindrical configuration, whereby said tool may be engaged with and disengaged from said core barrel. 15. The remotely controlled apparatus as claimed in claim 14, wherein there is included a tool mounting arm affixed to said third drive means to be driven thereby along said second dimension. 16. The remotely controlled apparatus as claimed in claim 15, wherein said tool mounting arm includes means for releasably grasping a plug to be selectively inserted within one of said plurality of flow holes. 17. The remotely controlled apparatus as claimed in claim 16, wherein the workpiece includes at least one work surface and there is further included sensing means for sensing the location of one of said plurality of flow holes and fourth drive means mounted on said support table for disposing said sensing means along said second dimension with respect to said core barrel. 18. The remotely controlled apparatus as claimed in claim 17, wherein there is further included a sensing means arm coupled to said fourth drive means for carrying said sensing means. 19. The remotely controlled apparatus as claimed in claim 18, wherein said core barrel includes at least one pointer disposed thereon in a fixed relation with corresponding of said plurality of flow holes, and there is further included indicator means mounted upon said tool carriage to permit alignment of said remotely controlled apparatus with respect to said corresponding of said plurality of flow holes. 20. The remotely controlled apparatus as claimed in claim 19, wherein said second drive means is mounted on said tool carriage in a fixed relationship with respect to said indicator means for moving said tool with respect to said indicator means as aligned with said pointer to align said tool with respect to one of said plurality of flow holes. 21. The remotely controlled apparatus as claimed in claim 20, wherein there is included encoder means mounted on said tool carriage and coupled to said tool for providing an output signal indicative of the relative position of said tool with respect to said indicator means. 22. The remotely controlled apparatus as claimed in claim 20, wherein there is included first encoder means mounted on said tool carriage and coupled to said support table for providing an output signal indicative of the relative position of said support table with respect to said indicator means, and second encoder means mounted on said support table and coupled to said tool for providing an output signal indicative of the movement imparted to said tool by said third drive means along said second dimension.
046997571	summary	BACKGROUND OF THE INVENTION The invention relates to nuclear fuel rods for use in nuclear reactors and comprising a gas tight sheath containing a stack of fuel pellets and means for holding the stack against an end plug of the sheath. The term "fuel" should be interpreted widely as designating not only materials used for producing energy by fission, but also fertile or neutron absorbing material. The fuel rods used up to present, particularly in reactors cooled by circulation of water or molten sodium, comprise a stack of oxide fuel pellets and holding means formed by a helical spring between the stack and an end plug. This solution has considerable drawbacks. The spring requires space and reduces the volume available for receiving the fission gases released by the fuel. Under irradiation, the material forming the spring is subject to relaxation which causes them to lose a fraction of their initial properties. Often, degradation during the operation in a reactor is such that the holding means no longer provide sufficient retention during fuel handling and transportation. Attempts have been made to solve this problem. French Pat. No. 2,529,371 describes a fuel rod whose holding spring is made from a material which exerts a retaining force which decreases when the temperature increases. The only advantage of this solution is that it reduces the stresses undergone by the fuel pellets during operation in the reactor. French Pat. No. 2,018,665 describes a rod in which the spring in replaced by a member in the form of a split bucket resiliently engaging the sheath. Such a member has drawbacks: it exerts a constant retention force and remains in position when the stack of pellets contracts, so that it no longer holds the stack in position during shipping. SUMMARY OF THE INVENTION It is an object of the invention to provide a nuclear fuel rod with improved holding means. It is a more specific object to provide retaining means which require less space and need not to consist of material having high resiliency for ensuring a substantially constant holding force whatever the swelling of the pellets under irradiation. For that, it uses the fact that it is essential to hold the stack only during transport and handling of the rods, since the rods are in a vertical position during operation in a reactor and the stack of pellets then rests quite naturally on the lower plug of the sheath. According to the invention, there is provided a rod of the above-defined type in which the holding means comprise a radially expansible element, of a shape at rest such that, when in the sheath and at the atmospheric temperature, it frictionally engages the inner surface of the sheath, and further comprise means responsive to the temperature for forcibly contracting said element and reducing its friction against the sheath when the temperature exceeds a predetermined value lower than the normal operating temperature in a reactor. The means responsive to the temperature are formed from a shape memory alloy having a transformation temperature intermediate between atmospheric temperature and the normal operating temperature in a reactor. Shape memory alloys date back to twenty years or so. Below a transition temperature, they can be deformed plastically. When heated above the transition temperature, they resume the shape which was given them during a metallurgical shaping treatment. Numerous alloys are known fulfilling that condition, having different transition temperatures. A description of such alloys may be found in "Some Applications of Shape Memory Alloys", Journal of Metals, June 1980, pp. 129-137. Among the alloys which may be used for implementing the invention, titanium-nickel alloys may be of interest. They transform from the austenitic phase to the martensitic phase and inversely at a temperature less than 400.degree. C., i.e. at a value lower than that reached in the plenum chamber of the fuel rods, at least in existing pressurized water reactors. Examples of such titaniumnickel alloys may be found in British Pat. No. 2,117,001. The stack of pellets is likely to shorten when the rod cools down and the temperature is below the transition temperature. Resilient means may advantageously be inserted between the radially expandable element and the end pellet of the stack, so as to accomodate possible variations in length at that time. The resilient means may comprise a washer in abutment against the stack, connected to the radially expandable element by a compression spring and by a second spring made from a shape memory alloy, having a transition temperature higher than the atmospheric temperature range and lower than that of the first spring, the second spring exerting, above its transition temperature, a retraction force greater than the force of the first one. The holding means of the invention use the fact that their function is no longer required during operation of the reactor. On the other hand, during cooling down of the reactor before fuel unloading, the reversibility of the shape memory of the alloy allows the radially expandable element to frictionally engage the sheath and to lock on the sheath. During reactor operation, since the expandable element is substantially free, it may follow the movements of the end pellet of the stack, due for example to swelling under irradiation. Welding of the end plug during manufacture is rendered easier since there is no spring in abutting relation with the plug.
	description	The present invention claims priority to and is related to U.S. patent applications Ser. No. 10/009,996 filed on Dec. 14, 2001, now U.S. Pat. No. 6,715,201 B1 which is the national stage of international application PCT/JP01/03246, filed on Apr. 16, 2001. The entire contents of each of these applications are incorporated herein by reference. The present invention relates to a repairing system for the emergency-repairing of a reactor vessel and in-pile structures in, for example, a nuclear power plant or the like, and a repairing method. Stuctures of a light-water reactor, such as a boiling-water reactor, are formed of materials having a sufficient corrosion resistance and high-temperature strength in an environment of high temperatures, such as austenitic stainless steels or nickel-base alloys. However, there is apprehension about the quality degradation of the materials of the members difficult to change of the structures due to exposure to a severe environment during the long-term operation of the plant or the detrimental irradiation with neutrons. Particularly, weld zones are subject to the potential danger of stress-corrosion cracking due to the sensitization of the materials by weld heat input and residual tensile stress. A shroud supporting fuel assemblies, among the structures, is particularly subject to the influence of neutrons produced by the fuel assemblies and is highly subject to stress-corrosion cracking. Various working systems for the inspection for soundness and preventive maintenance of structures highly subject to damaging danger have been invented and practically applied. However, since those working systems are intended to carry out work efficiently for entire weld lines, the working systems are large, need much time for preparatory work, are large in scale and complicated and need well-trained operators for operation. Thus, the conventional working systems are unable to take such steps as occasion demands and to meet the demands of occasion. The present invention has been made in view of those problems and it is therefore an object of the present invention to provide a partial-repairing system capable of dealing with various tasks somewhat efficiently, through, and of being inserted and installed in a reactor instantly, and highly maneuverable, and a repairing method. A modular submersible repairing system according to the present invention includes a working unit; and a base unit; wherein the working unit includes: at least one type of tool module repairing structures in a reactor, a scanning/pitching module being selectively connected to or disconnected from the tool module, and provided with a scanning/pitching shaft for scanning or pitching the tool module, a submersible fan module being selectively connected to or disconnected from the scanning/pitching module, and a first buoyant module for keeping an orientation of the tool module; the base unit includes: a manipulator module internally provided with an actuator driving mechanism, a adsorbing module being detachably mounted on the manipulator module and of adsorbing to a wall, and a second buoyant module for keeping an orientation of the manipulator module; each of at least the scanning/pitching module and the manipulator module is provided with a submersible connecting device being operated in water for engagement and disengagement; configuration and functions of the modular submersible repairing system can be changed or modified according to various purposes of work in the reactor by properly combining those modules; and the modules can be connected together in the reactor by remotely operating the submersible connecting devices. Preferred embodiments of the present invention will be described with reference to the accompanying drawings. FIG. 1 is a schematic perspective view of a modular submersible repairing system in a preferred embodiment according to the present invention. The modular submersible repairing system has a working unit including one of various types of tool modules 1 capable of repairing structures, a scanning/pitching module 2 capable of selectively scanning and positioning the tool module 1, a submersible fan module 3 capable of being selectively connected to or disconnected from the scanning/pitching module 2, and a buoyant module 4 (first buoyant module); and a base unit including a manipulator module 5, a adsorbing module 6 capable of being selectively connected to or disconnected from the manipulator module 5 and provided with suction cups 6a, a submersible fan 7, and a buoyant module 8 (second buoyant module). The tool modules 1 are used selectively according to the purpose of work. The scanning/pitching module 2 is provided with a scanning/pitching mechanism 2a for moving and scanning the tool module 1 mounted on the scanning/pitching module 2. The submersible fan modules 3 and 7 are provided with submersible fans 3a and 7a, respectively. The submersible fans 3a and 7a generate thrusts to press the submersible fan modules 3 and 7 against a wall, respectively. The buoyancies of the buoyant modules 4 and 8 are keeping it's orientation stably. The manipulator module 5 is provided with a pantographic extension mechanism 9. The scanning/pitching module 2 can be detachably joined to the free end of the extension mechanism 9. FIGS. 2(a) and 2(b) area side view and an elevational view, respectively, of the extension mechanism 9. Internally threaded nuts 10a and 10b attached to the upper and the lower base end, respectively, of the pantographic linkage of the extension mechanism 9 are screwed on a threaded shaft 11. The threaded shaft 11 has an upper threaded section 11a and a lower threaded section 11b provided with threads of the opposite hands, respectively. The upper nut 10a and the lower nut 10b are screwed on the upper threaded section 11a and the lower threaded section 11b, respectively. The threaded shaft 11 is interlocked through a bevel gear mechanism 12 to the drive shaft of a driving motor 13. The joints of the pantographic linkage include bearings 14. The pantographic linkage is extendible. Since the joints of the pantographic linkage includes the bearings 14, the pantographic linkage is able to bend to some extent in a direction perpendicular to a reference plane. The other end of the pantographic linkage is connected to a connecting member 15 connecting the scanning/pitching module 2 and the extension mechanism 9 so as to be vertically slidable on the connecting member 15. The driving motor 13 drives the threaded shaft 11 for rotation. Consequently, the nuts 10a and 10b are moved toward or away from each other to extend or contract the pantographic mechanism horizontally. The modules are detachable from each other. Some modules including the scanning/pitching module 2 and the manipulator module 5 are provided with, for example, a submersible connecting device. The submersible connecting devices are remotely operated in water for connection or disconnection. At least the scanning/pitching module 2 and the manipulator module 5 are provided with submersible connecting devices, respectively. The tool module 1 may be provided with a submersible connecting device. FIG. 3 is a schematic view of the submersible connecting device. For example, the scanning/pitching module 2 is provided a male connecting unit 18 including a taper member 16 tapering toward its free end, and a draw-bolt 17 fastened to the extremity of the taper member 16. The male connecting unit 18 projects horizontally from the scanning/pitching module 2. A key groove 19 is formed in the base part of the taper member 16 of the male connecting unit 18. Dints 20 are formed in an upper part of the scanning/pitching module 2. The hooks of a hoisting device, not shown, engage the dints 20. On the other hand, the manipulator module 5, to which the scanning/pitching module 2 is connected, is provided with a female connecting unit 21. A taper hole 22 complementary to the taper member 16 of the male connecting unit 18 is formed in a part of the manipulator module 5 facing the scanning/pitching module 2. A key 23 to be engaged in the key groove 19, and an ultrasonic distance measuring device 24 for measuring the distance between the scanning/pitching module 2 and the manipulator module 5 in a noncontact measuring mode are disposed near the open end of the taper hole 22. The female connecting unit 21 is provided with a gripping mechanism 25 capable of gripping the draw-bolt 17 and of pulling the male connecting unit 18 toward the female connecting unit 21. The gripping mechanism 25 is operated by a hydraulic cylinder actuator 26. A pneumatic locking device 28 is connected to one end of the hydraulic cylinder actuator 26. When the male connecting unit 18 is pulled into the gripping mechanism 25, the pneumatic locking device 28 engages a piston rod 27 included in the hydraulic cylinder actuator 26 to restrain the piston rod 27 from movement. A recess 29 is formed in an upper part of the manipulator module 5. A drawing claw engages the recess 29. When connecting the scanning/pitching module 2 and the manipulator module 5, the taper part 16 of the male connecting unit 18 is inserted in the taper hole 22 of the female connecting unit 21, the stopping members 25a of the gripping mechanism 25 are engaged with the draw-bolt 17, and the hydraulic cylinder actuator 26 is operated to draw the draw-bolt 17 into the taper hole 22. After the scanning/pitching module 2 and the manipulator module 5 have been thus connected, the pneumatic piston having the locking device 28 holds the piston rod 27 of the hydraulic cylinder actuator 26 fixedly to prevent the accidental disengagement of the male connecting unit 18 and the female connecting unit 21 of the submersible connecting device resulting from the faulty operation of the hydraulic cylinder actuator 26 due to faulty operations or loss of pressure applied to the hydraulic cylinder actuator 26 during work. FIGS. 4(a) and 4(b) are a side view and an elevational view, respectively, of a hoisting device 30 for suspending the module or a combination of the modules in water and for connecting the module or a combination of the modules to an existing module. A pair of hooks 32 are supported on a lower end part of the hoisting device 30. The hooks 32 are turned about horizontal axes, respectively, by a pneumatic cylinder actuator 31 to engage the same with or disengage the same from the dints 20 of the module. The hoisting device 30 is provided with an arm 34 capable of being advanced toward and retracted away from the module to be connected to another module, for example, the manipulator module 5, by a pneumatic cylinder actuator 33. A drawing claw 35 and a pushing claw 36 are supported on the arm 34. The claws 35 and 36 are connected pivotally by pin joints 37 to a claw support member 38 held on the arm 34. The claws 35 and 36 hung down from the claw support member 38 by their own weights. The drawing claw 35 is able to turn away when the arm 34 is moved in a pushing direction and is restrained from turning by a stopper 39 when the arm 34 is moved in a drawing direction. The pushing claw 36 is able to turn away when the arm 34 moves in the drawing direction and is restrained from turning by the stopper 39 when the arm 34 is moved in the pushing direction. The claws 35 and 36 and the claw support member 38 are provided with holes 40, 41 and 42, respectively. A pin is inserted in the holes 40 and 42 to hold the drawing claw 35 in a horizontal position when the drawing claw 35 is not used. A pin is inserted in the holes 41 and 42 to hold the pushing claw 36 in a horizontal position when the pushing claw 36 is not used. When connecting the modules together in water contained in the reactor by a remotely controlled operation, the drawing claw 35 is set in a vertical position, the pushing claw 36 is set in a horizontal position, a hoisting hook driving mechanism including a linkage is operated by the pneumatic cylinder actuator 31 to engage the hooks 32 in the dints 20 of the module 2 provided with the male connecting unit 18, and the module 2 is lowered. The module 2 is moved in the reactor so that the male connecting unit 18 of the module 2 approaches the female unit 21 of the module 5, and hoisting wires are controlled so as to insert the taper part 16 in the taper hole 22 of the module 5. The taper part 16 is inserted in the taper hole 22 deep enough to enable the drawing claw 35 to engage in the recess 29 of the module 5 by a manual operation. Then, the pneumatic cylinder actuator 33 is actuated to move the arm 34 in the drawing direction. Consequently, the drawing claw 35 engaged in the recess 29 draws the female connecting unit 21 forcibly toward the male connecting unit 18. Thus, the gripping mechanism 25 is made to grip the draw-bolt 17 by a remotely controlled operation. When disconnecting the modules from each other in water contained in the reactor and taking out the module 2 from the reactor, the drawing claw 35 set in a horizontal position and the pushing claw 36 set in a vertical position are inserted in the reactor, and the hooks 32 are engaged in the dints 20 of the module 2. Then, the gripping mechanism 25 is operated to release the draw-bolt 17 to disconnect the male connecting unit 18 from the female connecting unit 21. Generally, the taper part 16 cannot be removed from the taper hole 22 at this stage. Therefore, the arm 34 is moved in the pushing direction to push the female connecting unit 21 from the male connecting unit 18. The modular submersible repairing system thus constructed carries out work for the maintenance of the shroud of a reactor in the following manner. The modules of the base unit and the working unit are assembled in a vertical arrangement as shown in FIG. 5 such that the base unit and the working unit have the smallest horizontal cross sections, respectively, to build a modular submersible repairing system meeting restrictions placed on the dimensions of the modular submersible-repairing system by a space between jet pumps 45 placed in a space between a pressure vessel 43 and a shroud 44. The modular submersible repairing system is suspended and lowered to a predetermined position as shown in FIG. 6, the submersible fan module 7 of the base unit is operated to move the modular submersible repairing system to the outer surface of the shroud 44 by a thrust produced by the submersible fan module 7. Then, the modular submersible repairing system is held fixedly on the shroud 44 by the agency of the suction cups 6a of the adsorbing module 6. The modular submersible repairing system is kept always in a fixed vertical position by the agency of the buoyant module 8 while the modular submersible repairing system is lowered in the pressure vessel 43. The manipulator module 5 for work on the outer surface of the shroud 44 is provided with the pantographic extension mechanism 9. Since the pin joints of the extension mechanism 9 include the spherical bearings 14, the working unit can be moved along the outer surface of the shroud 44 into a space between the jet pumps 45 and the shroud 44 and can be moved near to an objective part. The submersible fan module 3 is operated while the extension mechanism 9 is extending, so that the working unit does not separate from the surface of the shroud 44 and moves along the surface of the shroud 44. The manipulator module 5 is locked after the working unit has been thus moved near to a desired position to complete the positioning of the working unit. Subsequently, the X- and the Y-shaft of the scanning/pitching module 2 are operated to carry out batch work. After the completion of the work, the foregoing procedure is reversed to take out the modular submersible repairing system from the reactor. When repairing the inner surface of the shroud 44, the height of an adjusting module 47 is considered with reference to the height of a defect in the inner surface of the shroud 44 from a core plate 46 (FIG. 8), and an adjusting module 47 of a length and a shape suitable for repairing work is selected. Referring to FIG. 7 showing the adjusting module 47, end members 47b and 47c are connected to an upper part and a lower part, respectively, of a module body 47a of a predetermined length with bolts 48 so that height is adjustable. The end members 47b and 47c are provided with connecting units 49a and 49b, respectively. Referring to FIG. 8, the base unit is built by connecting the manipulator module 5, the adjusting module 47 and a fixing module 50. The base unit is lowered through an opening of an upper grid plate 51 in the reactor by a cable of the hoisting device 30, and is inserted in a control rod guide pipe 53 held on the core plate 46. The orientation of the fixing module 50 is determined by engaging a locating pin, not shown, in a locating hole of the fixing module 50. A locking mechanism, not shown, included in the fixing module 50 is operated to fix the base unit in the control rod guide pipe 53. Then, the cable of the hoisting device 30 is disconnected from the base unit and is taken out of the reactor. Subsequently, the scanning/pitching module 2 combined with the submersible fan module 3 and the buoyant module 4 is suspended and lowered in the reactor by the hoisting device 30. The scanning/pitching module 2 is passed through an opening of the upper grid plate 51 other than that through which the base unit was passed, the scanning/pitching module 2 is moved near to the manipulator module 5 in cooperation with the operation of the arm 34, and the female connecting unit 21 of the manipulator module 5 and the male connecting unit 18 of the scanning/pitching module 2 are engaged, in which the engagement of the taper member in the taper hole is assisted by the drawing claw 35 of the hoisting device 30. Upon the confirmation of the connection of the scanning/pitching module 2 and the manipulator module 5 from a signal provided by the ultrasonic distance measuring device 24, the locking device 28 is actuated to prevent the faulty operation of the hydraulic cylinder actuator 26. Then, the hooks 32 of the hoisting device 30 is disengaged from the scanning/pitching module 2 and the hoisting device 30 is taken out of the reactor. Subsequently, the tool module 1 is suspended and lowered in the reactor by the hoisting device 30, and the female connecting unit of the scanning/pitching module 2 and the male connecting unit of the tool module 1 are engaged. After the modules have been thus connected, the manipulator module 5 is operated to move the working unit near to the objective part, the tool module 1 is pressed against the shroud by the agency of the submersible fan module 3, and the scanning mechanism of the scanning/pitching module 2 carries out batch work. As apparent from the foregoing description, according to the present invention, the shape and configuration of the repairing system can be changed according to the condition of the object of work and is capable of carrying out repairing work for repairing structures of a boiling-water reactor which places severe dimensional restrictions. Various modules provided with standardized connecting units can be used for the efficient operation of the modular repairing system. Since the modules can be connected in water by a remotely controlled operation, the proper modules can be assembled in the reactor, the dimensional restrictions can be relaxed.
	claims	1. A grid having wall elements absorbing electromagnetic radiation, wherein the wall elements include a mixture of a material which is flowable in the processing state and an absorption material which absorbs the electromagnetic radiation, and wherein the wall elements exhibit a double comb structure with webs extending on two sides from a base surface, wherein the base surface takes the form of an absorbent foil provided with perforation holes, and wherein the webs are connected from one side of the foil to the other through the perforation holes. 2. A grid as claimed in claim 1, wherein the absorption material is embedded in the mixture in the form of particles. 3. A grid as claimed in claim 1, wherein the material which is flowable in the processing state contains or consists of a thermoplastic polymer selected from the group of polypropylene, liquid crystal polymer, polyamide, polycarbonate and/or polyoxymethylene. 4. A grid as claimed in claim 1, wherein the absorption material comprises a heavy metal. 5. A grid as claimed in claim 1, wherein the wall elements are arranged alternately with lamellae of an absorbent material. 6. A detector having a grid for the absorption of X-rays, wherein the grid comprises wall elements, which wall elements include a mixture of a material which is flowable in the processing state and an absorption material which absorbs electromagnetic radiation, wherein the wall elements exhibit a double comb structure with webs extending on two sides from a base surface, wherein the base surface takes the form of an absorbent foil provided with perforation holes, and wherein the webs are connected from one side of the foil to the other through the perforation holes. 7. A detector as claimed in claim 6 including radiation absorbent lamellae, wherein the wall elements are arranged alternately with the lamellae. 8. A detector as claimed in claim 6 wherein the absorption material is embedded in the mixture in the form of particles. 9. A detector as claimed in claim 6 wherein the material which is flowable in the processing state contains or consists of a thermoplastic polymer selected from the group of polypropylene, liquid crystal polymer, polyamide, polycarbonate and/or polyoxymethylene. 10. A detector as claimed in claim 6 wherein the absorption material contains or consists of a heavy metal. 11. An imaging device for generating an image of an object or part of an object by X-radiation, comprising a detector having a grid for the absorption of X-rays, wherein the grid comprises wall elements, which wall elements include a mixture of a material which is flowable in the processing state and an absorption material absorbing electromagnetic radiation, wherein the wall elements exhibit a double comb structure with webs projecting on two sides from a base surface wherein the base surface takes the form of an absorbent foil provided with perforation holes, and wherein the webs are connected from one side of the foil to the other through the perforation holes. 12. An imaging device as claimed in claim 11 wherein the wall elements are arranged alternately with the lamellae. 13. An imaging device as claimed in claim 11 wherein the absorption material is embedded in the mixture in the form of particles. 14. An imaging device as claimed in claim 11 wherein the material which is flowable in the processing state contains or consists of a thermoplastic polymer selected from the group of polypropylene, liquid crystal polymer, polyamide, polycarbonate and/or polyoxymethylene. 15. A method of producing a grid having wall elements absorbing electromagnetic radiation, wherein the wall elements are produced by injection molding from a mixture of a material which is flowable in the processing state and an absorption material absorbing electromagnetic radiation, wherein the wall elements are produced in a double comb structure with webs projecting on two sides from a base surface, wherein the base surface takes the form of an absorbent foil provided with perforation holes, and wherein the webs are connected from one side of the foil to the other through the perforation holes.
	abstract	The present invention relates to a method for the generation of 223Ra of pharmaceutically tolerable purity comprising: i) preparing a generator mixture comprising 227Ac, 227Th and 223Ra in a first aqueous solution comprising a first mineral acid; ii) loading said generator mixture onto a DGA separation medium (e.g. resin); iii) eluting said 223Ra from said DGA separation medium using a second mineral acid in a second aqueous solution to give an eluted 223Ra solution; and iv) stripping the DGA separation medium of said 227Ac and 227Th by flowing a third mineral acid in a third aqueous solution through the DGA separation medium in a reversed direction; The invention further relates to high purity radium-223 formed or formable by such a method as well as pharmaceutical compositions comprising such radium-223 of pharmaceutical purity.
	description	Certain embodiments of X-ray sources according to the invention generate X-rays from a plasma produced by focusing pulsed laser light on a target material inside a reduced-pressure chamber. Other embodiments generate X-rays from a plasma produced by converting a target material into a plasma using an electrical discharge. An X-ray optical element is used to receive the X-rays from the plasma and guide the X-rays to a downstream X-ray optical system. First Representative Embodiment This embodiment is depicted in FIG. 1. The X-ray source is an LPX source. X-rays are produced from a plasma 603 generated by discharging a target gas from a nozzle 600 at ultrasonic velocity while irradiating the discharged target gas with a pulsed laser light 602. Discharge of target gas through the nozzle 600 is controlled by a pulse valve 601. The generated X-rays have a nearly uniform intensity distribution within a plane parallel to the laser-incidence plane (a horizontal plane in the figure). A paraboloidal mirror 604 having a focal point in the middle of the plasma is situated as shown in the figure. The paraboloidal mirror 604 reflects the X-rays emitted from the plasma 603 and produces a collimated X-ray light flux 611 having an axially symmetric intensity distribution. The paraboloidal mirror 604 also directs the X-ray light flux 611 toward a downstream optical system. The paraboloidal mirror 604 is coated with multiple thin-film layers so as to be reflective to X-rays having a specified wavelength. The multi-layer period varies in a controlled manner across the reflective surface of the mirror 604 so as to maximize the reflectivity of the mirror at various locations on the mirror surface. The axis of rotational symmetry of the paraboloid is oriented so as to pass through the center of the plasma 603. Thus, the rotational axis of the paraboloidal mirror 604 is coincident with the optical axis of X-rays reflected from the mirror 604 (this optical axis is the axis of symmetry of the X-ray source). A rotational actuator 605 is situated relative to the paraboloidal mirror 604 and is configured to rotate the mirror 604 about its axis of rotational symmetry. The rotational actuator 605 is mounted via linear stages 608, 609, 610 on tilt stages 606, 607. The tilt stages 606, 607 are oriented perpendicularly to one another. The linear stages 608, 609, 610 also are oriented perpendicularly to one another in three dimensions. The combination of the tilt stages 606, 607 and linear stages 608-610 are a representative example of various mechanisms that can be utilized for accurately positioning and rotating the mirror 604. Contact-type displacement sensors 612, 613 are mounted, with respective orientations that are perpendicular to each other, on the side and rear surfaces (in the figure) of the paraboloidal mirror 604. (The direction orthogonal to the plane of the page is not depicted.) The particular configuration of the depicted displacement sensor 612 is exemplary only. Any of various types of displacement sensors can be employed. To facilitate displacement sensing, the side and rear surfaces (in the figure) of the paraboloidal mirror 604 extend very accurately parallel and perpendicular, respectively, to the axis of rotational symmetry of the mirror 604. The output of each displacement sensor 612, 613 is routed to a computer or other processor (not shown but well understood in the art) as a representative controller. During rotation of the mirror 604, any change in the rotational axis of the mirror 604 is detected by the displacement sensors 612, 613 as a corresponding shift in mirror position. Data from the displacement sensors 612, 613 are processed by the computer. If the computer determines that the magnitude of shift exceeds applicable specifications, then the computer initiates actuation of the tilt stages 606, 607 and/or the linear stages 608, 609 to return the mirror position to within specification. Thus, by providing in this embodiment respective devices for detecting mirror position, for controlling mirror position, and for actuating a drive mechanism to restore proper mirror position, displacement of the optical axis of X-rays reflected by the rotating mirror 604 is maintained within specification so as to ensure that any flying debris deposited on downstream optical components is distributed uniformly about the axis. If the accuracy of the mirror-rotation mechanism is sufficiently high for maintaining axial displacement of the rotating mirror within the maximal angular spread of the X-ray light flux accommodated by the downstream optical system, then the devices for detecting mirror position, controlling mirror position, and actuating drive mechanisms to restore mirror position can be omitted. Second Representative Embodiment An X-ray source (LPX source) according to this embodiment is shown in FIG. 2. X-rays are generated by a plasma produced by irradiation of laser light on a gaseous target material. The X-ray source is contained within a chamber 100 defining an interior space evacuated by a vacuum pump (not shown but well understood in the art). The pressure of the interior space is reduced to a level at which X-rays radiating from the plasma are not absorbed or excessively attenuated en route. The target-gas-delivery device in this embodiment is a gas nozzle 101 (desirably made of an inert metal such as stainless steel) from which the target gas (e.g., krypton) is discharged. Discharge from the gas nozzle 101 is controlled by a pulse valve 113. Target gas discharged from the nozzle 101 that is not converted into the plasma is evacuated to the external environment through an evacuation port 104 located axially opposite the nozzle 101 and connected to the vacuum pump. Any other target gas circulating in the vacuum chamber 100 is evacuated through the vacuum port 104 by the vacuum pump discussed above. The laser is incident along an optical axis, passing through the center of the plasma 102, extending perpendicularly to the plane of the page of FIG. 2. I.e., the laser pulses are incident at the plasma 102 from below the plane of the page along an axis perpendicular to the plane of the page. Pulsed laser light emitted from the laser device (not shown but well understood in the art) is focused by a condenser lens (not shown) at a position 0.5 mm from the tip of the nozzle 101, along the axis of the nozzle, to produce the plasma 102. The shape of the plasma 102 is filamentous, with a length of approximately 300 xcexcm along the optical axis of the laser and approximately 100 xcexcm perpendicular to the optical axis of the laser. The plasma 102 is produced approximately 500 xcexcm toward the condenser lens from directly in front of the nozzle 101. A paraboloidal mirror 103 and the nozzle 101 are situated such that the plasma 102 is formed substantially at the focal point of the mirror 103. Regarding the X-rays emitted from the plasma 102, only those X-rays of a specified wavelength (e.g., 13 nm) are reflected by the mirror 103. To such end, the paraboloidal mirror 103 is coated with multiple thin-film layers. X-rays reflected from the mirror 103 are collimated and pass through a filter 110 that is opaque to visible light but transmissive to X-rays. By way of example, the filter 110 comprises a thin film of zirconium (Zr), 150 nm thick, formed on a mesh of nickel (Ni). The mesh is supported by a holder 111. The X-rays passing through the filter 110 propagate to a downstream X-ray optical system (not shown). The mirror 103 is supported by a stage assembly, comprising an annular ultrasonic motor 105 situated and configured to rotate the paraboloidal mirror 103 around its axis of rotational symmetry. The stage assembly also comprises three axially orthogonal stages 106, 107, 108 for determining and controlling the position of the mirror 103, and a tilting stage 109 for controlling the inclination of the mirror 103. The stages 108, 109 are mounted behind the mirror 103, and the stages 106, 107 are displaced laterally from the stages 108, 109. The stages 106-109 can be driven by respective motors or other actuators from outside the vacuum chamber 100. In this embodiment, a set of multiple (desirably three) semiconductor lasers and respective photodiodes is used for detecting the position and inclination of the paraboloidal mirror 103. The semiconductor lasers and photodiodes are disposed adjacent the paraboloidal mirror 103 at positions that do not block X-rays reflected by the mirror 103. The semiconductor lasers are positioned at 120xc2x0 relative to each other. The respective photodiodes also are positioned at 120xc2x0 from each other, but angularly between the lasers. This scheme is depicted in FIG. 3(A) so as to be understood readily, wherein FIG. 3(A) represents a view, from a location on the mirror axis but downstream of the mirror 103, toward the mirror 103. A laser beam from the semiconductor laser 201 strikes a point on the surface of the paraboloidal mirror 103 and is reflected toward a respective photodiode 204. Respective laser beams from the other two semiconductor lasers 202, 203 are likewise reflected by the surface of the mirror 103 toward respective photodiodes 205, 206. Hence, the points on the surface of the mirror 103 irradiated by the laser beams are at 120xc2x0 relative to each another. Each photodiode 204, 205, 206 has a respective light-reception surface 208 that is partitioned into four portions 208a-208d, as shown in FIG. 3(B). Each portion produces a respective electrical signal from respective incident light of the reflected laser beam. These electrical signals are routed to the computer (discussed above). In a situation in which the nozzle 101, the paraboloidal mirror 103, and the downstream optical system are all aligned with each other, the respective signals output from the photodiodes 204-206 (3 photodiodesxc3x974 portions each=12 signals) are received by, stored in, and processed by the computer. This situation represents an xe2x80x9cinitial statexe2x80x9d of the system. Upon starting up this X-ray source, rotation of the paraboloidal mirror 103 commences, as effected by the ultrasonic motor 105. The rotational velocity of the mirror is a function of the rate at which flying debris from the plasma adhere to and accumulate on neighboring structures. If the rate of production of flying debris by the plasma is low, then a low rotational velocity is permissible. Conversely, higher-velocity rotation is necessary if the rate of particle adhesion is high. By way of example, the LPX of this embodiment tends to emit low quantities of flying debris, so the rotational velocity of the mirror 103 can be one revolution per hour. In any event, if alignment of the mirror 103 shifts during rotation, then the positions at which the respective reflected laser beams from the semiconductor lasers 201-203 reach the respective photodiodes 204-206 change accordingly. These changes produce corresponding changes in the magnitudes of respective signals produced by the portions 208a-208d of the light-reception surface 208 in each photodiode 204-206. If the differences in electrical outputs from the portions 208a-208d of the light-reception surfaces in the photodiodes 204-206, relative to the initial conditions, exceeds predetermined thresholds, then the computer will detect an excessive misalignment of the mirror 103 and will cause the inclination stage 109 and the linear stages 106-108 to apply corrective positioning of the mirror 103 to return the electrical signals to within specifications. In addition, the direction of mirror shift can be ascertained from the signal changes in the four portions 208a-208d of the respective light-reception surface 208 of each photodiode 204-206, allowing the stages 106-109 to be actuated appropriately to correct the shift. Hence, changes in the optical axis of the X-rays reflected from the mirror 103 are maintained within specifications so as not to have any adverse effect on downstream optical systems, even while rotating the mirror 103. In addition, there is no loss in the axial symmetry of the X-rays reflecting from the mirror 103, even if the angular distribution of the flying debris is asymmetrical. In this embodiment, the X-ray filter 110 is located in the chamber 100. As a result, flying debris from the plasma also can accumulate on the filter 110. If the angular distribution of the flying debris is asymmetric, then the debris will accumulate asymmetrically on the filter 110. As a result, the flux of X-rays transmitted by the filter 110 will become asymmetric. Therefore, in this embodiment, an annular ultrasonic motor 112 (or analogous actuator) is installed on the perimeter of the holder 111 on which the X-ray filter 110 is mounted, so as to rotate the filter 110 around the center axis of the X-ray flux. The filter rotation prevents degradation of the symmetry of the transmitted X-ray flux, by ensuring that the quantity of flying debris accumulating on the filter 110 is axially symmetric relative to the X-ray optical axis. In this embodiment, since the rate of change in the transmissivity of the filter 110 is miniscule, even if some shift occurs in the rotational axis of the filter 110, a filter-position sensor is not normally necessary (and hence is not shown). By rotating the filter 110 as described above, variations in the transmission of X-rays through the filter 110 can be ameliorated (e.g., variances arising by variances in the thickness of the filter material and/or of the mesh support members). This is especially effective whenever the X-ray source of this embodiment is used for performing microlithographic exposures, as in soft X-ray (EUV) microlithography apparatus and methods. The respective rotational velocities of the mirror 103 and filter 110 may be equal, or they may be different according to the operating status of the X-ray source. In addition, the respective directions of rotation of the mirror and filter may be the same or different. Although a paraboloidal mirror 103 is used in this embodiment, it will be understood that the mirror alternatively can be a spherical mirror or an ellipsoidal mirror. The mirror also may be a rotationally symmetrical a spherical mirror. The mirror surface (whether spherical, paraboloidal, ellipsoidal, and/or a spherical) can be formed on a single substrate, or alternatively on a substrate divided into multiple segments conjoined into a single unit or situated adjacent one another. In this embodiment, the light-receiving surface 208 of each photodiode 204-206 was divided into four portions 208a-208d. However, the number of portions is not limited to four. Alternatively, each light-receiving surface 208 can be divided into two, three, or more portions, or not divided at all. The photodiodes 204-206 can be one-dimensional (as in photodiode arrays), or two-dimensional (as in CCDs). Although semiconductor lasers were used in this embodiment to measure displacements of the mirror 103, other measuring devices alternatively can be used such as contact-needle displacement gauges (see FIG. 1), over-current sensors, ultrasonic sensors, electrostatic capacity sensors, etc. The mirror can be disposed in any orientation relative to the plasma. FIG. 4 shows an example configuration employing a discharge-plasma X-ray source (dense-plasma focus, or DPF source). In FIG. 4, only the electrodes (anode 300, cathode 301) of the DPF source are shown, and the power supply is not shown. A multilayered paraboloidal mirror 305 is situated laterally adjacent the electrode. Also not shown are a mirror-drive mechanism and a device for detecting mirror position. If the mirror is planar it can be rotated using the direction of a normal ray as an axis. FIG. 5 shows a situation in which a multilayer planar mirror 404 is used, together with a gas-jet LPX used to generate X-rays from a plasma 402. The multilayer planar mirror 404 is rotated about the normal-ray axis AA. Not shown are a mirror-drive mechanism and a device for detecting mirror position. A laser beam 403 is focused at the location of the plasma 402. Whereas a multilayer mirror is used in the embodiments described above, a grazing-incidence mirror alternatively can be used for achieving full reflection of incident X-rays. An example configuration employing a grazing-incidence paraboloidal mirror 502 is shown in FIG. 6, used in conjunction with a DPF for generating the X-rays. In this figure, only the electrodes (anode 500, cathode 501) of the DPF source are shown; the power supply is not shown. The DPF source produces a plasma 504 at the location shown, relative to the mirror 502. In FIG. 6, the mirror 502 is rotated about its axis of symmetry (axis Bxe2x80x94B), which is the propagation axis of X-rays reflected from the mirror 502. (The mirror-drive mechanism and device for detecting mirror position are not shown.) Although the mirror 502 has a paraboloidal reflective surface, the mirror 502 alternatively can have an ellipsoidal reflective surface or a reflective surface having a combination of these profiles (e.g., a Walter mirror). Item 503 is an axial beam stop useful for producing a collimated beam. Although LPXs were used in several of the embodiments described above in which gas jets were used, LPXs employing mechanisms in which the target material is discharged in clusters, a liquid jet, liquid droplets, microdroplets, or microparticles alternatively can be used. The target material used for LPXs or discharge-plasma X-ray sources is not limited to krypton. Alternatively, the target material can be, e.g., xenon (Xe), carbon dioxide (CO2), or lithium (Li), or a mixture or compound of any of these substances. As an alternative to using a DPF source for generating X-rays, other configurations of discharge-plasma X-ray sources alternatively can be used. For example, a Z-pinch plasma source or a capillary-discharge plasma source can be used. By employing a rotating reflective optical element (onto which X-rays generated from the plasma are initially incident), or other rotating optical element in the vicinity of the plasma, flying debris will accumulate in an axially symmetrical fashion on the optical element, even if the flying debris is emitted from the plasma in a spatially irregular distribution. As a result, the axial symmetry of the X-ray flux reflected or transmitted by the optical element is maintained. Consequently, there is no decrease in the performance of a downstream optical system requiring an axially symmetrical X-ray flux, even if the X-ray source is operated for a long period of time. Whereas the invention has been described in connection with multiple representative embodiments, it will be understood that the invention is not limited to those embodiments. On the contrary, the invention is intended to encompass all modifications, alternatives, and equivalents as may be included within the spirit and scope of the invention, as defined by the appended claims.
	claims	1. A reflective particle tag reader system comprising:a read head assembly, the read head assembly comprising a camera, a plurality of illuminators, and a rigid frame portion for supporting the camera and the plurality of illuminators;the plurality of illuminators mounted to the frame and directed to illuminate a focal point located opposite the camera, the focal point being the location of the reflective particle tag;a computer having a display, a processor, a data communication input and output means, and a data storage device; the computer in data communication with the camera to receive and store one or more images of the reflective particle tag acquired by the camera; andthe computer configured to process video images and to quantify a positional alignment parameter and an angular alignment parameter of the reader with respect to the reflective particle tag;wherein a rapid burst of image frames is obtained in response to the positional alignment and the angular alignment parameters being within a predetermined tolerance and identity of the reflective tag is established between a first image set and a second image set. 2. The system of claim 1, wherein, the one or more images of the reflective particle tag further comprises a high-frequency sequence of camera images, the camera having an aperture in synchronization with a strobe frequency of the illuminators to obtain the high-frequency sequence of camera images of the reflective particle tag. 3. The system of claim 1, further comprising a bandpass filter connected to the camera, the bandpass filter configured to reject ambient illumination and pass light emitted in a narrow spectral band. 4. The system of claim 1, wherein a set of fiducials is projected on the display screen to determine alignment of the camera with respect to the reflective particle tag. 5. The system of claim 1, wherein a set of features for each image is calculated via an algorithm, descriptors defining the features, and focus measures; and wherein the set of features is stored on the computer. 6. The system of claim 1, wherein during alignment, the computer further comprises a vision system, the vision system configured to analyze each image recorded by the camera and acquire a set of image features associated with respective image; compare the acquired features to the stored features, and calculate a homography matrix to provide a lateral displacement and azimuthal rotation between the acquired image feature set and stored feature set. 7. The system of claim 6, wherein a pair crosshair fiducials is projected by the visions system on the display screen; a first fiducial representing the coordinate system of an image collected on an imager chip and a second fiducial representing a coordinate system of the stored feature set. 8. The system of claim 5, wherein alignment of the first and second fiducials provides alignment of three degrees of orientation and rotation; and wherein the remaining three degrees of freedom are aligned using the acquired image feature set and the corresponding stored set of focus measures. 9. The device of claim 8, wherein each of the focus measures is acquired at a different spatial location within the respective image; and wherein if the reader is positioned at an angle relative to the conditions that were used for the stored data, then better focus measures will be obtained for some portions of the image relative to other portions of the same image. 10. The system of claim 9, wherein a balance bubble fiducial is projected on the display to indicate relative balance of all the focus measures, and wherein all degrees of freedom are aligned when the first and second fiducials are matched and the balance bubble is centered over the first and second fiducials. 11. A non-transitory computer-readable storage medium having stored thereon instructions which, when executed by one or more processing units, cause the one or more processing units to perform a method for maintaining and verifying authenticity of a reflective particle tag comprising:illuminating a focal point located opposite the camera;placing a reflective particle tag at the focal point;receiving by a computer in data communication with a camera one or more images of the reflective particle tag acquired by the camera;processing video images and quantifying a positional alignment parameter and an angular alignment parameter of the reader with respect to the reflective particle tag;obtaining a rapid burst of image frames in response to the positional alignment and the angular alignment parameters being within a predetermined tolerance; andauthenticating an identity of the reflective tag between a first image set and a second image set. 12. The non-transitory computer-readable storage medium of claim 11, wherein the method further comprises:analyzing each frame recorded by a camera and rapidly acquiring a set of image features for the incoming image frame; comparing the acquired features to the stored set of features; andcalculating a homography matrix using features that provide a good match between the current and stored feature sets via the lateral displacement and azimuthal rotation between the acquired features and stored features. 13. The system of claim 1, wherein the system acquires a burst of ˜100 full resolution (2048×2048) frames at a rate of 90 frames per second. 14. The system of claim 1, wherein each illuminator of the plurality of illuminators project approximately F/2 beams.
	description	This application claims priority under 35 U.S.C. §119(e) from U.S. Provisional Application No. 61/352,484, entitled “Mechanical Control Rod Drive Shaft Unlatching Tool”, filed on Jun. 8, 2010, the disclosure of which is incorporated herein by reference. 1. Field of the Invention The present invention relates to nuclear power plants, such as, without limitation, pressurized water reactor (PWR) type nuclear power plants, and in particular to a tool for unlatching and removing a control rod drive shaft in a nuclear reactor vessel. 2. Description of the Related Art In nuclear power generation, a reactor vessel is the primary vessel wherein heat is generated for producing steam. The reactor vessel typically includes a flanged body having a flanged, removable upper closure head bolted atop its upper portion for forming a sealed enclosure. Fuel pellets, which are located within fuel assemblies, are positioned within the reactor vessel for producing a controlled nuclear fission reaction which, in turn, generates heat. The heat generated by the fission reaction heats borated water that is contained within the reactor vessel. Process piping, generally referred to in the art as a primary loop, is attached to the reactor vessel. The heated borated water flows out of the reactor vessel and passes through the primary loop to a steam generator for transferring its heat to a secondary loop, wherein steam is produced for ultimately producing electrical power, as is well known in the art. The borated water then returns to the reactor vessel via the primary loop where the above described process is repeated. In a pressurized water reactor (PWR), and in contrast to a boiling water reactor (BWR), pressure in the primary loop prevents the borated water from boiling within the reactor. The rate of the fission reaction taking place within each fuel assembly is regulated by means of an associated control rod assembly. Each control rod assembly is formed from an array of stainless steel tubes containing a neutron absorbing substance, such as silver, indium or cadmium. These stainless steel tubes (known as “rodlets” in the art) are suspended from a spider-like bracket, and a control rod drive shaft (CRDS) is connected to the spider-like bracket. Each CRDS is also coupled to a control rod drive mechanism (CRDM) carried by the closure head. Each CRDM is structured to either insert or withdraw the rodlets of the associated control rod assembly deeper into or farther out of the associated fuel assembly in order to modulate the amount of heat generated thereby. Periodically, nuclear reactors must be refueled, a process wherein a fraction of the fuel assemblies of the reactor are replaced. During the refueling of a nuclear reactor, the closure head is removed, the reactor vessel is flooded with water and the upper internals of the reactor vessel are removed. When this is done, however, the rodlets need to remain in place within the reactor vessel. Thus, prior to removing the upper internals of the reactor vessel, each CRDS (which is carried by the upper internals) must be disconnected from the cluster of rodlets to which it is attached so that the rodlets will not be carried away with the drive shaft but instead will remain in place. More specifically, each spider bracket has a grooved circular ferrule hub and the bottom of each CRDS has a pair of fingers that are received in the ferrule to couple the CRDS to the spider bracket. This connection must be released so that the CRDS can be removed from the reactor vessel while leaving the rodlets in place. For some time, a prior art CRDS unlatching tool has been used to release the connection between a CRDS and a spider bracket. That tool uses a first pneumatic mechanism to actuate a first set of latch fingers provided on a button shaft which engage a top portion of the CRDS and disengage the CRDS from the spider bracket (it causes the fingers to be released from the ferrule hub) and a second pneumatic mechanism to actuate a second set of latch fingers which engage an outer surface of the CRDS and allow it to be held by the tool while it is removed. The problem with this prior art tool is that it undesirably permits a tool operator to inadvertently actuate the air cylinders of the second pneumatic mechanism while the tool is latched onto the CRDS, thereby allowing the CRDS to fall out of the tool. As will be appreciated, this has the potential to cause costly damage to the CRDS itself and to nearby equipment and/or injury to nearby personnel. In one embodiment, a tool for unlatching a control rod drive shaft of a nuclear reactor vessel is provided that includes a support assembly structured to receive the control rod drive shaft in a first end thereof and a latching assembly, wherein the support assembly is received within the latching assembly in a manner wherein the latching assembly is moveable relative to the support assembly. The support assembly has a plurality of latch fingers positioned at the first end thereof and at least one pin positioned at a second end thereof, each of the latch fingers being movable between a latched position wherein the latch finger is structured to engage and hold the control rod drive shaft when the control rod drive shaft is received in the first end and an unlatched position wherein the latch finger is structured to not engage the control rod drive shaft when the control rod drive shaft is received in the first end. The latching assembly includes a first sleeve member at a first end thereof and a second sleeve member at a second end thereof, the second sleeve member having at least one slot, wherein the at least one pin is moveably received within the at least one slot, wherein the latching assembly is movable in an unlatching manner from a latched state to an unlatched state wherein the latching assembly slides relative to the support assembly in a first direction and causes the first sleeve member to engage each latch finger and move each latch finger from the latched position to the unlatched position and wherein the latching assembly is movable in a latching manner from the unlatched state to the latched state wherein the latching assembly slides relative to the support assembly in a second direction opposite the first direction and causes the first sleeve member to engage each latch finger and move each latch finger from the unlatched position to the latched position. These and other objects, features, and characteristics of the present invention, as well as the methods of operation and functions of the related elements of structure and the combination of parts and economies of manufacture, will become more apparent upon consideration of the following description and the appended claims with reference to the accompanying drawings, all of which form a part of this specification, wherein like reference numerals designate corresponding parts in the various figures. It is to be expressly understood, however, that the drawings are for the purpose of illustration and description only and are not intended as a definition of the limits of the invention. As used in the specification and in the claims, the singular form of “a”, “an”, and “the” include plural referents unless the context clearly dictates otherwise. Directional phrases used herein, such as, for example and without limitation, top, bottom, left, right, upper, lower, front, back, and derivatives thereof, relate to the orientation of the elements shown in the drawings and are not limiting upon the claims unless expressly recited therein. As employed, herein, the statement that two or more parts or components are “coupled” together shall mean that the parts are joined or operate together either directly or through one or more intermediate parts or components. As employed herein, the statement that two or more parts or components “engage” one another shall mean that the parts exert a force against one another either directly or through one or more intermediate parts or components. As employed herein, the term “number” shall mean one or an integer greater than one (i.e., a plurality). FIGS. 1 and 2 are schematic diagrams of a CRDS unlatching tool 2 according to one exemplary embodiment of the present invention. In FIG. 1, CRDS unlatching tool 2 is shown in a latched condition wherein is structured to engage the outer surface of a CRDS and securely hold the CRDS, and in FIG. 2, CRDS unlatching tool 2 is shown in an unlatched condition wherein a CRDS is not held by the tool. As described in greater detail herein, CRDS unlatching tool 2 provides a mechanical latching function which replaces the pneumatically operated latch gripper assembly of the prior art described above (i.e., the second pneumatic mechanism). More specifically, and as described in greater detail below, the CRDS unlatching tool 2 incorporates a mechanical interlock to prevent a tool operator from inadvertently unlatching a CRDS from CRDS unlatching tool 2 during operation. CRDS unlatching tool 2 includes a CRDS support assembly 4 that is provided within a mechanical latching assembly 6. As described in detail herein, CRDS support assembly 4 is structured to engage a top portion of the CRDS and disengage the CRDS from the spider bracket, and mechanical latching assembly 6 is structured to actuate latch fingers 8 which engage an outer surface of the CRDS and allow it to be held by the CRDS unlatching tool 2 while it is removed. Referring to FIGS. 3-6, CRDS support assembly 4 includes an upper support assembly 10 coupled to a lower support assembly 12. FIG. 3 is a side elevational view of upper support assembly 10 and FIG. 4 is a cross-sectional view of upper support assembly 10 taken along lines 4-4 of FIG. 3. FIG. 5 is a side elevational view of lower support assembly 12 and FIG. 6 is a cross-sectional view of lower support assembly 12 taken along lines 6-6 of FIG. 5. Referring to FIGS. 3 and 4, upper support assembly 10 includes a shroud 14 having a first inner tube portion 16 attached thereto. In addition, a pneumatic cylinder 18 having couplings 20, 22 attached thereto is coupled to the top of shroud 14. Dowel pins 19A and 19B extend from opposite sides of shroud 14, and a hole 21 is provided in the lower end of shroud 14. The functions of dowel pins 19A and 19B and hole 21 are described elsewhere herein. Shroud 14 also includes at least one window 24. Upper support assembly 10 further includes an upper button shaft member 26 that is moveably housed within first inner tube portion 16. Upper button shaft member 26 extends through shroud 14 and is operatively coupled to pneumatic cylinder 18 through coupling 28. Pneumatic cylinder 18, coupled to a pneumatic source through a valve assembly (not shown), is thus able to selectively drive upper button shaft member 26 within first inner tube portion 16 along the longitudinal axis thereof. A threaded adaptor 30 is provided at the distal end of first inner tube portion 16 and upper button shaft member 26. The function of threaded adaptor 30 is described elsewhere herein. Referring to FIGS. 5 and 6, lower support assembly 12 includes a lower housing 32 having an adaptor 34 coupled thereto which defines a CRDS receiving orifice 36. A shroud 38 having a window 40 is coupled to the other end of lower housing 32. A second inner tube portion 42 is coupled to shroud 38 through an adaptor 44. Lower support assembly 12 also includes a lower button shaft member 46 that is moveably housed within second inner tube portion 42. Lower button shaft member 46 is coupled to an actuator 48 having an actuator housing 52 by a coupling 50. As seen in FIG. 6, a spring 54 is provided within lower housing 32 and is structured to bias actuator 48 forward toward adaptor 34 and CRDS receiving orifice 36. Actuator 48 is operatively coupled to button fingers 56 such that when actuator 48 is pulled backwards against the spring bias as described elsewhere herein, button fingers 56 will be caused to pivot and extend through holes 58 provided in cylindrical member 60 of actuator 48 to grab and hold the top portion of the CRDS and disengage the CRDS from the spider bracket (as described elsewhere herein, it causes the lower fingers of the CRDS to be released from the ferrule hub of the spider bracket). In addition, the latch fingers 8 described elsewhere herein, which are structured to engage and hold the outside to the CRDS, are pivotably held within lower housing 32. The manner in which the latch fingers 8 are selectively actuated is described elsewhere herein. In the illustrated embodiment, lower housing 32 includes three latch fingers 8, although more or less latch fingers 8 may be provided in lower housing 32 within the scope of the present invention. A welded adaptor 62 is provided at the distal end of second inner tube portion 42 and lower button shaft member 46. The function of welded adaptor 62 is described elsewhere herein. Referring again to FIGS. 1 and 2, mechanical latching assembly 6 includes an upper latch member 64 and a lower latch member 66. Upper latch member 66 includes an upper latch housing 68. A bail plate 70 having a bail 72 is attached, preferably by welding, to the top end of upper latch housing 68. An upper tube 74 is attached, preferably by welding, to the bottom end of upper latch housing 68. Upper latch housing 68 has at least one window 76 provided therein. An inverted J-shaped slot 78A is provided on a first side of upper latch housing 68. A similar inverted J-shaped slot 78B is provided on a second side of upper latch housing 68 opposite the first side. A latch orifices 80 and 82 and an unlatch orifices 84 and 86, each structured to receive a respective end of a pin member 88, are also provided within upper latch housing 68 near the bottomed end thereof. The function of each of these components is described elsewhere herein. Lower latch member 66 includes a lower latch housing (sleeve) 90. A lower tube 92 is attached, preferably by welding, to the top end of lower latch housing 90. Lower latch housing 90 has windows 94A, 94B provided therein. In addition, three inverted L-shaped slots 96 are provided on the lower end of lower latch housing 90. As seen in FIGS. 1 and 2, the L-shaped slots 96 are structured to be in alignment with latch fingers 8. In the exemplary embodiment, CRDS unlatching tool 2 is assembled as follows. First, upper support assembly 10 is inserted into upper latch member 64 through upper tube 74 and lower support assembly 12 is inserted into lower latch member 66 through lower latch housing 90. When this is done, the end of upper support assembly 10 is allowed to extend slightly out of upper tube 74 and the end of lower support assembly 12 is allowed to extend slightly out of lower tube 92. Next, upper support assembly 10 and lower support assembly 12 are coupled to one another as shown in FIGS. 7 and 8. More specifically, lower button shaft member 46 and upper button shaft member 26 are coupled to one another using a dowel pin 9 as seen in FIGS. 7 and 8. Next, coupling halves 98A, 98B are provided around the junction of threaded adaptor 30 and welded adaptor 62 and secured to one another using any suitable means such as a number of screws. Next, upper latch member 64 and lower latch member 66 are slid toward another and secured to one another by bolting the two components together through flanges 100A, 100B that are provided, preferably by welding, on the ends of upper tube 74 and lower tube 92, respectively. In addition, when so assembled, each dowel pin 19A, 19B is received through a respective inverted J-shaped slot 78A, 78B. Also, each latch finger 8 is aligned with a respective L-shaped slot 96. The operation of CRDS unlatching tool 2 will now be described. During operation of CRDS unlatching tool 2, the various states thereof will be determined by two things: (i) the position of the latch fingers 8, i.e., whether they are portioned inward so as to engage and grip the outer surface of a CRDS (latched) or outward so as to be out of engagement with the outer surface of a CRDS (unlatched), and (ii) the position of the button shaft formed by upper button shaft member 26 and lower button shaft member 46, i.e., whether it is pneumatically driven up or down within CRDS support assembly 4. When the button shaft is up, button fingers 56 will be caused to pivot and extend through holes 58 provided in cylindrical member 60 of actuator 48 to grab and hold the top portion of the CRDS, and conversely, when the button shaft driven down, button fingers 56 will be caused to pivot out of holes 58. For purposes of describing operation of CRDS unlatching tool 2, the following discussion will commence with CRDS unlatching tool 2 in an unlatched, button down state as shown in FIGS. 2, 9, 10 and 11. In the unlatched, button down state, mechanical latching assembly 6 is in a raised, upward position such that the bottom portion of lower latch housing 90 is positioned toward the top end of lower housing 32. In this state, the bottom of longer portion of the L-shaped slots will engage the top of each latch finger 8 (at an upper 102 thereof) and cause it to extend outwardly beneath lower latch housing 90 and out of the interior chamber of lower housing 32. In addition, each dowel pin 19A, 19B will be positioned at and against the bottom terminal end of the associated J-shaped slot 78A, 78B (FIGS. 2 and 11). Furthermore, in this state, the position of upper latch housing 68 will cause unlatch orifice 86 to be aligned with and positioned over hole 21 of shroud 14. In the exemplary embodiment, a first end of pin member 88 is inserted into unlatch orifice 86 and hole 21 and a second end of pin member 88 is inserted into unlatch orifice 84. Pin member 88 thus acts as a locking mechanism that prevents movement of mechanical latching assembly 6 relative to CRDS support assembly 4 until the pin member 88 is removed. In the exemplary embodiment, pin member 88 is attached to bail 72 by a lanyard 106. Next, to remove a CRDS, CRDS unlatching tool 2 is placed over the CRDS in a manner wherein the CRDS is received through CRDS receiving orifice 36 into the interior chamber of lower housing 32. Pin member 88 is then removed from 21 and unlatch orifices 86 and 84. CRDS unlatching tool 2 is then moved to a latched, button down state as shown in FIGS. 1, 7, 12 and 13. This is done by lowering the mechanical latch assembly 6 (moving it to the right is FIG. 11), rotating it clockwise (by rotating bail 72), and lifting it slightly to latch. During this process, dowel pins 19A and 19B will traverse the length of the associated J-shaped slot 78A, 78B such that each ends up in the position shown in FIGS. 1 and 13 wherein it is positioned at and against the top, opposite terminal end of the associated J-shaped slot 78A, 78B in the notch formed thereby. Also during this process, the bottom portion of lower latch housing 90 will be moved toward the bottom end of lower housing 32 as shown in FIG. 1. During such movement, the bottom edge of lower latch housing 90 will engage a lower cam 104 on the outside of each latch finger 8 and force latch finger 8 into the interior chamber of lower housing 32 as shown in FIG. 12 so that the latch fingers 8 will engage and hold the CADS. Moreover, upper cam 102 of each latch finger will be received within the shorter portion of the L-shaped slots as seen in FIGS. 1 and 12. In this state, upper latch housing 68 will be positioned in a manner wherein latch orifice 82 is aligned with and positioned over hole 21 of shroud 14. The operator then inserts the first end of pin member 88 into latch orifice 82 and hole 21 and the second end of pin member 88 is inserted into latch orifice 80 to lock CRDS unlatching tool 2 in the latched state. Next, CRDS unlatching tool 2 is moved to a latched, button up state as shown in FIG. 14 by pneumatically driving the button shaft formed by upper button shaft member 26 and lower button shaft member 46 up toward the top of CRDS unlatching tool 2 as described elsewhere herein. When this is done, actuator 48 is pulled backwards against the spring bias as described elsewhere herein, and button fingers 56 are caused to pivot and extend through holes 58 provided in cylindrical member 60 of actuator 48 to grab and hold the top portion of the CRDS and disengage the CRDS from the spider bracket. With the CRDS disengaged from the spider bracket and held by the latch fingers 8, CRDS unlatching tool 2 and thus the CRDS it holds may be safely removed from the reactor vessel (e.g., using a hoist coupled to bail 72) and moved to a storage location. Spring 108 inside mechanical latching assembly 6 is used to support the weight of CRDS latch assembly 6 when CRDS unlatching tool 2 is moved by the operator. Once safely removed from the reactor vessel, the process just described above may be reversed to move the CRDS unlatching tool 2 back to the unlatched, button down state with pin member 88 in the UNLATCH position (in unlatch orifices 84 and 86) so that CRDS unlatching tool 2 can be separated from the CRDS (e.g., again using a hoist coupled to bail 72) and used to remove another CRDS. The mechanical latching and unlatching mechanism just described replaces the prior art design which required two valve operated air cylinders to move the latch assembly strictly up and down. Thus, it will be appreciated that CRDS unlatching tool 2 reduces the danger of a CRDS being dropped. More specifically, this danger is reduced because CRDS support assembly 4 hangs from mechanical latching assembly 6 with each dowel pin 19A, 19B in the notch at the end of the horizontal portion of the associated inverted L-shaped slot 78A, 78B. This feature will not permit the CRDS to be unlatched from the CRDS unlatching tool 2 unless the CRDS is resting on (i.e., uncoupled from) or in (i.e., coupled to) the control rod hub (the spider bracket) or the CRDS is seated in a CRDS storage stand location mounted, for example, on the wall of the refueling cavity. This action permits the mechanical latching assembly to be lowered slightly to move the inverted J-shaped slot 78A, 78B down and away from contact with dowel pins 19A, 19B. This is only possible when the operator removes the one-piece pin member 88 from the LATCH location (latch orifices 80 and 82). As noted above, pin member 88 is secured from dropping by lanyard 106. Although the invention has been described in detail for the purpose of illustration based on what is currently considered to be the most practical and preferred embodiments, it is to be understood that such detail is solely for that purpose and that the invention is not limited to the disclosed embodiments, but, on the contrary, is intended to cover modifications and equivalent arrangements that are within the spirit and scope of the appended claims. For example, it is to be understood that the present invention contemplates that, to the extent possible, one or more features of any embodiment can be combined with one or more features of any other embodiment.
045004498	description	DETAILED DESCRIPTION OF THE INVENTION In the method according to the invention a mole ratio of sodium to boron of either 0.25 or 0.7 is set by adding sodium hydroxide to the residues, corresponding to a pH-value ranges of 7.3 to 8.0 for a mole ratio of 0.25, and 9.8 to 10.2 for a mole ratio of 0.7. For, it has been found that, contrary to the embedment conditions up to now, for instance, with a mole ratio of about 1, the following disadvantages are avoided with the mole ratios according to the invention: The resistance of the bitumen to strong caustic solutions is relatively poor. Therefore, the leaching resistance of the end product (bitumen with embedded residues) is improved substantially by the invention because of the smaller amount of caustic solution. In addition, because of the reduced requirement for sodium hydroxide, the mole ratio according to the invention results in a reduction of up to 50% in waste volume. The composition of the embedded borates is substantially less aggressive, i.e. reactive with respect to the bitumen, so that the danger of liberation of flammable vapors, previously observed, is practically completely avoided. The maximum embedment quantity of the residues is obtained with a mole ratio of 0.7. There, in accordance with a further feature of the invention, the processing and storage are carried out at temperatures of at least 50.degree. C. and preferably 80.degree. C. in order to avoid crystallization, which occurs for this mole ratio at lower temperatures. Of course, unduly high temperatures such as would cause evaporation of the concentrate in storage should not be employed. For this purpose the containers used for storing the residues can be provided with heating means such as a steam jacket or coil, so that the heating to the above-mentioned temperatures can be carried out not later than when the sodium hydroxide is added in accordance with the invention. With the above-mentioned mole ratio of sodium to boron of 0.7, the pH value is between 9.8 and 10.2 at 20.degree. C. The solubility of the borate is 125 g boron/kg at 80.degree. C. and about 5 g boron/kg bitumen at 20.degree. C. The solids produced during the embedding in bitumen have approximately the following composition: EQU 2Na.sub.2 O.times.3B.sub.2 O.sub.3 .times.4H.sub.2 O. These solids are present in the form of a melt at the operating temperatures of the worm dryer used as the mixing apparatus, so that abrasion and salt incrustation in suction lines associated with the mixing device are minimized. Also the leaching resistance of the end product is very favorable because of the low solubility of these borates at low temperatures. With a mole ratio of sodium to boron of 0.25, on the other hand, the processing and storage temperature can be in the range of room temperatures. For this ratio, a pH-value of 7.3 to 8.0 is obtained at 20.degree. C. The solubility of the borates is about 100 g boron/kg at 80.degree. C. and 20 g boron/kg at 20.degree. C. The solids have the composition: EQU Na.sub.2 O.times.4B.sub.2 O.sub.3 .times.4H.sub.2 O. Operating with a sodium-boron mole ratio of 0.25 prevents any salt encrustation in exhaust domes and lines and reduces the wear of the dryer worms. The leaching resistance of the end products is good. The pH-value of the bound solids of about 7.5 would seem to ensure the leaching resistance even after extended periods of storage. For adjusting the sodium-boron mole ratio of 0.25, only small amounts of sodium hydroxide are required. An advantage is that the storage of the residues can be carried out at temperatures of 20.degree. C.
046438688	claims	1. A support arrangement in combination with a nuclear reactor, which comprises at least one fuel cell module, at least one control drive mechanism, and at least one pressure vessel head, wherein said support arrangement is located between said fuel cell module and said control drive mechanism and is supported by said pressure vessel head, said support arrangement comprising: a module support nut, engaged with said pressure vessel head and supported therefrom, including a downwardly depending screw threaded portion, and a shroud housing for said fuel cell module including a screw threaded portion engaged with said screw threaded portion of said support nut such that said shroud housing is suspended from said support nut and thus from said pressure vessel head, said module support nut and said shroud housing including a locking means for locking said nut and housing against relative rotation. a module support nut, engaged with said pressure vessel head and supported therefrom, including a downwardly depending screw threaded portion, and a shroud housing for the fuel cell module including a screw threaded portion engaged with said screw threaded portion of said support nut such that said shroud housing is suspended from said support nut and thus from said pressure vessel head, and an installation tool assembly adapted to be inserted through the bore in the control drive mechanism for said fuel cell module, said assembly including a lifting and preloading tool for engaging and lifting said shroud housing and a concentrically mounted torquing tool for engaging said module support nut and applying torque thereto. 2. A support arrangement as claimed in claim 1 wherein said locking means includes a spring mount spline lock member associated with said shroud housing and a spline lock member associated with said support nut, wherein said spline lock member of said shroud housing is normally biased into engagement with said spline lock member of said support nut. 3. A support arrangement as c1aimed in claim 1 wherein said module support nut includes a further spline member adapted to engage a torquing tool for applying torque to said support nut. 4. A support arrangement as claimed in claim 1 wherein said shroud member includes a gripping flange adapted to be engaged by a lifting and preloading tool for providing lifting and preloading of said support arrangement. 5. A support and installation arrangement in combination with a nuclear reactor which comprises at least one fuel cell module, at least one control drive mechanism, and at least one pressure vessel head, wherein said support arrangement is located between said fuel cell module and said control drive mechanism and is supported by said pressure vessel head, said support arrangement comprising: 6. A support and installation arrangement as claimed in claim 5 wherein said lifting and preloading tool includes an expandable gripping portion for, when expanded, gripping the shroud housing, and a central rod member for controlling expansion of said expandable gripping portion. 7. A support and installation arrangement as claimed in claim 6 wherein said shroud housing includes an inwardly extending flange with a downwardly facing surface and said gripping portion includes an upwardly facing support surface which, in the use of said lifting and preloading too, is adapted to be engaged by said downwardly facing surface of said flange. 8. A support and installation arrangement as claimed in claim 7 wherein said lifting and preloading tool includes an elongate lifting member which includes said gripping portion, and hydraulic means for controlling longitudinal movement of said lifting member. 9. A support and installation arrangement as claimed in claim 8 wherein said hydraulic means includes a hydraulic piston formed integrally with said lifting member. 10. A support and installation arrangement as claimed in claim 5 wherein said torquing tool includes expandable gripping means for, when expanded, gripping said module support nut and means for applying a torque to said gripping means. 11. A support and installation arrangement as claimed in claim 10 wherein said lifting and preloading tool includes an elongate lifting member positioned concentrically within said expandable gripping means of said torquing tool and including means for expanding said expandable gripping means. 12. A support and installation arrangement as claimed in claim 11 wherein said gripping means comprises an expandable sleeve, said torque applying means comprises a handle connected to the upper end of said sleeve, and wherein said arrangement further includes a spacer nut concentric with said sleeve and in abutment with the upper end thereof. 13. A support and installation arrangement as claimed in claim 12 wherein said means for expanding said expandable gripping means comprises an enlarged portion of said elongate lifting member located intermediate the ends thereof and wherein said module support nut includes an inwardly extending spline portion for engagement by said expandable gripping means. 14. A support and installation arrangement as claimed in claim 13 wherein said module support nut and said shroud housing include spline lock means for locking said support nut and housing against rotation and wherein said gripping portion of said lifting and preloading tool provides disengagement of said spline lock means from said shroud housing. 15. A support and installation arrangement as claimed in claim 14 wherein said spline lock means includes a spring loaded spline locking member and said shroud housing includes a bolt member which is positioned to be engaged by said gripping portion of said lifting and preloading tool and which, when engaged by said gripping portion, contacts said spring loaded spline locking member and causes disengagement thereof from the module support nut.
049824198	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a Potter-Bucky device for use in a radiation image recording apparatus, and more particularly to a Potter-Bucky device which cannot transmit adverse vibrations to other parts of an apparatus in which it is employed. 2. Description of the Prior Art There has been in wide use a radiation image recording apparatus in which a radiation image of an object is recorded on an X-ray film by exposing the X-ray film to radiation such as X-rays which have passed through the object. Further, there has been known a radiation image recording and reproducing system in which a radiation image is recorded and reproduced by the use of a stimulable phosphor instead of the X-ray film. When certain kinds of phosphors are exposed to radiation such as X-rays, .alpha.-rays, .beta.-rays, .gamma.-rays, cathode rays or ultra-violet rays, they store part of the energy of the radiation. Then when the phosphor which has been exposed to radiation is exposed to a stimulating ray, such as a laser beam, light is emitted from the phosphor in proportion to the amount of radiation energy which was stored by the phosphor. A phosphor exhibiting such properties is referred to as a stimulable phosphor. In a radiation image recording and reproducing system, a radiation image of an object such as the human body is recorded and reproduced by the use of such a stimulable phosphor. Specifically, as disclosed, for instance, in U.S. Pat. No. 4,258,264 and Japanese Unexamined Patent Publication No. 56(1981)-11395, a recording medium bearing thereon a stimulable phosphor layer is first exposed to radiation which has passed through an object in order to store a radiation image of the object in the stimulable phosphor layer, and then the stimulable phosphor layer is two-dimensionally scanned with a stimulating ray which causes it to emit light in a pattern corresponding to the stored radiation image. The light emitted from the stimulable phosphor layer upon stimulation thereof is photoelectrically detected and converted into an electric image signal, which is processed to reproduce the radiation image as a visible image on a recording medium such as a photosensitive material, a display system such as a CRT, or the like. This system is advantageous over conventional radiography which uses X-ray film in that a radiation image can be recorded over a much wider radiation energy exposure range. That is, it has been found that the intensity of light emitted from the stimulable phosphor upon stimulation thereof after it is exposed to radiation remains proportional to the energy of the radiation to which it was exposed for a very wide radiation energy intensity range. Accordingly, even if the energy intensity range of the radiation to which the stimulable phosphor is exposed varies substantially due to changes in the recording conditions, a visible radiation image independent of variations in the radiation energy intensity range can be obtained by choosing an appropriate gain when converting the light emitted from the phosphor into an electric signal. In radiation image recording systems in which X-ray film or a stimulable phosphor layer (both generically referred to as an image recording medium) is exposed to radiation which has passed through an object so as to record a radiation image of the object, a grid is sometimes disposed between the object and the image recording medium. That is, when a relatively thick part of an object, such as the chest of a human body, is radiographed, diffused radiation emitted from the object upon radiographing can deteriorate the quality of the radiation image obtained. In such cases, a grid device for absorbing the diffused radiation should be provided between the object and the recording medium. As is well known, a grid device comprises lead foils or the like arranged in parallel or in a grid and is disposed so as to overlap with the whole image recording area of the image recording medium. When such a grid device is kept stationary during the recording of a radiation image, fine stripes corresponding to the foils of the grid are projected onto the recording medium together with the radiation image of the object. Since the stripes are very fine, the stripes are almost invisible when the radiation image is recorded on X-ray film. However, when the stimulable phosphor layer is used to record the radiation image, the stripes are visible as moire fringes since the stimulable phosphor is more sensitive than the X-ray film and the radiation image stored in the stimulable phosphor layer is read out by scanning the stimulable phosphor layer with a light beam at very fine pitches. Particularly, in radiation image recording systems in which a stimulable phosphor layer is utilized and subtraction processing of image signals is involved, such as the systems disclosed in U.S. Pat. Nos. 4,710,875, 4,590,517 and Japanese Unexamined Patent Publication No. 58(1983)-163339, and the like, the stimulable phosphor layers storing therein different radiation images such as ones recorded before and after the infusion of a contrast medium, or digital image signal tapes bearing thereon image signals read out from the stimulable phosphor layers are subjected to translation processing and/or rotation processing in order to correct fluctuation in the position of the radiation images recorded on the stimulable phosphor layers. The fine stripes recorded on the stimulable phosphor layers interfere with each other during translation and rotation processing in such a way that they produce moire fringes in the finally reproduced image, thereby very adversely affecting diagnoses based on the reproduced image. As disclosed in EP-0114978, there has been proposed a Potter-Bucky device having a driving means for driving the grid back and forth at a high speed in parallel to the stimulable phosphor layer. This arrangement prevents part of the recording medium from being shielded by the foils of the grid during image recording, thereby preventing the formation of stripes on the recording medium and the production of moire fringes in the radiation image which has been read out, even if the recording medium uses a stimulable phosphor layer. However there is a problem in that, though the Potter-Bucky device can prevent production of the moire fringes, it is apt to transmit vibrations to other parts of the system due to the movement of the grid. Particularly, in the case of a so-called built-in type radiation image recording and reproducing system, in which an image recording device, an image read-out device and an erasing device are incorporated into a single unit and the recording medium having the stimulable phosphor layer is conveyed or circulated through the system, when vibrations are transmitted from the Potter-Bucky device in the image recording device to the image read-out device, if the image read-out device is carrying out the image read-out operation, the accuracy of the read-out operation deteriorates. SUMMARY OF THE INVENTION In view of the foregoing observations and description, the primary object of the present invention is to provide a Potter-Bucky device in which no vibration is transmitted to parts of an apparatus around the device even if the grid is moved. The Potter-Bucky device in accordance with the present invention comprises a grid which is disposed in a radiation image recording apparatus for exposing an image recording medium to radiation which has passed through an object in order to store a radiation image of the object on the recording medium and is supported by a support means between the object and the recording medium, and a grid driving means which reciprocates the grid parallel to the recording medium, and is characterized by having a balancer, which is connected to the support means and is reciprocated in synchronization with said grid but in the opposite direction, thereby compensating for displacement of the center of gravity of the Potter-Bucky device due to the reciprocation of the grid. With this arrangement, the vibration caused by the displacement of the center of gravity of the Potter-Bucky device due to the movement of the grid is compensated for by the movement of the balancer, and accordingly, no vibration is transmitted to parts of the apparatus around the Potter-Bucky device.
	summary
058964290	abstract	A method and apparatus is disclosed for inspecting a wall to evaluate the remaining thickness of the wall as well as the extent to which the wall has been infiltrated by another material. The disclosure discusses directing photons of radiation and/or neutrons into the wall and measuring and analyzing the radiation emitted from the wall as a result of Compton scattering, pair production, photoelectric absorption and/or neutron absorption. The invention is particularly well suited for inspecting a carbon hearth-wall liner of an iron-smelting blast furnace.
RE0315834	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring first to FIG. 1 there is illustrated the reactor pressure vessel 10 for a pressurized water reactor. Pressure vessel 10 extends generally in a vertical direction and has coolant inlet means such as nozzle 12 and coolant exit means such as nozzle 14 which provide entry and egress for coolant, in this instance water, passing through the reactor. A stainless steel core support barrel 16 is rigidly attached to and supported by pressure vessel 10. Core support barrel 16 contains the reactor core 18 supported upon core support assembly 20. Core support assembly 20 is rigidly affixed to core support barrel 16 at or near the lower end thereof and includes at its upper extent core support plate 22. An upper guide structure 24 is also contained in the core support barrel 16 and is rigidly affixed to the barrel above core 18, usually at the upper end of core support barrel 16. Upper guide structure 24 principally houses shrouds 26 for control element assemblies 28 and includes at its lower extent the upper core alignment plate 30. The shrouds 26 are joined to alignment plate 30 to form an integral structure with guide structure 24. Reactor core 18 is comprised of a plurality of vertically extending nuclear fuel assemblies 32 arranged in a substantially circular geometry. A typical reactor contains more than 200 of such fuel assemblies 32. The coolant flow path within pressure vessel 10 as indicated by arrows 34 is from inlet nozzle 12 downwardly between pressure vessel 10 and core support barrel 16 into the area of core support assembly 20 and upwardly through various coolant openings in core support plate 22 through the fuel assemblies 32 in core 18, and ultimately out through outlet nozzle 14. A typical fuel assembly 32, seen more clearly in FIG. 2, is comprised of five vertically extending parallel Zircaloy guide tubes 36 coextensive with one another and rigidly attached to upper and lower end fittings 38 and 40, respectively. Guide tubes 36 provide the vertical structural framework for fuel assembly 32. A plurality of rectangular Zircaloy spacer grids 42 are positioned at various elevations along guide tubes 36 and are welded thereto. Fuel rods 44 extend in parallel vertical arrangement within fuel assembly 32 and their pitch over the full length of the fuel assembly is maintained by spacer grids 42. A large number of fuel rods 44, for instance 176, are individually retained by compartments in the several spacer grids 42. A retention grid 46 welded to the upper portion of the lower and fitting 40 consists of spring strips interlocked in egg crate fashion and welded to perimeter strips. Overlapping spring fingers, formed within the spring strip, engage a machined groove in the lower end of each fuel rod 44. In this manner all rods 44 are both axially and laterally restrained. Lower end fitting 40 is comprised essentially of a lower end plate 48 having alignment posts 50 mechanically secured thereto and depending downwardly therefrom. Alignment posts 50 fit slideably into holes in the core support plate 22 and provide the necessary vertical support and lateral alignment of the lower end of the fuel assembly 32. In some core designs, the core support plate has the alignment posts affixed thereto and the posts are slideably engaged by holes in the fuel assembly lower end fitting. The length of alignment posts 50 is sufficient to ensure, that even in the event of maximum possible upward lifting of fuel assembly 32 during an accident, some part of alignment post 50 will be retained by core support plate 22 and prevent lateral movement of the fuel assembly. In order to facilitate installation and removal of each fuel assembly, the alignment posts 50 are not captured in a way which restricts their vertical movement. The upper end of each fuel assembly 32 includes an upper end fitting 38 which is mechanically secured to guide tubes 36 in a manner to be discussed more thoroughly below. Included within upper end fitting 38 are an upper end plate 52 extending transversely of the vertically extending guide tubes 36 and alignment means, such as upwardly extending alignment posts 54. As earlier mentioned an upper core alignment plate 30 affixed to core support barrel 16 extends horizontally of the core support barrel and is located slightly above the end plates 52 of the fuel assemblies 32 making up core 18. Holes in alignment plate 30 are positioned and sized to provide close slideable engagement of the alignment posts 54. In this manner alignment plate 30 provides the alignment of the upper ends of the fuel assemblies 32. In the preferred embodiment of the invention, alignment posts 54 are axially co-incident with guide tubes 36 and have passageways extending axially therethrough which are co-incident with the interior passages of the guide tubes. The control element assemblies 28 of the present embodiment each comprise five parallel elongated downwardly extending control fingers which are insertable into and withdrawable from the fuel assemblies 32 by way of guide tubes 36. As seen in FIG. 2, the alignment plate 30 has a single large hole extending therethrough at each control element assembly location. The hole is sized and shaped to allow total insertion of a control element assembly 28 into a fuel assembly 32 while laterally restraining the outer most of alignment posts 54. Because not all fuel assemblies 32 have control element assemblies 28, the alignment holes in plate 30 at those locations need not extend entirely through alignment plate 30 and each alignment post 54 may be received in an individual hole, as seen in FIGS. 5a and 5b. The distance between lower core support plate 22 and upper core alignment plate 30 is such that a gap or spacing "d" will exist between upper core alignment plate 30 and upper end plate 52. Gap "d" exists to accommodate mechanical tolerances and the thermal expansion differential exhibited between fuel assembly 32 and core support barrel 16 in the vertical direction. In the present instance wherein guide tubes 36 are more than 13 feet in length and are of Zircaloy with core support barrel 16 being of stainless steel, the gap d between end plate 52 and alignment plate 30 can vary 5/8 of an inch or more between the cold ane hot extremes of reactor operation. The slideable engagement of alignment posts 50 by the holes in core support plate 22 permit some upward movement of an entire individual fuel assembly 32 in the event that the hydraulic lifting forces of the coolant moving upwardly through core 18 are greater on the fuel assembly 32 than the weight of the fuel assembly. Even though a fuel assembly 32 might weigh as much as 1400 lbs., the upward hydraulic forces directed thereagainst in present reactors, particularly when the coolant is cold and has greatest density, is often sufficient to lift the fuel assembly upward from core support plate 22 and into contact with upper alignment plate 30. In order to prevent such lifting, the fuel assembly hold-down device of the invention is employed. According to the invention, the fuel assembly hold-down device includes coil spring means, such as coil springs 56, in compression and acting between alignment plate 30 and fuel assembly upper end plate 52 to provide a downward force against fuel assembly 32. The coil spring means, unlike an inherently low deflection device such as a leaf spring, is capable of a much greater range of deflection than a leaf spring of comparable size. Because size, economy and simplicity are important in providing an effective hold down for the fuel assemblies the coil spring means are preferred in providing the hold-down function. According to the preferred embodiment and as seen in FIG. 3, five alignment posts 54 axially co-incident with the five guide tubes 36 are arranged such that one alignment post and guide tube extend axially through the center of fuel assembly 32 and the other four are spaced radially outward therefrom and extend parallel thereto. Alignment posts 54, upper end plate 52 and guide tubes 36 are mechanically joined to form an integral structure. Upper end plate 52, seen most clearly in FIG. 3, is a cast metal structure having coolant openings 58 therein through which coolant exhausts from the fuel assembly. Additionally, five holes extend through the plate at the locations where the guide tubes 36 and the alignment posts 54 are joined to plate 52. As seen in FIG. 4, the lower ends of stainless steel guide posts 54 are pressed into the proper holes in end plate 52. The inner wall of alignment post 54 is recessed and threaded near its lower end and forms shoulder 59. The upper end of guide tube 36 is enlarged radially outwardly forming shoulder 61 for axial mating engagement with shoulder 59 in alignment post 54. A nut 62 disposed about guide tube 36 is inserted upwardly between guide tube 36 and alignment post 54 and is threaded into engagement with the threads in post 54 to maintain the axial engagement therebetween. Locking means 64 are affixed, as by welding to end plate 52 and engage nut 62 in a manner preventing unthreading rotation of the nut. This arrangement provides a unitary structure between alignment posts 54, end plate 52 and guide tubes 36. The central alignment post 54, because of the difficulty of access thereto may be threaded into plate 52 and guide tube 36 slideably engaged therewithin. Each of the alignment posts 54 is generally cylindrical in shape and extends several inches above upper end plate 52. The upper end portion of each alignment post 54, generally about 1 inch, is radially enlarged to form shoulder 66, the function of which is discussed below. The alignment hole or holes in alignment plate 30 are sized to accommodate close sliding engagement of the radially enlarged portion of each alignment post 54. The coil springs 56 used to provide the hold-down forces on the fuel assembly are coaxially disposed about one or more of the alignment posts 54. While it might be possible in some arrangements to use only one coil spring 56 about the central alignment post 54, a more desirable arrangement is obtained wherein coil springs 56 are coaxially disposed about each of the four outer alignment posts 54 and act against hold-down plate 68. This arrangement is only one of several possible but provides a symmetrical placement of the springs and allows each spring to be sized to provide only one-fourth of the overall downward force needed. Moveable hold-down plate 68 is slideably mounted on alignment posts 54 and is capable of movement between end plate 52 and the shoulder 66 on the alignment posts. Shoulders 66 restrict the upward travel of hold-down plate 68 relative to fuel assembly 32, making the plate an integral part thereof. The hold-down plate 68 of the invention includes hub portions 70 with apertures 72 extending vertically therethrough and sized to permit slideable passage therethrough of said alignment posts 54. The several hub portions 70 have legs 74 extending radially outward from them in a horizontal direction, and in the preferred embodiment legs 74 serve to interconnect hub portions 70 making up hold-down plate 68. The radially outer most portions of a hold-down plate 68 extend sufficiently beyond the alignment hole or holes in plate 30 to engage the lower surface of plate 30. Because hold-down plate 68 additionally serves as a lifting surface in handling fuel assembly 32 as will be discussed more thoroughly below, these lifting surfaces are preferably symmetrically disposed about the vertical center line of gravity of fuel assembly 32. The coil springs 56 which serve to provide the required hold-down forces are coaxially disposed about various ones of the alignment posts 54 between upper end plate 52 and the hold-down plate 68. Springs 56 are sized and preloaded to ensure that a net downward force of 150 lbs. will be maintained on fuel assembly 32 for all normal and anticipated transient flow and temperature conditions. The cyclic loads on the spring are minimal since the spring operates over an extended range, as seen in FIGS. 5a and 5b, only during reactor startup and shutdown conditions. In the example of the preferred embodiment, each spring 56 is fabricated of Inconel X750, has a free length of about 4.5 inch, an internal diameter of 1.5 inch, an external diameter of 1.8 inch, a wire size of 0.135 inch, and has 81/2 coils yielding a spring rate of 17 lbs. per inch. As disposed about alignment posts 54, each spring 56 is held captive in the event of a spring fracture at almost any point in the spring. As seen more clearly in FIGS. 5a and 5b, the hold-down plate 68 which is upwardly biased by coil springs 56 acts against the under side of upper core alignment plate 30 resulting in a downward force against upper end plate 52 and accordingly fuel assembly 32. During the fuel loading operation the various fuel assemblies 32 are loaded into the reactor and the upper guide structure 24 which includes upper alignment plate 30 rigidly affixed thereto, is then lowered into place and rigidly joined to core support barrel 16 at or near the top thereof. The vertical positioning of alignment plate 30 provides a gap at cold conditions between alignment plate 30 and upper end plate 52 having a distance, d.sub.1, The gap distance is smallest at reactor cold conditions and d.sub.1 will normally be about 2.4 inches. As seen in FIG. 5a, this gap distance is such that hold-down plate 68 is moved downwardly along alignment post 54 against the upward force of springs 56 resulting in a significant downward force against fuel assembly 32. This is the preload mentioned earlier. As the temperature in the core 18 increases during reactor operation, the differences between the linear thermal expansions of fuel assembly 32 and core support barrel 16 result in an increased gap distance, d.sub.2, and accordingly the spring biased hold-down plate 68 which acts upwardly against alignment plate 30 is permitted to slide upwardly along alignment posts 54 to the extent necessary. This somewhat reduces the downward force which the springs exert against fuel assembly 32, however, as previously mentioned the initial preload on springs 56 at the time of installing guide structure 24 is such that, even when maximum gap distance d.sub.2 occurs, there results a net downward force on the fuel assembly of about 150 lbs. The axial length of alignment post 54 above upper end plate 52 is sufficient to ensure that the alignment post will always be laterally engaged by alignment plate 30. Further still, the radially enlarged shoulder 66 of each of the several alignment posts 54 is sufficiently distant from upper end plate 52 that even for a maximum gap distance, d.sub.2, which occurs when the reactor is hot, the upper core alignment plate 30 will continue to exert a downward force against hold-down force against hold-down plate 68 and accordingly on fuel assembly 32. The shoulders 66 on alignment posts 54 restrict the upward travel of hold-down plate 68, as mentioned above, with the result that hold-down plate 68 forms a lifting surface for fuel handling means during fueling operations. A typical fuel assembly coupling member 76, as seen in FIG. 6, is employed to lift and move individual fuel assemblies 32 during fueling and refueling operations. Fuel coupling member 76 is usually positioned at the lower end of a rotatable and translatable shaft. Those fuel assemblies 32 which receive control element assemblies will normally be moved with the control elements fully inserted thereinto. This will dictate the number and geometry of slots 78 in coupling member 76. Typically, the upper or head portion of a control element assembly 28 appears to coupling member 76 as a vertical extension of hold-down plate 68. Accordingly, slots 78 are spaced at 90.degree. intervals about the cylindrical coupling member 76 and are axially long enough to receive hold-down plate 68 and a control element head portion, if present. The coupling member 76 is positioned above a fuel assembly 32 with slots 78 aligned with the legs 74 on hold-down plate 68. Coupling member 76 is then moved downwardly over legs 74 and rotated and raised slightly to supportingly engage legs 74 in seats 80. In this manner the fuel hold-down device of the invention serves also as the lifting surface for the fuel assembly during handling thereof. It should be appreciated that, while the fuel assembly hold-down device of the preferred embodiment is affixed to each fuel assembly and may also be used as a handling surface, the scope of the invention would also include a similar hold-down device employing coil springs wherein the alignment posts extend downwardly from the upper core alignment plate and into holes in the fuel assembly upper end fitting. While we have illustrated and described a preferred embodiment of the invention, it is to be understood that such a merely illustrative and not restrictive and that variations and modifications may be made therein without departing from the spirit and scope of the invention. We, therefore, do not wish to be limited to the precise details set forth but desire to avail ourselves of such changes as fall within the purview of the invention.
051204878	summary	BACKGROUND OF THE INVENTION This invention relates to measurement of plasma parameters in a tokamak. In particular, this invention departs from the prior art by enabling measurement of the dc toroidal electric field E in a tokamak plasma. Using this invention, a brief, deliberate perturbation of hot tokamak electrons produces a transient synchrotron radiation signal, in frequency-time space, and plasma parameters including the dc electric field can be inferred from the radiation response. The use of synchrotron emission to deduce plasma properties is an established and important technique. Generally, transient synchrotron emission is used for information on the electron temperature; recently there have been attempts to uncover further details of the electron momentum distribution function f. Prior art methods are limited, however, in that deduction of details of the electron distribution function is based on the synchrotron emission from the entire distribution of electrons; consequently, only one-dimensional data (in frequency) can be used to constrain f. Measurement of the dc parallel electric field has been unavailable using methods of the prior art. Typically less than a volt per meter in a tokamak, this field is far too small to be inferred through atomic phenomena, and cannot be measured directly by probes because the plasma is too hot. Its effect is manifest, however, in the dynamics of superthermal electrons--those that synchrotron radiate most profusely. It is the primary object of this invention to provide a method for measurement of the dc toroidal electric field E in a tokamak plasma. In the accomplishment of the foregoing object, it is another important object of this invention to provide a method for inferring other parameters of tokamak plasma from the transient radiation response produced by brief perturbation of the plasma. It is another important object of this invention to provide a method for distinguishing the steady dc electric field from noise. It is a further object of this invention to present a method for comparing parameter sets that might possibly explain an incremental transient signal and for estimating the informative worth of the data prior to obtaining it. Additional objects, advantages and novel features of the invention will become apparent to those skilled in the art upon examination of the following and by practice of the invention. SUMMARY OF THE INVENTION To achieve the foregoing and other objects, this invention comprises a method including a brief, this deliberate perturbation of hot tokamak electrons which produces a transient synchrotron radiation signal, in frequency-time space, and the inference, using very fast algorithms, of plasma parameters including the effective ion charge state Z.sub.eff, the direction of the magnetic field, and the position and width in velocity space of the brief heating, and, in particular, the dc toroidal electric field. In addition, this invention includes a method for comparing essentially all parameter sets that might possibly explain the transient signal, and, by simulating data, for estimating the informative worth of data prior to obtaining it.
	description	The present subject matter of the teachings described herein relates generally to a method and system for collecting 3He gas from heavy water moderated and/or cooled nuclear reactors. 3He is an isotope of helium with applications in many different industries. 3He can be formed by beta decay of tritium. One known source of 3He gas is the decay tritium in nuclear weapons. Another source of tritium is the irradiation of tritium producing burnable absorber rods (TPBARs) within light water nuclear reactors. Another source of tritium is heavy water nuclear reactors. Heavy water includes deuterium. Heavy water reactors, for example reactors that use heavy water as a moderator, coolant or both, may produce tritium as a result of thermal neutron activation of the deuterium in the heavy water. The heavy water can be detritiated, the tritium can be collected and 3He may be obtained as the tritium decays. Demand for 3He may exceed the supply of 3He from known production and/or collection methods. 3He may be commercially valuable. Therefore, there remains a need for an alternative apparatus and/or system for directly collecting 3He. This summary is intended to introduce the reader to the more detailed description that follows and not to limit or define any claimed or as yet unclaimed invention. One or more inventions may reside in any combination or sub-combination of the elements or process steps disclosed in any part of this document including its claims and figures. In accordance with one broad aspect of the teachings described herein, a method of collecting 3He from a nuclear reactor may include the steps of a) providing heavy water at least part of which is exposed to a neutron flux of the reactor, b) providing a cover gas in fluid communication with the heavy water, c) operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by β− decay and at least a portion of the 3He gas escapes from the heavy water and mixes with the cover gas, d) extracting an outlet gas stream, the outlet gas stream comprising a mixture of the cover gas and the 3He gas and e) separating the 3He gas from the outlet gas stream. The method may also include outputting a 3He gas stream for further processing and may include treating the outlet gas stream to provide a treated cover gas stream. The method may include mixing at least a portion of the treated cover gas stream into the cover gas in fluid communication with the heavy water. The step of extracting the outlet gas stream may be performed while nuclear reactor is operating and the outlet gas stream may be extracted as a generally continuous stream while nuclear reactor is operating. The step of separating the 3He gas from the outlet gas stream may be an on-line process that is performed while the nuclear reactor is operating. The step of separating the 3He gas from the outlet gas stream may include at least one of a thermal diffusion process, a fractional diffusion process, a heat flush process, a superleak process and a differential absorption process. The cover gas may contact the heavy water at a free surface interface. A method of collecting 3He from a nuclear reactor may include the steps of a) providing heavy water at least part of which is exposed to a neutron flux of the reactor, b) operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by β− decay and at least a portion of the 3He gas escapes from the heavy water, c) extracting an outlet gas including the 3He gas, and d) optionally, separating the 3He gas from any other gas in the outlet gas stream. According to another broad aspect of the teachings described herein, a system for collecting 3He may include a nuclear reactor having a vessel containing a heavy water and having a cover gas head space containing a cover gas above the heavy water. The reactor may have a gas outlet in communication with the cover gas head space. Operation of the nuclear reactor may result in thermal neutron activation of deuterium in the heavy water to produce tritium (3H) and at least some of the tritium may undergo β− decay to produce 3He gas that mixes with the cover gas. A gas extraction passage may be fluidly connected to the gas outlet of the vessel to extract a gas outlet stream through the gas outlet. The gas outlet stream may include the cover gas and the 3He gas mixed with the cover gas. A 3He separation apparatus may be fluidly connected to the gas extraction passage downstream gas outlet and may be operable to receive the gas outlet stream and separate the 3He gas from the cover gas. A gas inlet may be provided in the vessel and in communication with the cover gas head space. A cover gas supply passage may be coupled to the gas inlet of the vessel to supply the cover gas to the cover gas head space. The 3He separation apparatus may include a 3He outlet to output a separated 3He gas stream and a separate treated cover gas outlet to output a treated cover gas stream. The treated cover gas outlet of the 3He separation apparatus may be fluidly connected to the cover gas supply passage to re-introduce at least a portion of the treated cover gas stream into the cover gas head space. The gas outlet stream may be extractable as a generally continuous gas stream while the nuclear reactor is in operation. The cover gas provided above the heavy water may consist essentially of 4He. The 3He separation apparatus may include at least one of a thermal diffusion apparatus, a fractional diffusion apparatus, a heat flush apparatus, a superleak apparatus and a differential absorption apparatus. According to yet another broad aspect of the teachings described herein a moderator cover gas system for use with a nuclear reactor having a vessel containing heavy water may include a cover gas supply passage having a gas outlet connectable to a gas inlet on the vessel to supply a cover gas into the vessel. A gas extraction passage may have a gas inlet connectable to a gas outlet on the vessel to extract an outlet gas stream from within the vessel. The outlet gas stream may include a mixture of at least the cover gas and 3He gas. A gas separation apparatus may be connected to the cover gas flow passage downstream from the gas outlet on the vessel and operable to separate the 3He gas from the outlet gas stream. A fresh cover gas source may be fluidly connected to the cover gas supply passage to introduce cover gas consisting essentially of 4He into the interior of the vessel. The gas separation apparatus may include a first outlet to output the separated 3He gas and a second outlet to output a treated cover gas stream. The second outlet may be fluidly connected to the cover gas supply passage to feed at least a portion of the treated cover gas stream into the cover gas supply passage. Elements shown in the figures have not necessarily been drawn to scale. Further, where considered appropriate, reference numerals may be repeated among the figures to indicate corresponding or analogous elements. Various apparatuses or processes will be described below to provide an example of an embodiment of each claimed invention. No embodiment described below limits any claimed invention and any claimed invention may cover processes or apparatuses that differ from those described below. The claimed inventions are not limited to apparatuses or processes having all of the features of any one apparatus or process described below or to features common to multiple or all of the apparatuses described below. It is possible that an apparatus or process described below is not an embodiment of any claimed invention. Any invention disclosed in an apparatus or process described below that is not claimed in this document may be the subject matter of another protective instrument, for example, a continuing patent application, and the applicants, inventors or owners do not intend to abandon, disclaim or dedicate to the public any such invention by its disclosure in this document. Helium−3 (3He) is an isotope of helium, with 4He being the most common isotope of helium by a large factor. 3He has applications in a variety of industries including, for example, the nuclear safeguard, security, medical and oil and gas industries. For example, 3He can be used in neutron detector apparatuses that can be used to detect nuclear and radiological materials. Such neutron detector apparatuses may be used at border crossings, ports, airports and other points of entry into a country in an attempt to help detect smuggled and/or concealed nuclear material. In other examples, 3He may be used in combination with magnetic resonance imaging (MRI) to help provide visualization of a patient's lung capacity and function and/or may be used to help determine the rock porosity and/or presence of hydrocarbon reserves in the oil and gas industry. In the construction industry, neutron detectors utilizing 3He may be used to measure soil compaction and moisture content. 3He may also be used to obtain low refrigeration temperatures via dilution refrigeration. 3He gas can be produced by the decay of the radioactive isotope tritium (3H), which has a half life of 12.3 years. One source of 3He is tritium found in thermonuclear warheads. As the tritium decays it produces 3He. Tritium has also been produced through neutron irradiation of 6Li-containing tritium-producing-burnable-absorber rods (TPBARs) in light water nuclear reactors. However, quantities of tritium produced in this manner, and the resulting quantities of 3He produced by the decay of the tritium, may not be sufficient to satisfy 3He demand. As the demand for neutron detectors and other commercial uses of 3He gas increases the demand for 3He will also increase. Conventional sources of 3He, such as harvesting 3He from decaying tritium in nuclear warheads, may not be sufficient to meet increased 3He demands. Some current estimates suggest that the annual global demand for 3He gas now exceeds the current annual supply of 3He gas. For example, while there is a relatively small amount of data regarding the current use of and/or demand for 3He, is the inventors estimate that the production of 3He gas in the United States may be approximately 8,000 L/year, while the global demand for 3He gas is estimated to be about 65,000 L/year, or 65 m3/year. Accordingly, the inventors have identified a need for an alternative method of harvesting or collecting 3He gas. The inventors have discovered that heavy water nuclear reactors may be one viable source from which 3He gas may be directly extracted or collected. It has been discovered that potentially useable amounts of 3He are produced within the heavy water contained the reactors, as either a moderator, coolant or both, and that this 3He can be directly harvested or extracted from the reactors without first separating, collecting and/or storing tritium outside the reactor. This direct extraction of 3He may be used as an alternative to, or in combination with known tritium collection processes. Some examples of heavy water reactors include pressurized heavy water reactors (such as Canada Deuterium Uranium (CANDU™) reactors), reactors including a heavy water moderator, reactors that use heavy water as a coolant and reactors that use heavy water as both a coolant and a moderator. Whether utilized as a moderator and/or a coolant, heavy water that is present within the nuclear reactor may be subject to thermal neutron activation to produce tritium, and decay of such tritium may form 3He. Some commercial heavy water power reactors, such as CANDU™ reactors use heavy water (D2O) in the moderator and heat transport systems. In such reactors, the moderator may be contained within a calandria vessel, and a cover gas, such as a moderator cover gas, is provided within the calandria in fluid communication to with free surface(s) of the moderator. Heavy water free surfaces may be present within the calandria vessel, and may also be present at one or more locations in other process piping, vessels and other portions of a moderator system. The space above some or all of these free surfaces is filled with the moderator cover gas via a cover gas system. The free surfaces of the moderator heavy water do not need to be in communication with each other, and the moderator system may include multiple discrete regions in which moderator cover gas is in contact with a free surface of the heavy water. Optionally, all of the moderator cover gas can be circulated within a common cover gas system. Typically, the moderator cover gas is substantially pure helium gas (4He). For example, the cover gas may be at least 85% 4He and be at least 90%, at least 95% and/or at least 99% 4He by volume when the reactor is in use. When the reactor is in operation, one or more other gases and/or impurities may accumulate within the cover gas. For example, D2 may be produced from radiolysis of the heavy water and may collect in the cover gas. Similarly, small amounts of O2 (for example from radiolysis of heavy water, from O2 added to promote D2-O2 recombination and/or from air leaks), N2 (for example from air leaks), CO2 possibly containing trace 14C (for example produced as an activation product of 17O in the moderator), 41Ar (an activation product of 40Ar which may be an impurity in the helium cover gas) and other gases, such as T2 and DT may also accumulate in the cover gas. A moderator cover gas system is provided to circulate the cover gas within the reactor and within each gas head space, and may include any suitable components and/or apparatus to help control the concentration of impurities in the cover gas or otherwise process the cover gas, including, for example, cover gas preheaters, recombination units, scrubbers, catalytic converters and flame arresters. Optionally, the cover gas can be processed to help control the concentration of impurities in the helium cover gas within desirable design limits (for example, less than about 3% H2 by volume, less than about 2% O2 by volume, and less than about 4% D2 by volume). The cover gas system may also include one or more sources of fresh, pure helium, including, for example, helium bottle stations. CANDU™ and other heavy water reactors may generate tritium (3H) in the heavy water systems as a waste by-product during operation (for example, when used to generate electrical power). For example, tritium (3H) may be produced within the moderator through thermal neutron activation of deuterium (2H), via 2H(n,γ)3H, and in the heat transport system (or coolant) of the heavy water reactor. The neutron radiative capture reaction 2H(n,γ)3H is believed to be the dominant method of tritium production in heavy water reactors. Some existing heavy water reactor facilities are configured to extract tritium from the heavy water used in the moderator and heat transport systems, for example using heavy water detritiation plants, to help reduce operator dose and environmental emissions. In such installations, the elemental tritium removed from the pressurized heavy water heat transport systems can be stored as titanium tritide in stainless steel storage vessels as a waste product. Eventually, the tritium will decay producing 3He gas. Such storage vessels may or may not include mechanisms for off-line recovery of the 3He gas. As these known processes include separating and storing tritium from the reactors and then harvesting 3He from the stored tritium, they may be referred to as in-direct 3He extraction processes. While some 3He gas may be produced by the decay of tritium extracted from waste water storage tanks, the inventors believe that when heavy water reactors are in use a potentially useable quantity of 3He gas can be produced within the heavy water moderator and/or the heavy water coolant, and that at least some of the 3He gas can be directly extracted from these systems within the heavy water reactor (i.e. without first harvesting and/or storing tritium) For example, the inventors believe that 3He may be produced within the moderator and that at least some of the 3He present in the moderator can escape the liquid (e.g. via diffusion and/or via bubbling up to the free surfaces surface) and may collect in a moderator cover gas that is provided over the free surfaces of the moderator. The heavy water moderator may be contained at relatively low pressures (relative to the heavy water used in the heat transport system), and may be at approximately atmospheric pressure. Alternatively, or in addition, 3He may be produced within the heavy water coolant, and may escape the coolant and be collected in a cover gas provided in the heat transport system, including, for example the pressurized cover gas contained in the coolant pressurizer. Instead of, or in addition to, removing and treating tritium-carrying heavy water from the heat transport system of the reactor, the inventors have discovered that at least a portion of the moderator cover gas, and optionally the coolant cover gas, can be extracted from reactor and can be treated or processed to separate the 3He gas from the cover gas. Extracting and processing the moderator cover gas and/or coolant cover gas may help collect at least some of the 3He gas produced in the moderator liquid. Conventional methods of collecting and storing tritium-carrying heavy water from the moderator or heat transport systems do not capture the 3He gas that is directly released from the moderator liquid and coolant and accumulates in the cover gases. In some configurations, extracting 3He from the moderator cover gas system may be more desirable than extracting 3He from the coolant cover gas because the heat transport system is an important safety system and it may not be desirable to modify or interfere with such a system. The separated cover gas and/or extracted 3He gas can then be stored and/or sent for further processing. Separating 3He gas from the cover gas may, in some instances, be more desirable than processing waste heat transport heavy water and/or may help facilitate capture and collection of the 3He gas that is generated within the reactor and escapes from the heavy water prior to the collection, processing and/or storage of the heavy water. For example, in heavy water reactors, 3He gas may be formed in the moderator as a result of tritium β− decay. The 3He gas formed may also be converted back to tritium via the reaction 3He(n,p)3H, in the moderator. The thermal neutron absorption cross-section of the 3He(n,p)3H reaction, in which the product of the tritium β− decay is converted back to tritium, is believe to be about seven orders of magnitude greater than the cross-section for the reaction 2H(n,γ)3H. However, it is believed that because of the low solubility of 3He gas in the heavy water moderator, at least a portion of the 3He gas formed in the moderator may escape irretrievably into the moderator cover gas provided above the moderator, before this back conversion reaction can occur. The 3He gas then mixes with the 4He forming the cover gas. Because it has been found that at least some 3He gas escapes the moderator, it is believed that the residence time of the 3He gas in the moderator may be too short to convert a significant amount of the 3He gas formed in the moderator back into tritium. Therefore, it is believed that there is the potential to recover useable and possibly commercially significant quantities of 3He gas formed in the moderator of CANDU reactors by extracting and processing the moderator cover gas. An example of an estimate of 3He gas production in a typically CANDU reactor is set out below. If one assumes that the extent of the 3He(n,p)3H reaction is negligible compared to the rate of 3He formation by tritium β− decay, then the rate of 3He production in the moderator is given by the rate of tritium β− decay. Consequently, the 3He production rate (atoms·s−1) can be written as: ⅆ N 3 He ⅆ t = λ ⁢ ⁢ MN T ( 1 ) Where:M=Total heavy water inventory in the Moderator (kg),N3He=Total number of 3He atoms in the moderator at time t (atoms),NT=Number of tritium atoms in the moderator per kgD2O at time t (atoms·kg−1),t=Time (s), andλ=Tritium decay constant (s−1). To solve Equation 1, an estimate of the tritium concentration in the moderator as a function of time, t, is required. This can be done by using an example a mass balance for tritium in the moderator: ⅆ ⅆ t ⁢ ( MN T ) = φ ⁢ ⁢ σ ⁢ ⁢ N D ⁢ ma + F M ⁢ N 0 - λ ⁢ ⁢ MN T - L R ⁢ N T ( 2 ) ⅆ N T ⅆ t + N T ⁡ ( λ + L R M ) = φ ⁢ ⁢ σ ⁢ ⁢ N D ⁢ m M ⁢ a + F M M ⁢ N 0 ( 3 ) Where:FM=Make-up heavy water flow (kg·s−1),LR=Heavy water loss rate (kg·s−1),ND=Number of deuterium atoms in the moderator per kgD2O (atoms·kg−1),N0=Number of tritium atoms in the make-up heavy water per kgD2O (atoms·kg−1),a=Reactor capacity factor,m=Heavy water inventory under the neutron flux (kg),t=time (s),φ=Thermal neutron flux (neutrons·cm−2·s−1), andσ=Thermal neutron absorption cross-section (cm2) Equation 3 is derived based on the simplifying assumption that the conversion of 3He, the tritium β− decay product, back to tritium in the moderator is negligible. The general solution to Equation 3 is given by: N T = N ⁡ ( 0 ) ⁢ ⅇ - λ e ⁢ t + ( S + F M M ⁢ N 0 ) λ e ⁡ [ 1 = ⅇ - λ e ⁢ t ] ⁢ ⁢ Where ⁢ : ⁢ ⁢ S = φ ⁢ ⁢ σ ⁢ ⁢ N D ⁢ m M ⁢ a ⁢ ⁢ λ e = λ + L R M , and ⁢ ⁢ N ⁡ ( 0 ) = Initial ⁢ ⁢ tritium ⁢ ⁢ activity ⁢ ⁢ in ⁢ ⁢ the ⁢ ⁢ moderator . ( 4 ) The heavy water loss rate may vary from reactor to reactor and there is no single value that can be used to describe a standard CANDU 6 reactor. For this reason, the case where, N0=0 and LR=FM=0, was used to simplify Equation 4 as given by (it is believed that the simplifying assumptions used here lead to an overestimation of the tritium activity in the moderator): N T = N ⁡ ( 0 ) ⁢ ⅇ - λ ⁢ ⁢ t + S λ ⁢ ( 1 - ⅇ - λ ⁢ ⁢ t ) ( 5 ) Equation 5 based on the simplifying assumptions, described above, may overestimate the tritium activity in the moderator. The specific tritium activity (Bq·kg−1) in the heavy water at time t can now be written as:A=λNT=λN(0)e−λt+S(1−e−λt) (6)For N(0)=0.A=S(1−e−λt) (7) The tritium activity (A) in the moderator (Bq·kg−1) of a CANDU reactor can be estimated from Equation 6 provided that all the parameters in the equation are known. These values were used to calculate the tritium activity in the moderator of a CANDU 6 reactor as a function of time. The parameter values are shown in Table 1 [adopted from M. J. Song, S. H. Song, C. H. Jang, Waste Management, 15, 8, 593 (1995)]. TABLE 1Parameter Values in Equation 3 for a CANDU 6 ReactorParameterValueFMMake-up heavy water flow (kg · s−1)ReactorDependantLRHeavy water loss rate (kg · s−1)ReactorDependantMTotal heavy water inventory in the Moderator (kg)2.57 × 105 NTNumber of tritium atoms in the moderator perVariablekg D2O (atoms · kg−1)NDNumber of deuterium atoms in the moderator per6.01 × 1025kg D2O (atoms · kg−1) (The isotpic purity of heavywater in the moderator ≧99.75%)N0Number of tritium atoms in the make-up heavy0water per kgD2O (atoms · kg−1)aReactor capacity factor (A value of 85% is assumed)85%mHeavy water inventory under the neutron flux (kg)1.90 × 105 φThermal neutron flux (neutrons · cm−2 · s−1)2.30 × 1014σNeutron absorption cross-section (cm2) 4.19 × 10−28λTritium decay constant (s−1)1.78 × 10−9 The use of Equation 6 to estimate the evolution of the moderator tritium activity with time was validated by comparing the data calculated using Equation 6 and actual moderator activity data from two existing CANDU reactors. These comparisons are shown in FIG. 5 and FIG. 6, respectively. Considering the simplifying assumptions used in the derivation of Equation 6, the calculated moderator activity values follow the measured values reasonably well. As expected the calculated values were higher than the measured values due to a variety of assumptions used in the derivation of Equation 6, including for example no loss of tritium from the moderator, other than from decay, a capacity factor of 85% and flux of 2.30×1014. Also these results suggest that the contribution of reaction 3He(n,p)3H to the production of 3H in the moderator is not important. The results confirm the validity of Equation 6 and hence Equation 5 for use in estimating the tritium activity in the moderator in CANDU reactors, under the assumption that there is no loss of tritium from the moderator other than from decay. The 3He production rate in the moderator (Equation 1) can now be written as: ⅆ N 3 He ⅆ t = λ ⁢ ⁢ MN T = λ ⁢ ⁢ MN ⁡ ( 0 ) ⁢ ⅇ - λ ⁢ ⁢ t + SM ⁡ ( 1 - ⅇ - λ ⁢ ⁢ t ) ( 8 ) N 3 He = SMt + ( MN ⁡ ( 0 ) - SM λ ) ⁢ ( 1 - ⅇ - λ ⁢ ⁢ t ) ( 9 ) For the case where N(0)=0, Equation 9 simplifies to: N 3 He = SMt - ( SM λ ) ⁢ ( 1 - ⅇ - λ ⁢ ⁢ t ) ( 10 ) Equation 8 gives an upper-bound estimate of the production rate of 3He in the moderator for a CANDU 6 reactor and Equation 9 gives an upper-bound estimate of the total number of 3He atoms in the moderator as a function of time. FIG. 7 shows the upper-bound estimates for total 3He produced in the moderator and moderator tritium activity as a function of time. An estimate of the design life of current CANDU 6 reactors and the pressure tubes, at a capacity factor of 85%, may be 40 and 25 years, respectively. As FIG. 7 shows, over the design life of the pressure tubes, a typical CANDU 6 reactor generates about 12.7 m3 (STP) of 3He in the moderator, assuming that there is no loss of tritium from the moderator through heavy water leaks, replacement, detritiation, or evaporation to the moderator cover gas. This amounts to an upper-bound, average 3He production rate of ˜0.8 m3 (STP) per year. As the data show an amount of 3He (<<1 m3 (STP) per annum) may be available for recovery from a CANDU reactor from the moderator cover gas. The tritium activities in the moderators and the 3He production rates at different CANDU reactors in Canada are different. Table 2 shows the 3He production rate as a function of the tritium activity in the moderator for a typical CANDU 6 reactor. As the data show, even at the highest moderator activity used in the calculations, which is similar to the measured tritium activity in the Existing Reactor Two moderator in 2007, the 3He production rate is <0.7 m3(STP)·per year. TABLE 2Estimated 3He Production Rate per Year in a CANDU 6 ReactorModerator Activity (GBq · kg−1)3He Production Rate (m3(STP) · a−1)3700.117400.2211100.3414800.4518500.5622200.67 Based on the above, it is believe that the amount of 3He gas in the extracted moderator cover gas is significantly higher than the amount of naturally abundant 3He found in the helium cover gas. That is, the moderator cover gas within the calandria may be 3He enriched. If 3He gas is to be extracted from collected moderator cover gas, it may be desirable to capture a significant portion of the 3He enriched cover gas in the calandria, and preferably to capture substantially all of the 3He enriched cover gas, for processing. In operation, some of the moderator cover gas may escape from the calandria and/or the moderator cover gas system. If, for example, a daily helium loss rate of 30% of the total helium inventory in the moderator cover gas system is assumed, it is believed that recovering about 800-900 m3 per year of helium may help facilitate recovery of most of the 3He produced in a CANDU 6 reactor. Table 3 shows the concentration of 3He in the recovered helium gas, as a function of the average moderator activity. TABLE 3Estimated Concentraion of 3He in the Recovered Helium GasModerator Activity (GBq · kg−1)3He in Recovered Helium (ppm V)3701307402601110410148054018506702220800 Preferably, the cover gas system can be configured so that impurities in the moderator cover gas, including, for example, D2, O2, CO2, 14C, Ar, T2 and DT, can be removed from the helium cover gas before the cover gas is processed to separate the 3He gas from the 4He gas. Optionally, the extraction of the cover gas can be an off-line process, when the reactor is shut down and/or the cover gas system(s) are purged allowing substantially all of the enriched cover gas to be collected in a single batch. This may allow the cover gas to be batch processed to extract the 3He gas, which may be advantageous for some extraction process and/or apparatuses. It may also be desirable if the reactor is going to be shut down anyway (for example for service). Alternatively, the extraction of the cover gas can be an on-line process, in which a stream (optionally a generally continuous stream) of cover gas can be drawn from the calandria and/or the heat transport system (e.g. the pressurizer) while the reactor is in use. Optionally, in such a configuration the 3He gas can be separated from the extracted stream of cover gas using a real time or on-line process or separation apparatus. This may allow the 3He gas to be extracted while the reactor is in use. This may help facilitate substantially continuous collection of 3He gas Optionally, after the 3He gas has been separated, some or all of the cover gas extracted from the calandria can be recycled and reintroduced into the calandria. Cover gas from the heat transport system may also be treated and recycled. Recycling at least some of the cover gas may help reduce the amount of make up or replacement cover gas needed and/or may help increase the efficiency of the cover gas system. The 3He gas can be separated from the 4He cover gas using any suitable separation apparatus and/or separation technique. For example, a number of technologies have been used in the past for separating 3He from 3He+4He mixtures. Examples of some of the known methods include: 1. Thermal Diffusion 2. Fractional Distillation 3. “Heat-Flush” Method 4. “Super Leak” Method 5. Cryogenic Adsorption Thermal Diffusion Before the extraction of 3He from tritium decay started, there have been efforts to separate naturally abundant 3He from helium sources (extracted from air or natural gas). Thermal diffusion has one of the early technologies tested for use in enriching naturally abundant 3He in helium sources. Thermal diffusion is the relative motion of the components of a gaseous mixture or solution, which is established when there is a temperature gradient in a medium. Thermal diffusion in gases was theoretically predicted by Enskog on the basis of the kinetic theory of gases [D. Enskog, Physik Zeits, 12, 56 and 533 (1911)]. It was later discovered experimentally by Chapman and Dootson [S. Chapman, F. W. Dootson, Phil. Mag., 33, 248 (1917)]. Thermal diffusion sets up a concentration gradient with lighter molecules concentrating at the high-temperature side with heavier molecules concentrating in the low-temperature side leading to separation of components in a gaseous mixture. The concentration gradient in turn causes ordinary diffusion and the separation effect of thermal diffusion is balanced by the counteraction of the concentration diffusion. In a binary gaseous mixture at constant pressure, the total diffusion mass flux, Ji, for each component, i, in the absence of external forces, is given by: J i = - nD 12 ⁢ ∇ ⁢ C i + nD T ⁢ 1 T ⁢ ∇ T = - nD 12 ⁡ [ ∇ ⁢ C i - k T ⁢ 1 T ⁢ ∇ ⁢ T ] = - nD 12 ⁡ [ ∇ C i - α ⁢ ⁢ C 1 ⁢ C 2 ⁢ ∇ ⁢ ln ⁢ ⁢ T ] ( 11 ) Where: Ci=Concentration of species i (Ci=ni/n, 1=1,2) D12=Binary diffusion coefficient, DT=Thermal diffusion coefficient, n=Total number of molecules in unit volume (n=n1+n2) kT=Thermal diffusion ratio=D12/DT=αC1C2 α=Thermal diffusion constant In gaseous mixtures a does not generally exceed 0.4; and for mixture of isotopes, a typical value for α is ˜0.01. The value of kT depends in a complex manner on the molecular masses, effective molecule size, temperature, mixture composition, and on the laws of intermolecular interaction. The closer the intermolecular forces approach the laws of interaction between the elastic, solid, spheres, the greater is the value of kT; it also increases with increase in the molecule dimension and mass ratio. When molecules interact in accordance with the law for solid, elastic, spheres, kT is independent of the temperature and the heavier molecules gather, in this case, in the cold region (kT>0 for m1>m2>2 where m1 and m2 are the masses of the respective components), but if m1 and m2 are equal, then larger molecules move into a cold region. For other laws of intermolecular interaction kT can depend considerably on the temperature and can even change sign. Thermal diffusion became important as a method of separating isotopes or mixtures of gases when Clusius and Dickel invented the thermal-diffusion column [K. Klusius, G. Dickel, Naturwiss, 26, 546 (1938)]. The original thermal diffusion column theory was developed by Furry, Jones and Onsagar (FJO) [W. H. Furry, R. C. Jones, L. Onsager, Phys. Rev., 55, 1083 (1939)], [W. H. Furry, R. C. Jones, Phys. Rev., 69, 459 (1946)], [R. C. Jones, W. H. Furry, Rev. Mod. Phys., 18, 151 (1946)]. A thermal diffusion column, used for isotopes separation, essentially consists of a vertical tube maintained at a low temperature with a heated wire located in the central axis. Other variants include coaxial, tube-in-tube configuration, where the central tube is heated and maintained at a high temperature while outer tube is maintained at a lower temperature. In a thermal diffusion column, the lighter gas flows upwards near the central hot wire or tube and the heavier gas flows downwards near the outside cold wall, by convection. The temperature gradient across the tube causes a horizontal concentration gradient by thermal diffusion with the lighter molecules concentrating at the hot central wire or tube and the heavier molecules concentrating at the cold wall. These two effects are superimposed and the opposing convection currents carry the lighter molecule to the top and the heavier molecule to the bottom. The upward and downward gas flows are in counter flow resulting in a concentration gradient between the top and the bottom of the column greater than in a horizontal plane. The maximum separation factor that can be obtained in a column is limited by the remixing of gases caused by ordinary or concentration diffusion and by the convection currents. For a binary mixture, modified set of FJO equations describing the mass transport of species in the Thermal Diffusion column is given in [L. Hodor, Sep. Sci. Technol., 38, 5, 1229 (2003)]. The effectiveness of thermal diffusion as a means of separating 3He from oil-well helium (3He/4He=1×10−7) has been investigated by several groups. McInteer et al. [B. B. McInteer, L. T. Aldrich A. O, Nier, Phys. Rev. 74, 8, 946 (1948)], using a 3-column thermal diffusion system, were able to produce 14 cm3(STP) per day of 0.21% 3He using a 1.15×10−5% 3He feed, which corresponds to a separation factor of 1.83×104, under the test conditions. The thermal diffusion system used consisted of two 3.5 m-long coaxial, tube-in-tube columns in the front end. The first column had a hot wall diameter of ˜6.04 cm and cold wall diameter of 7.3 cm and the second column had a hot wall diameter of 3.5 cm and a cold wall diameter of 4.76 cm. The final column consisted of a 2.5 m-long hot-wire column of wire diameter 0.036 cm. The columns were operated at high pressure (0.69 and 0.88 MPa(g)). The separation factor achieved, under the tested conditions, was found to be a strong function of the product draw-off rate and it decreased with increasing draw-off. It was also found that the hot-wire column alone could have a separation factor in the order of 1×104. A thermal diffusion plant capable of producing 2 cm3 (STP) per week was also operated for several years at an existing establishment at Harwell, England for several years using a feed gas containing naturally abundant 3He in helium from air (1.2×10−4% 3He). The thermal diffusion system consisted of two identical coaxial, tube-in-tube columns (0.8 cm diameter hot wall×3 cm diameter cold wall×4.5 m high) at the front end and a hot-wire column (1.3 cm diameter cold wall×4.57 m high) at the back end. With the feed gas that is available from CANDU reactors, it is believed that an overall separation factor of about 1×103 to about 1×104 may be required to obtain a stream of 3He with 99.9% purity. However, for the recovery of 3He from the available feeds, a large volume of recovered helium gas (˜840 m3) needs to be processed. While thermal diffusion is a relatively straightforward method of isotope separation, as a separation process, it may have a low thermodynamic efficiency, requires multiple stages, large amounts of electrical power and long processing times. However, thermal diffusion may be suitable as a final stage of 3He enrichment since, a single hot-wire column could have a separation factor in the order of 1×104. Fractional Distillation The boiling points of 4He and 3He are 4.2 K and 3.9 K respectively and 3He+4He mixtures may be separated by distillation. Distillation of 3He+4He solutions is generally considered as a more efficient method than the thermal diffusion method for the separation of 3He isotope. The separation factor for thermal diffusion is proportional to the square root of the isotopic mass ratio which is fixed at 1.5 while the separation factor for distillation is proportional to the relative volatility ratio and the minimum separation factor for 3He+4He distillation is 2.5 at the critical temperature of 3.36 K, and increases with decreasing temperature. Several small-scale, batch distillation processes have been reported for the purification of 3He+4He mixtures relatively rich in 3He [W. R. Abel, A. C. Anderson, W. C. Black, J. C. Wheatley, Physics, 1, 337 (1967)], [V. N. Grigor'ev, B. N. Yesel'son, V. A. Mikheev, O. A. Tolkacheva, Soy. Phys., J.E.T.P., 25, 572 (1976)], [R. H. Sherman, Proceedings of the 10th International Conference on Low-Temperature Physics, Vol. 1, 188 (1966)], [R. P. Giffard, R. B. Harrison, J. Hatton, W. S. Truscot, Cryogenics, 7, 179 (1967)], [A. C. Anderson, Cryogenics, 8, 50 (1968)], [A. Tominaga, S. Kawano, Y. Narahara, J. Phys., D: Appl. Phys. 22, 1020 (1989)]. The 3He in the feed used in these studies varied from 10 to 99.9993% and 3He product purity varied from 99.99% to 99.9998%. The production rate of 3He in these processes was reported to be in the range 0.1 to 18.5 L·h−1. The objective of most of these processes was essentially to remove the 4He impurity traces in enriched 3He. These data, however, show the potential of the fractional distillation process to obtain very high purity 3He. A continuous distillation apparatus for the separation of 3He—4He mixtures is described in W. R. Wilkes, Advances in Cryogenic Engineering, 16, 298 (1970), the entirety of which is incorporated herein by reference. This system has been operated continuously for few hours at a time with a 3He—4He mixture containing 8.7% 3He in the feed while withdrawing a product containing 99.95% 3He and a raffinate containing 0.02% 3He. Based on the data obtained, the author concluded that at a feed rate ˜60 L(STP)·h−1 of this mixture, a 99.9% pure 3He product may be obtained at a rate of ˜5.2 L·h1. A continuous distillation process of this size is suitable as a final stage of enriching 3He in 3He+4He mixtures recovered from the moderator cover gas. However, this requires pre-enriching the 3He content in 3He+4He mixtures, recovered from the moderator cover gas from CANDU reactors, for use as feed for a distillation system of similar size. Heat Flush Method The “heat flush” method of 3He—4He isotopes separation is based on superfluid properties of 4He. The “heat flush” method exploits the property that 3He does not participate in the superfluid flow of 4He. It has been demonstrated that if heat is applied at one end of a vessel containing liquid helium below the lambda point (A) and refrigeration at the other end, the 3He flows with the normal liquid away from the heater and towards the cold end of the vessel [C. T. Lane, H. A. Fairbank, L. T. Aldrich, A. O. Nier, Phys. Rev. 73, 256 (1948)], [T. Soller, W. M. Fairbank, A. D. Crowell, Phys. Rev. 91, 1058 (1953)]. The λ point is the temperature below which normal fluid helium (helium I) transitions to the superfluid helium point (helium II). The λ temperature of 4He in 3He+4He solutions decreases with increasing 3He content in the solution. Consequently, at any given temperature there is a 3He concentration, above which 4He is no longer a superfluid. This, in effect, imposes a limitation on the enrichment of 3He that can be achieved using methods that exploit superfluid properties of 4He. The “heat flush” method has been used to enrich gas-well helium (˜1×10−7) by a factor of 130 [C. T. Lane, H. A. Fairbank, L. T. Aldrich, A. O. Nier, Phys. Rev. 73, 256 (1948)] and 3×104. In the latter case, up to 0.5% of 3He in 3He+4He mixtures were obtained at a rate of about 60-75 cm3 of enriched gas. A device combining the “heat flush” method with batch distillation, capable of enriching 3He from natural abundance level (˜1×10−8) to ˜99.5% has also been demonstrated [V. P. Peshkov, J. Exp. Theor. Phys, 30, 850 (1956), Translation: Soviet Physics, JETP, 3, 706 (1956)]. In this case, product of the “heat flush” step was ˜0.2% of 3He. There have been no reports of enrichment of 3He in 3He+4He mixtures above ˜1.5% directly by the “heat flush” method. A final enrichment up to 4% of 3He has been achieved by the “heat flush” method using pre-enriched mixture of 3He—4He up to 0.01% 3He by thermal diffusion method. While the “heat flush” method may not achieve enrichment of 3He significantly above few percent, it may be used as a suitable pre-enrichment process for separating 3He from the moderator cover gas recovered from CANDU reactors. Superleak Method The “superleak” method of separating 3He—4He gas mixtures is also based on the superfluid properties of 4He. The method of “superleak” is based on the ability of superfluid 4He to flow through capillaries or very narrow channels, while 3He cannot. This method has been used to partially enrich 3He present in atmospheric helium at an abundance ratio of 1.22×10−6. The “superleak” consisted of a ground glass joint with a channel width of ˜1 μm. Using ten parallel superleaks of dimension 1×10−4 cm, a 3He+4He mixture containing 2% 3He was enriched to 95%, in a single operation, while processing the initial gas mixture at a rate of 200 cm3 (STP) per hour. Since the “superleak” method is believed to be capable of enriching 3He to significant levels from relatively dilute 3He+4He mixtures in a single operation, the “superleak” process may be a suitable pre-enrichment process that could be coupled to a fractional distillation stage to achieve high-purity 3He. Cryogenic Adsorption Method The adsorption based separation of 3He—4He, at liquid helium temperature, is based on the differences in the adsorption energies of 3He and 4He on activated charcoal. This method has been used to remove trace amounts of 4He impurity (˜0.1%) in commercially available 3He. A reduction in the 4He impurity from 0.1% to <0.01% has been achieved after two passes through the column of 25 L of at a flow rate in the range 0.04-0.1 L·min−1. About 23 L of purified product and ˜2 L of 4He enriched gas were recovered. No details on the amount of charcoal used in the process or the 4He adsorption capacity of charcoal at liquid helium temperature are given. No other reports on using the cryogenic adsorption of 4He on charcoal at high 4He levels are found in the literature. This method, while simple, may be more appropriate for removing 4He impurity at trace levels in enriched 3He. Using one or more of the apparatuses and processes described above, and optionally using any other suitable apparatus and/or process, there are several processing options are available for consideration for the processing of moderator cover gas recovered from CANDU reactors to extract high purity 3He. Some examples of processing options include: 1. Pre-enrichment with “heatflush” method and final enrichment with distillation, 2. Pre-enrichment with “superleak” method and final enrichment with distillation, and 3. Pre-enrichment with “superleak” method and final enrichment with thermal diffusion. Referring to FIG. 1, a schematic representation example of a heavy water reactor, e.g. a CANDU reactor 100, includes a calandria 102 containing a heavy water moderator liquid 104 and a plurality of pressure tubes 106 extending through the calandria 102. A heat transport system 108 is used to circulate a cooling fluid 110 through the pressure tubes, and includes a pressurizer 109. A moderator a cover gas system 112 is used to circulate and optionally treat or process a moderator cover gas 114, and a coolant cover gas system 113 is used to circulate and optionally treat a coolant cover gas 137. Optionally, the moderator cover gas 114 and the coolant cover gas 137 may be the same gas, such as helium. The pressure tubes 106 may be of any suitable design and can contain one or more nuclear fuel bundles/rods 115. The reactor 100 can include any suitable number of pressure tubes 106, arranged in any suitable configuration. The pressure tubes 106 can be formed from any suitable material. The heat transport system 108 may be used to circulate a pressurized heavy water cooling fluid 110 through the pressure tubes 106. Incoming heavy water cooling fluid enters the tubes 106, illustrated by arrows 116, is heated by the fuel bundles 115 and exits the pressure tubes 106, illustrated by arrows 118 at an elevated temperature. The high temperature cooling fluid may then flow through any suitable heat exchanger, for example a boiler 120 that can be used to heat an incoming water stream 122 to generate a steam stream 124, which may in turn be used to drive any suitable turbine generator (not shown) and produce electrical power. The heat transport system 108 may include any suitable fixtures and components including, for example, valves, pumps, filters and any other suitable apparatus that is not illustrated in the present schematic drawing. The moderator liquid 104 is contained within the calandria 102 and surrounds the pressure tubes 106. A moderator system 119 circulates the moderator through the calandria, and can include any suitable piping, conduits, processing modules (such as a heat exchanger), valves, pumps and other such components. In the illustrated schematic, a moderator vessel 117 holds moderator liquid that is outside the calandria. The moderator liquid 104 may have exposed free surfaces 126 at a plurality of locations within the moderator system 115. For example, a free surface 126 is located toward the top of the calandria 102. A head space or plenum 128 is defined between the free surface 126 of the moderator fluid 104 and the upper wall 130 of the calandria. While illustrated as a single, continuous chamber, the head space 128 may be formed from two or more separate chambers or regions within the calandria 102, and need not be a single, continuous chamber. The size and shape of the head space 128 may be selected based on a variety of factors, including, for example the calandria size, the calandria shape, the configuration of the cover gas system 112 and the operating conditions of the reactor 100. A free surface 126a may also be formed within a head space 128a in vessel 117, and optionally within some of the conduits or piping of the moderator system 115. Each head space 128 and 128a may be filled with moderator cover gas, and may be in fluid communication with a common moderator cover gas system 112. Cover gas system features described in relation to the calandria 104 and head space 128 may also be included in vessel 117 and head space 128a, and analogous elements may be identified using analogous reference characters with an “a” suffix. The calandria 104 may include a gas inlet 132 and a gas outlet 134 that are in fluid communication with each other, for example via the cover gas head space 128, and that can be connected to any suitable cover gas system 112. While illustrated as a single port for clarity, the gas inlet 132 may include a plurality of discrete ports or openings in the calandria sidewall and the supplying conduit may have a corresponding number of branches and outlets. Similarly, the gas outlet 134 may include a plurality of separate ports or openings that are in communication with the cover gas head space 128, and connected to a common outlet passage. The cover gas 114 can flow into the head space 128 via the gas inlet 132 and can be extracted from the head space 128 via the gas outlet 134. Optionally, the gas inlet 132 and gas outlet 134 can include any suitable valve(s) or flow control mechanism to selectably adjust and/or limit the flow of cover gas 114 within the head space 128. The gas inlet 132 and gas outlet 134 may also include any other suitable equipment, including, for example, a flow meter and sensors. The cover gas system is used to supply cover gas 114 to the calandria 104 and to circulate the cover gas 114 through the head space 128. The cover gas system 112 can be of any suitable configuration, and may include any suitable components or apparatuses. In the illustrated example, the cover gas system 112 includes a cover gas supply passage 138 for supplying cover gas 114 to the head space 128, and a gas extraction passage 140 for extracting gas from within the head space 128. The passages 138, 140 may be formed from any suitable conduit members, including, for example pipes and ducts, and may be formed from any material that is suitable for use with a pressurized heavy water reactor. In the illustrated example, the cover gas supply passage 138 has an upstream or inlet end 142 and downstream or outlet end 144 that is spaced apart from the inlet end 142. The outlet 144 of the supply passage 138 is connectable to the gas inlet 132 on the calandria 104. In this configuration, cover gas 114 may be supplied into the head space 128 via the supply passage 138, as illustrated using arrows 146. When contained within the head space 128, the cover gas 114 is in contact with the free surface 126 of the moderator liquid 104. The inlet 142 of the cover gas supply passage 138 can be connected to any suitable supply or source of cover gas 114. In the illustrated example, the inlet 142 of the cover gas supply passage 138 is connected to a separation apparatus 148, as explained in greater detail below. Alternatively, the inlet 132 may be connected to a helium bottle (not shown) or other cover gas supply source. As explained in detail above, when the reactor 100 is operated a quantity of tritium may be produced within the moderator liquid. The tritium may then decay to produce 3He gas 150 in the moderator liquid 104. Due to the relatively low solubility of 3He gas in the heavy water moderator 104, at least a portion of the 3He gas 150 produced may form bubbles, diffuse out of the free surfaces 126 or otherwise escape from the moderator liquid, as illustrated using arrows 152. 3He gas bubbling out of the moderator liquid 104 can flow into the head space 128, and may become mixed with the cover gas 114 contained in the head space 128. Optionally, as explained above, the cover gas 114 introduced into the head space may be substantially pure helium (4He) gas. When the 3He gas 150, and other impurities and by-products as explained above, flow into the head space 128, the composition of the cover gas 114 may change from substantially pure helium (4He) to a mixture of gases. The mixture of gases may be extracted from the head space 128 via the gas outlet 134 as a gas outlet stream, represented by arrow 154. In this configuration the gas outlet stream 154 may include a mixture of the helium cover gas 114 and at least a portion of the 3He gas 150. In the illustrated example, an inlet end 156 of the gas extraction passage 140 is coupled to the gas outlet 134 of the calandria 104 to extract the gas outlet stream 150 from the head space 128. The gas outlet stream 154 can flow along the gas extraction passage 140, away from the head space 128, for further treatment and/or processing. One or more suitable gas treatment and/or processing apparatuses can be provided in the gas extraction passage 140, downstream from the head space 128. Some examples of suitable gas processing apparatuses are explained above. The processing apparatuses can be selected to process the gas outlet stream in a variety of different ways. In addition to, or as an alternative to known gas processing apparatuses, the cover gas system 112 includes a 3He gas separation apparatus 148 that is operable to separate 3He gas 150 from the mixture of gases forming the gas outlet stream 154. The 3He gas separation apparatus 148 may be of any suitable configuration and can include one or more gas separation modules. Some examples of suitable 3He gas separation apparatuses are explained in detail above, and any one of these apparatuses can be used alone and/or in combination with any one or more of the other apparatuses described herein or any other suitable apparatus. One or more suitable gas processing apparatus (for example apparatuses to remove other impurities, such as D2, O2, CO2 and 41Ar from the cover gas) may be provided upstream and/or downstream from the 3He separation apparatus. In the illustrated example, the 3He gas separation apparatus 148 includes a 3He gas outlet passage 160 to output a stream 162 of 3He gas separated from the outlet gas stream 154. The 3He gas outlet passage 160 can be connected to any suitable downstream apparatus including, for example, a storage container and/or secondary processing apparatus (not shown). Optionally, the 3He gas separation apparatus 148 may also include at least one other outlet to output non-3He gas streams. In the illustrated example, the 3He gas separation apparatus 148 includes a second outlet 164 for outputting a treated cover gas stream 166. The treated cover gas stream 166 may include the helium cover gas from which the 3He gas was separated, and may include other trace gases and/or impurities. In the illustrated example, the non-3He gas outlet 164 is coupled to the inlet 142 of the cover gas supply passage 138. In this configuration, treated cover gas 166 (i.e. cover gas that has had the 3He gas removed) can be re-used and recycled into the head space 128. Alternatively, the non-3He gas outlet 164 need not be coupled to the cover gas supply passage 138, and the non-3He gas exiting the separation apparatus 148 can be contained or disposed of in any suitable manner. Optionally, one or more additional gas treatment apparatuses can be provided upstream or downstream from the separation apparatus 148. Optionally, the cover gas system 112 can be operated as an on-line system, in which the gas outlet stream 154 can be drawn from the head space 128 while the reactor 100 is in use. In this configuration, the gas outlet stream 154 may be extracted at a generally uniform flow rate while the reactor 100 is in use. Alternatively, the flow rate of the gas outlet stream 154 may vary over time and/or in response to operating conditions of the reactor. Operating the cover gas system 112 in an on-line configuration may allow the 3He gas to continuously extracted from the head space 128 while the reactor 100 is in use. This may allow collection of 3He gas from active reactors and may help to minimize disruptions or alterations to the operating conditions of the reactor. Alternatively, the cover gas system 112 can be operated in an off-line or batch-type system. It is understood that only some aspects of the reactor 100 are illustrated in the present schematic. An operational reactor 100 incorporating one or more of the aspects of the present teaching may include any combination of suitable operating components, including, for example, control rods, light water condensate pumps, secondary cooling loops, fuel loading machines, a reactor containment building, a pressurizer, valves, pumps and any other suitable equipment. While treatment of the moderator cover gas 114, via a moderator cover gas system 112 is described in detail above, an analogous process may be used to extract 3He from the coolant cover gas 137. A coolant cover gas system 1112 may include some or all of the elements of the cover gas systems described herein, and/or may include additional elements not described above. The coolant cover gas system 1112 may be generally similar to the moderator cover gas system 112, and analogous elements are illustrated using like reference numerals indexed by 1000. The coolant cover gas system 1112 may include any suitable 3He separation apparatus 1148 that may be in fluid communication with head space 1128 within the pressurizer 109 (or at any other suitable location or locations within the coolant system 113). A cover gas extraction passage 1140 may transport the coolant cover gas 1128, including 3He mixed therein, for processing, and treated cover gas may be returned to the head space 1128 via the gas supply passage 1138. Referring to FIG. 2, another schematic example of a reactor 200 includes a calandria 202 and pressure tubes 206. The reactor 200 may be generally similar to reactor 100, and like elements are illustrated using like reference characters indexed by 100. The calandria 202 contains a heavy water moderator liquid 204 and a cover gas head space 228 is provided above the free surface 226 of the moderator liquid 204. A cover gas supply passage 238 is connected to gas inlet 232 to introduce the cover gas 236 into the head space 228, and a gas extraction passage 240 extends away from the gas outlet 234 to extract a gas outlet stream 254 from the head space 228. A 3He separation apparatus 248 is provided in the gas extraction passage, downstream from the gas outlet. In the illustrated example, the 3He separation apparatus 248 is a two-stage apparatus that includes a first separation module 270 and a second separation module 272 provided downstream from the first module 270. The first and second separation modules 270, 272 may be the same apparatus/process, or alternatively may be different apparatuses/processes. An intermediate conduit 274 extends from an outlet 276 on the first separation module 270 to an inlet 278 on the second module 272. A 3He gas outlet 280 on the second module 272 can form the 3He gas outlet for outputting a 3He gas stream 262. At least one, or both, of the first separation module 270 and second separation module 272 may also include one or more second outlet 264 for outputting a treated cover gas stream 266, or other gas output stream. Optionally, in the illustrated example, as illustrated using dashed lines, at least a portion 266a of the treated cover gas 266 stream may be recycled via a recycle passage 282 and re-introduced in the cover gas supply passage 238, upstream from the cover gas inlet 232 of the calandria 202. Optionally, one or more suitable treatment apparatuses may be provided in the recycle passage. In the illustrated configuration, two gas treatment apparatuses 284a and 284b are provided in the recycle passage 282. The gas treatment apparatuses 284a and 284b may be preheaters, recombination units, filters, separators, flame arresters or any other suitable apparatus. Passing the treated cover gas stream 266a through one or more treatment apparatuses 284a and 284b may help remove additional impurities from the cover gas and otherwise treat the cover gas so that it is suitable for re-introduction into the head space 228. This may help make the treated cover gas 266a more suitable for re-use. While illustrated as being provided in the recycle passage 282, the gas treatment apparatuses 284a and 284b may be provided in the gas extraction passage 240, and optionally, may be upstream from the 3He separation apparatus 248. Alternatively, or in addition to receiving recycled cover gas 266a, the gas supply passage 238 may be connected to any suitable external cover gas source (not shown). The cover gas 236 supplied to the head space 228 may comprise recycled cover gas 266a, fresh cover gas 246a or any suitable combination thereof. Referring to FIG. 3, another schematic example of a reactor 300 includes a calandria 302, pressure tubes 306 and a cover gas supply passage 338. A heavy water moderator liquid 304 is contained in the calandria 302 and a cover gas head space 328 covers the free surface 326 of the moderator 304. The reactor 300 may be generally similar to reactor 100, and like elements are illustrated using like reference characters indexed by 200 In the illustrated example, a gas treatment apparatus 384 is provided the extracted gas passage 340 upstream from the 3He separation apparatus 348. The gas outlet stream 354 can be extracted from the head space 328 and fed into the gas treatment apparatus 384. Impurities and other gases removed by the gas treatment apparatus 384 can be discharged via a first outlet 390 as an impurity gas stream 392. After treatment, a partially treated gas stream 394, for example comprising primarily 3He and 4He gases, can exit the gas treatment apparatus 384 via a second outlet 398 and can flow into the 3He separation apparatus 348. A stream of separated 3He gas 362 can exit the 3He separation apparatus 348 via a first outlet 400, and a stream of treated cover gas 366 can exit the 3He separation apparatus via a second outlet 402. In some configurations, based on the performance and characteristics of the gas treatment apparatus 384 and the 3He separation apparatus 348, the stream of cover gas 366 exiting the 3He separation apparatus 348 may be substantially pure 4He gas. Positioning one or more gas treatment apparatuses 384 upstream from the 3He separation apparatus 348 may facilitate removal of impurities and other gases from the gas outlet stream 354 before the gas outlet stream reaches the 3He separation apparatus 348. This may help prevent fouling or damage to the 3He separation apparatus 348. This may also help improve the efficiency of the 3He separation apparatus 348 and/or allow for the use of a particular 3He separation apparatus (providing a given separation process) that may not be suitable for use on a gas stream that includes impurities or a mixture of gases other than 3He and 4He gases. Optionally, a gas recycle passage 382 can be provided to recycle some or all of the cover gas 366 exiting the 3He separation apparatus 348 to the cover gas supply passage 338 for re-introduction into the head space 328. In all the configurations shown in FIGS. 1, 2 and 3, the devices and equipment for separating out 3He can be provided in a separate circuit from the main circuit, possibly in parallel with it, to maintain a separate circulation of the helium to the cover gas space above the moderator, so that any failure of equipment in the separate circuit should not compromise operation of the main circuit. As noted it may be preferable, in the separate circuit to provide elements to remove or otherwise process other contaminant gases. For example some form of igniter can be provided to ensure that any hydrogen or deuterium present is burned to form water. After removal or processing of such contaminants, the helium isotopes can be separated. Referring to FIG. 4, a method of collecting 3He from a pressurized heavy water nuclear reactor may begin a step 1000 with providing heavy water as either a moderator, coolant or both within a heavy water reactor. At step 1002 a cover gas can be provided within the reactor and may cover at least a portion of the free surface of the heavy water. Optionally, the cover gas may be substantially pure 4He, for example comprising at least 90% 4He by volume. If necessary, additional make-up cover gas may be added to the reactor from time to time, as needed, at step 1002a. At step 1004 the heavy water nuclear reactor may be operated to produce 3He gas, for example via decay of tritium in the heavy water. At least a portion of the 3He gas may escape from the heavy water and mix with the cover gas, at step 1006. At step 1008 an outlet gas stream is extracted from within the reactor. The outlet gas stream may include a mixture of the cover gas, the 3He gas and other trace gases and/or impurities. At step 1010 the 3He gas is separated from the outlet gas stream using a suitable separation apparatus. Optionally, the method can include the optional step 1010a of collecting or routing the 3He gas for further processing, and the optional step 1010b in which the collected 3He gas can be further treated or purified. At step 1012 a 3He gas stream may be output from the separation apparatus for further processing, and at step 1012 a treated cover gas stream may also be output from the separation apparatus. Optionally, at step 1014 at least a portion treated cover gas stream can be further processed or treated using a suitable gas treatment apparatus. At optional step 1016, at least a portion of the treated cover gas can be recycled by reintroducing at least a portion of the treated cover gas stream into the reactor. Optionally, some or all of steps 1000 to 1016 can be on-line steps performed while the heavy water nuclear reactor is operating. Alternatively, some or all of steps 1000 to 1016 can be off-line steps performed while the reactor is not operating. The step of separating the 3He gas from the outlet gas stream may include utilizing any suitable apparatus and/or carrying out any suitable process including, for example, at least one of a thermal diffusion process, a fractional diffusion process, a heat flush process, a superleak process and a differential absorption process. While heavy water reactors including both a heavy water moderator and heavy water coolant are illustrated, the 3He extraction apparatuses and methods described herein may be used on any suitable heavy water reactor, including, for example, reactors having a heavy water moderator and a non-heavy water coolant, reactors having a non-heavy water moderator and a heavy water coolant and reactors having a non-heavy water moderator, a non-heavy water coolant but that include some other type of heavy water circuit or system that is provided within the reactor such that 3He is formed in the heavy water system. What has been described above has been intended to be illustrative of the invention and non-limiting and it will be understood by persons skilled in the art that other variants and modifications may be made without departing from the scope of the invention as defined in the claims appended hereto.
	claims	1. An ion implanting apparatus, comprisinga vacuum chamber;an ion irradiating unit for irradiating ions into the vacuum chamber;a substrate holder for holding an object to be processed in an irradiating range of ions inside the vacuum chamber in which the ions are irradiated;an object to be measured, which is arranged in the irradiating range of ions inside the vacuum chamber;a temperature measuring unit for measuring the temperature of the object to be measured when the ions are irradiated on the object to be measured and on the object to be processed together; anda control unit which is connected to the temperature measuring unit,wherein the object to be measured and the object to be processed are both arranged in the ion irradiation range, when the ions are irradiated onto the object to be processed, the ions and neutralized particles are made incident to both of the object to be processed and the object to be measured, to raise temperatures,wherein a relationship between a temperature of the object to be measured and a number of atoms to be implanted into the object to be processed is stored in the control unit, When the ions are irradiated,wherein the number of atoms implanted into the object to be processed is being determined from the temperature measured by the temperature measuring unit, and when the determined number of atoms reaches the predetermined number of the atoms, irradiation of the ions on the object to be processed and the object to be measured is being terminated. 2. The ion implanting apparatus as set forth in claim 1, wherein the object to be measured is arranged spaced apart from the object to be processed. 3. The ion implanting apparatus as set forth in claim 2, wherein the temperature measuring unit includes a sensor contacted with the object to be measured, and the temperature of the object to be measured is being detected by the sensor.
055915647	description	DETAILED DESCRIPTION OF THE INVENTION According to the invention, electromagnetic radiation of shorter, or of significantly shorter, wavelength than visible light is used for lithography of integrated circuits. Given the inherent resolution problems associated with conventional visible light and near-visible light photolithographic techniques (discussed hereinabove), the use of shorter wavelength radiation sources is highly desirable. However, due to the failure of conventional optics to perform at these short wavelengths, it is necessary to employ near-field or direct-write, afocal imaging techniques (non-focusing or non-converging optically) with short wavelength radiation sources. In this manner, by avoiding the inherent resolution problems associated with the relatively long wavelengths of light, finer (smaller) features can be defined on a semiconductor wafer. For example, fine lines can be patterned in a layer of photoresist material on the wafer and, subsequently, lines can be etched into a layer underlying the photoresist. The invention takes advantage of the situation that the apparent `resolutions` (if you will) of etching techniques are typically much (e.g., orders of magnitude) finer than the resolution of light. Generally, "resolution" is the ability of a given medium to create patterns on another medium. As used herein, the term "lithography" means any technique of creating patterns on a semiconductor wafer (or on a photoresist layer on the wafer). By providing higher (than light) resolution lithography techniques, the invention affords the opportunity to create finer, more densely packed features and devices on a semiconductor device. For example, more transistors can be formed on a die of given area, and more conductive lines (e.g., polysilicon or metal) can be provided in a given area. The size (width) of conductive lines is a measure of process resolution (sometimes called "process geometry"). Present photolithographic techniques, limited as they are by the relatively long wavelengths of light, are limited to approximately 0.5 .mu.m. By using the techniques disclosed herein, which involve employing radiation of shorter or significantly shorter wavelengths than light, conductive lines that are smaller, and much smaller than 0.5 .mu.m, can be created on semiconductor devices. For example, lines having a width of less than "w" microns can readily be fabricated on semiconductor devices, where "w" is below 0.5, 0.4, 0.3, 0.2, 0.1, and smaller. Current densities in such "fine" lines evidently need to be controlled (or limited). In order to increase the current-carrying capability of such fine lines, it is also contemplated herein that the height (above the surface of the wafer) of such lines must be maximized. For example, a line having a width of 0.1 .mu.m can be formed that has a height "h" of 0.2 .mu.m, in which case the h:w "aspect ratio" of the line would be on the order of 2:1. According to an aspect of the invention, which is provided herein mainly as a "rule of thumb", fine conductive lines have a height:width (h:w) aspect ratio of at least "x" where "x" is 1.2:1, 1.3:1, 1.4:1, 1.5:1, 1.6:1, 1.7:1, 1.8:1, 1.9:1, 2:1, 2.5:1, 3.0:1, 3.5:1, 4.0:1, 4.5:1 or 5.0:1. Preferably, "x" is at least 2.0:1. In contrast to a technique that provides only narrow lines, without allowing for increased line density (see, e.g., U.S. Pat. No. 5,139,904, described above), the present invention allows for both finer lines and for packing fine lines closer together. In one embodiment of the invention, integrated circuit (semiconductor) lithography is performed using X-ray emissions as the actinic source. As described hereinabove, the use of short-wavelength radiation sources, such as X-rays, for semiconductor lithography is less wavelength resolution-limited than light for fabricating small-geometry (sub-micron cd) integrated circuits. The short wavelength (10.sup.-8 -10.sup.-11 meters) of X-ray radiation is well suited to providing the resolution required for the formation of very small sub-micron semiconductor features in semiconductor devices. Some efforts have been made to use X-ray lithography. However, such efforts are plagued with difficulties. Present image mask substrate materials include silicon carbide, polyimide, and silicon dioxide. These masking substrate materials suffer from a number of shortcomings relative to X-ray lithography. Among these are: 1) present X-ray image mask substrates exhibit extremely poor transparency to X-rays, yielding masks which provide poor contrast; 2) the poor transparency of present X-ray mask substrate materials forces the use of extremely thin substrates, resulting in very fragile image masks; 3) present X-ray mask substrate materials are subject to humidity-induced distortions, yielding image masks of poor stability, and causing unpredictable critical dimensions and feature positions in a resist material exposed by such image masks; 4) the expansion and adhesion characteristics of the opaque materials patterned on present mask substrates result in pattern-dependent distortions of the thin image mask substrates; and 5) present X-ray image mask substrate materials have poor "windows of transparency" (ranges of wavelengths for which they are transparent) which do not include the wavelength of the most desirable (shortest) X-ray wavelengths and the most intense available X-ray sources. Further, an image mask formed from these substrate materials is subject to an overall distortion within the "carrier" to which it is mounted due to different rates of thermal expansion between the image mask and the carrier. When heated, present X-ray image masks distort to an unacceptable level, thereby requiring exotic processing techniques. According to the invention, Beryllium metal (chemical symbol Be) is used as an X-ray image mask substrate. Beryllium has many desirable characteristics which make it quite suitable for use as an X-ray image mask substrate. These desirable qualities of beryllium include: 1) good transparency to X-rays; 2) viability of formation into thicker (than present) mask substrates due to its good X-ray transparency; 3) insensitivity to moisture-induced distortion; 4) excellent Young's modulus to resist distortion; 5) expansion characteristics more compatible with those of available opaque masking materials and carrier materials. 6) a wide window of transparency, permitting the use of high-intensity X-ray sources of the shortest wavelengths possible. According to the present invention, a Beryllium substrate is used as a planar image mask substrate upon which a patterned layer of X-ray opaque material is disposed. In conjunction with the use of a beryllium substrate for the X-ray exposure mask, the following materials make excellent X-ray opaqueing materials, for forming patterns on the image mask: gold (chemical symbol Au), tungsten (chemical symbol W), platinum (chemical symbol Pt), barium (chemical symbol Ba), lead (chemical symbol Pb), iridium (chemical symbol Ir), and rhodium (chemical symbol Rh). These materials have excellent opacity to X-rays, are relatively insensitive to moisture and are highly corrosion resistant. Further, these materials will adhere adequately to a beryllium substrate, thereby making the combination of these materials patterned on a beryllium substrate ideal for X-ray lithography of semiconductor devices. The thickness of the masking material can be empirically determined for each material (e.g., for gold) and based upon the parameters of the specific X-ray source employed. The mask pattern should be substantially opaque to the X-ray emissions, thereby providing high contrast on the image mask. Since, like beryllium, these masking materials (e.g., gold) are all metals, their thermal coefficients of expansion are highly compatible (substantially equal) to the thermal coefficient of expansion of the beryllium image mask substrate. This serves to reduce the amount of thermally-induced pattern-dependent mask distortion as compared to that experienced with present mask and substrate materials. The insensitivity of the inventive combination of masking materials and substrate to humidity serves to substantially eliminate pattern-dependent mask distortion due to humidity. According to the present invention, an X-ray mask is formed by providing a substantially planar beryllium substrate, and disposing upon the substrate a patterned opaque (to X-rays) layer of gold, tungsten, platinum, barium, lead, iridium or rhodium. The patterned layer may be disposed on either the "upstream" (towards the X-ray source) side of the beryllium substrate or on the "downstream" (towards the semiconductor wafer) side of the substrate. The mask is positioned in close proximity "upstream" of a sensitized wafer (a wafer with a layer of X-ray sensitive resist). The wafer is then exposed to an upstream X-ray source through the mask, causing a downstream "shadow" image of the mask to be formed on the surface of the resist. The close proximity of the image mask to the substrate sufficiently avoids dispersion of the illuminating radiation so that a "copy" of the image mask pattern is imaged onto the sensitized (e.g., with a layer of photoresist) wafer. The X-ray radiation is actinic with respect to the resist, and causes exposed areas of the resist to become chemically converted. After exposure to X-ray radiation, the unconverted areas of the resist (those areas of the resist which were "shadowed" by the mask) are chemically removed. (This described a "negative" mode of resist development. It is also possible to employ a "positive" resist chemistry, for which only the exposed areas of the resist are chemically removed). FIG. 2a is a cross-sectional view of semiconductor lithography apparatus, according to the present invention, showing an X-ray mask assembly 200a comprising an image mask 215a and a carrier 210 for the image mask 215a, according to the invention. A patterned opaque layer 230 of an X-ray masking material (e.g., gold), as described hereinabove, is disposed on the "upstream" surface (left hand surface, as shown in the Figure) of a planar beryllium substrate 220 to form the image mask 215a. The image mask 215a is fastened to the carrier 210 in any suitable manner. Preferably, the carrier 210 has a coefficient of expansion similar, or substantially identical, to that of the beryllium substrate 220. This is quite feasible since the beryllium itself is a metal. For example, the carrier 210 is also made of beryllium so that its expansion characteristics are identical to that of the beryllium substrate 220, effectively eliminating any thermally induced distortion of the mask 215a. The image mask 215a is positioned a "near field" distance "d" upstream of a front (left, as viewed in the Figure), sensitized (e.g., with photoresist) surface of a semiconductor wafer "W" such that the plane of the image mask 215a is substantially parallel to the front surface of the wafer "W". The sensitized wafer "W" has a layer of X-ray sensitive resist (not shown) disposed upon its front surface. The wafer "W" is exposed to actinic (relative to the resist) X-ray radiation 240, which first passes through the image mask. Preferably, the distance "d" is between 0.5 and 3.0 .mu.m, so that the image mask is sufficiently close to the front surface of the wafer to cause a pattern formed in the masking material 230 to be imaged onto the photosensitive material on the front surface of the wafer "W". At greater distances from the wafer, the imaging ability of the image mask would suffer, unless the irradiating energy were perfectly collimated. In contrast to the present invention, for photolithography using light as the irradiating source, employing a taking lens (see 124, FIG. 1) is effectively the only practical way of faithfully replicating a pattern from the image mask onto the surface of the wafer. FIG. 2b is a cross-sectional view of an alternate embodiment of an X-ray mask assembly 200b comprising an image mask 215b and carrier 210, according to the invention. A patterned opaque layer 230 of an X-ray masking material, as described hereinabove, is disposed on the "downstream" surface (right hand surface, as shown in the Figure) of a planar beryllium substrate 220 to form the mask 215b. As in the embodiment of FIG. 2a, the mask 215b is fastened to the carrier 210. In this case, however, this patterned opaque layer 230 is on the wafer side (downstream side) of the mask 215b. As described hereinabove with respect to FIG. 2a, the carrier 210 preferably has a coefficient of expansion similar to that of the beryllium substrate 220. As with the previous embodiment (of FIG. 2a), the image mask 215b is positioned a distance "d" upstream of the front surface of a sensitized semiconductor wafer "W" such that the plane of the mask is parallel to the front (planar) surface of the wafer. The sensitized wafer "W" has a layer of X-ray sensitive resist (not shown) disposed upon its front surface. The wafer "W" is exposed to actinic (relative to the resist) X-ray radiation 240 through the mask. Preferably, the distance "d" is between 0.5 and 3.0 .mu.m, so that the pattern formed by the masking material 230 is faithfully reproduced into the sensitized layer (e.g., of photoresist) on the semiconductor wafer "W". The near-field, afocal, imaging (shadow imaging) technique described hereinabove with respect to FIGS. 2a and 2b is analogous to photographic contact printing, in that the mask (analogous to a negative) is placed almost directly on the wafer (analogous to photographic paper) to faithfully reproduce an image from the image mask onto the wafer, without using optics (i.e, without focusing). In the case of semiconductor device fabrication, the image on the wafer creates a pattern in a photosensitive layer on the wafer, which through subsequent removal of all but the image portions of the photosensitive layer is used to create (e.g., by etching) patterns (e.g., lines) in underlying layers (not shown) on the wafer (such as an underlying layer of polysilicon). The term "photosensitive layer" is used herein to mean a layer of material that chemically reacts (converts) in the presence of actinic (chemical conversion causing) radiation, such as X-rays. Although the use of X-rays for semiconductor lithography is advantageous in terms of its inherent higher resolving ability (i.e., higher than light), inter alia, high-quality, fluent X-ray sources tend to be very expensive and consume a great deal of power. Hence, according to the present invention, gamma-rays can be employed (rather than X-rays) as illuminating sources for semiconductor lithography. As mentioned above, gamma-rays are even shorter in wavelength than X-rays. Hence, gamma-rays are even less resolution-limited than X-rays. Materials are available that are relatively transparent to gamma rays, and materials which are substantially opaque with respect to gamma rays. It is contemplated by this invention that these materials could be substituted for the materials described above, for making a gamma-ray based image mask for near field lithography (using a gamma ray source for illumination instead of an X-ray source). The structures and methods described herein with respect to FIGS. 2a and 2b, with such different materials, would be usable and are contemplated for gamma-ray, instead of X-ray lithography. In the main hereinbelow, the use of gamma-rays (radiation) for direct write lithography, rather than near field lithography is discussed. Generally, the use gamma-rays for lithography of integrated circuits has certain significant advantages, including: 1. extremely high source brightness (intensity), due to the extreme intensity of naturally occurring gamma-ray sources such as Cobalt-60, which is a passive source requiring no power; 2. high inherent resolution, due to short wavelength; 3. large depth of field, due to short wavelength; 4. resist materials can be fabricated (as discussed in greater detail hereinbelow) which exhibit a high cross section (absorptivity) with respect to gamma-rays, and the resist materials can be applied to a semiconductor wafer; 5. materials are available which are not only highly opaque to gamma rays, but which also emit secondary emissions (primarily in the form of photons) which can be used with conventional photoresist materials (as discussed in greater detail hereinbelow); 6. numerous chemistries are available which can be exploited to create gamma-ray sensitive resist materials (discussed in greater detail hereinbelow); 7. mechanisms are available (according to the present invention) for beam modification (as discussed in greater detail hereinbelow) and shuttering (as discussed in greater detail hereinbelow. The use of gamma-rays as an exposure (illumination) source for integrated circuit lithography depends, of course, on the use of suitable materials for the resist process. In lithographic processes, a resist layer on the surface of a semiconductor wafer is exposed through a mask to an actinic radiation (illumination) source. The resist responds, in the areas where it is exposed (illuminated), by chemically converting. The unconverted areas are then chemically removed. (This is a "negative" process. "Positive" processes also exist whereby the exposed areas are chemically removed. Both positive and negative type processes are contemplated.) The result is a patterned etch resistant layer. Conventional photolithography uses one of a number of organic resist materials (e.g., polystyrene, phenolformaldehyde, polyurethane, etc.). These materials are photo-sensitive (convert chemically when exposed to visible light radiation) and have good etch resist characteristics for subsequent wafer etching. These conventional organic resist materials are suitable for gamma-ray lithography in all respects except that their intrinsic absorbance of gamma-rays is very low. As is well known in the art, silicon integrated circuits, including CMOS circuits, have a finite tolerance to gamma-ray exposure. Excessive exposure to gamma-rays can cause field inversion in MOS FETs (Metal-Oxide-Semiconductor Field Effect Transistors) leading to excessive current leakage or outright device failure, and various other serious problems. It is necessary, therefore, to provide a gamma-sensitive resist material which either: a) requires very short exposure times, or b) provides a sufficiently high cross-section to gamma-rays (absorbs gamma-rays well enough) so that the underlying wafer is effectively "shielded" from the gamma radiation by the resist material itself. According to the present invention, it is recognized that gamma radiation, being an ionizing radiation, is capable of causing secondary emissions in various materials. For example, when tungsten is exposed to gamma radiation, gamma radiation is effectively completely absorbed by the tungsten, and the energy of the gamma radiation causes electrons to be "ripped" from their ordinary orbits (shells) in the atomic structure of the tungsten, thereby causing the tungsten to become ionized. In the process, the change of energy levels causes the ionized tungsten to emit photons of energy at a different wavelength (secondary emissions). As will be evident from the discussion hereinbelow, these secondary emissions are compatible with (actinic with respect to) essentially conventional photoresist materials. Tungsten, having a high inherent cross-section (absorptivity) with respect to gamma radiation, i.e., gamma radiation does not pass through it particularly well, makes an effective gamma radiation shield when present over the surface of a semiconductor device. For example, by applying a layer of tungsten over a conventional organic resist layer (on the upstream side of the resist), the secondary emission properties of the tungsten in response to gamma radiation may be utilized to "expose" the resist, while simultaneously shielding the underlying wafer from gamma radiation. The organic resist materials have a high cross-section to the secondary radiation. Various combinations of conventional photoresist materials with tungsten added into or onto the photoresist are disclosed hereinbelow (with respect to FIGS. 3a-3d), and for purposes of this discussion are termed "compound resists". They are all effectively converted when exposed to incident gamma radiation, and can all be formulated to effectively shield the underlying semiconductor device (if necessary). FIG. 3a is a cross-sectional view of a sensitized semiconductor wafer 300a illustrating an embodiment of a gamma-sensitive compound resist. A front surface "S" of a semiconductor wafer 310 is coated with a layer 320a of photoresist, for example conventional organic photoresist material such as polystyrene, phenolformaldehyde, polyurethane, etc.. Preferably, the layer of photoresist is applied as a planar layer, in any suitable manner. A layer 330a of a material with a high cross-section to gamma radiation, such as tungsten, boron or bromine, is disposed over the resist layer 320a. This may be referred to as a "secondary resist layer"--together the layers 320a and 330a forming a "compound resist" sensitizing the front surface "S" of the wafer 310 to incident gamma radiation. Preferably, the secondary resist layer 330a is formed as a film of tungsten, and is preferably of sufficient thickness to absorb all incident gamma radiation. However, the layer 330a must also be fairly thin, to allow secondary emissions to enter the underlying resist layer 320a. Gamma radiation 340, impinging upon the secondary resist layer 330a ionizes the material of the secondary resist layer 330a, causing scattered secondary photon emissions 345a having a different (generally longer) wavelength than that of the gamma radiation to be emitted into the resist layer 320a. The secondary emissions 345a are preferably of substantially shorter wavelength than visible light, and are employed to convert the underlying resist material 320a. In this manner, the lithography technique of the present invention is not as resolution bound as conventional photolithography. Preferably, the secondary emissions 345a are substantially shorter in wavelength than visible light, and are nevertheless capable of converting (acting actinically with respect to) the underlying photoresist 320a. The resist material 320a is highly absorbent of the secondary emissions 345a and thereby limits the exposure of the resist layer 320a to the secondary emissions 345a to a small area 360a about the point where the gamma radiation 340 strikes the secondary resist layer 330a. After chemically (or mechanically) stripping the secondary resist layer 330a, the unconverted areas of the resist layer 320a are removed, leaving an "island" of etch-resistant resist material over an area 350a of the semiconductor wafer 310. This describes forming a point feature on the wafer. A layer to be patterned, such as a layer of polysilicon, underlying the resist is omitted for illustrative clarity (in all of FIGS. 3a-3d). With a finely collimated gamma beam in fixed position, the wafer 310 can be "walked around" so that the beam 340 can describe and convert a line of photoresist. This would be a so-called "direct write" technique for semiconductor lithography. In the case that the gamma radiation were to impinge on the sensitized wafer 300a through a mask (not shown), two dimensional patterns could be formed directly on the photoresist 320a. This would be "near-field" semiconductor lithography (compare FIGS. 2a and 2b). As shown in FIG. 3a, the secondary emissions 345a tend to scatter, in other words be emitted in directions at an angle to the incident beam 340. This causes a limited amount of "blooming" (or de-focusing of a pattern through an image mask). However, the intensity and direction of the gamma radiation 340 cause the bulk of the secondary emissions 345a to be emitted substantially in the direction of travel of the incident gamma beam 340. Further, high absorbency of the resist layer relative to the secondary emissions limits the amount of blooming. Further, since the secondary emissions do not travel any significant distance, their divergence from the path of the incident gamma beam is relatively insignificant (they do not have an opportunity to go in the "wrong" direction for very far). By controlling the beam diameter (in direct write applications), or by adjusting the mask pattern (in near field applications), to compensate for any blooming, it is possible to accurately control the feature size (area 350a). Vis-a-vis direct write lithography techniques, evidently the photoresist can be patterned with a fineness--having a critical dimension (cd)--substantially approaching the diameter of the beam. Techniques for creating an extremely small diameter, collimated beam of gamma radiation are discussed hereinbelow. Another approach to making a gamma-sensitive resist from conventional organic resist materials is to make use of the same secondary emission property of a secondary material in a slightly different way. The compound gamma-resist material shown and described with respect to FIG. 3a was formed by depositing an overlying, upstream layer of a secondary emitter (330a) with a high cross-section to gamma radiation over the organic resist (320a). If instead the organic resist material is doped with the secondary emitter, a homogenous, gamma-sensitive, compound resist can be formed. FIG. 3b is a cross-sectional view of a sensitized semiconductor wafer 300b employing a doped type of gamma-sensitive compound resist. As in FIG. 3a, an underlying layer of material to form semiconductor features (such as a layer of polysilicon) is omitted for illustrative clarity. Beginning with a base of substantially conventional, preferably organic photoresist material 320b, the photoresist 320b is doped with particles of a material which will absorb gamma rays and emit secondary emissions. A representative particle 325b is illustrated, and functions in a manner similar to the layer 345a of FIG. 3a. Preferably, the doped, base resist layer 320b is a conventional organic resist material, such as polystyrene, phenolformaldehyde, polyurethane, etc., and it is doped with a secondary emitting dopant (e.g., dopant particle 325b) with a high cross-section to gamma radiation, such as tungsten, boron, or bromine. Gamma radiation 340 is shown impinging upon a representative secondary emitter dopant particle 325b, causing it to become ionized, resulting in scattered secondary photon emissions 345b having a different wavelength than that of the gamma radiation. The "base" (undoped) resist material is highly absorbent of the secondary emissions 345b and thereby limits the exposure of the resist layer 320b to the secondary emissions 345b to a small area 360b about the point where the gamma radiation 340 strikes the particle 325b. After exposure, the unconverted areas of the doped resist layer 320b are removed, leaving areas (points or lines) of converted compound resist over an underlying layer (e.g., polysilicon) on the front surface "S" of the wafer 310. The number of particles (e.g., 325b) required to be "mixed" into the base photoresist is determined by the intensity of the incident gamma radiation. The particles (e.g., 325b) may be uniformly distributed throughout the base photoresist, for example by mixing the particles into the photoresist material prior to applying the photoresist material to the surface of the wafer. On the other hand, the particles can be "implanted" into the surface of photoresist already applied to the surface of the wafer, in which case there will be a concentration gradient of particles more concentrated towards the surface of the photoresist (away from the wafer). Other gradients or non-uniform concentrations of particles are also contemplated. Although the base photoresist is most sensitive to the secondary emissions (345a, 345b), it bears mention that the base photoresist may also be somewhat sensitive to the direct effects of gamma-ray irradiation. However, it is preferred that the process parameters be adjusted so that little or no gamma radiation reaches the underlying semiconductor wafer 310. Hence, the dopant concentration (or thickness of the film 330a, in FIG. 3a), as well as the transparency of the photoresist with respect to gamma radiation, as well as the sensitivity of any underlying structures to gamma radiation must be taken into account when performing the semiconductor lithography techniques of the present invention. Again assuming that the gamma radiation reaches the sensitized wafer 300b through a mask (not shown), forming patterns of intense gamma radiation on the surface of the doped resist layer 320b, the scattering of the secondary emissions 345b causes a certain amount of "blooming" or de-focusing of the pattern, for the reasons described hereinabove. As before, the intensity and direction of the gamma radiation 340 cause the bulk of the secondary emissions 345b to be emitted substantially in the direction of travel of the gamma radiation. Even though the secondary emitter (particle 325b) is disposed within the organic resist as a dopant, it still has a high cross-section to gamma radiation and effectively shields the semiconductor wafer 310 from excessive exposure. Both of these embodiments of gamma-sensitive resist, i.e., the two-layer compound resist described with respect to FIG. 3a and the doped compound resist described with respect to FIG. 3b, provide the desired characteristics of sensitivity to gamma radiation and inherent gamma-ray shielding, thereby acting as an effective resist while preventing excessive exposure of the underlying semiconductor wafer to gamma rays. In the two-layer gamma-sensitive resist embodiment shown and described with respect to FIG. 3a, the secondary resist layer (330a) is not generally chemically sensitive to the incident radiation (340). It simply serves as a secondary emitter which serves to simultaneously block the incident gamma-radiation and to convert the incident radiation to another type of radiation which is actinic with respect to a resist layer (or base resist) and of which the underlying integrated circuitry (on wafer 310) is more tolerant. There are materials, however, suitable for use as a secondary emitter which are themselves chemically sensitive to exposure to gamma radiation. Further, some organic (and inorganic) resist materials are at least somewhat chemically sensitive to exposure to gamma radiation. Accordingly, it is possible to form multilayer gamma-sensitive resist coatings where the top layer (overlayer) provides secondary emissions and is also chemically converted by exposure to gamma-radiation. If the bottom layer (underlayer, between the overlayer and the wafer) is gamma-sensitive, then the use of an overlayer which does not completely block gamma radiation permits exposure of the underlayer by both direct (leaked through the overlayer) gamma radiation and by secondary emissions in the overlayer. Several benefits are derived from the use of multilayer gamma-sensitive resist. First, the use of dual chemistries in combination permits considerably greater flexibility and versatility in determining overall resist characteristics. Second, a multilayer resist tends to permit better planarization of the resist surface. (Planar layers in semiconductor devices are generally sought-after objectives.) It is well known in the art that a truly planar surface is more easily obtained in two steps (i.e., an extremely planar surface is easier to form on top of a surface which is already substantially planar) than in one step. Improved surface planarity of a resist coating tends to enhance the linewidth uniformity of patterns created by incident radiation. (Linewidth uniformity in semiconductor layers is a generally sought-after objective.) FIGS. 3c and 3d illustrate two embodiments of multilayer, gamma-sensitive, compound resists, according to the present invention. FIG. 3c is a cross-sectional view of a sensitized semiconductor wafer 300c employing a multi-layer gamma-sensitive resist (320c/330c). A front surface "S" of the semiconductor wafer 310 is coated with a primary resist layer 320c, which is sensitive to both gamma radiation and secondary emissions. On top of this primary layer 320c is disposed a secondary gamma-sensitive resist layer 330c of a material with a relatively high cross-section to gamma radiation, but which permits some gamma radiation to pass through it. Gamma radiation 340 impinging upon the secondary resist layer 330c ionizes the material of the secondary resist layer 330c, causing scattered secondary (photon) emissions 345c which enter the primary resist layer 320c with, generally, a different wavelength than that of the gamma radiation being emitted into the secondary resist layer 330c. A portion 340' of the incident gamma radiation 340 (indicated by dashed line and arrow) passes through the secondary resist 330c and into the primary resist 320c. The primary resist material is highly reactive to the secondary emissions 345c and thereby limits the exposure of the primary resist layer 320c to the secondary emissions 345c to a small area 360c about the point where the gamma radiation 340 strikes the secondary resist layer 330c. The primary resist 320c is also chemically sensitive to the "leaked" gamma radiation 340' (gamma radiation is actinic to the primary resist), a factor which enhances the chemical conversion of the primary resist 320c, improving contrast. After chemically "developing and stripping the unconverted areas of the primary and secondary resist layers 320c and 330c, respectively, an "island" of etch resist remains over an area 350c of the semiconductor wafer 310 as shown. Complete patterns may be formed on the resist layers, in the manner described above, and the resist pattern may be transferred to an underlying layer (not shown) as described above. FIG. 3d shows another arrangement of a sensitized semiconductor wafer 300d employing a multilayer gamma-sensitive resist. Again, a front surface "S" of semiconductor wafer 310 is coated with a primary resist layer 320d. Preferably, the resist layer 320d is a conventional organic resist material, such as polystyrene, phenolformaldehyde, polyurethane, etc.. A secondary gamma-sensitive resist layer 330d of a gamma-sensitive material with a high cross-section to gamma radiation is disposed over the resist layer 320d. Gamma radiation 340, impinging upon the secondary resist layer 330d simultaneously chemically convents and ionizes an area of the material of the secondary resist layer 330d, causing scattered secondary (photon) emissions 345d, generally having a different wavelength than that of the gamma radiation 340 being emitted into the secondary resist layer 330d. The secondary resist material 330d is highly absorbent of the secondary emissions 345d and thereby limits the exposure of the primary resist layer 320d to the secondary emissions 345d to a small area 360d about the point where the gamma radiation 340 strikes the secondary resist layer 330d. The high cross-section of the secondary resist layer 330d to gamma radiation prevents leakage of gamma radiation 340 through the secondary resist layer 330d into the primary resist layer 320d and the underlying wafer 310, thereby limiting the exposure of the wafer 310 to gamma radiation. After chemically developing and stripping the unconverted areas of the primary and secondary resist layers 320d and 330d, an "island" of etch resist remains over an area 350d of the semiconductor wafer 310 as shown. Patterning and processing is performed as described above. The compound resists described above with respect to FIGS. 3a-3d are useful for either near field or direct write lithography, both of which processes are afocal. Near field lithography utilizes an image mask in close proximity to the sensitized surface of the wafer, as described with respect to FIGS. 2a and 2b, and is preferably performed with X-ray radiation. Direct write lithography is preferably performed with gamma-rays, and requires a tightly focused or collimated beam of radiant energy directed to specific locations of a resist layer, thereby exposing and chemically converting those areas and forming patterns for processing lines and the like in layers underlying the resist layer. However, both X-ray and gamma-ray radiation may be used for either direct-write or near-field lithography, such combinations being contemplated herein as within the scope of the present invention. Because of the high inherent resolution capability of gamma radiation (short wavelength), the prospect of direct-write gamma lithography is very attractive. In order to accomplish this, however, it is necessary to provide, in addition to the gamma-sensitive resists described hereinabove, means for generating a tightly focused or collimated beam of gamma radiation, and means Suitable means for for "shuttering" or gating the beam. collimating and shuttering are described hereinbelow with respect to FIGS. 4 and 5a-c, respectively. According to the invention, a broad incident beam of radiation (or a radiant point source) can be concentrated and collimated, providing a very narrow, intense beam of radiation useful over a range of distances. This is accomplished by using a hollow, horn-shaped (or conical) afocal concentrator of the type schematically depicted in FIG. 4. FIG. 4 is a diagrammatic view of an afocal concentrator 400 for providing a very narrow, collimated beam of radiation. The afocal concentrator 400 has a tapered input (upstream) section 410 and an optional cylindrical output (downstream) section 450. The tapered section has a broad upstream mouth 420 (analogous to the bell of a trumpet) and a narrow opening 425 at an opposite downstream end thereof. The cylindrical section 450 has a diameter "do" equal to the diameter of the narrow opening 425, and is preferably formed contiguously therewith (i.e., the tapered and cylindrical sections are preferably formed as a unit structure). A relatively broad incident beam of radiation (e.g., gamma radiation) enters the mouth 420 of the tapered portion 410. (Such a beam could be generated by any suitable means including by a chemical radiant source, with or without a backing reflector.) The radiation beam is indicated by representative rays 440a and 440b entering opposite outer peripheral (circumferential) portions of the mouth 420. In the tapered section 410 as shown, the taper is approximately exponential, however any tapered form (e.g., a linear taper forming a conical shape), may be employed. An inner surface 415 of the afocal concentrator 400 is reflective of the incident radiation, and serves to reflect and concentrate the incident radiation towards the narrow opening 425 of the tapered section 410. For example, the tapered section could be formed of aluminum, nickel or chromium, or plated with the same on its inner surface (bore), to reflect and concentrate gamma rays. The cylindrical section 450, which should also have a highly reflective bore, serves to further collimate this concentrated radiation beam, providing an intense, narrow, collimated output beam 460 at an output end 455 thereof. In practice, the output beam 460 is not perfectly collimated and will diverge to some degree. However, over a first distance, d1, the output beam 460 remains roughly converged to within approximately the diameter `do` of the output end 455 of the cylindrical section 450 (or of the narrow opening 425 if the cylindrical section 450 is not used). Assuming a maximum useful (for direct write lithography) beam diameter "dm", the output beam 460 is useful over a distance of up to "d2" (d2>d1, as shown here) from the output end 450 of the cylindrical section (or of the narrow opening 425 if the cylindrical section 450 is not used). The longer the afocal concentrator 400, especially the longer the tapered portion of the concentrator, generally the better the collimation of the output beam 460 can be (i.e., long taper=less beam divergence). The dimension "d2" represents the useful effective (for direct write lithography) depth of field of the concentrator 400. In practice, for direct write semiconductor lithography, a radiant source, such as a pellet of Cobalt-60, is placed as close to the mouth of the concentrator as possible. As mentioned hereinabove, the emissions from the source can be directed more-or-less exclusively towards the mouth of the concentrator 400 by providing a reflector (compare 114, FIG. 1) upstream of the source (compare 112, FIG. 1). The resultant output beam 460 exiting the concentrator 400 is intense, adequately collimated, highly fluent (given a fluent source such as Cobalt-60), and very highly homogenous (minimal or negligible hot spots in the cross section of the beam 460 due to the many reflections experienced by the beam in the concentrator 400). Exemplary dimensions for the concentrator, for semiconductor lithography are: Preferably, the mouth of the concentrator is between 50 .mu.m and 60 .mu.m in diameter, and the output diameter "do" is less than 0.5 .mu.m in diameter. The output diameter "do" can be made as small as desired, for example 0.1 .mu.m, for converting extremely small areas of resist material on a semiconductor wafer (compare 360a-d in FIGS. 3a-d). Evidently, to form converted lines in the resist material, in a direct-write application, one or the other of the concentrator 400 or the semiconductor wafer must "walk around" in the plane of the wafer surface (or the appropriate differential angle may be utilized to target areas without perpendicular beam impingement). Given the relative complexities of walking around the concentrator or the wafer, it is preferred that the concentrator remain stationary and that the wafer be moved around in a plane (X-Y positioning). High resolution positioning platforms are available for "walking" the wafer around. Given a "passive" gamma source such as Cobalt-60, it is evident that a mechanism must be provided for gating (turning on and off) the output beam (460). Else, walking around the wafer would produce an endless line. A shutter mechanism for gating the output beam is described below with respect to FIGS. 5a-c. Although the rays 440a and 440b are shown in FIG. 4 as parallel rays entering the "bell" (mouth 420) of the concentrator 400, the rays of the incident beam need not be parallel (collimated). The concentrator will collimate input rays that are not parallel. However, the less parallel the input rays, the more collimation must be performed by the collimator. These factors need to be taken into account in the design of an overall lithography system. The better the initial collimation of the incident beam, the better the ultimate collimation of the output beam. One way to improve the input collimation is to position the source of the incident beam distant from the mouth of the concentrator (relative to the size of the concentrator. This serves to make the rays of the incident beam that actually enter the mouth of the concentrator more parallel with one another, thereby improving output collimation. If the source is positioned far away from the mouth of the collimator, it is also possible to use a cylindrical pipe (not shown) between the source and the mouth of the concentrator to help collect and direct radiation from the source. For example, a Cobalt-60 pellet could be placed in a closed end of a cylindrical tube, the closed end serving as an upstream reflector. The tube would be oriented coaxial to the concentrator, with its downstream open end placed adjacent the mouth of the concentrator. The inside surface of the tube would be highly reflective. In this manner, emissions of gamma radiation into the environment (other than towards the wafer) could be minimized. Although the instant application of the afocal concentrator 400 is to concentrate a gamma radiation beam, the same technique is applicable to any form of radiation source, including X-rays, UV light, and visible light. A major difference between such afocal concentrators for different radiation sources would be the material of which the inner surface of the concentrator (e.g., 415) is formed (or plated). The main requirement for the material of the inner surface of the concentrator is that it be reflective of the range of wavelengths in the incident beam. Aluminum is reflective of many different radiation wavelengths, including gamma radiation, and is suitable for gamma lithography. Nickel and Chromium are also suitably reflective materials. It is within the spirit and scope of the present invention that the afocal concentrator described hereinabove be applied to concentrate any suitable radiation source. The bore size of the concentrator also depends on the desired size of the output beam. Extremely small bore diameters can be formed by etching, ion milling, and other "machining" techniques which are known. Although the concentrator may be somewhat expensive to manufacture with precision, its cost will readily be amortized over the course of fabrication for a great number of semiconductor devices. And, as mentioned hereinabove, a passive source, such as Cobalt-60, provides a great deal of energy without consuming any external power. A shutter mechanism for gating the output beam from the concentrator is described hereinbelow with respect to FIGS. 5a-c. According to the invention, a surface acoustic wave (SAW) device operating as a shallow angle reflecting/scattering surface, can operate as a shutter element for X-ray or gamma-ray (or other) radiations. A thin, reflective film of, for example, aluminum, nickel, or chromium, is disposed over the surface of a Surface Acoustic Wave (SAW) device. Assuming that the SAW device is not activated, the reflective surface is substantially planar, and reflects any incident energy (e.g., a beam of gamma rays) at an angle equal and opposite to its angle of incidence. A tightly collimated beam approaching at a known shallow angle, will be reflected off of the reflective surface of the unactivated SAW device at a predictable shallow angle. If the SAW device is activated, however, the surface of the SAW device becomes distorted and deflects or scatters the incident beam. By providing a beam stop or an aperture, and positioning the beam stop or aperture such that radiation from the incident beam will pass the beam stop or aperture only when shallow-angle-reflected off of the surface of the SAW device, an effective shutter mechanism can be implemented. FIGS. 5a-c illustrate the operation of this Surface Acoustic Wave shutter device, as exemplary of a shallow-angle-reflection, distortable-surface shutter mechanism. SAW devices are generally known for other (than the shallow angle shutter disclosed herein) purposes, such as for imposing propagation delays on travelling waves, allowing particular wavefronts to be selectively "picked off" from the end of the SAW device. FIG. 5a is a side view of a shutter mechanism 500 employing a SAW device in its unactivated (planar, non-deformed surface) state. The shutter is formed of a Surface Acoustic Wave device 510 with a planar top surface upon which a reflective film 510a is disposed, and a strategically positioned beam-stop (or "knife edge") 520. A collimated (directional) incident beam 540 approaches the reflective surface 510a of the Surface Acoustic Wave device 510 at a shallow angle, and is reflected by the reflective surface 510a at an equal and opposite angle, forming a reflected beam 540a. The trajectory of the reflected beam is such that it misses the beam stop 520 and continues traveling along the same trajectory. FIG. 5b is a side view of the shutter mechanism 500 of FIG. 5a, with the Surface Acoustic Wave device in an activated (surface-deformed) state. Electrical stimulation of the Surface Acoustic Wave device causes surface waves 530 or distortions (shown greatly exaggerated) to be formed on the reflective surface 510b. For the same shallow angle incident beam 540 (compare FIG. 5a), these surface distortions 530 cause the reflected beam 540b to be scattered or diverted relative to the position of the reflected beam 540a from the unactivated Surface Acoustic Wave device 510. In other words, the beam 540b is reflected at a different angle off the distorted surface than the beam 540a is reflected from the undistorted surface of the SAW device. As a result, the reflected beam strikes the beam stop 520 and is blocked thereby such that the reflected beam does not exit the Surface Acoustic Wave shutter 500. By selectively energizing (activating) the SAW device, the beam 540 is effectively gated (turned on and off). This allows discrete lines (e.g., of converted resist material) to be formed on the surface of a semiconductor device. FIG. 5c is a greatly enlarged (magnified) view of a portion of the Surface Acoustic Wave shutter device 510, in the surface-distorted state shown in FIG. 5b, showing the point of reflection. As before, the distortions are shown greatly exaggerated. A reference line 550 indicates the location and angle of the undistorted surface (see, e.g., 510a). A tangent line 560 indicates the angle of the reflective surface 510b at the point of reflection. The incident beam 540, approaches the reflective surface 510b of the Surface Acoustic Wave device 510 at a shallow incident angle .THETA.ih relative to the horizontal (i.e., relative to the reference line 550). However, due to the distortion of the reflective surface 510b, the effective angle of incidence relative to the tangent line 560 is a steeper angle, shown as .THETA.i, where .THETA.i>.THETA.ih (as shown). As a result, the incident beam 540 is reflected as a reflected beam 540b at a reflection angle of .THETA.r=.THETA.i relative to the tangent line 560. The effective reflection angle .THETA.rh of the reflected beam 540b relative to the reference line 550 (horizontal plane) is even greater (as shown). By this mechanism, the incident beam 540 can be reflected such that it is either blocked or passed by a beam stop or aperture under electrical control. Note that the beam stop 520 effectively forms an "aperture" with the surface 510a of the Surface Acoustic Wave device 510. Alternatively, an aperture may be provided instead of a "knife-edge" style of beam stop 520. It is not necessary, according to the invention that the incident beam 540 be "cleanly" reflected in any particular direction. It is only necessary that the reflected beam 540b be reflected anywhere other than past the beam stop or aperture 520. It will readily be apparent to one of ordinary skill in the art that a magnetostrictive device may be substituted for the Surface Acoustic Wave device 510 to accomplish a similar result (directing and diverting a beam of incident radiation). Both magnetostrictive and Surface Acoustic Wave devices act as a sort of "surface distortion device" for the purposes of the present invention. Any device that can reflect an incident beam at one angle in one energized state (e.g., not energized), and reflect an incident beam at another angle in another energized state (e.g., energized), in conjunction with a beam stop or aperture allowing a reflected beam to pass only at a critical angle, can be employed in the instant inventive shutter mechanism. An advantage of using a surface distortion device, rather than a device which must physically be positioned, is that the response time of such surface distortion devices is relatively quick. This enables such a device, in conjunction with a beam stop or aperture, to be used as a high-speed shutter (e.g., no moving parts). It will also be readily apparent to one of ordinary skill in the art that this type of shutter device may be applied to radiation of a variety of wavelengths, including gamma-rays, X-rays, UV light, etc.. It is within the spirit and scope of the present invention that the SAW (or magnetostrictive) shutter device described hereinabove be applied as a high-speed shutter to any suitable form of radiation beam. FIGS. 6a and 6b are block diagrams of (gamma-ray) direct-write lithographic apparatus employing the techniques described hereinabove with respect to FIGS. 3a-d, 4, and 5a-c; or for a near-field lithographic apparatus employing the techniques described hereinabove with respect to FIGS. 2a-b, 3a-d 4 and 5a-c. FIG. 6a is a block diagram of a direct-write (or energy source for near-field; gamma-ray, x-ray or other radiation) lithography (lithographic) apparatus 600a, according to the present invention. A radiation source 610 provides a source of intense directional gamma-ray (or x-ray, or other radiation, collectively herein called "gamma-ray") radiation. A suitable passive gamma radiation source is Cobalt-60 which passively radiates intense gamma-ray radiation. A reflector (such as that shown and described as 114 with respect to FIG. 1) may be employed to improve the directionality and intensity of the source 610. Given any shutter 630, or other "on/off" mechanism, the beam 640 need not be very well collimated. Given a shutter 630 such as was described with respect to FIGS. 5a-5c, a collimator similar to that shown in FIG. 4 could be employed to direct the beam 640 into the shutter 630. Incident gamma-ray radiation 640 from the gamma-ray radiation source 610 enters a shutter device 630, such as the Surface Acoustic Wave shutter device shown and described as 500 with respect to FIGS. 5a-5c. The shutter device 630 serves to selectively gate (block or pass) the incident beam 640, resulting in a controlled gamma-ray beam 640a. The controlled gamma-ray beam 640a enters the mouth of an afocal concentrator 620, such as that shown and described (400) with respect to FIG. 4. The afocal concentrator narrows and intensifies the controlled beams 640a to provide a collimated beam 640b. A semiconductor wafer 650 is positioned a distance "d3" from the output of the afocal concentrator 620 such that the collimated beam impinges upon the front surface 655 thereof. The front surface 655 of the wafer 650 is coated with a layer 660 of gamma-sensitive resist, such as that shown and described as 320a-d (and optionally 330a, c-d) with respect to FIGS. 3a-3d, respectively. Preferably, the wafer is mounted to a movable carriage (not shown) for direct-write application, by which means the wafer 650 may be positioned such that the collimated beam 640b may be caused to impinge on any point on the resist layer 660. The on/off state of the collimated beam 640b may be effectively controlled by selectively activating and de-activating the shutter device 630. Preferably, the distance "d3" is approximately 5 .mu.m. Preferably, the distance "d3" should be no greater than the distance "d2" shown in FIG. 4. FIG. 6b is a block diagram of an alternate embodiment of a direct-write gamma-ray (or x-ray or similar radiation, collectively herein called "gamma-ray") lithographic apparatus 600b, according to the invention. As before, a radiation source 610, such as Cobalt-60, provides a source of intense directional gamma-ray radiation. A reflector (such as that shown and described as 114 with respect to FIG. 1) may be employed to improve the directionality and intensity of the source 610. Incident gamma-ray radiation 640 from the gamma-ray radiation source 610 enters the mouth of an afocal concentrator 620, such as that shown and described (400) with respect to FIG. 4. The afocal concentrator narrows and intensifies the incident gamma-ray radiation 640 to provide a collimated gamma-ray beam 640a'. The collimated gamma-ray beam 640a' enters a shutter device 630, such as the Surface Acoustic Wave shutter device shown and described as 500 with respect to FIGS. 5a-5c. The shutter device serves to selectively gate (block or pass) the collimated gamma-ray beam 640a', resulting in a collimated, controlled gamma-ray beam 640b'. A semiconductor wafer 650 is positioned a distance "d3" from the output of the afocal concentrator 620 such that the controlled collimated beam 640b' impinges upon the front surface thereof. The front surface 655 of the wafer 650 is coated with a layer 660 of gamma-sensitive resist, such as that shown and described as 320a-d (and optionally 330 a,c-d) with respect to FIGS. 3a-3d, respectively. Preferably, the wafer 650 is mounted to a movable carriage (not shown) for direct-write application, by which means the wafer 650 may be positioned such that the collimated controlled beam 640b' may be caused to impinge on any point on the resist layer 660. The on/off state of the collimated controlled beam 640b' may be effectively controlled by selectively activating and de-activating the shutter device 630. In this configuration, the shutter device 630 is between the output of the concentrator 620 and the wafer 650. Hence, the shutter 630 must be made small. Small SAW (or magnetostrictive) devices can be fabricated to meet this criteria. It is also possible (in any of the examples set forth herein) that the beam may be reflected off a suitable reflecting surface (not shown) so that it initially approaches the wafer 650 at an angle (e.g., parallel, or between parallel and normal) to the surface of the wafer) and is reflected by the reflector to ultimately impact the wafer at ninety degrees (normal) to the surface of the wafer. It is within the spirit and scope of the present invention that the inventive techniques described hereinabove be used either alone or in combination. By employing these techniques, viable forms of short-wavelength (e.g., gamma-ray or X-ray) afocal lithography may be realized. It should also be recognized that many of the techniques described hereinabove may be applied to other types of radiation, such as UV light or visible light.
047524413	abstract	An array of plural rod guides each housing therewithin corresponding pluralities of rods, as disposed in parallel axial relationship within the cylindrical sidewall of the inner barrel assembly of a pressurized water reactor, defines a plurality of peripheral regions between the inner circumferential surface of the cylindrical sidewall and the peripheral edges of the array. A plurality of modular formers is installed within the respective plurality of peripheral regions, at each of one or more banks of predetermined, respective elevations within the inner barrel assembly. Each modular former comprises a pair of parallel former plates having arcuate outer edges corresponding to the interior circumference of the cylindrical sidewall and chordlike inner edges contoured to correspond to the juxtaposed peripheral edge of the array, and a vertical column extending between and rigidly interconnecting the associated plates of a pair. Attachment means secured to the surfaces of the plates extend beyond the respective outer arcuate edges thereof and through holes provided therefore in the cylindrical sidewall and are welded thereto from the exterior surface of the sidewall. One more banks of modular formers, as required, establish the proper pressure drop of the core output flow from the lower barrel assembly so as to approach an axial flow condition within the inner barrel assembly, reducing turbulance and minimizing vibration in operation.
	abstract	A method is provided for characterizing spectrometric properties (e.g., peak reflectivity, reflection curve width, and Bragg angle offset) of the Kα emission line reflected narrowly off angle of the direct reflection of a bent crystal and in particular of a spherically bent quartz 200 crystal by analyzing the off-angle x-ray emission from a stronger emission line reflected at angles far from normal incidence. The bent quartz crystal can therefore accurately image argon Kα x-rays at near-normal incidence (Bragg angle of approximately 81 degrees). The method is useful for in-situ calibration of instruments employing the crystal as a grating by first operating the crystal as a high throughput focusing monochromator on the Rowland circle at angles far from normal incidence (Bragg angle approximately 68 degrees) to make a reflection curve with the He-like x-rays such as the He-α emission line observed from a laser-excited plasma.

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