Split (2)

train · 33.5k rows

patent_number stringlengths 0 9	section stringclasses 4 values	raw_text stringlengths 0 954k
	description	This claims the benefit to U.S. Provisional Patent Application No. 62/024,348, filed on Jul. 14, 2014, which is hereby incorporated by reference herein. The present disclosure relates generally to a filtered containment venting system, and more specifically to a dry filtered containment venting system including metal fiber filters and molecular sieves. In the unlikely and hypothetical situation of a beyond design basis event or a severe accident at a nuclear plant, the pressure within the nuclear reactor containment building could build up causing a potential for leakage or even containment failure. A filtered containment venting system (FCVS) allows for the release of the over-pressure while retaining fission products. FCVSs historically have been provided in two general categories—wet and dry. A wet FCVS uses a water solution as the primary method of capturing radioactivity. With a dry FCVS, no water is required to capture radioactivity. Dry FCVSs have a more simple design and have less pressure drop than wet FCVSs. However, dry FCVSs historically have issues with decay heat limitations and plugging potential. A dry FCVS for a nuclear reactor containment is provided. The dry FCVS includes a housing and a round and/or elongated aerosol filter inside the housing for removing contaminant aerosols from gas passing through the housing during venting of the containment. The housing includes at least one inlet portion configured for directing gas into the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The dry filtered containment venting system may be arranged and configured such that when a flow of gas through the outlet portion is closed off at least one of convective, radiant and conductive heat transfer removes decay heat of aerosols captured in the aerosol filter. The FCVS according to one aspect of the invention, to be used outside containment, may include an inlet portion and an outlet portion. The inlet portion includes a tube or pipe that expands into a bowl-like structure with a tubesheet opposite the inlet pipe. The outlet portion has a similar design. The tubesheet can have an internal chimney to allow for better heat removal. Alternatively, headers could be used in place of the tubesheets. A number of tubes extend between the inlet and outlet tubesheets/headers. Each of these pressure tubes may include an aerosol filter, preferably a metal fiber filter (MFF), and an iodine filter, preferably a molecular sieve (MS). The pressure tubes are positioned in a spaced arrangement, allowing air flow therebetween. This allows for radiant, conductive and/or convective heat transfer to remove decay heat and prevent the MFF and MS from reaching unsafe temperatures. Additionally, an air gap may be provided between the outside of the filter and the interior surface of the pressure tube. The air gap is sized to ensure optimal heat transfer is achieved, while being large enough to ensure that the process flow going into the filter is not sub-cooled. In addition to being to the air gap being equal to one third of the MFF diameter, the present design is such that the area of the hot surface (discharging heat) is less than the area of the cold surface (receiving heat). The FCVS according to another aspect of the invention, to be used inside containment, may include an aerosol filter for removing contaminants from gas passing therethrough during venting of the containment, a lower inlet portion for directing gas upward into a lower section of the aerosol filter during the venting of the containment, an upper inlet portion for directing gas downward into an upper section of the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The lower inlet portion and the upper inlet portion is arranged such gas flows in through the lower inlet portion upward through the lower and upper sections of the aerosol filter and out through the upper inlet portion when a flow of gas through the outlet portion is closed off so as to allow a forced convective cooling of the decay heat of aerosols captured in the aerosol filter, via a chimney effect. Some embodiments of the present invention are directed to a dry FCVS having a pressure tube design. Typical dry FCVSs reach elevated temperatures due to the collected radioactivity, which creates heat called “decay heat.” This decay heat can elevate the filter surface temperature to 270° C. or more, which is greater than the melting temperature of CsOH, an aerosol that is produced during severe accidents. As a result, the melted aerosol can coat the filter and block the flow (called filter cake melting). Additionally, the temperatures that can be reached by typical dry FCVSs are as high as 550° C., well above the hydrogen auto-ignition temperature, which could result in a fire or detonation. Furthermore, typical dry FCVS designs place the cooling pipes in the flow of the exiting gas. This can sub-cool the gas, making it wet, and compromising the filter and molecular sieve efficiencies. Another concern with typical dry FCVSs is that there is no way to clean the filters in place. Thus, these known FCVSs have limits to aerosol loading. Embodiments of present invention may provide improved dry FCVSs that are not subject to one or more of these shortcomings. In some embodiments, the FCVS includes a pressure tube design, which may remove heat more effectively. Furthermore, the embodiments of the pressure tube design may not cool within the flow path, so the filter efficiency is not compromised. Additionally, passive pressure pulsing can be added to the MFFs, which can keep them from plugging and allow for operation into molten concrete-corium interaction where dust loading can be very high. In another embodiment, the FCVS includes a chimney design to convectively remove decay heat when the FCVS is not venting. FIG. 1 shows a cross-sectional view of a dry FCVS 10 in accordance with an embodiment of the present invention. FCVS 10 includes a housing 11 having an inlet portion 12, an outlet portion 14 and pressure tubes 28. Inlet portion 10 includes an inlet tube or pipe 16 that expands into a bowl-shaped manifold 18 holding a tubesheet 20 opposite inlet pipe 16. Outlet portion 14 has a similar design, including an outlet tube or pipe 22 that expands into a bowl-shaped manifold 24 holding a tubesheet 26 opposite outlet pipe 22. A number of pressure tubes 28 extend between inlet and outlet tubesheets 20, 26 and together define a cylindrical shape that is axially sandwiched between tubesheets 20, 26. Each of these pressure tubes 28 houses a round elongated aerosol filter in the form of a cylindrical MFF 30 and an iodine filter in the form of a MS 32. In preferred embodiments, MFF 30 is positioned on an inlet side 34 of each pressure tube 28 and MS 32 is positioned on an outlet side 36 of each pressure tube 28. Pressure tubes 28 are positioned in a spaced arrangement, allowing spaces 38 for air flow therebetween and convective heat transfer with ambient air 39. The spaced arrangement allows for radiant and convective heat transfer to remove decay heat 50 and prevent MFF 30 and MS 32 from reaching unsafe temperatures. FCVS 10 also includes a schematically shown air source 60 for providing air to remove decay heat from MFF 30 when a flow of gas through outlet portion 14 is closed off, for example via a controllable valve 62 provided in outlet pipe 22. The air source 60 provides convective air flow through pressure tubes 28 to remove the decay heat of radioactive aerosols captured in MFF 30. Air source 60 may be connected to a cooling inlet port 64, which may be opened and closed by a controllable valve 66, formed in inlet portion 12 at manifold 18. Outlet portion 14 may also include a cooling outlet port 68, which may be opened and closed by a controllable valve 70, provided at manifold 24. When the outlet of pipe 22 is closed by valve 62, ports 64, 68 may be opened by respective valves 66, 70 such that cooling air from air source 60 flows into inlet portion 12, through pressure tubes 28 and out of outlet portion 14 via cooling outlet port 68. FCVS 10 is arranged and configured such that when a flow of gas through outlet portion 14 is closed off convective, radiant and conductive heat transfer removes decay heat of aerosols captured in MFFs 30. Convective heat transfer occurs via the outer surface of pressure tubes 28 and the surrounding air, radiant heat transfer occurs between MFFs 30 and pressure tubes 28 and conductive heat transfer occurs by conducting decay heat from MFFs 30 to pressure tubes 28. For the radiant heat transfer, in contrast to conventional systems, the cold surface area of each of pressure tubes 28, formed by the inner surface of the pressure tube 28, is greater that the hot surface area of each of pressure tubes 28, formed by the outer surface of the MFF 30, such that decay heat radiates from MFF 30 to pressure tubes 28. In other words, the design of FCVS 10 is such that relative surface area of the cold to hot surfaces for heat transfer is greater than 1. The other pressure tube embodiments—FCVSs 110, 210, 310—may be similarly designed. FIGS. 2a to 2d illustrate more detailed views of pressure tubes 28 shown in FIG. 1. FIG. 2a shows an enlarged view of a cross-section of one of pressure tubes 28 shown in FIG. 1 illustrated a gas flow therethrough, FIG. 2b shows a view of pressure tube 28 along the same cross-section as in FIG. 2a, FIG. 2c shows a cross-sectional view of pressure tube 28 along A-A in FIG. 2b and FIG. 2d shows a perspective view of inlet side 34 of pressure tube 28. Air gaps 40 may be provided between an outer surface 42 of MFF 30 and an inner surface 44 of pressure tube 28. Air gaps 40 are delimited circumferentially between spacers 43, which extend radially between outer surface 42 of MFF 30 and inner surface 44 of pressure tube 28, and are sized to ensure additional conductive heat transfer is achieved, while being not too large to ensure that the process flow going into MFF 30 is not sub-cooled. Spacers 43 may be formed of metal for conductive heat transfer by conducting decay heat from MMFs 30 to pressure tubes 28. As schematically shown in FIG. 2a, contaminated inlet gas 46 may enter inlet side 34 of pressure tube 28, enter air gaps 40 and pass radially into MFF 30 for aerosol filtering. The aerosol filtered gas exiting MFF 30 then passes through a hole 45 in a barrier 46 separating MFF 30 and MS 32, then enters MS 32 for iodine filtering. The aerosol filtered and iodine filtered outlet gas 48 then exits MS 32 and outlet side 36 of pressure tube 28 to enter into manifold 24 for merging with aerosol filtered and iodine filtered outlet gas leaving other tubes 48. The aerosol filtered and iodine filtered outlet gas 48 next exits outlet pipe 22 and is released outside of the nuclear reactor containment. As schematically shown in FIG. 2a, decay heat 50 is released to spaces 38 and ambient air 39 for convective heat transfer. MFF 30 captures fission products that would otherwise be vented outside of the containment building. MFF 30 may be formed of stainless steel sintered metal fibers. In one preferred embodiment, MFFs 30 are commercially available cartridges, lowering cost and allowing for easy installation and removal. For example, SINTERFLO sintered stainless steel filter cartridges from Porvair Filtration Group may be used. Alternative materials for the cartridge-type filters may also be utilized. MS 32 may be a cartridge filled with a media that absorbs iodine. For example, the media may be a zeolite coated with silver. The silver reacts with the iodine present in the vent gasses to capture the iodine and prevent it from being exhausted outside the containment building. The use of commercially available sieve media allows for a lower cost for the filtering hardware. If iodine capture is not required by an end user, the molecular sieve portion may be eliminated. By using multiple MFF 30/MS 32 sets, each in a respective pressure tube 28, each individual MFF 30/MS 32 set has its own pressure boundary, delimited by pressure tube 28, that is exposed to the ambient air. In this way the heat transfer to address decay heat does not have to pass across to the inside of a pressure vessel (with less ratio of surface area), as with conventional dry FCVS designs. Pressure tubes 28 each have sufficient surface area to expel the required heat. By providing several relatively smaller tubes, pressure tubes 28 are advantageously thin and still able to handle the same pressure as an equivalent thicker pressure vessel. Pressure tubes 28 can be sized based on plant configuration and to accommodate the desired heat transfer. A preferred inner diameter size for pressure tubes 28 is approximately 2 inches to approximately 10 inches, with a nominal inner diameter of 4 inches being more preferred. The wall thickness of pressure tube 28 is a function of diameter and pressure. With the 4 inch nominal inner diameter, 1/16 inch would be a preferred nominal wall thickness. FCVS 10 may allow for higher pressure operation than other dry systems that use HVAC-type enclosures. A typical HVAC FCVS uses a square casing and has an orifice plate before the system that drops the pressure to atmospheric, requiring a larger filter area since the steam/air mixture has expanded in volume. The small diameter pressure tubes 28 of the pressure tube FCVS 10 can be thin and still be able to handle the pressure, which is spread across the plurality of pressure tubes 28. Furthermore, in event of a hydrogen burn pressure spike, FCVS 10 may easily handle the pressure spike whereas an HVAC-type enclosure may fail. The integral MFF 30 and MS 32 in each pressure tube 28 eliminates the need for two separate vessels/enclosures—one for the MFF and another for the MS—of other dry FCVS designs. The pressure tube design of FCVS 10 allows for passive decay heat removal including the high decay heat load of multi-unit power plants. Decay heat is from radioactive decay of captured aerosols and iodine. Each MFF 30 is close to the respective pressure tube 28, and the pressure tube 28 is indirect communication with the ambient environment, so the path for heat transfer is short. With the pressure vessel design of known dry FCVSs, the heat must make it all the way to the pressure vessel surface and there is limited surface area. For known HVAC-type designs, the required enclosure is large with relatively little effective surface area, while cooling tubes are positioned within the process flow, which can sub-cool the flow. Thus, FCVS 10 is completely passive with no requirement to add water or chemicals. Plugging potential is significantly reduced by the increased surface area of the MFFs 10 and potential use of pressure pulse technology. The decay heat removal capability keeps the temperature below the auto-ignite temperature of hydrogen and also below the melting point of hydroscopic aerosols. The air gap and geometry are designed to ensure that during normal operation the heat loss does not impact performance, but during idle venting periods, the heat built up from decay heat can be released via a combination of radiant heat due to the higher temperature as well as natural convection cells created in the stagnant tubes (that is, pressure tubes that are not being used during an idle period), as well as conductively removed through spacers 43. The relatively small size of pressure tubes 28 allows for the possibility of cleaning MFFs 30 and MSs 32 in place. A nitrogen bottle system can be added to back purge pressure tubes 28 with a pressure pulse for less than 0.5 second to reverse clean the filter. FIG. 3 shows a partial cross-sectional view of an inlet portion 112 of a dry FCVS 110 according to another embodiment of the present invention. FCVS 110 may be configured in the same manner as FCVS 10 downstream of tubesheet 20. Inlet portion 112 includes a sump 114 for collecting aerosols from the pressure pulse backwashing process, which occurs by pulsing gas into the outlet side of each pressure tube 28. FIGS. 4a to 4c show a dry FCVS 210 in accordance with another embodiment of the present invention. FIG. 4a shows a cross-sectional view of FCVS 210, FIG. 4b shows a perspective view of FCVS 210 and FIG. 4c shows a cut-away perspective view of FCVS 210. FCVS 210 is formed in the same manner as FCVS 10, except that FCVS 210 is designed for internal tube nest cooling. More specifically, FCVS 210 includes a housing 211 having an inlet portion 212, an outlet portion 214 and pressure tubes 28. Inlet portion 212 includes an inlet tube or pipe 216 that expands into an annular manifold 218 holding an annular tubesheet 220 opposite inlet pipe 216. An inlet cooling tube 219 is imbedded in inlet portion 212 and passes through manifold 218 and tubesheet 220. Outlet portion 214 has a similar design, including an outlet tube or pipe 222 that expands into an annular manifold 224 holding an annular tubesheet 226 opposite inlet pipe 216. An outlet cooling tube 227, which forms an internal chimney, is imbedded in outlet portion 214 and passes through manifold 224 and tubesheet 226. Ambient air 39 enters into inlet cooling tube 219 and enters into an interior air space 240 formed within FCVS 210 along a center axis CA thereof. Ambient air flow 242 also enters radially toward center CA for convective removal of decay heat. The cooling air then passes out outlet cooling tube 227. As shown in FIG. 4a, but omitted from FIGS. 4b and 4c, FCVS 210, similar to FCVS 10, also includes a schematically shown air source 260 for providing air to remove decay heat from MFF 30 when a flow of gas through outlet portion 214 is closed off, for example via a controllable valve 262 provided in outlet pipe 222. The air source 260 provides convective air flow through pressure tubes 28 to remove the decay heat of radioactive aerosols captured in MFF 30. Air source 260 may be connected to a cooling inlet port 264, which may be opened and closed by a controllable valve 266, formed in inlet portion 212 at manifold 218. Outlet portion 214 may also include a cooling outlet port 268, which may be opened and closed by a controllable valve 270, provided at manifold 224. When the outlet of outlet pipe 222 is closed by valve 262, ports 264, 268 may be opened by respective valves 266, 270 such that cooling air from air source 260 flows into inlet portion 212, through pressure tubes 28 and out of outlet portion 214 via cooling outlet port 268. FIGS. 5a and 5b show a dry FCVS 310 according to another embodiment of the present invention. FIG. 5a shows a partial cross-sectional view of an inlet portion 312 of a dry FCVS 310 and FIG. 5b shows a perspective view of FCVS 310. Pressure tubes 28 may be arranged horizontally as shown in FIG. 5a or vertically as shown in FIG. 5b. In contrast to the FCVS 10, FCVS 310 includes an inlet header 320 at inlet portion 312 in place of tubesheet 20 and an outlet header 326 at outlet portion 314 in place of tubesheet 26. Headers 320, 326 are connected to respective inlet and outlet pipes 316, 322 and are cylindrically shaped. Pressure tubes 28 extend from one side of each header 320, 326 along substantially the entire length of headers 320, 326. Headers 320, 326 extend longitudinally in a direction 350 that is perpendicular to a direction 352 in which pressure tubes 28 extend longitudinally. Headers may also be used in place of a tubesheets in FCVS 10, 110 and 210 and the pressure tubes in such embodiment may also be arranged horizontally instead of vertically. Similar to FCVSs 10, 210, in a preferred embodiment, FCVS 310 is configured with additionally cooling ports, valves and an air source to provide convective air flow through pressure tubes 28 to remove decay heat from MFF 30 when a flow of gas through outlet portion 314 is closed off. FCVSs 10, 110, 210, 310 are configured for use outside of a containment building, or in a containment innerspace, as discussed for example below with respect to FIGS. 7a, 7b. FCVSs 10, 110, 210, 310 may allow for lighter construction than other (wet or dry) pressure vessel designs, that may require two heavy pressure vessels—one for filter and one for sieve. FCVS 10, 110, 210, 310 may also allow for easier modularization, by adding more channels to get the required flow and/or making several groups of pressure tubes 28 per tubesheets and/or headers. FIGS. 6a and 6b show cross-sectional view of a dry FCVS 410 in accordance with another embodiment of the present invention. FIG. 6c shows a cross-sectional view at A-A in FIG. 6a. Instead of utilizing the pressure tube design of FCVSs 10, 110, 210, 310, FCVS 410 has a chimney design for convective transfer of decay heat and is configured for use inside of a containment 420. FCVS 410 includes a dual inlet housing 411, which in this embodiment is formed of metal, having two inlet portions 412, 413 and an outlet portion 414. Lower inlet portion 412 includes an inlet tube or pipe 416 that expands into a manifold 418 and upper inlet portion 413 includes an inlet tube or pipe 417 that expands into a manifold 419. Inlet portion 412 is positioned vertically below outlet portion 413. Outlet portion 414 includes an outlet tube or pipe 422 arranged horizontally and for receiving aerosol and iodine filter gas from an outlet manifold 424 and exhausting gas out of containment 420. A plurality of round elongated aerosol filters in the form of longitudinally horizontally extending cylindrical MFFs 430 are arranged inside housing 411 vertically between inlets 412, 413, i.e., above lower inlet portion 412 and below upper inlet portion 413. An iodine filter in the form of a MS 432 is also arranged inside housing 411 horizontally between MFF 430 and outlet portion 414. In the exemplary embodiment shown in FIGS. 6a to 6c, as shown in FIG. 6c, FCVS 410 includes twenty-five MFFs 430 arranged in a five column, five row square arrangement in a spaced manner such that MFFs 430 are arranged distances from each other by space 435. In other embodiments, different numbers of MFFs 430 may be used and MFFs 430 may be arranged in different geometries. A flow of gases through FCVS 410 during normal venting of containment 420 is illustrated in FIG. 6a. During the venting of containment 420, two contaminated inlet gas streams 434, 436 may enter into FCVS 410 at the same time, flow through MFFs 430 and MS 432 and then exit FCVS 410 outside of containment 420. A first inlet gas stream 434 flows upwardly into lower inlet portion 412 and a second inlet gas stream 436 flows downwardly into upper inlet portion 413. A higher temperature and pressure of ambient air 438 inside containment 420 compared to an ambient air 440 outside containment 420 causes inlet gas streams 434, 436 to enter into FCVS 410 and exit FCVS 410 at outlet portion 414 into air 440 outside of containment 420. Lower inlet portion 412 is arranged for directing contaminated gas stream 434 upward into a lower section 442 of MFFs 430 during the venting of containment 420 and upper inlet portion 413 is arranged for directing contaminated gas stream 436 downward into an upper section 444 of MFFs 430 during the venting of containment 420. Gas stream 434 flows upward through inlet pipe 416 into manifold 418 and through a lowermost or bottom surfaces 446 of MFFs 430 while gas stream 436 flows downward through inlet pipe 417 into manifold 419 and through an uppermost or top surfaces 448 of MFFs 430. Contaminated gas entering into MFFs 430 passes through cylindrical outer surfaces 450 of MFFs 430. Filter 450 remove aerosol particles from the contaminated gas stream and define channels 452 therein for the flow of aerosol filtered gas 454. The aerosol filtered gas 454 then flows longitudinally with respect to channels 452 and horizontally out of channels 452 into the directly adjacent MS 432. At longitudinal ends 433 of MFFs 430 adjacent to MS 432, MFFs 430 are embedded in a tubesheet 431 that limits the airflow into MS 432 to the aerosol filtered gas 454. The aerosol filtered gas flowing horizontally through MS 432 is iodine filtered and then flows horizontally through manifold 424 and outlet pipe 422 to join ambient air 440 outside of containment 420. FIG. 6b illustrates a flow of gases through FCVS 410 when FCVS 410 is not venting, i.e., when a flow of gas through outlet portion 414 is closed off. When FCVS 410 is not venting, the pressure difference between ambient air 438 and ambient air 440 is not present and gas is not sucked downward through inlet portion 413 into housing 411. However, due to the temperature difference between inlet portion 412 and inlet portion 413, gas stream 434 enters inlet portion 412, passes through spaces 435 between MFFs 430 and exits inlet portion 413. Gas stream 434 enters upwardly between MFFs 430 at lower section 442 and out of upper section 444. More specifically, gas stream 434 enters upwardly into inlet pipe 416, through manifold 418 and past lower surfaces 446 of MFFs 430, then through spaces 435 between MFFs 430 and past of upper surfaces 448, into manifold 419 and out of outlet pipe 417. Lower inlet portion 412 and upper inlet portion 413 are accordingly arranged such gas flows in through lower inlet portion 412 upward past the lower and upper sections 442, 444 of MFF 430 and out through upper inlet portion 413 when a flow of gas through outlet portion 414 is closed off so as to allow a forced convective cooling of the decay heat of aerosols captured in MFF 430. FCVS 410 is arranged and configured such that when a flow of gas through outlet portion 14 is closed off convective and radiant heat transfer removes decay heat of aerosols captured in MFFs 430. Convective heat transfer occurs via the outer surface of MFFs 430 and air passing upward via the chimney effect through housing 411 and radiant heat transfer occurs between MFFs 30 and housing 411. FCVS 410 addresses over-pressurization of containment 420 in the event of a severe accident by using one or more MFFs 430 and MS 432 in dual-inlet housing 411, which allows for two inlet paths during venting, but creates a natural convective heat transfer path when not venting to remove decay heat due to the chimney effect of the dual-inlets. Dual-inlet housing 411 creates a chimney effect with one inlet higher than the other, so that during non-venting periods, the containment atmosphere actually cools the decay heat via convective heat transfer with significant capability to handle large heat loads to address all types of Reactor designs. The convective design of FCVS 410 allows for passive decay heat removal, with no requirement to add water or chemicals. FCVS 410 may handle removal of the high decay heat load of multi-unit CANDUs and BWR and PWR Nuclear Power Plants. Dual inlet housing 411 can also be installed inside containment 420 allowing for a non-pressure vessel enclosure which keeps the entire radioactivity inside containment 420 and eliminates any need for any external building. In an alternative embodiment, with two containment penetrations at different elevations and utilizing a pressure vessel design, FCVS 410 can also be installed exterior to containment 420. In preferred embodiments, commercially available cartridge MFFs, for example SINTERFLO sintered stainless steel filter cartridges from Porvair Filtration Group, and commercially available MS media are used in FCVS 410 to allow for a lower cost for the filtering hardware. The convective decay heat removal capability of FCVS 410 allows the ability to keep temperature below the auto-ignite temperature of hydrogen and also below the melting point of hydroscopic aerosols by designing the chimney effect within the temperature restrictions. Since the heat transfer is convective, then aerosol fouling related to radiant heat transfer emissivity that limits the effectiveness of other dry FCVS technologies is not an issue for FCVS 410. FIGS. 7a and 7b show an example of a nuclear reactor containment 500 to illustrate the placements of FCVSs 210 and 310. In these embodiments, nuclear reactor containment 500 includes an innerspace 502 that is sealed off from inside 504 of containment 500 and outside 506 of containment 500. In the embodiment shown in FIG. 7a, innerspace 502 includes four FCVSs 210. Inlets 508 of FCVSs 210 are connected to inside 504 of containment and outlets 510 of FCVSs 210 are connected to outside 506 of containment. In the embodiment shown in FIG. 7b, one FCVS 310 is shown in innerspace 502, with an inlet 512 connected to inside 504 of containment and an outlet 514 of FCVS 310 connected to outside 506 of containment 500. In the preceding specification, the invention has been described with reference to specific exemplary embodiments and examples thereof. It will, however, be evident that various modifications and changes may be made thereto without departing from the broader spirit and scope of invention as set forth in the claims that follow. The specification and drawings are accordingly to be regarded in an illustrative manner rather than a restrictive sense.
	claims	1. A thermoelectric generator with on-demand activation for use on a space vehicle, comprising:a fuel sample;a neutron source having electrical leads and constructed and arranged to emit neutrons into the fuel sample to initiate radioactive decay reactions in the fuel sample in response to the neutron source receiving an activation input at the electrical leads; anda thermoelectric converter coupled to the fuel sample to convert thermal energy from the radioactive decay reactions to electrical energy,the thermoelectric generator thus constructed and arranged to generate power for the space vehicle on demand in response to the neutron source receiving the activation input,wherein the fuel sample includes stable Bi209, and wherein the radioactive decay reactions include (i) a radioactive decay of Bi210 to Po210 (ii) a radioactive decay of Po210 to stable Pb206, andwherein the fuel sample further includes a catalyst to amplify neutron generation initiated by the neutron source. 2. A thermoelectric generator as in claim 1, wherein the catalyst includes beryllium formed in a thin film coating over the Bi209. 3. A thermoelectric generator as in claim 1, further comprising control circuitry coupled to the neutron source to provide the activation input to the neutron source and thereby to initiate the radioactive decay reactions in the fuel sample and conversion of thermal energy into electrical energy on demand. 4. A thermoelectric generator as in claim 3, further comprising a communication receiver coupled to the control circuitry to receive a remotely generated activation signal while the thermoelectric generator is deployed in outer space, wherein the control circuitry is further constructed and arranged to provide the activation input to the neutron source in response to the communication receiver receiving the remotely generated activation signal. 5. A thermoelectric generator as in claim 3, wherein the neutron source is constructed and arranged to emit neutrons into the fuel sample at a level that varies in relation to a pulsewidth of the activation input, and wherein the control circuitry is further constructed and arranged to output the activation input with different pulsewidths to initiate different levels of radioactive decay reactions in the fuel sample and thereby to cause the thermoelectric generator to generate different amounts of electrical energy. 6. A thermoelectric generator as in claim 5, further comprising a meter coupled to the thermoelectric converter to measure an electrical output level of the thermoelectric converter, wherein the meter is further coupled to the control circuitry to provide feedback to the control circuitry that varies in relation to the electrical output level of the thermoelectric converter, wherein the control circuitry is further constructed and arranged to detect, based on the feedback, when the electrical output level from the thermoelectric converter drops below a predetermined level and then to again provide the activation input again to the neutron source to reactivate the fuel sample to increase the electrical output level. 7. A thermoelectric generator as in claim 3, wherein the neutron source and the fuel sample are moveable relative to each other within the thermoelectric generator to expose different portions of the fuel sample to neutron emission, wherein the control circuitry is further constructed and arranged to provide the activation input to the neutron source multiple times to activate the different portions of the fuel sample in sequence. 8. A thermoelectric generator as in claim 3, further comprising a set of additional neutron sources disposed in relation to the fuel sample to expose different portions of the fuel sample to neutron emission, wherein each of the set of additional neutron sources is coupled to the control circuitry to receive a respective activation input from the control circuitry. 9. A thermoelectric generator as in claim 8, wherein the control circuitry is further constructed and arranged to apply activation inputs to the neutron source and the set of additional neutron sources in a timing sequence to expose the different portions of the fuel sample to neutron emission at different times, such that, as radioactive decay reactions in one portion of the fuel sample diminish over time, radioactive decay reactions in another portion of the fuel sample are increased to extend a service life of the fuel sample. 10. A thermoelectric generator as in claim 3, further comprising a set of additional fuel samples of a same initial composition as the fuel sample and a set of additional neutron sources each disposed in relation to a respective additional fuel sample to expose the set of additional fuel samples to neutron emission, wherein each of the set of additional neutron sources is coupled to the control circuitry to receive a respective activation input from the control circuitry to initiate radioactive decay reactions in the respective fuel sample and conversion of thermal energy into electrical energy on demand. 11. A thermoelectric generator as in claim 10, wherein the control circuitry is further constructed and arranged to apply activation inputs to the neutron sources in a timing sequence to expose the different fuel samples to neutron emission at different times, such that, as radioactive decay reactions in one fuel sample diminish over time, radioactive decay reactions in another fuel sample are increased to extend a service life of the thermoelectric generator. 12. A thermoelectric generator with on-demand activation for use in a space vehicle, comprising:multiple fuel samples each including Bi209;multiple neutron sources, each neutron source having electrical leads and disposed in relation to one of the fuel samples to emit neutrons into the respective fuel sample to initiate radioactive decay reactions in the fuel sample in response to the neutron source receiving an activation input at the electrical leads;multiple thermoelectric converters, each thermoelectric converter coupled to a respective one of the fuel samples to convert thermal energy from the radioactive decay reactions in the fuel sample to electrical energy; andcontrol circuitry coupled to each of the neutron sources to provide the respective activation input to each of the neutron sources, wherein the control circuitry is constructed and arranged to apply activation inputs to the neutron sources in a timing sequence to expose the respective fuel samples to neutron emission at different times, such that, as radioactive decay reactions in one fuel sample diminish over time, radioactive decay reactions in another fuel sample are increased to extend a service life of the thermoelectric generator,the thermoelectric generators thus constructed and arranged to generate power for the space vehicle on demand in response a respective neutron source receiving an activation input,wherein the radioactive decay reactions include (i) a radioactive decay of Bi210 to Po210 and (ii) a radioactive decay of Po210 to stable Pb206, andwherein the fuel sample further includes a catalyst to amplify neutron generation initiated by the neutron source. 13. A thermoelectric generator as in claim 1, wherein the fuel sample, the neutron source, and the thermoelectric converter are embodied together in a thermoelectric generator assembly, wherein the neutron source is disposed along a central axis of the thermoelectric generator assembly, and wherein the thermoelectric converter includes multiple thermoelectric converter elements disposed concentrically around the neutron source. 14. A thermoelectric generator as in claim 13, wherein the fuel sample is disposed concentrically around the neutron source between the neutron source and the thermoelectric converter elements. 15. A thermoelectric generator as in claim 14, wherein the thermoelectric generator assembly further includes an insulating layer disposed concentrically around the neutron source between the neutron source and the fuel sample.
043495052	abstract	A neutral beamline generator with energy recovery of the full-energy ion ponent of the beam based on magnetic blocking of electrons is provided. Ions from a positive ion source are accelerated to the desired beam energy from a slightly positive potential level with respect to ground through a neutralizer cell by means of a negative acceleration voltage. The unneutralized full-energy ion component of the beam exiting the neutralizer are retarded and slightly deflected and the electrons in the neutralizer are blocked by a magnetic field generated transverse to the beamline. An electron collector in the form of a coaxial cylinder surrounding and protruding axial a few centimeters beyond the neutralizer exit terminates the electrons which exit the neutralizer in an E x B drift to the collector when the collector is biased a few hundred volts positive with respect to the neutralizer voltage. The neutralizer is operated at the negative acceleration voltage, and the deflected full energy ions are decelerated and the charge collected at ground potential thereby expending none of their energy received from the acceleration power supply.
	abstract	A loading machine for simultaneously transferring fuel elements between a reactor core and a storage rack in a nuclear power station has a mast divided into individual mast parts. Each mast part has its own gripping and guiding devices for gripping and holding the individual fuel elements. The mast is provided on a trolley. At least one of the mast parts is movable horizontally on the trolley and is rotatable about its longitudinal axis. The individual fuel elements can be lifted out of the reactor core simultaneously, their position relative to each other can be changed, and the fuel elements can be set down on a storage rack or a workplace. A method for transferring fuel elements in a nuclear power station is also provided.
043107656	summary	BACKGROUND OF THE INVENTION This invention relates to neutron sources and more particularly to accelerator-type neutron tube sources having improved ionization sections. Accelerator-type neutron tube sources are employed in many applications. A well known application is in the radioactivity logging of wells penetrating subterranean formations. For example, in the art of neutron-neutron well logging a source of primary neutrons is employed to irradiate subterranean formations of interest. The resulting secondary radiation is measured by one or more detectors spaced axially from the source within the borehole. Such secondary irradiation may take the form of thermal neutrons, epithermal neutrons, or thermal neutron capture gamma rays. A logging tool of this type employed for porosity measurements is disclosed in U.S. Pat. No. 4,005,290 to Allen wherein the logging tool includes a neutron source and epithermal and thermal neutron detectors. In procedures such as porosity logging, the neutron source is a continuous source usually of a chemical type. Other well known radioactive well logging techniques involve the use of pulsed neutron sources. For example, in the art of radioactive assay well logging an assay tool is lowered into the well to the level of a formation to be assayed. The assay operation is then carried out by cyclically operating a neutron source in the tool in order to irradiate the formation under investigation with repetitive bursts of fast neutrons. In one assay procedure, disclosed in U.S. Pat. No. 3,686,503 to Givens et al, delayed fission neutrons emitted by uranium within the formation may be detected by a neutron detector. Another procedure, disclosed in U.S. Pat. No. 4,180,730 to Givens et al, involves detection of prompt fission neutrons emitted from uranium in the formation. Pulsed neutron logging techniques may also be employed in procedures in which radioactive decay rates are determined. Thus, the formation under investigation is irradiated with a burst of fast neutrons and the resulting neutron population is detected during successive or overlapping time windows. For example, U.S. Pat. No. 3,800,150 to Givens discloses a pulsed neutron logging technique in which epithermal neutron decay or thermal neutron decay is measured by employing time windows for detection which partially overlap one another. Neutron sources such as may be employed in radioactive logging procedures as described above may take the form of accelerator-type neutron tubes comprising a target section, a replenisher section, and an ionization section located between the target and the replenisher section. The replenisher section provides a source of accelerator gas to the ionization section where it is ionized and then accelerated to impact the target. The target is formulated of material which responds to the bombarding ions to produce neutrons. In a number of well known accelerator-type tube sources, heavy isotopes of hydrogen are employed as the accelerator gas and in the target. For example, the accelerator gas may take the form of deuterium or mixtures of deuterium and tritium and the target may include tritium molecules, deuterium molecules or mixtures of deuterium and tritium molecules. The so-called deuterium-tritium nuclear reaction is one commonly employed in an accelerator-type neutron tube to produce neutrons. In the replenisher section a filament or reservoir usually made of zirconium or titanium is electrically heated (under controlled conditions) to release deuterium gas previously adsorbed in the filament or reservoir. Zirconium and titanium have the property of adsorbing copious quantities of different gases such as hydrogen, deuterium, tritium, and other gases. These materials have the further property of releasing the hydrogen isotope gases under a controlled release condition when heated to about 300.degree. C. and at the same time retaining other gases that may have been adsorbed. The deuterium molecules are ionized in the ionizing section by the application of a positive voltage to an anode in the ionizing section. The deuterium ions are then accelerated by a large negative voltage, e.g. -100 KV, and impact the tritium target to produce a supply of neutrons. While various techniques may be employed in ionizing the accelerator gas, one ionization technique which is suitable particularly where the neutron source is operated at a low accelerator gas pressure and in a pulsed mode is the so-called Penning method. A Penning ion source comprises spaced cathodes and an anode located intermediate the cathodes. In a cold-cathode type Penning ion source, electrons are emitted from a cathode surface by field emission when a positive voltage pulse is applied to the anode. A magnet associated with the source functions to spiral the electrons thus increasing their flight path and increasing the statistical probability that they will collide with molecules of accelerator gas supplied to the ionization chamber. In a well designed Penning ion source, some of the electrons originating at one cathode surface will impact the other cathode surface and secondary electrons are emitted which also function to increase the ionization reaction. Such ion sources are well known to those skilled in the art and are described in Flinta, J. "Pulsed High-Intensity Ion Source", Part I; Pauli, R. and Flinta, J. "Pulsed High-Intensity Ion Source", Part II, Nuclear Instruments 2, pp 219-236 (1958). In a hot-cathode type Penning ion source, one cathode is a heated filament and initial electrons are supplied by thermionic emission from the filament. In all other respects, cold-cathode and hot-cathode Penning ion sources are essentially the same. Hot-cathode ion sources are also well known to those skilled in the art and one such source is described in Wood, J. and Crocker, A. "An Electrostatically Focused Ion Source And Its Use In A Sealed-Off D.C. Neutron Source", Nuclear Instruments And Methods 21, pp 47-48 ( 1963). SUMMARY OF THE INVENTION In accordance with the present invention, there is provided an accelerator-type neutron tube having a new and improved ionization section for ionizing the accelerator gas. The ionization section is located between the target section and replenisher section of the neutron tube and comprises an ionization chamber adapted to receive accelerator gas from the replenisher section. First and second cathodes are spaced from one another and have opposed active surfaces exposed to the interior of the chamber. Anode means are located at a position intermediate of the cathodes whereby in response to an applied positive voltage electrons 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 2 or more. In a further embodiment of the invention, the active surface of a first cathode member located adjacent to the replenisher section of the tube has a protuberant member extending axially into the chamber. A second cathode member spaced from the first cathode member in the direction of the target has an aperture therein along the axis of the protuberant portion. The anode member extends peripherally around the interior of the ionization chamber at a position intermediate the first and second cathode members. In yet a further embodiment of the invention, the ionization section includes an annular magnet extending around the exterior of the ionization chamber and enveloping the anode member. Means are provided which establish a high permeability magnetic flux path extending outwardly from the opposed poles of the magnet to the active surfaces of the cathode members.
	description	FIG. 1 illustrates a measurement path in an X-ray (testing) machine, not shown in detail. In a known manner, an X-ray source 1 generates X-ray radiation FX and radiates it onto an object 2 to be X-rayed, the object 2 being located on a transport device 3. A collimator or aperture arrangement 4a, 4b generates a primary beam FX1 preferably as a pencil beam. As dictated by the crystal-lattice structure of the material of the object 2 to be X-rayed, the primary beam FX1 is diffracted in a known way at a plurality of lattice points G (only one is shown here). As a result of the diffraction at the lattice points, at least part of the primary beam FX1 is deflected as the radiation FX1xe2x80x2 with a specific beam energy at an angle "THgr"M that is a function of the material. In a known manner, this condition is utilized to determine the material in the beam on the basis of the physical effect of X-ray diffraction (Bragg""s interference pattern). Through the predetermined of a specific angle "THgr"M, different energies are measured (in accordance with Bragg) and compared to known values. FIG. 2 illustrates a collimator 6, which can be used to determine the material or type of material of the object X-rayed with the primary beam FX1 (FIG. 1). The collimator 6 is a round-slot collimator with a crystal detector 11 disposed behind it. Both the collimator 6 and the detector 11 are used for a measurement employing X-ray diffraction. A blind-bore-like opening 7 that is integrated into the center of the collimator 6 acts as a central collimator. At a radial distance from the central opening 7, the collimator 6 has a conically-expanding round slot 10, which determines an angular path of the predetermined angle "THgr"M. Disposed inside the opening 7 are a first detection device 8 and, behind it at a defined distance, a second detection device 9. The surface of the crystal detector 11 is large enough to permit the scattering-cone beams (radiation FX1xe2x80x2) exiting through round slot 10 to be recorded. The crystal 11 has a preferably circular, X-ray-sensitive surface 12 that faces the collimator 6. The X-ray source is indicated by 1 in FIGS. 1-3. FIG. 3 depicts a further collimator 26, which has a first detection device 28 in a central, blind-bore-like opening 27, and a second detection device 29 at a defined distance from the first detection device, preferably at the rear end of the passage 27. The first detection device 28 is embodied as a detector for relatively lower X-ray energies, while the second detection device 29 is embodied as a detector for relatively higher X-ray energies. This collimator/detection device is used, for example, for conventional material detection. FIG. 4 schematically shows the inside structure of the collimator 6, 26 with the electrical connections for signal evaluation, which are necessary for adjustment. Apertured-diaphragm arrangements 13a, 13b and 14a, 14b, are disposed, respectively, in front of the detection device 8, 28 or the detection device 9, 29. These arrangements adapt the detection surface of the detection device 8, 28 or the detection device 9, 29 to the diameter of the primary beam FX1. Elements 15 and 17 represent signal-amplifier stages, which are necessary for amplifying the signals picked up at the respective detection devices 8, 28 and 9, 29. These amplifier stages 15 and 17 are connected to display units 16 and 18, respectively, and are also connected to a microprocessor (computer) 23. The additionally-illustrated crystal detector 11 is omitted if a simple collimator 26 is to be adjusted instead of a ring-slot collimator 6. A further embodiment according to FIG. 5 employs detection devices 8, 28, 9, 29, which permit the determination of the center of gravity of the intensity distribution of the X-ray or primary beam FX1 impacting them. The detectors can be numerous individual detectors, detector arrays, four-quadrant detectors or position sensitive detectors having multiple-segmented diodes. Amplifier stages that operate in parallel are disposed downstream of these detection devices 8, 28, 9, 29; for the sake of a clear overview, the stages are only indicated by a respective reference numeral 19 or 21. An amplifier of the respective amplifier stage 19, 21 is associated with each individual detector, each detector in the array and each diode in the detection devices 8, 28, 9, 29. A display unit 20 or 22 is disposed downstream of these amplifier stages 19, 21. The amplifier stages 19, 21 are also electrically connected to the microprocessor 23. The first collimator arrangement 13a, 13b and the second collimator arrangement 14a, 14b, or only the second collimator arrangement 14a, 14b from FIG. 4, can be omitted if the orientation is effected in a pencil beam (point beam) FX1. Regardless of the embodiment, the adjustment process is performed as follows: As shown in FIG. 6, the X-ray radiation emitted as a primary beam FX1 and observed inside the passage 7 prior to the adjustment may be located outside of the center of the first and second detection devices 8, 28 and 9, 29. A first point P1 that is predetermined for the adjustment and a second point P2 on the primary beam do not coincide with the respective center point (center) of the detection device 8, 28 or 9, 29. Consequently, the collimator 6, 26 must be oriented spatially, that is, in three planes, to reach its optimum spatial position without having any tilt relative to the primary beam FX1. In a first step of the adjustment, the collimator 6, 26 is moved, for example in a plane perpendicular or approximately perpendicular to the propagation direction of primary beam, until the generated signal is maximal in the first detection device 8, 28. The primary beam FX1 lies in the point P2, further outside of the center of the second detection device 9, 29. The signals generated in the detection devices 8, 28 and 9, 29 are displayed on the display unit 16 and 20, respectively. For the optimum orientation of the collimator 6, 26, it is necessary to orient the collimator toward the second point P2 in a second step. For the orientation toward this point, the collimator 6, 26 is rotated or adjusted in two independent planes about an imaginary point P3, which is preferably located near the center of the first detection device 8, 28, until the signal from the second detection device 9, 29 is also at its maximum. This rotational plane encompasses a pointed conical region (see arrows) originating from the point P3. After this second step, the maximum of the first rotational plane is established, effecting a first local pre-orientation of the collimator 6, 26. After the intensity maximum in the first plane has been determined, the rotation in the second rotational plane is initiated for the further spatial orientation. A point near the center of the first detection device 8, 28 is also selected for this orientation; the point P3 can serve as a reference. Also in this case, the collimator 6, 26 is rotated inside the pointed conical region such that the intensity maximum is established at the respective detection devices 8, 28 or 9, 29. This procedure is also to be performed in the third rotational plane, so the primary beam FX1 then impacts the centers of the first and second detection devices 8, 28, 9, 29 at a right angle. The order of the collimator moving and rotating steps may be varied depending, for example, on the initial alignment condition and alignment reproducibility. In an embodiment comprising a plurality of position-sensitive detectors, the amplitudes of the individual signals of the detection devices 8, 28 and 9, 29 are first analyzed step-wise in the microprocessor (computer) 23 and further processed. The current position of the beam center of the primary beam FX1 is determined from the individual signals and compared to the center of the respective detection device 8, 28 or 9, 29. The detection centers are readjusted to the beam center of the primary beam FX1 until the generated signal values are at their maximum. In this way, an offset between the respective detection center and the center of the primary beam can be more clearly identified and corrected in one step with the aid of a one-time calibration. In an embodiment comprising a four-quadrant detector in the detection device 8, 28 or 9, 29, the signals generated quadrant-wise are evaluated, with the detection center being established when the primary beam FX1 impacts the common intersection of the quadrants and generates signals of identical magnitude in each quadrant. Thus, simplified means are used to effect an optimum orientation of the collimator 6, 26 along a primary beam FX1. The collimator position in an X-ray testing machine (not shown in detail here) is sequentially varied with the aid of a regulating device (not shown in detail here) until both of the detection devices 8, 28 and 9, 29 are impacted maximally by the primary beam FX1. The collimator 6, 26 is rotated stepwise in the individual planes by mechanical elements installed in the X-ray machine, which are automatically controlled and regulated with a program controlled by the microprocessor (computer) 23. The intensity maxima determined in the first adjustment can be stored for future use. In the use of a calibrated collimator 6, 26 with a plurality of location-sensitive detectors, a one-time measurement can immediately reveal the offset between the primary-beam maximum and the detection center, and the method can correct the offset. It is further possible to use a plurality of detection devices that are disposed one behind the other inside the passage 7, but it is not necessary for the adjustment process itself, because this feature renders the method more costly. The invention now being fully described, it will be apparent to one of ordinary skill in the art that many changes and modifications can be made thereto without departing from the spirit or scope of the invention as set forth herein.
	summary
	summary
	summary
059404620	abstract	A method for packaging, compacting or storing spent control and absorber elements of light-water reactors for waste disposal, a device for preparing an elongated component of a spent control element, a light-water-cooled reactor and a coil of an originally elongated component are provided for storing absorber fingers or absorber sheets of spent control elements in storage containers in pressurized-water reactors or boiling-water reactors. The absorber fingers or absorber sheets are separated from the control elements. They are not cut into pieces (in which process large amounts of radioactive substances would be released), but instead they are only deformed, namely wound into coils which can be stacked in a space-saving manner in the storage containers.
	summary
	claims	1. A method for producing radioactive iodine comprising:providing a volume of non-enriched uranium (“NEU”) material in an irradiation chamber, the irradiation chamber having a gas inlet port and a gas outlet port;providing one or more non-moderating neutron-reflecting regions disposed in the irradiation chamber;irradiating the uranium material with neutrons provided by a compact, stand-alone neutron generator to cause fission reactions to occur in the uranium material, wherein:the neutrons are provided at an energy above a fast fission threshold for U-238,the one or more non-moderating neutron-reflecting regions are configured to increase the path length traveled by at least some of the neutrons before those neutrons leave the irradiation chamber,at least some of the fission reactions lead to the production of radioactive iodine, andat least some of the radioactive iodine sublimates in the irradiation chamber;during irradiation, introducing a carrier gas into the gas inlet port to mix with the sublimated radioactive iodine and form a gas mixture;withdrawing the gas mixture from the gas outlet port; andseparating at least some of the radioactive iodine from other components of the gas mixture. 2. The method of claim 1 wherein at least a portion of the gas mixture is returned to the gas inlet port after separation of the radioactive iodine. 3. The method of claim 1 wherein separating at least some of the radioactive iodine includes using a silver zeolite trap. 4. The method of claim 1, and further comprising, after an irradiation period:removing the NEU material from the irradiation chamber;dissolving the NEU material in a solvent, thereby releasing additional sublimated iodine as part of an additional gas mixture; andseparating at least some of the additional radioactive iodine from other components of the additional gas mixture. 5. The method of claim 1 wherein the NEU material consists of one of the following forms: solid material, crushed solid material, metallic shavings, metallic filings, sintered pellets, liquid solutions, molten salts, molten alloys, and slurries. 6. The method of claim 1 wherein the NEU material comprises material in granular form. 7. The method of claim 1 wherein the NEU material comprises material in molten form. 8. The method of claim 1 wherein the NEU material comprises material in solid form. 9. The method of claim 1 wherein the NEU material comprises material formed into multiple elongate rods. 10. The method of claim 1 wherein the compact, stand-alone neutron generator generates neutrons above an energy of 0.8 MeV. 11. The method of claim 1 wherein the one or more non-moderating neutron-reflecting regions disposed in the irradiation chamber comprise multiple elongated parallel non-moderating neutron-reflecting regions disposed within the irradiation chamber, wherein each of the multiple elongated parallel non-moderating neutron-reflecting regions has an interior bore for receiving the NEU material. 12. The method of claim 1 wherein the at least one non-moderating neutron-reflecting region disposed within the irradiation chamber comprises multiple non-moderating neutron-reflecting regions. 13. The method of claim 1 wherein the NEU material comprises depleted uranium. 14. The method of claim 1 wherein the compact, stand-alone neutron generator comprises a linear accelerator-type neutron generator. 15. The method of claim 12 wherein the at least one non-moderating neutron-reflecting region disposed within the irradiation chamber further comprises a non-moderating neutron-reflecting wall interior to the irradiation chamber that is exterior to the multiple non-moderating neutron-reflecting regions.
	abstract	A computer-based method for load testing a web application. The computer-based method includes selecting one or more uniform resource locator (URL) parameters, identifying the selected parameters by parameter type, and loading the web application with URLs created by randomly generating the selected URL parameters. An electronic system adapted to load test a web application utilizing random parameter generation.
	summary
051587408	description	DETAILED DESCRIPTION OF THE INVENTION In the following description, like references characters designate like or corresponding parts throughout the several views. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a conventional nuclear reactor fuel assembly, represented in vertically foreshortened form and generally designated by the numeral 10. Being the type use in a pressurized water nuclear reactor (PWR), the fuel assembly 10 basically includes a lower end structure or bottom nozzle 12 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 14 which project upwardly from the bottom nozzle 12. The assembly 10 further includes a plurality of transverse grids 16 axially spaced along the guide thimbles 14 and an organized array of elongated fuel rods 18 transversely spaced and supported by the grids 16. Also, the assembly 10 has an instrumentation tube 20 located in the center thereof and an upper end structure or top nozzle 22 removably attached to the upper ends of the guide thimbles 14. With such an arrangement of parts, the fuel assembly 10 forms an integral unit capable of being conventionally handled without damaging the assembly parts. As mentioned above, the fuel rods 18 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 16 spaced along the fuel assembly length. As seen in FIG. 1 and in greater detail in FIG. 2, each fuel rod 18 includes a plurality of nuclear fuel pellets 24 disposed in a stack in an elongated hollow cladding tube 26 having its opposite ends closed by top and bottom end plugs 28, 30 so as to hermetically seal the tube 26. Commonly, a plenum spring 32 is disposed within the cladding tube 26 between the top end plug 28 and the pellets 24 to maintain the pellets in a tight, stacked relationship within the rod 18. The fuel pellets 24 composed of fissile material are responsible for creating the reactive power of the nuclear reactor. A liquid moderator/coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract heat generated therein for the production of useful work. To control the fission process, a number of control rods 34 are reciprocally movable in the guide thimbles 14 located at predetermined positions in the fuel assembly 10. Fuel Rod End Plug Welding Method The top and bottom end plugs 28, 30 are typically girth welded to the opposite ends of the cladding tube 26. Also, the top end plug 28 typically contains an axial gas pressurization passage or bore 36 which is used after completion of the girth welding to evacuate gases from the interior of the tube 26 and then to fill the tube with a suitable inert gas before the bore 36 is sealed. Suitable welding apparatus for carrying out both girth and end seal welding is illustrated and described in U.S. Pat. No. 4,837,419 to Anthony E. Boatwright et al, assigned to the assignee of the present invention, the disclosure of which is incorporated herein by reference. Referring to FIGS. 3 and 4, the top end plug 28 includes a generally cylindrical body 38 having an exterior flat end face 40 with the axial bore 36 extending through the plug body 38 and terminating in an external opening 36A defined on the flat end face 40. The flat end face 40 is bounded by an exterior conical peripheral face 42. In accordance with the present invention, in order to facilitate the fabrication of a quality seal weld 44 at the outer portion of the bore 36 and its opening 36A, a trepan 46 is defined in the flat end face 40 of the end plug 28 about the external opening 36A of the axial bore 36. The trepan 46 is composed by an annular groove 48 encircling and spaced radially outward from the external opening 36A and an annular end face portion 50 extending between the annular groove 48 and the external opening 36A. The depth of the groove 48 and the diameter of the end face portion 50 encircled by the groove 48, that is, the components of the trepan 46, must bear a ratio to one another within a particular range to produce the end seal weld 44 of optimum quality. The ratio of the depth of the groove 48 to the diameter of the annular end face portion 50 of the trepan 46 is within a range of from about 1:3 to 1:6, and is preferably about 1:5. The ratio of the depth of the groove 48 to its width is within a range of from about 1:1 to 1:2. The range of ratios between the depth and diameter were developed by trial and error tests to determine which values achieve the best centering of the welding arc and most satisfactory melting of the portion 50. When the stated proportions are produced on an end plug, the geometry of the plug causes a centering of the weld puddle, and thereby uniform depth of penetration of the weld, resulting in a higher quality weld to close the seal hole and inner fuel rod integrity. Briefly, the theory of operation of the trepan 46 in producing an optimum quality seal weld 44 is as follows. The material of the annular end face portion 50 receives the welding arc and melts and flows outwardly. However, the melt flow collides with the outer wall 48A of the groove 48 which acts as a cold obstruction, compared to the melt temperature. The effect of the cold obstruction is to break up the surface tension of the melt and better centralize the weld 44. Referring to FIGS. 3-5, there is illustrated a portion of the arrangement of the cited U.S. patent for carrying out the end seal welding and thereby sealing of the axial bore 36 of the top end plug 28. The illustrated arrangement includes an axial welding electrode 52 and an end plug stop 54. The axial welding electrode 52 used to form the end seal weld 44 is positioned in axially spaced relation from the trepan 46 on the end face 40 of the end plug 28. The end plug stop 54 is positioned against the conical peripheral face 42 of the end plug 28 in outward radially spaced relation from the trepan 46. Therefore, the stop 54 does not engage in the groove 48 of the trepan 46. When an axially-directed welding arc is applied by the axial electrode 52 to the trepan 46 in the flat end face 40 of the end plug 28, centering of the arc on the trepan 46 results and a melting of the material of the trepan into the end face 40 is produced such that the trepan 46 of FIGS. 3 and 4 is replaced by a shallow concavity 56 extending across the region substantially encircled originally by the annular groove 48 of the trepan 46. The concavity 56 lies over the exterior of the resultant weld 44. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely an exemplary embodiment thereof.
043137911	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 illustrates an ultrasonic search unit 20. The search unit 20 includes an ultrasonic transducer element 21 and a strip carrier 22 that, as is best shown in FIG. 2, has mutually opposing faces 23, 24, and an aperture in which the transducer element 21 is suitably mounted. The transducer element 21, which is a polarized ferroelectric ceramic having an electrode deposited or fired on two of its surfaces, is aligned within the aperture so that one surface 25 is flush with face 23 of the strip carrier 22. The opposing surface of the transducer element is recessed within the aperture and faced by a sonic damping material 26. A decoupling isolating material 30 is placed between the perimeter of the aperture and the respective opposing surfaces of the element 21. The element 21 is secured within the aperture by an electrically non-conducting cement 31. The surface 25 of the transducer element 21 which is flush with face 23 of the strip carrier is grounded to the carrier. Grounding is accomplished by spot welding several conductors 32, or by other suitable means. The decoupling material 30 is disposed between the transducer element and carrier in order to minimize ultrasonic coupling therebetween. A coaxial cable 33, having an inner conductor 34 and an outer conductor 35, is attached to an edge 36 of the strip carrier. The inner conductor 34 is attached to the transducer element 21. The outer conductor 35 is attached to the strip carrier 22. The search unit 20 must be capable of freely traversing the limited clearances between the fuel elements or between a fuel element and a control element guide tube of a fuel assembly which may be spaced to within two millimeters of each other. Hence, the search unit 20, as well as its individual components, must be selected to satisfy specific dimensional requirements without compromising the ultrasonic characteristics needed to apply the principles of the detection technique. A specific example of a search unit constructed in accordance with the principles of the invention includes a transducer element fabricated from lead zirconate titanate, measuring approximately 2.5 millimeters wide, 12.5 millimeters long and 0.3 millimeters thick, mounted in an aluminum carrier. The transducer element is isolated from the perimeter of the aperture by a layer of cork. The front and back surfaces of the transducer element are coated with fired silver electrodes, and the surface that is flush with one face of the carrier is grounded to the adjacent aluminum at several points through small copper wires tack welded to both the aluminum and the silver electrode. A layer of the conducting epoxy resin may be spread over the copper wires and face of the transducer element at face of the carrier in order to present a smooth surface for insertion into the fuel assembly. The damping material 26 is composed of two grades of tungsten powder mixed in a low molecular weight polysulfide polymer. A specific damping material includes a mixture of a tungsten powder of an average particle size of 4.5 microns with a tungsten powder of an average particle size of 1.33 microns mixed with a low molecular weight polysulfide polymer called Thiokol LP-3, manufactured by the Thiokol Chemical Corporation, Trenton, New Jersey. A non-conducting epoxy resin is used to secure the transducer element within the aperture. The recessed surface of the ceramic is connected to the inner conductor of a coaxial cable which is disposed along the edge of the carrier. Other arrangements, shapes and materials can be used for the transducer element as long as the search unit is insertable between the components of the fuel assembly. In an alternate embodiment, for example, a hollow tubular carrier within which the coaxial cable is contained might be used. FIG. 3 shows, as a section of a fuel assembly, a schematic planar representation of a search probe 20 transversely aligned with the lower plenum of a fuel element 40. The transducer element, coupled to the fuel element 40 for transmitting ultrasonic energy into the fuel element, is energized by a pulser (not shown) to emit pulses at a predetermined rate and frequency. The sweep of an oscilloscope is synchronized to display the transmitted and reflected pulses. The reflected waves are received by the scope via the transducer. If the fuel element has not failed, then gas will be the only fluid present in the lower plenum. A high reflection coefficient at the metal-gas interface will prevent significant propagation of the ultrasound past the inner surface of the cladding. The response displayed on a conventional pulse echo instrument for a gas filled fuel element is shown as an oscillogram in FIG. 4 with time (t) plotted as the abscissa. The oscillogram of FIG. 4, and also FIGS. 6 and 7, is representative of the resulting display generated at a frequency of approximately seven megahertz wherein each division of the time scale is approximately three microseconds and the fuel element outside diameter is slightly below 0.5 inches. In FIG. 4, the transmitted signal is substantially mixed with the received signal reflected from the first or front gas-metal interface due to the low coefficient of transmission of the gas. If, in contrast, the fuel element has failed so that the lower plenum contains water, the reflection coefficient at the front interface will be significantly diminished. Thus, as schematically shown in FIG. 5, significant portions of the ultrasonic pulse will propagate through the liquid and be reflected at the back liquid-metal interface within the fuel element 40. Hence, a reflected signal of a relatively pronounced magnitude separated from the transmitted signal on the time scale will be displayed. The response displayed on a conventional pulse echo instrument for a defective, water filled fuel element is shown as an oscillogram in FIG. 6. A significant response occurs at approximately fifteen microseconds on the abscissa--this represents the echo received from the back wall. The lower plenum of a fuel element generally contains a helical spring member which may restrict the free passage of the ultrasound. This does not, however, present an insurmountable difficulty. If the width of the piezoelectric element, measured along the longitudinal axis of the fuel element is greater than the pitch of the helical spring, then sound will propagate to the far wall and return. FIG. 7 shows the typical response of a water filled element containing a spring. Conventional ultrasound instruments contain gating circuits that allow the extraction of signals during a selected period of time relative to an initial pulse. In addition, circuitry can be provided to produce an alarm signal only when the ultrasonic signal amplitude in the gated period exceeds a preset threshold level. If the gate is set to pass signals between twelve and fifteen microseconds on the abscissa, and if the amplitude threshold is set at line 1 of the ordinate, then the presence of water is detectable in a fuel element with or without springs. In operation, the search unit is inserted into the spacing between adjacent components of the fuel assembly. Irradiated fuel assemblies are generally maintained under water, for cooling and shielding purposes, during removal from a reactor and initially are stored in a spent fuel pool. Hence, it will be understood that the inspection of the fuel elements is effected under water. The transducer element is transversely aligned with the longitudinal axis of the fuel element to be examined. A pulse is then emitted from the transducer into the fuel element. A fuel assembly can be tested by insertion of the search unit into the bundle of fuel elements without any component disassembly. Hence, the assembly need only be removed from the reactor for inspection purposes. The technique can be expanded to use multiplexed transducers to examine all the fuel elements of a fuel assembly automatically and rapidly.
045267456	description	DESCRIPTION OF PREFERRED EMBODIMENTS In FIGS. 1, 2, 3 and 4, the numeral 1 designates a fuel box which is attached to a sleeve-formed base 2. In use of the fuel assembly in a boiling water reactor, the exterior surfaces of the fuel box and the base face the space within the reactor but outside the fuel assembly. The fuel box 1, which is composed of four mutually equal sheet-metal elements 1', interconnected by means of four vertical strips 1", surrounds sixty-four fuel rods 3, twelve smaller water tubes 4 and one larger, central water tube 5. Each water tube forms a vertical channel extending along the fuel rods and conducting a water flow along but being separated from the fuel rods. The water tubes are mechanically connected to each other by means of a plurality of elongated linking members 6, which are attached with their ends to the strips 1". The fuel rods 3 rest with their lower ends on a bottom lattice device or element 7, which rests on two vertical supporting plates 8, which are welded to a hollow, cruciform water distributing member 9, which is provided with connection openings for the water tubes 4 and 5. The connection openings are constructed with annular supporting surfaces, against which the water tubes 4 and 5 rest at their lower end surfaces. The water distributing member 9 is provided with at least one (e.g. four) substantially radial tube 10, which opens out at the side surface of the fuel assembly and through which both the interior cavity of the member 9 and the water tubes 4 and 5 are hydraulically connected to space located radially outside the base 2. The base 2 of the fuel assembly has a circular, downwardly-facing inlet opening positioned below tubes 10, as illustrated. The inlet opening is surrounded by a substantially annular end surface 13. A guide member 14, which consists of a ring 14" and a plurality of rods 14' attached thereto, is arranged below the sleeve-formed base 2. The fuel assembly shown has a vertical center line and is intended to be supported, together with three similar fuel assemblies, by a common supporting plate intended for four fuel assemblies, said supporting plate being constructed with a conical supporting surface and a circular throttling opening for each fuel assembly. Below the radial tubes 10 and below the water distributing member 9, the wall of the base 2 is provided with at least one through-hole 12 which opens out at the side surface of the fuel assembly and opens inwardly directly into the interior of base 2, so that the inlet opening is hydraulically connected to the reactor space located radially outside the base. When the flow paths of the described fuel assembly are dimensioned in such a way that an optimum by-pass flow through the water tubes 4 and 5 is obtained at full reactor power, it cannot be avoided that a certain amount of boiling takes place in the water tubes 4 and 5 when the reactor power is reduced to a minimum by reducing the speed of the circulating pumps. To a certain extent such boiling gives rise to void formation at the upper part of the water tubes, which results in the pressure being considerably reduced in the water distributor 9, water thus flowing in through the channels 10 from the gaps located between the fuel assemblies. At a fully acceptable value of the void formation, the corresponding increase of the water flow through the radial channels 10 will be so great that a state of equilibrium will be created.
	abstract	A method for flattening a sample surface by irradiating the sample surface with a gas cluster ion beam, generates clusters of source gas in a cluster generating chamber, ionizes the generated clusters in an ionization chamber, accelerates the ionized cluster beam in an electric field of an accelerating electrode, selects a cluster size using a magnetic field of a sorting mechanism, and irradiates the surface of a sample. An irradiation angle between the sample surface and the gas cluster ion beam is less than 30° and an average cluster size of the gas cluster ion beam is 50 or above.
	summary
	summary
054250714	abstract	An end plug is inserted into the end of a nuclear fuel pin cladding tube after loading mixed oxide fuel pellets into the tube from a containment area. During the pellet loading operation, a disposable sleeve and an end plug carrier protect the external surfaces of the cladding tube and the end plug from contamination. The sleeve, mounted on the end of the cladding tube, is inserted into a seal arranged in a wall of the containment area. The end plug, located within a recess in the plug carrier, is then inserted into the end of the cladding tube by sliding the carrier through the sleeve. Upon removal of the cladding tube, the sleeve and the carrier are retained in the seal for ejection into the containment area during the next pellet loading operation.
	abstract	Embodiments of the invention provide systems and methods for achromatically bending beam of charged particles by about 90° during radiation treatment. A system may include first, second, third, and fourth bending magnets serially arranged along the particle beam path. The first and fourth bending magnets are configured to generate a positive field gradient that defocuses the particle beam in the bend plane. The second and third bending magnets are configured to generate a negative field gradient that focuses the particle beam in the bend plane. The first, second, third, and fourth bending magnets collectively bend the particle beam by about 90°, e.g., by about 22.5° each.
043200281	summary	BACKGROUND OF THE INVENTION The problem of storage of nuclear waste products from both military and civilian sources is presently becoming so acute that further progress, particularly in the field of development of nuclear energy, is threatened. The solutions proposed thus far generally have been based on irretrievable disposal of the wastes. The problems and uncertainties associated with such disposal are clearly summarized in Geological Survey Circular 779, entitled "Geologic Disposal of High-Level Radioactive Wastes . . . Earth-Science Perspectives," by J. B. Bredehoeft et al. Accordingly to these authors, it is estimated that 476,000 spent fuel assemblies will be on hand by the year 2000, occupying 3000 m.sup.3, if processed as high-level waste, and an order of magnitude more in intermediate-level waste. The toxicity of these wastes is illustrated by the fact that the quantity of water that would be required to dilute the wastes on hand by year 2000 to levels considered safe is double the quantity of fresh water in global storage. The urgency of the problem is illustrated by the discovery that increasing levels of artificial radionuclides in Lake Erie and Ontario have been traced to Cattaraugus Creek, which flows by an interim waste-storage facility at West Valley, N.Y. (Industrial Research/Development, July 1978, page 46). The size of the space necessary to accommodate such waste would be only about 10 km.sup.3, but the penalty for error in selecting the placement and planning of the facilities to provide it would be most severe. Critical features to be considered are small faults or fracture systems, which are extremely difficult to detect. Although techniques for non-destructive characterization of the site have recently been proposed, these systems require further refinement. In addition to the problems that arise from the nature of the site prior to storage of the waste products therein, the effect of the mechanical, chemical and thermal disturbance arising from the waste products themselves must be provided for. Canisters of high-level waste may produce 5 kW of heat ten years after reprocessing and it may take as much as one hundred years for the rate to decrease to 0.5 kW. No adequate model of the effects of the hundred-year thermal output exists. The form of the waste during storage is, of course, of crucial importance. At the present time it appears to be well accepted that high-level wastes from reprocessing may be cast in the form of glass billets, the glass having a very low leachability (Marsily et al, 1977, Energy Research and Development Administration 1976, Page 7). Nevertheless, some contaminants will be released and the nature of the glass itself may change as a result of the radiation so that the rate of leaching will undoubtedly change with time. Large gaps exist in our knowledge of transport systems, that is, with respect to the course by which released contaminants may reach the biosphere. The problems in this area include lack of knowledge of fractures, natural or manmade, and of measurements of the effect of fluid head and permeability of the rock or salt within which the billets are stored. Data are largely unavailable so that the complete description of groundwater flow is a problem still awaiting solution. The goal, of course, is to contain the waste and prevent it from reaching the biosphere until it is no longer hazardous. Estimates have been made as to the length of time for which the waste must be stored so that it presents no further damage. Strontium 90 and cesium 137 will constitute 99% of the projected curie accumulation by the year 2020, but these will be reduced to one millionth the initial radioactivity in 600 years. However, the toxicity from iodine 129 and radium and the actinide elements will remain well over ten million years. Ferruccio Gera, 1975, "Geochemical Behavior of Long-Lived Radioactive Wastes: U.S. Energy Research and Development Administration, Oak Ridge National Laboratory, report ORNL TM-4481," at page 14 considered that assuring containment for longer than five million years is "clearly impractical since totally reliable geologic predictions of the detail required over such long time frames are beyond present capability." The science of geologic prediction is limited by the assumption of constancy of rates of processes and incomplete data; the data in U.S. seismic records go back for only two hundred years. Accordingly, "validating a waste-management model for the time spans of concern will never be possible," according to this author. Even routine predictions of one-hundred-year processes have varied from good to poor. Predictive models, an essential step in selecting a site and managing the waste, have components that are inherently unpredictable at present. Accordingly, these models will not give a single answer to the fate of radioactive waste in geologic repositories but rather a spectrum of alternative outcomes based on uncertain assumptions about the future. As aforenoted, present thinking is based on the concept of incorporating the nuclear waste in glass and then irretrievably storing the glass, as billets, in a suitable storage space, a substantial thickness of earth providing the shielding. In this concept the sole defense against the access of water, which could solubilize the radioactive material, leading to its transport into the biosphere, is the choice of a stable site. However, as discoveries of the last few years have emphasized, the earth's crust is not stationary. Earthquakes have been noted throughout recorded history, and are much more frequent along certain well-known fault lines, but severe earthquakes in previously "immune" regions are not unknown. More important is the fact that the surface of the earth consists of so-called tectonic plates which move about on the surface of the melt below and clash with each other. The collision between the plates gives rise to mountain ranges and to other types of severe deformation of the crust. Such actions take place both slowly over geologic times and abruptly. Unfortunately, the period over which the nuclear waste must be safeguarded is comparable with the geologic periods over which even the slow but severe changes in the shape of the earth crust may take place. The time of ten million years has already been mentioned above. It has been pointed out that improved reprocessing could reduce the period of concern to about 1000 years; this would reduce but not eliminate the uncertainties. To project the safety of high-level radioactive wastes irretrievably stored over such time periods is therefore impossible. Nevertheless, proposals continue to be put forward that call for storage of such glass billets in hard rock caverns or in salt mines which obviously have been free of ground water for long periods of time. Such proposals originate not only in the United States but in other countries as well. For instance, Frank Feates and Norman Keen of the technology division of Atomic Energy Research Establishment, Harwell, England, publishing in the New Scientist of Feb. 16, 1978, propose that liquid waste be converted into glass and encased in steel cylinders 60 cm in diameter and about 3 meters long, each containing about 1.4 tons of glass. They consider that the glass would probably require cooling for several years before final disposal. As the repository for the cylinders, emplacement on or under the ocean bed appears to be a very safe method, according to these authors. They also view favorably the study of disposal in salt formations or clay or hard rock formations, suggesting that the repository be at least 300 m deep so as to lie below the permafrost level in any future ice age. Even so, they do not propose that disposal be carried out at this time, but rather that further data be accumulated. They suggest that small bore holes be drilled so as to investigate fracture structure using television and various physical techniques. In addition to the problems of storage, there are beginning to be difficulties in the above-ground transport of nuclear wastes. Perceived dangers have led to laws and ordinances restricting passage of vehicles carrying radioactive cargo, for example, in New York, New Jersey and Connecticut (New York Times, Apr. 17, 1978). A number of writers have also raised the concern that fissile isotopes (uranium 235 and plutonium) from nuclear power cycles may, because of their usability in nuclear weapons, become targets for terrorists or blackmailers. This imposes the additional requirement of strict accountability for the wastes during disposal, a condition difficult to satisfy when large numbers of billets must be handled. Methods of preparing fused glass bodies and suitable compositions have received considerable attention, J. R. Grover et al in U.S. Pat. No. 3,321,409 proposing to mix a radioactive waste liquid with a dry powder in a container, driving off the water and heating the product to fusion to form a glass. Joseph Kivel et al in U.S. Pat. No. 3,364,148 manufacture a source of radioactive energy by enclosing an insoluble radioactive material in a heat-fusible, continuous matrix comprising at least 92% by weight of silica, the peripheral portion of the matrix being free of the radioactive material. H. D. Bixby in U.S. Pat. No. 3,249,551 teaches the disposal of high-level radioactive waste materials by mixing the waste materials in clay and firing the mixture to make a ceramic body; the ceramic body is then covered with a ceramic glaze. F. C. Arrance in U.S. Pat. No. 3,093,593 disposes of radioactive waste by mixing same with ceramic materials, adding water to the mixture, shaping into porous pieces, pre-firing the pieces to destroy the ion-exchange capacity of the ceramic materials, saturating the pieces with radioactive waste materials by absorption, drying and finally firing. Kuan-Han Sun et al describe the preparation of a radioactive fluophosphate glass composition and making glass fibers of same, in their U.S. Pat. No. 3,373,116. According to the inventors, the glass may be used either in the form of thin glass fibers or small glass particles as a fuel for nuclear reactors. W. W. Schulz et al U.S. Pat. No. 4,020,004 describe the manufacture of a borosilicate glass incorporating radioactive cesium. Werner Hild et al in U.S. Pat. No. 3,971,717 propose to make solid glass blocks containing radioactive wastes and then to place them in water in order to condition the water, said conditioning including sterilization and facilitation of the filterability of the sludge. As is evident from the above, the incorporation of radioactive waste materials in a glass, particularly of the borosilicate type, has received considerable attention, based, apparently, on the belief that the glass is essentially unattackable by ground water; it is still an open question whether glass is actually impervious to attack by water over period of time extending to millions of years. In fact, accelerated tests at high pressures and temperatures by G. J. McCarthy et al indicate, to the contrary, that such glass is subject, first, to discoloration and fissure formation and, finally, to fracture and crystallization (Chemical and Engineering News, June 1, 1978, page 28). In addition, the concept of storage under ground is also looked upon with more or less favor but, as mentioned above, it is beginning to be realized that the integrity of the storage region cannot be predicted with any degree of certainty, so that this solution to the problem of waste disposal must be regarded as flawed. Other modes of disposal, such as firing the waste into solar orbit have also been proposed, but these are not economically feasible at this time. It is evident that either a new approach or an improved approach, eliminating the aforenoted problems, is needed. SUMMARY OF THE INVENTION Nuclear waste is incorporated into a glass by any convenient or conventional method, and the molten mixture is drawn into fibers. Preferably, a pool of water is provided for relatively short-term storage of the fibers, during which the intensity of radiation will decrease rapidly, the disposal of the heat generated in the process being more readily effected at this stage of the processing that at a more advanced stage. The fibers are then made into a bundle or cable, the diameter of the fibers and the number of fibers in the cable being such that the cable is flexible and can be wound on a support. The cable is fed through an underground duct to a well-head and through a well leading deep into the earth to a storage chamber in which winding apparatus winds the flexible cable onto a support for long-term storage. A buffer device for the temporary storage of a portion of the cable is provided between the cable-fabricating plant and the duct to accommodate momentary inequalities between the rates of fabrication and transport through the duct; another such buffer device is installed, for a similar reason, at the well-head. The manufacturing and transport processes are remotely controlled throughout. During both transport and storage, critical information about the state of the cable is received from an in-cable monitoring system. The cable is so devised that it can be withdrawn from the support to the surface of the earth should the monitoring system indicate that the integrity of the storage chamber is endangered or breached. The cable may also be retrieved for harvesting isotopes or for incorporating additional nuclear waste after the activity of the cable has decreased. As long as no cause for concern or economic motive for retrieval exists, the cable may be left in place indefinitely. Feeding the cable to the waste-receiving facility and then to the storage chamber is facilitated by provision of leaders, that is, non-radioactive segments of the cable, at the forward and rearward ends of the cable. The flexibility of the cable is enhanced by making it in flattened form, that is, in the form of a belt. The cable may be color-coded for identification and may have a coating therearound for retention of glass fragments. Longitudinal variations in the radiation spectra may be used to provide further information for characterizing and identifying locations along the cable. Transfer of the nuclear waste from the fabrication plant and its buffer device to the well-head is facilitated by provision of an underground conduit through which the cable may be fed with the aid of the leaders. As is evident, a cable is much more readily transported automatically from the melting tank to the well-head than would be glass billets. With the nuclear waste in the form of a cable transportable through a conduit, the problem of transporting nuclear waste above ground is completely eliminated. It is similarly evident that the problem of accounting for a single cable is much less severe than that of keeping track of the thousands of billets it replaces. The integrity of the cable may be monitored by means of light pulses transmitted through optical fibers associated with the cable. The underground conduit is positioned far enough below the surface of the earth so that the earth serves to screen out all but a minimal portion of the radiation from the nuclear waste. The optimal depth for the duct is best determined either by measurements on simulated configurations or the adaptation of existing computer codes to a linear source model. The principles of shielding are discussed in L. Wang Lau, Elements of Nuclear Reactor Engineering, Gordon & Breach, New York, 1974, where the formal solution for such a model is given. However, rough calculations based on rules of thumb also given by Lau suffice to give an estimate on the depth required. These indicate that a depth equivalent to 2.5 meters of concrete will attenuate the gamma radiation by a factor of 10.sup.8 and the neutron flux by 10.sup.25. Beta and alpha particles and heavy ions are stopped much more readily than these. The actual depth required will depend on the nature of the soil covering the duct, and an additional safety factor to account for such eventualities as soil erosion and digging by animals and human intruders will be necessary, but a depth of 4 to 5 meters can be entirely satisfactory. Accordingly, an object of the present invention is a method of forming nuclear waste into glass fibers and then into a flexible cable, which can be stored retrievably in an underground repository. Another object of the present invention is a method of increasing the utilization of an underground repository for the storage of nuclear waste products by periodically retrieving said waste products after partial decrease in activity thereof and addition of further nuclear waste products. A further important object of the present invention is a cable of glass fibers containing nuclear waste products, the cable being sufficiently flexible to be wound on a support in a repository deep beneath the surface of the earth and being provided with means for retrieving said cable should such retrieval become desirable or necessary. A significant object of the present invention is a plant for incorporating nuclear waste products in glass fibers, forming said fibers into a flexible cable, transporting said cable underground to a well-head, dropping said cable into a repository deep below the surface of the earth, winding said cable onto a support, and retrieving said cable from said repository should such retrieval become desirable or necessary. Yet another important object of the present invention is a plant as described further having a pool of water equipped with fiber-handling devices, and having a buffer storage region in which said fibers can be stored immediately after being drawn, to wait out the initial period during which heat evolution is at a maximum and for removal of the heat generated during this period, prior to their assembly into said cable. A further significant object of the present invention is a unique monitoring system that continually tests the integrity and state of the cable by the passage and modulation of light pulses transmitted along optical fibers contained in the cables alongside the radioactive fibers. Still other objects and advantages of the invention will in part be obvious and will in part be apparent from the specification. The invention accordingly comprises the several steps and the relation of one or more of such steps with respect to each of the others, the apparatus embodying features of construction, combinations and arrangements of parts which are adapted to effect such steps, and the article which possesses the characteristics, properties and relations of elements, all as exemplified in the detailed disclosure hereinafter set forth, and the scope of the invention will be indicated in the claims.
	description	The present invention relates to the determination of dryout properties in nuclear light water reactors, more specifically in a boiling water reactor (BWR). The invention is in particular related to a method of determining the so-called R-factor, which is used when determining the dryout properties in a nuclear boiling water reactor. The R-factor(s) is a concept that is known to a person skilled in the art, and which accounts for the weighted local power influence on a nuclear fuel rod, including contributions from neighbouring fuel rods. The invention also concerns a processor configured for automatically determining the R-factor, a computer program product, a method of determining the critical power for a bundle of fuel rods, a nuclear energy plant, and a method of operating a nuclear energy plant. The fuel rods in a BWR core are grouped in bundles with spacers and usually also end plates to keep the fuel rods in each bundle in a predetermined geometry. The predetermined rod lattice may be regular or irregular and even change axially. The bundles are then enclosed by channels to direct the coolant flow upward and give the fuel arrangement mechanical and thermal hydraulic stability and facilitate handling and exchange of the fuel. A fuel assembly may comprise several (for example 4) bundles (sometimes also referred to as subbundles) of fuel rods. In other constructions, each fuel assembly includes only one bundle of fuel rods. The fuel bundle may vary considerably in size concerning the number of fuel rods and it may also contain special purpose rods such as tie rods, water rods and burnable absorber rods. The bundle may comprise both full length rods and so-called part length rods, which are essentially shorter than the full length rods. The fissile material enrichment can vary within the fuel rod and may also vary from fuel rod to fuel rod. The present invention is applicable to all of these fuel arrangements and their operation in the reactor. As is well known to a person skilled in the art, in a BWR a cooling medium in the form of water flows through the fuel assemblies, which contain the fuel rods. The purpose of the water is to cool the fuel rods and to act as a neutron moderator. A mixture of steam and water flows through the fuel bundle, providing cooling for the rods by convective and boiling heat transfer. As the steam quality (the steam content fraction) of the coolant increases, the flow pattern changes. At a certain point in the bundle, an annular flow pattern is formed. This implies existence of a thin liquid film on the surface of the rods, and a mixture of vapour and droplets in the channels between the rods. The existence of this film allows for efficient heat transfer from the rods to the coolant. This enables both effective steam generation and prevents the rods from overheating. The breakdown of this film is referred to as dryout. In a BWR, dryout should be avoided. Dryout deteriorates heat transfer from the fuel rods to the reactor cooling medium and therefore leads to an increased temperature of the walls of the fuel rods. The increased temperature can damage the fuel rods. If a BWR is operated at or above a certain high power, the so-called critical power (CP), dryout may thus occur. In order to avoid dryout, the reactor is therefore operated at a lower power, such that a certain safety margin exists, the so-called dryout margin. A measure of the dryout margin is the critical power ratio (CPR). The CPR can be defined as the following ratio:CPR=(critical power)/(actual power) The CPR can be calculated locally for a large number of points in the reactor core. The smallest value of the CPR in any point is called the minimum critical power ratio MCPR. In the following critical power and critical heat flux and critical steam quality are treated as synonymous or equivalent entities as there exist straight forward physical transformation laws between them in steady state operation. With the coolant flow and the inlet enthalpy known, the steam quality directly provides the fuel arrangement power with steam/water thermodynamic data and vice versa. Different methods of determining the critical power are known in the prior art. EP 1 775 732 A1, and the corresponding U.S. patent application Ser. No. 11/512,938, which are incorporated herein by reference, describe one such method. Independently of with which method the critical power is determined, it is usually necessary to take the above mentioned R-factor into account. The R-factor is also mentioned in the above cited EP 1 775 732 A1 (and the corresponding US patent application). According to the prior art, the R-factor for a certain fuel rod in a bundle of fuel rods is normally determined by taking the effect from neighbouring fuel rods into account and by using a predetermined weight function for the axial variation of the R-factor. This weight function is normally such that the upper levels of the fuel bundle have a higher weight than the lower levels. The weight function is common to all the fuel rods in the bundle. Furthermore, the levels above part length rods are normally compensated for by using very high additive constants. In this manner, according to the prior art, it is possible to determine an R-factor for every fuel rod in the bundle of fuel rods. The highest R-factor for the different fuel rods in the fuel bundle is taken as the R-factor for the whole bundle. This R-factor is used when determining the critical power ratio for the fuel bundle in question. An object of the present invention is to provide an improved method of determining the R-factor(s) in a nuclear light water reactor of the boiling water reactor kind. A further object is to provide such a method which takes the properties of the individual fuel rods better into account than according to the prior art. Another object is to provide such a method, which results in a more accurate R-factor or R-factors for a fuel bundle that includes part length fuel rods. These objects are achieved by a method as defined in claim 1. According to the invention, a local R-factor (Ri(z)) is thus determined for each fuel rod (i) in said bundle and for each of a plurality of levels (z) in the axial direction. Furthermore, the individual axial heat generation profile for a certain fuel rod (i) is taken into account when determining the local R-factors (Ri(z)) for said fuel rod (i). With this method, the R-factors, and thereby the dryout properties, can be determined more accurately than with previous methods. With the present invention it is thus not necessary to use the above described predetermined weight function for the axial variation of the R-factor. Such a predetermined weight function could not be optimized for each level for each fuel rod. However, with the present invention, the local R-factors can be determined accurately since the individual axial heat generation profiles for the fuel rods are used when determining the local R-factors. Furthermore, with the present invention also the local R-factors for shorter fuel rods (so-called part length rods) can be determined accurately, since the individual axial heat generation profile for the fuel rods are used when determining the local R-factors. With the present invention it is therefore not necessary to compensate for the shorter fuel rods by using the above mentioned very high additive constants. According to a preferred implementation of the method according to the invention, said number of levels is at least equal to 10. By using at least 10 levels, the R-factor profile is determined with acceptable accuracy. Although the local R-factors could be determined for an infinite number of levels (continuously in the axial direction), the local R-factors are preferable determined for a limited number of levels in the axial direction, in order to facilitate the determination. The number of levels could for example be between 15 and 50, preferably between 20 and 30, for example 25. Preferably, said bundle of nuclear fuel rods includes at least 15 fuel rods. The bundle can be a so-called subbundle and the number of fuel rods in this bundle may for example be 24. Alternatively, the bundle can include a larger number of fuel rods, for example all the fuel rods of a fuel assembly. According to a preferred manner of carrying out the method according to the invention, the method comprises the determination of a total R-factor (Rz) for each of said levels (z) for the whole bundle of fuel rods, wherein said total R-factor (Rz) at a level (z) is determined as the maximum of said local R-factors (Ri(z)) at said level (z) in said bundle of nuclear fuel rods. Such a total R-factor is convenient to use when determining the dryout properties for the fuel assembly or for the subbundle. According to a manner of carrying out the method according to the invention, the determination of said local R-factor (Ri(z)) at a level (z) for a certain fuel rod (i) includes the determination of the double integrated heat generation rate of said fuel rod (i) up to the level (z) relative to the double integrated average heat generation rate of all fuel rods in the bundle up to the level (z). Such a determination has been found to provide an advantageous manner of determining the local R-factors. It should be noted that the concept “integrated” does in this document not necessarily mean that the integration is continuous. Instead, according to a preferred embodiment, the “integration” is done for discrete levels. The integral symbol in the shown formulas can therefore instead be seen as a summation symbol concerning such discrete levels. Preferably, said determination also includes a normalisation by means of the integrated heat generation rate of said fuel rod up to the level (z) relative to the integrated average heat generation rate of all fuel rods in the bundle up to the level (z), wherein the normalised determination of the double integrated heat generation rate of said fuel rod (i) up to the level (z) relative to the double integrated average heat generation rate of all fuel rods in the bundle up to the level (z) is carried out by determining: [ ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q _ ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q i ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ] aor an equivalent expression, where z=axial position z′=integration variable representing the axial position z″=integration variable representing the axial position qi(z)=linear heat generation rate of rod i at the level z q(z)=average linear heat generation rate of all rods in the bundle at the level z z0=the axial position of bulk boiling boundary (zero steam quality limit) in said bundle a=a constant, 0<a<1 Such a normalisation facilitates the determination of the R-factor influence on the dryout properties. According to one manner of carrying out the method according to the invention, the determination of said local R-factor (Ri(z)) at a level (z) for a certain fuel rod (i) includes the determination of the sum of the integrated heat generation rate of said fuel rod up to the level (z) and a fraction of the integrated heat generation rates of fuel rods neighbouring said fuel rod up to the level (z), relative to the integrated average heat generation rate of all fuel rods in the bundle up to the level (z). Such a determination improves the accuracy of the determination of the local R-factors. The mentioned determination of the sum of the integrated heat generation rate of said fuel rod up to the level (z) and a fraction of the integrated heat generation rates of fuel rods neighbouring said fuel rod up to the level (z), relative to the integrated average heat generation rate of all fuel rods in the bundle up to the level (z) is preferably carried out by determining: ( ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + c ⁢ ∑ j ∈ S i ⁢ ( ∫ z 0 z ⁢ q j ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + d ⁢ ∑ k ∈ D i ⁢ ( ∫ z 0 z ⁢ q k ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b ( 1 + cN S i + dN D i ) ⁢ ( ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b or an equivalent expression, where the symbols are as explained above and where qj(z)=linear heat generation rate of side neighbouring fuel rod j at the level z qk(z)=linear heat generation rate of diagonal neighbouring fuel rod k at the level z Si=the set of side neighbouring fuel rods for the fuel rod i Di=the set of diagonal neighbouring fuel rods for the fuel rod i NSi=the number of side neighbouring fuel rods for the fuel rod i NDi=the number of diagonal neighbouring fuel rods for the fuel rod i b=a constant, 0<b<1 c=a constant, 0<c<1 d=a constant, 0<d<c This has proved to be an accurate an efficient manner of performing the determination. Preferably, the determination of said local R-factor at a level (z) for a certain fuel rod (i) is carried out by determining: R i ⁡ ( z ) = ( 1 + e i ) [ ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q _ ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q i ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ] a ⁢ ( ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + c ⁢ ∑ j ∈ S i ⁢ ( ∫ z 0 z ⁢ q j ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + d ⁢ ∑ k ∈ D i ⁢ ( ∫ z 0 z ⁢ q k ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b ( 1 + cN S i + dN D i ) ⁢ ( ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b or an equivalent expression, where the symbols are as explained in the previous embodiments, and where ei=the dryout sensitivity constant for the fuel rod i (also denoted “rod constant”) According to another aspect, the invention provides a processor configured for automatically determining the R-factor for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind. The reactor comprises a plurality of bundles of nuclear fuel rods, wherein the fuel rods in the bundle are arranged side by side, at least substantially parallel to each other, and extend essentially in an axial direction. The R-factor is a factor that accounts for the weighted local power influence on a fuel rod, including contributions from neighbouring fuel rods. According to the invention, the processor is configured with an input receiving data concerning the linear heat generation rate (qi(z)) of the different fuel rods (i) at the different levels (z) in said bundle and to determine the R-factor in accordance with any one of the preceding embodiments. Similarly, the invention provides a computer program product directly loadable into the internal memory of a computer, which computer program product comprises a computer program configured to carry out a method according to any one of the preceding embodiments of the method. With such a processor, and with such a computer program product, the advantages described above in connection with the method are achieved. The invention also provides a method of determining the critical power for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind, wherein said method includes the determination of an R-factor according to any one of the preceding embodiments of the method. The invention also provides a method of determining the critical power by means of the local steam quality at dryout (XDO) for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind. This method comprises the following: determining how the local steam quality at dryout (XDO) depends on the flow of the cooling medium through the nuclear fuel bundle (f1(G)), determining how the local steam quality at dryout (XDO) depends on the axial power profile of the nuclear fuel bundle (f2(I2)), determining how the local steam quality at dryout (XDO) depends on the R-factor of the nuclear fuel arrangement (f3(R)), determining how the local steam quality at dryout (XDO) depends on the pressure of the cooling medium in the nuclear fuel arrangement (f4(P)), and determining the local steam quality at dryout (XDO) on the basis of the previous determinations. This method is characterised in that the R-factor dependence is determined by using the method of any one of the embodiments of the method described above for determining the R-factor. The invention also provides a nuclear energy plant comprising a nuclear light water reactor of the boiling water reactor kind. The plant comprises a control unit arranged to carry out a method according to any one of the two preceding paragraphs. According to one embodiment of the nuclear energy plant, it includes operation parameter detectors, arranged to detect operation parameters of the nuclear reactor during operation, wherein the control unit is arranged to receive information concerning said operation parameters from the detectors and to use these operation parameters when carrying out the method. According to one embodiment, the control unit comprises control outputs arranged to control the operation of the nuclear reactor in dependence on said method carried out by the control unit. Furthermore, the invention concerns a computer program product directly loadable into the internal memory of a computer which can form part of the above defined control unit, which computer program product comprises a computer program configured to carry out a method according to any one of the above embodiments of the method of determining the critical power. Finally, the invention provides a method of operating a nuclear energy plant comprising a nuclear light water reactor of the boiling water reactor kind, wherein said method includes the following steps: provide information concerning operation parameters of the nuclear reactor, use this information in a method according to any one of the above embodiments of the method of determining the critical power, and control the operation of the nuclear reactor in dependence on the previous method step. All these different aspects of the invention have advantages corresponding to those described above. The different aspects of the invention can be used on a nuclear energy plant when in operation. However, the invention as defined in claims 1-13 can also be used before the nuclear energy plant is in operation, for example in order to determine the dryout properties of the nuclear energy plant before it is actually in operation. In this manner it can be ensured that the correct dryout margin is the case before the plant is actually in operation. FIG. 1 shows schematically a nuclear energy plant, which constitutes an embodiment of the present invention. The nuclear energy plant comprises a nuclear light water reactor of the boiling water reactor kind. The nuclear reactor has a reactor vessel 3 in which the reactor core 5 is located. As is known to a person skilled in the art, the reactor core 5 comprises a plurality of bundles of nuclear fuel rods, wherein fuel rods in the bundle are arranged side by side, at least substantially parallel to each other, and extend essentially in an axial direction. Water is fed to the reactor vessel 3 via a water inlet 7 with the help of a pump 9. The generated steam leaves the vessel 3 via an outlet 11. Control rods 13 can be moved relative to the core 5 with the help of a control rod drive unit 15. The nuclear energy plant has a control unit 17, which suitably includes a computer. This control unit 17 is arranged to carry out a method according to the invention. The control unit 17 can thus for example be arranged to calculate the margin to dryout for different parts of the core 5 of the nuclear reactor by being arranged (programmed) to carry out a method according to the invention for example for determining the critical power, and to thereby determine the R-factor for a bundle of nuclear fuel rods according to a method according to the invention. The control unit 17 can be connected to operation parameter detectors 19, arranged to detect operation parameters of the nuclear reactor during operation. The detectors 19 can directly or indirectly detect operation parameters such as the mass flow of the cooling medium (the water), the pressure of the cooling medium, the position of the control rods 13 in the reactor core 5 and the neutron flux in different parts of the core 5. It is known to a person skilled in the art how to detect such operation parameters of a nuclear reactor. The control unit 17 is thus arranged to receive information concerning said operation parameters from the detectors 19 and to use these operation parameters when carrying out the mentioned method, for example for determining a dryout margin in different parts of the reactor core 5. Based on the calculated dryout margin, a person responsible for the operation of the nuclear energy plant can increase or decrease the power with which the nuclear reactor operates. Alternatively, the control unit 17 can have control outputs 21 arranged to automatically control the operation of the nuclear reactor in dependence on said method carried out by the control unit 17. It should be noted that the concept “control unit” as used herein thus includes two possibilities: either the control unit 17 constitutes a supervision unit which supplies information to a person (the operator), who can then manually control the operation of the nuclear energy plant (an open loop), or the control unit 17 can include means for automatically controlling the nuclear energy plant (a closed loop). However, in both cases the control unit preferably includes means (e.g. a computer) arranged to automatically carry out a method according to the invention, in order to provide information concerning the dryout properties of the nuclear reactor. The person skilled in the art knows how to control the power of a nuclear reactor. This can for example be done by changing the mass flow of the cooling medium, with the help of the pump 9, or by changing the position of the control rods 13, with the help of the control rod drive unit 15. The outputs 21 from the control unit 17 can thus be arranged to change for example the mass flow of the cooling medium or the position of the control rods 13. The invention also provides a computer program product 23 directly loadable into the internal memory of a computer which can form part of the control unit 17. The computer program product 23 comprises a computer program recorded on a computer-readable medium, the program being configured to cause the computer to carry out a method according to the invention for determining the R-factor and/or the critical power for a bundle of nuclear fuel rods in the nuclear reactor. 23 may also refer to a processor configured for automatically determining the R-factor for a bundle of nuclear fuel rods in the nuclear light water reactor. As already mentioned, the R-factor is a factor that accounts for the weighted local power influence on a fuel rod, including contributions from neighbouring fuel rods. The processor is configured with an input receiving data concerning the linear heat generation rate (qi(z)) of the different fuel rods (i) at the different levels (z) in said bundle and to determine the R-factor in accordance with a method according to the invention. With reference to FIG. 2, an example of a method, according to the invention, of operating a nuclear energy plant will now be described. The nuclear light water reactor is of the boiling water reactor kind. Information is provided concerning operation parameters of the nuclear reactor as described above. This information is used in a method, according to the invention, of determining the critical power for a bundle of nuclear fuel rods in the nuclear reactor. The critical power can be determined in different manners, for example as described in the above mentioned EP 1 775 732 A1, and the corresponding U.S. patent application Ser. No. 11/512,938. The critical power may thus be determined by means of the local steam quality at dryout (XDO) for a bundle of nuclear fuel rods by: determining how the local steam quality at dryout (XDO) depends on the flow of the cooling medium through the nuclear fuel bundle (f1(G)), determining how the local steam quality at dryout (XDO) depends on the axial power profile of the nuclear fuel bundle (f2(I2)), determining how the local steam quality at dryout (XDO) depends on the R-factor of the nuclear fuel arrangement (f3(R)), determining how the local steam quality at dryout (XDO) depends on the pressure of the cooling medium in the nuclear fuel arrangement (f4(P)), and by determining the local steam quality at dryout (XDO) on the basis of the previous determinations. As explained in the above cited documents, the local steam quality at dryout XDO can be thus described as a function:XDO=f1(G)f2(I2)f3(R)f4(P)+optional terms When XDO or another suitable measure of the dryout property has been determined, the operation of the nuclear reactor can be controlled in dependence on the determined property, such that the nuclear reactor is operated with a sufficient safety margin. When determining the R-factor dependence, the R-factor is determined by using the method according to the present invention. One manner of doing this is described below. The R-factor is thus determined for a bundle of nuclear fuel rods in a nuclear light water reactor of the boiling water reactor kind, which reactor comprises a plurality of bundles of nuclear fuel rods. The fuel rods in the bundle are arranged side by side, at least substantially parallel to each other, and extend essentially in an axial direction. The bundle of nuclear fuel rods may for example include 24 fuel rods. For determining the R-factor for a bundle, first a local R-factor (Ri(z)) is determined for each fuel rod (i) in said bundle and for each of a plurality of levels (z) in said axial direction. The number of levels may for example be 25. The individual axial heat generation profile for a certain fuel rod (i) is taken into account when determining the local R-factors (Ri(z)) for said fuel rod (i). A total R-factor (Rz) for each of said levels (z) for the whole bundle of fuel rods can thereby be determined, wherein said total R-factor (Rz) at a level (z) is determined as the maximum of said local R-factors (Ri(z)) at said level (z) in said bundle of nuclear fuel rods. The determination of said local R-factor (Ri(z)) at a level (z) for a certain fuel rod (i) includes the determination of the double integrated heat generation rate of said fuel rod (i) up to the level (z) relative to the double integrated average heat generation rate of all fuel rods in the bundle up to the level (z). The determination also includes a normalisation by means of the integrated heat generation rate of said fuel rod up to the level (z) relative to the integrated average heat generation rate of all fuel rods in the bundle up to the level (z). Furthermore, the determination of said local R-factor (Ri(z)) at a level (z) for a certain fuel rod (i) includes the determination of the sum of the integrated heat generation rate of said fuel rod up to the level (z) and a fraction of the integrated heat generation rates of fuel rods neighbouring said fuel rod up to the level (z), relative to the integrated average heat generation rate of all fuel rods in the bundle up to the level (z). The determination of said local R-factor at a level (z) for a certain fuel rod (i) can thereby be is carried out by determining: R i ⁡ ( z ) = ( 1 + e i ) [ ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q _ ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ⁢ ∫ z 0 z ⁢ ∫ z 0 z ′ ⁢ q i ⁡ ( z ″ ) ⁢ ⁢ ⅆ z ″ ⁢ ⁢ ⅆ z ′ ] a ⁢ ( ∫ z 0 z ⁢ q i ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + c ⁢ ∑ j ∈ S i ⁢ ( ∫ z 0 z ⁢ q j ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b + d ⁢ ∑ k ∈ D i ⁢ ( ∫ z 0 z ⁢ q k ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b ( 1 + cN S i + dN D i ) ⁢ ( ∫ z 0 z ⁢ q _ ⁡ ( z ′ ) ⁢ ⁢ ⅆ z ′ ) b or an equivalent expression, where z=axial position z′=integration variable representing the axial position z″=integration variable representing the axial position qi(z)=linear heat generation rate of rod i at the level z q(z)=average linear heat generation rate of all rods in the bundle at the level z z0=the axial position of bulk boiling boundary (zero steam quality limit) in said bundle qj(z)=linear heat generation rate of side neighbouring fuel rod j at the level z qk(z)=linear heat generation rate of diagonal neighbouring fuel rod k at the level z Si=the set of side neighbouring fuel rods for the fuel rod i Di=the set of diagonal neighbouring fuel rods for the fuel rod i NSi=the number of side neighbouring fuel rods for the fuel rod i NDi=the number of diagonal neighbouring fuel rods for the fuel rod i a=a constant, 0<a<1 b=a constant, 0<b<1 c=a constant, 0<c<1, preferably 0<c<0.25 d=a constant, 0<d<c, preferably 0<d<0.125 ei=the dryout sensitivity constant for the fuel rod i (also denoted “rod constant”) The concepts used are known to a person skilled in the art. However, with reference to FIG. 3, it will now be explained what is meant by “side neighbouring fuel rods” and “diagonal neighbouring fuel rods”. FIG. 3 show schematically a cross section of a bundle of 24 fuel rods. Four such bundles may together form a fuel assembly. As an example for explaining side and diagonal neighbouring fuel rods, we may consider the fuel rod marked 31. This fuel rod has three side neighbouring fuel rods, i.e. the fuel rods 32, 33 and 34, and two diagonal neighbouring fuel rods, i.e. the fuel rods 35 and 36. The present invention thus provides methods and devices for determining the R-factor and dryout properties in advantageous manners. It is thereby possible to predict and control the operation of a nuclear fuel plant with higher accuracy, and thereby to operate the plant with a high efficiency, while making sure that the dryout margin is sufficient. The invention is not limited to the described embodiments, but can be varied within the scope of the claims. It should also be noted that a mathematical expression can normally be written in different manners and still have the same meaning, or approximately the same meaning. Consequently, the claims should not be seen as being limited to the exact mathematical expression defined in some of the claims. The claims are thus intended to cover equivalent expressions of the formula and alternative formulations that constitute approximations of the formulas. Such transformations are regularly done for numerical evaluations and can be tailored for high accuracy over a predetermined application range.
	abstract	A reactor cooling system for cooling a nuclear reactor using nitrogen comprising a refrigeration unit for cooling and compressing nitrogen gas into liquid nitrogen, a liquids storage tank to store liquid nitrogen, the tank in fluid communication with the refrigeration unit, a heat exchanger drop system in fluid communication with the liquids storage tank, adjacent to the nuclear reactor, wherein the nitrogen absorbs heat by becoming gaseous, a tank for receiving and holding nitrogen gas in fluid communication with the heat exchanger and in fluid communication with the refrigeration unit, and where the system is a closed-loop drop system.
	description	FIG. 1 is a sectional view, with parts cut away, of a boiling water nuclear reactor pressure vessel (RPV) 10. RPV 10 has a generally cylindrical shape and is closed at one end by a bottom head 12 and at its other end by a removable top head 14. A side wall 16 extends from bottom head 12 to top head 14. Side wall 16 includes a top flange 18. Top head 14 is attached to top flange 18. A cylindrically shaped core shroud 20 surrounds a reactor core 22. Shroud 20 is supported at one end by a shroud support 24 and includes an opposed removable shroud head 26. An annulus 28 is formed between shroud 20 and side wall 16. A pump deck 30, which has a ring shape, extends between shroud support 24 and RPV side wall 16. Pump deck 30 includes a plurality of circular openings 32, with each opening housing a jet pump 34. Jet pumps 34 are circumferentially distributed around core shroud 20. An inlet riser pipe 36 is coupled to two jet pumps 34 by a transition assembly 38. Each jet pump 34 includes an inlet mixer 40, and a diffuser 42. Inlet riser 36 and two connected jet pumps 34 form a jet pump assembly 44. Heat is generated within core 22, which includes fuel bundles 46 of fissionable material. Water circulated up through core 22 is at least partially converted to steam. Steam separators 48 separates steam from water, which is recirculated. Steam dryers 50 remove residual water from the steam. The steam exits RPV 10 through a steam outlet 52 near vessel top head 14. The amount of heat generated in core 22 is regulated by inserting and withdrawing a plurality of control rods 54 of neutron absorbing material, for example, hafnium. To the extent that control rod 54 is inserted adjacent fuel bundle 46, it absorbs neutrons that would otherwise be available to promote the chain reaction which generates heat in core 22. Control rod 54 couples with a control rod drive mechanism (CRDM) 58 to form a control rod apparatus 60 (shown in FIG. 3). CRDM 58 moves control rod 54 relative to a core support plate 64 and adjacent fuel bundles 46. CRDM 58 extend through bottom head 12 and is enclosed in a control rod drive mechanism housing 66. A control rod guide tube 56 extends vertically from control rod drive mechanism housing 66 to core support plate 64. Control rod guide tubes 56 restrict non-vertical motion of control rods 54 during control rod 54 insertion and withdrawal. Control rod guide tube 56 has a cruciform shape. In alternative embodiments control rod guide tube 56 can have other shapes, for example cylindrical, rectangular, or Y-shaped. FIG. 2 is a perspective side view of control rod 54. Control rod 54 includes at least one blade 70, a longitudinal tube 72 at the intersection of blades 70, a hub 74 coupled to blades 70, and a longitudinal axis 76 aligned with longitudinal tube 72. Control rod 54 further includes an upper end 80 and a lower end 82. In an exemplary embodiment, control rod 54 includes four radially extending blades 70 in a cruciform shape. In an alternative embodiment, control rod 54 includes other blade configurations, including for example, a Y-shaped blade configuration (not shown). Blades 70 intersect at longitudinal tube 72. Longitudinal tube 72 extends the length of control rod 54, including through hub 74. Hub 74 is integrally attached to blades 70 at lower end 82 of control rod 54. Any suitable material can be used for hub 74, for example, stainless steel XM19 or Nixe2x80x94Crxe2x80x94Fe alloy X-750. These alloys provide high strength and provide corrosion resistance in the environment of a boiling water nuclear reactor. Each blade 70 includes a first surface 84, a second surface 86, and a blade thickness 88 between first surface 84 and second surface 86. Blades 70 contain a neutron absorbing material (not shown) between first surface 84 and second surface 86 in a sealed, corrosion resistant condition allowing for an extended useful period. FIG. 3 is a schematic, partial cross sectional view of control rod apparatus 60. CRDM 58 includes a drive axis 90, an index tube 92 and a restraining device 94. Index tube 92 includes a first end 96 and an outer surface 98. A bayonet head 100 is secured to first end 96. Restraining device 94 engages index tube 92. In an exemplary embodiment, CRDM 58 is operated by a hydraulic motive system (not shown). CRDM 58 is operated to axially position index tube 92. In an alternative embodiment, a mechanical screw-type motive system (not shown) operates CRDM 58. Index tube 92 retractably extends through control rod drive mechanism housing 66 into control rod guide tube 56 to position control rod 54. Control rod apparatus 60 further includes a control rod coupling assembly 102 which releasably couples CRDM 58 and control rod 54, and is shown at lower end 82 (shown in FIG. 2) of control rod 54. FIG. 4 is a cross sectional view of an index tube 92. Index tube outer surface 98 includes an axial channel 104. Restraining device 94 includes a roller key 106 secured to control rod drive mechanism housing 66 slidably engaged in axial channel 104. Restraining device 94 engages index tube 92 so as to restrict rotational movement, while allowing vertical motion. FIG. 5 is an enlarged, schematic, partial cross-sectional view of control rod apparatus 60. Coupling assembly 102 releasably couples control rod 54 to CRDM 58. Coupling assembly 102 includes a shaft 120, a handle 122, and a bayonet socket 124. In one embodiment, coupling assembly 102 further includes an external hex nut 126 attached to hub 74 and an axial bearing 128. Axial bearing 128 is coupled to bayonet socket 124 and shaft 120. Axial bearing 128 abuts hub 74 to reduce friction between the control rod 54 and the coupling assembly 102. In another embodiment (not shown), shaft 120 secures directly to bayonet socket 124. Referring to FIGS. 2 and 5, shaft 120 includes a proximate end 130 and a distal end 132. Shaft 120 is received in and extends axially through longitudinal tube 72 of control rod 54. Shaft 120 is free to rotate in longitudinal tube 72. Proximate end 130 extends through hub 74. Distal end 132 of shaft 120 extends through control rod upper end 80. Handle 122 is threadedly coupled to distal end 132 of shaft 120. In alternative embodiments, handle 122 can be secured to shaft 120 by other suitable attachments, for example, by fasteners. Handle 122 is detachable from shaft 120 and includes a plate 134 and an opening 136. In alternative embodiments, handle 122 can include other configurations, for example, handle 122 can include other suitable shapes. For example, handle 122 can include a rod forming a closed loop (not shown) with shaft 120. Handle 122 further includes roller mechanisms 138 to facilitate operation between fuel bundles 46 (as shown in FIG. 1). In an alternative embodiment, roller mechanisms 138 are not included in handle 122. Referring to FIGS. 2 and 5, handle 122 further includes a first side 140 and a second side 142, and a handle thickness 144 between first side 140 and second side 142. Handle thickness 144 is about equal to or less than blade thickness 88 to facilitate use in reactor core 22. Handle 122 facilitates rotation of shaft 120 in longitudinal tube 72. Referring to FIG. 5, axial bearing 128 is secured to bayonet socket 124 and shaft 120. Bayonet socket 124 includes an upper end 146, a cylindrical body 148, and a coupling cavity 150. Axial bearing 128 is disposed between upper end 146 and hub 74 and rotatably engages hub 74 to facilitate rotation of bayonet socket 124. Axial bearing 128 is threadedly and detachably secured to bayonet socket 124. Axial bearing 128 is also threadedly and detachably secured to shaft 120. External hex nut 126 includes an outer portion 152 (shown in FIG. 6) attached to hub 74 and an inner wall 154 circumferentially enclosing cylindrical body 148 of bayonet socket 124. Cylindrical body 148 is free to rotate within hex nut 126 while bayonet socket 124 is retained by hex nut 126. Axial bearing 128 is secured to bayonet socket upper end 146. Axial bearing 128 is disposed between upper end 146 and hub 74 and rotatably engages hub 74 to facilitate rotation of bayonet socket 124. Axial bearing 128 is threadedly and detachably secured to bayonet socket 124. Axial bearing 128 is also threadedly and detachably secured to shaft 120. FIG. 6 is a side view of bayonet head 100 engaged with bayonet socket 124. Referring to FIGS. 5 and 6, bayonet socket 124 further includes an internal engagement flange 156 comprising four arcuate segments 158 forming an engagement aperture 160. Upper end 146 is secured to axial bearing 128, which is threadedly secured to proximate end 130 of shaft 120. In another embodiment, upper end 146 of bayonet socket 124 is threadedly secured to proximate end 130 of shaft 120 without axial bearing 128. Each segment 158 subtends slightly less than 45 degrees of radial arc. In alternative embodiments, as described below, different numbers of segments, with different arc spans can be used. Each segment 158 includes an internal face 162. Internal faces 162 are arcuate to facilitate engagement with bayonet head 100. Bayonet socket 124 is fabricated from any suitable material including, for example, stainless steel XM19 or Nixe2x80x94Crxe2x80x94Fe alloy X-750. These alloys provide high strength, permitting minimum size and weight of bayonet socket 124, and provide corrosion resistance in the environment of a boiling water nuclear reactor. Bayonet head 100 extends from index tube first end 96. Bayonet head 100 includes four members 170 in a cruciform arrangement. FIG. 7 is a top view of bayonet socket 124 disengaged from the bayonet head 100. Each member 170 subtends slightly less than 45 degrees of radial arc, complementary to segments 158. Each member 170 also includes a lower surface. 172. In alternative embodiments, a different number of members is used, including for example, two members 170 each subtending about 90 degrees of radial arc, complementary to two segments 158. Lower surfaces 172 are convex to facilitate engagement with internal faces 162 of socket 124. In another embodiment, lower surfaces 172 and internal faces 162 include other complementary shapes. Bayonet head 100 is also fabricated from any suitable material including, for example, stainless steel XM19 or Nixe2x80x94Crxe2x80x94Fe alloy X-750. These alloys provide high strength, permitting minimum size and weight of bayonet head 100, and provide corrosion resistance in the environment of a boiling water nuclear reactor. In operation, coupling assembly 102 facilitates a secure coupling of control rod 54 to CRDM 58 that precludes inadvertent uncoupling of control rod 54 from CRDM 58, while allowing uncoupling of control rod 54 from CRDM 58 for maintenance without rotation of control rod 54. Using standardized procedures for reactor maintenance, and with top head 14 and other components removed, a tool (not shown) is lowered to grasp handle 122. The tool then rotates handle 122, rotating shaft 120 in longitudinal tube 72 and rotating bayonet socket 124 about bayonet head 100. Restraining device 94 restricts rotation of index tube 92 and bayonet head 100. As handle 122 is rotated segments 158 disengage from members 170. segments 158 complete disengagement after about 45 degrees of rotation. FIG. 7 shows bayonet socket 124 and bayonet head 100 in a disengaged condition. Control rod 54 is uncoupled from CRDM 58 and is lifted from control rod guide tube 56. Installation and coupling of control rod 54 to CRDM 58 requires a similar operation. Supported by a tool (not shown) grasping handle 122, control rod 54 is lowered into control rod drive tube 56. Bayonet head 100 is received through engagement aperture 160 into coupling cavity 150 of bayonet socket 124. Internal faces 162 of segments 158 are below members 170. Handle 122 is rotated by the tool, causing rotation of shaft 120, and thus rotation of bayonet socket 124. Internal faces 162 rotate to engage lower surfaces 172. Rotation of about 45 degrees aligns handle 122 substantially co-planar with blade 70 and completes engagement of segments 158 with members 170. FIG. 6 shows bayonet head 100 engaged in bayonet socket 124. It is to be understood that the present invention is not limited to a bayonet socket with four segments coupling to a bayonet head with four members. Alternative configurations using other complementary arrangements of segments and members could be utilized. The above described coupling assembly 102 facilitates installation and removal of control rods 54 while providing reliable coupling of control rod 54 and CRDM 58. Coupling assembly 102 ensures retention and control of control rod 54 while facilitating rotation of bayonet socket 124 during maintenance procedures. Coupling assembly 102 is particularly advantageous where rotation of control rod 54 is restricted after fuel bundles 46 are removed. In addition, coupling assembly 102 facilitates replacement of control rod 54 when less than all adjacent fuel bundles 46 are removed, and does not require removal of CRDM 58. Furthermore, coupling assembly 102 can improve reliability and reduce maintenance time, as compared to a conventional control rod apparatus, with an overall reduction in maintenance cost and reduced outage time. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims.
	summary
045308144	abstract	Apparatus for superheating steam. In accordance with one aspect of the invention, the apparatus is provided with two banks of inclined tubes extending upwardly from an outlet header to respective inlet headers. The banks of tubes are disposed in the flow path of main steam through the apparatus and provide a flow of vapor for adding superheat to the main steam. In accordance with another aspect of the invention, the tubes extend upwardly to closed ends whereby a lower header acts as both a vapor inlet and condensate outlet. In accordance with another aspect of the invention, the vapor is provided by a fossil fuel-fired vapor generator to superheat main steam which has been provided by a nuclear steam generator.
	summary
	claims	1. A ventilation system for an operating space accessible to operators in a nuclear installation, the ventilation system comprising:an external inlet;a supply air line guided from said external inlet to the operating space;a first fan connected in said supply air line;a first inert gas adsorber column connected in said supply air line;an external outlet;an exhaust air line guided from the operating space to said external outlet;a second fan connected in said exhaust air line;a second inert gas adsorber column connected in said exhaust air line;a switching device for interchanging roles of said first and second inert gas adsorber columns;a circulating-air line;a CO2 adsorber column connected in said circulating-air line; anda circulating-air fan connected in said circulating-air line, said circulating-air line leading away from and back to the operating space, said second fan being able to be connected into said circulating-air line as said circulating-air fan. 2. The ventilation system according to claim 1, wherein said circulating-air line having an inlet side connected to said exhaust air line and an outlet side connected to said supply air line. 3. The ventilation system according to claim 1, wherein said first fan is disposed upstream of said first inert gas adsorber column, viewed in a direction of flow of supply air. 4. The ventilation system according to claim 3, further comprising at least one of a throttle or an air dryer connected into said supply air line between said first fan and said first inert gas adsorber column. 5. The ventilation system according to claim 1, wherein said second fan is disposed downstream of said second inert gas adsorber column, viewed in a direction of flow of exhaust air. 6. The ventilation system according to claim 1, further comprising a throttle connected into said exhaust air line upstream of said second inert gas adsorber column, viewed in a direction of flow of exhaust air. 7. The ventilation system according to claim 1, further comprising:an iodine filter connected into said supply air line; andan aerosol filter connected into said supply air line. 8. The ventilation system according to claim 7, wherein said iodine filter and said aerosol filter are disposed upstream of said first fan, viewed in a direction of flow of supply air. 9. The ventilation system according to claim 1, further comprising a stand-alone power supply module. 10. The ventilation system according to claim 1, wherein said switching device contains a plurality of 3-way valves. 11. A method for operating a ventilation system for an operating space accessible to operators in a nuclear installation, which comprises the steps of:connecting into a supply air line being guided from an external inlet to the operating space, a first fan and a first inert gas adsorber column;connecting into an exhaust air line being guided from the operating space to an external outlet, a second fan and a second inert gas adsorber column;providing a switching device for interchanging roles of the first and second inert gas adsorber columns;flowing supply air through one of the inert gas adsorber columns and the one inert gas adsorber column thus being loaded with radioactive inert gases, and exhaust air simultaneously flowing through the other inert gas adsorber column and the other inert gas adsorber column thus being backwashed, and in which roles of the first and second inert gas adsorber columns are interchanged by switching as soon as an adsorption capacity of a currently loaded inert gas adsorber column is exhausted; andproviding a circulating-air line, into which a CO2 adsorber column and a circulating-air fan are connected, the circulating-air line leads away from and back to the operating space, pressure being built up at least in one of the inert gas adsorber columns by means of the first fan, and, simultaneously, CO2 reduction being carried out by the CO2 adsorber column in a circulating-air mode. 12. The method according to claim 11, wherein, while the pressure is being built up, the operating space is ventilated exclusively by circulating air decontaminated from CO2. 13. The method according to claim 11, which further comprises simultaneously feeding supply air to the operating space by means of at least one of the inert gas adsorber columns, and CO2 reduction is carried out by the CO2 adsorber column in the circulating-air mode. 14. The method according to claim 11, which further comprises using the first fan as the circulating-air fan.
	claims	1. A packaging system for radioactive materials, which comprises, starting from the inside and working outwards:(1) a vial with closure for accommodating the radioactive material,(ii) a first casing to be opened, enclosing the vial, and essentially made from a transparent material, which has a sufficient or appropriate capture cross-section for shielding at least a part of the emitted radiation, and(iii) a second casing to be opened, made from a material with a high capture cross-section (Z) for essentially shielding the remaining radiation, the second casing enclosing the first casing,wherein the second casing can be opened by removing the cover so that when the casing is opened, the entire vial is essentially visible through the transparent first casing. 2. The packaging system as claimed in claim 1, in which the first casing is provided in the form of a cylinder, which is closed by cover plates at one or both ends. 3. The packaging system as claimed in claim 1, in which the second casing is provided in the form of a cylinder, which is closed by cover plates at both ends. 4. The packaging system as claimed in claim 1, in which the first casing encloses the vial in a tight fit. 5. The packaging system as claimed in claim 1, in which the second casing encloses the first casing in a tight fit. 6. The packaging system as claimed in claim 2, in which the cover plate(s) of the first is or are respectively made from the same material as the respective cylinder wall. 7. The packaging system as claimed in claim 1, in which the first casing can be opened by removing a cover and a top part of the vial is exposed when the casing is opened. 8. The packaging system as claimed in claim 7, in which the first casing is provided in the form of a cylinder closed by at least a top cover plate, the cover comprises the top cover plate and a part of the cylinder wall and the distance of the opening from the top cover plate is shorter than the distance of this opening from the bottom cover plate when the cover is removed. 9. The packaging system as claimed in claim 1, in which the second casing is provided in the form of a cylinder closed by cover plates, the cover comprises the top cover plate and a part of the cylinder wall and the distance of the opening from the top cover plate is greater than the distance of this opening from the bottom cover plate when the cover is removed. 10. The packaging system a claimed in claim 9, in which only the bottom cover plate of the second casing is essentially left on the packaging system when the second casing is opened by removing the cover. 11. The packaging system for as claimed in claim 1, in which the first and/or second casing has a mating shoulder or a thread. 12. The packaging system as claimed in claim 1, in which the vial is made from glass. 13. The packaging system as claimed in claim 1, in which the vial is a vial with a pointed base or a vial with a flat base. 14. The packaging system as claimed in claim 1, in which the vial is a primary packaging means authorized for drugs. 15. The packaging system as claimed in claim 1, in which the vial is closed by a stopper. 16. The packaging system as claimed in claim 1, in which the first casing is made from a transparent plastic of a sufficient thickness essentially to shield a β-radiation emitted by a radioactive solution. 17. The packaging system as claimed in claim 16, in which the plastic is selected from the group consisting of polyethylene, polypropylene, polycarbonate, polystyrene, polyethylene terephthalate, polyacrylate, polymethacrylate, copolymers containing them and mixtures thereof. 18. The packaging system as claimed in claim 1, in which the second casing is made from a metal or a metal alloy with a high Z of a sufficient thickness essentially to shield the remaining radiation completely. 19. The packaging system as claimed in claim 18, in which the metal or metal alloy is selected from the group consisting of Al, Ag, An, Pb, Cd, Ce, Cr, Co, Cu, Fe, Hg, Hf, Bi, In, Mg, Mn, Mo, Nb, Ni, Pd, Pt, Pr, Re, Rh, Sn, Si, Ta, Ti, Tb, Th, V, W, Y, Yb, Zn, Zr, Al/Mg, Al/Cu, Al/Cu/Mg, Al/Mg/Si, Al/Cr, Tinal alloy BB, copper alloys such as brass and bronzes, iron alloys such as Fe/Cr, Fe/Ni, Fe/Cr/Ni, Fe/Cr/Al, nickel alloys Ni/Ti, Ni/Cr, and Nitinol, platinum alloys, titanium alloys such as Ti/Al, Ti/Al/V and Ti/Mo, Woods alloys, Inconel, tungsten alloys such as Densimed and mercury alloys such as amalgams. 20. The packaging system as claimed in claim 1, in which the first casing has a cut-out above the closure of the vial. 21. The packaging system as claimed in claim 20, in which the first casing is provided in the form of a cylinder closed by at least a top cover plate, which has a centrally disposed opening. 22. The packaging system as claimed in claim 1, in which these materials are selected from the group consisting of solutions, powder, particles, granulate, lyophilisate, liposomes, nano-particles, emulsions or suspensions. 23. The packaging system as claimed in claim 1, in which the radioactive materials are selected front β-emitters, γ-emmiters and/or X-radiation emitting material, which has a maximum particle energy of at least 500 keV in the case of β-radiation (Εβmax) and/or a photon energy in the range of from 20 to 100 keV in the case of γ-radiation and/or X-radiation. 24. The packaging system as claimed in claim 23, in which the radioactive material contains the nuclides Sr-90, Y-90, Y-86, Sr-89, Tm-170, P-32, Ca-45, Cl-36, Ce-144, Tb-160, Ta-182, Tl-204, W-188, Re-188, Ir-192, Pd-103, Se-75, J-125, S-35, Lu-177, Ho-166, Re-186, Te-125m, Tc-99m or mixtures thereof. 25. The packaging system as claimed in claim 1, in which the first casing encloses the vial and a bottom insert disposed underneath the base of the vial in a tight fit. 26. The packaging system as claimed in claim 1, in which the vial is closed by a stopper and a flanged cap.
049873091	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a radiation therapy unit with a beam of rays propagating from a focal point along a beam axis and a radiator head arranged on the beam axis, having the following features: (a) the radiator head comprises a double-focus multi-leaf collimator; PA1 (b) the multi-leaf collimator exhibits a plurality of adjacently arranged diaphragm plates which in each case have two side faces, two front faces and an inside and an outside face; PA1 (c) means are provided for displacing each individual diaphragm plate. PA1 each side face of each diaphragm plate forms a part of a surface area of a cone, all such cones having both a common cone axis which extends perpendicularly to the beam axis through the focal point and a common cone point which coincides with the focal point, and PA1 means are provided for guiding the diaphragm plates so that each diaphragm plate performs a pure rotation about the cone axis during its displacement. 2. Discussion of Background Radiation therapy units are used in medicine for therapeutically treating tumors by means of high-energy photons or electrons. In this connection, it is of importance that the beam of rays generated by the unit has accurately defined characteristics with respect to field limiting. In this respect, radiation therapy units having multi-leaf collimators of the type initially mentioned are particularly suitable. Such a unit is known, for example, from U.S. Pat. No. 4,672,212. Instead of a conventional pair of collimator blocks, a multi-leaf collimator is used there. In order to avoid unwanted half shadows at the edge of the radiation field, the multi-leaf collimator is focussed twice. For this purpose, the individual diaphragm plates are segments of a circular ring and have a cross section of the shape of an equal-sided trapezoid. The multi-leaf collimator composed of the diaphragm plate is thus circularly bent in two directions which are perpendicular to one another and thus imitates a part of a spherical shell (only approximately, however). Each diaphragm plate is provided at its rear end with a rod pointing upwards. The diaphragm plate is advanced and retracted on this rod by means of a motor. The problem of this multi-leaf collimator lies in the fact that an unwanted leakage radiation occurs between the individual diaphragm plates. A further disadvantage lies in the complicated drive arrangement. This is because the diaphragm plates move on curved paths which change from plate to plate. The type of coupling proposed between motor and diaphragm plate leads to a non-linear relationship between plate advance and motor speed. The published European patent application EP-0,259,989 A1 describes quite a different multi-leaf collimator. This is used in a conventional radiation therapy unit in addition to the two pairs of collimator blocks, for shielding sensitive organs of the patient. The multi-leaf collimator consists of a plurality of laterally adjoining diaphragm plates. The essentially rectangular plates are slightly rounded on their front and have straight-line guide slots at the side faces. They run along straight tracks and are driven by associated motors by means of flexible cables. Although this multi-leaf collimator has a linear drive arrangement, it can only be used in conjunction with the conventional collimator blocks due to the lack of double focussing. SUMMARY OF THE INVENTION Accordingly, it is an object of this invention to create a radiation therapy unit of the type initially mentioned, which allows great freedom in shaping the radiation field. At the same time, the radiation field should be free of half shadows and leakage radiation. Finally, the radiator head should have as low a constructional height as possible. According to the invention, the solution consists in the fact that, in a radiation therapY unit of the tYpe initially mentioned, The core of the invention lies in the fact that the diaphragm plates are shaped in such a manner that their side faces adjoin one another in a form-closing manner and that, at the same time, the double focussing is retained during the displacement of the diaphragm plates. In contrast to the known double-focus multi-leaf collimator, each diaphragm plate has its individual shape which is given by its relative position in the entire package. In a particularly advantageous embodiment, the outside faces of all diaphragm plates arranged laterally adjacently have, overall, as an enveloping surface a part of a surface area of an outer cylinder the axis of which is the cone axis. The means for displacing the diaphragm plates engage the outside faces. This makes the drive arrangement particularly simple. The means for displacing preferably comprise for each diaphragm plate a toothed rail which is mounted on the outside face of the diaphragm plate, a worm-rack gear engaging this toothed rail and a stepping motor which actuates the gear. This ensures a linear relation between the speed of the stepping motor and the plate advance. It is particularly advantageous if the radiator head exhibits two multi-leaf collimators which are arranged above one another and are aligned perpendicularly to one another and if a matrix ionization chamber, by means of which the multi-leaf collimators are monitored, is arranged on the beam axis opposite to the radiator head. Further advantageous embodiments are obtained from the dependent claims.
055442102	summary	TECHNICAL FIELD This invention relates to a pressure vessel apparatus for containing fluid under high temperature and pressure. The apparatus disclosed herein has particular application for use as a nuclear reactor vessel and incorporates safety features. BACKGROUND ART My U.S. Pat. No. 5,217,681, issued Jun. 8, 1993, discloses a prestressed pressure vessel safety enclosure used as a pressure safety enclosure for a nuclear reactor pressure vessel or other primary system vessel containing fluid or gaseous material under high pressure. The special pressure vessel enclosure comprises a first pressure vessel containment assembly surrounding the primary pressure vessel. A pair of first upper and lower pressure vessel jackets are adapted to enclose and be spaced apart, respectively, from the upper and lower portions of the first pressure vessel containment assembly with the rims of the jackets adapted to be slidable and sealed with respect to the first pressure vessel containment assembly. The spaces between the jackets and pressure vessel containment assembly are filled with a high boiling point, low melting point metal. Upper and lower ring girders, connected to each other by tension tendon members, in conjunction with upper and lower jacket bearing plates and skirts are used to apply force to the respective upper and lower jackets for moving the jackets toward or away from each other. My U.S. Pat. No. 5,465,280, issued Nov. 7, 1995, discloses a nuclear reactor vessel employing bellows in the construction thereof which operate as fluid barriers, confine lead material filler, and allow for relative movement of structural components of the apparatus in a controlled manner. Additionally, the apparatus disclosed in U.S. Pat. No. 5,465,280 incorporates connector tendons of a specialized construction, incorporating two sets of tendons, one of which is prestressed almost to yield point, and the other of which is prestressed to a lesser degree. U.S. Pat. No. 5,204,054, issued Apr. 20, 1993, discloses a nuclear reactor system pressure vessel comprising a steel inner liner part, an intermediate insulative layer part and an outer concrete encasing part pre-stressed by cable tendons located inside the casing. Use of the pre-stressed construction allows for construction of pressure vessels of larger size. The outer vessel part can be of a cast single piece structure or it can be an integrated concrete segment assembled structure embodying pre-stressing cable tendons arranged in various orientations to effect pre-stressing. Further, the major portion of the pressure vessel can be disposed below grade to lessen the presence of vessel structure in a nuclear system containment. Cooling passages are provided in the pressure vessel to carry off reactor decay heat as well as heat in the concrete outer vessel part. Applicant is aware of a publication entitled Recent Investigations and Tests With the BBR Winding System for Circumferential Prestressing of Concrete Vessels and Containments authored by K. Schutt and F. E. Speck, published in 1993 by Elsevier Science Publishers B. V. in SMiRT-12 Transactions. The publication discloses the use of elongated reinforcement elements in the form of continuously wound prestressing wire strands or bands which are wound about the periphery of large prestressed concrete pressure vessels for nuclear power stations. The prestressed strands are applied in layers spirally wound over the whole width of channels formed at the outer periphery of the concrete pressure vessel, requiring much less space and making them easier to install and inspect compared to cable tendons. The following publications and United States patents are also believed to be representative of the state of the prior art: U.S. Pat. No. 3,433,382, issued Mar. 18, 1969, U.S. Pat. No. 3,775,251, issued Nov. 27, 1973, U.S. Pat. No. 4,192,718, issued Mar. 11, 1980, U.S. Pat. No. 3,445,971, issued May 27, 1969, U.S. Pat. No. 3,512,675, issued May 19, 1970, U.S. Pat. No. 3,653,434, issued Apr. 4, 1972, U.S. Pat. No. 3,606,715, issued Sep. 21, 1971, U.S. Pat. No. 5,229,067, issued Jul. 20, 1993, U.S. Pat. No. 5,047,201, issued Sep. 10, 1991, U.S. Pat. No. 4,859,402, issued Aug. 22, 1989, U.S. Pat. No. 4,650,642, issued Mar. 17, 1987, and U.S. Pat. No. 4,032,397, issued Jun. 28, 1977. Applicant has authored a paper entitled Prestressed Safety Enclosure (PSE) with Metallic Cushion for New or Existing Reactor Pressure Vessels, published in SmiRT 11 Transactions Vol. SD2 (August, 1991). DISCLOSURE OF INVENTION The present invention relates to pressure vessel apparatus defining a pressure vessel interior for containing fluid under pressure. The invention has particular application to nuclear reactor system pressure vessels, especially large size vessels. Existing pressure vessel designs have two major limitations which tend to restrict the overall safety and capacity of the vessel: 1. Diameter and pressure capacity of the vessel head is limited by the number, size and capacity of the studs tying the head to the vessel body. This limitation in turn limits the number and spacing of the various penetrations which can be located there. PA1 2. Diameter and pressure capacity of the vessel main body and top and bottom is limited by the maximum vessel shell thickness as dictated by manufacturing or fabrication technologies. For nuclear reactor vessels this limitation imposes upper limits on core size and aspect ratio and makes any system integrated design difficult, restricting the space available for steam separation or heat exchangers, spent fuel storage, and a large water reservoir. The invention described below overcomes these limitations. The pressure vessel of the present invention incorporates a number of features contributing to the structural stability and safety of such pressure vessels. In particular, such features contribute to the containment of radioactive or toxic materials within the pressure vessel in the event of generation of high pressures therein, for example due to a core melt event, steam explosions, and/or hydrogen explosions. The vessel is always (except during extreme accident) in a state of three dimensional compression. Relative movement between certain structural components thereof due to temporary high pressure conditions within the vessel interior is permitted to avoid structural failure and emission of dangerous materials from the pressure vessel. The pressure vessel apparatus of the present invention includes a vessel main body having a bottom and an outer peripheral wall extending upwardly from the bottom and defining a vessel main body top opening. A vessel top body having an outer peripheral wall is positioned on the top opening of the outer peripheral wall of the vessel main body to form a joint therebetween. The vessel top body defines a vessel top body bottom opening communicating with the vessel main body top opening, and the vessel top body defines a top opening. A vessel head is positioned on the vessel top body and covers the top opening. First securement means secures the vessel top body to the vessel main body. Second securement means secures the vessel head to the vessel top body. The second securement means comprises a plurality of double-ended, elongated tendons under tension extending between and secured to the vessel top body and the vessel head. The outer peripheral walls of the vessel main body and the vessel top body define a plurality of spaced throughbores extending vertically alongside and spaced from the pressure vessel interior. The first securement means includes a plurality of double-ended tendons under tension extending through the throughbores and secured to the vessel main body and the vessel top body. The tendons secured to the vessel main body and the vessel top body permit movement of the vessel top body away from the vessel main body due to a pressure surge or explosion caused by core melt or other accident within the pressure vessel interior before the tendons secured to the vessel top body and the vessel head permit movement of the vessel head away from the vessel top body due to the pressure surge within the pressure vessel interior. Other features, advantages, and objects of the present invention will become apparent with reference to the following description and accompanying drawings.
	description	The present invention relates to a method for producing an iodine radioisotopes fraction, in particular of I-131, comprising steps of: (i) Alkaline (or based) dissolution of enriched uranium targets by obtaining an alkaline (or based) slurry containing aluminium salts, uranium and isotopes generated by the fission of enriched uranium and a gaseous phase of Xe-133, (ii) Filtration of said alkaline (or based) slurry in order to isolate, on the one hand, a solid phase containing the uranium, and on the other hand, an alkaline (or based) solution of molybdate and salts of iodine radioisotopes, (iii) Adsorption of said salts of iodine radioisotopes on an alumina resin doped with silver and recovery of said alkaline (or based) solution of molybdate depleted of iodine radioisotopes, in particular of I-131, passing through said alumina resin doped with silver, and (iv) Recovery of said iodine radioisotopes fraction, in particular of I-131. Such a method is well-known and described in the document “Preparation and characterization of silver coated alumina for isolation of iodine-131 from fission products. Mushtaq et al.—Journal of Engineering and Manufacturing Technology, 2014”. According to this document, the highly enriched uranium targets are processed for the purpose of producing radioisotopes of molybdenum-99 and radioisotopes of iodine-131 by alkaline dissolution. As mentioned above, the alkaline slurry is then filtered and the alkaline liquid phase (filtrate) is loaded on an alumina resin doped with silver. A fraction containing the iodine radioisotopes, in particular of iodine-131, is recovered by elution of the alumina column doped with silver by sodium thiosulfate (Na2S2O3). According to this document, the recovered fraction containing the iodine radioisotopes, in particular iodine-131, is not sufficiently pure and must also be distilled for medical applications. Elution with sodium thiosulfate should lead to the recovery of about 90% of the iodine radioisotopes, in particular of iodine-131, loaded on the alumina column doped with silver. Unfortunately, this document is silent regarding the overall purification yield. Although it details the elution yields with respect to the total quantity of iodine loaded on the column, the document does not give any information about the iodine purification yield with respect to the basic alkaline resulting from the dissolution of the targets. Another method for producing an iodine radioisotope fraction, in particular of iodine-131 is described in the document “Reprocessing of irradiated Uranium 235 for the production of Mo-99, I-131, Xe-133 radioisotopes. J. Salacz—revue IRE tijdschrift, vol 9, No 3 (1985)”. According to this document, processing the products of the fission of uranium for the purpose of producing short-lived radioisotopes involves highly restrictive working conditions. These particularly restrictive working conditions involve having to work in shielded cells using robotic arms, or working outside shielded cells using the handling devices of the production chain to operate robotic arms. Once the methods for processing the targets containing highly enriched uranium are well established and secured to ensure very low or no environmental pollution, the production of radioisotopes method is clearly fixed. The smallest change to these methods is, if possible, to be avoided in order not to disrupt the production scheme, because when the environmental pollution level is considered to be secured, each change is considered as a new risk to manage in order to achieve a new satisfactory design of environmental constraints. Furthermore, the method is conducted in cells that are fitted with portholes of led-shielded glass several tens of centimetres thick, through which articulated arms, robotic or not, are operated from the outside. Several cells follow each other. In each cell, a part of the method is carried out. A first cell is dedicated to dissolve the targets of highly enriched uranium. Once the liquid phase containing the soluble products of uranium fission is recovered through filtration, including the radioisotope of Mo-99, it is transferred to the second cell where it is acidified to enable, during the exothermic acidification step, a gaseous release of iodine. The solution from which the iodine is released, is heated and bubble-stirred to release iodine in a gaseous form. The gas containing the iodine radioisotopes is then captured using a platinised asbestos trap. Iodine radioisotopes, in particular of I-131, are then desorbed from the platinised asbestos trap and sent to the cell where they undergo chemical purification by distillation. The iodine radioisotope yields, in particular of I-131, described in this document are of about 80 to 90%. 10 to 20% of the iodine radioisotopes, in particular of I-131, remain in the acidified liquid phase and contaminate the other radioisotopes. Thus, according to this document, the selectivity of the iodine isolation for the production thereof is not optimal. Furthermore, during the exothermic acidification, although the temperature of the acidified liquid phase increases, it is also necessary to provide further heating and bubble-stirring to try to recover a maximum of iodine radioisotopes, in particular of I-131. This heating causes the evaporation of the nitrates resulting from the acidification with nitric acid, thereby contaminating the iodine radioisotopes, in particular of I-131 in a gaseous form, which is problematic as it interferes with the marking process of subsequent biological molecules. There is therefore a need to provide a method enabling to produce iodine with a better yield, by reducing environmental hazards and by securing and reducing potential releases of iodine in the ventilation system, but also where the production selectivity is improved to increase the purity of the iodine radioisotopes fractions, in particular of I-131. The purpose of the invention is to overcome the disadvantages of the state of the art by providing a method enabling to improve the purity of the produced iodine by acting on the selectivity of the production operations while reducing environmental hazards. To overcome this issue, a method is provided according to the invention and as described at the beginning wherein said recovery of said iodine radioisotopes fraction, in particular of I-131, comprises a washing of the alumina resin doped with silver with a solution of NaOH at a concentration comprised between 0.01 and 0.1 mol/l, preferably between 0.03 and 0.07 mol/l and more preferably of about 0.05 mol/l, and an elution of the iodine radioisotopes, in particular of I-131 by a thiourea solution presenting a thiourea concentration comprised between 0.5 mol/l and 1.5 mol/l, preferably comprised between 0.8 and 1.2 mol/l, more preferably of about 1 mol/l, with the collection of an eluate containing said iodine radioisotopes, in particular I-131, in a thiourea solution. By performing this fixation step on a column of alumina doped with silver, about 90% of the iodine radioisotopes contained in the alkaline (or based) solution of molybdate and of the salts of iodine radioisotopes are fixed on the alumina resin doped with silver. According to the present invention, the alumina column is manufactured according to the disclosures of document “Preparation and characterization of silver coated alumina for isolation of iodine-131 from fission products. Mushtaq et al.—Journal of Engineering and Manufacturing Technology, 2014”, with the exception that the silver is reduced with hydrazine instead of sodium sulphate. The impregnation rate of the alumina resin by silver is of at least 4, preferably of at least 5, more preferably of about 5.5% by weight of silver with respect to the total weight of non-doped alumina. By performing an elution with thiourea according to the present invention, it was revealed, surprisingly, that the rate of iodine radioisotopes, in particular of eluted iodine-131, with respect to the total content of iodine radioisotopes, in particular of iodine-131 loaded on the alumina column, was greater than 90%, and even greater than 95% in activity. In addition, the elution using thiourea is quicker and carries out a narrower elution peak, thereby increasing the selectivity of the purification of iodine radioisotopes, in particular of iodine-131, while also reducing to a minimum the presence of other radioisotopes in the eluate of the alumina column doped with silver. In addition, according to the present invention, the volume of the washing solution is configured to be optimised and sufficiently delayed with respect to the passage of molybdenum through the column, for example with the presence of Mo-99 radioisotopes that would otherwise contaminate the eluate of iodine radioisotopes, in particular of I-131, but not too much to prevent the loss of iodine radioisotopes, in particular of iodine-131. Consequently, in the method according to the present invention, the selectivity of the iodine recovery, in particular of iodine-131, is improved along with the environmental safety, by the adsorption of iodine radioisotopes, in particular of iodine-131, on an alumina resin doped with silver, rather than imperatively having to pass the total quantity of iodine radioisotopes, in particular of iodine-131, of the alkaline solution of molybdate and the salts of iodine radioisotopes in a gaseous phase, to recover the totality of iodine radioisotopes, in particular of iodine-131, via a gas trap. In an advantageous embodiment, said uranium targets are low enriched uranium targets. Although the method according to the present invention applies to all types of targets, in particular to highly enriched uranium targets, but also to low enriched targets, the embodiment based on low enriched-enriched uranium targets is preferred. Indeed, the production of radioisotopes for medical applications has long relied on highly enriched uranium. Highly enriched uranium (HEU) is challenging in terms of worldwide safety considerations as it is relatively vulnerable to terrorist organisations and because of the potential thereof in the development of a nuclear weapon. Although the many facilities producing radioisotopes for medical applications feature sturdy security measures, minimising the use of highly enriched uranium in civilian applications is a significant act contributing to reducing the danger of proliferation. Despite improved efficiency in the production of radioisotopes from HEU, both in financial terms and in environmental terms, the conversion of the method for producing radioisotopes from HEU is significantly restricted by the USA, which remains the main source of uranium as a crude material. The United States has just taken all the necessary measures to promote the use of LEU by implementing compensatory measures accompanying the use of radioisotopes produced from low enriched uranium (LEU), by introducing limits to the acquisition and delivery of HEU, or by introducing penalties on the use of Mo-99 produced from HEU. In this context, there is therefore a need to develop a method for producing fractions containing a radioisotope of I-131 that make it possible to achieve a satisfactory compromise in terms of the economic efficiency of the production method, while reducing the use of highly enriched uranium. Unfortunately, given that the quantity of radioisotopes is directly linked to the quantity of uranium-235, and for the purpose of guaranteeing the same procurement level of pure I-131 medical isotopes, low enriched uranium-based targets contain overall much more uranium that highly enriched uranium targets, and therefore contain much more unusable matter (up to 5 times more). It is therefore advantageous, according to the present invention, to implement the method to process low enriched uranium targets, despite the presence of contaminants that are very different from those produced by highly enriched uranium targets, thereby increasing environmental safety, while maintaining/improving purity by acting on the selectivity with respect to iodine radioisotopes, in particular I-131, and by maintaining the qualitative criteria of iodine radioisotopes fractions, in particular of I-131. Advantageously, according to the present invention, the method further comprises, before said filtration an addition of alkaline-earth nitrate, more particularly of strontium, calcium, barium, preferably of barium and sodium carbonate to said alkaline slurry. Indeed, according to the present invention, it was possible to create a method that can be used industrially, by optimising the selectivity of the production of iodine radioisotopes, in particular of iodine-131, with acceptable yields and with improved environmental safety, and wherein also, despite the presence of 5 times more unusable matter, the production of radioisotopes of Mo-99 enables to achieve the required purity for a medical application and also improves environmental safety (both for the environment and for the operators). It has been demonstrated in the method according to the present invention that the alkaline dissolution of the targets, generating a slurry with a much higher concentration of solid unusable matter, but also of contaminants of the liquid portion of the slurry, could be efficiently filtered by the addition of alkaline-earth nitrate, more particularly of strontium, calcium, barium, preferably of barium and sodium carbonate. Indeed, when the alkaline-earth nitrate, more particularly of strontium, calcium, barium, preferably of barium, is added to the slurry, along with sodium carbonate, insoluble carbonates are formed, such as for example of barium, but also of strontium and other carbonates that serve as a filtrating medium during filtration, thereby preventing the clogging of the pores of the fibreglass filter. This has made it possible to achieve a significant reduction of the filtration time. According to the present invention, the filtration time of the slurry was reduced by 4 to 6 hours to a reduced time comprised between 30 minutes and 2 hours, based on the amount of targets involved in the dissolution. This is already significantly higher than with a method using highly enriched uranium-based targets (where the filtration time is typically from 10 to 20 minutes), but this method represents a possibility of industrial implementations which, otherwise, would not have existed without excessively increasing the production cost of radioisotopes produced by the fission of uranium 235. With low enriched uranium-based targets, the solid phase content of the slurry is 5 times higher. In addition, typically, these targets are based on an aluminium and uranium alloy, in particular in the form of UAl2, although other forms of the alloy are also present (such as UAl3, UAl4, etc.). Low enriched uranium-based targets contain less than 20% by weight of uranium 235 with respect to the total weight of uranium present in the target. Highly enriched uranium-based targets contain more than 90% by weight of uranium 235 with respect to the total weight of uranium present in the target. Consequently, the enriched uranium content is proportionally and significantly reduced (by a factor of about 5). Furthermore, by working with an alloy, among others with UAl2, it is possible to increase the uranium density present in the target, which clearly improves the production yield, but also creates other impurities, such as magnesium, which affect the method for producing radioisotopes of Mo-99 for medical applications. Indeed, the increase of the uranium density of the uraniferous nucleus has imposed the replacement of pure A5 aluminium with a harder alloy. Indeed, with this increased density, in the case of pure A5 being used, the integrity of the targets (and the absence of deformation thereof) during the production thereof would not be guaranteed. It is therefore not the use of UAl2 that brings Mg as impurity, but the fact that the uranium UAl2 alloy is denser and the fact that the total quantity of uranium has increased, which requires the use of an aluminium alloy, containing the Mg, for the production of the targets. Consequently, in the method according to the present invention, although the content in highly radioactive waste has increased, it has been possible not only to filter the slurry in an industrially-usable time, but also to eliminate the impurities brought by the use of a uranium and aluminium alloy in the slurry. In particular, in the method according to the present invention, the contamination of the Mo-99 radioisotope fraction by the Sr-90 radioisotope is reduced as it precipitates with the carbonate brought to the slurry. This is of the utmost importance as the radiotoxicity of the Sr-90 radioisotope is very high because of the combination of the extended physical half-life thereof (radioactive half-life: 28.8 years), the high-energy beta decay thereof and the long biological half-life thereof (bone tropism). It is therefore very important to reduce this impurity to minimise the potential long-term side effects for the patient. In addition, although it is relatively essential, the filtration adjuvant used in the method according to the present invention does not affect the fixation of the iodine on the silver-coated alumina column, on the contrary, given the already-reduced presence of contaminants in the source, the present invention reveals that it is possible to produce, in a profitable and efficient manner, on the one hand, a Mo-99 radioisotope from low enriched uranium, without the radioisotope fraction being ultimately less pure, thereby satisfying the criteria of the European Pharmacopoeia, despite the massive presence of a much greater quantity of waste and contaminants that are difficult to eliminate, such as magnesium, but also wherein, on the other hand, the risk of the presence of strontium in the Mo-99 radioisotope fraction is largely reduced, but in which about 90% of the iodine present in the alkaline slurry is collected on the alumina column doped with silver after the filtration. In a first advantageous embodiment of the method according to the present invention, the method further comprises an acidification of said eluate containing said iodine radioisotopes, in particular I-131 in a thiourea solution by the addition of a buffer solution, in particular a solution of phosphoric acid with a concentration comprised between 0.5 and 2 mol/l, preferably between 0.8 and 1.5 mol/l, and more preferably of about 1 mol/l, with a recovery of an acidified solution of iodine radioisotope salts, in particular of I-131. According to the present invention, the iodine radioisotopes, in particular iodine-131 are acidified for the purpose to be pre-purified and separated from most of the contaminants, including the thiourea, used beforehand to recover the iodine from the silver-coated alumina. In the scope of the present invention, the term “effluent of the resin” is used to describe the mobile phase that passes through the resin and leaves the chromatography column. In a preferred embodiment of the present invention, the method further comprises a purification of said acidified solution of iodine radioisotope salts, in particular of I-131, said purification comprising a loading of said acidified solution of iodine radioisotope salts, in particular of I-131 on an ion-exchange column, a washing of said ion-exchange resin with water, an elution of said ion-exchange resin with NaOH at a concentration between 0.5 and 2.5 mol/l, preferably between 0.8 mol/l and 1.5 mol/l and particularly preferably of about 1 mol/1, with a recovery of said iodine radioisotopes fraction, in particular of I-131, in a solution of NaOH. Advantageously, said ion-exchange resin is a weak anion resin. In another embodiment of the method according to the invention, the method also comprises an acidification of the alkaline solution of molybdate depleted of iodine radioisotopes, in particular of I-131 passing through said alumina resin doped with silver, with formation of an acid solution of molybdenum salts and release of residual iodine radioisotopes, in particular of I-131, in the form of gas for the purpose of the recovery thereof. In this variant of the method according to the present invention, as mentioned above, the quantity of iodine radioisotopes, in particular of iodine-131, recovered by adsorption on the alumina column doped with silver is of about 90% by activity with respect to the total activity of iodine radioisotopes, in particular of iodine-131. The residual 10% of iodine radioisotopes, in particular of iodine-131, are still present in the alkaline molybdate solution previously passed through said alumina column doped with silver. Consequently, recovering in a separate step the residual iodine is advantageous for two reasons. Firstly, the iodine thus recovered can be enhanced in the form of an iodine radioisotopes fraction, in particular of iodine-131, and secondly because the presence of residual iodine in the alkaline molybdate solution generates the environmental hazard of having these iodine radioisotopes, in particular iodine-131, being released in the ventilation system, which is also connected to the chimney. Consequently, isolating the iodine at this stage represents a profitability potential in the scope of the method according to the present invention, but also reduces the environmental risk associated with the iodine in the method according to the present invention. Preferably, in another advantageous embodiment of the method according to the present invention, the method further comprises, before said acidification of the alkaline molybdate solution depleted of iodine radioisotopes, in particular of I-131 passing through said alumina resin doped with silver, a cooling of the alkaline molybdate solution depleted of iodine radioisotopes, in particular of I-131 passing through said alumina resin doped with silver, to a temperature below or equal to 60° C., preferably below or equal to 55° C., more particularly below or equal to 50° C. In this manner, it was surprisingly observed that the purity and yield of the produced iodine radioisotopes fractions, in particular of I-131, were improved. According to the present invention, it was highlighted that to solve this problem relating to the difficulty in controlling the massive release of iodine at high temperatures, simply cooling the aqueous alkaline phase resulting from the filtration before acidification to a temperature below or equal to 60° C., preferably below or equal to 55° C., more particularly below or equal to 50° C., favours the solubility of the iodine in the acid solution of molybdenum salts. In this manner, owing to the fact that the solubility of the gases decreases with the increase of the temperature, the cooling of the aqueous alkaline phase resulting from the filtration enables a slower volatilisation of the iodine, and therefore prevents the sudden release thereof when the acid is added. Indeed, when iodine is brought to the iodine trap suddenly, the capture of the iodine is negatively impacted, while the cooling enabling a controlled release improves the yield of capture by the trap. During acidification, the temperature of the acid solution of molybdenum salts increases progressively and makes it possible for an equally progressive release of the iodine towards the trap, which favours the capture thereof, unlike the massive release of the iodine. Consequently, according to the present invention, it is possible to improve the production yield of iodine radioisotopes, in particular of I-131, from aluminium targets containing highly enriched uranium very simply, by cooling the filtrate to prevent the massive release of iodine in the iodine trap during the acidification to a temperature of about 50° C., and in any case below 60° C. The filtrate is therefore acidified by concentrated nitric acid. The iodine radioisotopes are then released during the acidification in far greater quantities. In a specific embodiment of the present invention, the method further comprises, after acidification, heating of the acid solution of molybdenum salts to a temperature greater than 93° C., preferably greater than or equal to 95° C., preferably between 96° C. and 99° C., but preferably below 100° C., accompanied by air bubbling to optimise the release of iodine in a gaseous form, at a precisely determined moment, during and after acidification. Advantageously, in the method according to the present invention, said recovery of the iodine radioisotopes, in particular I-131 upon the release thereof is carried out by a transfer of the iodine radioisotopes, in particular I-131 in the form of gases in a pipe connected at one end to an acidifier wherein the acidification occurs and at the other end to a closed container containing an aqueous phase and a surrounding medium, said transfer of iodine radioisotopes, in particular I-131 in the form of a gas being carried out so as to result directly in the aqueous phase wherein the iodine radioisotopes, in particular I-131, in the form of gas pass through the aqueous phase and escape in the form of bubbles in the surrounding medium of the aqueous phase, contained in the closed container. In this manner, the nitrates that might be present in the form of aerosols, as well as other gaseous species soluble in water, such as nitrogen oxides, are solubilised and eliminated from the iodine radioisotopes, in particular from I-131, in the form of a gas. Also, in another embodiment of the present invention, said closed container is connected by a pipe to a second closed container that contains an NaOH trap and wherein the surrounding medium of the aqueous phase is transferred from the closed container to the second closed container containing the NaOH trap in the form of a solution at a concentration from 2 to 4, in particular of about 3 mol/l, with discharge of the surrounding medium containing the iodine radioisotopes, in particular I-131 of the pipe into the solution of the NaOH trap, with solubilisation of the iodine radioisotopes, in particular I-131 in the form of gas into iodide of iodine radioisotopes, in particular I-131 in the aqueous solution of the NaOH trap. The iodine radioisotopes, in particular I-131 are thus dissolved in the NaOH aqueous solution at an NaOH concentration from 2 to 4 mol/l, preferably of 3 mol/l, and form a crude iodine solution. In a preferred embodiment of the method according to the present invention, the aqueous solution of the NaOH trap containing the iodides of the iodine radioisotopes, in particular I-131, forms a crude iodine solution, which is then purified by a second acidification to form gaseous iodine. The crude solution is transferred to an iodine purification cell. The crude solution is then acidified by H2SO4+H2O2 to again produce the gaseous iodine, which is captured in NaOH 0.2 M bubblers. This solution is called the “stock solution”, and it is then packaged in sealed vials, depending on the orders. Alternately, the iodine radioisotopes fraction, in particular of I-131 in an NaOH solution containing the iodides of the iodine radioisotopes, in particular of I-131, forms a crude iodine solution and is then purified by a second acidification, preferably carried out in the presence of H2SO4 and H2O2 to again produce the gaseous iodine. Then, preferably, the gaseous iodine is captured in NaOH 0.2 M bubblers to form said fraction containing a radioisotope of iodine-131. In an advantageous embodiment, said iodine radioisotopes fraction, in particular of I-131 in an NaOH solution and the aqueous solution of the NaOH trap containing the iodides of iodine radioisotopes, in particular I-131, are collected and purified together by a second acidification. Other embodiments of the method according to the invention are indicated in the appended claims. The invention also relates to an iodine radioisotopes fraction, in particular of I-131 conditioned in a solution of NaOH having a radiochemical purity in iodine radioisotopes, in particular of I-131 greater than 97%, preferably of at least 98%, more particularly of at least 98.5% of the activity present in the chemical iodide form of said radioisotope of the I-131 with respect to the total activity of said radioisotope of I-131 in all the forms thereof in said fraction. More specifically, said solution of iodine radioisotopes, in particular of I-131, is conditioned in sealed vials, said sealed vials being enclosed in individual shielded containers. Advantageously, the iodine radioisotopes fraction, in particular of I-131, presents a nitrate content of below 30 g/l. In an advantageous version, the iodine radioisotopes fraction, in particular of I-131, is obtained by the method according to the present invention. Other embodiments of the fraction according to the invention are indicated in the appended claims. Other characteristics, details and advantages of the invention will become apparent from the description given hereafter, with reference to the examples and not limited thereto. When the uranium 235 is bombarded with neutrons, it forms fission products with a smaller mass and which are themselves unstable. These products generate, through a decay chain, other radioisotopes. In particular, it is by this process that the Mo-99, Xe-133 and I-131 radioisotopes are produced. The low enriched uranium-based targets contain an aluminium alloy containing uranium. The content of enriched uranium with respect to the total weight of uranium is at most of 20%, and typically of around 19%. The low enriched uranium targets are dissolved during the alkaline dissolution phase in the presence of NaOH (at about 4 mol/l or more) and of NaNO3 (at about 3.5 mol/l). During the dissolution, a slurry is formed along with a gaseous phase of Xe-133. The slurry contains a solid phase mainly formed from uranium and hydroxides of fission products and a liquid phase of molybdate (MoO4−) and of iodine-131 in the form of iodine salts. The volume of the alkaline dissolution phase increases with the amount of targets, given the very high content of unusable products after dissolution of the targets. The dissolution of the aluminium of the target is an exothermic reaction. The gaseous phase of xenon is recovered by capture using a xenon trap. When the xenon is eliminated, a solution of alkaline-earth nitrate, more particularly of strontium, calcium, barium, preferably of barium, is then added to the slurry at a concentration of between 0.05 mol/l and 0.2 mol/l and in a quantity of 2 to 6 litres, depending on the number of targets. Sodium carbonate is also added at a concentration comprised between 1 mol/l and 1.5 mol/l, preferably of about 1.2 mol/l, and in a quantity of 100 to 300 ml, depending on the number of dissolved targets. The slurry is then diluted with water in a volume of 2 to 6 litres, depending on the number of targets, to make it possible for the transfer thereof to the subsequent step. The slurry containing the liquid phase and the basic phase is then filtered through a fibreglass filter with a porosity comprised between 2 and 4 μm, preferably of about 3 μm. The solid phase is washed twice with a volume of water of 900 ml, recovered and possibly sent upstream from the method for a subsequent alkaline dissolution. The filtrate (recovered alkaline liquid phase containing the Mo-99, I-131, I-133, I-135, Cs-137, Ru-103, Sb-125 and Sb-127 fission products) is recovered, along with the aluminate formed by the alkaline dissolution of the aluminium targets, which is soluble in a basic pH. Aluminium is soluble both in an acid and in an alkaline medium. However, it is insoluble when the pH ranges from 5 to 10. At this stage, the filtrate is loaded on an alumina column doped with silver in order to fix the iodine and recover an alkaline filtrate depleted of iodine-131. The alumina column doped with silver is washed with a volume of about 500 ml of caustic soda at a concentration of about 0.05 mol/l. The impregnation rate of the alumina resin contained in the alumina column is about 5.5% by weight. The iodine is fixed selectively by reaction with the silver doping present at the surface of the alumina to form an insoluble silver iodide. The alumina column doped with silver is preferably positioned in between two reactors. The reactor downstream from the alumina column doped with silver is placed under a controlled vacuum, which enables the transfer of the liquid onto the column at a flow rate of about 250 ml/min. The yields of the iodine capture are of about 95%. The alumina column doped with silver is then eluted with a thiourea solution with a concentration comprised between 0.5 mol/l and 1.5 mol/l, preferably of about 1 mol/l. The eluate contains iodine coming from the column. The eluate is then brought to an acid pH by adding a buffer mixture, in particular of phosphoric acid, in order to obtain an acid solution of iodine salts. The acid solution of iodine salts is then loaded on an ion-exchange column, in particular on a weak anion resin column pre-processed in a non-oxidising acid medium, in particular with phosphoric acid. In this manner, in terms of safety, in this advantageous embodiment of the method according to the present invention, the activity of the iodine fixed on the ion-exchanging resin is transferred from one cell to the next in a solid form. The ion-exchange column on which the iodine is fixed is then eluted with NaOH at a concentration of between 0.5 mol/l and 2.5 mol/l, preferably of about 1. The iodine radioisotopes are thus transformed into iodide and solubilised in the NaOH. The fraction containing the iodine radioisotopes undergoes a first purification step using the second acidification. The collected filtrate must then be acidified. However, the acidification also causes the release of heat. Consequently, prior to acidification, the filtrate is cooled to a temperature of about 50° C. Indeed, as disclosed in the document “Form and Stability of Aluminium Hydroxide Complexes in Dilute Solutions” (J. D. Hem and C. E. Roberson—Chemistry of Aluminum in Natural Water—1967), the behaviour of aluminium in a solution is complex and the transformation reactions of the Al3+ ion into the precipitated hydroxide form and the aluminate soluble form are subject to a certain amount of kinetics. The formation of metastable solids is known and the conditions of equilibrium are sometimes difficult to achieve, even with long reaction times. Aluminium oxides and hydroxides form different crystalline structures (bayerite, gibbsite, etc.) that are chemically similar but differ in terms of solubility. The experimental conditions of temperature, concentration and speed of addition of the reagents significantly affect the results. The reaction that governs the equilibrium between the various forms of aluminium is as follows during acidification: As the medium is highly radioactive and at a high temperature because of the alkaline dissolution, but also because of the exothermic character of neutralisation during the acidification step, the addition of acid would form, in localised sites, acid overconcentration that would lead to local heating by the neutralisation reaction, and to the formation of insoluble aluminium forms or with slow aluminium salts re-dissolution kinetics. However, given the restrictions of the method described in the state of the art, given that the reaction environment has a high temperature, given that it is highly radioactive and difficult to access, it is not possible to maintain the stirring to avoid these local sites of aluminate concentration at high temperature. The effects of acid overconcentration must be avoided for two main reasons. On the one hand, the formation of aluminium salt precipitates significantly risks clogging the installation, which reduces the production yield, and on the other hand it also creates a health risk considering the high radioactivity of the reaction mixture. Indeed, it is not simple, and maybe not even possible, to intervene manually to unclog the installation, but furthermore, this could only be done to the detriment of the production yield. Consequently, the filtrate is cooled so as to avoid the precipitation of the aluminium salts during the acidification at a temperature of about 50° C., and in any case of below 60° C. The filtrate is therefore acidified with concentrated nitric acid. The acidified filtrate is heated to a temperature greater than 93° C., preferably greater than or equal to 95° C., preferably between 96° C. and 99° C., but preferably of less than 100° C., and maintained in a bubbling state. In a first embodiment of the present invention, the acidification makes it possible to carry out a solution with an acid pH in order to fix the Mo-99 radioisotope on the alumina column (in the presence of an excess of acid of about 1 M). The acidified liquid phase, depleted of iodine, is then loaded onto an alumina column, which is conditioned in nitric acid at a concentration of 1 mol/l. The Mo-99 is adsorbed on the alumina while most of the contaminant fission products are eliminated in the effluent of the alumina column. The alumina column on which the Mo-99 radioisotope is fixed, is washed with nitric acid at a concentration of 1 mol/l, with water, with sodium sulphite at a concentration of about 10 g/l and finally once again with water. The washing effluent is discarded. The alumina column is then eluted with NaOH at a concentration of about 2 mol/l and then with water. The eluate recovered from the alumina column forms the first eluate of the Mo-99 radioisotope in the form of molybdate. In a preferred embodiment of the method according to the present invention, the first eluate of the column is kept for a period of between 20 and 48 hours. After this predetermined period, the alumina column is once again eluted with NaOH at a concentration of about 2 mol/l and then with water, prior to the washing thereof. The eluate recovered from the new elution forms the second eluate of the Mo-99 radioisotope in the form of molybdate. At this stage, the first eluate of the Mo-99 radioisotope can be collected with the second eluate of the Mo-99 radioisotope and forms a single eluate which will undergo further purification steps. Alternately, each first and second eluate is treated individually in subsequent purification steps, in the same manner. For more simplicity, below, the eluate of the Mo-99 radioisotope will be referred to, to describe the first eluate of the Mo-99 radioisotope or the second eluate of the Mo-99 radioisotope, or both together. The eluate of the Mo-99 radioisotope of the alumina column is then loaded onto a second chromatography column containing a high anion ion-exchange resin pre-processed in water. The ion-exchange column is then eluted with nitrate using a solution of ammonium nitrate at a concentration of about 1 mol/l. The recovered eluate therefore comprises the Mo-99 radioisotope in a fraction containing ammonium nitrate. The solution of ammonium nitrate containing the radioisotope of Mo-99 is then loaded on an activated carbon column with a 35-50 mesh, which can also be doped with silver to recover any trace amounts of iodine. The activated carbon column on which the Mo-99 radioisotope is fixed is then washed with water and eluted with a solution of NaOH at a concentration of about 0.3 mol/l. The elution of the activated carbon column makes it possible for the recovery of a solution of Na299MoO4 in NaOH and to keep any iodine possibly captured on the column at a preferred concentration of 0.2 mol/l, which will then be packaged and prepared for delivery. In a particular embodiment of the invention, the solution of Na299MoO4 in NaOH at a preferred concentration of 0.2 mol/l is loaded onto an alumina resin in a Mo-99/Tc-99 generator or on a resin of titanium oxide to make it possible for the generation of a technetium-99 radioisotope for nuclear medicine. In a second advantageous embodiment of the method according to the present invention, the acidification enables to achieve a solution with an acid pH to fix the Mo-99 radioisotope on the titanium oxide column (in the presence of an excess of acid 1 M). The acidified liquid phase, depleted of iodine, is then loaded onto a titanium oxide column, processed in nitric acid at a concentration of 1 mol/l. The Mo-99 is adsorbed on the titanium oxide, while most of the contaminant fission products are eliminated in the effluent of the titanium oxide column. The titanium oxide column on which the Mo-99 radioisotope is fixed is washed with nitric acid at a concentration of 1 mol/l, with water, with sodium sulphite at a concentration of about 10 g/l and finally once again with water. The washing effluent is discarded. The titanium oxide column is then eluted with NaOH at a concentration of about 2 mol/l and then with water. The eluate recovered from the titanium oxide column forms the first eluate of the Mo-99 radioisotope in the form of molybdate, and comprises about 90% or more of the Mo-99 initially present. In a preferred embodiment of the method according to the present invention, the first eluate of the column is kept for a period of between 20 and 48 hours. After this predetermined period, the elution of the titanium oxide column is continued with NaOH at a concentration of about 2 mol/l and forms an elution tail containing the Mo-99 radioisotope, in the form of molybdate. At this stage, the first eluate of molybdate and/or said molybdate eluate tail are collected or not and acidified with a solution of sulphuric acid at a concentration comprised between 1 and 2 mol/l, preferably of 1.5 mol/l, thereby forming an acidified fraction of the pure Mo-99 radioisotope, in the form of molybdenum salts. For more simplicity, below, the eluate of the Mo-99 radioisotope will be referred to, in the form of molybdate to describe the first eluate of the Mo-99 radioisotope or the tail of the molybdate eluate, or both together. The eluate of the Mo-99 radioisotope of the titanium oxide column is then loaded onto a second chromatography column containing a weak anion ion-exchange resin pre-processed in water. The ion-exchange column is then eluted with nitrate using a solution of ammonium nitrate at a concentration of about 1 mol/l. The recovered eluate therefore comprises the Mo-99 radioisotope in a fraction containing ammonium nitrate. The solution of ammonium nitrate containing the radioisotope of Mo-99 is then loaded on an activated carbon column with a 35-50 mesh, which can also be doped with silver to recover any trace amounts of iodine. The activated carbon column on which the Mo-99 radioisotope is fixed is then washed with water and eluted with a solution of NaOH at a concentration of about 0.3 mol/l. The elution of the activated carbon column makes it possible for the recovery of a solution of Na299MoO4 in NaOH and to keep any iodine possibly captured on the column at a preferred concentration of 0.2 mol/l, which will then be packaged and prepared for delivery. In a specific embodiment of the invention, the solution of Na299MoO4 in NaOH at a preferred concentration of 0.2 mol/l is loaded onto an alumina resin in a Mo-99/Tc-99 generator or on a resin of titanium oxide to make it possible for the generation of a technetium-99 radioisotope for nuclear medicine. During the formation of the slurry, the uranium fission products are released, some in a soluble form, others in a gaseous form. This is, among others, the case of xenon and krypton, which are therefore in a gaseous phase. The gaseous phase escapes from the liquid medium and remains contained in the sealed container wherein the dissolution occurs. The sealed container comprises a gaseous phase output connected to a device for the recovery of noble gases, isolated from the outside environment, as well as an input for a flushing gas. The gaseous phase contains ammonia (NH3) that comes from the reduction of the nitrates and from the main gaseous fission products, which are Xe-133 and Kr-85. The dissolution is a highly exothermic reaction, imposing two large refrigerants. However, water vapour is present in the gaseous phase. The gaseous phase is transported by a carrier gas (He) towards the device for recovering noble gases. In a first variant, the recovery of xenon is carried out as follows: The gaseous phase leaves the sealed container of alkaline dissolution and is brought towards the device for the recovery of noble gases. The gaseous phase containing, among others, the radioisotope Xe-133 is first passed through a molecular sieve to eliminate the ammonia (NH3) and the water vapour. Then, the gaseous phase is passed through silica gel to eliminate all trace amounts of residual water vapour. The gaseous phase is then brought to the cryogenic trap. In a second advantageous variant according to the present invention, the gaseous phase is adsorbed on zeolite, in particular on a titanosilicate or on an aluminosilicate doped with silver, preferably on Ag-ETS-10 or on Ag-chabazite. It is then marketed directly on the zeolite, or desorbed in heated conditions and sent towards a cryogenic trap. The gaseous phase containing, among others, the radioisotope Xe-133 is therefore brought to the cryogenic trap in a U-shaped tube immersed in liquid nitrogen (i.e. at −196° C.) contained in a shielded container, through stainless steel shavings. The stainless steel 316 shavings and manufactured from a stainless steel 316 rod, with a diameter of between 1.5 and 2 cm and with a length of between 10 and 20 cm, preferably between 14 and 18 cm, and more particularly of about 16 cm, using a four-flute end milling cutter with a diameter of 16 mm and a hydraulic vice. The speed of the milling machine comprising the abovementioned milling cutter is of 90 rpm and the travel speed thereof is of 20 mm/min. The cutting depth of the milling cutter is of about 5 mm. The stainless steel shavings have an average weight comprised between 20 and 30 mg/shaving, preferably between 22 and 28 mg/shaving, and a non-packed bulk density comprised between 1.05 and 1.4. The stainless steel shavings have an average length of 7 mm, an average diameter of about 2.5 mm and a thickness of about 1.7 mm. The U-shaped tube comprises a quantity of comprised between 90 g and 110 g. The volume of stainless steel 316 shavings comprised in the U-shape tube is totally immersed in liquid nitrogen. The radioisotope Xe-133 from said gaseous phase containing the radioisotope Xe-133 is then captured by liquefaction of said Xe-133 by said cooled stainless steel shavings that capture the Xe-133 by condensation. The liquefaction temperature of the Xe-133 is of about −107° C. Consequently, the gaseous Xe is condensed to a liquid form on the stainless steel shavings. However, as the liquefaction temperature of the Kr-85 is of about −152° C., there is a significantly smaller quantity of Kr trapped in the liquid nitrogen trap, and the residual Kr is collected in specific traps with the gases resulting from the method described herein, namely the gaseous phase substantially depleted of Xe-133, among others. Once the Xe-133 has been captured in the liquid nitrogen trap, the ducts are purged, the injection of liquid nitrogen is cut and the trap is brought into contact with a vacuum bulb, the volume of which is 50 times greater than the volume of shavings contained in the liquid nitrogen trap. The liquid nitrogen trap, in a closed circuit including the collection tube, is thus brought to ambient temperature. After warming, 99% of the Xe-133 initially present in a gaseous form is present in the bulb. In a variant of the method according to the present invention, the residual iodine radioisotopes, in particular of I-131, that were not captured by the alumina resin doped with silver prior to acidification, are then recovered during the acidification of the alkaline slurry, which makes it possible to obtain a solution with an acid pH that is able to fix the radioisotope of Mo-99 on the alumina column, the acidification also releasing iodine radioisotopes for the purpose of the recovery thereof. The recovery of the iodine can then be performed during and after the acidification of the pre-cooled alkaline filtrate. The iodine radioisotopes are released by heating of the acidified filtrate to a temperature greater than 93° C., preferably greater than or equal to 95° C., preferably between 96° C. and 99° C., but preferably below 100° C., and maintained in a bubbling state to increase the release of iodine in a gaseous form. When the acidified filtrate is heated, a gaseous phase is formed, containing the iodine radioisotopes along with an evaporated portion of the filtrate. The acidifier comprises a gaseous phase outlet pipe immersed in a closed container containing water. Another tube exits this closed container. The aqueous phase therefore leaves the acidifier and is left to bubble in the water contained in the closed container. In this manner, the portion of the filtrate that was evaporated is dissolved in the water contained in the closed container, while the insoluble portion, namely the iodine radioisotopes, remains above the water surface, in the closed container, and exits therefrom through the outlet pipe of the closed container and travels towards a second closed container, which is a trap containing NaOH at a concentration of 3 mol/l. The iodine radioisotopes are then transformed into iodide and solubilised in the NaOH contained in the iodine trap, where it forms a crude iodine solution. In a preferred embodiment of the method according to the present invention, the aqueous solution of the NaOH trap containing the iodides of the iodine radioisotopes, in particular of I-131, is then purified by a second acidification. The crude solution is transferred to an iodine purification cell. The crude solution is then acidified by H2SO4+H2O2 to produce again the gaseous iodine, which is captured in NaOH 0.2 M bubblers. This solution is called the “stock solution”, and it is then packaged in sealed vials contained in a shielded enclosure to be shipped to the customer. It is understood that the present invention is by no means limited to the embodiments described above and that many modifications may be made thereto without departing from the scope of the appended claims.
	summary
043604951	claims	1. The target arrangement for spallation-neutron sources wherein a horizontal proton beam continuously impinges on the target, the arrangement comprising: a wheel having an annular volume of target material arranged thereon having an outer periphery upon which the beam impinges, a jacket overlying the target material in spaced relation thereto, and a window of low mass number metal joining the jacket and overlaying the outer periphery of the target material in spaced relation with respect thereto; shaft means for mounting the wheel to rotate in one direction about a vertical axis perpendicular to the horizontal proton beam with the target material positioned in intersection with the proton beam, and means for cooling the target material with a liquid coolant from a coolant source, the cooling means comprising: a plurality of grooves on the upper and lower surfaces of the target material, the grooves in one of the surfaces following involute curves which advance away from the direction of rotation of the wheel as the curves progress toward the periphery of the wheel and the grooves in the other surface following involute curves which advance toward the direction of rotation of the wheel as the curves progress toward the periphery of the wheel, the grooves communicating with the space between the window and periphery of the target material whereby liquid coolant flows from the grooves which advance away from the direction of rotation to the space between the window and outer periphery and then through the grooves which advance toward the direction of rotation while being confined by the jacket, the cooling means further including inlet and outlet means concentric with the shaft and connected to the grooves and the source of liquid coolant said grooves being of substantial equal lengths to provide for uniform heat removal from the entire target material. 2. The target arrangement of claim 1 wherein the wheel has a diameter of approximately 2.5 meters. 3. The target arrangement of claim 1 wherein the window is made of one metal selected from the group consisting of Al, Zr or Ti. 4. The arrangement of claim 1 wherein the inlet and outlet means for the coolant are connected to the source of coolant at a location above the wheel. 5. The target arrangement of claim 1 wherein the grooves which advance away from the direction of rotation are in the upper surface of the target material while the grooves which advance toward the direction of rotation are in the lower surface of the target material.
045129495	abstract	Direct power shape monitoring in parallel with precision power monitoring is effected by use of gamma sensors in the fuel core and signal processing that includes compensation for slow signal response and takes advantage of a substantially direct relationship of sensor signal to linear power generation rate. Continuous readout from the direct power shape monitor is available during readout interruptions in the precision monitor.
	abstract	An in-line dissolved gas removal membrane-based apparatus for removing dissolved hydrogen and fission gases from the letdown stream from a reactor coolant system.
	description	The present invention relates to a method of controlling a turbine equipment and a turbine equipment, particularly relates to a method of controlling a turbine equipment and a turbine equipment preferably used in a closed cycle gas turbine for circulating a working fluid in a closed system by using an atomic reactor or the like for a heat source. In accelerating a speed of a generator equipment of a gas turbine or the like of a background art, a motor provided exclusive for starting and electric apparatus SFC (Static Frequency Converter) for using a generator as a motor are used. The apparatus are used only in starting, and therefore, in order to achieve a reduction in plant cost, the apparatus having capacities as small as possible have been used. Therefore, in a current state, a revolution number is increased up to 30% of a rated revolution number by using the apparatus exclusive for accelerating a speed, thereafter, fuel supply is started and the revolution number is increased by an acceleration torque of the turbine per se. However, according to the closed cycle gas turbine of circulating the working fluid in the closed system by using the atomic reactor or the like for the heat source, a temperature elevating rate is restricted by a restriction imposed on a reactor main body (for example, 100° C./h) and rapid temperature elevation is difficult. Therefore, according to a starting method using a starting procedure similar to that of the gas turbine of the background art, a problem that time is taken for elevating the speed of the turbine to the rated revolution number is posed. On the other hand, according to a method of elevating a speed up to a rated revolution number by using only a starting apparatus, a capacity required for the starting apparatus is increased, and therefore, a problem of increasing plant cost is posed. As a method of resolving the above-described problem, a technology of operating an amount of filling helium circulated in a closed system has been proposed (refer to, for example, the Publication of Japanese Patent No. 3020853). According to a technology described in the Publication of Japanese Patent No. 3020853, by reducing the amount of filling helium more than a filling amount in a rated operation in accelerating the speed of the turbine, a drive torque required for elevating the speed of the turbine can be reduced and a capacity required for the starting apparatus can be reduced. However, according to the technology described in the Publication of Japanese Patent No. 3020853, time is taken in operating the helium filling amount, and therefore, a problem that the technology is not pertinent for being used in an acceleration operation which is finished in a comparatively short time period. Further, according to a speed reduction gear or the like for transmitting a drive torque generated by a motor or the like to a turbo unit comprising a compressor and a turbine, a drawback of fretting or the like is prevented from being brought about by being loaded with a necessary minimum torque. However, when the drive torque required for accelerating the speed of the turbine is reduced as described above, also a torque loaded to the reduction gear or the like is also reduced to pose a problem that there is a concern of bringing about a drawback of fretting or the like by deviating load sharing of the gear from a standard value by a self weight of the gear, or shifting a contact position of teeth from a standard position. The invention has been carried out in order to resolve the above-described problem and it is an object thereof to provide a method of controlling a turbine equipment and a turbine equipment capable of carrying out a starting operation by controlling a load applied on a speed reducing portion while complying with a restriction imposed on an apparatus provided at a turbine equipment. In order to achieve the above-described object, the invention provides a following means. A method of controlling a turbine equipment of the invention is characterized by a method of controlling a turbine equipment including a compressing portion of compressing a working fluid, a turbine portion driven to rotate by the working fluid, and a circulating flow path of circulating the working fluid at least between the compressing portion and the turbine portion, the method including a speed accelerating step of increasing a revolution number by driving to rotate the compressing portion and the turbine portion by a motor by way of a speed reducing portion, a load detecting step of detecting a load applied to the speed reducing portion by a load detecting portion, and a bypass flow rate controlling step of increasing a flow rate of the working fluid bypassed from a delivery side to a suction side of the compressing portion when an absolute value of the detected load is equal to or smaller than an absolute value of a predetermined value and reducing the flow rate of the bypassed working fluid when the absolute of the load is equal to or larger than the absolute value of the predetermined value. According to the invention, in operating to accelerate a speed of the compressing portion and the turbine portion, by controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressor based on the load applied to the speed reducing portion, the load applied to the speed reducing portion is controlled to the predetermined value. For example, in comparison with the method of controlling the flow rate of the bypassed operating fluid based on a time period elapsed from starting to operate to accelerate the speed, a control of the load applied to the speed reducing portion becomes accurate. That is, when the absolute value of the load applied to the speed reducing portion is equal to or smaller than the absolute value of the predetermined value, by an increase in the flow rate of the bypassed working fluid, the flow rate of the working fluid passing the compressor is increased. When the flow rate of the passing working fluid is increased, a torque necessary for driving the compressor is increased, and therefore, a load applied to the speed reducing portion arranged between the motor and the compressor is increased and the load applied to the speed reducing portion is controlled to the predetermined value. On the other hand, when the absolute value of the load applied to the speed reducing portion is equal to or larger than the absolute value, by reducing the flow rate of the bypassed working fluid, the flow rate of the working fluid passing the compressor is reduced. When the flow rate of the passing working fluid is reduced, the torque necessary for driving the compressor is reduced, and therefore, the load applied to the speed reducing portion arranged between and the compressor is reduced, and the load applied to the speed reducing portion is controlled to the predetermined value. Further, the load applied to the speed reducing portion is controlled by controlling only the flow rate of the bypassed working fluid, and therefore, even in a case of a turbine equipment provided with a heat source of restricting a temperature accelerating rate of, for example, an atomic reactor or the like of heating the working fluid, the load applied to the speed reducing portion can be controlled to the predetermined value while complying with a restriction of the temperature elevating rate or the like imposed on the heat source. According to the invention, it is preferable that the bypass flow rate controlling step includes a first calculating step of calculating the bypassed flow rate based on the detected load and the predetermined value, a second calculating step of calculating a bypass flow rate necessary for preventing surging of the compressor from being brought about based on a pressure ratio between the suction side and the delivery side of the compressor, and a corrected revolution number of the compressor calculated based on a temperature of the working fluid sucked to the turbine portion, a selecting step of selecting the bypass flow rate having a larger flow rate from the bypass flow rates calculated by the first and the second calculating steps, and a flow rate controlling step of controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressing portion to the selected bypass flow rate. According to the invention, the bypass flow rate having the larger flow rate is selected from the bypass flow rate of controlling the load applied to the speed reducing portion to the predetermined value and the bypass flow rate for preventing surging of the compressor from being brought about to be controlled to the bypass flow rate of selecting the flow rate of the bypassed working fluid, and therefore, not only the load applied to the speed reducing portion is prevented from being smaller than the predetermined value but surging of the compressor is prevented from being brought about. A method of controlling a turbine equipment of the invention is characterized by a method of controlling a turbine equipment including a compressing portion of compressing a working fluid, a turbine portion driven to rotate by the working fluid, and a circulating flow path of circulating the working fluid at least between the compressing portion and the turbine portion, the method including a speed accelerating step of increasing a revolution number by driving to rotate the compressing portion and the turbine portion by a motor by way of a speed reducing portion, and a bypass flow rate controlling step of reducing a flow rate of the working fluid bypassed from the delivery side to the suction side of the compressing portion with an elapse of a time period after starting to increase a revolution number. According to the invention, in operating to accelerate speed of the compressing portion and the turbine portion, by controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressor based on the elapsed time period after starting to operate to accelerate speed, the load applied to the speed reducing portion is controlled to the predetermined value. For example, in comparison with the method of controlling the flow rate of the bypassed working fluid based on the load applied to the speed reducing portion, the control of the load applied to the speed reducing portion is facilitated. That is, in starting to operate to accelerate speed, the temperature of the working fluid circulating through the compressing portion and the turbine portion is low, the torque necessary for driving the compressing portion or the like is small, and therefore, also the load applied to the speed reducing portion is small. Therefore, in starting to operate to accelerate speed, the flow rate of the working fluid passing through the compressor is ensured without reducing the flow rate of the bypassed working fluid and the load applied to the speed reducing portion is ensured. Thereafter, when time has elapsed after starting to operate to accelerate the speed, the temperature of the working fluid circulating to the compressing portion and the turbine portion becomes high, the torque necessary for driving the compressing portion or the like is increased and the load applied to the speed reducing portion is increased. Therefore, an increase in the torque necessary for driving the compressor is restrained and an increase in the load applied to the speed reducing portion is restrained by reducing the flow rate of the bypassed working fluid with an elapse of time after starting to operate to accelerate speed. In the invention, it is preferable that the bypass flow rate control step includes a first calculating step of calculating the bypass flow rate based on the elapsed time period after starting to increase the revolution number, a second calculating step of calculating the bypass flow rate necessary for preventing surging of the compressor from being brought about based on a pressure ratio between the suction side and the delivery side of the compressor, and a corrected revolution number of the compressor calculated based on a temperature of the working fluid sucked to the compressor, a selecting step of selecting the bypass flow rate having a larger flow rate from the bypass flow rates calculated by the first and the second calculating steps, and a flow rate controlling step of controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressing portion to the selected bypass flow rate. According to the invention, the bypass flow rate having the larger flow rate is selected from the bypass flow rate based on the elapsed time period after starting to increase the revolution number and the bypass flow rate for preventing surging of the compressor from being brought about to control to the bypass flow rate selecting the flow rate of the bypassed working fluid, and therefore, not only the load applied to the speed reducing portion is prevented from being smaller than the predetermined value and surging is prevented from being brought about at the compressor. A turbine equipment of the invention is characterized in being provided with a compressing portion of compressing a working fluid, a turbine portion driven to rotate by the working fluid, a circulating flow path of circulating the working fluid at least between the compressing portion and the turbine portion, a bypass flow path of bypassing the working fluid from a delivery side to a suction side of the compressing portion, a flow rate control portion of controlling a flow rate of the working fluid flowing through the bypass flow path, a motor of driving to rotate the compressing portion and the turbine portion by way of a speed reducing portion in starting, and a control portion for executing the control method according to any one of Claim 1 through Claim 4. According to the invention, by carrying out the controlling method of the invention by the control portion, the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressor is controlled and the load applied to the speed reducing portion is controlled. Further, the load applied to the speed reducing portion is controlled by controlling only the flow rate of the bypassed working fluid, and therefore, even in a case of a turbine equipment provided with a heat source restricting a temperature elevating rate of, for example, an atomic reactor or the like of heating the working fluid, the load applied to the speed reducing portion can be controlled to the predetermined value while complying with a restriction imposed on the heat source of the temperature elevating rate or the like. According to the method of controlling the turbine equipment and the turbine equipment, in operating to accelerate speed of the compressing portion and the turbine portion, by controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressor based on the load applied to the speed reducing portion, there is achieved an effect of capable of carrying out a starting operation of controlling the load applied to the speed reducing portion while complying with the restriction imposed on the apparatus provided at the turbine equipment. According to the method of controlling the turbine equipment and the turbine equipment of the invention, by controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the compressor based on the elapsed time period after starting to operate the accelerating of the speed in operating to accelerate speed of the compressing portion and the turbine portion, there is achieved an effect capable of carrying out the starting operation restricting the load applied to the speed reducing portion while complying with the restriction imposed on the apparatus provided at the turbine equipment. A power generating equipment including a closed cycle gas turbine according to a first embodiment of the invention will be explained in reference to FIG. 1 through FIG. 5. FIG. 1 is a schematic view for explaining a constitution of a power generating equipment according to the embodiment. According to the embodiment, an explanation will be given by applying the invention to a power generating equipment including a gas turbine for circulating a helium gas constituting a working fluid in a closed circulating system (closed style) and using an atomic reactor as a heat source of heating a compressed working fluid. As shown by FIG. 1, a power generating equipment (turbine equipment) 1 is mainly provided with a turbine portion 3, a low pressure compressor (compressing portion) 4, a high pressure compressor (compressing portion) 5 and a speed reduction gear portion (speed reducing portion) 6 arranged on the same rotating shaft 2, a generator (motor) 7 connected to the speed reduction gear portion 6, an atomic reactor 8 for heating a working fluid compressed by the high pressure compressor 5, and a circulating flow path for circulating a working fluid in an order of the atomic reactor 8, the turbine portion 3, the low pressure compressor 4 and the high pressure compressor 5. As shown by FIG. 1, the turbine portion 3 is arranged at the rotating shaft 2 and driven to rotate by the high temperature high pressure working fluid supplied from the reactor 8. The working fluid is connected to be able to flow between the atomic reactor 8 and the turbine portion 3 and between the turbine portion 3 and the low pressure compressor 4 by the circulating flow path 9. As shown by FIG. 1, the low pressure compressor 4 is arranged at the rotating shaft 2 for compressing the working fluid by using a rotational drive force supplied by way of the rotating shaft 2. The working fluid is connected to be able to flow by the circulating flow path 9 between the turbine portion 3 and the low pressure compressor 4 and between the low pressure compressor 4 and the high pressure compressor 5. As shown by FIG. 1, the high pressure compressor 5 is arranged at the rotating shaft 2 for compressing the working fluid to a higher pressure by using a rotational drive force supplied by way of the rotating shaft 2. The working fluid is connected to be able to flow by the circulating flow path 9 between the low pressure compressor 4 and the high pressure compressor 5 and between the high pressure compressor 5 and the atomic reactor 8. As shown by FIG. 1, the atomic reactor 8 is arranged between the high pressure compressor 5 and the turbine portion 3 for supplying the high temperature high pressure working fluid to the turbine portion 3 by heating the high pressure working fluid delivered from the high pressure compressor 5. The working fluid is connected to be able to flow by the circulating flow path 9 between the high pressure compressor 5 and the atomic reactor 8 and between the atomic reactor 8 and the turbine portion 3. As shown by FIG. 1, the speed reduction gear portion 6 connects the rotating shaft 2 and the generator 7 to be able to transmit the rotational drive force for transmitting the rotational drive force from the rotating shaft 2 to the generator 7, or from the generator 7 to the rotating shaft 2 while converting a revolution number. The speed reduction gear portion 6 is constituted by a combination of a plurality of gears and various combination styles can be used. For example, although a planetary gear can be used as the speed reduction gear portion 6, the invention is not particularly limited thereto. FIG. 2 is a block diagram for explaining a control of the power generating equipment of FIG. 1. As shown by FIG. 1, the speed reduction gear portion 6 is provided with a torque meter (load detecting portion) 11 for measuring a torque applied to the speed reduction gear portion 6. As shown by FIG. 2, the torque measured by the torque meter 11 is outputted to a feedback control portion 51. As shown by FIG. 1, the generator 7 is connected to the speed reduction gear portion 6 to be able to transmit the rotational drive force and is driven to rotate by the turbine portion 3 by way of the rotating shaft 2 and the speed reduction gear portion 6 to generate a power when the power generating equipment 1 is brought into an operating state. On the other hand, in starting the power generating equipment 1, the turbine portion 3, the low pressure compressor 4 and the high pressure compressor 5 are driven to rotate by way of the rotating shaft 2 and the speed reduction gear portion 6 by using a power supplied from outside. As shown by FIG. 1, the circulating flow path 9 is a flow path of circulating the working fluid among the reactor 8, the turbine portion 3, the low pressure compressor 4 and the high pressure compressor 5. The circulating flow path 9 is provided with a regenerating heat exchanger 21 for carrying out heat exchange between the working fluid flowing out from the turbine portion 3 and the working fluid delivered from the high pressure compressor 5, a cooler 22 for carrying out heat exchange between the working fluid sucked to the low pressure compressor 4 and sea water, an intermediate cooler 23 for carrying out heat exchange between the working fluid delivered from the low pressure compressor 4 and seat water. As shown by FIG. 1, the regenerating heat exchanger 21 is a heat exchanger for heating the working fluid flowing to the atomic reactor 8 by recovering heat from the working fluid flowing out from the turbine portion 3. The regenerating heat exchanger 21 is arranged between the turbine portion 3 and the low pressure compressor 4 and between the high pressure compressor 5 and the atomic reactor 8. As shown by FIG. 1, the cooler 22 is a heat exchanger for radiating heat of the working fluid flowing out from the regenerating heat exchanger 21 to sea water. The cooler 22 is arranged between the regenerating heat exchanger 21 and the low pressure compressor 4. Further, the cooler 22 of a constitution of radiating heat of the working fluid to seat water as described above may be used, or a heat exchanger of a constitution of radiating heat of the working fluid to other medium may be used, and the invention is not particularly limited. As shown by FIG. 1, the intermediate cooler 23 is a heat exchanger of radiating heat of the working fluid delivered from the low pressure compressor 4 to sea water. The intermediate cooler 23 is arranged between the low pressure compressor 4 and the high pressure compressor 5. Further, the intermediate cooler 23 of a constitution of radiating heat of the working fluid to sea water as described above may be used, the heat exchanger of a constitution of radiating heat of the working fluid to other medium may be used, the invention is not particularly limited. Further, as shown by FIG. 1, the circulating flow path 9 is provided with a first bypass flow path (bypass flow path) 31 for increasing a flow rate of the working fluid sucked to the low pressure compressor 4, and a second bypass flow path (bypass flow path) 32 for controlling an amount of filling the working fluid, that is, a flow rate of the working fluid flowing at inside of the circulating flow path 9 and increasing a flow rate of the working fluid sucked to the low pressure compressor 4 and the high pressure compressor 5. As shown by FIG. 1, the first bypass flow path 31 is a flow path of recirculating a portion of the working fluid flowing out from the intermediate cooler 23 to between the low pressure compressor 4 and the cooler 22. In other words, the first bypass flow path 31 is a flow path one end of which is connected to the circulating flow paths 9 between the intermediate cooler 23 and the high pressure compressor 5 and other end of which is connected to the circulating flow path 9 between the cooler 22 and the low pressure compressor 4. The first bypass flow path 31 is provided with a first bypass valve 36 for controlling a flow rate of the recirculating working fluid. As shown by FIG. 1, the first bypass valve 36 is a valve arranged at the first bypass flow path 31 for controlling a flow rate of the working fluid flowing in the first bypass flow path 31. In other words, the first bypass valve 36 is a valve of controlling the flow rate of the working fluid sucked to the low pressure compressor 4 for preventing surging at the low pressure compressor 4 from being brought about by controlling the flow rate. Although according to the embodiment, an explanation will be given by being applied to an example of arranging the two first bypass valves 36 in parallel, a number of the first bypass valves 36 may be larger or smaller than two and is not particularly limited. As shown by FIG. 1, the second bypass flow path 32 is a flow path of filling the working fluid to one or both of a delivery side of the high pressure compressor 5 and a suction side of the low pressure compressor 4 and a flow path of recirculating a portion of the working fluid delivered from the high pressure compressor 5 to between the regenerating heat exchanger 21 and the cooler 22. In other words, the second bypass flow path 32 is a flow path one end portion of which is connected to between the high pressure compressor 5 and the regenerating heat exchanger 21 and other end portion of which is connected between the cooler 22 and the low pressure compressor 4. The second bypass flow path 32 is provided with a first buffer tank 41 and a second buffer tank 42 connected to outside working fluid filling system and a second bypass valve (flow rate controlling portion) 43 arranged between the first buffer tank 41 and the second buffer tank 42. The first buffer tank 41 is a tank arranged on a side of the high pressure compressor 5 of the second bypass flow path 32. The second buffer tank 42 is a tank arranged on a side of the cooler 22 of the second bypass flow path 32. When the working fluid is filled to the circulating flow path 9, the working fluid is filled from the working fluid filling system by way of one or both of the first buffer tank 41 and the second buffer tank 42. On the other hand, when the flow rate of the working fluid sucked to the low pressure compressor 4 and the high pressure compressor 5 is controlled, a portion of the working fluid delivered from the high pressure compressor 5 is made to flow in an order of the first buffer tank 41 and the second buffer tank 42 to be recirculated to between the regenerating heat exchanger 21 and the cooler 22. As shown by FIG. 1, the second bypass valve 43 is a valve arranged at the second bypass flow path 32 between the first buffer tank 41 and the second buffer tank 42 for controlling the flow rate of the working fluid flowing in the second bypass flow path 32. In other words, the second bypass valve 43 is a valve for controlling the flow rate of the working fluid sucked to the low pressure compressor 4 and the high pressure compressor 5 in operating the power generating equipment 1 and is a valve for controlling a torque applied to the speed reduction gear portion 6 in starting. As shown by FIG. 2, the second bypass valve 43 is inputted with a control signal for controlling a valve opening degree from the feedback control portion 51. Although according to the embodiment, an explanation will be given by being applied to an example of arranging two of the second bypass valves 43 in parallel, a number of the second bypass valves 43 may be larger or smaller than two and is not particularly limited. Further, as shown by FIG. 2, the power generating equipment 1 is provided with the feedback control portion (control portion) 51 for controlling the opening degree of the second bypass valve 43 based on a torque signal constituting a value of a torque measured by the torque meter 11. An explanation will be given of a control of an opening degree of the second bypass valve 43 of the feedback control portion 51. Next, power generation at the power generating equipment 1 comprising the above-described constitution will be explained. When an operation, that is, power generation is carried out at the power generating equipment 1, as shown by FIG. 1, the high pressure working fluid flowing to the atomic reactor 8 is further heated by absorbing heat generated at the atomic reactor 8 to flow out from the atomic reactor 8 to the circulating flow path 9 as the working fluid of, for example, about 900° C. The working fluid flows to the turbine portion 3 from the circulating flow path 9. The turbine portion 3 generates the rotational drive force from an energy provided to the flowing high temperature high pressure working fluid to transmit the generated rotational drive force to the rotating shaft 2. The rotational drive force is transmitted from the rotating shaft 2 to the speed reduction gear portion 6 and is transmitted from the speed reduction gear portion 6 to the generator 7. A revolution number of the rotating shaft 2 is reduced to a revolution number pertinent for driving the generator 7 by the speed reduction gear portion 6. The generator 7 generates power by being driven to rotate by the transmitted rotational drive force. On the other hand, a temperature of the working fluid is lowered to about 500° C. when flowing out from the turbine portion 3 and the working fluid flows to the regenerating heat exchanger 21 by way of the circulating flow path 9. At the regenerating heat exchanger 21, heat exchange is carried out between the working fluid flowing from the turbine portion 3 and the working fluid delivered from the high pressure compressor 5 mentioned later and the working fluid flows out from the regenerating heat exchanger 21. The working fluid flowing out from the regenerating heat exchanger 21 flows to the cooler 22 by way of the circulating flow path 9 to carry out heat exchange between the working fluid and sea water, cooled to about 20° C., thereafter, flows out from the cooler 22. The working fluid flowing out from the cooler 22 is sucked to the low pressure compressor 4 by way of the circulating flow path 9. The low pressure compressor 4 compresses the sucked working fluid by the rotational drive force supplied from the turbine portion 3 by way of the rotating shaft 2 to deliver to the circulating flow path 9. The working fluid delivered from the low pressure compressor 4 flows to the intermediate cooler 23 by way of the circulating flow path 9 to carry out heat exchange between the working fluid and sea water, cooled to about 20° C., thereafter, flows out from the intermediate cooler 23. The working fluid flowing out from the intermediate cooler 23 is sucked to the high pressure compressor 5 by way of the circulating flow path 9. The high pressure compressor 5 compresses the working fluid compressed by the low pressure compressor 4 to a higher pressure by the rotational drive force supplied from the turbine portion 3 by way of the rotating shaft 2 to deliver to the circulating flow path 9. The working fluid delivered from the high pressure compressor 5 flows to the regenerating heat exchanger 21 by way of the circulating flow path 9, carries out heat exchange between the working fluid and the working fluid flowing out from the turbine portion 3, heated to, for example, about 450° C. and flows out to the circulating flow path 9. The working fluid flowing out from the regenerating heat exchanger 21 flows to the atomic reactor 8 by way of the circulating flow path 9 to repeat the above-described procedure. When a flow rate of the working fluid flowing in the circulating flow path 9 is small, in other words, when a flow rate of the working fluid flowing to the low pressure compressor 4 is small, the first bypass valve 36 is opened to prevent surging at the low pressure compressor 4 from being brought about. That is, by opening the first bypass valve 36, a portion of the working fluid delivered from the low pressure compressor 4 and flowing out from the intermediate cooler 23 flows to the circulating flow path 9 between the cooler 22 and the low pressure compressor 4 by way of the first bypass flow path 31. Therefore, the flow rate of the working fluid flowing to the low pressure compressor 4 is increased in comparison with a flow rate of circulating the working fluid in a total of the circulating flow path 9 to prevent surging at the low pressure compressor 4 from being brought about. On the other hand, when the flow rate of the working fluid flowing to the low pressure compressor 4 and the high pressure compressor 5 is small, the second bypass valve 43 is opened to prevent surging at the low pressure compressor 4 and the high pressure compressor 5 from being brought about. That is, by opening the second bypass valve 43, a portion of the working fluid delivered from the high pressure compressor 5 flows to the circulating flow path 9 between the regenerating heat exchanger 21 and the cooler 22 by way of the second bypass flow path 32, the first buffer tank 41 and the second buffer tank 42. Therefore, the flow rate of the working fluid flowing to the low pressure compressor 4 and the high pressure compressor 5 is increased in comparison with the flow rate of circulating the working fluid in the total of the circulating flow path 9 to prevent surging at the low pressure compressor 4 and the high pressure compressor 5 from being brought about. Further, when an amount of filling the working fluid circulating in the circulating flow path 9 is small, the working fluid is filled to inside of the circulating flow path 9 from the working fluid filling system connected by way of the first buffer tank 41 and the second buffer tank 42. Next, a control in starting the power generating equipment 1 constituting a characteristic of the embodiment will be explained. In starting the power generating equipment 1, as shown by FIG. 1, a power is supplied from outside to the generator 7. The generator 7 supplied with the power generates the rotational drive force as a motor to drive to rotate the turbine portion 3, the low pressure compressor 4 and the high pressure compressor 5 by way of the speed reduction gear portion 6 and the rotating shaft 2. FIG. 3 is a graph for explaining a change over time of the revolution number in starting the power generating equipment of FIG. 1 and a change over time of an opening degree of the second bypass valve. When the power generating equipment 1 is started, as shown by FIG. 3, the low pressure compressor 4, the high pressure compressor 5 and the like are driven to rotate by a rotational speed of about 300 rotations per minute and the rotational speed of about 300 rotations per minute is maintained until first predetermined time T1 at which a speed accelerating instruction is inputted. At this occasion, as shown by FIG. 1 and FIG. 2, the torque applied to the speed reduction gear portion 6 is measured by the torque meter 11, and a value of the measured torque, that is, the torque signal is inputted to the feedback control portion 51. The value of the torque measured by the torque meter 11 is more proximate to 0 than a target torque (predetermined value), and therefore, the feedback control portion 51 outputs a control signal of opening the second bypass valve 43. When the second bypass valve 43 is opened, a portion of the working fluid delivered from the high pressure compressor 5 is recirculated to the low pressure compressor 4 by way of the second bypass flow path 32, the first buffer tank 41 and the second buffer tank 42. In other words, the flow rate of the working fluid compressed by the low pressure compressor 4 and the high pressure compressor 5 is increased and the drive torque necessary for driving the low pressure compressor 4 and the high pressure compressor 5 is increased. Then, also the torque applied to the generator 7 constituting the motor, and the speed reduction gear portion 6 arranged between the low pressure compressor 4 and the high pressure compressor 5 is increased. FIG. 4 is a graph for explaining a change over time of the torque applied to the speed reduction gear portion in starting the power generating equipment of FIG. 1. From starting to the first predetermined time T1, the rotational speed of the low pressure compressor 4, the high pressure compressor 5 or the like is low, further, also the temperature of the atomic reactor 8 is brought into a low state, and therefore, the control signal of opening the second bypass valve 43 is outputted from the feedback control portion 51 until the second bypass valve 43 is fully opened. The torque from starting to the first predetermined time T1 shown in FIG. 4 indicates the torque applied to the speed reduction gear portion 6 in a state of fully opening the second bypass valve 43. That is, from starting to the first predetermined time T1, the torque measured by the torque meter 11 is included in a range from 0 to a negative first predetermined torque (predetermined value) −Q1. Here, a positive torque in FIG. 4 indicates a value of a torque applied to the speed reduction gear portion 6 when the generator 7 is driven to rotate by the turbine portion 3, and a negative torque indicates a value of a torque applied to the speed reduction gear portion 6 when the turbine portion 3, the low pressure compressor 4 and the high pressure compressor 5 are driven to rotate by the generator 7. Further, a region between the positive first predetermined torque Q1 to the negative first predetermined torque −Q1 in FIG. 4 is a region in which a value of the torque applied to the speed reduction gear portion 6 is small and there is a high possibility of bringing about fretting in a gear or the like constituting the speed reduction gear portion 6. FIG. 5 is a flowchart for explaining a control in starting the power generating equipment of FIG. 1. When the speed accelerating instruction is inputted and an increase in the rotational speed of the low pressure compressor 4 or the like is started, that is, when the first predetermined time T1 has elapsed, as shown by FIG. 3, the rotational speed of the low pressure compressor 4, the high pressure compressor 5 or the like is increased with an elapse of time and is accelerated to about 6000 rotations per minute of a rated rotational number at the second predetermined time T2 (step S1 (speed accelerating step)). When the rotational speed is increased, also the flow rate of the working fluid compressed by the low pressure compressor 4 and the high pressure compressor 5 is increased, and therefore, also the drive torque necessary for driving the low pressure compressor 4 and the high pressure compressor 5 is increased. Therefore, also the torque applied to the speed reduction gear portion 6 measured by the torque meter 11 is increased (step S2 (load detecting step)). The feedback control portion 51 outputs the control signal of controlling the opening degree of the second bypass valve 43 in accordance with an inputted torque signal (step S3 (bypass flow rate controlling step). That is, when an absolute value of the torque applied to the speed reduction gear portion 6 is smaller than an absolute value of the target torque −Q, the feedback control portion 51 outputs the control signal of opening the second bypass valve 43 to carry out the control of making the torque applied to the speed reduction gear portion 6 proximate to the target torque −Q. On the other hand, when the absolute value of the torque applied to the speed reduction gear portion 6 is larger than the absolute value of the target torque −Q, the feedback control portion 51 outputs the control signal of closing the second bypass valve 43 to carry out a control of making the torque applied to the speed reduction gear portion 6 proximate to the target torque −Q. According to the above-described constitution, by controlling the flow rate of the working fluid bypassing from the delivery side of the high pressure compressor 5 to the suction side of the of the low pressure compressor 4 based on the torque applied to the speed reduction gear portion 6 in operating to accelerate the speed of the low pressure compressor 4, the high pressure compressor 5 and the turbine portion 3, the torque applied to the speed reduction gear portion 6 is controlled to the target torque −Q. Thereby, the control of the torque applied to the speed reduction gear portion 6 becomes accurate in comparison with, for example, a method of controlling the flow rate of the bypassing working fluid based on a time period elapsed from starting to operate to accelerate the speed. That is, when the absolute value of the torque applied to the speed reduction gear portion 6 is equal to or smaller than the absolute value of the target torque −Q, the flow rate of the working fluid passing through the low pressure compressor 4 and the high pressure compressor 5 is increased by increasing the flow rate of the bypassing working fluid. When the flow rate of the passing working fluid is increased, the torque necessary for driving the low pressure compressor 4 and the high pressure compressor 5 is increased, and therefore, the torque applied to the speed reduction gear portion 6 arranged between the generator 7 and the low pressure compressor 4 and the high pressure compressor 5 is increased and the torque applied to the speed reduction gear portion 6 is controlled to the target torque −Q. On the other hand, when the absolute value of the torque applied to the speed reduction gear portion 6 is equal to or larger than the absolute value of the target torque −Q, the flow rate of the working fluid passing through the low pressure compressor 4 and the high pressure compressor 5 is reduced by reducing the flow rate of the bypassing working fluid. When the flow rate of the passing working fluid is reduced, the torque necessary for driving the low pressure compressor 4 and the high pressure compressor 5 is reduced, and therefore, the torque applied to the speed reduction gear portion 6 arranged between the generator 7 and the low pressure compressor 4 and the high pressure compressor 5 is reduced and the torque applied to the speed reduction gear portion 6 is controlled to the target torque −Q. Further, the torque applied to the speed reduction gear portion 6 is controlled by controlling only the flow rate of the bypassing working fluid, and therefore, even in a case of the power generating equipment 1 provided with the heat source the speed accelerating rate of which is restricted of, for example, the atomic reactor 8 or the like of heating the working fluid, the torque applied to the speed reduction gear portion 6 can be controlled to the target torque −Q while complying with a restriction of the speed accelerating rate or the like. Further, although according to the above-described embodiment, an explanation has been given by being applied to the example of measuring the load applied to the speed reduction gear portion 6 by using the torque meter 11 as the torque applied to the speed reduction gear portion 6, not only the load is measured as the torque but a displacement of a gear constituting the speed reduction gear portion 6 may be measured as the load and the invention is not particularly limited thereto. Further, although according to the above-described embodiment, an explanation has been given by being applied to the example in which the compressor is constituted by two stages, that is, the low pressure compressor 4 and the high pressure compressor 5, the compressor may be constituted by one stage or may be 3 stages or more and the invention is not particularly limited thereto. Although according to the above-described embodiment, an explanation has been given by being applied to the example of controlling the opening degree of the second bypass valve 43 by the feedback control portion 51, an opening degree of the first bypass valve 36 may be controlled by the feedback control portion 51, and the invention is not particularly limited thereto. Although according to the above-described embodiment, an explanation has been given by being applied to the example of controlling the second bypass valve 43 in starting the power generating equipment 1, a similar control may be carried out when the power generating equipment 1 is stopped, or the flow rate of the working fluid flowing in the circulating flow path 9 is small, and the invention is not particularly limited thereto. Next, a second embodiment of the invention will be explained in reference to FIG. 6 through FIG. 8. Although a basic constitution of a power generating equipment of the embodiment is similar to that of the first embodiment, the embodiment differs from the first embodiment in a method of controlling the first bypass valve. Therefore, according to the embodiment, only control of the first bypass valve will be explained in reference to FIG. 6 through FIG. 8, and an explanation of other constituent elements or the like will be omitted. FIG. 6 is a block diagram for explaining the control of the power generating equipment according to the embodiment. Further, constituent elements the same as those of the first embodiment are attached with the same notations and an explanation thereof will be omitted. As shown by FIG. 6, a control portion 150 in a power generating equipment 101 of the embodiment is provided with the feedback control portion 51 for controlling the opening degree of the second bypass valve 43, and a program control portion (control portion) 151 for controlling the opening degree of the first bypass valve (flow rate control portion) 36. The program control portion 151 controls the opening degree of the first bypass valve 36 based on the speed accelerating instruction for instructing to accelerate the rotational speed of the low pressure compressor 4 and the like in starting the power generating equipment 1. A control of the opening degree of the first bypass valve 36 at the program control portion 151 will be explained as follows. Next, a control in starting a power generating equipment 101 constituting a characteristic of the embodiment will be explained. Further, although in starting, the feedback control portion 51 and the program control portion 151 respectively control the opening degrees of the second bypass valve 43 and the first bypass valve 36, the control of the opening degree of the second bypass valve 43 by the feedback control portion 51 is similar to the control in the first embodiment, and therefore, an explanation thereof will be omitted. Further, power generation at the power generating equipment 101 is similar to the power generation according to the first embodiment, and therefore, an explanation thereof will be omitted. FIG. 7 is a graph for explaining a change over time of the revolution number and a change over time of the first and the second bypass valve opening degrees in starting the power generating equipment of FIG. 6. Further, a graph V1 in FIG. 7 indicates the opening degree of the first bypass valve 36 and a graph V2 indicates the opening degree of the second bypass valve 43. When the power generating equipment 101 is started, as shown by FIG. 7, the low pressure compressor 4, the high pressure compressor 5 and the like are driven to rotate at a rotational speed of about 300 rotations per minute and the rotational speed of about 300 rotations per minute is maintained until the first predetermined time T1 at which the speed accelerating instruction is inputted. At this occasion, the program control portion 151 outputs a control signal of fully opening the first bypass valve 36. When the first bypass valve 36 is opened, a portion of a refrigerant delivered from the low pressure compressor 4 and passing the intermediate cooler 23 is recirculated to the suction side of the low pressure compressor 4 by way of the first bypass flow path 31. In other words, the flow rate of the working fluid compressed by the low pressure compressor 4 is increased and the drive torque necessary for driving the low pressure compressor 4 is increased (refer to FIG. 1). Then, also the torque applied to the speed reduction gear portion 6 arranged between the generator 7 constituting a motor and the low pressure compressor 4 is increased. FIG. 8 is a flowchart for explaining the control in starting the power generating equipment of FIG. 6. Thereafter, when the rotational speed of the low pressure compressor 4 or the like is started by inputting the speed accelerating instruction (step S1 (speed accelerating step)), the program control portion 151 outputs a control signal of reducing the opening degree of the first bypass valve 36 in accordance with an elapsed time period after inputting the speed accelerating instruction (step S13 (bypass flow rate control step)). That is, when the rotational speed of the low pressure compressor 4 or the like is increased with an elapse of time, also the flow rate of the working fluid compressed by the low pressure compressor 4 is increased, and therefore, also the drive torque necessary for driving the low pressure compressor 4 is increased with an elapse of time. Therefore, the program control portion 151 outputs the control signal of reducing the opening degree of the first bypass valve 36 in proportion to the elapse of time after inputting the speed accelerating instruction. When the opening degree of the first bypass valve 36 is reduced, the flow rate of the working fluid flowing through the first bypass valve 36 is reduced and the flow rate of the working fluid recirculated to the low pressure compressor 4 is reduced. Therefore, an increase in the flow rate of the working fluid compressed by the low pressure compressor 4 is restrained and also an increase in the drive torque necessary for driving the low pressure compressor 4 is restrained. According to the above-described constitution, in operating to accelerate the speed of the low pressure compressor 4, the high pressure compressor 5, and the turbine portion 3, by controlling the flow rate of the working fluid bypassed from the delivery side to the suction side of the low pressure compressor 4 based on the elapsed time period after starting to operate to accelerate the speed, the torque applied to the speed reduction gear portion 6 can be controlled to the target torque −Q. Thereby, the control of the load applied to the speed reduction gear portion 6 is facilitated in comparison with, for example, a method of controlling the flow rate of the bypassed working fluid based on the torque applied to the speed reduction gear portion 6. That is, in starting to operate to accelerate speed, a temperature of the working fluid circulated to the low pressure compressor 4, the high pressure compressor 5 and the turbine portion 3 is low, the torque necessary for driving the low pressure compressor 4 or the like is small, and therefore, also the load applied to the speed reduction gear portion 6 is small. Therefore, in starting to operate to accelerate speed, the torque applied to the speed reduction gear portion 6 can be ensured by ensuring the flow rate of the working fluid passing through the low pressure compressor 4 without reducing the flow rate of the bypassed working fluid. Thereafter, when time has elapsed after starting to operate to accelerate speed, the temperature of the working fluid circulated to the low pressure compressor 4, the high pressure compressor 5 and the turbine portion 3 becomes high, the torque necessary for driving the low pressure compressor 4 or the like is increased and the torque applied to the speed reduction gear portion 6 is increased. Therefore, by reducing the flow rate of the bypassed working fluid with an elapse of time after starting to operate to accelerate speed, an increase in the torque necessary for driving the low pressure compressor 4 can be restrained and an increase in the torque applied to the speed reduction gear portion 6 can be restrained. Further, by respectively controlling the opening degrees of the second bypass valve 43 and the first bypass valve 36 by using the feedback control portion 51 and the program control portion 151, the control of the torque applied to the speed reduction gear portion 6 is facilitated. That is, a total capacity of the first and the second bypass valves 36, 43 can be used for controlling the torque applied to the speed reduction gear portion 6, and therefore, it is not necessary to constitute either one of the first bypass valve and the second bypass valve 43 by large-sized formation or increase a number of pieces thereof. Therefore, a change or an addition of operation end used for controlling the torque applied to the speed reduction gear portion 6 is dispensed with and an increase in initial cost of the power generating equipment 101 can be restrained. Next, a third embodiment of the invention will be explained in reference to FIG. 9 through FIG. 11. Although a basic constitution of a power generating equipment of the embodiment is similar to that of the first embodiment, the embodiment differs from the first embodiment in a method of controlling the second bypass valve. Therefore, according to the embodiment, only the method of controlling the second bypass valve will be explained in reference to FIG. 9 through FIG. 11 and an explanation of other constituent element or the like will be omitted. FIG. 9 is a schematic view for explaining a constitution of a power generating equipment according to the embodiment. FIG. 10 is a block diagram for explaining a control of the power generating equipment of FIG. 9. Further, constituent elements the same as those of the first embodiment are attached with the same notations and an explanation thereof will be omitted. As shown by FIG. 9 and FIG. 10, a power generating equipment 201 of the embodiment is further provided with a second pressure ratio measuring portion 261 for measuring a pressure ratio of a working fluid pressure on a suction side and a working fluid pressure on a delivery side of the high pressure compressor 5, a temperature measuring portion 262 for measuring a temperature of the working fluid flowing to the turbine portion 3, and a corrected revolution number calculating portion 263 for calculating a corrected revolution number. As shown by FIG. 9, the second pressure ratio measuring portion 261 is a measuring portion for measuring a ratio of a pressure of the working fluid sucked to the high pressure compressor 5 and the pressure of the working fluid delivered from the high pressure compressor 5. The pressure ratio measured by the second pressure ratio measuring portion 261 is inputted to a second surging control portion 251 of a control portion 250, as shown by FIG. 10. As shown by FIG. 9, the temperature measuring portion 262 is a measuring portion of measuring a temperature of the working fluid flowing to the turbine portion 3. As shown by FIG. 10, a temperature measured by the temperature measuring portion 262 is inputted to the corrected revolution number calculating portion 263. As shown by FIG. 10, the corrected revolution number calculating portion 263 calculates a corrected revolution number N1 by the following equation based on the temperature Ti inputted from the temperature measuring portion 262 and an actual revolution number N of the turbine portion 3.N1=N/√{square root over ( )}(Ti) The corrected revolution number N1 calculated by the corrected revolution number calculating portion 263 is inputted to the second surging control portion 251. Further, as shown by FIG. 10, the control portion 250 of the power generating equipment 201 of the embodiment is provided with the feedback control portion 51 for controlling an opening degree V2 of the second bypass valve 43 based on the torque measured by the torque meter 11, the second surging control portion (control portion) 251 for controlling the opening degree V2 of the second bypass valve 43 by calculating a suction flow rate for preventing surging of the high pressure compressor 5 from being brought about, and a second selecting portion 252 for selecting a control signal having a large opening degree V2 of the second bypass valve 43 from control signals outputted from the feed back control portion 51 and the second surging control portion 251. As shown by FIG. 10, the second surging control portion 251 calculates the suction flow rate for preventing surging of the high pressure compressor 5 from being brought about based on the pressure ratio inputted from the second pressure ratio measuring portion 261 and the corrected revolution number N1 inputted from the corrected revolution number calculating portion 263, calculates the opening degree V2 of the second bypass valve 43 based on the calculated suction flow rate and outputs the control signal of controlling the opening degree V2. The control signal outputted from the second surging control portion 251 is inputted to the second selecting portion 252. Here, the calculated suction flow rate for preventing surging from being brought about is a flow rate of adding a predetermined allowance (margin) to the suction flow rate of bringing about surging in the high pressure compressor 5. Therefore, the calculated opening degree V2 of the second bypass valve 43 is an opening degree for making the working fluid of the flow rate added with the above-described margin flow to the high pressure compressor 5. As shown by FIG. 10, the second selecting portion 252 selects the control signal having the larger opening degree from the opening degrees V2 of the second bypass valve 43 applied to the control signal inputted from the feedback control portion 51 and the control signal inputted from the second surging control portion 251. The selected control signal is outputted from the second selecting portion 252 to the second bypass valve 43. Next, a control in starting the power generating equipment 201 constituting a characteristic of the embodiment will be explained. Further, although in starting, the feedback control portion 51 and the second surging control portion 251 respectively output the control signals by calculating the opening degree of the second bypass valve 43, calculation of the opening of the second bypass valve 43 by the feedback control portion 51 or the like is similar to that in the case of the first embodiment, and therefore, an explanation thereof will be omitted. Further, power generation at the power generating equipment 201 is similar to power generation according to the first embodiment, and therefore, an explanation thereof will be omitted. FIG. 11 is a flowchart for explaining a control in starting the power generating equipment of FIG. 10. In starting the power generating equipment 201, as shown by FIG. 11, calculation of the bypass flow rate in the feedback control portion 51, and calculation of the opening degree V2 of the second bypass valve 43 (step S21 (first calculating step)), and calculation of the bypass flow rate at the second surging control portion 251, and calculation of the opening degree V2 of the second bypass valve 43 (step S22 (second calculating step)) are executed independently from each other. As shown by FIG. 10, the second surging control portion 251 of the control portion 250 is inputted with the pressure ratio of the working fluid on the suction side and the delivery side of the high pressure compressor 5 is inputted from the second pressure ratio measuring portion 261, and the corrected revolution number N1 is inputted from the corrected revolution number calculating portion 263. The second surging portion 251 calculates the suction flow rate for preventing surging of the high pressure compressor 5 from being brought about based on the inputted pressure ratio and the inputted corrected revolution number N1. The flow rate for preventing surging from being brought about in the second surging control portion 251 is calculated based on a table or the like stored previously to the second surging control portion 251. The second surging control portion 251 further calculates the opening degree V2 of the second bypass valve 43 based on the calculated suction flow rate and outputs the control signal of controlling the opening degree of the second bypass valve 43 to the second selecting portion 252. As shown by FIG. 10, the second selecting portion 252 is inputted with the control signal of controlling the opening degree V2 of the second bypass valve 43 from the second surging control portion 251 and inputted with a control signal of controlling the opening degree V2 of the second bypass valve 43 also from the feedback control portion 51. The second selecting portion 252 selects the control signal having the larger opening degree of the second bypass valve 43 from the inputted control signals and outputs the selected control signal to the second bypass valve 43 (step S23 (selecting step)). At the second bypass valve 43, the opening degree V2 is controlled based on the inputted control signal and the flow rate of the working fluid flowing to the second bypass valve 32 is controlled (step S24 (flow rate controlling step)). According to the above-described constitution, by selecting the bypass flow rate having a larger flow rate from the bypass flow rate of controlling the torque applied to the speed reduction gear portion 6 to the target torque −Q and the bypass flow rate for preventing surging from being brought about in the high pressure compressor 5, the flow rate of the bypassed working fluid is controlled to the selected bypass flow rate, and therefore, not only the torque applied to the speed reduction gear portion 6 can be prevented from being smaller than the target torque −Q but surging can be prevented from being brought about in the high pressure compressor 5. Particularly, even when there is brought about a situation in which surging is liable to be brought about in the high pressure compressor 5 by a disturbance, surging of the high pressure compressor 5 can be prevented from being brought about. Next, a fourth embodiment of the invention will be explained in reference to FIG. 12 through FIG. 14. Although a basic constitution of a power generating equipment of the embodiment is similar to that of the third embodiment, the embodiment differs from the first embodiment in a method of controlling the first bypass valve. Therefore, according to the embodiment, only the method of controlling the first bypass valve will be explained in reference to FIG. 12 through FIG. 14, an explanation of other constituent element or the like will be omitted. FIG. 12 is a schematic view for explaining a constitution of the power generating equipment according to the embodiment. FIG. 13 is a block diagram for explaining a control of the power generating equipment of FIG. 12. Further, constituent elements the same as those of the third embodiment are attached with the same notations and an explanation thereof will be omitted. As shown by FIG. 12 and FIG. 13, the power generating equipment 301 of the embodiment is further provided with a first pressure ratio measuring portion 361 for measuring a pressure ratio of a working fluid pressure on a suction side and a working fluid pressure of a delivery side of the low pressure compressor 4. As shown by FIG. 12, the first pressure ratio measuring portion 361 is a measuring portion for measuring a ratio of a pressure of the working fluid sucked to the low pressure compressor 4 and a pressure of the working fluid delivered from the low pressure compressor 4. As shown by FIG. 13, the pressure ratio measured by the first pressure ratio measuring portion 361 is inputted to a first surging control portion 351 of a control portion 350. Further, as shown by FIG. 12, the control portion 350 of the power generating equipment 301 of the embodiment is provided with the program control portion 151 for controlling the opening degree V1 of the first bypass valve 36 based on the torque measured by the torque meter 11, the first surging control portion (control portion) 351 for controlling the opening degree V1 of the first bypass valve 36 by calculating the suction flow rate for preventing surging of the low pressure compressor 4 from being brought about, and a first selecting portion 352 for selecting a control signal having a larger opening degree of the first bypass valve from control signals outputted from the program control portion 151 and the first surging control portion 351. As shown by FIG. 12, the first surging control portion 351 calculates a suction flow rate for preventing surging of the low pressure compressor 4 from being brought about based on the pressure ratio inputted from the first pressure ratio measuring portion 361 and the modified revolution number N1 inputted from the modified revolution number calculating portion 263, calculates the opening degree V1 of the first bypass valve 36 based on the calculated suction flow rate and outputs the control signal for controlling the opening degree. The control signal outputted from the first surging control portion 351 is inputted to the first selecting portion 352. Here, the calculated suction flow rate for preventing surging from being brought about is a flow rate of adding a predetermined allowance (margin) to the suction flow rate for bringing about surging at the low pressure compressor 4. Therefore, the calculated opening degree V1 of the first bypass valve 36 is an opening degree by which the working fluid of the flow rate added with the above-described margin flows to the low pressure compressor 4. As shown by FIG. 12, the first selecting portion 352 selects the control signal having the larger opening degree of the opening degrees V1 of the first bypass valve 36 applied to the control signal inputted from the program control portion 151 and the control signal inputted from the first surging control portion 351. The selected control signal is outputted from the first selecting portion 352 to the first bypass valve 36. Next, a control in starting the power generating equipment 301 constituting a characteristic of the embodiment will be explained. In starting the power generating equipment 301, similar to the third embodiment, the second selecting portion 252 is inputted with signals for controlling the opening degree V2 of the second bypass valve 43 from the feedback control portion 51 and the second surging control portion 251 and the control signal selected by the second selecting portion 252 is inputted to the second bypass valve 43. FIG. 14 is a flowchart for explaining the control in starting the power generating equipment of FIG. 12. Simultaneously with the control of the opening degree of the second bypass valve 43, as shown by FIG. 14, calculation of the bypass flow rate at the program control portion 151, and calculation of the opening degree of the first bypass valve 36 (step S31 (first calculating step)), and calculation of the bypass flow rate at the first surging control portion 351, and calculation of the opening degree of the first bypass valve 36 (step S32 (second calculating step)) are executed independently from each other. As shown by FIG. 13, the first surging control portion 351 of the control portion 350 is inputted with the pressure ratio of the working fluid on the suction side and on the delivery side of the low pressure compressor 4 from the first pressure ratio measuring portion 361 and inputted with the corrected revolution number N1 from the corrected revolution number calculating portion 263. The first surging control portion 351 calculates the suction flow rate for preventing the surging of the low pressure compressor 4 from being brought about based on the inputted pressure ratio and the inputted corrected revolution number N1. The flow rate for preventing surging from being brought about in the first surging control portion 351 is calculated based on a table or the like previously stored to the first surging control portion 351. The first surging control portion 351 calculates the opening degree V1 of the first bypass valve 36 based on a calculated suction flow rate and outputs the control signal of controlling the opening degree of the first bypass valve 36 to the first selecting portion 352. As shown by FIG. 13, the first selecting portion 352 is inputted with the control signal of controlling the opening degree V1 of the first bypass valve 36 from the first surging control portion 351 and inputted with the control signal of controlling the opening degree V1 of the first bypass valve 36 also from the program control portion 151. The first selecting portion 352 selects the control signal having the larger opening degree of the first bypass valve 36 and outputs the selected control signal to the first bypass valve 36 (step S33 (selecting step)). At the first bypass valve 36, the opening degree is controlled based on the inputted control signal and the flow rate of the working fluid flowing through the first bypass flow path 31 is controlled (step S34 (flow rate controlling step)). According to the above-described constitution, the bypass flow rate having the larger flow rate is selected from the bypass flow rate based on an elapsed time period after starting to increase the revolution number of the low pressure compressor 4 or the like and the bypass flow rate for preventing surging of the low pressure compressor 4 from being about, the flow rate of the bypass working fluid is controlled to the selected flow rate, and therefore, not only the torque applied to the speed reduction gear portion 6 can be prevented from becoming lower than the target torque −Q but surging of the low pressure compressor 4 can be prevented from being brought about. Particularly, surging of the low pressure compressor 4 can be prevented from being brought about even when there is brought about a situation in which surging of the low pressure compressor 4 is easy to be brought about by a disturbance. Further, the technical range of the invention is not limited to the above-described embodiments but can variously be modified within the range not deviated from the gist of the invention. For example, although according to the above-described embodiments, an explanation has been given by applying the invention to the example of controlling the second bypass valve 43 by the feedback control portion 51 and controlling the first bypass valve 36 by the program control portion 151, conversely, the first bypass valve 36 may be controlled by the feedback control portion 51, the second bypass valve 43 may be controlled by the control portion 151 and the invention is not limited thereto.
	summary
051329972	abstract	An X-ray spectroscopic analyzing apparatus which comprises a source of X-rays, a first analyzing crystal for diffracting the X-rays from the X-ray source, and a second analyzing crystal for diffracting the X-rays from the X-ray source and also for passing therethrough a diffracted X-ray component from the first analyzing crystal. The first and second analyzing crystals are so disposed and so positioned as to permit the diffracted X-ray components of different wavelengths to travel along a single path towards a sample to be analyzed. On an optical path extending between the X-ray source and the sample, a filtering means for cutting a portion of the X-rays which has a wavelength shorter than a predetermined wavelength.
	abstract	The X-ray detection apparatus is equipped with an X-ray irradiation unit, an X-ray detector, a movable collimator and a shield for blocking X-rays. The shield blocks X-rays, which are to enter the X-ray detector directly from the X-ray irradiation unit. The shield also blocks fluorescent X-rays and scattered X-rays generated by irradiation of the collimator with X-rays. In such a manner, it is possible to prevent X-rays other than fluorescent X-rays from the sample S from being detected by the X-ray detector. The shield is joined with the collimator, so that the collimator and the shield move as a unit. It is possible to locate the shield even in a downsized X-ray detection apparatus.
052456412	claims	1. For use in a storage pool for nuclear fuel assemblies, said pool containing a coolant and having a pool floor, a fuel rack module comprising a base plate to be disposed generally horizontally on said floor, a plurality of cells welded to said base plate, each said cell having elongated wall means for defining a transverse cross-sectional area for receiving a fuel assembly and a volume for accommodating said received fuel assembly, the outer walls of said cells of said plurality being wholly within the boundaries of said base plate, and a plurality of receptacles structured uniquely to accommodate the plate-engaging mechanism of lifting apparatus, said receptacles being spaced uniformly on said base plate for receiving and locking said plate-engaging mechanism for lifting the base plate alone or the base plate with said cells secured thereto, said receptacles being wholly within the periphery of said base plate and each said receptacle including a hole in said base plate having a contour such as to receive the lower end of a fuel assembly with slots extending from the periphery of said hole and a block secured to the underside of said base plate having a hole and slots coextensive with the hole and slots in said base plate, said hole and slots being dimensioned to pass said plate-engaging mechanism to the underside of said block, and said receptacle also having means in said block angularly displaced from said slots for locking said engaging mechanism, the blocks of all said receptacles being wholly within the boundary of said base plate. 2. For use in a storage pool for nuclear fuel assemblies, said pool containing a coolant and having a pool floor; a fuel rack module comprising: a base plate to be disposed generally horizontally on said floor, a plurality of cells mounted on said base plate, each said cell having elongated wall means defining a transverse cross-sectional area for receiving a fuel assembly and having a volume for accommodating said received fuel assembly, the outer walls of said cells of said plurality being wholly within the boundaries of said base plate, and means, in said base plate wholly within the periphery of said base plate, structured uniquely to accommodate lifting apparatus for lifting said module, said uniquely-structured means including at least one opening in the base plate and also including at least one member connected to the underside of said base plate, said member having an opening coincident with the opening in said base plate, said coincident opening being shaped to pass the plate engaging mechanism of the lifting apparatus and said member also including means for locking said lifting apparatus after it has passed through said openings. 3. For use in a storage pool for nuclear fuel assemblies, said pool containing a coolant and having a pool floor; a fuel rack module comprising a generally rectangular base plate to be disposed generally horizontally on said floor, a plurality of cells mounted on said base plate, each said cell having elongated wall means defining a transverse cross-sectional area for receiving a fuel assembly and having a volume for accommodating said received fuel assembly, the outer walls of said cells of said plurality being wholly within the boundaries of said base plate, and means, in said base plate wholly within the periphery of said base plate, structured uniquely to accommodate lifting apparatus for lifting said module, said uniquely-structured means including, near each corner of said base plate a member secured to the underside of said base plate, each said member and the base plate having coincident openings shaped to pass the plate-engaging mechanism of the lifting apparatus and each said member also including at each corner, means for locking said lifting apparatus, each of said members and said locking means being symmetrically disposed with respect to said base plate.
054209021	abstract	A fuel assembly includes a cluster of mutually parallel fuel rods. A fuel assembly channel laterally surrounds the cluster of fuel rods and has a substantially rectangular cross section and flat channel walls. Grid-like spacers having meshes formed therein each receive a respective one of the fuel rods for guiding the fuel rods in a plurality of axial positions. At least one support spring laterally supports each respective one of the fuel rods in the mesh guiding the fuel rod. Each of the spacers have inner ribs being aligned parallel to the fuel rods and outer peripheral ribs opposite the channel walls. At least some of the inner ribs are fastened to the peripheral ribs, and the outer peripheral ribs are joined together only by the inner ribs.
055641024	abstract	An object material to be melted (radioactive liquid waste and glass material) 16 is charged into the interior of a cold-crucible induction melting apparatus 10, and a conductor 18 the melting point of which is higher than that of the glass material is inserted into a melting furnace 12. A high-frequency current is supplied to a coil 14 so as to generate heat in the conductor and indirectly heat the glass material. The conductor is withdrawn after a part of the glass material has been put in a molten state. The glass material as a whole is thereafter kept in a molten state while maintaining the induction heating by the molten glass material. The conductor inserted into the melting furnace is, for example, a silicon carbide rod. The surface of the molten material which contacts the inner surface of the melting furnace turns into a solid layer (skull) due to cooling, so that the molten material does not directly contact the refractories. This enables the high-temperature corrosion of the melting furnace to be prevented.
	claims	1. A method of making a gap between first and second objects have a predetermined value, said method comprising the steps of:a first introduction step of introducing light to a first entry window on the first object;a first detection step of detecting a first total intensity of light from a first exit window on the first object with respect to each of a plurality of values of the gap, the first exit window being at such a position that light would enter through the first entry window of the first object, reflect off the second object, and then enter the first exit window if the gap has the predetermined value;a second introduction step of introducing light into a second entry window on the first object;a second detection step of detecting a second total intensity of light from a second exit window on the first object with respect to each of a plurality of values of the gap, the second exit window being at such a position that light would enter through the second entry window of the first object, reflect off the second object, and then enter the second exit window if the gap has the predetermined value;an obtaining step of obtaining a position of the second object, with respect to a direction of the gap, where the gap has the predetermined value based on a plurality of total intensities of light detected in said first detection step and a plurality of total intensities in said second detection step; anda positioning step of positioning the second object at the position obtained in said obtaining step,wherein said first and second introduction steps respectively introduce light to a diffraction grating, as the first and second entry windows, common to said first detection step and said second detection step. 2. A method according to claim 1, wherein said first detection step detects a total intensity of light from a first diffraction grating in the first exit window, and said second detection step detects a total intensity of light from a second diffraction grating in the second exit window. 3. A method according to claim 1, wherein said first detection step includes a step of: varying a position of the second object with respect to the direction of the gap while light is introduced to the first entry window on the first object and a total intensity of the light from the first exit window on the first object is detected. 4. A method according to claim 1, wherein said second detection step includes a step of varying a position of the second object with respect to the direction of the gap while light is introduced to the second entry window on the first object and a total intensity of the light from the second exit window on the first object is detected. 5. An apparatus for making a gap between first and second objects, said apparatus comprising:a light introducing unit configured to introduce light to a first entry window formed in the first object;a first detection unit configured to detect a total intensity of light from a first exit window formed in the first object with respect to each of a plurality of values of the gap, the first exit window being at such a position that light would enter through the first entry window of the first object, reflect off the second object, and then enter the first exit window if the gap has a predetermined value;a second detection unit configured to detect a total intensity of light from a second exit window formed in the first object with respect to each of a plurality of values of the gap, the second exit window being at such a position that light would enter through a second entry window of the first object, reflect off the second object, and then enter the second exit window if the gap has the predetermined value; anda positioning mechanism configured to obtain a position of the second object, with respect to a direction of the gap, where the gap has the predetermined value based on a plurality of total intensities of light detected by said first detection unit and a plurality of total intensities of light detected by said second detection unit, and to position the second object at the obtained position,wherein said light introducing unit is configured to introduce light to a diffraction grating, as the first and second entry windows, common to said first detection unit and said second detection unit. 6. An apparatus according to claim 5, wherein said first detection unit is configured to detect a total intensity of light from a first diffraction grating in said first exit window, and said second detection unit is configured to detect a total intensity of light from a second diffraction grating in the second exit window.
050948032	abstract	A steam generator utilized for a liquid-metal coolant reactor comprises an outer shell body in which an electromagnetic pump is arranged. The electromagnetic pump comprises a hollow cylindrical iron core provided with comb-shaped portion at an outer peripheral surface thereof and an annular stator coil means assembled in the comb-shaped portion of the cylindrical iron core. A main passage of liquid metal is formed on a side on which the stator coil of the iron core is assembled and a cooling bypass passage is formed at substantially the central portion of the cylindrical iron core in a vertically penetrating fashion.
	description	This application is a U.S. National Stage of International Application No. PCT/RU2018/000900, filed on Dec. 28, 2018, and published as WO 2020/067920 on Apr. 2, 2020, titled “Device for Confining Nuclear Reactor Core Melt,” which claims priority to RU 2018133765 filed on Sep. 25, 2018. Each application, publication, and patent listed in this paragraph are hereby incorporated by reference in their entireties. The invention relates to nuclear engineering, in particular, to systems that ensure the safety of nuclear power plants (NPP), and can be used in severe accidents that lead to the core meltdown, nuclear reactor pressure vessel destruction and the release of the melt into the space of the NPP containment. The greatest radiation hazard is represented by accidents with core meltdown, which can occur in various combinations of failures (destruction of equipment components) of active and passive safety systems and normal operation systems, or in conditions of the total power loss of the NPP, and the inability to supply power within the time period established by the NPP design to ensure the emergency core cooling. In case of such accidents occurrence, the core meltdown—corium, melting the core internals and the reactor vessel, flows out of it and, due to the residual heat generation remaining in it, can violate the integrity of the NPP containment' the last barrier to the release of radioactive products into the environment. To eliminate this, it is necessary to confine the corium that has flowed out of the reactor vessel and ensure its continuous cooling, until the complete crystallization of all corium components. This function is performed by the water-cooled nuclear reactor core melt cooling and confinement system, which prevents damage to the NPP containment and, thereby, protects the population and the environment from radiation exposure in severe accidents of nuclear reactors. The device (1) for confining nuclear reactor core melt comprising a melt trap, which is installed in the reactor vessel bottom and provided with a cooled containment as the multilayer vessel, a filler for the melt dilution, placed in the specified multi-layered vessel, the bottom support consisting of horizontal, sectional, solid or split, embedded plate mounted on a multilayer vessel in the concrete of the reactor pit, a cylindrical vertical tube connecting the body of the melt trap with a bond plate by means of clamps, and fasteners, is already known. The drawback of the device is low reliability, due to the fact that when a peak of non-axisymmetric melt discharge into the melt trap body, the melt trap can shift under the influence of shock loads and tip over the vertical wall of the reactor pit, which will lead to the release of the melt outside the melt trap. The melt confining system (2), mounted in a reactor pit consisting of a support surface and side walls, comprising a vessel for the melt, and upper supports mounted on the protrusions of the side walls of the reactor pit, is already known. The drawback of the system is low reliability, due to the fact that when a peak of the melt discharge into the melt trap body, the upper supports are deformed, that leads to the melt trap fall on the lower surface of the reactor pit and its overturning on the vertical wall of the reactor pit, which will lead to the melt release outside the melt trap. The technical result of the claimed invention is to increase the reliability of the device for confining nuclear reactor core melt. The objects to be solved by the claimed invention are to eliminate the overturning of the melt trap of the device for confining nuclear reactor core melt when exposed to non-axisymmetric shock loads and the melt releases out of the body. The objects are solved due to the fact that the device for confining nuclear reactor core melt comprising a melt trap, which is installed in the reactor vessel bottom and provided with a cooled containment as the multilayer vessel, a filler for the melt dilution, placed in the specified multi-layered vessel, the upper support, the bottom support consisting of horizontal, sectional, solid or split, embedded plate mounted on a multilayer vessel in the concrete of the reactor pit, according to the invention, the horizontal sectional, solid or split embedded plate comprises radial supports, the melt trap comprising radial supports, based on the radial support of the horizontal sectional, solid or split embedded plate, the radial supports of the horizontal radial section, solid or split embedded plate and the radial supports of the melt trap body are connected through fasteners, while the radial supports and the clamps have oval holes, the upper support comprises turnbuckles, mounted in pairs on the upper part of the melt trap body so that the longitudinal axis of each radial support of the melt trap bottom support passes in projection at an equispaced distance from the fitting location of the paired turnbuckles installed tangentially to the melt trap body and connecting the melt trap body with the reactor pit vertical wall, while the fasteners have holes made in the form of hyperbolic surfaces. One characteristic feature of the claimed invention is the upper support, which consists of paired turnbuckles that are located on the melt trap outer body such a way that the longitudinal axis of the radial support passes in the projection at an equispaced distance from the places where the turnbuckles are fitted. Another characteristic feature is that the turnbuckles are mounted tangent to the melt trap body. One more characteristic feature of the claimed invention are fasteners with holes made in the form of hyperbolic surfaces. This type of turnbuckles arrangement provides: free thermal radial expansions of the melt trap body in the turnbuckles plane (in the horizontal plane) due to the tangent release of the turnbuckles in the melt trap body lugs, in which any radial expansions of the melt trap body lead only to a change in the plane angle of the turnbuckle tangent position relative to the melt trap body generating line. Thus, the risk of turnbuckles formability with loss of their performance and the risk of cracking or destruction of the melt trap body is eliminated; non-exceedance of the radial pullout strength effect on embedded parts in the concrete reactor pit (controlled loading) due to the distribution of the radial shock load between all the turnbuckles. In this case, a part of the turnbuckles will work for compression, part—for stretching in the turnbuckles plane. In this case, the horizontal shock load leads to planar vibrations of the melt trap body flange, in which all turnbuckles work alternately for tension and compression in the area of elastic deformations of the turnbuckles, up to the attenuation of the planar vibrations; reducing the non-axisymmetric impact on the bottom support of the melt trap body with non-axisymmetric axial (vertical) shock loading of the body in the flange area due to the distribution of the axial shock load between all the turnbuckles. In this case, those turnbuckles in the area of which the effect of a non-axisymmetric axial shock load has appeared do not provide mechanical resistance to the shape change of the melt trap body flange. Thus, the melt trap body flange, in the area of which the axial impact has appeared, redistributes the shock load along its perimeter, redistributing the axial impact into two additional components with the formation of both azimuth (along the perimeter of the body) and radial (planar) vibrations. A part of the impact in the form of axial elastic vibrations of the melt trap body does not affect the turnbuckles, azimuth vibrations are damped by elastic deformations of the turnbuckles, and radial vibrations propagating in the plane of the turnbuckles are alternately damped by them, as when a radial shock load is damped; non-exceedance of the effect of azimuth pullout strength on embedded parts in the concrete reactor pit during seismic effects on the melt trap body (damping torsional vibrations of the body flange) due to the alternate operation of the turnbuckles for tension and compression under the influence of flat torsional vibrations from the side of the flange of the melt trap body flange. The vibrations damping is provided by absorbing the energy of the turnbuckles elastic deformations, up to the attenuation of torsional vibrations; maintaining the melt trap body flange integrity, embedded parts of the reactor pit and the upper support during axial thermal expansion of the melt trap body by ensuring the turnability of the turnbuckles fork-plugs in the axial (vertical) plane, which is provided by the hyperbolic surface of the turnbuckles mounting holes in the fork-plugs of the melt trap body and in the fork-plugs of vertical embedded plates installed in the reactor pit. Execution of the hyperbolic surface of the holes in the fasteners may be performed both on the melt trap body and on embedded plates. The claimed invention is functioning as follows. FIG. 1 shows the device (1) for confining nuclear reactor core melt containing melt (3) trap (2), which is installed in the reactor vessel bottom (18) and provided with a cooled containment as the multilayer vessel, a filler (4) for the melt (3) dilution, placed in the specified multi-layered vessel, the bottom support (5) consisting of the radial supports (6) positioned on the external side of the melt (3) trap (2) bottom part body (7) and based on the radial supports (8) of the horizontal embedded plate (9) which are connected with fasteners (10), the upper support (11), that comprises turnbuckles (12), mounted in pairs on the upper part of the melt (3) trap (2) body (7) so that the longitudinal axis (B) of each radial support (6) of the melt (3) trap (2) bottom support (5) passes in projection at an equispaced distance from the fitting location of the paired turnbuckles installed (13) tangentially to the melt (3) trap (2) body (7) and connecting the melt (3) trap (2) body (7) with the reactor pit vertical wall (14); the cross section (C) of the turnbuckle attachment to the trap body, which is also illustrated in FIG. 4. FIG. 1a shows an enlarged view of the upper support (11) with turnbuckles (12), and the fitting location of the paired turnbuckles (13), with additional details being illustrated in FIG. 2. FIG. 1b shows a top view of a portion of the upper support and bottom support and illustrates how the installation sites of paired turnbuckles can be equidistant (e.g., a distance (A)) from the longitudinal axis (B) of a lower radial support of the bottom support. As illustrated, the longitudinal axis (B) of the lower radial support is perpendicular to the sheet and directed upwards as also illustrated in FIG. 1. FIG. 1c shows an example of a solid embedded plate of the bottom support. FIG. 1d shows an example of a split embedded plate of the bottom support that is divided, for example, into two concentric parts. FIG. 1e shows an example of a sectional embedded plate of the bottom support, which can be, for example, in the form of 4 sections. FIG. 2 shows an example of a pair of turnbuckles (12) mounted on a flange (17) at the upper part of the melt trap (2) body. As shown on FIG. 3, FIG. 4a, and FIG. 4b the pull rods (15) of the turnbuckles (12) have holes (16) made in the form of hyperbolic surfaces, in which the axes (19) of the fasteners (20) of the upper support (11) are installed. When changing the position of the pull rods (15) of the turnbuckles (12) connecting the body (7) to the fitting locations (13) of the paired turnbuckles (12), the pull rods (15) rotate in the axial plane passing through the axis of each turnbuckles (12). At the moment of the reactor vessel destruction, the core melt (2) under the action of hydrostatic and overpressure begins to flow into the double body (7) of the melt trap and comes into contact with the filler (4). In the case of a non-axisymmetric peak of the melt (2) discharge, for example, 60 tons of superheated steel for 30 seconds, the main shock load falls on the side inner wall of the body (7) of the melt (3) trap (2). As shown on FIG. 2, in this case, those turnbuckles (12a), in the area of which the effect of a non-axisymmetric axial shock load has appeared, do not have mechanical resistance to the shape of the flange (17) of the body (7). Thus, the body flange (17), in the area of which the axial impact has appeared, redistributes the shock load along its perimeter, redistributing the axial impact into two additional components with the formation of both azimuth (along the perimeter of the body (7)) and radial (planar) vibrations. A part of the impact in the form of axial elastic vibrations of the body (7) does not affect the turnbuckles (12a), azimuth vibrations are damped by elastic deformations of the turnbuckles (12b), and radial vibrations propagating in the plane of the turnbuckles (12a) are alternately damped by them, as when a radial shock load is damped. The radial shock load is damped as follows. A part of the turnbuckles (12a) will work for compression, part—for stretching in the turnbuckles (12) plane. In this case, the horizontal shock load leads to planar vibrations of the body (7) flange (17), in which all turnbuckles (12) work alternately for tension and compression in the area of elastic deformations of the turnbuckles (12), up to the attenuation of the planar vibrations. The use of the upper support together with the bottom support in the device for confining nuclear reactor core melt made it possible to completely eliminate the probability of the melt release outside the melt trap body by excluding its overturning, even when exposed to a non-axisymmetric shock load. 1. Russian Patent No. 2398294, IPC G21C 9/00, priority dated Apr. 15, 2009. 2. Japanese Patent JP2010271261, IPCG21C9/00, priority dated May 25, 2009.
043022855	description	DETAILED DESCRIPTION OF THE INVENTION Referring to FIG. 1 a neutron activation analysis installation forming the subject of the present invention is designed for quantitative determination of the chemical composition of various materials and quick non-destructive test in the production of metals, alloys, semiconductor and other materials to obtain a desired chemical composition of semi-finished and finished products. The proposed installation comprises a neutron generator 1 whose target chamber 2 communicates through a transport means 3 with a test sample receiving and loading assembly 4 which, in its turn, communicates with an impurity concentration measuring unit 5. The impurity concentration measuring unit includes a detector 6, measuring equipment 7 and a minicomputer 8 and communicates with the receiving and loading unit 4 over a through channel 9. The channel 9 has a through lateral port 10 communicating on one side with the input of an irradiated sample surface layer removal unit 11, an irradiated sample distribution assembly 12 being arranged on the other side of the port 10, the distribution assembly 12 represents an air cylinder 13 (FIG. 2) with a hollow rod 14 having a bar 15 arranged along the axis thereof, said bar carrying on its end a sample receiver 16. The bar 15 is disposed in a manner allowing its rotation along the longitudinal axis thereof and reciprocating motion through the port 10 in the through channel 9. In one extreme position the bar 15 does not reach the channel 9 (ref. I of FIG. 2 shown with a dashed line) leaving it vacant to enable a sample 17 falling from a capsule 18 (FIG. 2) of the receiving and loading assembly 4 to pass along the through channel 9 into the detector 6 (FIG. 1). In the intermediate position (ref. II of FIG. 2) the sample receiver 16 secured to the bar 15 is found in the channel 9, thus blocking the latter. In the other extreme position (ref. III shown with a dashed line) the bar 15 with the sample receiver 16 gets into the surface layer removal unit 11 after passage via the port 10 in the through channel 9. The reciprocating motion of the bar 15 with the sample receiver 16 and the 180-degree turn of the bar 15 in the through channel 9 are accomplished by supplying air to the air cylinder 13 through holes a and b. Under its pressure the rod 14 with the bar 15 carrying the sample receiver 16 moves in either direction. The bar 15 is turned about its axis by means of a turning mechanism composed of a piston 19 contained within the rod 14 encompassing the bar 15, secured on the bar 15 with a key 20 in a manner allowing sliding motion along the axis of the bar 15 and coupled to the rod 14 by means of a carrier 21 which is rigidly fixed on the piston 19 in a manner allowing its motion through a screw slot 22 in the rod 14. The air displacing the piston 19 is supplied into the cavity of the rod 14 from the air cylinder 13 through a hole "c". A lever 23 enables installation of the bar 15 in position I or II. In the event of an oxygen content determination the detector 6 may, for example, represent a device based on two scintillation units with large lead-shielded NaI (T1) crystals. The measuring equipment 7 may include five discriminating amplifiers, four recomputation devices, and two coincidence circuits. In doing oxygen content analysis, for example, use is made of two sample activity measuring channels, one neutron flux monitoring channel used during irradiation of samples, and one neutron flux test channel. In doing nitrogen content analysis use is made of five channels and two coincidence circuits. The minicomputer 8 processing measurement data may include a keyboard computer and a matching unit which interrogates scales, feeds data into the computer and initiates computation instructions in accordance with the preset algorithm. For example, an oxygen content in the sample in accordance with the preset algorithm is determined from the formula ##EQU1## where .eta..sub.x =oxygen content in the test sample, % by weight; .eta..sub.0 =oxygen content in the reference sample, % by weight; PA1 N.sub.x =number of counts for the sample; PA1 N.sub.1x =number of background counts for the sample; PA1 N.sub.0 =number of counts for the reference sample; PA1 N.sub.10 =number of background counts for the reference sample; PA1 M.sub.x =number of sample monitor counts; PA1 M.sub.0 =number of reference sample monitor counts; PA1 M.sub.x =weight of the sample, g; PA1 M.sub.0 =weight of the reference sample, g; and PA1 K=coefficient accounting for a difference in absorption of 16.sub.N isotope gammas in the test and reference samples. For clarity, FIG. 3 presents a perspective view of the irradiated sample distribution assembly 12 with the surface layer removal unit 11 and a portion of the channel 9. Turning now to FIG. 4 the irradiated sample surface layer removal unit 11 comprises at least three communicating chambers 24, 25, 26 arranged successively in the direction of reciprocating motion of the bar 15, the position of the last chamber 24 corresponding to extreme position III of the bar 15. The sample receiver 16 is protected by a cylindrical guard 27 to preclude the penetration of a reagent from one of the chambers (24 to 26) to the other during backward motion of the bar 15. Two vessels 28 containing the reagents are provided for each chamber (24 to 26), one vessel being used to treat the sample with a required reagent, while the other vessel is used for draining the reagent. The air cylinder 13 is provided with air locks 29 whose cavities communicate with the cavity of the air cylinder 13 through ports 30 disposed along the periphery thereof. The number of the air locks 29 suits the number of partitions between the chambers 24 to 26, each lock being designed to brake and stop the bar 15 with the sample receiver 16 in one of the chambers (25 or 26) during its backward motion by discharging the air through outlet connections d, e, respectively. The ports a, b and connections d, e are closed and opened by electromagnetic valves 31 a, b, d and e (letter designations of the valves correspond to letter designations of the respective holes). The operation of the valves 31 is controlled by a timer (not shown in the drawings) activated on signals from a photosenser 32 arranged in the channel 9. Referring to FIG. 5 the sample 17 is contained within the receiver 16 placed in one of its extreme position in the chamber 24. In FIG. 6 the ports 30 are distributed along the periphery of the air cylinder 13 and the air lock 29 is shown with the outlet connection e. The neutron activation analysis installation forming the subject of the present invention operates in the following manner. Before operation, it is necessary to estimate the purity of the test sample 17 as regards an impurity content. If the impurity content has a concentration exceeding 5.multidot.10.sup.-3 % by weight and no surface removal is required after irradiation prior to measuring the sample activity, the lever 23 should be set to position I so that the through channel 9 is unblocked to allow passage of the sample to the detector 6. The test sample 17 is enclosed in the capsule 18 which is then placed in the receiving and loading assembly 4. The transport means 3 is used to deliver the sample 17 from the receiving and loading assembly to the target chamber 2 of the neutron generator 1 wherein the sample 17 is irradiated. The same transport means 3 delivers the irradiated sample 17 to the receiving and loading assembly 4 whence it goes over the through channel 9 to the detector 6. The sample activity is measured by the measuring equipment 7, the impurity content is calculated from formula (1) using the minicomputer 8 and a presentation of the result is provided. If the sample 17 is pure or highly pure having, for example, an oxygen content less than 5.multidot.10.sup.-3 % by weight, the lever 23 should be set to position II so that the sample receiver 16 blocks the through channel 9. The transport means 3 delivers the sample 17 with the capsule 18 to the target chamber 2 of the neutron generator 1 wherein it is irradiated. At a preset time after irradiation the transport means 3 delivers the sample 17 with the capsule 18 to the receiving and loading assembly 4. Therefrom the sample 17 removed from the capsule 18 passes over the through channel 9 to the sample receiver 16. As the sample 17 passes over the channel 9, the photosensor 32 furnisches a signal causing the electromagnetic valve 31 a to open. From said valve the compressed air is supplied through the port a to the air cylinder 13, thus pushing the rod 14 with the internal bar 15 whose end mounts the receiver 16 with the sample 17 via the through port 10 into the irradiated sample surface layer removal unit 11. The receiver 16 with the sample 17 is placed in the extreme chamber 24 under the inlet connection coupled to the vessel 28 filled with the reagent required to treat the irradiated sample 17 with a view to removing its surface layer. When the receiver 16 is installed in the chamber 24, the running reagent in a uniform manner the surface layer from the irradiated sample 17 after which it is drained into the second vessel 28 through the outlet connection. At a preset time after the treatment of the sample in the chamber 24 is completed the compressed air is supplied through the open electromagnetic valve 31b and the respective port b to the air cylinder 13, thus pushing the rod 14 with the bar 15 and the receiver 16 which is stopped in the next chamber 25 to enable further treatment or washing of the sample 17 in the receiver 16 with running reagent or water. As this happens, the guard 27 closes the port through which the chambers 24 and 25 communicate. To stop the receiver in the chamber 25, the rod 14 is braked by discharging the air from the air lock 29 through the connection e and the electromagnetic valve 31e. At a preset time after the treatment of the sample 17 in the chamber 25 is completed, the valve 31e closes and the air coming through the port b pushes the rod 14 until the receiver 16 with the sample 17 stops in the chamber 26 to enable further treatment and blowing of the sample 17 with air. The receiver 16 with the sample 17 is stopped in the chamber 26 by discharging the air from the second air lock 29 through the connection d and the electromagnetic valve 31g. In this case, the guard 27 closes the two ports through which the chambers 24 to 26 communicate. At a preset time after the treatment of the sample 17 in the chamber 26 is completed, the valve 31d closes and and the compressed air is supplied through the valve 31b and the port b to the air cylinder 13, thus pushing the rod 14 until the receiver 16 with the sample 17 enters the through channel 9. When the receiver 16 with the sample 17 is placed in the through chanel 9, the compressed air is supplied from the air cylinder 13 through the port c to the cavity of the rod 14, thus pushing the piston 19 which slides along the axis of the bar 15. Since the piston 19 is coupled to the rod 14 by means of the carrier 21 rigidly fixed on the piston 19 in a manner allowing its motion through the screw slot 22 in the rod 14, the bar 15 with the sample receiver 16 makes a 180-degree turn thanks to the screw slot 22 whereby the sample 17 is removed from the receiver 16 and supplied to the detector 6 over the through channel 9. Next, the sample activity is measured by the equipment 7 and the minicomputer comutes in accordance with the preset algorithm (say, formula (1) an oxygen content and feeds the test data to a printer or a display unit. The minicomputer also allows computing errors of a randomly chosen set of data or a single test. The entire impurity determination process is invariably short, say, from 1.5 to 3 min in doing oxygen content analysis, its duration being dependent upon the half-life of a given radioisotope. A short impurity determination time permits monitoring the entire process of fabricating semi-finished products to a high accuracy. Another advantage of the proposed neutron activation analysis installation over the prior art is that it holds much promise as regards sensitivity, accuracy, use of a still greater number of elements for impurity content analysis, automation of the entire test process and fast data output by sound signalling, visual presentation or printing. The aforesaid advantage is associated with the use of a high-current neutron generator with a minimum flux of 5.multidot.10.sup.12 neutron/s employing deuterium-tritium beams and a tritium-fed target and also of a minicomputer and up-to-date integrated circuits, which is generally a space-saving factor allowing further miniaturization. Using neutron moderators the hereinproposed installation is capable of operating not only with direct-action accelerators generating monochromatic .about.14 MeV neutrons but also with slow and thermal reactors. Furthermore, the installation forming the subject of the present invention makes it possible to do volume, surface and correlation analyses.
	summary
053612868	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In accordance with the method of the invention, a nozzle cleaning tool/fixture assembly 6 or TBF cleaning tool/fixture assembly 8 is attached to an inlet mixer 46 by way of a clamping fixture 2 (see FIG. 3). UHP water is supplied to the cleaning tool by an UHP pump 66 via one of a plurality of umbilicals 68 unwound from a hose reel 70. Low pressure water (e.g., 600 psi) is supplied by a pneumatically operated intensifier pump 72, which is controlled by hydraulic control panel 74. Electrical power and sensing for the system motors is supplied by another umbilical connected to an electrical junction box 76 mounted on hose reel 70. Central computer control system 78 controls and monitors the position and orientation of the cleaning tool, and activates and deactivates the supply of UHP and low-pressure water to the cleaning tool and related fixtures. Electrical power is optionally supplied to UHP pump 66 and to the monitoring system 80 (including a TV monitor, a character generator and a video cassette recorder) by a transformer 82 if 460 V 60 Hz or 380 V 50 Hz is unavailable. Hose reel 70, pump 72, hydraulic control panel 74 and monitoring equipment 80 are installed on the refueling platform 84, which is translatable along a pair of tracks 86. UHP pump 66, computer control system 78, transformer 82 and tracks 86 reside on the refueling floor. During UHP cleaning, scale buildup debris dispersed in the water inside the inlet mixer is removed from the internal surfaces of the inlet mixer 46 and sucked out by filter/pump 88 via inlet suction line 90. The pump operates at low pressure (i.e., about 100 psi). The filter collects the debris. Discharge pump 92 pumps the filtered water back into the pool. Referring to FIG. 4, a clamping fixture 4 comprises a pair of clamp arms 100, each of which is driven by a pair of clamp cylinders 102 (only one cylinder of each pair is visible). Clamp arms 100 clamp onto the inlet mixer 46 in response to controlled low-pressure water received via an umbilical 101 (see FIG. 10). A base 104 with a pair of locating pins 106 and 106' slides on rollers 108 relative to the frame of the clamping fixture. Base 104 is locked in a desired position by actuation of a base lock cylinder 107 using low-pressure water received via an umbilical 111 (shown in FIG. 6). In the locked state, base lock cylinder secures a locking pin 105 extending from base 104. Before the clamping fixture can be clamped, it must precisely positioned relative to the inlet mixer, since in later operations the clamping fixtures provides the sole support for the nozzle and TBF cleaning fixtures which must guide the respective cleaning tools through a secondary inlet opening 62. Positioning of the clamping fixture relative to the inlet mixer and of the sliding base relative to the clamping fixture is accomplished by a locating fixture 4 (see FIG. 5). Locating fixture 4 has a mount 112 which pilots into the clamping fixture between locating pins 106 and 106'. A safety cable 124 connects umbilical 118 to locating fixture 4. A pair of recesses 114 and 114' on locating fixture 4 receive the locating pins 106 and 106' respectively (see FIG. 4). The locating fixture is latched into this position by means of latches 110 and 110' which are mechanically linked to a latching cylinder 116. In response to low-pressure water received via umbilical 118, the latches 110 and 110' lock the locating fixture to the sliding base 104 (see FIG. 4) for sliding relative to clamping fixture 2. The nozzle cleaning and TBF cleaning fixtures (described in detail below) have the same latching mechanisms for interchangeable mounting on the clamping fixture. The locating fixture 4 provides the azimuthal and axial positioning of the clamping fixture 2 by means of a locating finger 120 manipulated by a finger cylinder 122 driven by low-pressure water. When finger cylinder 122 is retracted, the locating finger 120 is extended and vice versa. As best seen in FIG. 6, the clamping fixture is positioned so that locating finger 120 in its extended position extends into a secondary inlet aperture 62 of the inlet mixer 46. The locating finger 120 is then swung toward its retracted position, so that it bears against the inner surface of the throat section 54 (see FIG. 2A). As the locating finger continues to swing relative to the body of the locating fixture, the inner surface of the throat section blocks further movement, causing the locating finger to pull the clamping fixture 2 into contact with the inlet mixer. Clamp arms 100 are then clamped around the inlet mixer as shown in FIG. 4. Thereafter the locating finger continues to pull the sliding base 104 until a pair of locating stops (not shown) are contacted. Then sliding base 104 is locked in place. The locating fixture 4 is unlatched and lifted by a grapple hook 128 (see FIG. 6) which couples with a lifting eye 126 (see FIG. 5) on the locating fixture. To clean the inlet mixer nozzles 52, a nozzle cleaning tool/fixture assembly 6 is lowered by grapple hook 128 into position on the clamping fixture, as shown in FIG. 7. Assembly 6 comprises a nozzle cleaning tool 130 and a nozzle cleaning fixture 132. Grapple hook 128 is hooked into a lifting eye of an low-pressure coolant inlet ("LPCI") adapter 142. Adapter 142 is a rigid boomerang-shaped member that bypasses the LPCI coupling 140 of a BWR, which is an obstacle to correct positioning of nozzle cleaning fixture 132. Fixture 132 has guide slots 134 which guide tool 130 from an uppermost position (shown by dashed lines) to a cleaning position inside the inlet mixer (shown by solid lines) in FIG. 7. The guide slots are shaped to provide the precise path of travel to enable the nozzle cleaning tool 130 to enter the inlet mixer via a secondary inlet opening (62 in FIG. 2B). The nozzle cleaning tool 130 is used to clean the internal surfaces of the inlet mixer nozzles 52. Referring to FIGS. 8A and 8B, tool 130 has a cleaning head 144 which can travel along a circular orbit for the purpose of positioning the cleaning head under the nozzle 52 to be cleaned. Then the cleaning head 144 is raised in a direction parallel to the axis of that orbit to position the cleaning head inside the nozzle. Referring to FIG. 9A, the nozzle cleaning tool 130 has an indexing motor 149, located inside indexing drive 150, which drives rotation of a positioning arm 152 about an axis of bevel gear 154a by means of gears 155, drive shaft 156, universal joint 157 and bevel gear 154b. The indexing motor is electrically powered via umbilical 138 (see FIG. 7). A rotation sensor (resolver) 151 is mounted on the back of the indexing motor 149 to provide angular position feedback via umbilical 138. The UHP water is supplied via UHP hose 136 (see FIG. 7) to a UHP supply port 161 (see FIG. 8A). UHP water supply port 161 is connected to a swivel 162 (see FIG. 9A), which in turn supplies UHP water to one end of UHP tube 163. At its other end UHP tube 163 is connected to a swivel inside swivel housing 164. The UHP water then flows into UHP feed tube 160 (see FIG. 9B) via transfer tube 165 and channel 166. Positioning arm 152 carries a stationary nut 149 which is threadably coupled to a lead screw 148 having a cleaning head 144 mounted on the top end thereof. The axis of lead screw 148 is parallel to the axis of rotation of bevel gear 154a. Positioning arm 152 has a length such that the distance between the axis of rotation of bevel gear 154a and the lead screw axis is equal to the pitch radius of the circular array of nozzles 52. Thus, rotation of positioning arm 152 enables cleaning head 144 to be oriented under any one of nozzles 52. Referring to FIG. 9C, a second electrically powered motor 168, located inside a traveling housing 146 (see FIG. 8A), drives rotation of lead screw 148. A rotation sensor (resolver) 173 is mounted on the back of the lead screw drive motor 168 to provide cleaning head travel feedback via umbilical 138 (see FIG. 7). Rotation of lead screw 148 in either direction causes the cleaning head 144 to move up or down so that it can enter or exit an inlet mixer nozzle 52. Cleaning head 144 has a nozzle 158 (see FIG. 9B) which directs the UHP waterjet 159 (see FIG. 8C). The UHP water is supplied to nozzle 158 by way of stationary feed tube 160, which has at least one cross hole 167. The UHP water exits cross hole 167 and enters the internal volume of body 236 either directly or via translating UHP tube 238. Translating tube 238, which surrounds stationary UHP feed tube 160 and forms an annular space therebetween, is coupled to (by a threaded port joint) and in fluid communication with body 236, so that UHP water flows from feed tube 160 to body 236 via translating tube 238 when body 167 has been elevated to a point beyond the elevation of cross hole 167. A sliding seal 240 prevents leakage between tubes 160 and 238. A swivel joint 242 having two high-pressure seals 232 is in fluid communication with the internal volume of body 236 via a side port 234 and with the hollow shaft 230 (see FIG. 9B) of lead screw 148 via channel 244, completing the path of UHP water from feed tube 160 to nozzle 158. Swivel joint 242, UHP tube 238, body 236, motor 168 and resolver 173 are all located inside a traveling housing 146 which translates in unison with the lead screw. As lead screw 148 rotates, a UHP waterjet 159 exits nozzle 158 on cleaning head 144 (see FIG. 8C). The waterjet 159 scans a spiral path on the internal surface of the inlet mixer nozzle 52. To clean the throat (54), barrel (56) and flare (58) sections of the inlet mixer (see FIG. 2), a TBF cleaning tool/fixture assembly 8 is lowered by grapple hook 128 into position on the clamping fixture 2, as shown in FIG. 9. Assembly 8 comprises a TBF cleaning tool 170 and a TBF cleaning fixture 172. Fixture 172 hangs on the LPCI adapter 142 and is coupled to TBF fixture umbilical 192 which supplies low-pressure water for feeding the TBF cleaning tool 170 into the inlet mixer 46. Umbilical 192 is carried on grapple cable 129, which supports the LPCI adaptor 142. An umbilical assembly 194 comprises a TBF rotation motor 196 (with associated resolver) and an umbilical 198 connected to the TBF cleaning tool 170. A lifting eye 204 for a grapple hook is provided to lift the umbilical assembly in the event that drive sprocket motor 192 fails. Umbilical 198 includes a hose 200 for supplying UHP water to the TBF cleaning tool, a hose 201 for supplying low-pressure water to the centralizing arms and a rotation drive cable 202 for rotating the TBF cleaning tool (see FIG. 12A). (The rotation sensor cable is not visible in FIG. 12A.) Rotation drive cable 202 is driven by TBF rotation motor 196 (see FIG. 10). The mechanism for feeding TBF cleaning tool 170 into the inlet mixer 146 is built into the TBF cleaning fixture 172 and comprises a TBF rotation motor 204 which drives a drive sprocket 206 by way of a gearbox 208 and a drive chain 210 (see FIG. 10). A rotation sensor (resolver) 205 coupled to motor 204 provides feedback to the central computer. TBF tool umbilical 198 is encased in a semi-flexible jacket 212 that is designed to provide support so the UHP conduit can be rotated and pushed up and down the TBF fixture during installation and cleaning. A support roller 214 is mounted on a pivotable member 216 which is biased to urge jacket 212 into contact with drive sprocket 206. The jacket 212 has means for engaging the teeth on drive sprocket 206, whereby the jacket 212 and TBF cleaning tool 170 coupled thereto are displaced in response to rotation of drive sprocket 206. A pair of alignment slides 218 orient the TBF cleaning tool 170 at the proper angle for insertion into the inlet mixer. A guide roller 220 guides the TBF cleaning tool 170 into a secondary inlet opening 62. After insertion through a secondary inlet opening, TBF cleaning tool 170 is centered inside the inlet mixer by first and second pluralities of centralizing arms 178a and 178b, which respectively provide circumferentially distributed points of support at first and second elevations. In accordance with the preferred embodiment depicted in FIG. 11B, each plurality comprises three centralizing arms pivotably mounted to extend at equal angular intervals (i.e., 120 deg). Each centralizing arm has a roller at its terminus: rollers 224a for arms 178a and rollers 224b for arms 178b. The rollers 224a and 224b are arranged to roll axially over the internal surfaces of the inlet mixer as the TBF cleaning tool 170 is lowered in increments by the drive motor 204, thereby reducing friction between the cleaning tool and the inlet mixer. Referring to FIG. 13, arms 178a and 178b are pivotably mounted on a stationary housing 258. Arms 178b are extended by a low-pressure water-driven piston 264, which carries pins 256 for coupling with respective recesses on arms 178b. Piston 264 is connected to a piston 265 which slides in housing 258, thereby compressing a spring 260. Spring 260 urges a piston 254, which slides on threaded shaft 250, to a position whereat pins 256' on piston 254 couple with respective recesses on arms 178a to extend the latter. Arms 178a and 178b pivot independently until further rotation is prevented by abutment with an internal surface of the inlet mixer. Upon venting of the low-pressure water, the piston 264 is pushed back to its starting position by spring 262, thereby retracting arms 178b. Likewise, threaded shaft 250, connected to piston 264, retreats and nut 252 overcomes the resistance of spring 260 to return piston 254 to its starting position, thereby also retracting arms 178a. TBF cleaning tool 170 has one (or more) UHP water-jet nozzle 174 incorporated in the end of a rotor arm 176. The rotor arm 176 is pivotably mounted on a rotating swivel housing 222, which is rotated by rotation drive cable 202. A rotation sensor (resolver) 219 coupled to swivel housing 222 provides feedback to the central computer. Since the UHP waterjet in the TBF cleaning tool exits from an offset position relative to the swivel housing axis, the thrust of the jet acts to assist in rotation of the swivel housing. The jet thrust is sufficiently high that the TBF rotation drive cable 202 acts to slow down or maintain the desired rotation speed. At each incremental axial position of the waterjet nozzle 174, the swivel housing/rotor arm assembly is rotated 360.degree. . By repeating this sequence of incremental axial advancement and 360.degree. rotation, the internal surfaces of the throat, barrel and flare sections of the inlet mixer can be cleaned by the UHP waterjet exiting nozzle 174. Rotor arm 176 is pivotable in an azimuthal plane relative to swivel housing 222. When the TBF cleaning nozzle 174 is being used to clean the flare section 58 (see FIG. 2) of the inlet mixer, it is desirable that the angle of inclination of rotor arm 176 be varied in dependence on the flare section radius to ensure that nozzle 174 will be maintained in proximity to the internal surface to be cleaned. This is accomplished by mechanically linking centralizing arms 178b to rotor arm 176, as described below. Referring to FIG. 12A, the TBF cleaning tool comprises a rotating sleeve 223 which is mounted on the piston housing 258 via bearings (not shown). Swivel housing 222 and rotating sleeve 223 rotate in unison, whereas swivel housing 222 is displaceable relative to rotating sleeve 223. Swivel housing 222 is coupled to piston 264 (see FIG. 13) so that the swivel housing displaces toward rotating sleeve 223 as the piston is driven to extend the centralizing arms. Rotor arm 176 is pivotable about a pivot 221 mounted on swivel housing 222. The end of rotor arm 176 remote from nozzle 174 is coupled to rotating sleeve 223 by a mechanical linkage 225. Thus, as swivel housing 222 displaces in unison with piston 264, the rotating sleeve maintains one end of mechanical linkage 225 stationary. Thus, the other end of mechanical linkage 225 is displaced relative to pivot 221, causing rotor arm 176 to pivot as a function of piston displacement, e.g., from angular position B to angular position A. The rotor arm 176 and centralizing arms 178a and 178b are disposed in retracted positions (as shown by dashed lines in FIG. 10) to facilitate insertion of the TBF cleaning tool into the inlet mixer. The preferred embodiments have been described in detail for the purpose of illustration only. Variations and modifications of the disclosed embodiments will be apparent to any skilled mechanical engineer. For example, it will be apparent that the number of centralizing arms in each plurality can be more than three. Also the TBF cleaning tool could be provided with more than one rotor arm. Further, motors driven electrically can be replaced by motors driven by low-pressure water. All such variations and modifications are intended to be encompassed by the claims appended hereto.
	abstract	A process of an extreme ultraviolet lithography is disclosed. The process includes receiving an extreme ultraviolet (EUV) mask, an EUV radiation source and an illuminator. The process also includes exposing the EUV mask by a radiation, originating from the EUV radiation source and directed by the illuminator, with a less-than-three-degree chief ray angle of incidence at the object side (CRAO). The process further includes removing most of the non-diffracted light and collecting and directing the diffracted light and the not removed non-diffracted light by a projection optics box (POB) to expose a target.
060552883	summary	BACKGROUND OF THE INVENTION The invention relates to a nuclear reactor vessel and more particularly to a nuclear reactor adapted to reduce the rate of degradation of structural members in its core region. A pressurized water nuclear reactor vessel in a commercial electric power generating plant recirculates an aqueous solution generally known as the "primary coolant" (principally containing small variable amounts of boric acid and lithium hydroxide and substantially saturated with hydrogen) through the reactor core region of the vessel at temperatures of up to about 500.degree. F. or more and at pressures of up to about 2250 psi or more. The principal function of the primary water is to transfer heat generated by fuel assemblies in the reactor core region to one or more near-by heat exchangers for generating steam to drive turbines and thereby generating electric power. In a commercial boiling water reactor, the steam is generated in the reactor vessel itself. In other reactor designs, the core region may be cooled by liquid sodium or by gaseous coolants instead of aqueous solutions. The primary coolant also cools the internal structural members in the high temperature irradiated core region of the reactor vessels containing the fuel assemblies. Recent ultrasonic inspections in commercial pressurized water reactors outside of the United States have indicated that the baffle/former bolts which conventionally fasten internal baffle plates (which support the fuel assemblies and guide the primary coolant through the core region) with internal former plates (which maintain the baffle plates in place in the reactor vessel cores) may be susceptible to cracking. Contemporaneous inspections of "up flow" pressurized water reactors having water (i.e., primary coolant) cooled joint designs including vertically oriented flow holes machined through the former plates, which holes interconnect with the bolt holes like those shown in U.S. Pat. No. 4,069,102, did not indicate any degradation of their baffle/former bolts. Periodic inspections of existing baffle/former bolts in commercial "down flow" commercial reactors in the United States have not indicated significant degradation which would require the nuclear power industry to replace the bolts or otherwise backfit such reactors. Now, the nuclear power industry is considering the desirability of backfitting operating reactors in order to better cool the baffle/former bolts and thereby to provide an additional temperature margin in order to reduce their susceptibility to cracking. However, a backfit of an existing nuclear reactor vessel along the lines of the teaching of U.S. Pat. No. 4,069,102 will be very difficult because the former plates are physically located behind the baffle plates and therefore are inaccessible. Such a backfit will require either that radially directed flow holes be machined into the bolt holes of former plates from the baffle side while the vessel internals remain assembled in a submerged reactor vessel, or alternatively that the internals be removed from the reactor vessel, disassembled, and the former plates then machined remotely from the reactor vessel. Both of these backfit methods are very difficult, costly, time-consuming processes. In addition, the cost of providing replacement power will be high. Alternatively, replacement of the existing reactor internals would be very costly, although not as time-consuming. U.S. Pat. No. 4,069,102 also proposes a bolt design having an internal bore extending down the centerline of the bolt shaft for cooling the bolt. However, such bores make the bolts difficult to ultrasonically inspect and may significantly affect the strength of the bolt. SUMMARY OF THE INVENTION It is an object of the present invention to provide an improved reactor vessel design having an improved baffle/former joint which provides effective cooling of the baffle/former bolts. In addition, the inventor has found that the baffle plate/former plate joint design of U.S. Pat. No. 4,069,102 makes it possible under some operating conditions for impurities to buildup and precipitates and particulates to deposit in the annulus of the baffle plates under the bolt heads. The build-up of particulates and precipitates under the bolt heads may then cause the temperature of the bolt heads to rise and may even cover the crevices formed by the bolt heads with the baffle plates, which can accelerate degradation due to stress corrosion cracking or to crevice corrosion. Thus, it is a further object of the present invention to provide a design which effectively washes the undersurfaces of the bolt heads to reduce the local concentrations of dissolved and/or particulate chemicals or materials. With these objects in view, the present invention resides in a nuclear reactor vessel having a baffle-barrel assembly for supporting fuel assemblies in a core region and for guiding a fluid flowing through the core region when the vessel is in service during power operation. The baffle-barrel assembly includes: a baffle plate having a countersunk hole with a relatively large diameter and a smaller diameter baffle plate bolt hole extending from the countersunk hole; a former plate having a bolt hole aligned with the baffle plate hole; and a bolt fastening the two plates together. The bolt has a head portion with an undersurface disposed in the countersunk hole and a shank extending from the underside of the head portion into the aligned bolt holes. The bolt head portion defines at least in part a fluid flow passageway extending externally of the shank interconnecting the countersunk hole with the baffle plate bolt hole. Advantageously, such a bolt can be readily ultrasonically tested and its strength is not compromised by a passageway through the shaft. Also, when the pressure vessel is in service, primary coolant fluid can flow through the short passageway and thereby cool the bolt head. Importantly, the coolant fluid also washes the region under the bolt head so that precipitates will not be deposited and trapped at the crevice formed by the underside of the bolt head portion and the baffle plate as a result of the generation of steam or of the accumulation of other solid particulates from temporarily trapped fluids during startup, power or shutdown operations. In addition, existing nuclear reactor vessels may be readily backfitted in accordance with the present invention by: removing the nuclear reactor vessel from service; and then replacing one or more of the existing baffle plate/former plate bolts described above each with a baffle plate/former plate bolt having a head portion defining at least in part a fluid flow passageway external of the shaft. Then, the primary coolant flowing through the reactor vessel may wash the underside of the replacement bolt heads when the reactor vessel is returned to service.
047132116	abstract	A high temperature pebble bed nuclear reactor having a medium power capacity of 300 to 500 MW.sup.e is equipped with two different shut-down arrangements comprising reflector rods used exclusively for scram. The total shut-down reactivity of the reflector rods is proportioned in order to prevent the excessive cooling of the reactor core folowing scram. In this manner, the reactor core is rendered subcritical (in the event of accidents, for example, or at the beginning of any operating state) yet capable of returning to criticality at a reduced level of temperature and power output following the removal of heat. The use of all of the reflector rods for scram is effected only in the event of reactivity accidents. For all other scram incidents, only a portion of the reflector rods are used.
	summary
	claims	1. A method of X-ray examination of a body comprising:forming a flat fan vertical beam of X-ray radiation of low intensity by passing an X-radiation of low intensity from a radiation source through a collimator connected to said source with a telescopic bar;scanning with said fan flat vertical beam of X-radiation due to movement of said collimator and a receptor of X-radiation in a horizontal plane in relation to the body;receiving X-radiation transmitted by the body;converting X-radiation into visible light radiation which is further converted into electronic signals; andshaping and analyzing the electronic signals. 2. A method as in claim 1, wherein the movement of said flat vertical beam of X-radiation is provided by means of moving a collimator in said horizontal plane with a permanent ratio of moving, speeds of said collimator and said receptor of X-radiation. 3. A method as in claim 2, wherein movement of said collimator is provided by means of a step motor. 4. A method as in any of claims from 1 to 3 wherein movement of said receptor of X-radiation is provided by means of a step motor with synchronization of the movement of a collimator and said receptor of X-radiation being effected due to maintaining a pre-defined ratio of rotation frequencies of both step motors. 5. A method as in claim 1, wherein visible light radiation generated from X-radiation received at each scanning moment is directly converted into digital signals. 6. An apparatus for X-ray examination of a body, comprising:a carrier for positioning the body;an information processing device;a source of X-radiation of low intensity and a holder with positioned on said holder a vertical collimator and a receptor of X-radiation;said receptor being a vertical array of X-radiation detectors, each of said detectors comprising a first device for converting X-radiation transmitted by the body into visible light radiation and an adjoining second device for converting visible light radiation into an electronic signal;said apparatus being supplied by two guides, and said collimator and said receptor of X-radiation being autonomously movable along said guides; andthe source of X-radiation rotating around a vertical axis of said source, and the source of X-radiation being connected by a telescopic bar to the collimator. 7. An apparatus as in claim 6, wherein said collimator is made in the form of at least one pair of parallel plates. 8. An apparatus as in claim 6, wherein said two guides being first and second guides, said first guides for moving said receptor of X-radiation and said second guides for moving said collimator, and the guides for moving said receptor of X-radiation and the guides for moving se id collimator being positioned on the holder horizontally. 9. An apparatus as in claim 6, wherein the collimator is supplied with a drive mechanism having a step motor with at least one set of vertical metal plates. 10. An apparatus as in claim 6, wherein the receptor of X-radiation is supplied with a drive mechanism having a step motor, where movement of the receptor is synchronized with movement of the collimator by maintaining a pre-defined ratio of rotational speeds of said step motor. 11. An apparatus as in claim 6, wherein the holder is positioned horizontally, said holder moving parallel to itself in relation to a stationary fixed carrier for supporting; the body. 12. An apparatus as in claim 6, wherein the second device of said X-radiation detector is made for converting visible light radiation directly into a digital signal. 13. A method of X-ray examination of a body comprising:passing said body through a holder, said holder being defined by a first substantially vertical rack for securing a radiation receptor, a second substantially vertical rack for securing a collimator, and a member connected between said vertical racks, each of said vertical racks and said member having a long axis, each long axis being mutually coplanar;pre-shaping a stationary fan-flat vertical beam of X-radiation of low intensity with said collimator;receiving X-radiation transmitted by the body;converting said X-radiation into visible light radiation which is further converted into electronic signals;generating an analysis of the electronic signals; andsaid beam of X-radiation being positioned such that the horizontal plane traversing the bottom of the body cuts off said beam by 2–5 degrees. 14. A method as in claim 13, wherein the beam of X-radiation is shaped at a scattering angle in a vertical plane of 37–43 degrees. 15. A method as in claim 13, wherein visible light radiation generated from X-radiation received at any scanning moment is directly converted into digital signals. 16. An apparatus for X-ray examination of a body, comprising:a carrier for positioning the body;an information processing device;a source of X-radiation of low intensity and a holder with positioned thereon a vertical collimator and a receptor of X-radiation;said receptor being a vertical array of X-radiation detectors, each of said detectors comprising a first device for converting X-radiation transmitted by the body into visible light radiation and an adjoining second device for converting visible light radiation into an electronic signal;said holder being π-shaped and being positioned vertically;said π shaped holder comprising a first substantially vertical rack for securing said radiation receptor, a second substantially vertical rack for securing said collimator and a top member connected between said vertical racks, each of said vertical racks and said top member having a long axis, each long axis being mutually coplanar;the carrier for positioning the body being movable in a horizontal plane between the racks of the holder transversely to a plane of said holder, andsaid beam of X-radiation being positioned such that the horizontal plane traversing the bottom part of the body cuts off said beam by 2–5 degrees. 17. An apparatus as in claim 16, wherein the collimator is secured inside said second rack. 18. An apparatus as in claim 16, wherein the source of X-radiation is positioned in an outward part of said second rack. 19. An apparatus as in claim 18, wherein the source of X-radiation is positioned 20–50 percent higher than the level of the carrier. 20. An apparatus as in one of claims 16 and 17 to 19, wherein space between the source of X-radiation and a second rack is covered by an additional housing, made in the form of a pyramid with a base closely adjoining said rack and an angle at the top equal to a largest angle of scattering of said beam. 21. An apparatus as in claim 20, wherein said apparatus is provided with at least one additional collimator made in the form of a pair of parallel plates positioned vertically inside said additional housing. 22. An apparatus as in claim 16, wherein a receptor of X-radiation is comprised of at least two parts with the upper one or said parts making up 60–70 percent of the total height of said receptor of X-radiation and positioned at the angle of 4–6 degrees in relation to a vertical plane. 23. An apparatus as in claim 16, wherein the upper bar between said vertical racks is made in the form of four rods passing through respective holes at corners of the four flat rectangular plates positioned pair-wise at one third of the rod's length adjacent each end of said rods with ends of said rods being used for securing to said vertical racks. 24. An apparatus as in claim 16, wherein the second device of said X-radiation detector is made for converting visible light radiation directly to a digital signal.
062366998	summary	BACKGROUND OF THE INVENTION The present invention relates to nuclear power plants, and in particular, to on-line monitoring of control rod positions relative to regulatory requirements for short term, long-term, and transient insertion limits. Commercial nuclear power plants are subject to comprehensive regulatory compliance covering virtually every phase of nuclear reactor operation. Many of these regulatory constraints are manifested in the form of so-called "Technical Specifications", which are an integral part of the operating license for the power plant. Each vendor of a nuclear steam supply system (NSSS), achieves compliance with the technical specifications, by formulating and justifying operating procedures for approval by the regulatory authorities. In pressurized water nuclear power plants (PWR plants), one type of Technical Specification concerns the accumulated time during which control rods are present in the reactor core. As is well known, control rods serve two important functions. The extent of insertion directly affects the gross power level in the reactor core. Another function is to control the local distribution of power in the core, thereby avoiding high localized power peaking, relative to the average power generated in the core. The prolonged insertion of particular control rods in the core, especially during periods of relatively high power, can have two detrimental effects long term. First, the pattern of fuel consumption can be distorted to the extent that upon removal of these rods, new, previously unpredicted local power peaking or power oscillations may arise. Furthermore, control rods can prematurely lose effectiveness over time. It is also well known that individual control rods can be ganged together as an assembly for insertion and removal by a single drive mechanism, and that a plurality of assemblies, such as four or eight, can be controlled as a group for substantially simultaneous movement into and out of the core. Four or five groups are typically programmed for staggered insertion and withdrawal from the core (unless, of course, all groups are to be dropped simultaneously to trip, or "scram", the reactor). For purposes of the present disclosure, a cluster of control rods which are moved by a single drive mechanism, are referred to as a "control element assembly" (CEA), whereas a plurality of CEA's which are controlled for substantially simultaneous movement into and out of the reactor, are referred to as a "CEA group". According to one approach for compliance with Technical Specifications, plant Limiting Conditions of Operation (LCO) are established to impose operational constraints with regard to CEA rod group insertions and thereby assure that the design bases which underlie the Technical Specifications are not violated. These limitations are typically characterized in terms of restrictions imposed on CEA rod group insertions between the Long Term Steady State Insertion Limit (LTSSIL) and the Transient Insertion Limit (TIL). These restrictions are typically expressed in terms of either clock hours, or effective full power days (EFPD) of exposure. An EFPD is the exposure equivalent of 24 hours at the licensed full power operation of the reactor. In addition, restrictions are imposed upon exceeding the Short Term Steady State Insertion Limit (STSSIL) under certain conditions. In a PWR, all CEA's are typically out of (above) the core at full steady state power, and are inserted downwardly into the core to reduce power level. Typical examples of limiting conditions of operation are set forth in the following Table 1. TABLE 1 ROD GROUP APPLICA- OPERATIONAL LIMITATION BILITY (LCO) CRITERIA Regulating Insertion between STSSIL and Limit to 4 hours per 24 TIL hour interval Regulating Insertion between LTSSIL and Limit to 5 EFPD per 30 TIL EFPD interval Regulating Insertion between LTSSIL and Limit to 14 EFPD per TIL 365 EFPD interval Regulation Insertion beyond the STSSIL Take Prescribed Action with COLSS out of service within 1 hour Part Strength Insertion between LTSSIL and Limit to 7 EFPD per 30 TIL EFPD interval Part Strength Insertion between LTSSIL and Limit to 14 EFPD per TIL 365 EFPD interval These restrictions limit the duration (in terms of hours) that CEA rods can be positioned between the STSSIL and the TIL, and the amount of CEA exposure which can be accumulated (in terms of Effective Full Power Hours) for insertions between the LTSSIL and the TIL. The graph of FIG. 20 depicts typical operational regions bounded by these insertion limits. The LTSSIL is a position limit in which there is no restriction for CEA rod insertions which are above this position. However, CEA rod insertions below this position and bounded by the TIL are constrained to the limits of CEA exposure as noted in Table 1. The STSSIL is a position limit below (i.e., greater than) the LTSSIL in which further restrictions on insertion (time duration--as opposed to CEA exposure accumulations) are imposed on CEA rod insertions which are below this position and bounded by the TIL. These limits are noted in Table 1. The TIL is a position limit below the STSSIL which CEA rod insertions must not exceed. This limit is designed to allow for plant maneuvering using CEA insertions (as long as the CEA's do not go below this limit and as long as they maintain the CEA exposure and time limit durations for insertion as previously noted). Should CEA's be inserted below the TIL, the plant annunciator system normally outputs an alarm message and the operator must then take corrective action (such as - restore the GEA rods to within the prescribed limits within a defined time period; or reduce plant thermal power). It is conventional to identify groups of CEA's beginning with number 1 and proceeding, e.g., to number 5 according to the order in which they are withdrawn from the core in a zero power condition at which all CEA groups are fully inserted. The corollary is that in the initial condition where the reactor core is at full power, with all rods out (the most desirable operating condition), Group 5 is the first to be inserted, followed by 4, 3, etc. The Long-Term Steady State Insertion Limit is shown in FIG. 20 as a vertical line extending through range of 1.0-0.2 power fraction and (when projected) intersecting the Group 5 insertion representation bar at an insertion distance of 108 inch (274 cm), out of a total group rod length of 150 inches (381 cm). Because Group 4 and subsequent groups follow in staggered relationship, it is clear that whenever Group 5 is positioned in the core anywhere within the Steady State Insertion Limit, no other Groups are in the core. It is evident that Group 4 does not begin entering the core, until Group 5 is at the 60 inch (152 cm) withdrawal position (i.e., 90 inches (229 cm) of insertion). The Short Term Steady State Insertion Limit for Group 5 is also shown in FIG. 20 as a vertical line which has an upper limit at a power fraction of 0.75 and extends downward to 0.25, and intersects the Group 5 bar at the 60 inch (152 cm) withdrawal position. Thus, it can be appreciated from FIG. 20, that the Group 5 Short Term Steady State Insertion Limit, would not be accompanied by a Short Term Steady State Insertion Limit for any other Group. On the other hand, the Transient Insertion Limit allows for a variety of CEA insertion configurations including the fifth and fourth Groups fully inserted and the third Group inserted at the 60 inch (152 cm) withdrawal position. Not all configurations are permitted at every power level, however, i.e., the greater extent of Group insertion, the lower the permitted power level even during a transient. Thus, it may be appreciated that the LCO's impose concurrent limitations on insertion. For example, even if the CEA groups have not reached the limit of 5 EFPD per 30 EFPD interval, for insertion between the LTSSIL and the TIL, desirable repositioning of the Groups may be foreclosed by the further requirement that insertion between the STSSIL and the TIL must not exceed 4 hours per 24 hour interval. The foregoing operational requirements are presently maintained by manual surveillance. The inventor has concluded that this approach has the following shortfalls which are remedied by the present invention: 1. Manual monitoring is cumbersome and prone to human error. PA1 2. There is no automatic method to display and analyze the monitored data which, in turn, reduces the situational awareness for the operator of the existing accumulated CEA group exposures relative to the operational limits. PA1 3. There is no automatic early notification of approach to operational limits so that corrective action can be taken prior to exceeding an operational limit. PA1 4. There is no automatic alarm notification when the operational limits are exceeded so that corrective action may be immediately initiated. PA1 5. The resolution of the manually recorded data is coarse. PA1 6. Manual recording of accumulated EFPD and hours for CEA rod group exposures does not conveniently lend itself to monitoring a contiguous data interval or window. This may result in the selection of discrete monitoring intervals which are sequential. Such discrete monitoring intervals can lead to potential circumscribing of the intent of the operational limits. For example, the restriction of 5 EFPD per 30 EFPD will seemingly be satisfied by two sequential monitoring intervals in which 4.5 EFPD exposure occurs during the last 4.5 days of the first monitoring interval (of 30 EFPD) and in which 4.5 EFPD exposure occurs during the first 4.5 days of the following monitoring interval (of 30 EFPD). Each monitoring interval seemingly satisfies the restriction of 5 EFPD per 30 EFPD interval but, in fact, 9 contiguous EFPD of exposure have occurred. If the starting period of the first monitoring interval was advanced 5 EFPD, then the total EFPD exposure for the first monitoring interval would have been 9 EFPD (rather than 4.5 EFPD) which exceeds the operational limit. In this example, the operational limits were either complied with or violated depending upon the happenstance of when the start of a discrete monitoring interval was chosen. SUMMARY OF THE INVENTION According to the present invention, these deficiencies in conventional techniques are overcome by a method and apparatus, in which the incremental effective exposure for each CEA group is computed commensurate with core power, for each time increment at which each group is within the position range where an administrative limit is imposed. The increments of effective exposures for each group are accumulated, and the accumulated effective exposure for each group is compared with the administrative limit for each group. This comparison is then displayed to the reactor operator. The displaying of the comparison to the reactor operator, preferably provides for continuous monitoring, alarming, and reporting of accumulated group exposure, expressed in terms of hours and effective full power days relative to the established operational limits. Although the administrative limits are preferably LCO's, other administrative limits, whether or not based directly on the plant Technical Specifications, can provide the applicable limits. In a further preferred embodiment, the display provides graphical information utilizing a "rolling wheel" and "sliding bar" format. In a still further preferred embodiment, a display sectoring mode is included. In yet further embodiments, query and predictive modes of operation, pre-alarm notification upon approach to applicable limits, and alarm notification upon exceeding applicable limits, are also provided. In the predictive mode, the effect on the LCO's of a planned power maneuver is assessed. If insufficient EFPD margin is available, a projection is made as to when suitable margin will be regained to allow the maneuver to occur while maintaining compliance with the LCO's. In the pre-alarm feature, an early warning of an indication of approach to an LCO limit regarding accumulated EFPD is displayed, so that action can be taken to avoid an actual violation of the LCO. The invention is preferably implemented to receive as continuous inputs: the current plant power level; the CEA Group positions; and the operational status of the Core Operating Limit Supervisory System (COLSS). The COLSS determine automatically and on-line, the gross thermal power level of the core. One implementation of such as system is described in U.S. Pat. No. 3,752,735, issued Aug. 19, 1973, and U.S. Pat. No. 4,330,367 issued May 18, 1982, the disclosures of which are hereby incorporated by reference. An internal clock maintains an accurate time base so that plant EFPD may be calculated as a function of the current plant power level, accumulated time, and licensed full power level. Accumulated time (in terms of hours), is also maintained employing the internal clock. Utilizing the positions of the Regulating and Part Strength rod groups, and the internally calculated EFPD and accumulated time, the system continuously determines the exposure for these groups whenever they are inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit. The exposures are determined for contiguous monitoring intervals which are defined by the Limiting Conditions for Operations (see examples in Table 1). The computed exposures are then continuously compared with the operational criteria. In addition, the positions of the Regulating groups are continuously compared with the Short Term Steady State Insertion Limit whenever the applicable LCO's are exceeded (such as whenever the Core Operating Limit Supervisory System is out of service). For such occurrences, excursions beyond the Short Term Steady State Insertion Limit are annunciated and the time remaining to take corrective action, in accordance to the Technical Specifications for operations, is displayed. For cases in which the Limiting Conditions for Operation are not applicable (such as for a Reactor Power Cutback event) an inhibit signal prevents unwanted exposure accumulations or spurious alarm messaging. A reactor power cutback system of the type mentioned herein, is described in U.S. Pat. No. 4,075,059 issued Feb. 21, 1978, the disclosure of which is hereby incorporated by reference. It should be appreciated that the use of part strength CEA's is an option, and the implementation of the invention follows the same procedures for part strength CEA's as for regulating CEA's. As the term is used herein, regulating CEA's is meant to include all the Groups which are normally controlled for sequential insertion and removal, as depicted in FIG. 20, for the purpose of regulating power and/or power distribution during power generation in the reactor. The reactor may also have additional control rods which are not normally intended for regulating purposes, but which are available for rapid shutdown or extended zero power outages. These features of the invention provides significant advantages over conventional techniques. Automatic calculation and continuous display of accumulated time (hours) and accumulated Effective Full Power Days (EFPD) of CEA rod group exposure relative to the Limiting Conditions for Operation (LCO) for insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit, is provided. Real-time monitoring of plant power and CEA rod group positions allows automatic and continuous calculation and updating of Effective Full Power Days and rod group exposures. This simplifies the operator workload and provides timely information relative to monitoring compliance with operational limitations on CEA rod group insertions and assists with the planning of future CEA rod group maneuvers. Continuous comparison of Regulating rod group positions with the Short Term Steady State Insertion Limit under applicable conditions as noted within the LCO's (such as whenever COLSS is out of service), provides automatic notification of exceeding the Limiting Condition for Operations for Regulating rod groups. Graphical representation of accumulated time (hours) and Effective Full Power Days (EFPD) relative to the LCO's, utilizing unique "Rolling Wheel" and "Sliding Bar" display formats, is intuitive. The display formats provide the user with an easily understood representation of the accumulated time and accumulated EFPD exposure for CEA rod groups relative to the operational limits as defined by the LCO's. The display formats are designed to accommodate a contiguous monitoring interval in which old exposure data is continuously discarded (rolls off for the "Rolling Wheel" format or slides off for the "Sliding Bar" format) while new data is continuously added (rolls on for the "Rolling Wheel" format or slides on for the "Sliding Bar" format) for the monitoring intervals as defined by the LCO's. These graphical displays provide a spatial representation of accumulated rod group exposure for a contiguous monitoring interval which is readily understandable to the end user. The displays improve the situational awareness and comprehension of the existing accumulated rod group exposures and readily indicates when exposure margin can be regained. The Sector mode which is associated with the graphical displays allows users to define "sectors" within the "Rolling Wheel" and "Sliding Bar" displays to be expanded and thus examined at higher resolutions. The ability to "sector" to finer resolutions allows finer detail to be observed, for the interval of interest, than can normally be displayed on a Video Display Unit. The Predictor Mode of operation allows the effect of a planned CEA Rod Group maneuver on the LCO's to be assessed in advance of performing the actual maneuver. This minimizes the likelihood of exceeding the operational limits for CEA rod group exposure. If insufficient time (hours) or EFPD margin is available, the Predictor Mode projects when suitable margin will be regained to allow the maneuver to occur while maintaining compliance with the LCO's. Various "what if" scenarios can be investigated using the Predictor Mode. The Query Mode allows the user to recall historic information and to determine when a certain level of accumulated exposure (in terms of hours and/or EFPD) will "roll off (slide off)" and be regained as usable margin. This allows the user to review previously recorded information and to determine when accumulated exposure margin (expressed in terms of hours and/or EFPD) will be regained which serves as an advanced planning tool. In the event of a system outage, the Update mode allows the system to be recalibrated to the current operational conditions. In event of a system outage the Update mode allows the user to enter the appropriate time--power--and rod group exposure history for the outage interval in order to recalibrate the system to the current operational conditions. Thus, the system can account for outages and be immediately reinserted into service when the system is restored to operation. The Summary Report allows the user to observe the accumulated exposure and remaining exposure margin for all Regulating and part Strength rod groups on a single display. Provides the user with an overall assessment of the current accumulated exposure and remaining exposure margin utilizing a single convenient display page. This alleviates the necessity of searching through multiple display pages to obtain an overall assessment of the current operational status. The pre-alarm notification alerts the user to an impending approach to an established Limiting Condition for Operation. Advanced notification of an impending limit excursion provides the user with time to take corrective action before the limit is actually exceeded. Alarm notification alters the user to any excursion beyond an established LCO boundary. Such alarming alerts the operator that an operational limit has been exceeded to that he may take appropriate action as called forth within the Technical Specifications for plant operations. Time remaining for completion of corrective action is displayed whenever an Alarm is annunciated (via exceeding an LCO boundary). Display of such information provides the operator with a convenient assessment of the progress of corrective action(s) relative to requisite "Completion Times" as stated within the Technical Specifications for plant operation. In cases for which the Limiting Conditions for Operation are not applicable (such as for a Reactor Power Cutback event) an inhibit signal prevents unwanted exposure accumulations or spurious alarm messaging. A contiguous monitoring interval is maintained for calculating rod group exposure relative to the LCO's for insertions between the Long Term Steady State Insertion Limit and the Transient Insertion Limit, rather than sequential discrete intervals. A contiguous monitoring interval avoids potential ambiguity in determining compliance with the LCO's.
056404344	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a miniaturized nuclear utilizing improved pressure tube structural members. More particularly, the present invention relates to a new miniaturized nuclear reactor utilizing novel structural members that are used to support the loads and stresses of multiple nuclear reactor fuel channel pressure tubes in a confined area. 2. Description of the Prior Art Nuclear power plants traditionally have been designed for achieving long term, safe, and reliable performance. To assure safety, the plants incorporate systems and procedures representing a studied anticipation of emergency conditions. design approaches will have considered theories or premises which may include, for example, design redundancies which are challenged by updated rules of performance as operating experience with nuclear power progresses. Thus, investigators in this power field continuously are called upon to develop improved analytic models of operation exhibiting improved bounding of operational factors and to further achieve higher levels of safety in view of changing rules of safety related performance. Because of the necessarily extensive time interval involved in developing or constructing a new nuclear power facility, for example such an effort may encompass ten years or more, and further in view of the numerous nuclear power facilities now in operation, these investigators typically are called upon to meet new rule criteria by modification of longexisting facilities. Retrofitting procedures can be quite extensive, calling for revised electrical power supplies, major valving replacements, and the like. The nuclear industry has evolved a variety of reactor types. One such type finding substantial field use performs to produce steam for turbine drive within the reactor core itself and is referred to as a boiling water reactor (BWR). The reactor heated water of the BWR serves not only as working fluid, but also as a reaction moderator, and along with other parameters, its proper supply and application within the system necessarily has been the subject of safety requirements or rule generations by government regulatory agencies such as the Nuclear Regulatory Commission (NRC). Typically, the general structure of a BWR nuclear system will include an upstanding reactor vessel which incorporates a lower reactor core structure beneath which are control rod drives. Above the core are, in order, a steam separator assembly and a steam dryer assembly leading to a steam outlet, above the reactor is a shield wall and outwardly of that a drywell. A pressure suppression chamber (wetwell), being torroidal in shape, is located below and encircling the drywell. In more typical BWR installations, water coolant is heated in the reactor core to rise within the reactor vessel as a two-phase mixture of water and steam. This dual phase mixture then passes upwardly through the steam separator assembly and steam dryer structure to enter the steam line leading to a turbine. Following turbine drive, the steam is condensed to water and returned to the reactor by relatively large condensate and feedwater pumps of a feedwater system. The feedwater enters the downcomer region of the reactor, where it is mixed with the water returning from the steam separator and drying functions. The water in the downcomer region is circulated through the reactor core via the vertically oriented recirculation pumps which direct flow to the vertical jet pumps located between the core shroud and vessel wall (downcomer annulus). In typical fashion, two distinct recirculation loops with corresponding recirculation pumps are employed for this recirculation function. In the event of some form of breakage or excursion generating malfunction, referred to as a "loss-of-coolant accident" (LOCA), designers anticipate that the relatively higher temperature-higher pressure water within the reactor will commence to be lost. A variety of safety systems and procedures may then be invoked both for containment and for thermal control of this LOCA. For the latter, thermal control, safety designs recognize that, while loss of the water moderator terminates the core reaction to eliminate a possibility of a nuclear incident, the momentum of generated heat or the residual energy within the reactor will remain of such magnitude as to require a cooling control to avoid for example, core melt down. In general, the amount of water within the containment system is more than adequate for this purpose, for example that contained in the suppression pool, or additionally, the condensate storage tank. To apply this water coolant for the safety purpose, a variety of safety related techniques or "emergency core cooling systems" (ECCS) have been developed to accommodate the LOCA. For example, core spray (CS) systems and low pressure coolant injection (LPCI) installations have been evolved in a variety of configurations. The LPCI system incorporates, for example, four pumps which are activated by a safety system in the event of a coolant loss. Where the loss of coolant is of sufficient extent, and the vessel pressure remains high, for example in the event of a small pipe break then, an automatic safety system will function to depressurize the reactor vessel permitting the relatively lower pressure water supply pumps to operate to introduce water to the reactor. Because the recirculation system earlier described is ideally structured for this purpose, generally it is used by the LPCI system for water introduction under ECCS conditions. Safety designs heretofore have recognized, however, that a recirculation loop may be broken under a LOCA condition. Thus, the pumping of water into that loop under such a LOCA condition may have no effectiveness. Accordingly, the LPCI systems have been equipped with a recirculation loop selection feature termed "loop selection logic" to avoid such conditions. This safety control detects the broken recirculation loop and initiates a procedure injecting water into the redundant, intact recirculation loop by actuating appropriate LPCI injection valves. Experience with such LPCI loop selection features have shown them to be complex and difficult to test and maintain. Under more current rule-based requirements, the design must accommodate for such occurrences as valve failure and the like. However, to function more effectively under current rules, procedures for retrofitting existing facilities to update them are elaborate and quite expensive, implementation involving such activities as recabling, pump reconnection activities and the like. Thus, an approach has been sought by investigators which offers operators the opportunity to eliminate the requirement for a loop selection logic regimen and associated costs therewith while improving the reliability of the LPCI system. Numerous innovations for structural member for nuclear reactor pressure tubes have been provided in the prior an that are described as follows. Even though these innovations may be suitable for the specific individual purposes to which they address, they differ from the present invention as hereinafter contrasted. In U.S. Pat. No. 3,584,903 titled ROLLED CHANNEL JOINTS by inventor James David Prichard, a strong and leak-free hub assembly for use with the pressure tubes of a nuclear reactor is disclosed in which the hub includes a hard insert having at least one groove formed in it, the hardness of the insert being greater than the hardness of the tubular element with which it is joined. Typically, the hub is formed of stainless steel, the insert is formed of surface hardened stainless steel and the tubular element is a zirconium-niobium alloy. The insert has a hardness greater than the hardness of the tubular element. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system. In U.S. Pat. No. 4,555,361 titled METHOD OF REDUCING THE VOLUME OF SOLID RADIOACTIVE WASTE by inventor Leo P. Buckley et at., combustible, solid radioactive waste, such as paper, plastics, rubber, cloth and wood are reduced in volume to ash residue using pyrohydrolysis, a method which combines pyrolysis of the waste in a vessel at temperatures in the range of 500.degree. to 700.degree. C. and gasification of residual carbon with superheated steam. Pressures of 1.0 to 3.5 Mpa are used with steam flows in the range 4 to 50 grams/second/cubic meter so that carbon containing components of the waste are removed as gaseous decomposition products in the form of carbon monoxide and hydrogen leaving an ash residue. The present invention differs from the above described patent due to the features of a method of reducing the volume of ash produced whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance. In U.S. Pat. No. 4,627,069 titled JOULE MELTER FOR THE PROCESSING OF RADIOACTIVE by inventor Keith B. Harvey et at., the joule melter has an outer cylindrical electrode which forms the outer wall of the melt containment, an inner cylindrical electrode which protrudes upward in the containment and forms the outlet for the melt, thus, also determining the depth of the melt. A non-conducting sealing material forms a base plug between the electrodes. A cylindrical electrically conductive baffle is located between the electrodes and includes an opening which allows the melt to flow from near the outer electrode where the melt material is first inserted into the melter, to the inner electrode which is the outlet. In addition to the inner and outer electrodes, the baffle may be connected to a power supply to modify the currents flowing at each of the electrodes. The present invention differs from the above described patent due to the features of melting the radioactive waste whereas the present invention describes utilizing glass and other impurities which when reacted with the fuel pellets form a less radioactive substance. In U.S. Pat. No. 3,837,397 rifled TUBE BUNDLE ASSEMBLY by inventor Michael J. Pettigrew, a robe bundle assembly, for example, a heat exchanger tube bundle or a nuclear fuel element tube bundle, comprises a bundle of laterally spaced tubes, a frame around the outermost tubes, and a lattice of wire cables with their ends held against lateral displacement by the frame and the tubes in the lattice interstices. The cables are deflected round a portion of each tube to space the tubes from one another, and the cables are preferably tensioned against the frame for this purpose. The present invention differs from the above described patent due to the features of the cable matrix whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein. In U.S. Pat. No. 5,213,757 rifled METHOD FOR FIXING A SPRING PACKAGE TO A TOP NOZZLE IN A FUEL ASSEMBLY OF A NUCLEAR POWER REACTOR by inventor Lennart Ohman, a method of fixing a spring package to a top nozzle in a fuel assembly of a nuclear reactor wherein the fuel assembly comprises fuel rods, guide tubes and spacers arranged in a bundle between a top nozzle and a bottom nozzle wherein a T-shaped slot in milled out in a clamp which is welded to or forms an integral part of the top nozzle for receiving one end of the spring package, the end of the spring package is then inserted into the slot and the end is then fixed in the slot by means of a locking pin. The present invention differs from the above described patent due to the features of the bundle whereas the present invention describes utilizing a moderator comprising separate compartments within which the calandria tubes and fuel channel pressure tubes are contained. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein. In U.S. Pat. No. 5,213,755 titled LOW PRESSURE COOLANT INJECTION MODIFICATION FOR BOILING WATER REACTORS by inventor David M. Kelly et al., a conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite quantifies from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,788,033 titled CALANDRIA by inventor Luciano Veronesia calandria for use in conducting the hot coolant of a nuclear reactor transversely. The calandria includes an upper plate and a lower plate which support tubes. The plates and tubes are enclosed in a shell which extends above the upper plate and has a supporting flange. The lower plate has holes for transmitting coolant into the region between the plates. The shell has openings whose boundaries mate with the outlet nozzles of the reactor. The tubes are of stainless steel and are dimensioned so that they have mass, stiffness and strength such that they are not subject to failure by the transverse flow of the coolant even at a high velocity. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,788,032 rifled REACTOR WITH FLOW GUIDANCE IN THE UPPER INTERNALS by inventor Jacques Baujat et al., a nuclear reactor has a pressure resistant vertical vessel with inlet and outer pipes situated at the same horizontal level. It also includes internals having a barrel supporting the core and defining with the vessel a down flow path for the coolant from the inlet pipes towards a space under the core and upper internals defining a flow path for the coolant leaving the core, above the latter, and flowing towards the outlet pipes. The upper internals include dividing walls defining circumferentially distributed volumes located at the common level of the pipes and each over a different angular sector. Some volumes belong to the initial part of the down going coolant path and the others force part at least of the coolant leaving the core to follow a path which is successively directed upwardly then curving towards the outlet pipes. The invention is particularly suitable for use in pressurized water reactors. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,759,904 rifled PRESSURIZED WATER REACTOR HAVING IMPROVED CALANDRIA ASSEMBLY by inventor James E. Gillet et al., a calandria assembly is received within the pressure vessel of a nuclear reactor system, at an elevation corresponding to the level of the outlet nozzles of the vessel, and receives pressurized coolant traveling in an axial flow direction within the vessel and turns same to a radial direction for exit though the outlet nozzles. Hollow tubes mounted in parallel relationship at opposite ends to first and second plates of the calandria in conjunction with a cylindrical skin of cylindrical configuration joining the first and second plates of the calandria, present a redundant structure introducing the potential of thermal stresses, which are limited by selection of the pattern of flow holes in the lower plate and the provision of flexible annular weld joints of J-shaped configuration between the lower ends of the calandria tubes and the lower, second calandria plate. The present invention differs from the above described patent due to the features of the coolant system whereas the present invention describes utilizing a moderator comprising separate compartments having coolant systems flowing therein within which the calandria tubes and fuel channel pressure tubes are contained and are maximally cooled due to the novel features of the present invention. In addition, the separation of the calandria tubes and the fuel channel pressure tubes are accomplished by a novel support system explicitly described herein which achieve this cooling maximization. In U.S. Pat. No. 4,284,475 rifled WEAR SLEEVE FOR CONTROL ROD GUIDE TUBE by inventor Andrew J. Anthon, a wear sleeve for a guide robe in a nuclear fuel assembly, and a method of installing the sleeve. The sleeve is an elongated metal cylinder having an upper portion adapted to be suspended from the upper end of the guide tube, and a lower portion adapted to be permanently deformed into interference fit with the walls of the guide tube whereby the sleeve may be secured against vertical movement. The method of installing the sleeve includes the steps of suspending the sleeve from the upper end of the guide tube, then expanding a selected lower surface of the sleeve until the sleeve is permanently deformed, whereby an interference fit between the sleeve and robe is formed. The present invention differs from the above described patent due to the features of the joint assembly, the present invention utilizes a threaded joint connector to join a fuel channel pressure tube to the reactor system. Numerous innovations for structural member nuclear reactors utilizing nuclear reactor pressure tubes have been provided in the prior art that are adapted to be used. Even though these innovations may be suitable for the specific individual purposes to which they address, they would not be suitable for the purposes of the present invention as heretofore described. SUMMARY OF THE INVENTION The present invention describes new shapes of fuel tubes. The advantages are that the fuel tubes are stronger and less brittle, there is more surface contact area for heat exchange to take place and therefore, the tubes' new shape is more efficient. For purposes of miniaturization, the fuel tube can be made from any type of material-metal, metal alloys, ceramic, glass, fiberglass, carbon-graphite, epoxy and/or plastic composites or a combination of these materials with or without reinforcements. The surfaces could be enameled, coated, lined and/or cladded. The shapes would be most applicable to the miniaturization, but could be used in the larger scale reactors, USING HIGH ENRICHED URANIUM FUEL. In the present invention, novel fuel bundles for use in miniaturized reactor are described. The novel fuel tube design is to solve partially the problem of disposal of the spent fuel. The secondary benefit is the increased safety of operation of the reactor, in case of accidental meltdown. The present invention describes novel Support pads to hold the fuel tubes in place. The pads are of different shapes and sizes. Pads provide continuous support, intermediate support, are integral with Structural Member and are inserted inside the fuel tube. The pads are applicable to miniaturization. All surfaces and parts of the reactor and/or fuel tubes could be coated, cladded, enameled or lined. The rolled joint connection at the end of the fuel channel pressure tube (FCPT) was developed to facilitate removal and replacement of the fuel channel pressure tube (FCPT). This is important in design of miniaturized reactors and in maintenance of all integral pans contained therein. The advantages of the novel Structural Member Metal Tubes are as follows: 1. PROTECTION OF SURFACE BY LINING OR COATING The interior surface and other integral parts contained therein can be further protected by adding a protective coating or lining of the interior surface of the member metal tube, to prevent irradiation of the metal tube from the FCPT. The coating or lining of the interior surface should be of a material inert to irradiation to provide positive protection. The use of coating, lining, etc. is novel and can be implemented because of the novel design of the novel Structural Metal Member. The deflection and bending stresses inherent therein would be nominal with the present design of the invention. Therefore, the coating or lining would not develop cracks, peel or other structural and/or functional defects. In the prior art, routinely, the calandria tube would deflect and bend to the extent that some coating or lining could not have been used. This obvious disadvantage would be overcome by the present invention. 2. FUEL CHANNEL PRESSURE TUBE (FCPT) The present invention reduces friction for movement of expansion and/or rotation of the fuel channel pressure tubes as follows: A) FCPT made from ceramic or any Irradiation-Inert Material such as glass, fiberglass, carbon-graphite, epoxy, metal alloys, or plastic composites in accord with the following features: 1. Coat exterior surface of the FCPT to reduce friction around the FCPT. PA1 2. Provide steel bends where the FCPT comes in contact with intermediate support pads. PA1 3. Coat the steel bends. PA1 1. Coat exterior surface of the FCPT to reduce friction. PA1 2. Coat to prevent irradiation of support pads and spacers. PA1 3. Coat the ends of the FCPT to prevent irradiation of the tube extension at connection with FCPT. PA1 1. The support pads and separators could be made of metal and/or metal alloys. PA1 2. The metal should be coated at contact with the FCPT to reduce friction. PA1 3. The pads should be grooved or have depressions to allow for circulation and cooling. PA1 1. The support pads and separators should be made out of material ( ceramic, glass, etc.) are inert to irradiation from FCPT. PA1 2. The metal support pads and superstars should be coated or lined with material inert to irradiation from FCPT. PA1 3. The surface of support pad and separators in contact with FCPT should be coated to reduce friction. PA1 4. The pads should be grooved or have depressions to allow for circulation and cooling. PA1 A) providing first and second water flow paths from the source of water coolant to respective first and second recirculation loops; PA1 B) providing low pressure coolant injection pumps actuable or pumping water from the source through the first and second water flow paths; PA1 C) providing a valve arrangement actuable from a closed to an open condition for effecting flow within the first and second water flow path actuating the valve arrangement in response to the safety output to permit water coolant flow simultaneously in each first and second water flow path; actuating the low pressure coolant injection pumps in response to the safety output; and PA1 D) restricting the flow of the water coolant in each first and second water flow path to a predetermined fluid flow rate selected to deliver the predetermined quantity of water coolant to each respective first and second independent recirculation loops, said flow rate being selected as effective for carrying out the emergency cooling of the reactor core from one water flow path. B) FCPT Made From Metal Subject to Irradiation in accord with the following features: 3. NEW SUPPORT PADS AND SPACERS The support pads inserts are one-piece made full length (20) feet of FCPT to be inserted into the new structural metal member inside the tubes. The support pads once inserted to be fastened to the new structural metal member. The fastener(s) should prevent the sliding of the pad out of position. The spacers could be intermittently spaced and do not have to be the full length. They would be held in position by being attached to the full length support pad or attached to the new structural metal member. The support pads could be an integral part of the new structural metal member. The configuration where the web of member penetrates to the inside of the tube as depicted in the drawings. The part projecting part inside the tube to be shaped as support pad or as spacer depending on location. 4. SUPPORT PADS AND SEPARATOR FOR FCPT A) FCPT Made of Ceramic, etc. B) FCPT Made of Metal C) FUEL BUNDLES-SUPPORT PADS INSIDE THE FCPT The fuel bundles inside the FCPT rest directly on the bottom of the robe. The fuel bundles should rest on pads to protect the surface of the FCPT from abrasion, wear and tear. The abrasion is caused by the fuel bundles sliding during loading and unloading and due to the elongation of the FCPT and vibration, etc. The pads would have a shape of rails (two) full length of the FCPT secured at ends against moving out of position. The pads should be used with the FCPT made from metal, ceramic, glass, etc. The pads would also protect the ceramic and glass FCPT from Chipping and cracking. 5. FCPT MADE OUT OF GLASS Advantages of using glass for making the fuel channel pressure tubes. The use would increase safety and reduce radiation emission in case of a meltdown. The glass, during the extreme heat due to meltdown, would melt. The melted glass would encapsulate the fuel bundles and pellets. This would reduce radiation emission from the nuclear fuel and contamination of parts of the reactor. It would minimize the damage to the FCPT effected by meltdown and allow for repairs of the reactor by replacement of the FCPT affected by the meltdown. 6. THE NOVEL REACTOR UNIT The new reactor unit would house four, six, or eight FCPT within it, and be used as a reactor. The unit will be a self contained miniature reactor. The exterior shape is the reactor can be round square, rectangular triangular and polygonal and/or any combination thereof. The tubes as shown in FIG. 8 are novel calandria tubes resting on support pads. Inside the calandria tubes are fuel channel pressure tubes. The use of glass for FCPT would have the advantage of safety, and reduction of emission of radiation during a melt down. In case of meltdown, the metal reactor would be encased to prevent radiation passing to the exterior and placed inside a concrete vault similar to a transformer vault in case of malfunction and/or meltdown the radiation will be contained therein. When the fuel is used up (spent) it will be removed and replaced to provide continuous service of the miniature reactor during normal usage. The present invention of the novel nuclear reactor has support pads for Calandria Tubes. The support pads as shown in FIG. 7 could be used to support the calandria tubes. The load of the fuel bundles inside the FCPT would be transferred to the Calandria tubes and from the Calandria tubes to the new unit reactor. In addition, the pads could be in the shape of two rails on which the bundles could readily slide. The present invention describes novel spent fuel disposal. The fuel pellets of spent fuel could be encapsulated in melted glass for disposal. This could be done individually or in bundles. The glass encapsulated fuel would be encased in concrete blocks to be stacked up in storage. The blocks would be made from contaminated (material) concrete, ceramic, and recast into blocks. The most radioactive is the fuel having a protective shield from a low contaminated material made in shapes for easy shipping, handling and storage. It would be fully automated, requiring no handling by humans. The orderly fashion of disposal would require less space, be economical and would not represent danger to the surrounding area. When the fuel is spent, the fuel bundles are removed from the reactor. The spent pellets should be removed from the bundles. The pellets should intentionally undergo a meltdown, and in the process, some contaminant be added to prevent reprocessing the spent uranium into a bomb grade material (national security reasons) and the pellets should be encapsulated with a glass coating to reduce radiation emission. The pellets should be placed in a storage container. This container should be manufactured from radiation contaminated material. The container could be of metal and/or concrete. The size that could be handles for transport and to put on shelf or warehouse. The process described could be fatty automated and done by remote control. The advantages are that the radiation contaminated material would be utilized and the waste disposal will be done in an orderly and controlled manner. It would reduce the amount of waste, reduce the space to store, and would reduce the amount of radiation from the spent fuel. The controlled and orderly manner of handling and storage would increase safety and protect the environment. The present invention describes disposal of spent fuel being encased in melted down glass. Could be used to dispose of nuclear wastes. The product would be radioactive glass blocks that would have to be stored for safety. The glass blocks would be stable and would not be radioactive molecules leaking. This would be stable for a very, very long time. The present invention describes a novel, state of the art fuel bundle. The fuel bundle is approximately 20 inches long and 4 5/8 inches in diameter. The Cylindrical fuel pellets are approximately 3/4 inches long and 1/2 inch in diameter. The fuel element is a metal fuel. In the present invention, the pads are of different shapes and sizes. Pads provide continuous support, intermediate support, are integral with Structural Member and are inserted inside the tube. The pads are applicable to miniaturization. Another feature of the present invention, is all surfaces and parts could be coated, cladded, enameled, or lined as stated in the text of the first patent and this application. An additional feature of the present invention is the rolled joint connection at the end of the fuel channel pressure tube was developed to facilitate the removal and replacement of the fuel channel pressure tube. This is important in design of miniaturized reactors and in maintenance of all other sizes of reactors, and applies to the use of the fuel channel pressure tube made of all materials. The spent fuel after second use would be less radioactive. It would pose a lesser problem of storage and handling. One benefit of the present invention is in making use of a currently discarded material namely spent fuel in highly radioactive state. An additional benefit of the present invention is in reduction of storage volume of highly radioactive spent fuel. The spent fuel should be used as fuel for "heating" the hot water produced by the second use of fuel in the unminiaturized reactor, would be passed through a heat exchanger and returned to the reactor. The heated water from the heat exchanger could be used to heat apartment and/or office buildings and/or generate electricity and/or generate heat for green houses to produce. The heated water could be convened to steam and utilized as mechanical energy. The spent fuel after first use is still highly radioactive, but not sufficient for production of electricity. The spent fuel when used the second time is less radioactive, and the reactor would also operate at a lower pressure. Still another feature of the present invention is addressed to a structural member for nuclear reactor pressure tubes and method which provides effective insertion of water coolant within the recirculating loops of conventional boiling water reactors, but without resorting to complex loop selection logic. Through analysis by modeling and the like of the requirements of the a structural member for nuclear reactor pressure tubes in terms of time for complete coolant injection and in terms of the required quantity of injected fluid, flow rates of injection are derived and requisite quantities of coolant are determined and identified such that the a structural member for nuclear reactor pressure tubes process is controlled through the simple approach of utilizing flow rate controlling hydraulic resistance within coolant injection conduits. Those hydraulic resistances may be implemented with a conventional orifice, the size and shape of which determines desired flow rates or by the throttling of a valve within the injection conduit achieving the equivalent result. Under the process, cross tie conduits and associated cross tie valving otherwise used for recirculation loop selection for coolant injection are not activated, but merely remain in an open condition under the new method and system, necessary a structural member for nuclear reactor pressure tubes modifications are achieved without resort to the complicated system and instrumentation otherwise required for loop selection with a minimum of hardware perturbation, rewiring or repiping. As another feature, the invention provides a structural member for nuclear reactor pressure tubes having a low pressure coolant injection system for a nuclear power facility of a variety having a boiling water reactor, having a reactor core and normal operating pressure, first and second recirculation loops including respective first and second recirculation pumps and actuable discharge valves, a suppression pool water source, a condensate storage tank, and a safety system responsive to a loss-of-coolant accident to generate a safety output. The system includes first and second low pressure coolant injection pumps having suction inputs and discharge outputs and actuable to pump water. A supply conduit arrangement is provided for coupling the suction inputs of the first and second low pressure coolant injection pumps in fluid flow communication with the suppression pool. First and second coolant injection conduits are provided which are coupled with respective discharge outputs of the first and second low pressure coolant injection pumps and to respective first and second recirculation loops. First and second hydraulic resistance components within respective first and second coolant injection conduits are provided for restricting the flow of water coolant therein to a predetermined fluid rate selected to deliver a predetermined quantity of water coolant to each of the first and second recirculation loops, the flow rates being selected as effective for carrying out the emergency cooling of the reactor core from one coolant injection conduit. A control arrangement is provided which is responsive to the safety output for actuating the first and second low pressure coolant injection pumps. As another feature, the invention provides a method for injecting low pressure cooling water into the boiling water reactor of a nuclear power facility having a source of emergency core cooling water, first and second independent recirculation loops normally circulating water through the core of the reactor for steam generation and a safety system responsive to a loss-of-coolant accident to generate a safety output for effecting the supply of at least a predetermined quantity of water coolant to the reactor, comprising the steps of: As another feature, the invention provides a low pressure coolant injection system for a nuclear power facility of a variety having a boiling water reactor with a reactor core, and normal operating pressure, first and second recirculation loops including respective first and second recirculation pumps and actuable discharge valves, a suppression pool water source, a condensate storage tank, and a safety system responsive to a loss-of-coolant accident to generate a safety output. The system includes first and second low pressure coolant injection pumps having suction inputs and discharge outputs and actuable to pump water. A supply conduit arrangement is provided for coupling the suction inputs of the first and second low pressure coolant injection pumps in fluid flow communication with the suppression pool and further includes a cross fie conduit arrangement for selectively interconnecting the discharge outputs of the first and second low pressure coolant injection pumps. First and second coolant injection conduits are provided which are coupled with respective discharge outputs of the first and second low pressure coolant injection pumps and to respective first and second recirculation loops. First and second low pressure coolant injection valves are provided within respective first and second coolant injection conduits and are actuable between closed and open orientations. Further provided are first and second hydraulic resistance devices within respective first and second coolant injection conduits for restricting the flow of water coolant therein to a predetermined fluid rate selected to deliver a predetermined quantity of water to each of the first and second recirculation loops, the flow rate being selected as effective for carrying out the emergency cooling of the reactor core from one coolant injection conduit. A cross tie valve arrangement is provided within the cross tie conduit which is actuable between open and closed conditions for selectively directing the outputs of the first and second low pressure coolant injection pumps to one of the first and second recirculation loops through select first and second coolant injection conduits. A control arrangement is provided which is responsive to the safety output for actuating the first and second low pressure coolant injection pumps, the first and second low pressure coolant injection valves and retaining the cross tie arrangement in the open condition in the presence of the safety output. The invention, accordingly, comprises the system and method possessing the construction, combination of elements, arrangement of parts and steps which are exemplified in the following description. Accordingly, it is an object of the present invention to provide a new structural member with metal fuel channel pressure tubes that reduce moment, reaction and deflection stresses at the ends of the metal pressure tubes. More particularly, it is an object of the present invention to provide a new structural member that will reduce the incidence of cracks developing in the metal of the fuel channel pressure tubes. The new structural members with ceramic fuel channel pressure tubes reduces moment, reaction and deflection stresses at the end of the ceramic pressure tube. The ceramic pressure tube is not affected by irradiation and growth of its diameter as the metal tube is. In keeping with these objects, and with others which will become apparent hereinafter, one feature of the present invention resides, briefly stated, in the ability to use ceramics instead of metal as the pressure tubes. When the structural member for nuclear reactor pressure tubes is designed in accordance with the present invention, stress of the pressure tube is greatly reduced, if not eliminated. In accordance with another feature of the present invention, the invention provides for the use of ceramic pressure tubes by providing full length support without deflection for ceramic brittle material. Another feature of the present invention is that the new structural member would be made to house four, six or eight, etc. pressure tubes within it. The new structural member would act as a Calandria for all the pressure tubes within. The advantage would be that the new structural member would act as a unit that would nave its own controls as to the flow of gas or heavy water. It could be taken out of service for maintenance or pressure tube replacement, while the reactor would remain in operation. Yet another feature of the present invention is the support pads which cradle the pressure tubes and prevent sideways movement of the tube. Accordingly, it is a general object of the present invention to provide the reduction of stresses in Calandria and pressure tubes. It is a more particular object of the present invention to provide continuous and intermittent support for the pressure tubes. An object of the present invention is to provide the prevention of cracks in the pressure tubes. A further object of the present invention is to eliminate deflection and sag in Calandria and pressure tubes. A still further object of the invention is to provide the use of materials for pressure tubes that withstand irradiation, high temperatures, etc (ceramic). A further object of the present invention is to allow for replacement of pressure tubes without shutting down the reactor. Accordingly, it is an object of the present invention to provide the End Plates of glass or metal will be formed with depressions to fit and accept the ends of the Fuel Elements. More particularly, it is an object of the present invention to provide a Hollow tube to be placed between the End Plates for the length of the fuel bundle. A rod or wire will be threaded through the tube and through a hole in the end plates. After the Fuel Elements will be in place in the End Plates the rod or wire will be reissued and anchored to hold the bundle together. In keeping with these objects, and with others which will become apparent hereinafter, one feature of the present invention resides, briefly stated, in the end of the tube or rod there will be a spacer plate. The spacer Plate will be in contact with the inside face of the End Plates. The stress of the rod or wire will hold the bundle together, but it will not put stress on the Glass Fuel Elements. When the fuel bundle is designed in accordance with the present invention, after the bundle is removed from the reactor, the end plates can be separated from the bundle and reused. The Fuel Elements with the spent fuel could be removed and sent to storage. Still another feature of the present invention is that The End Plate, made of glass or metal will be formed to leave cup-like indentions to fit to accept the ends of the Fuel Elements. Yet still another feature of the present invention is that The End Plate, made of glass, will be (welded) attached to the Fuel Elements by molten glass. Still yet another feature of the present invention is that the End Plate holding the fuel elements together will also be made of glass. Another feature of the present invention is that the Fuel Elements will be assembled into a bundle. Yet another feature of the present invention is that the Fuel Elements will be made of glass and filled with pellets. Still another feature of the present invention is that At Each end, a plate is welded to the Fuel Elements, holding them together as a bundle. Yet still another feature of the present invention is that Approximately Thirty-Seven of the Fuel Elements form a cylindrical Fuel Bundle. Still yet another feature of the present invention is that the Fuel Pellets are stocked end to end inside the cylindrical Fuel Element container and sealed. Another feature of the present invention is that the Fuel element is a metal Fuel Sheathing, a cylinder of approximately twenty inch length and 5/8 inch diameter. Yet another feature of the present invention is that Cylindrical Fuel pellets approximately 3/4 inches long and 1/2 inch in diameter. Still another feature of the present invention is that The Fuel Bundle is approximately twenty inches long and 4 5/8 inches in diameter. The novel features which are considered characteristic for the invention are set forth in the appended claims. The invention itself, however, both as to its construction and its method of operation, together with additional objects and advantages thereof, will be best understood from the following description of the specific embodiments when read and understood in connection with the accompanying drawing. BRIEF LIST OF REFERENCE NUMERALS UTILIZED IN THE DRAWING 10--miniaturized nuclear reactor utilizing improved pressure robe structural members 10 PA0 12--calandria tube 12 PA0 12A--calandria robe coating 12A PA0 12B--calandria robe lining 12B PA0 12C--calandria robe cladding 12C PA0 14--fuel channel pressure tube 14 PA0 14A--fuel channel pressure tube coating 14A PA0 14B--fuel channel pressure robe lining 14B PA0 14C--fuel channel pressure tube cladding 14C PA0 16--fuel bundle support pad 16 PA0 16A--fuel bundle support pad spacer 16A PA0 16B--fuel bundle support pad strap 16B PA0 17--fuel compartment pressure robe 17 PA0 18--fuel channel pressure tube pad 18 PA0 18A--fuel channel pressure tube pad vertical spacer 18A PA0 18B--fuel channel pressure tube pad end 18B PA0 18C--fuel channel pressure robe pad horizontal spacer 18C PA0 20--moderator 20 PA0 22--horizontal interior support pad 22 PA0 22A--horizontal interior support pad proximal end 22A PA0 22B--horizontal interior support pad distal end 22B PA0 22C--horizontal interior support pad groove 22C PA0 22D--horizontal interior support pad concave 22D PA0 22E--horizontal interior support pad coating 22E PA0 22F--horizontal interior support pad lining 22F PA0 22G--horizontal interior support pad cladding 22G PA0 24--vertical support pad 24 PA0 24A--vertical support pad proximal end 24A PA0 24B--vertical support pad distal end 24B PA0 24C--vertical support pad grove 24C PA0 24D--vertical support pad concave 24D PA0 26--fuel bundle 26 PA0 28--angular support pad 28 PA0 28A--angular support pad top member 28A PA0 28B--angular support pad bottom member 28B PA0 30--horizontal exterior support pad 30 PA0 30A--horizontal exterior support pad end 30A PA0 30B--horizontal exterior support pad fastener 30B PA0 30C--horizontal exterior support pad concave 30C PA0 40--fuel bundle 40 PA0 40AA--first fuel bundle proximal end plate 40AA PA0 40AAA--first fuel bundle proximal end plate fuel element end fastener 40AAA PA0 40AAB--first fuel bundle proximal end plate port 40AAB PA0 40AAC--first fuel bundle proximal end plate indent 40AAC PA0 40AAD--first fuel bundle proximal end plate opening 40AAD PA0 40BA--second fuel bundle distal end plate 40BA PA0 40BAA--second fuel bundle distal end plate fuel element end fastener 40BAA PA0 40BAB--second fuel bundle distal end plate port 40BAB PA0 40BAC--second fuel bundle distal end plate indent 40BAC PA0 40BAD--second fuel bundle distal end plate opening 40BAD PA0 40C--fuel element 40C PA0 40D--fuel bundle support 40D PA0 40DA--fuel bundle support proximal end 40DA PA0 40DB--fuel bundle support proximal end spacer 40DB PA0 40DC--fuel bundle support distal end 40DC PA0 40DD--fuel bundle support distal end spacer 40DD PA0 40DE--fuel bundle support spacer tube 40DE PA0 40DF--fuel bundle support rod 40DF PA0 40DG--fuel bundle support nut 40DG PA0 42--reactor wall 42 PA0 44--reactor wall interior horizontal 44 PA0 46--reactor wall interior vertical 46 PA0 48--joint connector 48 PA0 48A--joint connector 48A PA0 50--joiner ring 50 PA0 50A--joiner ring thread 50 PA0 52--service tube 52 PA0 112--second calandria tube 112 PA0 112A--second calandria tube compartments 112A PA0 113--second fuel channel pressure tube support pad 113 PA0 113A--second fuel channel pressure tube support pad end 113A PA0 113B--second fuel channel pressure tube support pad spacer 113B PA0 113C--second fuel channel pressure tube support pad concave 113C PA0 113D--second fuel channel pressure tube support pad convex 113D PA0 113E--second fuel channel pressure tube support pad groove 113E PA0 113F--second fuel channel pressure tube support pad opening 113F PA0 114--second fuel channel pressure tube 114 PA0 114A--second fuel channel pressure tube compartment 114A SECOND EMBODIMENT
053409965	claims	1. A radiation image recording apparatus in which a stimulable phosphor sheet provided with a layer of a stimulable phosphor is exposed to radiation, which carries information about an image, an the image is thereby stored on the stimulable phosphor sheet, said stimulable phosphor exhibiting such properties that, when it is exposed to radiation and is then exposed to stimulating rays, it emits light in proportion to the amount of energy stored thereon during its exposure to the radiation, wherein the improvement comprises the provision of a member, which is provided with a layer of a phosphor and which is located such that the layer of said phosphor is close to or in close contact with said stimulable phosphor sheet located at the position that is exposed to the radiation, said phosphor exhibiting such properties that, when it is exposed to the radiation, it produces fluorescence having wavelengths falling within the stimulation wavelength range of said stimulable phosphor. 2. A radiation image recording apparatus as defined in claim 1 wherein said stimulable phosphor is represented by the formula BaFX:Eu.sup.2+ wherein X is a halogen. 3. A radiation image recording apparatus as defined in claim 2 wherein said phosphor, which produces the fluorescence, is represented by the formula Gd.sub.2 O.sub.2 S:Tb.sup.3+. 4. A radiation image recording apparatus as defined in claim 2 wherein said phosphor, which produces the fluorescence, is represented by the formula Y.sub.2 O.sub.2 S:Tb.sup.3+. 5. A radiation image recording apparatus as defined in claim 1 wherein said radiation is X-rays. 6. A radiation image recording apparatus as defined in claim 1 wherein said stimulating rays are a laser beam.
	summary
	summary
048045152	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a method and apparatus for processing the signals generated by sensors monitoring selected parameters in a complex process such as a nuclear reactor. In particular, it relates to such a method and apparatus which utilizes a plurality of independent, digital, signal processors arranged in a number of redundant channel sets with each signal processor in each channel set generating one or more digital signals suitable for use in a process protection system and analog signals suitable for use in surveillance and control systems, but with related signals generated by different, independent signal processors to enhance system reliability. 2. Prior Art In a complex process, such as a nuclear power plant, numerous sensors are provided to measure various physical conditions in the process, such as for example, pressures, temperatures, flows, levels, radiation, and the state of various components, such as, the position of valves and whether a pump is operating or not. These measurements are generally used to perform three different functions: process control, surveillance and protection. Process control involves automatic or semi-automatic regulation of process conditions to achieve the desired result. Surveillance encompasses monitoring of process conditions to determine that the desired results are being achieved. Protection is concerned with automatic response to abnormal conditions in the process to prevent operating conditions from exceeding predetermined design limits and to take steps to mitigate the adverse effects of operation outside the design limits. In the case of a nuclear power plant in particular, the protection function is the most demanding of the three. In order to assure reliability of the protection system, redundant sets of critical sensors are provided. In order to improve the availability of the plant, correlation between the signals produced by the redundant sensors is made a prerequisite to initiation of the response to thereby reduce the probability of spurious interruption of normal operations. For instance, typically four redundant sets of sensors are provided, and an indication by at least two out of the four sensors is required to actuate the emergency or safety system. Some of the critical process conditions can be measured directly, such as pressurizer pressure in the case of a pressurized water reactor (PWR). Others are calculated from measured parameters, such as the departure from nucleant boiling ratio, (DNBR) in the PWR. In either case, the existing condition is compared with a preselected limiting value, and if the limit is exceeded, a digital signal is generated. These digital signals will be referred to as protection system actuation signals and include trip signals which are used to activate a system which shuts down or "trips" the reactor and engineered safeguard actuation signals which are used to initiate the operation of other plant emergency systems as is well known in the art. Since more than one such actuation signal is required to initiate the response, they are referred to as "partial trips" or "partial engineered safeguard actuation signals". In the typical prior art system, the sensor signals are grouped for processing in channel sets with each channel set including one sensor signal from each set of redundant sensor signals, although in instances where a particularly expensive sensor is required to generate a signal, such a signal may not be included in every channel set. As previously mentioned, a common arrangement is to provide four redundant sensors for most parameters, which therefore, are arranged in four channel sets for processing. In the prior art systems, each channel set includes a number of analog circuits each of which converts the applied sensor signal(s) to the appropriate range, calculates the desired parameter from the measured values where necessary, compares the resultant signal with a selected limit value and generates a protection system actuation signal when the limit is exceeded. Typically, the inputs to the analog circuits are provided with surge protection, electrical isolation and a buffer stage. The outputs of the analog circuits are bistables which provide a fail safe indication of a partial trip or engineered safeguard actuation signal by remaining active under normal conditions and by going inactive when the appropriate limit is exceeded. In the typical prior art protection system, the four partial trip and partial engineered safeguard actuation signals from each channel set for each parameter are applied to two redundant logic circuits which each perform the selected voting logic, such as two out of four as previously mentioned, on the partial protection system actuation signals. If two out of four of the corresponding partial actuation signals in either of the two logic circuits are inactive, appropriate emergency and safety control systems are actuated. An example of a prior art protection system is shown in commonly assigned U.S. Pat. No. 3,888,772. This system includes a semi-automatic tester for the voting logic which is described in commonly owned U.S. Pat. No. 3,892,954. To test the voting logic, the partial protection system actuation signals are removed from the voting logic for all of the actuation functions in one logic train and then an operator manually positions a selector switch so that preprogrammed test signals are rapidly and automatically applied to one logic module in the train being tested. Upon the completion of the test, the operator advances the selector switch to the next logic module. The duration of the test signals is so short that the actuation devices do not have time to react to the actuation signals generated and monitored by the tester, however, as an extra precaution, and to provide the capability of manually generating test signals, bypass breakers can be provided to avoid undesired actuation of the emergency and safety actions. A more recent form of an integrated protection system for a nuclear power plant is described in commonly owned U.S. Pat. No. 4,434,132 entitled "Power Supply with Nuclear Reactor". In this system, the redundant partial actuation signals generated by analog circuits are applied to four separate logic trains each of which performs the voting logic. The voting logic in each logic train or channel set is carried out in part by a microcomputer which exchanges information on partial actuations with a microcomputer in each of the other channel sets through fiber optic data links. If one of the sensors is out of service, its logic module in the assigned channel set can be bypassed singly, and the voting logic in the other channel sets for that actuation function only is changed by the associated microcomputer to two out of three. Each of the channel sets also contains a second microprocessor which monitors the status of the other channel sets and initiates bypassing of the entire channel set during testing of the individual actuation functions. A modification of this integrated protection system utilizing pulse logic is described in commonly owned U.S. patent application Ser. No. 546,604 filed on Oct. 17, 1983 and entitled "Pulsed Multichannel Protection System with Saturable Core Magnetic Logic Units". All of these prior art systems utilize analog circuits for generating the partial trip and partial engineered safeguard actuation signa1s wilh a11 the attendant shortcomings of such circuits including: size, cost, power consumption, heat generation, stability, limited life and inflexibility. SUMMARY OF THE INVENTION The present invention overcomes the shortcomings of the prior art by replacing the analog circuits with a plurality of independent, digital signal processors for generating the trip and engineered safeguard actuation signals in each channel set. At least some of the digital signal processors in each channel set generate more than one protection system actuation signal, however, the actuation functions are allocated between processors in a channel set so that related actuation functions are assigned to different processors. More specifically, some of the actuation functions are a primary indication of a particular event in the complex process and others are secondary indications. The primary indication provides rapid response to a particular abnormal condition while the secondary indication, which also initiates the required action, may be slower in response. For instance, in the event of a break in the primary coolant system of a PWR pressurizer pressure, which will drop instantly, is a primary indication of this emergency condition which requires immediate action. Reactor coolant temperature is a secondary indication of this event, but would not be effected as rapidly. According to the invention, the partial trip signals based upon pressurizer pressure and reactor coolant temperature are generated in separate independent signal processors so that failure of either signal processor leaves the other to initiate a reactor trip. Each of the trip signals is applied to a separate output line having its own protection grade isolation so that failures do not propogate from one part of the system to another. Each of the digital signal processors cyclically performs each of the assigned trip and engineered safeguard actuation functions utilizing the applied sensor signals. In order to continually check the operation of each individual processor, a tester/bypass unit in each channel set repetitively generates a ramp signal which is applied as an additional sensor input to each signal processor in the channel set. The signal processors all generate a test actuation signal when the ramp signal reaches a preset value. Failure of a processor to generate the test actuation signal in response to the ramp signal generates an alarm signal. The tester/bypass unit also, upon command, automatically individually tests each of the operational trip and engineered safeguard actuation functions in the channel set. This is accomplished by replacing the sensor signals with test signals selected to generate the trip or engineered safeguard actuation signal. The function is bypassed during this type of testing by continuously generating a normal output signal on the associated output line while the trip or engineered safeguard actuation signal is applied to the test/bypass unit for confirmation of function operability. The test is completed very rapidly and the remaining actuation functions in the channel set remain on line during the test. Upon completion of the test, the test signal is raised to an extreme value to verify that it has in fact been disconnected. As another aspect of the invention, the measured or calculated process parameter signals generated by the digital signal processors in generating the protection system actuation signals are outputed as analog signals for use in the reactor control and/or surveillance systems and by the plant computer. In order to eliminate the necessity for individual isolation circuits in the output lines for each one of these analog signals, the analog signals in each channel set are transmitted serially over a data link system to a common output device through a single isolation circuit. Sequencing of this data transfer is controlled by a common communications digital signal processor. As a practical matter, the digital signal processors in each channel set may be divided into groups each with its own communications digital signal processor for transmitting the parameter signals over a data link and through a common isolation circuit to the one common parameter signal output device.
044877412	abstract	Fuel assemblies of a nuclear reactor are transferred during fueling or refueling or the like by a crane. The work-engaging fixture of the crane picks up an assembly, removes it from this slot, transfers it to the deposit site and deposits it in its slot at the deposit site. The control for the crane includes a strain gauge connected to the crane line which raises and lowers the load. The strain gauge senses the load on the crane. The signal from the strain gauge is compared with setpoints; a high-level setpoint, a low-level setpoint and a slack-line setpoint. If the strain gauge signal exceeds the high-level setpoint, the line drive is disabled. This event may occur during raising of a fuel assembly which encounters resistance. The high-level setpoint may be overridden under proper precautions. The line drive is also disabled if the strain gauge signal is less than the low-level setpoint. This event occurs when a fuel assembly being deposited contacts the bottom of its slot or an obstruction in, or at the entry to the slot. To preclude lateral movement and possible damage to a fuel assembly suspended from the crane line, the traverse drive of the crane is disabled once the strain-gauge exceeds the low-level setpoint. The traverse drive can only be enabled after the strain-gauge signal is less than the slack-line set-point. This occurs when the line has been set in slack-line setting. When the line is tensioned after slack-line setting, the traverse drive remains enabled only if the line has been disconnected from the fuel assembly.
047541474	summary	BACKGROUND OF THE INVENTION (1) Field of the Invention The present invention relates to a collimator for a beam of radiation which allows rapid adjustment of the cross-section of the beam for different patients or other surfaces to be irradiated. In particular, the present invention uses bundles of adjustable clamped rods to shape the cross-section of the beam. (2) Prior Art Collimator devices for radiation are well known to the prior art. Most of them provide beams with a fixed cross-section. The problem is that this configuration is cumbersome where the cross-section of the beam has to be changed such as in patient treatment with the radiation in a defined area of the body. Variable collimator devices are also well known to the prior art. These devices provide an interfering member in the beam. U.S. Pat. No. 4,463,266 to Brahme describes a collimator which uses wedge shaped slabs mounted on an arcuate surface which moves into the beam so as to define the beam. This device is complicated and expensive to build because of the precision of the fit of the wedges relative to each other. Other prior art includes camera or iris type lenses which uniformly change the diameter beam and thus are not suitable where the beam is to have an irregular cross-section. One prior art example is U.S. Pat. No. 4,534,052 to Milcamps. None of this prior art provides a means for rapidly and simply adjusting the beam cross-section non-uniformly. OBJECTS It is therefore an object of the present invention to provide a variable radiation collimator which allows for rapid and non-uniform adjustment of the beam cross-section, particularly from patient to patient. Further it is an object of the present invention to provide a collimator which is inexpensive to construct and operate to vary the beam cross-section. These and other objects will become increasingly apparent by reference to the following description and the drawings.
	claims	1. A method comprising:operating a pressurized water reactor (PWR) wherein the operating includes circulating primary coolant in a pressure vessel upward through a nuclear reactor core that includes a plurality of fuel assemblies and a plurality of control rod drive assembly guide structures suspended above the plurality of fuel assemblies, wherein each fuel assembly includes a plurality of fuel rods containing a fissile material;during the operating, accommodating upward strain of the fuel assemblies and downward strain of the control rod drive assembly guide structures by a gap between the tops of the fuel assemblies and the bottoms of the suspended control rod drive assembly guide structures,wherein the control rod drive assembly guide structures are suspended in the pressure vessel from a support member disposed above the control rod drive assembly guide structures. 2. The method of claim 1, further comprising:during the operating, not applying a downward force other than gravity against the fuel assemblies. 3. A method of operating a pressurized water reactor including a pressure vessel having a plurality of fuel assemblies and a plurality of control rod guide assemblies suspended above the plurality of fuel assemblies, comprising the steps of:operating the pressurized water reactor;circulating primary coolant in the pressure vessel upward through the plurality of fuel assemblies;during the operating, accommodating upward strain of the fuel assemblies and downward strain of the control rod drive assembly guide structures by a gap between the tops of the fuel assemblies and the bottoms of the control rod drive assembly guide structures,wherein the control rod drive assembly guide structures disposed in the pressure vessel are suspended from a support element disposed above the control rod drive assembly guide structures. 4. The method of claim 3, further comprising:during the operating, not applying a downward force other than gravity against the fuel assemblies.
	claims	1. A cask cushioning body attached to both ends of a cask storing therein spent fuel and absorbing an impact applied to the cask, the cask cushioning body comprising:an end-surface side member in which a plurality of plates made of steel are formed at a distance between plate surfaces of the plates that face each other, and in which the plate surfaces of the plates are arranged along an end surface of the cask; anda circumferential-surface side member that forms a cylindrical body made of the steel, one end of which is connected to a periphery of the end-surface side member, and that is arranged along an end-portion outer-circumferential surface of the cask, whereinan impact absorber that absorbs the impact by deforming is provided outside of the end-surface side member and the circumferential-surface side member;the cask cushioning body is formed so that a size between an outer circumferential surface of a lid constituting the cask and an inner circumferential surface of the circumferential-surface side member is larger than a size between an outer circumferential surface of a main body of the cask to which the lid is fixed and the inner circumferential surface of the circumferential-surface side member. 2. The cask cushioning body according to claim 1, wherein the end-surface side member includes a plurality of end-surface reinforcing members provided to connect the plates to each other. 3. The cask cushioning body according to claim 1, wherein the end-surface side member includes at least one plate having an opening formed in a central portion thereof, andthe end-surface side member includes a plurality of central reinforcing members in a region between the plate with the opening and the other plates. 4. The cask cushioning body according to claim 3, wherein the impact absorber is inserted into a region surrounded by the central reinforcing members and the plate provided with the opening. 5. The cask cushioning body according to claim 1, whereinthe circumferential-surface side member includesa protruding portion in which peripheries of the plates in the end-surface side member protrude outward of the cylindrical body over an entire circumference of one end of the cylindrical body,a flange portion protruding outward over an entire circumference of other end of the cylindrical body, anda plurality of circumferential-surface reinforcing members connecting the protruding portion to the flange portion, and arranged on an outside surface of the cylindrical body. 6. The cask cushioning body according to claim 5, wherein the impact absorber is inserted into a region surrounded by the protruding portion, the flange portion, and the circumferential-surface reinforcing members. 7. The cask cushioning body according to claim 1, wherein a region of the end-surface side member in which the plates face each other covers a bolt for fixing the lid of the cask. 8. The cask cushioning body according to claim 7, the region of the end-surface side member in which the plates face each other covers an air-supply/exhaust, water-feed/drainage or pressure monitoring unit provided on the lid. 9. The cask cushioning body according to claim 1, wherein an impact absorbing material that absorbs the impact by deforming is filled in the end-surface side member between regions in which the plates face each other. 10. The cask cushioning body according to claim 1, whereinthe impact absorber is formed by a combination of a plurality of wood blocks, andthe impact absorber includesa first impact absorber group that is provided along a circumference of other end of the circumferential-surface side member, and that is made of a first material that absorbs the impact in a direction parallel to the end surface of the cask,a second impact absorber group that is provided around one end of the circumferential-surface side member, along the outer circumference of the end-surface side member, and adjacent to the first impact absorber group, and that is made of a second material that is lower in a compression strength than the first material and that absorbs the impact in the direction parallel to the end surface of the cask,a third impact absorber group that is provided in an external corner of the impact absorber along the outer circumference of the end-surface side member and adjacent to the second impact absorber group, and that is made of a third material that is lower in the compression strength than the second material and that absorbs the impact in a direction orthogonal to or inclined with respect to the end surface of the cask,a fourth impact absorber group that is provided along inner circumferences of the second and third impact absorber groups and adjacent to the second and third impact absorber groups, and that is made of the third material that absorbs the impact in the direction orthogonal to the end surface of the cask, anda fifth impact absorber group that is provided inside of a circumference of the fourth impact absorber group, and that is made of the third material that absorbs the impact in the direction parallel to the end surface of the cask. 11. The cask cushioning body according to claim 1, whereina mounting hole into which a bolt for fixing the cushioning body to the cask is inserted is provided in the impact absorber, andthe mounting hole is expandable and contractable in a depth direction of the mounting hole.
054250633	abstract	A process for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- for radiotracer synthesis is disclosed. The process includes producing [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- simultaneously by exposing a low-enriched (20%-30%) [.sup.18 O]H.sub.2 O target to proton irradiation, sequentially isolating the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- from the [.sup.18 O]H.sub.2 O target, and reducing the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- to [.sup.13 N]NH.sub.3. The [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [.sup.18 O]H.sub.2 O, and sequential elution of [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [ .sup.18 F]F.sup.- fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- from a single irradiation of a single low-enriched [.sup.18 O]H.sub.2 O target.
	description	The present invention relates generally to nuclear reactors and, more particularly, is concerned with reducing the pressure drop across the bottom nozzle of a nuclear fuel assembly. The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated and in heat exchange relationship with a secondary circuit for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump, and a system of pipes which are connected to the vessel form a loop of the primary side. For the purpose of illustration, FIG. 1 shows a simplified nuclear primary system, including a generally cylindrical reactor pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid reactor coolant, such as water, is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18 typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. An exemplary conventional reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertical, co-extending fuel assemblies 22, for the purposes of this description, the other vessel internal structures can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals' function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 2), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel 10 through one or more inlet nozzles 30, flows down through an annulus between the reactor vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through lower support plate 37 and lower core plate 36 upon which the fuel assemblies are seated and through and about the fuel assemblies. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces cause an upward force on the fuel assembly whose movement is restrained by the upper internals including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44. The upper internals 26 are supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate 40, primarily by a plurality of support columns 48. A support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plate 40. Rectilinearly moveable control rods 28, which typically include a drive shaft 50 and a spider assembly 52 of neutron poison rods (shown and described more fully with regard to FIG. 3), are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and the top of the upper core plate 40. The support column 48 arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally effect control rod insertion capability. FIG. 3 is an elevational view, represented in vertically shortened form, of a fuel assembly being generally designated by reference character 22. The fuel assembly 22 is the type used in a pressurized water reactor and has a structural skeleton which, at its lower end includes a bottom nozzle 58. The bottom nozzle 58 supports the fuel assembly 22 on lower core plate 36 in the core region of the nuclear reactor. In addition to the bottom nozzle 58, the structural skeleton of the fuel assembly 22 also includes a top nozzle 62 at its upper end and a number of guide tubes or thimbles 84 which align with guide tubes 54 in the upper internals. The guide tubes or thimbles 84 extend longitudinally between the bottom and top nozzles 58 and 62 and at opposite ends are rigidly attached thereto. The fuel assembly 22 further includes a plurality of transverse grids 64 axially spaced along and mounted to the guide thimbles 84 and an organized array of elongated fuel rods 66 transversely spaced and supported by the grids 64. Also, the assembly 22, as shown in FIG. 3, has an instrumentation tube 68 located in the center thereof that extends between and is captured by the bottom and top nozzles 58 and 62. With such an arrangement of parts, the fuel assembly 22 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 66 in the array thereof in the assembly 22 are held in spaced relationship with one another by the grids 64 spaced along the fuel assembly length. Each fuel rod 66 includes a plurality of nuclear fuel pellets 70 and is closed at its opposite ends by upper and lower end plugs 72 and 74. The pellets 70 are maintained in a stack by a plenum spring 76 disposed between the upper end plug 72 and the top of the pellet stack. The fuel pellets 70, composed of fissile material, are responsible for creating the reactive power of the reactor. The cladding which surrounds the pellets functions as a barrier to prevent fission byproducts from entering the coolant and contaminating the reactor system. To control the fission process, a number of control rods 78 are reciprocally moveable in the guide thimbles 84 located at predetermined positions in the fuel assembly 22. Specifically, a rod cluster control mechanism 80, positioned above the top nozzles 62 of selected fuel assemblies, supports a plurality of the control rods 78. The control mechanism has an internally threaded cylindrical hub member 82 with a plurality of radially extending flukes or arms 52 that form the spider previously noted with regard to FIG. 2. Each arm 52 is interconnected to a control rod 78 such that the control rod mechanism 80 is operable to move the control rods vertically in the guide thimbles 84 to thereby control the fission process in the corresponding fuel assembly 22, under the motive power of a control rod drive shaft 50 which is coupled to the control rod hub 80, all in a well-known manner. It is desirable to have a balanced flow across the reactor core, i.e., substantially the same pressure drop across each of the fuel assemblies, so that some of the fuel assemblies do not operate at higher temperatures than other fuel assemblies. Power output is limited by the hottest operating fuel element. Reducing pressure drop without comprising other beneficial features provides the fuel assembly designer the opportunity to add other features to make up for the reduced pressure drop, that can, for example, promote mixing which will enhance heat transfer that can translate into increased power output of the core. The bottom nozzles of the fuel assemblies, which include a horizontal top plate with a large number of flow through holes, is a significant contributor to that pressure drop. That is especially true for debris filter bottom nozzles which require that the holes be small enough to prevent the passage of debris that could damage the fuel rod cladding such as is described in U.S. Pat. No. 7,822,165, assigned to the Assignee of this application. Any modification that reduces the pressure drop across the fuel assemblies without otherwise adversely impacting the operation of the reactor core is desirable. Therefore, it is an object of this invention to reduce the pressure drop across the fuel assemblies by modifying the design of the bottom nozzles to alter the shape of the flow through holes. It is a further object of this invention to achieve that reduction in pressure drop by gradually changing the lateral flow area on either or both the upstream and downstream sides of the perforated flow plate of the bottom nozzle. These and other objects are achieved by a nuclear fuel assembly having a plurality of elongated nuclear fuel rods with an extended axial length. At least a lower most grid supports the fuel rods in an organized array having unoccupied spaces defined therein adapted to allow flow of fluid coolant therethrough and past the fuel rods when the fuel assembly is installed in a nuclear reactor. A plurality of guide thimbles extend along the fuel rods through and supporting the grid. A bottom nozzle is disposed below the lower most grid, below lower ends of the fuel rods and supports the guide thimbles. The bottom nozzle has openings therethrough to allow the flow of fluid coolant into the fuel assembly. The bottom nozzle includes a substantially horizontal plate supported orthogonal to the axis of the fuel rods. The horizontal plate has an upper face directed substantially toward the lower most grid and a lower face on an underside of the horizontal plate with the openings extending therethrough for the flow of coolant. At least some of the openings in the lower face have a funnel-like first appendage respectively extending below the lower face, around at least some of the openings in the lower face with an opening at the first appendage's substantially lowest extent having a larger diameter than a diameter of the opening in the lower face. An internal wall of the first appendage substantially gradually decreases in diameter from the opening at the first appendages substantially lowest extent until the wall of the first appendage transitions to the opening in the lower face. In one preferred embodiment, a lip in the opening in at least some of the first appendage's substantially lowest extent has a scalloped contour and preferably, the scalloped lip has a plurality of spaced depressions, resembling the contour of an egg receptacle in an egg carton and more preferably all of the lip of the opening at the first appendage's substantially lowest extent has such a scalloped contour. In one embodiment, the nuclear fuel assembly includes a funnel-like second appendage that extends up from at least some of the openings in the upper face with an opening at the second appendage's substantially highest extent having a larger diameter than a diameter of the opening in the upper face. An internal wall of the second appendage substantially, gradually increases in diameter from the transition at the opening in the upper face to the second appendage's substantially highest extent. In this latter embodiment, a lip of the opening at the second appendage's substantially highest extent has a scalloped contour. Desirably, the second appendage is at least partially recessed within the opening in the upper face. In one embodiment the highest extent of the second appendage terminates below the lower ends of the fuel rods and desirably, the highest extent of the second appendage is smaller than the lowest extent of the first appendage. At least some of the openings in the bottom nozzle substantially align with the unoccupied spaces in the lower most grid. In general, the internal wall of the first appendage gradually decreases the lateral flow area axially through the first appendage as the first appendage transitions to the opening in the lower face. The internal wall of the second appendage gradually increases the lateral flow area axially through the second appendage as the second appendage transitions from the opening in the upper face to the unoccupied flow spaces defined within the organized array of fuel rods. The present invention relates to a bottom nozzle 58 for a fuel assembly which, in addition to supporting the fuel assembly 22 on the lower core plate 36, also contains features which function to reduce the pressure drop across the nozzle. This can be appreciated from FIG. 3. The bottom nozzle includes a support means, for example, the skirt 56 shown in FIG. 3. The support means, skirt 56 in this embodiment, includes a plurality of corner legs 60 for supporting the fuel assembly 22 on the lower core plate 36. A generally rectangular, planar plate 86 is suitably attached to the upper surface of the support skirt 56. In the nozzle plate 86 of this embodiment, a large number of relatively small holes are provided to accommodate the passage of coolant from below the plate 86 to and through the lower most grid 88. These holes may be small enough to trap debris to shield the fuel element cladding from damage as described in U.S. Pat. No. 7,822,165, though it should be appreciated that this invention can provide a benefit to most any type of flow through hole in a fuel assembly seeking to minimize pressure drop. This invention recognizes that a significant portion of the pressure drop associated with the bottom nozzle flow plate 86 is due to abrupt changes in flow area. This advanced bottom nozzle concept incorporates “egg-crate” type features on either or both the upstream and downstream sides of the bottom nozzle flow plate 86 to gradually change the lateral flow area in the flow through path through the flow plate 86. FIGS. 4 through 8 show a portion of one embodiment of a flow plate 86 incorporating the features claimed hereafter. FIG. 4 shows an isometric view of a portion of the flow plate 86 with parts of the flow through holes 90 broken away to observe the interior of the flow through holes. FIG. 5 is the isometric view illustrated in FIG. 4 with the fuel rod end plugs 74 shown in position above the flow plate 86. FIG. 6 is a side view of a portion of the flow plate shown in FIG. 5. FIG. 7 is a bottom plan view of the flow plate shown in FIG. 5; and FIG. 8 is a top plan view of the portion of the flow plate shown in FIG. 4. On the upstream side (i.e., from the underside of the flow plate 86), which can best be appreciated from FIG. 7, streamlined “egg-crate” protrusions 92 gradually reduce the lateral flow area to minimize form losses associated with the rapid contraction that the coolant flow must undergo as it enters the perforated flow plate at the entrance to the flow holes 90. These “egg-crate” protrusions 92 also eliminate high pressure pockets of recirculating flow below each fuel rod location. The protrusions 92 are funnel-like extensions of the openings of the flow through holes 90 with a lip 98 that surrounds an opening on the lower most extent of the protrusions 92 having depressions 94 that in one embodiment are approximately equally spaced around its circumference; though it should be appreciated that the depressions need not be equally spaced to obtain some reduction in pressure drop. The depressions in the lip 94 form a scalloped contour. Additionally, though the protrusions 92 extending on either side of the flow plate 86 are shown to be approximately the same height, the height may vary over the surface of the plate and still obtain a reduction in pressure drop. On the downstream side (i.e., between the flow plate 86 and the fuel rods), streamlined “egg-crate” protrusions 96 gradually increase the lateral flow area to minimize form losses associated with the rapid expansion and contraction that the coolant experiences in the transition from the flow plate 86 to the fuel rod bundle. Due to the close proximity of the fuel rod bottom end plugs 74, the downstream “egg-crate” protrusions are recessed in the upper face of the plate 86 to interface with the fuel rods 66. There are no changes to the axial elevations of the fuel rods. The development of advanced fabrication techniques such as additive manufacturing makes the manufacture of this design more convenient, though it should be appreciated that traditional manufacturing techniques can also be employed. Though the egg-crate protrusion design has been applied to the upper and lower surfaces of the flow through plate 86 it should be appreciated that either of these designs may be employed alone to obtain some reduction in pressure drop or together to minimize the pressure drop for maximum benefit. Furthermore, an additional reduction in pressure drop can be achieved employing the embodiment illustrated in FIGS. 9-13. This embodiment retains the streamlined flow passages unique to the foregoing embodiment, which has the flow through holes substantially aligned with the unoccupied spaces between the lowermost grid and the fuel rod, but adds an additional flow path substantially in-line with the fuel rods. The additional flow holes 100 are of a similar design to the other flow holes 90, but are positioned directly under the fuel rods, are preferably smaller in diameter and have a set of standoffs 102 supporting the fuel rods and allowing the coolant flow to exit the bottom nozzle. The standoff may be the peaks of the scalloped lips of the appendages and ensure that the fuel rods don't block the flow holes during operation. Because the additional flow holes 100 are directly under the fuel rods they provide a “no-line-of-sight” path for the flow which helps minimize debris from passing thru the bottom nozzle yet help reduce the overall loss coefficient of the bottom nozzle by providing an additional flow path. Testing of this added feature showed a significant improvement over the embodiment employing the appendages without the additional flow holes in-line with the fuel rods. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
061887498	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows diagrammatically an X-ray examination apparatus 1 according to the invention. The X-ray source 2 emits an X-ray beam 3 in order to irradiate an object 4. As a result of differences in the X-ray absorption in the object 4, for example a patient to be radiologically examined, an X-ray image is formed on an X-ray-sensitive surface 15 of the X-ray detector 5 which is arranged opposite the X-ray source. A highvoltage power supply unit 51 supplies the X-ray source 2 with an electric high voltage. The X-ray detector 5 of the present embodiment is an image intensifier pick-up chain which includes an X-ray image intensifier 16 for converting the X-ray image into a light image on an exit window 17, and also includes a video camera 18 for picking up the light image. The entrance screen 19 acts as an X-ray-sensitive surface of the X-ray image intensifier which converts incident X-rays into an electron beam which is imaged onto the exit window by means of an electron optical system 20. The incident electrons generate the light image on a phosphor layer 45 of the exit window 17. The video camera 18 is coupled to the X-ray image intensifier 16 by means of an optical coupling 22, for example a lens system or an optical fiber coupling. The video camera 18 derives an electronic image signal from the light image, said image signal being applied to a monitor 23 in order to visualize image information in the X-ray image. The electronic image signal may also be applied to an image processing unit 24 for further processing. Between the X-ray source 2 and the object 4 there is arranged the X-ray filter 6 for local attenuation of the X-ray beam. The X-ray absorptivity of individual filter elements 7 of the X-ray filter 6 is adjusted by means of an adjusting unit 50. The adjusting unit 50 is coupled to the high-voltage power supply unit 51 so that the X-ray filter 6 can be adjusted on the basis of the intensity of the X-ray beam 3 emitted by the X-ray source. FIG. 2 is a diagrammatic representation of the X-ray filter of the X-ray examination apparatus according to the invention. The X-ray filter includes a system of approximately parallel row ducts 11 which are filled with an X-ray transparent liquid 12. The X-ray filter also includes a system of approximately parallel column ducts 13 which are filled with an X-ray absorbing liquid 14. The row ducts extend approximately perpendicularly to the column ducts in the example shown. A suitable X-ray absorbing liquid is, for example a solution of a lead salt, for example lead nitrate, lead dithionate or lead perchlorate in demineralized water, or liquid mercury. A suitable X-ray transparent liquid is, for example an oil which mixes only poorly with water. The filter elements 7 in the form of capillary tubes are provided between the row ducts 11 and the column ducts 13 in such a manner that each time a filter element is connected to a row duct 11 by way of an end 30 and to a column duct 13 by way of its other end 31. More specifically, an individual capillary tube is connected, by way of a first valve 32 and via the relevant row duct 11, to a first pump 41 and, by way of a second valve 33 and the relevant column duct 13, to a second pump 42. Each of the capillary tubes is provided with a piston 34 which keeps the X-ray absorbing liquid separated from the X-ray transparent liquid. The capillary tubes have a cross-section with a dimension of approximately 1 mm. The pistons in the example shown in FIG. 2 are formed by small balls, but other bodies can also be used as pistons. The pistons accurately fit in the relevant capillary tubes so that leakage of X-ray transparent and X-ray absorbing liquid between the piston and the wall of the capillary is avoided. The pistons are made, for example of an X-ray transparent material such as glass, anorganic oxides such as aluminium oxide (Al.sub.2 O.sub.3) and silicon dioxide SiO.sub.2 or polymers such as polycarbonate. In order to achieve suitable sealing for the liquids and/or a suitable degree of friction, it is advantageous to provide the pistons with a coating layer of, for example aluminium oxide (Al.sub.2 O.sub.3) or polyimide. The row ducts 11 and the column ducts are connected to a pressure control system. The pressure control system includes the first pump 41, the row valves 32, via which the first pump 41 is connected to the individual row ducts 11, and the column valves 33, via which the second pump 42 is connected to the individual column ducts 13. Preferably, electronically controllable row and column valves are used. The pumps 41, 42 and the row and column valves 31, 32 are controlled by means of a control unit 43. To this end, the control unit 43 is connected, via bus connections 44, 45, to control inputs of the row and column valves. Furthermore, the control unit is connected to control inputs of the pumps 41 and 42. It is to be noted, however, that use can be made of a single pump instead of two separate pumps, but in that case the control unit 43 must ensure that the row valves 32 are closed when only the column ducts 13 are to be pressurized, and that the column valves 33 are closed when only the row ducts 11 are to be pressurized. The X-ray absorbing liquid and the X-ray transparent liquid in the individual row and column ducts can be pressurized by means of the pump(s), the control unit 43 and the row and column valves. The amount of X-ray absorbing liquid in the capillary tubes can be adjusted on the basis of the liquid pressure in the row and column ducts whereto the relevant capillary tube is connected. The pumps 41, 42 and the control unit 43 form part of the adjusting unit 50. Only a small amount of time is required to open the valves and to displace the pistons under the influence of the liquid pressure so as to adjust the X-ray filter. It has been found that the X-ray filter can be adjusted within 40-50 ms, or even within 10 ms, depending on the liquid pressure. The adjustment of the X-ray filter can be readily canceled by opening all valves of the ducts containing the X-ray transparent liquid, being the row ducts 11 in the example shown in FIG. 2. The capillary tubes extend approximately parallel to the X-ray beam. Using a 5 molar lead salt solution and capillary tubes having a length of from approximately 5 to 6 mm, a 100-fold attenuation of the X-ray beam can be achieved and the X-ray absorption of individual capillary tubes may deviate by a factor of 20. Cylindrical pistons can also be used instead of balls. Such cylindrical pistons offer slightly more friction with respect to the wall of the capillary tubes. Because of this friction, the pistons can remain in their respective positions until liquid pressure is applied. The row and column ducts can be comparatively simply formed in a plate of glass, quartz, silicon or a polymer by chemical etching. All references cited herein are incorporated herein, as well as the priority document European Patent Application 98200179.4 filed Jan. 23, 1998, by reference in their entirety and for all purposes to the same extent as if each individual publication or patent or patent application was specifically and individually indicated to be incorporated by reference in its entirety for all purposes.
	summary
	description	The upper end part of a fuel rod for a pressurized-water nuclear reactor, denoted generally by the reference 1, is illustrated in FIG. 1 and in FIG. 2. The rod 1 comprises, in particular, tubular cladding 2 made of zirconium alloy which contain fuel pellets 3 and which is closed at its upper end, represented in FIG. 1 by a plug 4 constituting the upper plug of the fuel rod. The plug 4 is generally made up of zirconium alloy and comprises a part 4a which is engaged, virtually without clearance, in the end part of the bore 2. After filling the cladding 2 with fuel pellets 3, the closure plugs of the cladding, such as the upper plug 4, are engaged in the end parts of the cladding, the sealed attachment of the plug being provided by a weld line such as 5 made by melting the material of the plug and of the cladding, along a circular line. A spring for holding the fuel pellets is inserted between the upper plug 4 and the end of the column of fuel pellets 3. The upper plug 4, which is axisymmetric about an axis 6, comprises an end part 4b opposite, in the direction of the axis 6, to an end 4a engaged in a coaxial arrangement inside the cladding 2. The outside of the end part 4b of the plug 4 has, successively in the axial direction 5, from the upper end of the plug, a frustoconical plug, a shoulder and a cylindrical part along which the diameter of the plug is at a minimum. The minimum-diameter part of the plug and the shoulder make it possible to engage and to fasten the jaws of a gripper tool, during fuel rod maintenance operations. The plug 4 is traversed, in the direction of the axis 6, along a whole length, by a channel 7 which will be denoted subsequently as a filling channel, this channel 7 making it possible to fill the cladding of the rod with an inert pressurized gas, as will be explained below. The filling channel 7 of the plug 4 comprises several successive parts, the diameters of which generally decrease from the end 4a of the plug engaged in the cladding 2 to the outer end part 4b of the plug. At an upper end or outer end, the channel 7 comprises a terminal part 7a of frustoconical shape or a seal weld flared from the bottom upwards, i.e. from the inside to the outside of the plug, which emerges on the upper end surface 8 of the plug along a circular inlet opening, the center of which is on the axis 6 common to the plug and the cladding 2. The seal weld 7a of the channel 7, along which the channel 7 is closed by a weld, provides the junction between a part 7b of the channel 7 of minimum cross section and the circular inlet opening of the channel 7 on the upper end surface 8 of the plug 4. The small-diameter part 7b may have a diameter of about 0.6 mm and the circular inlet opening of the terminal frustoconical part 7a a diameter of about 1.3 mm. As illustrated in FIG. 2, after having filled the cladding of the rod with an inert pressurized gas, such as helium, the channel 7 is sealed shut in its upper terminal part, by a weld 10 obtained by melting the material of the plug in the central region of its upper end part, at the periphery of the seal weld 7a and consisting, for example, of three laser spot welds. In FIG. 3, the upper plug 4 of the fuel rod 2 is filled and sealed in a filling and welding apparatus 9. The filling and welding apparatus mainly comprises a enclosure, one wall of which comprises an opening for the passage of a fuel rod 2 in a horizontal arrangement, the upper end of the rod being inserted into the enclosure, so that the upper plug 4 of the fuel rod 2 engages with a valve 11, the closure member of which makes it possible to close or to open the terminal end of the filling channel of the plug. The enclosure of the filling and welding apparatus 9 comprises a part located opposite the part for introducing the rod 2, on the walls of which are mounted a laser beam welding device 12 and an optical assembly 13 making it possible to position the laser beam in order to weld the sealed closure of the filling channel of the upper plug 4 of the fuel rod 2 and which is also used to implement the inspection method by analyzing the optical image according to the invention. A mirror 14 for reflecting and focusing the laser welding beam emitted by the laser welding device 12 is placed inside the second part of the enclosure 9. The device 12 comprises an optical fiber 15 connected to a laser source and a collimator 16, so as to send, by the reflection and focusing mirror 14, a laser beam of axial direction, on the end part of the filling channel 7 constituting the seal weld. To allow a sealed closure of the filling channel by a weld 10 under satisfactory conditions, the laser beam may be directed accurately onto the center of the circular inlet opening of the filling channel 7 of the upper plug 4. In order to adjust the position of the laser welding beam, the optical illumination and adjustment assembly 13, which comprises an illumination device 17 and an optical sighting assembly comprising a reticule, is used so as to identify the direction along which the laser shot is sent. In the case of methods of adjusting the welding according to the prior art, adjusting the position of the upper plug of the rod is carried out manually and verified visually. In the case of the method according to the invention, the illumination device 17 for sighting and adjusting the laser shot is used in order to inspect, on digitized images, the position of the center of the inlet opening of the filling channel with respect to the laser beam and the compliance of the inlet opening, in order to determine whether the welding of the filling channel, after the rod is filled with pressurized helium, can be carried out satisfactorily. The weld, consisting of spot welds, is also inspected by analysis of digital images. For this purpose, a digital camera 18a connected to a microcomputer 20 comprising a screen for displaying images 19 is combined with an optic 18 directed in the axis of the enclosure 9 along which the fuel rod 2 is engaged. The microcomputer 20 comprises a video acquisition card and a digital input/output card enabling the microcomputer 20 to communicate with a controller managing the laser welding device 12. The microcomputer 20 receives inspection orders from the controller, via the digital input/output card, and verdicts established from the results of the inspection are sent by the microcomputer 20 to the controller for controlling the laser welding device 12. Information relating to the rod and the welding conditions is transmitted to the microcomputer 20 by the controller of the laser welding device 12. A first step of the method for inspecting the sealed closure by welding the filling channel of the upper plug of a rod is in determining the position and the size, i.e. the diameter of the inlet opening 8a of the filling channel 7, consisting of the circular outline along which the filling channel 7 opens on the end surface 8 of the upper plug 4 of the rod. This inlet opening constitutes the large diameter end of the chamfer of the terminal part 7a of the filling channel. The inspection is performed on the filling and welding apparatus as described, during a phase in the course of which the fuel rod is filled with pressurized helium, the valve 11 being open. An image is acquired of the end surface 8 of the plug and of the inlet opening 8a of the terminal part 7a of the filling channel using the optic 18 and the digital camera 18a, the optical image obtained being digitized by the digital camera 18a and transmitted in digital form to the inspection microcomputer 20. The image obtained may be made visible on the screen 19 of the microcomputer 20, as shown, for example, in FIGS. 6, 7 and 8. The search operations performed on the circular opening of the seal weld will be explained with respect to FIG. 4. As a result of illuminating the end of the upper plug of the rod in an axial direction, the end surface 8 of the rod appears as a light region and the terminal part 7a of the filling channel 7 as a dark region, the light region and the dark region being separated one from the other by a substantially circular line 8a constituting the inlet end of the seal weld. Analysis of a digital image of the end part of the upper plug allows determination of the position of the center of the circular opening 8a with respect to the center of a parameterizable sighting marker, marking the position of the laser beam with which the closure is sealed by welding the filling channel 7. Ideally, the position of the sighting marker and the position of the center of the inlet opening of the plug should be coincident. The sighting marker 21 comprises a vertical axis and a horizontal axis which define the center of the sighting marker. It is positioned visually when the laser welding station is correctly adjusted. The edges of the opening 8a are sought on the image along the horizontal and vertical axes of the reference sighting marker 21. The edges of the hole 8a are sought along the horizontal axis of the sighting marker 21 or first axis. A number N is chosen, which corresponds to the number of rows of the image which will be used on each side of the reference 21 in order to define a mean row along which the edges of the opening 8a will be sought. FIG. 5 illustrates the variations of the gray levels between the black and the white, along the mean row determined above. The distances on the X-axis, i.e. along the row of the image, are expressed in image points or pixels. A parameterized threshold value is chosen. The threshold value is illustrated by the straight line 22 of FIG. 5. The right edge of the image then the left edge are determined along the mean row and by comparing the pixels to the threshold value. From the right edge and from the left edge defined on the graph, the position of the center of the hole of the seal weld is determined, for example with respect to an edge of the image as illustrated by the segment 23 and the diameter of the seal weld as shown by the segment 24. The edges of the hole of the seal weld are then sought along a second axis consisting of the vertical axis 25 (see FIG. 4) passing through the center previously defined or the first center. The edges of the inlet opening of the seal weld are determined in the same way as above, but using the columns of the image instead of the rows. In this way, a second center of the circular inlet opening and the diameter along the vertical axis are defined. The edges of the circular inlet opening 8a of the seal weld are then determined, along the horizontal axis 26 or third axis passing through the second center defined by the search on the vertical axis. The position of the center of the circular opening 8a and the diameter of the circular opening along the horizontal axis are deduced therefrom, this third determination of the center of the opening supplying the actual center taken into account. The values obtained for the diameter along the vertical axis 25 and along the horizontal axis 26 are checked for consistency, that is to say that they do not differ by a value greater than a chosen threshold value. The pixels are then converted into millimeters and the diameter of the circular opening 8a is compared with threshold values defined by the xe2x80x9cminimum diameter and maximum diameterxe2x80x9d parameters of the opening 8a. The position of the center of the circular opening 8a defined by its distance to the center of the sighting marker 21 is compared with a threshold value corresponding to a xe2x80x9ctolerance of center positionxe2x80x9d parameter. The results are displayed on the screen of the microcomputer 20, as shown in FIGS. 6, 7 and 8. In each case, a first value is displayed in millimeters corresponding to the position of the center of the circular opening 8a, a second value is displayed in millimeters corresponding to the diameter of the circular opening 8a and a verdict indicating the compliance or the noncompliance of the measured values is displayed. Where the measured values comply, an execution command is transmitted to the controller of the laser beam welding device 12. Otherwise, a blocking command is transmitted. In the case of FIG. 6, the values identified comply, the distance between the center of the opening 8a and the center of the sighting marker is less than a chosen threshold value and the diameter measured being between the acceptable minimum diameter and the acceptable maximum diameter. In FIG. 7, the opening 8a is offset with respect to the center of the sighting marker 21, such that the distance between the center of the opening 8a and the center of the sighting marker is greater than the chosen threshold value. A verdict of noncompliance is therefore transmitted. On the other hand, the diameter in this case is between the maximum and minimum threshold values. In FIG. 8, the diameter measured on the image of the opening 8a is less than the minimum threshold value. A verdict of noncompliance is transmitted. In addition, the position of the center of the opening 8a is slightly offset with respect to the position of the sighting marker, the distance between the center of the opening 8a and the center of the sighting marker is however less than the threshold value. A fault verdict is also transmitted when it has not been possible to mark the edges of the opening 8a in the course of the searches as described above. In the case of a verdict of compliance, a welding execution command is sent to the controller of the laser welding device 12. The welding is performed by the pulsed laser which melts the metal of the upper plug in a peripheral region of the terminal part 7a or seal weld of the filling channel 7. A spot weld is produced closing off the inlet of the small-diameter part 7b of the filling channel 7, then two successive pulses are produced to form the weld 10. On its upper surface, the weld 10 is in the shape of a dish or crater due to the flow and to the deposition of the metal in the hollow central part of the plug. After welding, since the fuel weld is in position in the filling and welding apparatus 19, it is possible to inspect for the present and characteristics of the spot welds. For this purpose, the end surface of the rod 8 is illuminated, using the illumination apparatus 17, thereby sending light in an axial direction onto the surface of the rod, and images are acquired of the end surface 8 of the rod using the optic 8 and the digital camera 8a. The digitized image is transmitted to the microcomputer 20. The image of the end part of the plug can be displayed on the screen, as illustrated, for example, in FIGS. 11, 12 and 13. The search operations performed on the spot weld will be explained with respect to FIG. 9. On the image, the end surface 8 of the plug and the reflection 27 produced by the light reflected by the crater of the central part of the spot weld 10 appear as the light part. By analyzing the digitized image, the distance from the center of the reflection 27 to the center of the parameterizable sighting marker 21, which is shown on the screen, is determined. The position of the center of the reflection 27, that is to say the distance from this center to the center of the sighting marker, is compared with a xe2x80x9ctolerance of center positionxe2x80x9d parameter and the diameter of the reflection is compared with threshold values defined by xe2x80x9cminimum and maximum diameterxe2x80x9d parameters. The horizontal mean rows are constructed around the reference, then a maximum number of pixels greater than the threshold, and which be positioned in the reflection, are determined. The edges of the reflection 27 are then determined along a first axis consisting of the horizontal axis passing through the position found above. For this purpose, a mean row is defined, in the way indicated above and a graph is drawn representing the variation of the gray levels along the mean row, as shown in FIG. 10. A threshold value, represented by the horizontal straight line 28 in FIG. 10, is defined. The right edge and the left edge of the reflection are determined by comparing the successive pixels along the mean row with the threshold value. A first position of the center of the reflection is calculated with respect to an edge of the image, this position being represented by the segment 29 in FIG. 10. A first diameter of the reflection between the right and left edges is also calculated, this diameter being represented by the segment 30 in FIG. 10. The edges of the reflection along a second axis consisting of a vertical line passing through the first center defined above are determined by carrying out a search identical to the previous one but on the columns of the image. The consistency of the values obtained are verified for the diameter of the reflection 27 along the first and second axes. The pixels are converted into millimeters and the diameter is compared with the xe2x80x9cminimum and maximum diameterxe2x80x9d threshold values and the position, with the xe2x80x9ctolerance of center positionxe2x80x9d parameter. The results are displayed on the screen of the microcomputer, as shown in FIGS. 11, 12 and 13. In each case, one of the items of information displayed in digital form relates to the position and the other to the diameter of the reflection. The position of the reflection corresponds to the distance in millimeters between the center of the reflection and the center of the parameterizable sighting marker forming the reference. The position of the center of the reflection corresponds to the position of the center of the spot weld. In the case of FIG. 11, both the position and the diameter of the reflection are satisfactory. A verdict of compliance is then displayed. In the case of FIG. 12, the center of the reflection 27 is offset with respect to the center of the reference 21 by a value greater than the predetermined threshold value. A verdict of noncompliance is therefore transmitted. However, the diameter of the reflection is satisfactory, that is to say between the minimum admissible value and the maximum admissible value. In the case of FIG. 13 there is no reflection 27, which results in the absence of the spot weld in the seal weld of the filling channel 7 of the upper plug of the rod. A verdict of noncompliance is transmitted. In the case where a verdict of compliance is transmitted, the fuel rod may be considered as correctly welded and may be accepted at the end of manufacturing. The method according to the invention also makes it possible to determine the depth over which the weld has been carried out and, in particular, when only the first spot weld has been carried out, a very small reflection coming from the bottom of the seal weld is observed. In this case, the absence of the second and third spot welds is easily detected from the measurement of the diameter of the reflection. Examination of the graph giving the gray levels along a reference line passing through the reflection of the spot weld also makes it possible to determine the position of the center of the dish of the spot weld which is denoted by the reference 31 in FIG. 10. Analysis of the optical images produced by the microcomputer 20 is carried out using software. The invention makes it possible, in the apparatus for filling and sealing the rod, to inspect the diameter of the opening of the seal weld of the upper plug of the rod and its positioning with respect to the laser welding beam, so as to determine whether or not it is possible to carry out the sealed closure by welding the rod filled with pressurized helium. Secondly, the invention also makes it possible to verify that the spot weld has been carried out in a compliant manner. All the operations are carried out in parallel on the filling and welding apparatus. The method according to the invention makes it possible to avoid any operation of maintaining the rods between the filling and sealed closure apparatus and an inspection station and the verdict relating to the compliance of the spot weld is available from the end of the welding operation. Information relating to the operation as a whole (positioning, welding and inspection) may be saved on a hard disk and can be exploited subsequently in the form of databases. Finally, the illumination used to acquire images is a standard illumination which can be obtained by market available items. The invention is not limited strictly to the embodiments which have been described. It is possible to process the digitized images of the end surface of the upper plug before and after welding, by methods other than those which have been described, for example for determining the edges of the inlet opening of the filling channel and of the reflection of the spot weld. Finally, the method according to the invention is applicable to any nuclear fuel rod comprising an upper plug traversed by a channel for filling the rod with an inert pressurized gas.
045487839	abstract	A stopper apparatus for use in blocking the unvalved nozzle of a cooling fluid filled nuclear reactor pressure vessel includes a plug, typically in the shape of a frusto-conical member having inflatable gaskets for sealing against a seat of generally unknown surface characteristics and means for positioning and urging the plug into position to seal the nozzle. The plug is typically positioned by suspension cables whereby the apparatus can be temporarily inserted and removed from the pressure vessel. The urging means is generally a two-way hydraulically driven jack controlled by remotely-actuated hydarulic lines. The apparatus is a tool which permits temporary sealing of a submerged outlet in a reactor vessel to permit maintenance on a fluid recirculation loop.
	description	The disclosed and claimed concept relates generally to nuclear power generation equipment and, more particularly, to a detection apparatus and a method of use that are intended to evaluate the neutron absorption capability of control elements of BWR and PWR nuclear reactors. Various types of nuclear power generation systems are known to exist and are particularly known to include boiling water reactors (BWRs) and pressurized water reactors (PWRs). A BWR includes, among other components, a number of blades that are used as control elements that absorb neutrons and control the nuclear reaction within a reactor of a nuclear installation. As employed herein, the expression “a number of” and variations thereof shall refer broadly to any non-zero quantity, including a quantity of one. One such blade is depicted in an exemplary fashion in FIG. 1 at the numeral 6. Such blades and are configured to be of an elongated cruciform shape having a hub and four wings that protrude from the hub. The wings are structured to each be received between adjacent pairs of fuel assemblies to absorb neutrons in the water within which the fuel assemblies are situated. During normal operation of the BWR, the a substantial fraction of the blades typically are at least partially withdrawn from being situated between the fuel assemblies and can be, as needed, advanced further into a position situated between the adjacent pairs of fuel assemblies. The balance of the blades will be positioned partially to fully withdrawn as determined by the core reactivity management requirements as a function of the core depletion with exposure and the reactor thermal margin requirements. Such further insertion of the blades causes the blades into the reactor fuel region results in absorbing a greater number of neutrons, thereby slowing or controlling the nuclear reaction in a known fashion. The BWR control blades each include a plurality of elongated hollow passages that are filled with boron or other neutron absorbing substance and are capped. In one such type of blade design, the passages are oriented substantially perpendicular to the longitudinal extent of the hub and are formed by drilling holes in the edges of the wings. The holes are formed with boron or another neutron absorbing substance and are capped. In another such blade design, the blades are formed by providing elongated hollow tubes that each have an elongated passage that is filled with boron or another neutron absorbing and that is then capped at the ends, and then the tubes are laid side-by side and affixed together in any of a variety of fashions such as by covering the tubes with a sheet of metal to form the elongated cruciform shape. Over time, the passages in the blades that contain the boron or other neutron absorbing substance can develop defects in any of a variety of fashions. Such a defect will then allow the highly pressurized cooling water for the reactor into the space reserved for the absorbing material. The water intrusion into the neutron absorbing substance generally has the effect of washing away the absorbing material within the defective absorber tube, thereby leaving a region within the control bade partially or completely devoid of a neutron absorbing substance. The consequence is that the blade in the vicinity of such a breached passage does not absorb neutrons as required. This loss of absorption becomes significant from a safety and operational perspective when the defective portion of the control blade is large enough to either diminished the overall ability of the plurality of control blade insertions to extinguish the nuclear chain reaction within the reactor or results in unacceptable reactor power distributions and resulting loss of thermal margin. Over time, any given blade can have one or more locations thereon where the neutron absorbing substance is lacking due to such a defect, at which locations the blade does not absorb neutrons, and rather allows neutrons to pass directly therethrough. While a certain number of zones and/or a certain distribution of zones where neutrons pass directly through the blade can be deemed acceptable, the overall ability of the blade to absorb neutrons and to slow or otherwise control a nuclear reaction must not be compromised. As such, blades must be non-destructively examined once there are either visual or operational defect indications to assess their continued ability to effectively absorb neutrons and thereby meet the design and safety requirements. In one such prior inspection methodology, a neutron absorbing film is applied on one side of a wing and a neutron source is applied at the opposite side of the wing. Neutrons that are not absorbed and that rather pass through the wing at any given location thereon leave a telltale indication on the film at that location. A neutron source that has been used in such procedures employs californium 252, which is a neutron source that constantly emits neutrons and thus requires special shielding during transport and during certain portions of its use during an inspection procedure. After extensive exposure of the blades to the neutrons from the neutron source, the film would then be removed and examined, much in the way of X-ray film, for regions that include the telltale indication where neutrons had passed through the wing. The neutron absorption capability of the wing is then evaluated based upon an evaluation of the film. The use of such film has been cumbersome for a number of reasons, including the extended time frame to achieve meaningful film exposure, that is required to quantify the blade effectiveness, the special shielding of the californium 252 neutron source that is required, and the extensive time that transpires between initiating the procedure and obtaining the results of the procedure. Improvements would thus be desirable. An improved detection apparatus is usable to detect the neutron absorption capability of a control element of a nuclear installation and includes a neutron radiograph apparatus and a robot apparatus. The neutron radiograph apparatus includes an neutron emission source, a detector array, and a mask apparatus. The neutron emission source is advantageously switchable between an ON state and OFF state and employs an accelerator that emits a deuterium particle at either another deuterium particle or a tritium particle to result in the generation of helium isotopes and excess neutrons by a nuclear fusion reaction. The neutron emission source thus does not require shielding during transport and is energized to the ON state only after being submerged to a certain depth in the fuel inspection pool, which avoids any need for additional personnel or apparatus shielding. The neutron emission source is situated at one side of a wing and generates a neutron stream, the detector array is situated on an opposite side of a wing, and the neutron emission source and detector array are robotically advanced along the wing. The detector array is monitored in real time to determine whether the detector array emits an output signal that is representative of a detection of a defect by measuring the neutron stream passing through the blade. If the detector array does not measure excessive transmission of neutrons through the wing anywhere along the length of the wing, that wing is deemed to have passed the inspection and to be acceptable. On the other hand, if the detector array at a given location on the wing detects the transmission of neutrons through the wing, various masks of the mask apparatus are positioned between the neutron emission source and the detector array to more specifically identify the position on the blade where the neutrons are passing through. Each mask is in the form of a plate-like neutron absorber having a small opening formed therein. By employing one or more masks and manipulating the one or more masks with robotic manipulators, the detector array can determine the precise position(s) on each wing where neutrons are detected as passing through the wing. Such processing can occur rapidly and, when completed, the results can be evaluated to determine whether or not that wing is passed or whether it is in need of replacement or repair. Accordingly, an aspect of the disclosed and claimed concept is to provide an improved detection device that is usable to detect and evaluate the neutron absorption capability of a control element of a nuclear installation. Another aspect of the disclosed and claimed concept is to provide an improved method of operating such a detection apparatus in the detection and evaluation a neutron absorption capability of a control element of a nuclear installation. Another aspect of the disclosed and claimed concept is to provide such a device and method that enable the neutron absorption properties of a blade of a control element to be evaluated in real time. Another aspect of the disclosed and claimed concept is to provide such an improved detection apparatus and method wherein an neutron emission source is capable of being switched between an OFF state and an ON state and which therefore does not require shielding during transportation or during use and rather relies upon the water in a fuel inspection pool for any shielding that may be needed when the neutron emission source is in its ON state. Another aspect of the disclosed and claimed concept is to provide such an improved detection apparatus and method that allow for the rapid evaluation of the neutron absorption properties of a control element of a nuclear installation by providing a mask apparatus having a number of masks that can be used individually and in various combinations to enable a region of a blade at which neutrons have been detected as passing therethrough to be evaluated using progressively finer evaluation devices that permit a rapid evaluation of various portions of the blade with an appropriately high resolution without requiring evaluation of the entire blade at the same high resolution, thereby saving time. As such, an aspect of the disclosed and claimed concept is to provide an improved detection apparatus that is usable to detect a neutron absorption capability of a control element of a nuclear installation, the detection apparatus can be generally stated as including a processor apparatus which can be generally stated as including a processor and a storage, a robot apparatus in communication with the processor apparatus and that can be generally stated as including a number of manipulators, a neutron radiograph apparatus that can be generally stated as including an neutron emission source, a detector array, and a mask apparatus, the neutron radiograph apparatus being structured to receive the control element generally between the neutron emission source and the detector array, the neutron emission source being switchable between an ON state and an OFF state, the neutron emission source in the ON state being in an electrically energized condition structured to generate a neutron stream, the neutron emission source in the OFF state being in an electrically de-energized condition structured to output no meaningful neutron stream, the detector array being structured to detect an unabsorbed portion the neutron stream that passes without being absorbed through the control element, the detector array being further structured to generate an output signal that is representative of the unabsorbed portion the neutron stream; and the mask apparatus being movable by at least a first manipulator of the number of manipulators among a number of positions, a position of the number of positions being that in which the mask apparatus is disposed at least partially between the neutron emission source and the detector array, another position of the number of positions being that in which the mask apparatus is removed from between the neutron emission source and the detector array. Other aspects of the disclosed and claimed concept are provided by an improved method of operating the aforementioned detection apparatus to detect a neutron absorption capability of a control element of a nuclear installation wherein the nuclear installation has a pool of water available to accomplish control element inspection. The method can be generally stated as including receiving into the inspection pool of water with the neutron emission source in the OFF state, submerging the neutron emission source in the OFF state in the pool of water to a predetermined water depth, and switching the neutron emission source from the OFF state to the ON state when the control element inspection is to commence subject to the depth of the neutron emission source in the pool of water meeting or exceeding the predetermined water depth to provide adequate shielding to allow the neutron emission source to emit a neutron stream. The method potentially may further include receiving the detector array and the mask apparatus into the pool of water, receiving at least a portion of the control element generally between the neutron emission source and the detector array, and monitoring the detector array for the possible outputting therefrom of an output signal that would be representative of an unabsorbed portion the neutron stream passing without being absorbed through the at least portion of the control element. Similar numerals refer to similar parts throughout the specification. An improved detection apparatus 4 in accordance with the disclosed and claimed concept is depicted generally in FIGS. 1 and 3-4 and it is depicted schematically in FIG. 2. The detection apparatus 4 is advantageously usable to inspect a control element 6 of a nuclear installation 8 for its capability of absorbing neutrons. FIGS. 1 and 3-4 depict the control element 6 having been removed from what would be its usual position situated among a plurality of fuel assemblies of the nuclear installation 8 and having been received in a water pool 30 such as used fuel pool although any pool of water can be used to receive the control elements 6 for the inspection method that is set forth in greater detail below and that is performed using the detection apparatus 4. The detection apparatus 4 can be said to include a neutron radiograph apparatus 12, a robot apparatus 14, and a processor apparatus 16. Additionally, the detection apparatus 4 includes an input apparatus 18 that provides input signals to the processor apparatus 16 and an output apparatus 20 that receives output signals from the processor apparatus 16. Certain of the components, such as the processor apparatus 16, the input apparatus 18, and the output apparatus 20, or portions thereof, by way of example, may be situated on a control panel 21 that may be situated remotely from the robot apparatus 14 and which may be connected with the robot apparatus 14 via an umbilical 23. Other configurations of the various components will be apparent. The processor apparatus 16 can include a number of various components and can be generally in the nature of a general purpose computer that includes a processor such as a microprocessor or other processor and storage such as RAM, ROM, EPROM, EEPROM, FLASH, and the like and that function in the nature of internal storage on a computer. The processor apparatus 16 includes a number of routines 22 that are stored in the storage and that are executable on the processor to cause the processor apparatus 16 and thus the detection apparatus 4 to perform certain operations such as will be set forth in greater detail below. The input apparatus 18 can include any of a number detection devices or other input devices and can include the input terminals that are connected with the processor apparatus. The output apparatus 20 can include any of a wide variety of output devices and may include, for instance, visual displays, audio transducers, and data output devices, etc., without limitation. As can be understood from FIGS. 1 and 3, the control element 6 is of an elongated cruciform shape that can be generally described as including a central hub 24 and a plurality of wings 28A, 28B, 28C, and 28D, which may be collectively or individually referred to herein with the numeral 28, and each of which extends from the hub 24. Each of the wings 28 has formed therein a number of passages 39, a small number of which are represented in dashed lines on the wing 28B, that are filled with a boron containing material or other neutron-absorbing material that causes the control element 6 to absorb neutrons. It is noted that all of the wings 28 have the passages 39 formed therein, and the passages 39 are situated along substantially the entirety of the length of each wing 28. As will be set forth in greater detail below, the detection apparatus 4 is advantageously usable to assess whether any one or more of the passages 39 that are formed in each of the wings 28 have lost their boron or other neutron absorbing material to render the control element 6 at any one or more positions thereon transparent to neutrons and thus to not absorb neutrons or to have reduced neutron absorption as such positions. The neutron radiograph apparatus 12 can be said to include an neutron emission source 32, a detector array 36, and a mask apparatus 38. As a general matter, the neutron emission source 32 is switchable between an ON state and an OFF state. In the ON state, the neutron emission source 32 is in an electrically energized condition and is configured to output a neutron stream 50 such as is depicted generally in FIG. 4. In the OFF state, the neutron emission source 32 is in an electrically de-energized condition and outputs no meaningful stream of neutrons. The neutron emission source can be viewed as being shielded on all sides except the side that faces a wing 28 of the control element 6 and thus emits the neutron stream 50 generally only in one direction, namely that which faces toward the wing 28. The neutron emission source 32 can be said to include an accelerator 42, a number of first particles 44, and a number of second particles 48. The accelerator 42 is a particle accelerator which, when electrically energized, has a potential difference of multiple hundred kilovolts. The ability to switch the accelerator 42 between an energized ON state and a de-energized OFF state advantageously enables the neutron emission source 32 to be switched between the ON and OFF states. In the exemplary embodiment depicted herein, the first particles 44 are a number of deuterium particles. Also in the exemplary embodiment depicted herein, the second particles 48 can be either deuterium particles or tritium particles. When the accelerator 42 is in its electrically energized ON state, it accelerates the first particles 44 to cause them to impact the second particles 48 to cause nuclear fusion of the deuterium and/or tritium particles and thereby produce neutrons that form the neutron stream 50. In the situation wherein the second particles 48 are deuterium particles, the collision of one of the first particles 44 with one of the second particles 48 produces helium three (He3, which includes two protons plus a neutron in the nucleus) plus an additional neutron that is emitted as a part of the neutron stream 50. In the situation where the second particles 48 are tritium particles, the collision of one of the first particles 44 with one of the second particles 48 result in the generation of helium four (He4, which includes two protons and two neutrons in the nucleus) plus an additional neutron that is emitted as a part of the neutron stream 50. The detector array is a device that is configured to detect neutrons and, in particular, the neutron stream 50 and can be viewed as likewise being shielded on all sides except for the side that faces toward a wing 28. The detector array 36 can be any of a wide variety of devices that detect neutrons including but not limited to an ion chamber containing or coated with BF3, He3, enriched uranium (fission chamber) or other such isotopes that absorb neutrons and thereby promptly release a charged particle such as an alpha or beta particle, proton, deuteron, or fission fragments that can be subsequently collected and measured as a charge pulse or integrated into a current. The detector array 36 desirably will have a high neutron sensitivity and also have the ability to reject gamma radiation that is necessarily a component of a normal, irradiated control element. When the detector array 36 detects a neutron signal that exceeds a certain threshold, meaning a detection of portion of the neutrons from the neutron stream 50 that meets or exceeds a predetermined threshold, the detector array 36 outputs an alert signal that is received at the input apparatus 18 as an input to the processor apparatus 16. In response to the reception of such an alert signal originating from the detector array 36, the routines 22 may initiate further processing in the location where the detector array 36 is situated. Additionally or alternatively, the routines 22 may cause the processor apparatus 16 to generate some type of an output signal that is communicated to the output apparatus 22 and that generates an output that is detectable by a technician, for example. The mask apparatus 38 can be said to include a plurality of masks, at least some of which are represented by the four exemplary masks that are shown in FIGS. 1 and 3 and which are indicated at the numerals 52A, 52B, 52C, and 52D, and which may be collectively or individually referred to herein with the numeral 52. As will be explained below, the masks 52 can be said to additionally include another mask 85 that is depicted in FIG. 5C and may include other masks without limitation. The masks 52 are each of a plate-like configuration and are formed of a neutron absorbing material such as cadmium or other appropriate material. The masks 52A, 52B, 52C, and 52D each have an opening formed therein that is indicated, respectively, at the numerals 56A, 56B, 56C, and 56D, and which can be collectively or individually referred to herein with the numeral 56. In the depicted exemplary embodiment, the openings 56 are each in the form of an elongated slot wherein the material of the corresponding mask 52 has been removed, which thus permits neutrons to flow through the opening 56 if such a portion of the neutron stream 50 exist in the vicinity of the opening 56. The masks 52 with the openings 56 formed therein can each additionally or alternatively each be referred to as being a collimator. The robot apparatus 14 can be said to include a support 58 and a plurality of manipulators that are situated on the support 58 and that are indicated at the numerals 62A, 62B, 62C, 62D, 62E, and 62F, which may be collectively or individually referred to herein with the numeral 62. The manipulators 62 are each robotic manipulators having one end situated on the support 58 and another end opposite thereto carrying a component of the neutron radiograph apparatus 12. In the depicted exemplary embodiment, the manipulators 62A, 62B, 62C, and 62D carry at the free end thereof a corresponding mask 52A, 52B, 52C, and 52D, respectively. The manipulator 62E carries the neutron emission source 32, and the manipulator 62F carries the detector array 36. The manipulators 62 are each robotically operated and thus are operable independently of one another to move the masks 52 and the neutron emission source 32 and the detector array 36 independently of one another. It is expressly noted that the neutron radiograph apparatus 12 and the robot apparatus 14 are depicted herein as including only four masks 52 (plus the mask 85) and four corresponding manipulators 62 that carry the masks 52, but it is understood that any number of manipulators 62 and any number of masks 52 can be employed depending upon the needs of the particular application, as will be set forth in greater detail below. The openings 56A, 56B, 56C, and 56D in the depicted exemplary embodiment are each of an elongated approximately rectangular shape having a length 64A, 64B, 64C, and 64D, respectively and a width 68A, 68B, 68C, and 68D, respectively. As mentioned above, the masks 52 themselves absorb neutrons, but the openings 56 permit neutrons to flow therethrough. The robot apparatus 14 additionally includes some type of a robotic tractor mechanism that is situated on the support 58 and which engages the control element 6 on a wing 28 thereof or otherwise and moves the support 58 with respect to the control element 6 or vice versa. In this regard, it is understood that detection apparatus 4 or the support 58 or both may include additional structures or support elements that are situated at the base of the water pool 30 and which are configured to receive and to carry thereon the control element 6 when it is received in the water pool 30. Alternatively, the control element 6 can simply be received in the water pool 30 and can have the support 58 received thereon, with the tractor then being robotically operated to move the support 58 along the longitudinal extent of the control element 6. In use, the detection apparatus 4 with the neutron emission source 32 in its OFF state is received in the water pool 30 and is submerged therein until the neutron emission source 32 is situated at a predetermined depth 82 (FIG. 1) within the water pool 30. At the predetermined depth 82, the neutron emission source 32 can be switched to its ON state because the water in the water pool 30 provides sufficient shielding from the neutron stream 50 that no additional shielding is required to protect personnel from the neutron stream 50. Advantageously, therefore, the detection apparatus 4 relies upon the shielding that already exists in the water pool 30 to shield personnel from the neutron stream 50 when the neutron emission source is in its ON state. The ability to switch the neutron emission source between the ON and OFF states avoids the need for other shielding when the neutron emission source 32 is being transported from one location to another and prior to the neutron emission source 32 being submerged to the predetermined depth 82. That is, the neutron emission source 32 during transport and prior to being submerged to the predetermined depth 82 is simply in its OFF state and is not placed in its ON state until it is submerged to a depth in the water pool 30 at or below the predetermined depth 82. When the support 58 is situated on the control element 6 or vice versa, the manipulators 62E and 62F are operated to cause the neutron emission source 32 and the detector array 36 to be situated at opposite surfaces of the wing 28 that is to be inspected, which is the wing 28A in FIGS. 1, 3, and 4. Since the neutrons exiting the neutron emission source 32 are at too high of an energy state to be detected by the detector array 36, the neutron emission source 32 must be spaced away from the surface of the wing 28 by a sufficient distance to enable the neutrons to be thermalized and thus slowed sufficiently that the neutrons of the neutron stream 50 that pass through the wing 28 can be absorbed by the control element or, in the case of a defect, be detected by the detector array 36. At the initial stage of the detection operation, the manipulators 62 that have a mask 52 situated thereon are arranged such that the masks 52 are positioned away from the neutron emission source 32 and the detector array 36, such as is depicted generally in FIG. 1. The neutron emission source 32 is then switched to its ON state if it is not already in such a state, and the detector array 36 output signal is monitored to determine whether there is a defect at any given position on the wing 28. If no output is detected from the output 36, the tractor on the support 58 is operated to cause the support 58 and thus the manipulators 62 and the neutron radiograph apparatus 12 situated thereon to be advanced along the longitudinal extent of the control element 6. If no signal is detected from the detector array 36 at any point along the longitudinal extent of the wing 28 that is being inspected, the wing 28 is considered to be good, meaning that it has no regions thereon where neutrons from the neutron stream 50 pass unabsorbed through the wing 28, and rather the wing 28 has absorbed all portions of the neutron stream 50 without permitting any portions of the neutron stream 50 to be detected by the detector array 36. The process can then be repeated for the other wings 28 on the control element 6 until all of the wings 28 have been inspected. On the other hand, the detector array 36 may, at some location along the wing 28, generate an output signal that is received by the input apparatus 18 and is provided as an input to the processor apparatus 16 whereupon one of the routines 22 will view the signal originating from the detector array 36 as being indicative of the detection of a portion of the neutron stream 50 that has passed unabsorbed through the wing 28 at such location. An exemplary set of passages 69 are depicted in FIGS. 1 and 4 and demonstrate that the neutron emission source 32 and the detector array 36 are at any given time adjacent a plurality of the passages 69 that are formed in the wing 28 that is being evaluated. As such, an output signal from the detector array 36 that the detector array 36 has detected a portion of the neutron stream 50 passing through the wing 28 does not, of itself, indicate which of the one or more passages 69 that are situated adjacent the detector array 36 may be breached and may have lost its boron or other neutron absorbing substance. It is noted, however, that the mere detection of an output from the detector array 36 signaling the passage through the wing 28 of an unabsorbed portion of the neutron stream 50 does not automatically result in rejection of the wing 28 or the control element 6 as being defective and in need of replacement. Rather, if the result of the analysis presented herein is that a given control element 6 has no more than a predetermined number of failed positions (i.e., meaning positions thereon where neutrons are not being absorbed), and if the distribution of such positions is sufficiently scattered, the control element 6 can be declared to be effective and not in need of replacement. It is necessary, however, to determine the extent and location of any such positions on the control element 6 where neutrons are not being absorbed and are passing through the control element 6. As noted above, however, the detection of an output signal from the detector array 36 does not necessarily indicate which of the passages 69 is/are in a state of partial or total failure. As such, whenever the detector array 36 provides an output signal that is representative of a neutron signal being detected at a position on the wing 28 where neutrons are passing unabsorbed therethrough, the mask apparatus 38 is operated to more finely analyze the region of the wing 28 where the neutron emission source 32 and the detector array 36 are disposed in order to determine with greater specificity exactly what positions on the wing 28 are permitting neutrons to pass therethrough. Advantageously, therefore, the mask apparatus 38 is deployed, such as is depicted generally in FIG. 3 by operating one or more the manipulators 62 to position one or more of the masks 52 between the neutron emission source 32 and the detector array 36 to block from the detector array 36 all neutrons except those neutrons from the neutron stream 50 that are passing through the opening 56 that is formed in each of the masks 52 that have been deployed in such a masking fashion. While the masks 52 are individually deployable by the corresponding manipulators 62, it can be understood that any individual deployed mask 52 will block all of the neutrons of the neutron stream 50 except in the location where the opening 56 is situated, which may be a region having the length 64 and the width 68 of the corresponding opening 56 of the mask that is being situated between the neutron emission source 32 and the detector array 36. However, if a plurality of the masks 52 are deployed and positioned such that the openings 56 overlie one another, the portion of the neutron stream 50 that can be received by the detector array 36 is only that portion of the neutron stream 50 that passes through the wing 28 and that also passes through the region where the corresponding openings 56 are overlying one another. Such a cooperating plurality of masks 52 can be referred to herein as being a mask system 70, and the overlying portions of the openings 56 of the masks 52 of the mask system 70 can be said to form an orifice 72. If the openings 56 are oriented perpendicular to one another, the resulting orifice 72 is much smaller than the cross-sectional area of the detector array 36, meaning that relatively smaller portions of the wing 28 (i.e., smaller than the cross section dimensions of the detector array 36) can be separately inspected to determine whether each such position on the wing 28 is defective and is permitting neutrons from the neutron stream 50 to pass therethrough. As can be understood from FIG. 5A, the mask system 70 has the orifice 72, which can be understood to be much smaller than the detector array 36 that is overlying the wing 28. It is noted that FIGS. 5A, 5B, 5C, and 5D depict in phantom lines an exemplary outline of the detector array 36 as it would be situated overlying the wing 28, but with the detector array 36 having been removed in order to better visualize the overlying openings 56 in the masks 52. As can be understood from FIG. 5A, the orifice 72 can be said to have a first dimension 76 (which, in the exemplary embodiment depicted herein, is equal to the width 68A) and a second dimension 78 (which, in the depicted exemplary embodiment, is equal to the width 68B). With the masks 52A and 52B of that particular mask system 70 being situated as depicted in FIG. 5A, the detector array 36 is monitored to determine whether it generates an output signal that would be representative of neutrons from the neutron stream 50 passing through the orifice 72 which overlies a given position 84A on the wing 28. If a signal is detected from the detector array 36, this fact is recorded in the processor apparatus 16. That is, the processor apparatus 16 would record the fact that neutrons were detected at the position 84A on the wing 28. On the other hand, if neutrons were not detected when the mask system 70 was positioned as depicted in FIG. 5A, the processor apparatus 16 may (and likely would) record the fact that neutrons were not detected at the position 84A on the wing 28. The mask system 70 then can be manipulated to move the orifice 72 to a different position, such as the position 84B that is depicted in FIG. 5B. As can be understood from a comparison of FIGS. 5A and 5B, it can be understood that the orifice 72 was moved from the first position 84A of FIG. 5A to the second position 84B of FIG. 5B by employing the manipulator 62B to translate the mask 52 in the vertically downward direction from the perspective of FIGS. 5A and 5B when going from FIG. 5A to FIG. 5B. With the mask system 70 as arranged in FIG. 5B, the detector array 36 is monitored to determine whether it is generating an output signal that would be indicative of neutrons passing through the orifice 72 at the position 84B. Whether or not an output signal is detected from the detector array 36 when the orifice 72 overlies the position 84B is recorded in the processor apparatus 16. In this regard, it is understood that any number of intermediate positions between the position 84A of FIG. 5A and the position 84B of FIG. 5B can also be evaluated for the detection of neutrons by the detector array 36 and such detection (or non-detection) recorded in the processor apparatus 16, as appropriate. The mask apparatus 38 thus can be manipulated until all of the positions on the wing 28 that are adjacent the detector array 36 have been evaluated by overlying the orifice 72 over such positions and detecting whether or not the detector array 36 has generated an output signal that is indicative of neutron passage at such position on the wing 28. In so doing, it may be necessary to employ different masks, such as is depicted in FIG. 5C wherein the mask 52B is combined with the other mask 85 of the mask apparatus 38 to form another mask system 70A having an orifice 73 overlying another position 84C on the wing 28. For instance, the mask 52A may have been withdrawn and the mask 85 deployed in its place. The other mask 85 was not depicted in FIGS. 1 and 3 for reasons of simplicity of disclosure. The mask 85 is manipulated separately by its own manipulator of the robot apparatus 14 that is likewise not depicted in the accompanying drawings for reasons of simplicity of disclosure. In the depicted exemplary embodiment, the orifice 73 of FIG. 5C has the same first and second dimensions 76 and 78 as the orifice 72 of FIGS. 5A and 5B. It can be seen that the mask 85 has its own opening formed therein that is positioned thereon at approximately the middle thereof rather than being situated closer to the end thereof (as is the case with the mask 52A), and this enables an end 87 of the mask 85 to extend beyond the edge of the opening 56B. That is, the various masks (such as are indicated at the numerals 52, 85, etc. and which can include other masks that are not expressly depicted herein) are configured such that in various combinations they can completely block the passage of neutrons from the neutron stream 50 that may pass through the wing 28 other than the neutrons that additionally pass through the orifice 72 or 73, by way of example. It can be understood that the movement of one or more of the masks 52 with respect to other masks 52 and the like can be programmed into one of the routines 22 and executed by the processor apparatus 16 and the robot apparatus 14 that is connected therewith. Likewise, the selection of the various masks 52 and 85 and other such masks that may not be expressly depicted herein and their use in combination with other such masks 52 can likewise be programmed into one of the routines 22. It thus can be understood from the foregoing that the detector array 36 situated at the location depicted generally in FIG. 5A had generated an output signal, indicating that neutrons were passing from the neutron stream 50 through the wing 28 at one or more positions on the wing 28 directly adjacent the detector array 36 and were being detected by the detector array 36. As such, the masks 52A and 52B were deployed and were positioned as is depicted in FIG. 5A. The mask system 70 was manipulated such that the orifice 72 thereof was successively positioned, in a raster-like fashion, across all of the positions (84A, 84B, 84C, etc.) between the detector array 36 and the wing 28 to determine at which of such positions the detector array 36 again produced an output that is indicative of neutrons. In other words, the detector array 36 positioned as situated in FIGS. 5A-5C detected neutrons at some position on the wing 28 adjacent the detector array 36, and the various masks 52 were thus deployed between the detector array 36 and the wing 28 and were positioned to move the orifice 72 among all of the positions (such as the positions 84A, 84B, and 84C) to determine with greater specificity the particular position(s) on the wing 28 adjacent the detector array 36 where neutrons were actually passing through the wing 28. By way of example, it may have been determined that a positive signal was received from the detector array 36 only at the position 84C in FIG. 5C. This would indicate that when the detector array 36 was situated in the location on the control element 6 that is depicted in FIGS. 5A-5C, and when the detector array 36 initially generated its output signal representative of detecting neutrons somewhere along its cross-sectional area, it really was detecting neutrons somewhere within the position 84C. This would indicate that the wing 28 has a failed region within position 84C. Depending upon the sizes of the transverse dimensions of the orifice 73, the determination, as in FIG. 5C, that the position 84C was the source of neutrons may be sufficiently precise information that no further analysis is needed. On the other hand, it may be decided that merely identifying position 84C to the processor apparatus 16 is insufficiently accurate to evaluate the overall ability of the control element 6 to absorb neutrons, whether simply at the location or overall, and that more specific and fine detail may be warranted, as is depicted generally in FIGS. 5D and 6. For instance, it may be decided to remove the masks 52B and 58 from being situated adjacent the detector array 36 and to instead deploy the masks 52C and 52D to form another mask system 70B whose openings 56C and 56D are relatively narrower than those of the mask system 70A. That is, the widths 68C and 68D are relatively narrower than the widths 68B and the width of the opening in the mask 58. As such, the openings 56C and 56D overlaid as in FIG. 5D may form a relatively smaller orifice 83 having relatively smaller first and second dimensions 76A and 78A (FIG. 6), i.e., relatively smaller than the first and second dimensions 76 and 78 of the orifices 72 and 73. The mask system 70B can be manipulated such that the orifice 83 is successively positioned along a plurality of further positions 88 that are depicted in FIG. 6 and are themselves each discrete areas within the position 84C that was identified in FIG. 5C as having neutrons passing therethrough from the neutron stream 50. As such, the masks 52C and 52D can be manipulated by the manipulators 62C and 62D such that the resultant orifice 83 moves in a raster-like fashion across the position 84C among the various further positions 88 until a number of the further positions 88 are detected to have a neutron signal passing therethrough. Whether or not the detector array 36 output its alarm signal coincident with any of the further positions 88 is recorded in the processor apparatus 16. The exemplary further positions 88 where neutrons were detected are indicated schematically in FIG. 6 with cross-hatching and are represented at the numerals 90A, 90B, 90C, 90D, and 90E, which may be collectively or individually referred to herein with the numeral 90. If the detection of a neutron signal at the further positions 90 is information that is sufficiently detailed and precise, the evaluation of the position 84C can end and the detector array 36 can be moved to another location along the length of the control element 60. Alternatively, if more detailed analysis of the further positions 90 is desired, further masks 52 having even smaller openings formed therein can be deployed to evaluate in a raster-like fashion a number of smaller areas within each of the further positions 90, and the resulting output signals from the detector array 36 coincident with such smaller areas recorded in the processor apparatus 16, until sufficiently detailed information regarding exactly where on the wing 28 the neutron stream 50 is passing through is obtained. It is reiterated that all such outputs from the detector array 36 are received in real time at the processor apparatus 16. It is understood that the initial use of the detector array 36 without the mask apparatus 38 until the detector array 36 provides an output, and then the responsive use of masks 52 having progressively smaller openings 56 at the positions (84A, 84B, and 84C, for example) and further positions 84 and 90 (further by way of example) where neutron signals were detected saves inspection time and results in a rapid inspection procedure. While the entire wing 28 can be evaluated at the finest possible orifice size, such evaluation of the entire wing with such fineness would take an excessively long period of time and likely would be unnecessary. However, by performing a relatively coarse analysis by using the masks 52A and 52B with relatively larger openings 56A and 56B to identify, on a coarse basis, the various positions 84 where a neutron signal is detected, and by then performing a relatively finer evaluation using the masks 52C and 52D having the relatively smaller openings 56C and 56D, the relatively finer evaluation is done only at the positions 84 where it is known that a neutron signal exists. As such, the various positions where no signal is detected can be rapidly evaluated and ignored using the relatively coarse analysis afforded by the masks 52A and 52B, and the relatively finer analysis can be deployed only where neutron signals are known to have been detected. The use of such progressively finer evaluation at the positions where neutron signals are detected provides for a greater efficiency and minimized time waste. It is understood that the various masks 52 that are used may be dependent upon the configuration of the passages 69 in the control element 6. As mentioned above, the passages 69 that are depicted in an exemplary fashion in FIG. 1 in the wing 28B (and which are representative of the passages 69 that are situated along substantially the entirety of the lengths of all of the wings 28) represent one known way of configuring the passages 69 with the boron or other neutron absorbing substance contained therein, i.e., perpendicular to the longitudinal extent of the hub 24. As noted above, an alternative design exists wherein the passages that contain the boron or other neutron absorbing materials are oriented parallel with the longitudinal extent of the hub 24. It thus may be desirable to provide masks of different configurations having openings with different sizes and/or shapes and/or orientations that are optimized to evaluate such an alternative arrangement of passages. An alternative mask system 170 is depicted in FIGS. 7A-7C. The alternative mask system 170 is configured such that its opening 156 can selectively have a plurality of different dimensions. This is accomplished by configuring the mask system 170 to include a plate-like first member 154 having a notch 160 formed therein and by further providing a plate-like second member 166 which is movable with respect to the first member 154. In this regard, a separate robotic tractor may be provided on the mask system 170 to move the second member 166 with respect to the first member 154 or vice versa. As can be understood from FIGS. 7A-7C, the opening 156 is of a fixed length 164, but the width is variable between, for instance, a relatively wider width 168A (such as is depicted in FIG. 7A), a relatively smaller width 168B (such as is depicted in FIG. 7B), and a further smaller width 168C (as is depicted in FIG. 7C). The various widths 168A, 168B, and 168C are exemplary and can be larger, smaller, etc. without departing from the spirit of the present disclosure. By way of example, the mask system 170 potentially could take the place of the mask 52A being removed and replaced with the mask 52C having the relatively narrower opening 56C. Likewise, another instance of the mask system 170 could take the place of both the masks 52B and 52D and would avoid having to remove the mask 52B to be replaced with the mask 52D. Rather, the mask system 170 could be retained in place adjacent the detector array 36 and simply manipulated to make its opening 156 relatively narrower, such as by progressively changing it such that its width goes from that depicted in FIG. 7A and is indicated at the numeral 168A until the width is reduced to that indicted at the numeral 168C in FIG. 7C. The mask system 170 could be used with additional instances of the mask system 170 or with other masks 52 without limitation. It can be understood that the mask system 170 could take other forms that permit not only the width of the opening 156 to be changed, but also the length 164 can be changed as well as the position of the opening 156 thereon can likewise be changed. Any of a number of configurations of additional mask systems that provide various opening positions or dimensions and both can be envisioned. An improved method in accordance with the disclosed and claimed concept is depicted generally in FIGS. 8A and 8B. Processing can be said to begin, as at 204, where the neutron radiograph apparatus 12 in its OFF state is received in the inspection water pool 30 of the nuclear installation 8 and is submerged until the neutron emission source 32 is at or below the predetermined depth 82. The control element 6 can then be positioned, as at 208, in the water pool 30 such that a wing 28 or other portion of the control element 6 is situated generally between the neutron emission source 32 and the detector array 36. The neutron emission source 32 can then be switched from its OFF state to its ON state, as at 212, by energizing the accelerator 42 to cause it to emit the neutron stream 50. The support 52 can then be incrementally moved with respect to the control element 6 or vice versa while the detector array 36 is monitored for the possible outputting of a signal from the detector array 36 that would be representative of an unabsorbed portion of the neutron stream 50 passing unabsorbed through a portion of the control element 6. Such an output, if detected, could be received by the input apparatus 18 and provided as an input signal to the processor apparatus 16. If, as at 220, no signal above a threshold value is detected from the detector array 36, processing can continue, as at 216, where the neutron radiograph 12 is incrementally moved farther along the control element 6. On the other hand, if at 220 a signal is received from the detector array 36, an output can be generated, as at 224. The output, as noted above, can be received by the input apparatus 18 and can result in the outputting of an alarm signal or the triggering of a routine 22 to manipulate the mask system 38, or both. For example, and as at 228, a portion of the mask apparatus 38 can be moved from a location that is disposed generally away from the neutron emission source 32 and the detector array 36 to another location situated generally between the neutron emission source 32 and the detector array 36. Processing can then continue, as at 232, where the control element 6 is evaluated using the mask system 38 by using an orifice 72 that is of an initial size and that is moved in a raster-like fashion about that region of the control element 6 while monitoring the detector array 36 for additional signals that would be representative of an unabsorbed portion of the neutron stream 50 passing unabsorbed through the control element 6 and through the orifice 72. As at 236, the processor apparatus 16 can record the one or more positions on the control element 6 at which the additional signals were received from the detector array 36. If it is determined, as at 240, that greater fineness in the evaluation is not needed, the analysis can be terminated, as at 244, and the results of the analysis can be output for expert evaluation. However, if it is determined at 240 that further fineness in the evaluation is desired, processing can continue, as at 248, where a further evaluation of the positions at which the signals were detected with the detector array 36 can be carried out by using the mask system and an orifice 83 of a relatively smaller size. Again, such analysis would be performed in a raster-like fashion as set forth above, or can be carried out in other fashions without departing from the spirit of the present disclosure. Recordation can occur, as at 252, as to any one of more finely defined positions on the control element 6 at which further signals have been received from the detector array 36. It is understood that the procedure noted above can be performed in any of a variety of different orders depending upon the need for efficiency. For example, it is possible that the wing 28 may be evaluated in its entirety using the neutron emission source 32 and the detector array 36 without deployment of the mask apparatus 38 in order to find the various locations thereon where neutron signals are detected. Afterward, another pass can be made using the masks 52A and 52B having the relatively larger openings 56A and 56B, and such mask system 70 can be moved in a raster-like fashion across the various locations that were detected and stored during the first pass when the mask apparatus 38 was not employed. The various positions where signals are detected with the masks 62A and 62B in place would be recorded. This could occur across the entirety of the length of the wing that is being evaluated. Then, perhaps, a further pass can be performed across the entirety of the length of the control element using the relatively finer masks 52C and 52D, and only those positions where signals have been detected and stored would be the subject of further analysis by the mask system 70A. Other variations will be apparent. It thus can be understood that the detection apparatus and the method described herein permit rapid evaluation of the control element 6 and permits a determination to be made whether the control element is in need of replacement or whether the control element 6 can be redeployed in the nuclear installation 8. It is possible that variations of the detection apparatus 4 can be provided wherein multiple wings 28 are simultaneously evaluated using multiple neutron emission sources and detector arrays. The neutron emission source 32 advantageously can be switched between an OFF condition and an ON condition, thereby avoiding the need for shielding when the emissions source 32 is being transported from one location to another. As mentioned above, known californium neutron emission sources cannot be switched off, and thus the neutron emission source 32 advantageously avoids the need for the separate, heavily shielded shipping cask required to transport a californium source without the presence of a 5 meter deep large pool of water. Additionally, the evaluation using the detector array 36 and the mask system 38 provides real time analysis of the wing 28, which enables greater efficiency in directing the evaluation, and it also avoids the need for prolonged exposure of film to a neutron source. Additionally, the configuration of the mask apparatus 38 to provide progressively smaller orifices to provide finer evaluation only in those positions where such finer evaluation is needed saves evaluation time and results in a rapid overall evaluation of the control element 6. Further advantages will be apparent. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
046612904	summary	BACKGROUND OF THE INVENTION The present invention relates to an apparatus for reducing the volume of various solid waste materials discharged from homes, factories, nuclear power plants, for example, to produce compacted solid masses for landfill or storage, to such an apparatus further comprising its accessory facilities and also to a method for cleaning said apparatus. Various solid waste materials are discharged from homes, factories, power plants, and other facilities of today. For example, solid wastes such as pieces of plastics, metal, glass, and other materials are discharged from homes and factories, and radioactive wastes such as rags, polyethylene sheets, paper, concrete pieces, steel members, high-performance filters, heat insulation are discharge from nuclear power plants. The aforesaid solid wastes discharged from homes and factories are processed in different ways, which raise their own problems. For example, the discharged pieces of plastics are burned in incinerators. However, the incinerators tend to be melted by heat generated by burning of plastics, clogged by the molten plastics and damaged by local overheating. Furthermore, the incinerators produce harmful gases such as chlorine and dioxin. Landfill of polystyrene foam pieces, polyethylene sheets, plastics bags is disadvantageous in that since these materials are bulky, the cost of transportation thereof is high, and they tend to be exposed onto the landfill surface after buried and then scattered due to winds, resulting in environmental pollution. Various methods have been proposed to recover and reuse waste plastics for effective exploitation of resources. However, since urban trash includes a wide variety of materials, it is more costly to classify the different trash materials than to recover the waste plastics. It has been proposed in Japanese Patent Publication No. 57-11273 to produce large solid masses containing inorganic particles fixed together with a melt of thermoplastic waste materials by adding a granular or particulate inorganic material such as sand, crushed stone, or ash to the thermoplastic waste materials under heat. This process is however not suitable to granulate metals, fabric pieces and the like, and hence is required to be effected after the metals, fabrics and the like have been separated from the other materials. The wastes contaminated by a radioactive material in nuclear power plants are normally packed in polyvinyl bags which are placed in drum cans for storage, sometimes after having been classified into combustible and noncombustible materials. Waste materials such as high-performance filters composed of a wood material, a filter aid (inorganic), a metal plate and the like which are joined together are required to be disassembled into individual parts which should then be sorted. This process is complicated and gives the workers a greater chance to get exposed to radiation. The drum cans are stored in storage houses. Since the available storage spaces in the storage houses are becoming smaller than expected these days, the combustible radioactive wastes are burned and the produced ash is stored in drum cans or mixed with cement and solidified as stable solid masses. The burning process is used for processing the bulky materials such as polyethylene sheets, polyvinyl bags, rags and paper wastes, and hence is capable of reducing the volumes of these waste materials. For this reason, the burning process is widely employed to process the waste materials discharged from nuclear power plants, facilities using radioisotopes, and other similar facilities. However, when a large quantity of plastic materials are burned in an incinerator, the incinerator is liable to get damaged, and the incinerator system is required to be equipped with an exhaust gas processing apparatus which itself produces a secondary waste material. In addition, the cost of installation of said apparatus is high. Another process of treating radioactive solid waste materials is to employ a press for compacting the wastes into smaller volumes. There is developed a high-pressure compacting apparatus for compacting waste materials under a pressure ranging from 1000 to 3000 kg/cm.sup.2. According to this compacting process, polyethylene sheets, polyvinyl bags, paper wastes, rags and similar waste materials can be compacted into masses free of free spaces or air pockets trapped therein. Although the compacting process is one of effective methods, produced compacted masses are combinations of heterogeneous materials that are not desired for a long-term storage. SUMMARY OF THE INVENTION It is an object of the present invention to provide an apparatus for compacting solid waste materials discharged from homes and factories or radioactive solid waste materials discharged from nuclear power plants and facilities using radioisotope, to solidify the waste materials into masses which can be pelletized, stuffed into drum cans, or shaped into prismatic forms for stable landfill or storage, without producing secondary waste materials. Another object of the present invention is to provide such an apparatus further comprising an accessory facility for stably controlling the operation of the foregoing compacting apparatus. Still another object of the present invention is to provide such an apparatus further comprising an accessory facility for shaping materials extruded from the foregoing compacting apparatus, cutting off the shaped materials, and charging the cut-off materials into containers. A still further object of the present invention is to provide such an apparatus further comprising an accessory facility for processing an exhaust gas from the foregoing compacting apparatus for use as effective cooling means without requiring extra equipment and producing unwanted secondary waste materials. A yet still further object of the present invention is to provide such an apparatus further comprising an accessory facility for cleaning the foregoing compacting apparatus without requiring extra equipment and producing unwanted secondary waste materials. According to the present invention, there is provided an apparatus for compacting a solid waste material, comprising a hollow cylindrical body having a charging inlet for charging the solid waste material therethrough into the hollow cylindrical body, a heating portion for heating the solid waste material in the hollow cylindrical body, and a discharging outlet for discharging the solid waste material from the hollow cylindrical body; a rotatable shaft disposed in the hollow cylindrical body and having a helical screw blade thereon, the screw blade and an inner wall surface of the hollow cylindrical body being spaced from each other with a clearance provided therebetween for allowing the solid waste material in the hollow cylindrical body to form a bridge therein; support means for radially movably supporting an end of the rotatable shaft; a prime mover coupled to the end of the rotatable shaft through the support means; and an outlet nozzle coupled to the discharging outlet for compressing the solid waste material discharged from the discharging outlet. The compacting apparatus can compact various solid waste materials including plastics discharged from homes, factories, nuclear power plants, and other facilities, and compacted solid waste materials can be solidified into solid masses or pelletized. The solidified masses produced by the compacting apparatus are coated with plastics, and can be buried directly if they are derived from urban trash. Harmful metals contained in dry cells and other discarded devices are confined in the solidified masses to guard against environmental pollution. Since the solid waste materials are processed in a confined spaced at a temperature below 280.degree. C., mercury contained in dry cells and the like and dioxin are not discharged from the compacting apparatus. The exhaust gas from the compacting apparatus can easily be processed as only the charging inlet is open. If radioactive solid waste materials are so small in quantity that they are negligible depending on the amount and type of the nuclide, they can be stuffed into a drum can or container which can be buried. The solid waste material as it is pelletized and stored in drum cans can finally be melted. The compacting apparatus of the invention can be carried on a truck or otherwise mobilized so that it can be moved to a desired processing location. If a chopper is disposed in or near the charging inlet such as a hopper of the hollow cylindrical body, then the conventional supply device composed of a chopper and a conveyor can be dispensed with, and the overall apparatus is compact in size. For processing radioactive solid waste materials, the compacting apparatus is useful in preventing environmental pollution. According to the present invention, the compacting apparatus may further comprise an accessory facility for controlling rotation of the rotatable shaft depending on a current flowing through the prime mover or a torque imposed on the prime mover. With this controlling method, the screw blade and the prime mover such as a motor can be prevented from being damaged or broken and can be continuously operated safely when the solid waste material is compacted. According to the present invention, the compacting apparatus may further comprise an accessory facility for producing a solid mass comprising a plurality of shaping tubes into each of which an extruded mass from the compacting apparatus can be forcibly charged; means for closing one of the shaping tubes at a time when the extruded mass is charged into said shaping tube; means for moving over the shaping tubes; and means for ejecting the extruded mass out of said one shaping tube. The method of producing a solid mass by the use of the accessory facility may comprise the steps of positioning a shaping tube with an end thereof opening toward the discharging outlet; closing an opposite end of the shaping tube with closing means; forcibly charging an extruded mass from the apparatus into the shaping tube; thereafter rotating the rotatable shaft in a reverse direction to cut off the extruded mass in the vicinity of the discharging outlet; moving over the shaping tube filled with the extruded mass to cool the same; and ejecting the shaped and cooled extruded mass out of the shaping tube. With the above producing facility and method, well shaped solid masses are successively produced at a constant rate, and can automatically be treated to thereby save labor power. For processing radioactive waste materials, the automatic treatment of the solid masses is effective in preventing workers from being exposed to radioactivity. The cost of the facility is low since it is not necessary to add a special cutter for cutting the extruded mass. According to the present invention, the compacting apparatus may further comprise an accessory facility for processing an exhaust gas discharged from the compacting apparatus, the processing method comprising the steps of filtering the exhaust gas produced by the apparatus to collect dust particles therefrom; thereafter discharging the exhaust gas through an exhaust gas line; and cooling a cooling medium for an induction heating coil for heating the heating portion, with the exhaust gas in a heat exchanger disposed on the exhaust gas line. According to this exhaust gas processing method, the exhaust gas produced when compacting the solid waste material is used for cooling the cooling medium for the induction heating coil, and hence no extra equipment such as conventional cooling tower and fan coil is not required. A fan used in the gas processing system is employed both to discharge the exhaust gas and to cool the cooling medium, and hence is an energy saver. The dust particles collected by the filter used in the gas processing system can be processed by the compacting apparatus without producing an undesired secondary waste material. According to the present invention, there is further provided a method for cleaning the compacting apparatus, the method comprising the step of charging a cleaning material into the hollow cylindrical body to force a remaining solid waste material out of the hollow cylindrical body. Since no other materials than the solid waste material and sand as the cleaning material, the remaining solid waste material can be cleared away without producing a secondary waste material and without requiring additional cleaning equipment. Therefore, the compacting apparatus is inexpensive and can easily be operated for cleaning. The cleaning process can prevent a reduction of the compacting efficiency due to melted plastics deposited in the apparatus. The above and other objects, features and advantages of the present invention will become more apparent from the following description when taken in conjunction with the accompanying drawings in which preferred embodiments of the present invention are shown by way of illustrative example.
054815754	abstract	A method and a device for detecting oscillations in the core of a boiling-water nuclear reactor (BWR) comprising a plurality of neutron detectors, wherein instability is detected based on oscillations in the output signals of the neutron detectors. For each one of a number of selected neutron detectors, oscillations in the output signals of the neutron detectors are detected. The oscillations are detected on the basis of an oscillation signal (SS) which indicates that an oscillation criterion is fulfilled. An alarm signal for remaining oscillations Ka is generated if the oscillation signal (SS) remains during one delay interval (T3). A first alarm signal (LR) is generated if the oscillation signal (SS) indicates that the oscillation criterion is fulfilled at least once, during each of a predetermined number of consecutive alarm intervals (T1). An alarm signal for intermittent oscillations (Ra) is generated by blocking the first alarm signal (LR) when the alarm signal for remaining oscillations (Ka) is present. (FIG. 5)
	claims	1. A fuel bundle for use in a channel type heterogeneous reactor core of a heavy water reactor, the fuel bundle comprising:a filled central displacement tube; anda plurality of thorium based fuel pins surrounding the filled central displacement tube,wherein the plurality of thorium based fuel pins are arranged in a single ring around the filled central displacement tube. 2. The fuel bundle of claim 1, wherein the filled central displacement tube is filled with ZrO2, MgO, BeO, graphite or stagnant D2O coolant. 3. The fuel bundle of claim 1, wherein there are 21 radially positioned thorium based fuel pins surrounding the filled central displacement tube. 4. The fuel bundle of claim 1, wherein there are 35 radially positioned thorium based fuel pins surrounding the filled central displacement tube. 5. The fuel bundle of claim 1, wherein the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 3% or higher. 6. The fuel bundle of claim 1, wherein the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2 +ThO2) with a UO2 content of 35% or higher. 7. The fuel bundle of claim 1, wherein the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 2% or less. 8. The fuel bundle of claim 1, wherein the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2+ThO2) with a UO2 content of 30% or less. 9. A method of using a thorium based fuel comprising:placing the fuel bundle of claim 1 in a channel type heterogenous reactor core of a heavy water reactor; andconducting a burnup of the thorium based fuel. 10. The fuel bundle of claim 1, wherein the filled central displacement tube is filled with stagnant D2O coolant. 11. A fuel bundle for use in a channel type heterogeneous reactor core of a heavy water reactor, the fuel bundle comprising:a filled central displacement tube; anda plurality of thorium based fuel pins surrounding the filled central displacement tube,wherein all of the plurality of thorium based fuel pins have an identical composition, andwherein the plurality of thorium based fuel pins are arranged in a single ring around the filled central displacement tube. 12. The fuel bundle of claim 11, wherein the filled central displacement tube is filled with stagnant D2O coolant.
052672786	summary	A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the Patent and Trademark Office patent file or records, but otherwise reserves all copyright rights whatsoever. BACKGROUND OF THE INVENTION The present invention relates to apparatus and methods for monitoring and controlling the operation of commercial nuclear power plants. Conventionally, commercial nuclear power plants have a central control room containing equipment by which the operator collects, detects, reads, compares, copies, computes, compiles, analyzes, confirms, monitors, and/or verifies many bits of information from multiple indicators and alarms. Conventionally, the major operational systems in the control room have been installed and operate somewhat independently. These include the monitoring function, by which the components and the various processes in the plant are monitored; control, by which the components and the processes are intentionally altered or adjusted, and protection, by which a threat to the safety of the plant is identified and corrective measures immediately taken. The result of such conventional control room arrangement and functionality can sometimes be information overload or stimulus overload on the operator. That is, the amount of information and the variety and complexity of the equipment available to the operator for taking action based on such extensive information, can exceed the operator's cognitive limits, resulting in errors. The most famous example of the inability of operators to assimilate and act correctly based on the tremendous volume of information stimuli in the control room, particularly during unexpected or unusual plant transients, is the accident that occurred in 1978 at the Three Mile Island nuclear power plant. Since that event, the industry has focused considerable attention to increasing plant operability through improving control room operator performance. A key aspect of that improvement process is the use of human engineering design principles. Advances in computer technology since 1978 have enabled nuclear engineers and control room designers to display more information, in a greater variety of ways, but this can be counterproductive, because part of the problem is the overload of information. Improving "user friendliness" while maintaining the quantity and type of information at the operator's disposal has posed a formidable engineering challenge. SUMMARY OF THE INVENTION It is thus an object of the present invention to provide apparatus and method for nuclear power plant control and monitoring operations having the characteristics of concise information processing and display, reliable architecture and hardware, and easily maintainable components, while eliminating operator information overload. This objective should be accomplished while achieving enhanced reliability, ease of operation, and overall cost effectiveness of the control room complex. The solution to the problem is accomplished with the present invention by providing a number of features which are novel both individually and as integrated together in a control complex. The complex includes six major systems: (1) the center panels, (2) the data processing system (DPS), (3) the discrete indication and alarm system (DIAS), (4) the component control system consisting of the engineered safeguard function component controls (ESFC) and the process component controls (PCC), (5) the plant protection system (PPS), and (6) the power control system (PCS). These six systems collect data from the plant, efficiently present the required information to the operator, perform all automatic functions and provide for direct manual control of the plant components. The control complex in accordance with the invention provides a top-down integrated information display and alarm approach that supports rapid assessment of high level critical plant safety and power production functions; provides guidance to the operator regarding the location of information to further diagnose high level assessments; and significantly reduces the number of display devices relative to conventional nuclear control complexes. The complex also significantly reduces the amount of data the operator must process at any one time; significantly reduces the operational impact of display equipment failures; provides fixed locations for important information; and eliminates display system equipment used only for off normal plant conditions. It is known that the nuclear steam supply system can be kept in a safe, stable state by maintaining a limited set of critical safety functions. The present invention extends the concept of the critical plant safety functions to include critical plant power production functions, in essence integrating the two functions so that the information presentation to the operator supports all high level critical plant functions necessary for power production as well as safety. The information display hierarchy in accordance with the invention includes a "big board" integrated process status overview screen (IPSO) at the apex, which provides a single dedicated location for rapid assessment of key information indicative of critical plant power production and safety functions. Further detail on the sources and trends of normal or abnormal parameter changes are provided by the DIAS. Both IPSO and the DIAS provide direct access and guidance to additional system and component status information contained on a hierarchy of CRT display pages which are driven by the DPS. The IPSO continually displays spatially dedicated information that provides the status of the plant's critical safety and power production functions. This information is presented using a small number of easily understood symbolic representations that are the results of highly processed data. This relieves the operator of the burden of correlating large quantities of individual parameter data, systems or component status, and alarms to ascertain the plant functional conditions. The IPSO presents the operator with high level effects of lower level component problems. The IPSO relies primarily on parameter trend direction, e.g., higher, lower, an alarm symbol color and shape, to convey key information. These are supplemented by values for selected parameters. The IPSO presents consolidated, simplified information to the operator in relatively small quantities of easily recognized and understood information. Furthermore, the IPSO compensates for the disadvantage inherent in recent industry trends towards presenting all information serially on CRTs, by enabling the operator to obtain an overview, or "feel" of the plant condition. Display of plant level overview on a large-format dedicated display addresses two additional operational concerns. First, operator tasks often require detailed diagnostics in very limited process areas. However, maintaining concurrent awareness of plant-wide performance is also necessary. Rather than relying on multiple operators in the control room to monitor respective indicators and the like on spatially separated panels, the IPSO can be viewed from anywhere in the control room and thus provides an operator a continuous indication of plant performance regardless of the detailed nature of the task that may be requiring the majority of his attention. In the preferred implementation, IPSO supports the assessment of the power and safety critical functions by providing for each function, key process parameters that indicate the functional status. For each function, key success paths are selected with the status of that success path displayed. The IPSO clearly relates functions to physical things in the plant. The critical functions are applied to power production, normal post trip actions, and optimal functional recovery procedures. The second level in the display information hierarchy in accordance with the present invention is the presentation of plant alarms from the DIAS. A limited number of fixed, discrete tiles are used with three levels of alarm priorities. Dynamic alarm processing uses information about the plant state (e.g., reactor power, reactor trip, refueling, shut-down, etc.) and information about system and equipment status to eliminate unnecessary and redundant alarms that would otherwise contribute to operator information overload. The alarm system provides a supplementary level of easily understood cueing into further information in the discrete indicators, CRTs and controls. Alarms are based on validated data, so that the alarms identify real plant process problems, not instrumentation and control system failures. The alarm features include providing a detailed message through a window to the operator upon the acknowledgment of an alarm and the ability to group the alarms without losing the individual messages. The tiles can dynamically display different priorities to the operator. The acknowledgment sequence insures that all alarms are acknowledged while at the same time reducing the operator task loading by providing momentary tones, then continuous alarm, followed by reminder tones to insure that the alarms are not forgotten. The operator has the ability to stop temporarily alarm flashing to avoid visual overload, and resume the flashing to insure that the alarm will eventually be acknowledged. The discrete indicators in the DIAS provide the third level of display in the hierarchy of the present invention. The flat panel displays compress many signal sources into a limited set of read-outs for frequently monitored key plant data. Signal validation and automatic selection of sensors with the most accurate signal ranges are also employed to reduce the number of control panel indicators. Information read-outs are by touch-screen to enhance operator interaction and include numeric parameter values, a bar form of analog display, and a plot trend. Various multi-range indicators are available on one display with automatic sensor selection and range display. The automatic calculation of a valid process representation parameter value, with the availability of individual sensor readings at the same display, avoids the need for separate backup displays, or different displays for normal operation versus accident or post-accident operation. Moreover, in another preferred feature of the invention, the parameter verification automatically distinguishes failed or multiple failed sensors, while allowing continued operation and accident mitigation information to the operator even if the CRT display is not available. Furthermore, the normal display information can be correlated to a qualified sensor, such as that used for post-accident monitoring purposes. At the information display level associated with control of specific components, dynamic "soft" controllers are provided with component status and control signal information necessary for operator control of these components. For the ESFC system, this information includes status lamp, on-off controls, modulation controls, open-closed controls, and logic controls. For the PCCS, the information includes confirm load, set points, operating range, process values, and control signal outputs. In the fourth level of the information hierarchy, dynamic CRT display pages are complementary to all levels of spatially dedicated control and information and can be accessed from any CRT location in the control room, technical support center, or emergency operations facility. These displays are grouped into a three level hierarchy that includes general monitoring (level 1), plant component and systems control (level 2), and component/process diagnostics (level 3). Display implementation is driven by the DPS and duplicates and verifies all discrete alarm and indicator processing performed in the DIAS. In the preferred implementation of the invention, the indicator, alarm, and control functions for a given major functional system of the plant are grouped together in a single, modularized panel. The panel can be made with cutouts that are spatially dedicated to each of the displays for the indicators, alarms, controls, and CRT, independent of the major plant functional system. This permits delivery, installation, and preliminary testing of the panels before finalization of the plant specific logic and algorithms, which can be software modified late in the plant construction schedule. This modularization is achievable because the space required on the panel is essentially independent of the major plant functional system to which the panel is dedicated. Both the alarms and indicators can be easily modified in software. The number of indicators and alarm tiles that can be displayed to the operator are not significantly limited by the available area of the panel, so that standardization of panel size and cutout locations for the display windows is possible.
	description	The present application hereby claims priority under 35 U.S.C. §119 on German patent application number DE 10 2010 062 192.7 filed Nov. 30, 2010, the entire contents of which are hereby incorporated herein by reference. At least one embodiment of the invention generally relates to a 2D collimator for a radiation detector and/or a method for manufacturing a 2D collimator of this kind. Scattered radiation is basically caused by the interaction between the object of interest and primary radiation emanating from the focus of a radiation source. Because of this interaction, it is incident on a radiation converter of a radiation detector from a different spatial direction from that of the primary radiation and causes artifacts in the reconstructed image. To reduce the detected scatter component in the detector signals, the radiation converters are therefore preceded by collimators. Such collimators have absorber elements whose surfaces are aligned radially to the focus of a radiation source in a fan-like manner so that only radiation from a spatial direction in line with the focus can be incident on the radiation detector. Even a slight tilt or incorrect positioning of the collimator relative to a radiation converter can cause shadowing of the active regions of the radiation converter, resulting in distortion, i.e. a reduction in the achievable signal-to-noise ratio. A particular challenge for designing a radiation detector is therefore to produce a collimator of very high mechanical strength so that positioning accuracies to within a few μm can be maintained. These stability requirements are particularly important when the collimator is used in a CT scanner, due to the centrifugal forces acting on the collimators during rotation. In addition, the radiation detectors increasingly have a higher z-coverage in order to enlarge the scan field of view. This increases the width to be spanned by the collimators in the z-direction, thereby increasing the risk of collimator instability. Due to the enlargement of the radiation detector in the z-direction and in the case of dual-source systems in which two source/detector systems disposed in one scanning plane and offset by a fixed angle in the φ-direction are operated simultaneously to obtain projections, not only scatter suppression along the φ-direction is required but also collimation in the z-direction. Collimators which suppress scatter in one spatial direction only, usually in the φ-direction, are termed one-dimensional (1D) collimators. Collimators producing a collimating effect in two spatial directions are accordingly known as two-dimensional (2D) collimators. To meet the stability requirements for a 1D collimator, in the known case as described in the publication DE 10 2007 051 306 A1, absorber elements aligned along a z-direction are segmented and mounted in a housing. Segmentation of the absorber elements is performed with the aim of reducing the manufacturing costs while at the same time meeting tighter engineering tolerances. The mechanical stability of the 1D collimator is provided by using a housing in which the plate-shaped absorber elements are precisely aligned and mounted. As a supporting structure, the housing comprises two bridge-like frame sections which are mechanically fixed by a plug-in connection. Housing shapes are also disclosed wherein the frame sections run alongside the absorber elements in each case. However, the disadvantage of both types of housing is that the frame sections are in the beam path of X-ray radiation to be detected. Due to the nature of their material, the frame sections cannot be completely transparent to X-ray radiation, which means that providing mechanical stability via the housing involves unwanted attenuation of the X-ray radiation and additional scatter generation. This disadvantage is particularly apparent in the case of bridge-shaped housings where the edges of the absorber elements are spanned by the frame sections in one plane. Circumferential frame sections also have the disadvantage that the absorber elements can only be lined up with pitch discontinuities because of an intervening wall. A 2D collimator is described, for example, in DE 10 2005 044 650 A1. It has a two-dimensional structure with cellular radiation channels. In the disclosed case, the lamellar absorber elements are interconnected cruciformly in a form-fit manner by corresponding slits in the absorber elements to be connected. 2D collimators are also known which are produced by laser sintering of radiation-absorbing metal powder or by stacking a plurality of cast or injection-molded individual gratings made of tungsten-powder-filled polymers. The 2D collimators are also segmented into individual 2D collimator modules to reduce the manufacturing cost/complexity and narrow the manufacturing tolerances, the segment size usually corresponding to the segment size of the radiation converter's detector tile mounted in a detector module. To construct the 2D collimator and produce a mechanically stable arrangement of the 2D collimator modules, these are glued directly to the respective detector tiles. However, in the event of a defect, glued-on 2D collimator modules cause warping both of the 2D collimator module and of the detector tiles, as nondestructive removal is generally no longer possible. In addition, the detector tiles are subjected to corresponding centrifugal forces by the glued-on 2D collimator modules during rotation. In at least one embodiment of the invention, a 2D collimator for a radiation detector is implemented, the collimator including high mechanical stability, so as to create the preconditions for easy, low-cost maintenance of the radiation detector while at the same time preventing detector signal interference caused by interaction with the 2D collimator. In at least one embodiment of the invention, a method is specified for producing such a 2D collimator. In at least one embodiment of the invention, a 2D collimator is disclosed for a radiation detector and a method is disclosed for producing a 2D collimator. Advantageous embodiments of the invention are set forth in the respective sub-claims. In at least one embodiment, the invention is based on the recognition that 2D collimator modules, with their cellular structure of radiation channels constituting radiation detector elements, have a very high intrinsic stability or rather intrinsic rigidity which can be used for constructing a bridge-like 2D collimator without using a supporting structure. At least one embodiment of the inventive 2D collimator for a radiation detector accordingly comprises 2D collimator modules arranged in series, wherein adjacent 2D collimator modules are glued together to establish a fixed mechanical connection to facing module sides, and wherein the outer 2D collimator modules on the free-remaining module side have a retaining element for mounting the 2D collimator opposite a detector mechanism. At least one embodiment of the invention is also achieved by an inventive method for producing a 2D collimator having at least above described 2D collimator modules disposed in a collimation direction, said method comprising: a) providing a plurality of the 2D collimator modules, b) applying a layer of adhesive to at least one side of the respective 2D collimator module, and c) inserting the 2D collimator elements in a precision tool at a position provided for the respective 2D collimator module. In the figures, parts producing an identical effect are provided with the same reference characters. In the case of recurring elements in a figure, in some cases only one element is provided with a reference character for reasons of clarity. The representations in the figures are schematic and not necessarily drawn to scale, and the scales may vary between figures. Various example embodiments will now be described more fully with reference to the accompanying drawings in which only some example embodiments are shown. Specific structural and functional details disclosed herein are merely representative for purposes of describing example embodiments. The present invention, however, may be embodied in many alternate forms and should not be construed as limited to only the example embodiments set forth herein. Accordingly, while example embodiments of the invention are capable of various modifications and alternative forms, embodiments thereof are shown by way of example in the drawings and will herein be described in detail. It should be understood, however, that there is no intent to limit example embodiments of the present invention to the particular forms disclosed. On the contrary, example embodiments are to cover all modifications, equivalents, and alternatives falling within the scope of the invention. Like numbers refer to like elements throughout the description of the figures. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments of the present invention. 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 when an element is referred to as being “connected,” or “coupled,” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected,” or “directly coupled,” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between,” versus “directly between,” “adjacent,” versus “directly adjacent,” etc.). The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of example embodiments of the invention. As used herein, the singular forms “a,” “an,” and “the,” are intended to include the plural forms as well, unless the context clearly indicates otherwise. As used herein, the terms “and/or” and “at least one of” include any and all combinations of one or more of the associated listed items. It will be further understood that the terms “comprises,” “comprising,” “includes,” and/or “including,” when used herein, 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. It should also be noted that in some alternative implementations, the functions/acts noted may occur out of the order noted in the figures. For example, two figures shown in succession may in fact be executed substantially concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Spatially relative terms, such as “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, term such as “below” can 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 are interpreted accordingly. Although the terms first, second, etc. may be used herein to describe various elements, components, regions, layers and/or sections, it should be understood that these elements, components, regions, layers and/or sections should not be limited by these terms. These terms are used only 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 the present invention. FIG. 1 shows the basic structure of a CT scanner 24. The CT scanner 24 comprises a radiation source 25 in the form of an X-ray tube from whose focus 26 an X-ray fan beam 27 emanates. The X-ray fan beam 27 penetrates an object of interest 28, or a patient, and is incident on a radiation detector 20, in this case an X-ray detector. The radiation source 25 and the radiation detector 20 are disposed opposite one another on a gantry (not shown here) of the CT scanner 24, said gantry being rotatable in a φ-direction about a system axis Z (=patient axis) of the CT scanner. The φ-direction therefore represents the circumferential direction of the gantry and the z-direction the longitudinal direction of the object of interest 28. During operation of the CT scanner 24, the radiation source 25 and the radiation detector 20 disposed on the gantry rotate around the object 28, X-ray images of the object 28 being obtained from different projection directions. For each X-ray projection, the radiation detector 20 is impinged by X-ray radiation which has passed through the object 28 causing it to be attenuated. The radiation converter 29 in turn generates signals corresponding to the intensity of the incident X-ray radiation. The radiation converter is subdivided into individual detector elements 30 for locally resolved capture of the X-ray radiation. In this concrete example embodiment, signal generation takes place in two stages using a photodiode array 31 which is optically linked to a scintillator array 32. It would likewise be possible to use a directly converting radiation detector based on a semiconductor material. From the signals captured by the radiation detector 20 in this way, a processing unit 33 then calculates in per se known manner one or more two- or three-dimensional images of the object which can be displayed on a display unit 34. The primary radiation emanating from the focus 26 of the radiation source 25 is scattered in the object 28 (among other things) in different spatial directions. In the detector element 30, this so-called secondary radiation produces signals which cannot be differentiated from the primary radiation signals required for image reconstruction. Unless further action is taken, the secondary radiation would therefore result in misinterpretations of the detected radiation and hence considerable impairment of the images obtained using the CT scanner 24. In order to limit the effect of the secondary radiation, using 2D collimators 1 according to an embodiment of the invention essentially only the portion of the X-ray radiation emanating from the focus, i.e. the primary radiation component, is allowed to pass unhindered to the radiation converter 20, whereas the secondary radiation is ideally completely absorbed by absorber surfaces of the absorber elements 13, 15 shown in FIG. 4 both in the φ-direction and in the z-direction. In FIG. 1 the radiation detector 20 is shown without a visible detector mechanism 11 in which the 2D collimators 1 and the radiation converter 20 are incorporated in a mutually decoupled manner. The design of the radiation detector 20 with the detector mechanism 11 will be explained in greater detail in connection with FIG. 3. The 2D collimator 1 according to an embodiment of the invention is shown in FIG. 2 in a perspective view. It comprises a total of four 2D collimator modules 2, 3 arranged is series in the z-direction. The 2D collimator modules 2, 3 are glued together at their respective end face, i.e. module side 5, typically using an epoxy adhesive. Because of the cellular structure and associated high intrinsic rigidity of the 2D collimator modules 2, 3, this glued connection 4 means that, even in the case of large widths to be spanned in the z-direction, the thus constructed 2D collimator 1 possesses a strength which, even during rotation of the CT scanner 24 when rotationally-induced centrifugal forces are applied, results in no interference in the detector signal due to shadowing effects. The intrinsic strength can also be increased still further by using special manufacturing processes. For example, a particularly high intrinsic strength can be achieved if the 2D collimator modules 2, 3 are produced in one piece using what is known as rapid manufacturing. This involves selective laser sintering using radiation-absorbing metal powder, e.g. of tungsten, molybdenum or tantalum. Facing module sides 5 are of different design as illustrated in FIG. 4 which shows a 2D collimator module 2 by way of example. Thus it would be possible, for example, in the case of adjacent 2D collimator modules 2, for an absorber surface 12 to be glued to edges 14 of absorber elements 15, i.e. connecting pieces, running perpendicularly thereto. However, facing module sides 5 of adjacent 2D collimator modules 2, 3 can also be of identical construction. The respective module side 5 can be delimited facewise by an absorber element 13 running parallel thereto, so that two absorber surfaces 12 are glued together in each case. Because of the large surfaces, a very firm connection 4 is established between adjacent 2D collimator modules 2, 3. The edge absorber elements 13 which are bonded together can be made smaller than the absorber elements inside the 2D collimator module 2, 3 in order to compensate for the added thickness in the assembled state and can be typically only half as thick as adjacent absorber elements. Located at the free module sides 6 are angled retaining elements 7 which are attached to the respective module side 6 by a glued connection 4. The 2D collimator 1 is aligned and connected to the detector mechanism 11 via the retaining elements 7. The retaining element 7 comprises corresponding fastening devices 8 and adjustment devices 9, 10. In this example, a drilled hole 8 is used to fasten the 2D collimator 1 to the detector mechanism 11 via a screwed connection. A bearing surface 10 disposed on the underside of the respective retaining element 7 is used to adjust, i.e. align, the 2D collimator 1 in the radiation incidence direction 18. The external contour 9 of the retaining element 7 provides at least one device for adjusting or more specifically aligning the 2D collimator 1 in the z-direction and in the p-direction. Other forms of adjustment or fastening are self-evidently also conceivable. The 2D collimator 1 can be easily manufactured by a tool in which recesses are provided for precise positioning of the 2D-collimation modules 2, 3. The recesses are implemented such that, by inserting the 2D collimator element 2, 3 corresponding to the recess, alignment is effected such that, in the installed state, the radiation channels 35 are aligned to the focus 26 of the radiation source 25. FIG. 3 shows a perspective view of a section of the radiation detector 20 with a 2D collimator 1 according to an embodiment of the invention incorporated therein. The radiation detector 20 is subdivided into different detector modules 22, the term detector module 22 being understood as meaning the 2D collimator 1 and radiation converter module 21 as an entity. The radiation converter module 21 is in turn segmented into different detector tiles 23 which are disposed in a row in series along the z-direction. The 2D collimator 1 spans the entire radiation converter module 21 in the z-direction in a self-supporting manner. Each 2D collimator module 2, 3 is aligned to a specific detector tile 23 of the radiation converter module 21. The 2D collimator 1 is aligned in the radiation incidence direction 18 via the respectively provided bearing surface 10 of the retaining element 7, which bearing surface rests against a supporting surface 19 of precisely dimensioned pins 36. The fastening can be established by way of a screwed connection via the hole 8 drilled in the respective retaining element 7, into which hole a screw 37 disposed on the detector mechanism 11 engages. The external contour 9 of the respective retaining element 7, which contour is used as at least one device of adjustment in the z-direction and in the φ-direction, engages in corresponding recesses 38 in the detector mechanism 11. The radiation converter module 21 is incorporated in the detector mechanism 11 in a decoupled manner from the 2D collimator 1, thereby facilitating replacement of the respective component 1, 21. An embodiment of the inventive 2D collimator for a radiation detector accordingly comprises 2D collimator modules arranged in series, wherein adjacent 2D collimator modules are glued together to establish a fixed mechanical, connection to facing module sides, and wherein the outer 2D collimator modules on the free-remaining module side have a retaining element for mounting the 2D collimator opposite a detector mechanism. Different spatial arrangements of the 2D collimator elements are conceivable here. In the simplest case, a plurality of 2D collimator modules are arranged one-dimensionally in series in a row in the z-direction. The directions specified in respect of the 2D collimator relate to a normally used coordinate system of the CT scanner for correct use of the 2D collimator in the installed condition. As the 2D collimator modules are glued directly to one another, no additional supporting structures are required for producing a required rigidity, i.e. mechanical stability, thereby enabling positioning accuracies to within a few micrometers to be maintained during rotation of a CT scanner. In particular, no housing with bridge-like or circumferential frame sections is necessary. As a result, in comparison to the known collimators of bridge-type design, artifacts or disturbances in the detector signals caused by interaction of the incident radiation with the supporting elements are completely eliminated. Glued connections can be implemented with layer thicknesses of a few nanometers, so that the resulting gap between the 2D collimator modules has no measurable negative effect on signal generation. Dispensing with the housing also means that the 2D collimator is less expensive to manufacture because of the lower complexity. In addition, a continuous pitch of the 2D collimator modules disposed in the arc direction, i.e. φ-direction, can be achieved. The 2D collimator decoupled from the radiation converter is integrated into the radiation detector by way of the retaining elements provided at the edge. There is therefore no fixed mechanical connection between the radiation converter and the 2D collimator, thus making it possible to replace one component without destroying the respective other component. The 2D collimator according to an embodiment of the invention therefore also reduces the maintenance work involved in replacing a component. The module sides are preferably implemented such that an absorber surface, running parallel to the module side, of an absorber element of one 2D collimator module is glued to edges of perpendicularly thereto running absorber elements of the other 2D collimator module. In this context, an absorber element is to be understood as meaning a plate-like or lamellar basic element with which scattered radiation in respect of a direction running perpendicular to its surface is reduced for a row of detector elements of one detector element side. With this configuration, in particular an unbroken structure running continuously over the collimation direction can be produced in which no dead zones or heavy shadowing occur at seams or joints between adjacent 2D collimator modules. Alternatively, the sides of the modules are preferably implemented such that absorber surfaces, running parallel to the module side, of an absorber element of the 2D collimator modules are glued together. In this case the contact surface and therefore the achievable strength of the connection between the 2D collimator modules is maximized. To prevent unwanted shadowing of the radiation converter at the interface between the 2D collimator modules, the connecting pieces, i.e. the absorber elements, used to establish a connection can be made half as thick as the absorber elements disposed in the inner region of the 2D collimator module. In an advantageous embodiment of the invention, for mutually aligning the adjacent 2D collimator modules, at least one projection is disposed on one facing module side, said projection engaging in at least one recess in the corresponding other module side, thereby ensuring simple and at the same time precise mutual alignment of the 2D collimator modules. The respective retaining element has at least one fastening device for fixing the 2D collimator to a detector mechanism and/or as at least one adjustment device for positioning the 2D collimator in the collimation direction with respect to the detector mechanism, preferably in the form of a drilled hole. At least one device for fastening and/or adjustment can therefore be implemented in a simple and high-precision manner. Adjustment with respect to the detector mechanism would be possible, for example, using at least one alignment device in the form of a guide pin, whereas the position of the 2D collimator module can be simultaneously fixed by a screwed connection when it is in the aligned state. The respective retaining element preferably has a bearing surface as an adjustment device for positioning the 2D collimator with respect to a detector mechanism in a radiation incidence direction, the bearing surface coming to rest against a support surface of the detector when the 2D collimator is incorporated in a detector mechanism in the radiation incidence direction. Such a bearing surface constitutes a particularly easy to implement at least one adjustment device which can be produced with very tight manufacturing tolerances. In another advantageous embodiment of the invention, at least the outer 2D collimator modules are manufactured in one piece with the retaining elements. This allows the 2D collimator modules to be produced in a single manufacturing process, reduces the design complexity and increases collimator stability. The 2D collimator modules are preferably produced in a rapid manufacturing process, preferably by selective laser sintering. Rapid manufacturing is a manufacturing process in which a component is built up layer by layer from powder material using physical and/or chemical effects. In each production step, a new layer can be applied selectively, very precisely and thinly to the existing structure, so that the absorber elements can be produced with great accuracy in terms of their width, height and position. This process is based on layer data which can be easily generated directly from 3D surface data of the kind available in CAD systems. At least one embodiment of the invention is also achieved by an inventive method for producing a 2D collimator having at least above described 2D collimator modules disposed in a collimation direction, said method comprising: a) providing a plurality of the 2D collimator modules, b) applying a layer of adhesive to at least one side of the respective 2D collimator module, and c) inserting the 2D collimator elements in a precision tool at a position provided for the respective 2D collimator module. If the outer 2D collimator modules cannot be produced with the retaining elements as a single element, at least one embodiment of the method advantageously comprises: d) Gluing the retaining elements to the outer 2D collimator modules. In at least one embodiment, Step a) advantageously also comprises: a1) Producing the 2D collimator modules using a rapid manufacturing process, preferably by selective laser sintering. To summarize: At least one embodiment of the invention relates to a 2D collimator 1 for a radiation detector 20 with 2D collimator modules 2, 3 arranged in series, wherein adjacent 2D collimator modules 2, 3 are glued together to establish a fixed mechanical connection 4 to facing module sides 5, and wherein, on their free-remaining side 6, the outer 2D collimator modules 3 have a retaining element 7 for mounting the 2D collimator 1 opposite a detector mechanism 11. This creates the preconditions for decoupled integration into the radiation detector 20 with respect to the radiation converter module 21 and therefore for low-cost/complexity maintenance of the radiation detector 20 while at the same time preventing detector signal interference caused by the interaction of incident radiation with the 2D collimator 1. At least one embodiment of the invention also relates to method for manufacturing such a 2D collimator 1. The patent claims filed with the application are formulation proposals without prejudice for obtaining more extensive patent protection. The applicant reserves the right to claim even further combinations of features previously disclosed only in the description and/or drawings. The example embodiment or each example embodiment should not be understood as a restriction of the invention. Rather, numerous variations and modifications are possible in the context of the present disclosure, in particular those variants and combinations which can be inferred by the person skilled in the art with regard to achieving the object for example by combination or modification of individual features or elements or method steps that are described in connection with the general or specific part of the description and are contained in the claims and/or the drawings, and, by way of combinable features, lead to a new subject matter or to new method steps or sequences of method steps, including insofar as they concern production, testing and operating methods. References back that are used in dependent claims indicate the further embodiment of the subject matter of the main claim by way of the features of the respective dependent claim; they should not be understood as dispensing with obtaining independent protection of the subject matter for the combinations of features in the referred-back dependent claims. Furthermore, with regard to interpreting the claims, where a feature is concretized in more specific detail in a subordinate claim, it should be assumed that such a restriction is not present in the respective preceding claims. Since the subject matter of the dependent claims in relation to the prior art on the priority date may form separate and independent inventions, the applicant reserves the right to make them the subject matter of independent claims or divisional declarations. They may furthermore also contain independent inventions which have a configuration that is independent of the subject matters of the preceding dependent claims. Further, elements and/or features of different example embodiments may be combined with each other and/or substituted for each other within the scope of this disclosure and appended claims. Example embodiments being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as a departure from the spirit and scope of the present invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
	description	This application claims priority of Chinese Patent Application No. 201810960823.7 filed on Aug. 22, 2018, the entire contents of each of which are hereby incorporated by reference. The present disclosure generally relates to a medical treatment system and more specifically relates to methods and systems for correcting position errors of one or more leaves of a multi-leaf collimator (MLC) in a radiotherapy procedure. A multi-leaf collimator (MLC) is widely used for collimating radiation beams emitted from a radiation source in radiotherapy systems. The radiation beams collimated by an MLC may be projected to the tumor and an area formed by the projected radiation beams may comply with the shape of a tumor to prevent healthy tissues around the tumor from being radiated. Therefore, the positioning accuracy of the leaves in the MLC is important for precise radiotherapy. At present, the leaves of an MLC are generally driven by motors to move forward or backward to scale a radiation field. In this way, position errors may occur in the movement of the leaves and influence the positioning accuracy of the leaves in the MLC. It is desirable to provide systems and methods for determining an offset value of each of the leaves in an MLC for correcting the position errors. According to an aspect of the present disclosure, a method for correcting position error of a leaf is provided. The method may be implemented on at least one machine each of which has at least one processor and storage. The method may include determining a first position for each of the plurality of leaves, information associated with the first position including a first movement direction and a first angle, wherein a movement of the each of the plurality of leaves along the first movement direction is configured to move toward or away from a center of the radiation field; determining an offset value associated with the first position based on the first angle and the first movement direction; and determining a target position of the each of the plurality of leaves based on the offset value. In some embodiments, the determining a first position for each of the plurality of leaves may include obtaining an angle of a gantry corresponding to the first position of the each of the plurality of leaves; obtaining an angle of a collimator corresponding to the first position of the each of the plurality of leaves, wherein the MLC is mounted in the collimator and rotates along with the collimator; and determining the first angle of the each of the plurality of leaves based on the angle of the gantry and the angle of the collimator. In some embodiments, the determining a first position for each of the plurality of leaves may include obtaining a first velocity relating to the driving component; in response to a determination that the first velocity relating to the driving component is lower than a first threshold, determining the first movement direction as a backward movement direction, the each of the plurality of leaves being configured to move away from the center of the radiation field along the backward movement direction; and in response to a determination that the first velocity relating to the driving component is greater than a second threshold, determining the first movement direction as a forward movement direction, the each of the plurality of leaves being configured to move toward the center of the radiation field along the forward movement direction. In some embodiments, the determining a target position of the each of the plurality of leaves based on the offset value may include subtracting the offset value from a preprogrammed position of the each of the plurality of leaves. In some embodiments, the information associated with the first position may include a first main encoder value, and the determining an offset value associated with the first position based on the first angle and the first movement direction may include obtaining a first reference offset value associated with the first position of the each of the plurality of leaves from a pre-determined offset table; obtaining a first main encoder value corresponding to the first position of the each of the plurality of leaves, the first main encoder value being acquired by the main encoder; obtaining a second main encoder value corresponding to a second position of the each of the plurality of leaves, the second main encoder value being acquired by the main encoder, and the second position being a position at where a movement direction of the each of the plurality of leaves changes from a second movement direction to the first movement direction; and determining the offset value associated with the first position based on the first movement direction, the first reference offset value, and a difference between the first main encoder value and the second main encoder value. In some embodiments, the determining the offset value associated with the first position based on the first movement direction, the first reference offset value, and a difference between the first main encoder value and the second main encoder value may include if the each of the plurality of leaves moves away from the center of the radiation field along the first movement direction, designating a minimum value between the first reference offset value and a sum of the difference between the first main encoder value and the second main encoder value and a second reference offset value associated with the second position as the offset value associated with the first position; and if the each of the plurality of leaves moves toward the center of the radiation field along the first movement direction, designating a maximum value between the first reference offset value and a sum of the second reference offset value associated with the second position and the difference between the first main encoder value and the second main encoder value as the offset value associated with the first position. In some embodiments, the determining an offset value associated with the first position based on the first angle and the first movement direction may include if the each of the plurality of leaves moves toward the center of the radiation field along the first movement direction and the first angle is equal to 0 degrees, designating the offset value associated with the first position as 0. In some embodiments, the determining a target position of the each of the plurality of leaves based on the offset value may include obtaining a first main encoder value corresponding to a first position of each of the plurality of leaves acquired by the main encoder; and correcting the first main encoder value based on the offset value to obtain the target position of the each of the plurality of leaves. In some embodiments, the correcting the first main encoder value based on the offset value may include adding the offset value to the first main encoder value to obtain the target position of the each of the plurality of leaves. According to an aspect of the present disclosure, a method for correcting position error of a leaf is provided. The method may be implemented on at least one machine each of which has at least one processor and storage. The method may include determining a first position for each of the plurality of leaves, information associated with the first position including a first movement phase, wherein a movement of the each of the plurality of leaves moves in the first movement phase is configured to move toward or away from a center of the radiation field; determining an offset value associated with the first position based on the first movement phase; and determining a target position of the each of the plurality of leaves based on the offset value. According to an aspect of the present disclosure, a system for correcting position errors for a multi-leaf collimator (MLC) is provided. The MLC may include a plurality of leaves to shape a radiation field, each of the plurality of leaves being associated with a driving component including a main encoder. The system may include at least one storage device storing executable instructions, and at least one processor in communication with the at least one storage device, when executing the executable instructions, causing the system to determine a first position for each of the plurality of leaves, information associated with the first position including a first movement direction and a first angle, wherein the each of the plurality of leaves moves toward or away from a center of the radiation field along the first movement direction; determine an offset value associated with the first position based on the first angle and the first movement direction; and determine a target position of the each of the plurality of leaves based on the offset value. Additional features will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following and the accompanying drawings or may be learned by production or operation of the examples. The features of the present disclosure may be realized and attained by practice or use of various aspects of the methodologies, instrumentalities and combinations set forth in the detailed examples discussed below. In the following detailed description, numerous specific details are set forth by way of examples in order to provide a thorough understanding of the relevant disclosure. However, it should be apparent to those skilled in the art that the present disclosure may be practiced without such details. In other instances, well-known methods, procedures, systems, components, and/or circuitry have been described at a relatively high-level, without detail, in order to avoid unnecessarily obscuring aspects of the present disclosure. Various modifications to the disclosed embodiments will be readily apparent to those skilled in the art, and the general principles defined herein may be applied to other embodiments and applications without departing from the spirit and scope of the present disclosure. Thus, the present disclosure is not limited to the embodiments shown, but to be accorded the widest scope consistent with the claims. The terminology used herein is for the purpose of describing particular example embodiments only and is not intended to be limiting. As used herein, the singular forms “a,” “an,” and “the” may be intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprise,” “comprises,” and/or “comprising,” “include,” “includes,” and/or “including,” 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. It will be understood that the term “system,” “engine,” “unit,” “module,” and/or “block” used herein are one method to distinguish different components, elements, parts, section or assembly of different level in ascending order. However, the terms may be displaced by another expression if they achieve the same purpose. Generally, the word “module,” “unit,” or “block,” as used herein, refers to logic embodied in hardware or firmware, or to a collection of software instructions. A module, a unit, or a block described herein may be implemented as software and/or hardware and may be stored in any type of non-transitory computer-readable medium or another storage device. In some embodiments, a software module/unit/block may be compiled and linked into an executable program. It will be appreciated that software modules can be callable from other modules/units/blocks or themselves, and/or may be invoked in response to detected events or interrupts. Software modules/units/blocks configured for execution on computing devices (e.g., processor 310 as illustrated in FIG. 3) may be provided on a computer-readable medium, such as a compact disc, a digital video disc, a flash drive, a magnetic disc, or any other tangible medium, or as a digital download (and can be originally stored in a compressed or installable format that needs installation, decompression, or decryption prior to execution). Such software code may be stored, partially or fully, on a storage device of the executing computing device, for execution by the computing device. Software instructions may be embedded in firmware, such as an EPROM. It will be further appreciated that hardware modules/units/blocks may be included in connected logic components, such as gates and flip-flops, and/or can be included of programmable units, such as programmable gate arrays or processors. The modules/units/blocks or computing device functionality described herein may be implemented as software modules/units/blocks but may be represented in hardware or firmware. In general, the modules/units/blocks described herein refer to logical modules/units/blocks that may be combined with other modules/units/blocks or divided into sub-modules/sub-units/sub-blocks despite their physical organization or storage. The description may apply to a system, an engine, or a portion thereof. It will be understood that when a unit, engine, module or block is referred to as being “on,” “connected to,” or “coupled to,” another unit, engine, module, or block, it may be directly on, connected or coupled to, or communicate with the other unit, engine, module, or block, or an intervening unit, engine, module, or block may be present, unless the context clearly indicates otherwise. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. These and other features, and characteristics of the present disclosure, as well as the methods of operation and functions of the related elements of structure and the combination of parts and economies of manufacture, may become more apparent upon consideration of the following description with reference to the accompanying drawings, all of which form a part of this disclosure. It is to be expressly understood, however, that the drawings are for the purpose of illustration and description only and are not intended to limit the scope of the present disclosure. It is understood that the drawings are not to scale. The flowcharts used in the present disclosure illustrate operations that systems implement according to some embodiments of the present disclosure. It is to be expressly understood, the operations of the flowcharts may be implemented not in order. Conversely, the operations may be implemented in inverted order, or simultaneously. Moreover, one or more other operations may be added to the flowcharts. One or more operations may be removed from the flowcharts. Provided herein are systems and components for medical diagnostic and/or treatment. In some embodiments, the diagnostic and treatment system may include a radiotherapy system. The radiotherapy system may include a treatment plan system (TPS), an image-guided radiotherapy (IGRT) system, etc. Merely by way of example, the image-guided radiotherapy (IGRT) system may include, for example, a CT guided radiotherapy system, an MRI guided radiotherapy system, etc. An aspect of the present disclosure relates to a system and method for correcting position errors of a multi-leaf collimator (MLC) including a plurality of leaves to scale a radiation field. For each of the plurality of leaves, the system may determine a current position denoted by a first movement direction and a first angle. The movement of each of the plurality of leaves along the first movement direction may be configured to expand or narrow the radiation field. Then, an offset value associated with the current position of the each of the plurality of leaves may be determined based on the first angle of the each of the plurality of leaves and the first movement direction of the each of the plurality of leaves. Further, a target position of the each of the plurality of leaves may be determined according to the offset value and the first movement direction of the leaf. It should be noted that the radiotherapy system 100 described below is merely provided for illustration purposes, and not intended to limit the scope of the present disclosure. For persons having ordinary skills in the art, a certain amount of variations, changes, and/or modifications may be deducted under the guidance of the present disclosure. Those variations, changes, and/or modifications do not depart from the scope of the present disclosure. FIG. 1 is a schematic diagram illustrating an exemplary radiotherapy system 100 according to some embodiments of the present disclosure. As shown, the radiotherapy system 100 may include a radiotherapy device 110, a processing device 120, storage device 130, one or more terminal(s) 140, and a network 150. In some embodiments, the radiotherapy device 110, the processing device 120, the storage device 130, and/or the terminal(s) 140 may be connected to and/or communicate with each other via a wireless connection (e.g., the network 150), a wired connection, or a combination thereof. The connections between the components in the radiotherapy system 100 may vary. Merely by way of example, the radiotherapy device 110 may be connected to the processing device 120 through the network 150, as illustrated in FIG. 1. As another example, the radiotherapy device 110 may be connected to the processing device 120 directly. As a further example, the storage device 130 may be connected to the processing device 120 through the network 150, as illustrated in FIG. 1, or connected to the processing device 120 directly. As still a further example, the terminal(s) 140 may be connected to the processing device 120 through the network 150, as illustrated in FIG. 1, or connected to the processing device 120 directly (as indicated by the bidirectional arrow in the dashed line shown in FIG. 1), or connected to the radiotherapy device 110 directly or through the network 150. The terminal(s) 140 may be omitted. The radiotherapy device 110 may perform radiotherapy treatment on at least one part of a subject. In some embodiments, the radiotherapy device 110 may include a single modality apparatus, for example, an X-ray therapy apparatus, a Co-60 teletherapy apparatus, a medical electron accelerator, etc. In some embodiments, the radiotherapy device 110 may be a multi-modality (e.g., two-modality) apparatus to acquire a medical image relating to the at least one part of the subject and perform radiotherapy treatment on the at least one part of the subject. The subject may be biological or non-biological. For example, the subject may include a patient, a man-made object, etc. As another example, the subject may include a specific portion, organ, and/or tissue of the patient. For example, the subject may include head, neck, thorax, cardiac, stomach, blood vessel, soft tissue, tumor, nodules, or the like, or a combination thereof. In some embodiments, the subject may include a region of interest (ROI), such as a tumor, a node, etc. In some embodiments, the radiotherapy device 110 may include a gantry to which a treatment head may be connected. The treatment head may include a radiation source 112 and a multi-leaf collimator (MLC) 114. The radiation source 112 may emit radiation beams to a subject. The MLC 114 may be configured to collimate radiation beams emitted from the radiation source 112. In some embodiments, the MLC 114 may include a plurality of leaves to shape a radiation field. The plurality of leaves may be driven by one or more driving components (e.g., motors) to move to specific positions to expand or narrow the radiation field. Due to the mechanical factors associated with the one or more driving components (e.g., motors), an actual position of a leaf may be inconsistent with an ideal position specified by the processing device 120, which causes a position error between the actual position and the ideal position. The position error of a leaf may be caused by the backlash error relating to a driving component, leaf deformation, leaf positioning, horizontality, perpendicularity, or the like, or a combination thereof. In some embodiments, the driving component may include a main encoder configured to acquire a position of a leaf driven by the driving component. The main encoder may be used to determine the position of the leaf based on a parameter (e.g., a rotation velocity, a rotation count, etc.) of the driving component (e.g., a motor). For example, if the diving component drives a leaf to move via rotations of a motor, the main encoder may be configured to acquire a rotation count of the motor. Then, the position of the leaf may be determined based on the rotation count of the motor. In some embodiments, the main encoder may include an encoder of the motor (also referred to as motor encoder), a potentiometer mounted on the shaft end of the motor, or any other position measurement device. In some embodiments, each of the plurality of leaves in the MLC 114 may be coupled with an auxiliary encoder. In some embodiments, the auxiliary encoder may include a grating-rule displacement sensor, a Hall sensor, a potentiometer or any other position measurement device. More descriptions of the position measurement device may be found, for example, Chinese Application No. 201510581866.0, the contents of which are hereby incorporated by reference. More descriptions of the MLC 114 may be found elsewhere in the present disclosure (e.g., FIG. 2 and the descriptions thereof). The processing device 120 may process data and/or information obtained from the radiotherapy device 110, the storage device 130, and/or the terminal(s) 140. For example, the processing device 120 may determine a movement direction of a leaf in the MLC 114. As another example, the processing device 120 may determine an angle of each of a plurality of leaves based on an angle of the gantry of the radiotherapy device 110 and an angle of a collimator. As a further example, the processing device 120 may determine an offset value of each of a plurality of leaves in the MLC 114, and determine a position of the each of a plurality of leaves in the MLC 114 according to the determined offset value. In some embodiments, the processing device 120 may be a single server or a server group. The server group may be centralized or distributed. In some embodiments, the processing device 120 may be local or remote. For example, the processing device 120 may access information and/or data from the radiotherapy device 110, the storage device 130, and/or the terminal(s) 140 via the network 150. As another example, the processing device 120 may be directly connected to the radiotherapy device 110, the terminal(s) 140, and/or the storage device 130 to access information and/or data. In some embodiments, the processing device 120 may be implemented on a cloud platform. For example, the cloud platform may include a private cloud, a public cloud, a hybrid cloud, a community cloud, a distributed cloud, an inter-cloud, a multi-cloud, or the like, or a combination thereof. In some embodiments, the processing device 120 may be implemented by a mobile device 400 having one or more components as described in connection with FIG. 4. The storage device 130 may store data, instructions, and/or any other information. In some embodiments, the storage device 130 may store data obtained from the radiotherapy device 110, the processing device 120, and/or the terminal(s) 140. In some embodiments, the storage device 130 may store data and/or instructions that the processing device 120 may execute or use to perform exemplary methods described in the present disclosure. In some embodiments, the storage device 130 may include a mass storage, removable storage, a volatile read-and-write memory, a read-only memory (ROM), or the like, or any combination thereof. Exemplary mass storage may include a magnetic disk, an optical disk, a solid-state drive, etc. Exemplary removable storage may include a flash drive, a floppy disk, an optical disk, a memory card, a zip disk, a magnetic tape, etc. Exemplary volatile read-and-write memory may include a random access memory (RAM). Exemplary RAM may include a dynamic RAM (DRAM), a double date rate synchronous dynamic RAM (DDR SDRAM), a static RAM (SRAM), a thyristor RAM (T-RAM), and a zero-capacitor RAM (Z-RAM), etc. Exemplary ROM may include a mask ROM (MROM), a programmable ROM (PROM), an erasable programmable ROM (EPROM), an electrically erasable programmable ROM (EEPROM), a compact disk ROM (CD-ROM), and a digital versatile disk ROM, etc. In some embodiments, the storage device 130 may be implemented on a cloud platform as described elsewhere in the disclosure. In some embodiments, the storage device 130 may be connected to the network 150 to communicate with one or more other components in the radiotherapy system 100 (e.g., the processing device 120, the terminal(s) 140, etc.). One or more components in the radiotherapy system 100 may access the data or instructions stored in the storage device 130 via the network 150. In some embodiments, the storage device 130 may be part of the processing device 120. The terminal(s) 140 may be connected to and/or communicate with the radiotherapy device 110, the processing device 120, and/or the storage device 130. For example, the terminal(s) 140 may obtain a processed image from the processing device 120. As another example, the terminal(s) 140 may obtain image data acquired via the radiotherapy device 110 and transmit the image data to the processing device 120 to be processed. In some embodiments, the terminal(s) 140 may include a mobile device 140-1, a tablet computer 140-2, . . . , a laptop computer 140-N, or the like, or any combination thereof. For example, the mobile device 140-1 may include a mobile phone, a personal digital assistant (PDA), a gaming device, a navigation device, a point of sale (POS) device, a laptop, a tablet computer, a desktop, or the like, or any combination thereof. In some embodiments, the terminal(s) 140 may include an input device, an output device, etc. The input device may include alphanumeric and other keys that may be input via a keyboard, a touch screen (for example, with haptics or tactile feedback), a speech input, an eye tracking input, a brain monitoring system, or any other comparable input mechanism. The input information received through the input device may be transmitted to the processing device 120 via, for example, a bus, for further processing. Other types of the input device may include a cursor control device, such as a mouse, a trackball, or cursor direction keys, etc. The output device may include a display, a speaker, a printer, or the like, or a combination thereof. In some embodiments, the terminal(s) 140 may be part of the processing device 120. The network 150 may include any suitable network that can facilitate the exchange of information and/or data for the radiotherapy system 100. In some embodiments, one or more components of the radiotherapy system 100 (e.g., the radiotherapy device 110, the processing device 120, the storage device 130, the terminal(s) 140, etc.) may communicate information and/or data with one or more other components of the radiotherapy system 100 via the network 150. For example, the processing device 120 may obtain image data from the radiotherapy device 110 via the network 150. As another example, the processing device 120 may obtain user instruction(s) from the terminal(s) 140 via the network 150. The network 150 may be and/or include a public network (e.g., the Internet), a private network (e.g., a local area network (LAN), a wide area network (WAN)), etc.), a wired network (e.g., an Ethernet network), a wireless network (e.g., an 802.11 network, a Wi-Fi network, etc.), a cellular network (e.g., a Long Term Evolution (LTE) network), a frame relay network, a virtual private network (VPN), a satellite network, a telephone network, routers, hubs, switches, server computers, and/or any combination thereof. For example, the network 150 may include a cable network, a wireline network, a fiber-optic network, a telecommunications network, an intranet, a wireless local area network (WLAN), a metropolitan area network (MAN), a public telephone switched network (PSTN), a Bluetooth™ network, a ZigBee™ network, a near field communication (NFC) network, or the like, or any combination thereof. In some embodiments, the network 150 may include one or more network access points. For example, the network 150 may include wired and/or wireless network access points such as base stations and/or internet exchange points through which one or more components of the radiotherapy system 100 may be connected to the network 150 to exchange data and/or information. This description is intended to be illustrative, and not to limit the scope of the present disclosure. Many alternatives, modifications, and variations will be apparent to those skilled in the art. The features, structures, methods, and other characteristics of the exemplary embodiments described herein may be combined in various ways to obtain additional and/or alternative exemplary embodiments. For example, the storage device 130 may be a data storage including cloud computing platforms, such as public cloud, private cloud, community, and hybrid clouds, etc. As another example, the radiotherapy system 100 may further include a treatment planning system. However, those variations and modifications do not depart the scope of the present disclosure. FIG. 2A is a schematic diagram illustrating an exemplary multi-leaf collimator (MLC) 200 according to some embodiments of the present disclosure. As shown in FIG. 2A, the MLC 200 may include a leaf assembly 220 including multiple leaves (e.g., a leaf 220-1, a leaf 220-2, a leaf 220-i, . . . , a leaf 220-n, etc.), a carriage 240, and a driving assembly 260. Each of the multiple leaves may move in the carriage 240 independently, for example, move toward the center of a radiation field or move away from the center of the radiation field. The center of the radiation field may be a geometric center of the radiation field formed by the multiple leaves. The driving assembly 260 may include multiple motors (e.g., a motor 260-1, a motor 260-2, a motor 260-i, . . . , a motor 260-n, etc.) associated with the multiple leaves (e.g., the leaf 220-1, the leaf 220-2, the leaf 220-i, . . . , the leaf 220-n, etc.). Each of the multiple motors (e.g., the motor 260-1, the motor 260-2, the motor 260-i, . . . , the motor 260-n, etc.) may drive a corresponding leaf of the multiple leaves (e.g., the leaf 220-1, the leaf 220-2, the leaf 220-i, . . . , the leaf 220-n, etc.) to move independently in the carriage 240 to form the radiation field. Each one of the multiple leaves (e.g., the leaf 220-1, the leaf 220-2, the leaf 220-i, . . . , the leaf 220-n, etc.) may be driven by one of the multiple motors to move toward the center of the radiation field. Further, each one of the multiple leaves (e.g., the leaf 220-1, the leaf 220-2, the leaf 220-i, . . . , the leaf 220-n, etc.) may be driven by one of the multiple motors to move away from the center of the radiation field. FIG. 2B is a schematic diagram illustrating an exemplary control system of a multi-leaf collimator (MLC) according to some embodiments of the present disclosure. As illustrated in FIG. 2B, the control system may include a compensation table module 202, a position controller 204, a velocity controller 206, a velocity feedback module 216, and a position feedback module 218. The position controller 204 may receive a position command associated with a preprogrammed position of a leaf 214 and control the velocity controller 206 based on the position command. In some embodiments, the position command associated with the predetermined position of the leaf 214 may be modified based on an offset value associated with the leaf 214. The offset value associated with the leaf 214 may be determined by and/or obtained from the compensation table module 202 according to process 800 as illustrated in FIG. 8. In some embodiments, the offset value associated with the leaf 214 may be determined based on the current position of the leaf 214 acquired by the position feedback module 218 from an encoder 208 (or a main encoder) and/or a Hall sensor 212 (or an auxiliary encoder) according to process 900 as illustrated in FIG. 9 and/or process 1000 as illustrated in FIG. 10. The velocity controller 206 may determine a reference rotation velocity of the motor 210 based on the position command. The velocity controller 206 may control the motor 210 to rotate according to the reference rotation velocity within a certain time to drive the leaf 214 associated with the motor 210 to move to the preprogrammed position. The encoder 208 associated with the motor 210 may acquire a rotation count of the motor 210 to determine a current or real-time position of the leaf 214. The Hall sensor 212 may be coupled with the leaf 214. The Hall sensor 212 may acquire a signal relating to the current or real-time position of the leaf 214. The velocity feedback module 216 may be connected with the encoder 208 and/or the Hall sensor 212. The velocity feedback module 216 may be configured to determine a current rotation velocity of the motor 210 based on the rotation of the motor 210 acquired by the encoder 208. Alternatively or simultaneously, the velocity feedback module 216 may be configured to determine a current movement velocity of the leaf 214 based on the signal acquired by the Hall sensor 212. The velocity controller 206 may acquire the current rotation velocity of the motor 210 and/or the current movement velocity of the leaf 214. Then, the velocity controller 206 may modify the driving force of the motor 210 based on a difference between a referenced rotation velocity and the current rotation velocity of the motor 210 and/or a difference between a reference velocity and the current movement velocity of the leaf 214. The position feedback module 218 may be connected to the encoder 208 and/or the Hall sensor 212. The position feedback module 218 may be configured to determine a current position of the leaf 214 based on the rotation count of the motor 210 acquired by the encoder 208 and a current movement velocity of the leaf 214 based on magnetic field fluctuation signals acquired by the Hall sensor 212 (e.g., an encoder). Then, the position controller 204 may determine an offset value associated with the current position of the leaf 214 determined based on the encoder 208 and/or the Hall sensor 212. The position command may be modified based on the offset value associated with the current position of the leaf 214. More descriptions of the MLC may be found, for example, Chinese Publication No. 104667427A entitled “Leaf position monitoring device for a multi-leaf collimator (MLC), an MLC, and a radiotherapy device (, , , ).”, the contents of which are hereby incorporated by reference. This description is intended to be illustrative, and not to limit the scope of the present disclosure. Many alternatives, modifications, and variations will be apparent to those skilled in the art. The features, structures, methods, and other characteristics of the exemplary embodiments described herein may be combined in various ways to obtain additional and/or alternative exemplary embodiments. For example, the compensation table module 202 may be omitted. As another example, the velocity feedback module 216 and the position feedback module 218 may be integrated into one single module. However, those variations and modifications do not depart the scope of the present disclosure. FIG. 2C is a section diagram illustrating an exemplary treatment head of a radiotherapy device according to some embodiments of the present disclosure. As illustrated in FIG. 2C, the treatment head may include a radiation source 230, a primary collimator 250, one or more filters 252, a collimator 290, or any other component (e.g., a chamber between the one or more filters 252 and the collimator 290). The radiation source 230 may generate and/or emit radiation beams to a subject. The radiation source may include an accelerator 232, a target 234, or any other component (not shown). The primary collimator 250 may be configured to limit or collimate high energy beams (e.g., X-rays) emitted from the radiation source so that only those traveling parallel to a specified direction are allowed to pass through the primary collimator 250. The one or more filters 252 may be configured to adjust the distribution of the radiation impinging upon the subject. The one or more filters 252 may include a flattening filter, a bowtie filter, a wedge filter, or the like, or any combination thereof. The collimator 290 may be configured to shape a radiation field. The collimator 290 may include a Y-JAW 291, a X-JAW 292, an MLC 293 including a leaf assembly 2931 and a carriage 2934, or any other components. The leaf assembly 2933 may include multiple leaves. The MLC 293 may also include a driving assembly (not shown). Each of the multiple leaves may move in the carriage 2934 independently, for example, move toward the center of a radiation field or move away from the center of the radiation field to shape the radiation field. The MLC 293, the Y-JAW 291, and the X-JAW 292 may be form the radiation field, cooperatively. The MLC 293 may be mounted in the collimator 290. The collimator 290 may rotate. The MLC 293 may rotate along the rotation of the collimator 290. More descriptions for the MLC 293 may be found in elsewhere in the present disclosure (e.g., FIG. 2A and the descriptions thereof). FIG. 3 is a schematic diagram illustrating exemplary hardware and/or software components of an exemplary computing device 300 on which the processing device 120 may be implemented according to some embodiments of the present disclosure. As illustrated in FIG. 3, the computing device 300 may include a processor 310, storage 320, an input/output (I/O) 330, and a communication port 340. The processor 310 may execute computer instructions (e.g., program code) and perform functions of the processing device 120 in accordance with techniques described herein. The computer instructions may include, for example, routines, programs, objects, components, data structures, procedures, modules, and functions, which perform particular functions described herein. For example, the processor 310 may process data obtained from the radiotherapy device 110, the storage device 130, terminal(s) 140, and/or any other component of the radiotherapy system 100. In some embodiments, the processor 310 may include one or more hardware processors, such as a microcontroller, a microprocessor, a reduced instruction set computer (RISC), an application specific integrated circuits (ASICs), an application-specific instruction-set processor (ASIP), a central processing unit (CPU), a graphics processing unit (GPU), a physics processing unit (PPU), a microcontroller unit, a digital signal processor (DSP), a field programmable gate array (FPGA), an advanced RISC machine (ARM), a programmable logic device (PLD), any circuit or processor capable of executing one or more functions, or the like, or a combinations thereof. Merely for illustration, only one processor is described in the computing device 300. However, it should be noted that the computing device 300 in the present disclosure may also include multiple processors. Thus operations and/or method steps that are performed by one processor as described in the present disclosure may also be jointly or separately performed by the multiple processors. For example, if in the present disclosure the processor of the computing device 300 executes both operations A and B, it should be understood that operation A and operation B may also be performed by two or more different processors jointly or separately in the computing device 300 (e.g., a first processor executes operation A and a second processor executes operation B, or the first and second processors jointly execute operations A and B). The storage 320 may store data/information obtained from the radiotherapy device 110, the storage device 130, the terminal(s) 140, and/or any other component of the radiotherapy system 100. In some embodiments, the storage 320 may include a mass storage, removable storage, a volatile read-and-write memory, a read-only memory (ROM), or the like, or a combination thereof. For example, the mass storage may include a magnetic disk, an optical disk, a solid-state drive, etc. The removable storage may include a flash drive, a floppy disk, an optical disk, a memory card, a zip disk, a magnetic tape, etc. The volatile read-and-write memory may include a random access memory (RAM). The RAM may include a dynamic RAM (DRAM), a double date rate synchronous dynamic RAM (DDR SDRAM), a static RAM (SRAM), a thyristor RAM (T-RAM), and a zero-capacitor RAM (Z-RAM), etc. The ROM may include a mask ROM (MROM), a programmable ROM (PROM), an erasable programmable ROM (EPROM), an electrically erasable programmable ROM (EEPROM), a compact disk ROM (CD-ROM), and a digital versatile disk ROM, etc. In some embodiments, the storage 320 may store one or more programs and/or instructions to perform exemplary methods described in the present disclosure. The I/O 330 may input and/or output signals, data, information, etc. In some embodiments, the I/O 330 may enable user interaction with the processing device 120. In some embodiments, the I/O 330 may include an input device and an output device. Examples of the input device may include a keyboard, a mouse, a touch screen, a microphone, or the like, or a combination thereof. Examples of the output device may include a display device, a loudspeaker, a printer, a projector, or the like, or a combination thereof. Examples of the display device may include a liquid crystal display (LCD), a light-emitting diode (LED)-based display, a flat panel display, a curved screen, a television device, a cathode ray tube (CRT), a touch screen, or the like, or a combination thereof. The communication port 340 may be connected to a network (e.g., the network 150) to facilitate data communications. The communication port 340 may establish connections between the processing device 120 and the radiotherapy device 110, the storage device 130, and/or the terminal(s) 140. The connection may be a wired connection, a wireless connection, any other communication connection that can enable data transmission and/or reception, and/or a combination of these connections. The wired connection may include, for example, an electrical cable, an optical cable, a telephone wire, or the like, or a combination thereof. The wireless connection may include, for example, a Bluetooth™ link, a Wi-Fi™ link, a WiMax™ link, a WLAN link, a ZigBee link, a mobile network link (e.g., 3G, 4G, 5G, etc.), or the like, or a combination thereof. In some embodiments, the communication port 340 may be and/or include a standardized communication port, such as RS232, RS485, etc. In some embodiments, the communication port 340 may be a specially designed communication port. For example, the communication port 340 may be designed in accordance with the digital imaging and communications in medicine (DICOM) protocol. FIG. 4 is a schematic diagram illustrating exemplary hardware and/or software components of an exemplary mobile device 400 on which the terminal(s) 140 may be implemented according to some embodiments of the present disclosure. As illustrated in FIG. 4, the mobile device 400 may include a communication platform 410, a display 420, a graphics processing unit (GPU) 430, a central processing unit (CPU) 440, an I/O 450, a memory 460, and a storage 490. In some embodiments, any other suitable component, including but not limited to a system bus or a controller (not shown), may also be included in the mobile device 400. In some embodiments, a mobile operating system 470 (e.g., iOS™, Android™, Windows Phone™, etc.) and one or more applications 480 may be loaded into the memory 460 from the storage 490 in order to be executed by the CPU 440. The applications 480 may include a browser or any other suitable mobile apps for receiving and rendering information relating to image processing or other information from the processing device 120. User interactions with the information stream may be achieved via the I/O 450 and provided to the processing device 120 and/or other components of the radiotherapy system 100 via the network 150. To implement various modules, units, and their functionalities described in the present disclosure, computer hardware platforms may be used as the hardware platform(s) for one or more of the elements described herein. A computer with user interface elements may be used to implement a personal computer (PC) or any other type of work station or terminal device. A computer may also act as a server if appropriately programmed. FIG. 5 is a block diagram illustrating an exemplary processing device 120 according to some embodiments of the present disclosure. The processing device 120 may include an acquisition module 502, a control module 504, a processing module 506, and a storage module 508. At least a portion of the processing device 120 may be implemented on a computing device as illustrated in FIG. 3 or a mobile device as illustrated in FIG. 4. The acquisition module 502 may acquire data. In some embodiments, the data may be acquired from the radiotherapy device 110, the storage device 130, and/or the terminal(s) 140. In some embodiments, the data may include parameters (e.g., an angle) relating to an MLC, movement information (e.g., movement directions, movement phases, etc.) relating to a leaf in an MLC, an offset table associated with an MLC, instructions, or the like, or a combination thereof. The instructions may be executed by the processor(s) of the processing device 120 to perform exemplary methods described in the present disclosure. In some embodiments, the acquired data may be transmitted to the processing module 506 for further processing, or stored in the storage module 508. The control module 504 may control operations of the acquisition module 502, the processing module 506, and/or the storage module 508, for example, by generating one or more control parameters. For example, the control module 504 may control the acquisition module 502 to acquire data (e.g., an angle of an MLC, an angle of a gantry, etc.). As another example, the control module 504 may control the processing module 506 to generate an image relating to a subject. As a further example, the control module 504 may control the processing module 506 to implement a radiotherapy treatment plan for the subject. In some embodiments, the control module 504 may receive a real-time command or retrieve a preprogrammed command provided by a user (e.g., a doctor) to control one or more operations of the acquisition module 502 and/or the processing module 506. For example, the control module 504 may adjust the acquisition module 502 and/or the processing module 506 to determine the angle of a leaf according to the real-time command and/or the preprogrammed command. In some embodiments, the control module 504 may communicate with one or more other modules of the processing device 120 for exchanging information and/or data. The processing module 506 may process data provided by various modules of the processing device 120. For example, the processing module 506 may determine an offset value for each of leaves in a collimator (e.g., a multi-collimator in the radiotherapy device 110). As another example, the processing module 506 may generate a position instruction based on the offset value for each of the leaves in the collimator (e.g., a multi-leaf collimator in the radiotherapy device 110). The storage module 508 may store information. The information may include programs, software, algorithms, data, text, number, images, and some other information. For example, the information may include image data (e.g., a radiological image, an optical image, etc.), motion or position data (e.g., a speed, a displacement, an acceleration, a spatial position, etc.) relating to a component in the radiotherapy device 110 (e.g., the couch), instructions, or the like, or a combination thereof. In some embodiments, the storage module 508 may store program(s) and/or instruction(s) that can be executed by the processor(s) of the processing device 120 to acquire data, determine a spatial position of at least one part of a subject. In some embodiments, one or more modules illustrated in FIG. 5 may be implemented in at least part of the radiotherapy system 100 as illustrated in FIG. 1. For example, the acquisition module 502, the control module 504, the processing module 506, and/or the storage module 508 may be integrated into a console (not shown). Via the console, a user may set parameters for scanning a subject, controlling imaging or treatment processes, controlling parameters for the reconstruction of an image, etc. In some embodiments, the console may be implemented via the processing device 120 and/or the terminal(s) 140. FIG. 6 is a block diagram illustrating an exemplary processing module 506 according to some embodiments of the present disclosure. The processing module 506 may include a movement determination unit 602, an angle determination unit 604, an offset determination unit 606, a position adjustment unit 608, and a storage unit 610. At least a portion of the processing module 506 may be implemented on a computing device as illustrated in FIG. 3 or a mobile device as illustrated in FIG. 4. The movement determination unit 602 may be configured to determine a movement direction or a movement phase of the leaf. In some embodiments, the movement determination unit 602 may determine the movement direction or movement phase of the leaf based on the measurements of a main encoder (e.g., a motor encoder) connected with a driving component and/or an auxiliary encoder (e.g., a Hall sensor) associated with the leaf. For example, when an initial phase is an unknown phase, and the movement determination unit 602 may obtain a first measurement value and a second measurement value of the leaf through the Hall sensor in two adjacent sampling periods. If the difference between the first measurement value and the second measurement value is larger than a first preprogrammed count (e.g., 2 count, etc.), the leaf may be determined in a forward movement phase. If the difference between the first measurement value and the second measurement value is less than a second preprogrammed count (e.g., 0 count, etc.), then the leaf may be determined in a backward movement phase. In some embodiments, the movement determination unit 602 may be configured to determine a movement direction or a movement phase of the leaf based on a velocity of the leaf and/or a driving component associated with the leaf. For example, if the velocity of the driving component is less than a velocity threshold, the leaf may move away from the center of a radiation field. The angle determination unit 604 may be configured to determine an angle of the leaf. In some embodiments, the angle of the leaf may be determined by the angle determination unit 604 based on an angle of the gantry and an angle of a collimator. The collimator may be configured to support the MLC. More descriptions of the determining the angle of the leaf based on the angle of the gantry and the angle of the collimator may be found in operation 704 and the descriptions thereof. The offset determination unit 606 may be configured to determine the target offset of the leaf. In some embodiments, the offset determination unit 606 may determine the offset value of the leaf based on a preprogrammed offset table. In some embodiments, the offset value of the leaf may be determined based on measurement values acquired by a main encoder of a driving component associated with the leaf and an auxiliary encoder associated with the leaf. The position adjustment unit 608 may be configured to adjust the position of the leaf. In some embodiments, the position adjustment unit 608 may determine a target position of the leaf based on the offset value determined by the offset determination unit 606 associated with the position of the leaf. Further, the target position of the leaf may be determined by subtracting the offset value from a preprogrammed position of the leaf. The preprogrammed position of the leaf may be set by a user via the terminal device 140, or according to a default setting of the radiotherapy system 100, such as a treatment plan of a subject. In some embodiments, the position adjustment unit 608 may determine an actual position of the leaf based on the offset value determined by the offset determination unit 606 associated with the position of the leaf. Further, the actual position of the leaf based on the offset value determined by summing the current position of the leaf and the offset value. The current position of the leaf may be determined using the main encoder. The storage unit 610 may store information relating to, for example, determining the reference offset value, adjusting the position of the leaf, etc. The information may include programs, software, algorithms, data, text, number, and some other information. In some embodiments, the information relating to determining the reference offset may include data for determining the reference offset, algorithms for determining the reference offset value, parameters for determining the reference offset value, etc. The storage unit 610 may be a memory that stores data to be processed by processing devices, such as CPUs, GPUs, etc. In some embodiments, the storage unit 610 may be a memory that may be accessible by one or more GPUs or maybe a memory that is only accessible by a specific GPU. It should be noted that the above description of the processing module 506 is merely provided for the purposes of illustration, and not intended to limit the scope of the present disclosure. For persons having ordinary skills in the art, multiple variations or modifications may be made under the teachings of the present disclosure. However, those variations and modifications do not depart from the scope of the present disclosure. For example, the movement determination unit 602 and the angle determination unit 604 may be integrated into one single unit. FIG. 7 is a flowchart illustrating an exemplary process 700 for determining a target position of a leaf according to some embodiments of the present disclosure. In some embodiments, one or more operations of process 700 illustrated in FIG. 7 may be implemented in the radiotherapy system 100 illustrated in FIG. 1. For example, the process 700 illustrated in FIG. 7 may be stored in the storage device 130 in the form of instructions, and invoked and/or executed by the processing device 120 (e.g., the processor 310 of the computing device 300 as illustrated in FIG. 3, the GPU 430 or CPU 440 of the mobile device 400 as illustrated in FIG. 4). In 702, a movement direction of a leaf corresponding to a current position of the leaf in a multi-leaf collimator (MLC) may be determined. Operation 702 may be performed by the processing module 506. As used herein, the current position may be also referred to as a first position. In some embodiments, the movement direction of the leaf may include one of a backward movement direction and a forward movement direction. As used herein, the movement direction of a leaf may be considered as the backward movement direction if the leaf moves away from the center of a radiation field. The movement direction of a leaf may be considered to as the forward movement direction if the leaf moves toward the radiation field. In some embodiments, the movement direction of the leaf may correspond to a current or real-time movement direction when the leaf is at the current position. In some embodiments, the movement direction of the leaf may correspond to a preprogrammed movement direction of the leaf. The preprogrammed movement direction of the leaf may be determined based on an initial position of the leaf and a preprogrammed position of the leaf in a radiotherapy planning associated with the MLC. As used herein, the preprogrammed position of the leaf may refer to a default (or a commanded) position defined by a main encoder in the driving component. The preprogrammed position of the leaf may be set by a user via the terminal device 140, or according to a default setting of the radiotherapy system 100, such as a treatment plan associated with the MLC. In some embodiments, the movement direction of the leaf may be determined based on a rotation direction of a motor associated with the leaf in the driving component of the MLC. In some embodiments, the rotation direction of the motor may include one of a first rotation direction and a second rotation direction. When the motor rotates with the first rotation direction (e.g., a clockwise direction or anti-clockwise direction), the rotation of the motor may cause the leaf to move away from the center of the radiation field (i.e., with the backward movement direction). When the motor rotates with the second rotation direction (e.g., a clockwise direction or anti-clockwise direction), the rotation of the motor may cause the leaf to move toward the center of the radiation field (i.e., with the forward movement direction). In some embodiments, the rotation direction of the motor may be determined based on a rotation velocity of the motor when the leaf is at the current position. Further, if the rotation velocity of the motor is less than a first velocity threshold, the rotation direction of the motor may be determined as the first rotation direction. If the rotation velocity of the motor is greater than a second velocity threshold, the rotation direction of the motor may be determined as the second rotation direction. In some embodiments, the rotation velocity of the motor may be preprogrammed based on the preprogrammed position of the leaf. In some embodiments, the rotation velocity of the motor may be a current or real-time rotation velocity determined based on an encoder value acquired by a main encoder associated with the motor when the leaf is at the current position. For example, in adjacent sampling periods (also referred to as adjacent calculation cycles) of the main encoder, two measurements may be acquired by the main encoder corresponding two different positions of the leaf. The rotation velocity of the motor may be determined based on the two measurements and the sampling period. In some embodiments, the rotation direction of the motor may be determined based on the preprogrammed position of the leaf and the current position of the leaf. For example, if a distance between the preprogrammed position of the leaf and a reference point (e.g., the front end A of the MLC 200 as shown in FIG. 2A) is greater than a distance between the current position of the leaf and the reference point, the rotation direction of the motor may cause the leaf to move toward the preprogrammed position, i.e., move toward the center of the radiation field. If the distance between the preprogrammed position of the leaf and the reference point is less than the distance between the current position of the leaf and the reference point, the rotation direction of the motor may cause the leaf to move toward the preprogrammed position, i.e., move away from the center of the radiation field. The first velocity threshold may be a constant lower than or equal to zero, and the second velocity threshold may be a constant equal to or greater than zero. The first velocity threshold and/or the second velocity threshold may be determined by a user or according to a default setting of the radiotherapy system 100. In 704, an angle of the leaf corresponding to the current position of the leaf may be determined. Operation 704 may be performed by the angle determination unit 604. The angle of the leaf may be determined based on an angle of a collimator for supporting the MLC (e.g., the MLC 114) and an angle of a gantry of a radiotherapy device (e.g., the radiotherapy device 110) including the MLC. The MLC may be mounted on the collimator and rotate with the collimator. Further, the angle of the leaf may be determined according to Equation (1) as described below:sin(α)=sin(β)*cos(θ) (1)where, α represents an angle of a leaf; β represents an angle of a gantry of a radiotherapy device, and θ represents an angle of a collimator in the gantry of the radiotherapy device. As used herein, an angle of a leaf, an angle of a gantry of a radiotherapy device, and an angle of a collimator may be described in the coordinate systems of IEC (International Electrotechnical Commission) specifications. According to Equation (1), when the angle of the gantry is 0 degrees and the angle of the collimator is 90 degrees, the angle of the leaf may be determined to be 0 degrees. When the angle of the gantry is 90 degrees and the angle of the collimator is 0 degrees, the angle of the leaf may be determined to be 90 degrees. When the angle of the gantry is 45 degrees and the angle of the collimator is 45 degrees, the angle of the leaf may be determined to be 30 degrees. In some embodiments, the angle of the gantry and the angle of the MLC may be determined according to a default setting of the radiotherapy system 100, such as a treatment planning associated with the radiotherapy system 100. In some embodiments, the angle of the gantry and the angle of the collimator may be obtained from a control system associated with the collimator. In some embodiments, the angle of the gantry and the angle of the collimator may be determined using a measurement device, for example, an angle sensor associated with the gantry and/or the collimator. In 706, an offset value of the leaf may be determined based on the angle of the leaf and the movement direction of the leaf. Operation 706 may be performed by the offset determination unit 606. In some embodiments, a reference offset value of the leaf may be determined based on the angle of the leaf and the movement direction of the leaf. The offset value of the leaf may be determined based on a reference offset value. The reference offset value may obtain from an offset table (e.g., the Offset Table 1 as described in FIG. 8) associated with the MLC. In some embodiments, the offset table may include a plurality of reference offset values associated with each of a plurality of leaves under different angles and/or different movement directions. For example, each of the plurality of reference offset values of each of the plurality of leaves may be determined based on angles and movement directions of each of the plurality of leaves in the MLC using a distance measurement device and a main encoder of the driving component (e.g., a motor) associated with the each of the plurality of leaves. The distance measurement device may include a laser sensor, a dial indicator, a dial-gauge, etc. In some embodiments, the offset table may include a plurality of reference offset values associated with each of a plurality of leaves under different angles, different movement directions, and different positions. For example, each of the plurality of reference offset values may be determined based on angles, movement directions, and/or positions of each of a plurality of leaves in the MLC using a laser sensor and a main encoder of the driving component (e.g., a motor) associated with the each of the plurality of leaves. A reference offset value corresponding to the leaf may be determined from the offset table according to the movement direction of the leaf determined in 702, the angle of the leaf determined in 704, and/or the current position of the leaf described in 702. Then, the offset value of the leaf may be determined based on the reference offset value corresponding to the leaf. More descriptions for determining the offset value of the leaf based on the offset table may be found in FIG. 8, and the descriptions thereof. In some embodiments, the offset value of the leaf may be determined based on a movement phase of the leaf. In some embodiments, a current movement phase of the leaf may be determined based on measurement values of the main encoder (e.g., a motor encoder) of the driving component and/or the auxiliary encoder (e.g., a Hall sensor) associated with the leaf. The current movement phase of the leaf may be also referred to as a first movement phase. The offset value of the leaf may be determined based on the current movement phase of the leaf. For example, the current position of the leaf as described in 702 may be denoted by a first main encoder value acquired by the main encoder. A phase transition position of the leaf at where the leaf moves from a prior movement phase to the current movement phase may be denoted by a second main encoder value acquired by the main encoder. The offset value of the leaf may be determined based on a difference between the first main encoder value and the second main encoder value and a reference offset value associated with the phase transition position of the leaf. In some embodiments, the offset value of the leaf may be determined based on the current angle of the leaf and the current movement phase of the leaf. More descriptions for determining the offset value of the leaf based on the movement phase of the leaf may be found in FIGS. 9-11 and the descriptions thereof. In 708, a target position of the leaf may be determined based on the offset value. Operation 708 may be performed by the position adjustment unit 608. As used herein, the target position may be denoted by a main encoder value of the main encoder of the driving component associated with leaf. In some embodiments, the target position may refer to a position determined by calibrating, based on the offset value, the preprogrammed position of the leaf as described in 702. For example, the position adjustment unit 608 may adjust the preprogrammed position of the leaf based on the offset value associated with the current position of the leaf to obtain the target position of the leaf. Further, the target position of the leaf may be determined by subtracting the offset value from the preprogrammed position of the leaf. For example, the leaf is at the current position of 10 millimeters from the reference point (e.g., the front end A of the MLC 200 as shown in FIG. 2A) and in a forward movement direction, and the preprogrammed position of the leaf is 50 millimeter from the reference point. If the offset value of the leaf associated with the current position of 10 millimeters from the reference point determined in 706 is −2 millimeter, then the target position of the leaf may be a difference of the preprogrammed position and the offset value, which is equal to 52 millimeters. The motor may rotate in the second rotation direction to cause the leaf to move in the forward movement direction to the position of 50 millimeters from the reference point. As another example, the leaf is at the current position of 50 millimeters from the reference point and in a backward movement direction, and the preprogrammed position of the leaf is 20 millimeter from the reference point. If the offset value of the leaf associated with the current position of 50 millimeters from the reference point determined in 706 is 1.5 millimeter, then the target position of the leaf may be a difference between the preprogrammed position and the offset value, which is equal to 18.5 millimeters. The motor may rotate in the first rotation direction to cause the leaf to move in the backward movement direction to the position of 20 millimeters from the reference point. In some embodiments, an actual position (i.e., the target position) of the leaf corresponding to the current position of the leaf may be determined based on the offset value of the leaf. The target position corresponding to the current position may refer to a position determined by calibrating the current position based on the offset value. For example, the position adjustment unit 608 may adjust the current position of the leaf based on the offset value to obtain the target position (i.e., the actual position) of the leaf. Further, the target position (i.e., the actual position) of the leaf corresponding to the current position may be determined by summing the offset value and the current position of the leaf acquired by the main encoder. For example, the leaf is at the current position of 10 millimeters from the reference point (e.g., the front end A of the MLC 200 as shown in FIG. 2A) and in the forward movement direction. If the offset value of the leaf associated with the current position of 10 millimeters determined in 706 is −2 millimeter, then the target position (i.e., the actual position) of the leaf may be a sum of the current position and the offset value, which is equal to 8 millimeters. As another example, the leaf is at the current position of 50 millimeters from the reference point in the backward movement direction. If the offset value of the leaf associated with the current position of 50 millimeters determined in 706 is 1.5 millimeter, then the target position (i.e., the actual position) of the leaf may be a sum between the current position and the offset value, which is equal to 51.5 millimeters. It should be noted that the above description is merely provided for the purposes of illustration, and not intended to limit the scope of the present disclosure. For persons having ordinary skills in the art, multiple variations or modifications may be made under the teachings of the present disclosure. However, those variations and modifications do not depart from the scope of the present disclosure. For example, operation 706 may be divided into at least two operations. Operations 702 and 704 may be performed simultaneously. Further, a current position of a leaf may be determined. The current position of the leaf may be denoted by the movement direction of the leaf and the angle of the leaf. FIG. 8 is a flowchart illustrating an exemplary process 800 for determining an offset value of a leaf based on the angle of the leaf and a movement direction of the leaf according to some embodiments of the present disclosure. In some embodiments, one or more operations of process 800 illustrated in FIG. 8 may be implemented in the radiotherapy system 100 illustrated in FIG. 1. For example, process 800 illustrated in FIG. 8 may be stored in the storage device 130 in the form of instructions, and invoked and/or executed by the processing device 120 (e.g., the processor 310 of the computing device 300 as illustrated in FIG. 3, the GPU 430 or CPU 440 of the mobile device 400 as illustrated in FIG. 4). Operation 706 may be performed according to process 800. In 802, a first reference offset value of a leaf associated with a current position of the leaf may be obtained based on a movement direction and an angle of the leaf. Operation 802 may be performed by the acquisition module 502. The movement direction of the motor and the angle of the leaf may be determined as described in connection with operations 702 and 704 in FIG. 7. In some embodiments, the first reference offset value of the leaf may be determined from an offset table (e.g., Offset Table 1 as shown below) associated with an MLC including the leaf based on the movement direction of the leaf and the angle of the leaf. The offset table (e.g., Offset Table 1 as shown below) may include a plurality of reference offset values. Each of the plurality of reference offset values may correspond to one of a plurality of leaves in the MLC (e.g., the MLC 114 as shown in FIG. 1) under a specific angle and a specific movement direction. In some embodiments, the offset table of the MLC may be determined by and/or stored in the radiotherapy system 100. The acquisition module 502 may obtain the reference offset value of the leaf and/or the offset table of the MLC from the storage device 130, the storage 320, the storage 490, the storage module 508, the storage unit 610, and/or any other external storage. In some embodiments, one of the plurality of reference offset values in the offset table associated with one of the plurality of leaves in the MLC may be determined using a distance measurement device and a main encoder of a driving component (e.g., a motor) when the one of the plurality of leaves is moving in a specific movement direction (e.g., the forward movement direction or the backward movement direction) under a specific angle (e.g., 0 degrees, 45 degrees, 90 degrees, etc.). The main encoder of the driving component (e.g., a motor) associated with the one of the plurality of leaves may be configured to acquire the position of the one of the plurality of leaves that may be inaccurate caused by a position error (e.g., the backlash error) as described elsewhere in the present disclosure. The distance measurement device may be configured to acquire an exact position of one or more of the plurality of leaves. The position (i.e., inaccurate position) of a leaf acquired by the main encoder of the driving component of the leaf may be inaccurate with respect to the position (i.e., exact position) of the leaf acquired by the distance measurement device. Further, the one of the plurality of reference offset values of the one of the plurality of leaves may be determined based on a difference between the exact position and the inaccurate position of the one of the plurality of leaves. As shown in Offset Table 1, an MLC may include 120 leaves. Each of the plurality of leaves may have a specific leaf ID. Each of the plurality of leaves in the MLC may correspond to different reference offset values under different angles and different movement directions (e.g., the forward movement direction or backward movement direction shown in Offset Table 1). The angle of each of the plurality of leaves may be in a range from −90 degrees to 90 degrees. As used herein, the angle of a leaf is 0 degree if the leaf and a driving component (e.g., a motor) associated with the leaf are both parallel to the horizontal plane. The angle of a leaf is less than 0 degree if the leaf is located above a driving component (e.g., a motor) associated with the leaf. The angle of a leaf is greater than 0 degrees if the leaf is located below a driving component (e.g., a motor) associated with the leaf. In some embodiments, reference offset values of a specific leaf may be measured under a series of coherent angle values (e.g., every 10 degrees) between −90 degrees to 90 degrees. In some embodiments, reference offset values of a specific leaf may be measured under a series of distributed and random angle values (e.g., 10 degrees, 5 degrees, 2 degrees, etc.). As shown in Offset Table 1, the offset value is measured every 10 degrees of the leaf angle. And with a specific degree of the leaf, reference offset values of the specific leaf is measured in two different movement directions, the forward movement direction (i.e. Forward in Offset Table 1) and the backward movement direction (i.e. Backward in Offset Table 1). OFFSET TABLE 1Leaf Leaf IDangleLeaf ID1. . .60. . .12090ForwardBackward80ForwardBackward. . .0ForwardBackward. . .−80ForwardBackward−90ForwardBackward In some embodiments, a relationship between the reference offset value and the current position of a specific leaf in the MLC may be further determined when the leaf with a specific angle is moving at a specific direction. For example, when the leaf with the specific angle is moving at the specific movement direction, multiple positions of the specific leaf may be acquired using the distance measurement device and the main encoder of a driving component (e.g., a motor) respectively. Each of the multiple positions may be denoted by a main encoder value acquired by the main encoder and a measurement value acquired by the distance measurement device. Each of multiple reference offset values associated with the each of the multiple positions of the specific leaf may be determined based on a difference between the main encoder value acquired by the main encoder and the measurement value acquired by the distance measurement device. The relationship between the reference offset value and the position of the specific leaf when the leaf with the specific angle is moving at the specific direction may be determined based on the multiple reference offset values and the multiple positions using, for example, a polynomial fitting technique (e.g., a binomial fitting algorithm). Then, the reference offset value of the leaf associated with the current position may be determined based on the current position and the relationship between the reference offset value and the position of the leaf corresponding to the movement direction and the angle. In some embodiments, the offset table may include multiple relationships between the reference offset value and the position of each of the plurality of leaves under different angles and movement directions. In some embodiments, the offset table may include multiple reference offset values of each of the plurality of leaves under different angles, different movement directions, and different positions. A current reference offset value of a leaf corresponding to the current position may be obtained from the offset table based on the current position of the leaf, the angle of the leaf and the movement direction of the leaf. In 804, a first main encoder value acquired by a main encoder of a driving component associated with leaf may be obtained, which may be associated with the current position of the leaf. The operation 804 may be performed by the acquisition module 502. The first main encoder value may represent the current position of the leaf. In some embodiments, the first main encoder value may be obtained from the MLC 114, the storage device 130, the storage module 508, the storage unit 610, or any other external storage. In 806, a second main encoder value acquired by the main encoder of the driving component associated with the leaf may be obtained, which may be associated with a prior position of the leaf when a movement direction of the leaf (or a rotation direction of the motor) changes. Operation 806 may be performed by the acquisition module 502. The second main encoder value may represent the prior position of the leaf. As used herein, the prior position of the leaf may be also referred to as a direction transition position of the leaf. In some embodiments, when the leaf is at the current position, the movement direction of the leaf may be a first direction including one of the forward movement direction or the backward movement direction as described elsewhere in the present disclosure (e.g., FIG. 6 and the descriptions thereof)). When the leaf was at the prior position, the movement direction of the leaf may change from a second direction (e.g., the forward movement direction or the backward movement direction as described elsewhere in the present disclosure (e.g., FIG. 6 and the descriptions thereof)) to the first direction. For example, as shown in FIG. 11, if the leaf is moving in the forward movement direction, the current position of the leaf may correspond to a point on section (a) or section (d) of the movement curve in FIG. 11. The prior position of the leaf may correspond to the transition point Tc. If the leaf is moving in the backward movement direction, the current position of the leaf may correspond to a point on section (b) or section (c) of the movement curve in FIG. 11. The prior position of the leaf may correspond to the transition point Ta. In some embodiments, the second main encoder value may be obtained from a component (e.g., the position feedback module 218) of the control system, the storage device 130, the storage module 508, the storage unit 610, or any other external storage. In 808, a determination may be made as to whether the movement direction of the leaf moves in the backward movement direction. Operation 808 may be performed by the offset determination unit 606. If it is determined that the movement direction of the leaf corresponding to the current position is the backward movement direction as described elsewhere in the present disclosure, process 800 may proceed to operation 810. If it is determined that the movement direction of the leaf corresponding to the current position is the forward movement direction as described elsewhere in the present disclosure, process 800 may proceed to operation 812. In 810, a minimum value among the first reference offset value and the sum of a second reference offset value and the difference between the first main encoder value and the second main encoder value may be designated as an offset value of the leaf associated with the current position. Operation 810 may be performed by the offset determination unit 606. The second reference offset value may correspond to the prior position. In some embodiments, when the movement direction of the leaf is the backward movement direction at the current position, the offset value of the leaf may be determined according to Equation (2) below:Offset Px=min((Encoder Ta−Encoder Px+Offset Ta),Reference offset Px), (2),where, Encoder Ta represents a main encoder value (e.g., the second main encoder value) associated with a prior position of a leaf acquired by a main encoder of a driving component when the movement direction of the leaf changes from the forward movement direction to the backward movement direction, Encoder Px represents a main encoder value (e.g., the first main encoder value) associated with a current position of the leaf acquired by the main encoder of the driving component, Offset Ta denotes the second reference offset value associated with the prior position (e.g., transition point Ta as shown in FIG. 11) of the leaf, and Reference offset Px represents a first reference offset value obtained from an offset table when the leaf is moving with the backward movement direction. In 812, a maximum value among the first reference offset value and the sum of the second reference offset value and the difference between the first main encoder value and the second main encoder value may be designated as an offset value of the leaf associated with the current position. Operation 812 may be performed by the offset determination unit 606. In some embodiments, when the movement direction of the leaf is the forward movement direction at the current position, the offset value of the leaf may be determined according to Equation (3) below:Offset Px=max((Encoder Tc−Encoder Px+Offset Tc),Reference offset Px), (3),where Encoder T2 represents a main encoder value (e.g., the second main encoder value) acquired by a main encoder of a driving component associated with a prior position of a leaf when the movement direction of the leaf changes from the backward movement direction to the forward movement direction, Encoder Px represents a main encoder value (e.g., the first main encoder value) associated with a current position of the leaf acquired by the main encoder of the driving component, Offset Tc denotes the second reference offset value associated with the prior position (e.g., transition point Tc as shown in FIG. 11)) of the leaf, and Reference offset Px represents a first reference offset value obtained from an offset table when the leaf is moving with the forward movement direction. It should be noted that the above description is merely provided for the purposes of illustration, and not intended to limit the scope of the present disclosure. For persons having ordinary skills in the art, multiple variations and modifications may be made under the teachings of the present disclosure. However, those variations and modifications do not depart from the scope of the present disclosure. For example, operations 804 and 806 may be performed simultaneously or in a reverse order than that illustrated in FIG. 8. As another example, process 800 may further include process a first signal associated with the current position of the leaf acquired by the main encoder of the driving component to obtain the first main encoder value. As still another example, process 800 may further include process a second signal associated with the prior position of the leaf acquired by the main encoder of the driving component to obtain the second main encoder value. FIG. 9 is a flowchart illustrating another exemplary process 900 for determining an offset value of a leaf according to some embodiments of the present disclosure. In some embodiments, one or more operations of process 900 illustrated in FIG. 9 may be implemented in the radiotherapy system 100 illustrated in FIG. 1. For example, the process 900 illustrated in FIG. 9 may be stored in the storage device 130 in the form of instructions, and invoked and/or executed by the processing device 120 (e.g., the processor 310 of the computing device 300 as illustrated in FIG. 3, the GPU 430 or CPU 440 of the mobile device 400 as illustrated in FIG. 4). Operation 706 may be performed by process 900 as illustrated in FIG. 9. In 902, a current movement phase of a leaf corresponding to a current position of the leaf may be determined. Operation 902 may be performed by the movement determination unit 602. In some embodiments, the current movement phase of the leaf may be also referred to as a first movement phase when the leaf is at the current position (i.e., a first position). The current movement phase may be one of the four phases including a first phase, a second phase, a third phase, and a fourth phase. The leaf may be determined in the first phase (also referred to as a forward-movement phase) when the leaf is moving toward the center of a radiation field. The leaf may be determined in the second phase (also referred to as a backward consume phase) when the leaf is static relative to a carriage of an MLC including the leaf and configured to move away from the center of the radiation field. The leaf may be determined in the third phase (also referred to as a backward movement phase) when the leaf is moving away from the radiation field. The leaf may be determined in the fourth phase (also referred to as a forward consume phase) when the leaf is static relative to the carriage of the MLC including the leaf and configured to move toward the center of the radiation field. More descriptions of the four phases may be found elsewhere in the present disclosure (e.g., FIG. 16A, and the descriptions thereof). In some embodiments, the movement determination unit 602 may determine the current movement phase of the leaf based on measurements of a main encoder (e.g., a motor encoder) of a driving component and/or an auxiliary encoder (e.g., a Hall sensor) associated with the leaf in two adjacent sampling periods (also referred to as adjacent calculation cycles). As used herein, a measurement of each of the main encoder and/or the auxiliary encoder may include a count of signals acquired by each of the main encoder and/or the second encoder which may be used to determine a position of the leaf. Further, the current movement phase of the leaf may be determined based on a count difference of at least one of the main encoder and/or the second encoder in two adjacent sampling periods (also referred to as adjacent calculation cycles). For example, when the leaf is moving in the forward movement phase, if the count difference between two measurements of the main encoder in adjacent sampling periods is less than a first threshold (e.g., −5, or a constant less than −5), the movement determination unit 602 may determine that the movement phase of the leaf changes and the leaf is moving into the backward consume phase. If the count difference between two measurements of the main encoder in adjacent sampling periods is greater than a second threshold (e.g., −5, or a constant less than −5), the movement determination unit 602 may determine that the movement phase of the leaf changes and the leaf is moving into the forward consume phase. As another example, when the movement phase of the leaf is unknown, if the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is larger than a third threshold (e.g., 2, or a constant greater than 2), the movement determination unit 602 may determine the current movement phase of the leaf as moving in the forward movement phase. If the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is less than a fourth threshold (e.g., −2, or a constant less than −2), the movement determination unit 602 may determine the current movement phase of the leaf as the backward movement phase. As still another example, when the leaf is moving in the backward consume phase, if the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is larger than the fifth threshold (e.g. 0), the movement determination unit 602 may determine that the current movement phase of the leaf changes and the leaf is moving into the forward movement phase. If the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is less than the fifth threshold (e.g. 0), the movement determination unit 602 may determine that the current movement phase of the leaf changes and the leaf is moving into the backward movement phase. When the leaf is moving in the forward consume phase, if the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is larger than the fifth threshold (e.g. 0), the movement determination unit 602 may determine that the current movement phase of the leaf changes and the leaf is moving into the forward movement phase. If the count difference between the two measurements of the auxiliary encoder in adjacent sampling periods is less than the fifth threshold (e.g. 0), the movement determination unit 602 may determine that the current movement phase of the leaf changes and the leaf is moving into the backward movement phase. In 904, a reference offset value may be determined, which is associated with a prior position of the leaf at where a movement phase of the leaf changes from a prior movement phase to the current movement phase. Operation 904 may be performed by the offset determination unit 606. As used herein, the prior position may be also referred to as a phase transition position or transition point (e.g., transition points T1, T2, T3, and T4 as shown in FIG. 16A) of the leaf. The prior movement phase may be also referred to as a second movement phase. The prior position may correspond to a transition point closest to the current position of the leaf. For example, if the current movement phase of the leaf determined in 902 is the backward consume phase, the prior movement phase of the leaf may be the forward movement phase. Then the prior position may correspond to a second transition point (e.g., transition point T2 as shown in FIG. 16A) from the forward movement phase to the backward consume phase. If the current movement phase of the leaf determined in 902 is the backward movement phase, the prior the movement phase of the leaf may be the backward consume phase. Then the prior position may correspond to a third transition point (e.g., transition point T3 as shown in FIG. 16A) from the backward consume phase to the backward movement phase. If the current movement phase of the leaf determined in 902 is the forward movement phase, the prior movement phase of the leaf may be the forward consume phase. Then the prior position may correspond to a first transition point (e.g., transition point T1 as shown in FIG. 16A) from the forward consume phase to the forward movement phase. If the current movement phase of the leaf determined in 902 is the forward consume phase, and the prior movement phase of the leaf is the backward movement phase. Then the prior position may correspond to a fourth transition point (e.g., transition point T4 as shown in FIG. 16A) from the backward movement phase to the forward consume phase. The reference offset value of the leaf associated with the prior position of the leaf may be an offset value when the leaf is moving at the prior position. In some embodiments, if the leaf is in the forward consume phase, the reference offset value of the leaf may be equal to an offset value corresponding to the fourth transition point (e.g., transition point T4 as shown in FIG. 16A) from the backward movement phase to the forward consume phase. If the leaf is in the backward consume phase, the reference offset value of the leaf may be equal to an offset value corresponding to the second transition point (e.g., transition point T2 as shown in FIG. 16A) from the forward movement phase to the backward consume phase. In 906, a first main encoder value acquired by a main encoder of a driving component associated with the leaf may be obtained, which is associated with the current position of the leaf. Operation 906 may be performed by the offset determination unit 606. In some embodiments, the first main encoder value may represent the current position of the leaf. In some embodiments, the first main encoder value may be obtained from the main encoder (e.g., the encoder 208) directly. In some embodiments, the first main encoder value may be obtained from the storage device 130, the storage module 508, the storage unit 610, or any other external storage. In 908, a second main encoder value acquired by the main encoder may be obtained, which is associated with the prior position of the leaf. Operation 908 may be performed by the offset determination unit 606. In some embodiments, the second main encoder value acquired by the main encoder may represent the prior position of the leaf. In some embodiments, the second main encoder value may be obtained from the storage device 130, the storage module 508, the storage unit 610, or any other external storage. In 910, the offset value may be determined based on the reference offset value and a difference between the first main encoder value and the second main encoder value. Operation 910 may be performed by the offset determination unit 606. In some embodiments, the offset value of the leaf associated with the current position may be determined based on the reference offset value and the difference between the first main encoder value and the second main encoder value. Further, the offset value of the leaf associated with the current position of the leaf may be equal to a sum between the reference offset value and the difference between the second main encoder value and the first main encoder value as described by Equation (4) below:Offset Px=Encoder Tx−Encoder Px+Reference Offset Tx, (4),where, Offset Tx represents an offset value corresponding to a current position of a leaf, Encoder Tx represents a main encoder value corresponding to a prior position when a movement phase of the leaf changes from a prior movement phase to a current movement phase, also referred to as the second main encoder value For example, Encoder Tx may correspond to the transition position T2′ or T2 as shown in FIG. 19B. Encoder Px represents a main encoder value corresponding to the current position, also referred to as the first main encoder value, and Reference Offset Tx represents a reference offset value corresponding to the prior position. If the current movement phase of the leaf is the backward consume phase, and the prior movement phase is the forward movement phase, the offset value (i.e., Offset Px) of the leaf associated with a current position in the backward consume phase may be determined as described by Equation (5) below:Offset Px=Encoder T2−Encoder Px+Reference Offset T2 (5),where Offset Px represents an offset value of the leaf associated with a current position in the backward consume phase, Encoder T2 represents a main encoder value acquired by the main encoder associated with a position when the movement phase of the leaf changes from the forward movement phase to the backward consume phase, i.e., a second main encoder value, and Encoder Px represents a main encoder value acquired by the main encoder associated with the current position in the backward consume phase, i.e., a first main encoder value. Reference Offset T2 represents a reference offset value of the second transition point T2 that may be an offset value of the position when the movement phase of the leaf changes from the forward movement phase to the backward consume phase. According to Equation (5), if the current position corresponds to the third transition point when the movement phase of the leaf changes from the backward consume phase to the backward movement phase, the offset value of the current position may be denoted by Equation (6):Offset Px=Offset T3=Encoder T2−Encoder T3+Offset T2 (6). If the current movement phase of the leaf is the forward consume phase, and the prior movement phase is the backward movement phase, an offset value (i.e., Offset Px) of the leaf associated with the current position in the forward consume phase may be determined as described by Equation (7) below:Offset Px=Reference Offset T4−(Encoder Px−Encoder T4) (7).where Offset Px represents an offset value of the leaf associated with the current position in the forward consume phase, Encoder T4 represents a main encoder value (i.e., the second main encoder value) acquired by the main encoder associated with a position when the movement phase of the leaf changes from the backward movement phase to the forward consume phase, Encoder Px represents a main encoder value (i.e., the first main encoder value) acquired by the main encoder associated with the current position in the forward consume phase, and Reference Offset T4 represents a reference offset value of the fourth transition point T4 that may be an offset value of the position when the movement phase of the leaf changes from the backward movement phase to the forward consume phase. According to Equation (7), if the current position includes the position when the movement phase of the leaf changes from the forward consume phase to the forward movement phase, i.e., the first transition point T1, the offset value of the current position may be denoted by Equation (8):Offset Px=Offset T1=(EncoderT4−EncoderT1+OffsetT4) (8). In operation 912, the offset value of the current position of the leaf may be determined based on the reference offset value of the prior position. Operation 912 may be performed by the offset determination unit 606. When the leaf is at the prior position, a movement phase of the leaf changes from a prior movement phase to the current movement phase. The prior position may correspond to a transition point closest to the current position of the leaf. More descriptions for the prior position of the leaf may be found in operation 904. In some embodiments, if the leaf is in the forward movement phase or the backward movement phase, a backlash between the leaf and the driving component may be unchanged, such that the offset value for removing the backlash may be a constant. In other words, the offset value of the leaf may be unchanged when the leaf is moving in the forward movement phase or the backward movement phase. In some embodiments, if the leaf is in the forward movement phase, the offset value of the current position may be equal to an offset value of a phase transition position (e.g., first transition point T1 as shown in FIG. 16A) from the forward consume phase to the forward movement phase, denoted by Offset T1, and may be equal to an offset value of a phase transition position (e.g., second transition point T2 as shown in FIG. 16A) from the forward movement phase to the backward consume direction, denoted by Offset T2. In other words, the offset value of the current position in the forward movement phase may be denoted by Offset Px=Offset T1=Offset T2. In some embodiments, if the leaf is in the backward movement phase, the offset value of the current position may be equal to an offset value of a phase transition position (e.g., third transition point T3 as shown in FIG. 16A) from the backward consume phase to the backward movement phase, denoted by Offset T3, and may be also equal to an offset value of a phase transition position (e.g., fourth transition point T4 as shown in FIG. 16A) from the backward movement phase to the forward consume phase, denoted by Offset T4. In other words, the offset value of the current position in the backward movement phase may be denoted by Offset Px=Offset T3=Offset T4. In some embodiments, if the angle of the leaf is 0 degree and the leaf is in the forward movement phase, the offset value of the leaf may be equal to zero, which may be denoted by Offset Px=Offset T1=OffsetT2=0. In some embodiments, the angle of the leaf may change along with the leaf moves. The change of the angle of the leaf may cause a change of backlash error, and the offset value of the leaf needs to be modified. If the angle of the leaf is about 0 degree, the angle change of the leaf does not exceed a threshold, and the leaf is in the forward movement phase, the offset value of the current position in the forward movement phase may be equal to 0 that may be denoted by Offset Px=Offset T1=Offset T2=0. If the leaf is in the backward consume phase, the offset value of the current position in the backward consume phase may be denoted by Offset Px=Encoder T2−Encoder Px. If the leaf is in the backward movement phase, the offset value of the current position in the backward movement phase may denoted by Offset Px=Offset T3=Offset T4=Encoder T2−Encoder T3. If the leaf is in the forward consume phase, the offset value of the current position in the forward consume phase may denoted by Offset Px=(Encoder T2−Encoder T3)−(Encoder Px−Encoder T4). If the angle change of the leaf exceeds a threshold, the offset value of the current position of the leaf may be determined based on the reference offset value of the prior position. In some embodiments, if the leaf is in the forward movement phase or the backward movement phase, whether the offset value of the current position needs to be modified may be determined by determining whether the angle change of the leaf exceeds the threshold. In some embodiments, if the leaf is in the forward movement phase or the backward movement phase, whether the offset value of the current position needs to be modified may be determined based on measurements of a main encoder and an auxiliary encoder. More descriptions for determining whether the offset value of the current position needs to be modified may be found in FIG. 10 and the descriptions thereof. It should be noted that the above description is merely provided for the purposes of illustration, and not intended to limit the scope of the present disclosure. For persons having ordinary skills in the art, multiple variations and modifications may be made under the teachings of the present disclosure. However, those variations and modifications do not depart from the scope of the present disclosure. For example, process 900 may include the operation of obtaining the values acquired by the Hall sensor. In some embodiments, operations 906 and 908 may be omitted. The offset value of the leaf associated with the current position may be equal to the reference offset value. FIG. 10 is a flowchart illustrating an exemplary process 1000 for determining a reference offset value of a leaf according to some embodiments of the present disclosure. In some embodiments, one or more operations of process 1000 illustrated in FIG. 10 may be implemented in the radiotherapy system 100 illustrated in FIG. 1. For example, the process 1000 illustrated in FIG. 10 may be stored in the storage device 130 in the form of instructions, and invoked and/or executed by the processing device 120 (e.g., the processor 310 of the computing device 300 as illustrated in FIG. 3, the GPU 430 or CPU 440 of the mobile device 400 as illustrated in FIG. 4). Operation 904 may be performed according to process 1000 as illustrated in FIG. 10. According to process 1000, the offset value of a leaf may be influenced by the angle change of the leaf moving in the forward movement phase or backward movement phase. However, in the forward consume phase or backward consume phase, the effect of the angle change of the leaf on the offset value of a leaf may be neglected. In 1002, a determination may be made to as whether an angle change of a leaf exceeds a threshold in a current movement phase. Operation 1002 may be performed by the offset determination unit 606. If it is determined that the angle change of the leaf exceeds the threshold in the current movement phase, process 1000 may proceed to operation 1004. If it is determined that the angle change of the leaf does not exceed the threshold in the current movement phase, process 1000 may proceed to operation 1010. For example, as shown in FIG. 19B, if the angle change of the leaf does not exceed a threshold in the forward movement phase, the offset value of the leaf at the current position may be equal to the offset value of the transition point T1 i.e., Offset T1. If the angle change of the leaf exceeds the threshold in the forward movement phase, a moving curve of the leaf in the forward movement phase may change from section (a) to section (a)′, and the offset value of the leaf at the current position needs to be modified. As another example, as shown in FIG. 19B, if the angle change of the leaf does not exceed a threshold in the backward movement phase, the offset value of the leaf at the current position may be equal to the offset value of the transition point T3 i.e., Offset T3. If the angle change of the leaf exceeds the threshold in the backward movement phase, a moving curve of the leaf in the backward movement phase may change from section (c) to section (c)′, and the offset value of the leaf at the current position needs to be modified. In some embodiments, whether the angle change of the leaf exceeds the threshold may be determined by determining whether an angle change value of the leaf between a reference position and the current position exceeds a threshold. Further, if the angle change value of the leaf exceeds the threshold, the offset value of the leaf at the current position may be modified. For example, the angle change of the leaf may be determined based on an angle change value of the leaf between the prior position and the current position. In some embodiments, whether the angle change of the leaf exceeds the threshold may be determined by determining whether an angle change value of the leaf at the current position with respect to a prior sampling period exceeds the threshold. In some embodiments, the angle of the leaf may be determined according to Equation (1) as described in FIG. 7. When the leaf is at the reference position Tr, the angle of the leaf may be αTr. When the leaf is at the current position Px, the angle of the leaf may be αPx. The angle change value of the leaf may be denoted by Δα=αPx−αTr. Whether the angle change value Δα exceeds the threshold may be determined and the offset value of the leaf may be determined according to the determination that the angle change value Δα exceeds the threshold. In some embodiments, whether the angle change of the leaf exceeds the threshold may be determined based on measurements of a main encoder of a driving component and an auxiliary encoder associated with the current position of the leaf and measurements of a main encoder of a driving component and an auxiliary encoder associated with the reference position of the leaf. For example, if the main encoder includes a motor encoder and the auxiliary encoder includes a Hall sensor, whether the angle change of the leaf exceeds the threshold may be determined based on a change of a measurement Encoder Px of the motor encoder (or the main encoder) associated with the current position Px with respect to a measurement of the motor encoder (or the main encoder) associated with the reference position Tr and a change of a measurement Hall Px of the Hall sensor (or the auxiliary encoder) associated with the current position Px with respect to a measurement of the Hall sensor (or the auxiliary encoder) associated with the reference position Tr. The reference position may include a phase transition position, a position in a prior sampling period, a position between the phase transition position and the position in a prior sampling period, etc. For example, a first main encoder value and a first auxiliary encoder value of the current position of the leaf may be obtained. The first main encoder value may be acquired by the main encoder when the leaf is at the current position and the first auxiliary encoder value may be acquired by the auxiliary encoder when the leaf is at the current position. A second main encoder value and a second auxiliary encoder value of the phase transition position of the leaf may be obtained. The second main encoder value may be acquired by the main encoder when the leaf is at the phase transition position and the second auxiliary encoder value may be acquired by the auxiliary encoder when the leaf is at the phase transition position. A first difference between the first main encoder value and the second main encoder value. A second difference between the first auxiliary encoder value and the second auxiliary encoder value. Whether the offset value of the leaf needs to be modified may be determined based on the first difference and the second difference. For example, whether the offset value of the leaf needs to be modified may be determined based on a difference between the first difference and the second difference or a ratio of the first difference and the second difference. If the difference between the first difference and the second difference or the ratio of the first difference and the second difference exceeds a threshold, the angle change of the leaf exceeds the corresponding threshold and the offset value of the leaf needs to be modified may be determined. If the difference between the first difference and the second difference or the ratio of the first difference and the second difference does not exceed the threshold, the angle change of the leaf does not exceed the corresponding threshold and the offset value of the leaf does not need to be modified may be determined In 1004, an offset value associated with a reference position in the current movement phase may be determined. Operation 1004 may be performed by the offset determination unit 606. The reference position may be a specific position in the current movement phase. For example, the reference position may be a phase transition position from a prior movement phase to the current movement phase. As another example, the reference position may be a position in a prior sampling period. As still another example, the reference position may be a position between the phase transition position and the position in the prior sampling period. If the leaf is in the forward movement phase, the offset value of the reference position may be equal to Offset T1. If the leaf is in the backward movement phase, the offset value of the reference position may be equal to Offset T3. In 1006, measurement values of the main encoder and the auxiliary encoder may be obtained when the leaf is at the current position and the reference position respectively. Operation 1006 may be performed by the offset determination unit 606. In some embodiments, the main encoder may be a motor encoder. The auxiliary encoder may be a Hall sensor. A first motor encoder value Encoder Px and a first Hall value Hall Px corresponding to the current position Px may be obtained. The first motor encoder value Encoder Px may be acquired by the motor encoder and the first Hall value Hall Px may be acquired by the Hall sensor. A second motor encoder value Encoder Tr and a second Hall value Hall Tr corresponding to the reference position Tr may be obtained. The second motor encoder value Encoder Tr may be acquired by the motor encoder and the second Hall value Hall Tr may be acquired by the Hall sensor. In some embodiments, the first motor encoder value Encoder Px, the first Hall value Hall Px, the second motor encoder value Encoder Tr, and the second Hall value Hall Tr may be obtained from the storage device 130, the storage module 508, the storage unit 610 or any other storage device. In 1008, the offset value associated with the current position of the leaf may be determined based on the offset value of the reference position and the measurement values of the main encoder and the auxiliary encoder corresponding to the current position and the reference position respectively. Operation 1008 may be performed by the offset determination unit 606. As used herein, the second first ideal reference offset value associated with the prior position may also refer to as a reference offset value associated with the prior position not including the error caused by the Hall sensor. For example, the main encoder may be a motor encoder. The auxiliary encoder may be a Hall sensor. When the leaf is at the reference position, the offset value of the leaf at the reference position may be determined based on measurement values of the motor encoder and the Hall sensor according to Equation (9):Offset Tr=(Hall Tr−δ)−Encoder Tr (9),where Offset Tr refers to an offset value of the leaf at the reference position Tr, Hall Tr refers to a measurement value of the Hall sensor when the leaf is at the reference position, i.e., the second Hall value, Encoder Tr refers to a measurement value of the motor encoder when the leaf is at the reference position, i.e., the second motor encoder value, and δ refers to a feedback error of the Hall sensor caused by poor linearity and repeatability of the Hall sensor. When the leaf is at the current position, the offset value of the leaf at the current position may be determined based on measurement values of the motor encoder and the Hall sensor according to Equation (10):Offset Px=(Hall Px−δ)−Encoder Px (10),where Offset Px refers to an offset value of the leaf at the current position Px, Hall Px refers to a measurement value of the Hall sensor when the leaf is at the current position, i.e., the first Hall value, Encoder Px refers to a measurement value of the motor encoder when the leaf is at the current position, i.e., the first motor encoder value, and δ refers to the feedback error of the Hall sensor caused by poor linearity and repeatability of the Hall sensor. According to Equations (9) and (10), the offset value of the leaf at the current position may be determined according to Equation (11):Offset Px=Offset Tr+(Hall Px−Hall Tr)−(Encoder Px−Encoder Tr) (11). If the leaf is in the forward movement phase, the offset value of the reference position Offset Tr may be equal to Offset T1. If the leaf is in the backward movement phase, the offset value of the reference position Offset Tr may be equal to Offset T3. According to Equation (11), the feedback error of the Hall sensor may be removed. In 1010, an offset value of the leaf at a prior position may be determined as the offset value of the leaf at the current position. Operation 1010 may be performed by the offset determination unit 606. More descriptions for determining the offset value of the leaf at the prior position may be found in FIG. 9 and the descriptions thereof. FIG. 11 is a schematic diagram illustrating an exemplary movement curve of a leaf according to some embodiments of the present disclosure. The movement curve of the leaf may be obtained when the leaf has an angle of 0 degrees as shown in FIG. 14C, also referred to that the leaf is parallel to the horizontal plane. As illustrated in FIG. 11, the horizontal axis (i.e., X-axis) represents a position of the leaf acquired by a main encoder associated with the leaf. The vertical axis (i.e., Y-axis) represents the accurate position of the leaf acquired by a distance measurement device, for example, a laser sensor. The solid line denotes that the leaf is moving along the backward movement direction, i.e., moving away from the center of a radiation field. The dotted line denotes that the leaf is moving along the forward movement direction, i.e., moving toward the center of the radiation field. Thus, two transition points associated with the movement direction of the leaf are shown in FIG. 11. Ta represents the transition point from the forward movement direction to the backward movement direction. Tc represents the transition point from the backward movement direction to the forward movement direction. The movement curve of the leaf includes four sections corresponding to four movement phases of the leaf respectively. Section (a) corresponds to a first phase (also referred to as the forward movement phase) of the leaf that the leaf is moving toward the center of the radiation field. Section (b) corresponds to a second phase (also referred to as backward consume phase) that the leaf is static relative to a carriage of an MLC and configured to move away from the center of the radiation field. Section (c) corresponds to a third phase (also referred to as the backward movement phase) that the leaf is moving away from the center of a radiation field. And section (d) corresponds to a fourth phase (also referred to as forward consume phase) that the leaf is static relative to the carriage of the MLC and configured to move toward the center of the radiation field. FIG. 12 is a schematic diagram illustrating an exemplary movement curve of a leaf according to some embodiments of the present disclosure. The movement curve of the leaf is obtained when the leaf is located upward a motor associated with the leaf (e.g., the leaf 1442 with an angle of 90 degrees as shown in FIG. 14B). As shown in FIG. 12, section (a) of the movement curve in FIG. 11 shifts to section (a′) causing the shortening of section (b) and section (d), as well as the shifting of the transition point from Ta to Ta′. FIG. 13 is a schematic diagram illustrating another exemplary movement curve of a leaf according to some embodiments of the present disclosure. The movement curve of the leaf is obtained when the leaf is located downward a motor associated with the leaf (e.g., the leaf 1444 with an angle of −90 degrees as shown in FIG. 14B). As shown in FIG. 13, section (c) of the movement curve in FIG. 11 shifts to section (c′) causing the shortening of section (b) and section (d), as well as the shifting of the transition point from Tc to Tc′. FIG. 14A is a schematic diagram illustrating an exemplary gantry of the radiotherapy device 110 in a sectional view according to some embodiments of the present disclosure. In FIG. 14A, a gantry 1420 may include a collimator 1440. An MLC may be mounted on the collimator 1440 and rotate along the collimator 1440. The angle of the gantry (gantry angle) is 90 degrees relative to the horizontal plane as described by a plane formed by X-axis and Y-axis. Z-axis denotes a vertical direction. The direction denoted by arrow “a” corresponds to a direction toward the collimator 1440. Along the direction denoted by the arrow “a”, the collimator 1440 may be arranged in different angles relative to the horizontal plane, for example, 90 degrees, 0 degrees, etc. FIG. 14B and FIG. 14C are schematic diagrams illustrating exemplary leaves of the multi-leaf collimator 1440 in a sectional view according to some embodiments of the present disclosure. As shown in FIG. 14B, the angle of the gantry 1420 is 90 degrees relative to the horizontal plane. The angle of the collimator 1440 is 0 degrees relative to the horizontal plane. Then, the leaf 1442 and the leaf 1444 are vertical to the horizontal plane. The leaf 1442 has an angle of 90 degrees. The leaf 1444 has an angle of −90 degrees. As shown in FIG. 14C, the angle of the collimator 1440 is 90 degrees relative to the horizontal plane, and the leaf 1442 and the leaf 1444 are parallel to the horizontal plane. The angle of the leaf 1442 and the leaf 1444 may be determined based on the angle of the collimator 1440 and the angle of the gantry 1420 according to Equation (1) as described in FIG. 7. FIG. 15 is a schematic diagram illustrating an exemplary relationship between the angle of a leaf and movement curve of the leaf according to some embodiments of the present disclosure. The angle of a leaf (also referred to as leaf angle) may range from −180 degree to 180 degrees. According to large amounts of experiments, the movement curve of the leaf is distributed according to the leaf angle. The movement curves of the leaf with a leaf angle in the range of 0 degrees and 90 degrees are similar to that of the leaf with a leaf angle in the range of 90 degrees to 180 degrees. The movement curves of the leaf with a leaf angle in the range from 0 degrees to −90 degree are similar to that of the leaf with a leaf angle in the range from −90 degree to −180 degree. In a first range between −90 degree and −10 degree, the movement curves of the leaf are similar to the movement curve with the leaf angle of −90 degree, which means the backlash errors of the leaf with a leaf angle in the first range are approximately equal to the backlash error when the leaf angle is −90 degree. In a second range between −10 degree and 10 degree, the movement curves of the leaf are similar to the movement curve with the leaf angle of 0 degree, which means the backlash errors of the leaf with a leaf angle in the second range are approximately equal to the backlash error when the leaf angle is 0 degree. In a third range between 10 degree and 90 degree, the movement curves of the leaf are similar to the movement curve with the leaf angle of 90 degree, which means the backlash errors of the leaf with a leaf angle in the third range are approximately equal to the backlash error when the angle of the leaf is 90 degree. FIG. 16 is a schematic diagram illustrating an exemplary movement curve of a leaf according to some embodiments of the present disclosure. The movement curve of the leaf may be obtained when the leaf has an angle of 0 degrees. The horizontal axis (i.e., X-axis) shows a position of the leaf which means the position of the leaf being acquired by a main encoder associated with the leaf. The vertical axis (i.e., Y-axis) shows a Hall position of the leaf which means the position of the leaf being acquired by a Hall sensor associated with the leaf. The solid line denotes that the leaf moves away from the center of a radiation field (i.e., the backward movement direction). The dotted line denotes that the leaf moves toward the center of the radiation field (i.e., the forward movement direction). As shown in FIG. 16, the movement curve of the leaf includes four sections corresponding to four movement phases of the leaf respectively. Section (a) corresponds to a first phase (also referred to as the forward movement phase) of the leaf that the leaf is moving toward the center of the radiation field. Section (b) corresponds to a second phase (also referred to as backward consume phase) that the leaf is static relative to a carriage of an MLC and configured to move away from the center of the radiation field. Section (c) corresponds to a third phase (also referred to as the backward movement phase) that the leaf is moving away from the center of the radiation field. And section (d) corresponds to a fourth phase (also referred to as forward consume phase) that the leaf is static relative to the carriage of the MLC and configured to move toward the center of the radiation field. Thus, four transition points associated with four movement phases are shown in FIG. 16. T1 represents a first transition point from the forwarding consume phase to the forward movement phase. T2 represents a second transition point from forward movement phase to the backward consume phase. T3 represents a third transition point from the backward consume phase to the backward movement phase. And T4 represents a fourth transition point from the forward consuming phase to the forward movement phase. FIG. 17 is a schematic diagram illustrating exemplary movement phases a leaf according to some embodiments of the present disclosure. As shown, exemplary movement phases of a leaf may include a forward movement phase 1704, a backward movement phase 1706, a backward consume phase 1708, and a forward consume phase 1710 as described elsewhere in the present disclosure (e.g., FIG. 9 and FIG. 16, and the descriptions thereof). An initial phase of the leaf is unknown corresponding to an initial position of the leaf. The movement phases of the leaf may be determined based on a count difference of a Hall sensor and/or an encoder associated with the leaf between two adjacent calculation cycles, such as kth and (k−1)th. For example, the leaf is determined in the forward movement phase 1704 if a count difference of a Hall sensor associated with the leaf between two adjacent calculation cycles (i.e., Δhall=hall (k)−hall (k−1)) is greater than 2 counts. If the count difference Δhall is less than −2 counts, the leaf is in the backward movement phase 1706. When the leaf is in the forward movement phase 1704, the movement phase of the leaf changes from the forward movement phase 1704 to the backward consume phase 1708 if a count difference of an encoder associated with the leaf between two adjacent calculation cycles (i.e., Δenc=enc(k)−enc(k−1)) is less than −5. When the leaf is in the backward movement phase 1706, the movement phase of the leaf changes from the backward movement phase 1706 to the forward consume phase 1710 if the count difference Δenc is greater than 5 counts. When the leaf is in the backward consume phase 1708, the movement phase of the leaf changes from the backward consume phase 1708 to the forward movement phase 1704 if the count difference Δhall is greater than 0 counts. On the contrary, the leaf gets into the backward movement phase 1706 with the count difference shall less than 0 counts. When the leaf is in the forward consume phase 1710, the movement phase of the leaf changes from the forward consume phase 1710 to the forward movement phase 1704 if the count difference shall is greater than 0 counts. On the contrary, the leaf gets into the backward movement phase 1706 with the count difference shall less than 0 counts. FIG. 18A is a schematic illustrating an exemplary angle relationship between a leaf, collimator, and gantry according to some embodiments of the present disclosure. The leaf moves with the rotation of a gantry 1820 and a collimator 1840. As shown in FIG. 18A, an angle of the collimator 1840 is 0 degrees relative to the horizontal plane described by a plane formed by X-axis and Y-axis, which is as same as the angle of the gantry 1820. Z-axis denotes a vertical direction. The angle of the leaf 1842 and the leaf 1844 are determined as 0 degrees relative to the horizontal plane based on the angle of the collimator 1804 and the angle of the gantry 1820 according to Equation (1) as described in FIG. 7. FIGS. 18B-18E are schematics illustrating exemplary backlash error of a multi-leaf collimator (MLC) according to some embodiments of the present disclosure. A leaf may be driven by a driving component including a ball screw and a motor (not shown). The ball screw may include a screw 1860 and a nut 1880. The screw 1860 and the nut 1880 may be configured with a plurality of teeth (e.g., a screw tooth 1862 on the screw 1860 and a nut tooth 1882 on the nut 1880). The backlash error may exist between the two teeth on the screw 1860 and the nut 1880 respectively when the motor changes rotation direction. FIGS. 18B-18E show backlash errors relating to the driving component of the leaf corresponding to the four movement phases as described elsewhere in the present disclosure (e.g., FIG. 16A). As shown in FIG. 18B, the leaf is in the forward movement phase driven by the motor associated with the leaf. Two teeth (e.g., the screw tooth 1862 and the nut tooth 1882) on the screw 1860 and the nut 1840 respectively are contacted by each other. There is no backlash between the two teeth (e.g., the screw tooth 1862 and the nut tooth 1882) on the screw 1860 and the nut 1880 respectively. As shown in FIG. 18C, the motor changes the rotation direction and the leaf is in the backward consume phase. There is a backlash error (e.g., distance d1) between two teeth (e.g., the screw tooth 1862 and the nut tooth 1884 adjacent to the nut tooth 1882) on the screw 1860 and the nut 1880 respectively. As shown in FIG. 18D, the leaf is in the backward movement phase driven by the motor associated with the leaf. Two teeth (e.g., the screw tooth 1862 and the nut tooth 1884) on the screw 1860 and the nut 1880 respectively are contacted by each other. There is no backlash between the two teeth (e.g., the screw tooth 1862 and the nut tooth 1884 adjacent to the nut tooth 1882) on the screw 1860 and the nut 1880 respectively. As shown in FIG. 18E, the motor changes the rotation direction and the leaf is in the forward consume phase. There is a backlash error (e.g., distance d2) between two teeth (e.g., the screw tooth 1862 and the nut tooth 1882) on the screw 1860 and the nut 1880 respectively. FIG. 19A is a schematic illustrating an exemplary angle relationship between a leaf, collimator, and gantry according to some embodiments of the present disclosure. The leaf moves with the rotation of a gantry and a collimator 1920. As shown in FIG. 19A, an angle of the collimator 1920 is 0 degrees relative to the horizontal plane, which is perpendicular to the gantry 1802. The angle of the leaf is determined as 90 degrees relative to the horizontal plane based on the angle of the collimator 1920 and the angle of the gantry according to Equation (1) as described in FIG. 7. FIG. 19B is a schematic diagram illustrating an exemplary movement curve of a leaf described in FIG. 19A according to some embodiments of the present disclosure. When an angle of the leaf changes from 0 degrees to 90 degrees in the forward movement phase, the movement curve of the leaf corresponding to the forward movement phase moves from section (a) to section (a)′. The second transition point T2 also moves to point T2′. When an angle of the leaf changes from 0 degrees to −90 degrees in the backward movement phase, the movement curve of the leaf corresponding to the backward movement phase moves from section (c) to section (c)′. The third transition point T3 also moves to point T3′. Moreover, certain terminology has been used to describe embodiments of the present disclosure. For example, the terms “one embodiment,” “an embodiment,” and/or “some embodiments” mean that a particular feature, structure or characteristic described in connection with the embodiment is included in at least one embodiment of the present disclosure. Therefore, it is emphasized and should be appreciated that two or more references to “an embodiment” or “one embodiment” or “an alternative embodiment” in various portions of this specification are not necessarily all referring to the same embodiment. Furthermore, the particular features, structures or characteristics may be combined as suitable in one or more embodiments of the present disclosure. Further, it will be appreciated by one skilled in the art, aspects of the present disclosure may be illustrated and described herein in any of a number of patentable classes or context including any new and useful process, machine, manufacture, or composition of matter, or any new and useful improvement thereof. Accordingly, aspects of the present disclosure may be implemented entirely hardware, entirely software (including firmware, resident software, micro-code, etc.) or combining software and hardware implementation that may all generally be referred to herein as a “unit,” “module,” or “system.” Furthermore, aspects of the present disclosure may take the form of a computer program product embodied in one or more computer readable media having computer readable program code embodied thereon. A computer readable signal medium may include a propagated data signal with computer readable program code embodied therein, for example, in baseband or as part of a carrier wave. Such a propagated signal may take any of a variety of forms, including electro-magnetic, optical, or the like, or any suitable combination thereof. A computer readable signal medium may be any computer readable medium that is not a computer readable storage medium and that may communicate, propagate, or transport a program for use by or in connection with an instruction execution system, apparatus, or device. Program code embodied on a computer readable signal medium may be transmitted using any appropriate medium, including wireless, wireline, optical fiber cable, RF, or the like, or any suitable combination of the foregoing. Computer program code for carrying out operations for aspects of the present disclosure may be written in a combination of one or more programming languages, including an object oriented programming language such as Java, Scala, Smalltalk, Eiffel, JADE, Emerald, C++, C#, VB. NET, Python or the like, conventional procedural programming languages, such as the “C” programming language, Visual Basic, Fortran 2103, Perl, COBOL 2102, PHP, ABAP, dynamic programming languages such as Python, Ruby and Groovy, or other programming languages. The program code may execute entirely on the user's computer, partly on the user's computer, as a stand-alone software package, partly on the user's computer and partly on a remote computer or entirely on the remote computer or server. In the latter scenario, the remote computer may be connected to the user's computer through any type of network, including a local area network (LAN) or a wide area network (WAN), or the connection may be made to an external computer (for example, through the Internet using an Internet Service Provider) or in a cloud computing environment or offered as a service such as a Software as a Service (SaaS). Furthermore, the recited order of processing elements or sequences, or the use of numbers, letters, or other designations, therefore, is not intended to limit the claimed processes and methods to any order except as may be specified in the claims. Although the above disclosure discusses through various examples what is currently considered to be a variety of useful embodiments of the disclosure, it is to be understood that such detail is solely for that purpose and that the appended claims are not limited to the disclosed embodiments, but, on the contrary, are intended to cover modifications and equivalent arrangements that are within the spirit and scope of the disclosed embodiments. For example, although the implementation of various components described above may be embodied in a hardware device, it may also be implemented as a software only solution, for example, an installation on an existing server or mobile device. Similarly, it should be appreciated that in the foregoing description of embodiments of the present disclosure, various features are sometimes grouped in a single embodiment, figure, or description thereof for the purpose of streamlining the disclosure aiding in the understanding of one or more of the various inventive embodiments. This method of disclosure, however, is not to be interpreted as reflecting an intention that the claimed subject matter requires more features than are expressly recited in each claim. Rather, inventive embodiments lie in less than all features of a single foregoing disclosed embodiment. In some embodiments, the numbers expressing quantities or properties used to describe and claim certain embodiments of the application are to be understood as being modified in some instances by the term “about,” “approximate,” or “substantially.” For example, “about,” “approximate,” or “substantially” may indicate ±20% variation of the value it describes, unless otherwise stated. Accordingly, in some embodiments, the numerical parameters set forth in the written description and attached claims are approximations that may vary depending upon the desired properties sought to be obtained by a particular embodiment. In some embodiments, the numerical parameters should be construed in light of the number of reported significant digits and by applying ordinary rounding techniques. Notwithstanding that the numerical ranges and parameters setting forth the broad scope of some embodiments of the application are approximations, the numerical values set forth in the specific examples are reported as precisely as practicable. Each of the patents, patent applications, publications of patent applications, and other material, such as articles, books, specifications, publications, documents, things, and/or the like, referenced herein is hereby incorporated herein by this reference in its entirety for all purposes, excepting any prosecution file history associated with same, any of same that is inconsistent with or in conflict with the present document, or any of same that may have a limiting affect as to the broadest scope of the claims now or later associated with the present document. By way of example, should there be any inconsistency or conflict between the description, definition, and/or the use of a term associated with any of the incorporated material and that associated with the present document, the description, definition, and/or the use of the term in the present document shall prevail. In closing, it is to be understood that the embodiments of the application disclosed herein are illustrative of the principles of the embodiments of the application. Other modifications that may be employed may be within the scope of the application. Thus, by way of example, but not of limitation, alternative configurations of the embodiments of the application may be utilized in accordance with the teachings herein. Accordingly, embodiments of the present application are not limited to that precisely as shown and described.
	summary
	summary
	description	The present application is a national phase entry under 35 U.S.C § 371 of International Application No. PCT/EP2018/068429 filed Jul. 6, 2018, which claims priority from French Application No. 1756372 filed Jul. 6, 2017, all of which are hereby incorporated herein by reference. The present invention relates to the field of reactor vessels of nuclear power plants. The present invention relates more specifically to improved centering pins for centering reactor cores in the nuclear reactor vessels. The present invention particularly applies to all reactors of the pressurized-water nuclear power plants of second or third generation, for which the core is cooled using a heat transfer fluid circulating through the core. The general structure of nuclear power plants is known to those skilled in the art. This general structure is illustrated in the appended FIG. 1 which shows the vessel 10 housing a core 20. Such a power plant comprises a primary circuit 30 and a secondary circuit 40. The heat produced by the fission of an enriched material in the core 20 of the reactor heats up the pressurized water of the primary circuit 30. The heat from the primary circuit 30 evaporates the water from the secondary circuit 40 on passage through a steam generating exchanger 42. A turbine-generator set 60, 62 converts the mechanical power of the steam into electrical power. FIG. 1 furthermore schematically represents in 22 control rods, in 32 a pressurizer, in 34 a primary pump, in 12 a sealed enclosure, in 64 a condenser and in 66 an extraction pump. In a conventional manner per se, as illustrated in the appended FIG. 2, taken from document EP-1003171, the core 20 is centered in the vessel 10 by centering pins 100. Thanks to this structure, before entering the plenum, i.e. the lower part 13 of the vessel, the cooling water (heat transfer fluid) flows into the annular space 11 defined between the shell 20 of the core and the outer wall 10 of the vessel. The cooling water which flows around the centering pins 100 and enters the plenum 13 is then redirected toward the core 20 through a bottom plate 14 of the core. FIG. 3 represents a partial view in vertical section of the lower part of a known vessel 10 and particularly illustrates centering pins 100. FIG. 4 represents a view at an enlarged scale of such a centering pin 100 in accordance with the state of the art. In FIG. 3, the arrows designate the progress of cooling water in the annular space 11 then in the plenum 13. As can be seen in FIGS. 3 and 4, the pins 100 known in the related art are of vertical and horizontal square section. Therefore, they have a general shape of a rectangular parallelepiped. The inventors have observed that the shape of these known centering pins 100 disturbs the flow of the cooling water injected into the plenum 13 of the vessel bottom. The thus generated instabilities contribute to the heterogeneity of the core 20 cooling flow and have a strong impact on the heat and neutron exchanges of the assembly. A heterogeneous flow at the inlet of the core 20 modifies the operation of the core. Therefore, there is an impact on the heat and neutron exchanges of the assembly, on the mechanical loads and thus a negative effect on the efficiency and the lifetime of an assembly. The heat exchange between the water and the assemblies of the core depends on the flow rate. A high flow rate creates a strong heat transfer. At the same time, the water acts as a moderator for the neutron reaction. A strong flow rate of the water increases the intensity of the neutron reaction and thus increases the fuel consumption and the temperature of the assembly. When the distribution of the flow rate layer varies between neighboring assemblies, there is a temperature differential between the assemblies. The temperature induces a mechanical deformation (expansion) of the assembly and temperature differences between neighboring assemblies can cause mechanical loads. These mechanical loads can damage an assembly or set of assemblies. A very heterogeneous flow rate layer at the inlet of the core has therefore an impact on the heat and neutron exchanges of the assembly, on the mechanical loads and thus a negative effect on the efficiency and the lifetime of an assembly. Attempts to solve the problem thus posed have already been proposed. For example, it has been proposed to add structures in the vessel bottom 13. However, the most effective shapes have a lot of space requirement for the flow and the pressure loss of the primary circuit 30 is increased. This requires a re-dimensioning of the pumps of the primary circuit 30. This is a significant change in the design of a nuclear power plant. In conclusion, none of the solutions proposed so far is totally satisfactory. The object of the present invention is to improve the state of the art. A first object of the invention is to improve the hydrodynamics of the fluid flow of the primary circuit in the core of a nuclear power plant. Another object of the invention is to propose solutions without imposing significant changes to the design of a nuclear power plant. The aforementioned objects are achieved within the framework of the present invention thanks to the profiling of the centering pins for centering a core of a nuclear power plant in a reactor vessel. The profiling is characterized by a hydrodynamic profile on the upstream and downstream faces of the pin to reduce instabilities when the coolant is circulating around the pin and by the fact that the height of the upper hydrodynamic profile located upstream in the direction of flow of the fluid is in the order of magnitude of the horizontal thickness of the pin. According to other advantageous characteristics of the invention: hydrodynamic profiles are provided on the upstream face and on the downstream face of the pin, the height of the hydrodynamically-profiled lower part is in the order of magnitude of twice the horizontal thickness of the pin, the upper hydrodynamic profile has a general shape of a dihedral, the lower hydrodynamic profile has a shape of a pyramid, the lower hydrodynamic profile extends up to the vessel bottom. The present invention also relates to a reactor vessel of a nuclear power plant equipped with a plurality of centering pins of the aforementioned type. The present invention also relates to reactors of nuclear power plants including such a vessel and such pins. As indicated above, the general structure of a nuclear power plant is known to those skilled in the art. It will therefore not be described in detail. It is simply recalled that the core is centered in the vessel by centering pins. The appended FIGS. 5, 6 and 7 show a pin 110 according to the present invention. More specifically, the pin 110 is defined by a radially inner axial face 112 adjacent to the core 20, which conforms to the shape of the core 20 at this level and which is generally stepped. The pin 110 also includes a radially outer axial face 114 adjacent to the vessel 10 and which conforms to the shape of the vessel at this level. Thus, the face 114 is preferably curved. It converges toward the radially inner axial face 112 downwards. The central part 111 of each pin 110 is completed with two radial faces 116, 118 generally planar and radial relative to the vertical central axis of the core. As indicated above, according to the invention, each pin 110 is equipped with a hydrodynamic upper profile 120 and a hydrodynamic lower profile 130, disposed respectively above and below the central part 111. The hydrodynamic upper profile 120 located upstream in the direction of flow of the fluid entering the annular space 11, has the shape of a dihedral 121. More specifically, this dihedral 121 is formed of two symmetrical main facets 126, 128 which extend the radial faces 116 and 118 upwards. The facets 126, 128 are preferably generally flat and meet at a radial ridge 129 relative to the axis of the core. The upper profile 120 is completed with two radially inner 122 and outer 124 secondary facets respectively adjacent to the outer face of the core and to the inner face of the vessel. The lower hydrodynamic profile 130 has the general shape of a pyramid 131. It comprises two symmetrical main facets 136, 138 which extend the radial faces 116 and 118 downwards. The radially outer edges 135 of the main facets 136, 138 follow the contour of the vessel 10 and meet at the radially inner edge 137 at a point of convergence 139. The lower hydrodynamic profile 130 also comprises two radially inner 132 and outer 134 secondary facets. The radially outer 135 and radially inner 137 edges of the main facets 136, 138 converge to join the two secondary facets 132 and 134 of the pyramid 131 composing the lower profile, at the tip 139. Of course, the present invention is not limited to the particular embodiment that has just been described, but extends to all variants according to its spirit. Particularly, the hydrodynamic profiles 120 and 130 can be of angular or rounded geometry. The best possible embodiment of a hydrodynamic device is a shape that eliminates any stagnation space (low-flow area) in the vicinity of the centering pins 110. The present invention makes it possible to reduce or even totally eliminate turbulence at the bottom of the vessel. The invention makes it possible to stabilize the flow on its passage at the centering pins 110 and therefore reduce the heterogeneity of the flow upstream of the core and thus make the distribution of the heat transfer fluid more homogeneous in the core. FIG. 3 proposes an example of a pin 110 consisting of angular geometries. For this case, the pin 110 is defined by two parameters b and c. The parameter b (illustrated in FIG. 7) corresponds to the vertical height of the upper profile 120. The parameter c (illustrated in FIG. 7) corresponds to the vertical height of the lower profile 130. These parameters b and c are defined in relation to the thickness a of the centering pin 110 (illustrated in FIG. 6) considered in a horizontal section and along the radially inner edge of the central part 111 of the pin 110. In the example illustrated in FIGS. 5 to 7, the height b of the hydrodynamic upper part 120 of the pin 110 is in the same order of magnitude (0.87 times) as the thickness a of the centering pin 110. The height c of the hydrodynamic lower part 130 of the pin 110 is about twice (1.93 times) the thickness a of the centering pin 110. This shape is chosen to make the centering pin 110 hydrodynamic for the water flowing down through the annular space 11. The centering pin 110 with the hydrodynamic devices 120 and 130 thus takes the shape of a wing. The upper part 120 of the pin 110 represents a leading edge of the wing and the lower part 130 represents a trailing edge of the wing. The inventors have found that: If the height b of the upper part 120 tends to zero, the device 120 has no more effect. If the height b of the upper part 120 becomes too large, the device 120 blocks the rotation of water in the annular space 11. The rotation of water in the annular space is however significant because it mixes the water coming from the cold branches. This is an effect that stabilizes the circulation in the event of flow rate imbalance that may occur between cold branches. With the choice of a height b in the order of magnitude of the thickness a of the centering pin 110, an effective hydrodynamic shape is obtained without blocking the mixture in the annular space. The height c of the lower part 130 of the device, in the order of twice the thickness a of the centering pin 110, is chosen so that the pin is extended vertically up to the bottom of the vessel. The device is made less effective if it does not extend up to the bottom of the vessel. The device 130 can induce space requirement and even a source of instability if it takes a larger shape that does not follow the vertical axis and approaches the center of the vessel. Within the framework of the invention, it is meant by “same order of magnitude” a maximum variation of more or less 25% relative to a reference value.
	abstract	A target supply unit includes a nozzle unit having a through-hole to allow a target material to be outputted therethrough. A cover is provided to cover the nozzle unit, the cover having a through-hole to allow the target material to pass therethrough. A discharge device is included to pump out gas inside a space defined by the cover.
053393465	abstract	Submicron device fabrication entailing ringfield x-ray pattern delineation is facilitated by use of a condenser including a faceted collector lens. The collector lens is constituted of paired facets, symmetrically placed about an axis of a laser-pumped plasma source. Each of the members of a pair produce an image of the entire illumination field so that inhomogeneities in illumination intensity are compensated within each composite image as produced by a particular pair.
055457967	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to FIG. 1 of the drawings, one embodiment of a contaminated waste storage container 10, made solely of contaminated metal, is shown. This container has no interior or exterior concrete support or shielding associated with it. The metal for the container is made by melting radioactive rods, tubing, metal nuclear components and the like in a metal melting furnace as described in detail later in this specification. In the furnace, the bottom, lower-grade radioactive melt is separated from a top higher-grade radioactive slag phase. This bottom molten metal is cast into a container form including an associated lid, and cooled. The metal melting process involves expensive, regulated equipment, and so containers made by this process should serve special disposal needs of utilizing radioactive feed metal having at least a 130 Bq/g activity. This container could be used and transported under a variety of circumstances, but would be particularly advantageous where melting, casting the container, filling with contaminated waste material and burial or storage would be in the same site, so that feed metal having as high as or over 370 Bq/g activity can be used, saving tremendous storage or burial volume for very "hot" metal. FIG. 2 shows one embodiment of a structure or article such as slab wall 2 which can be made of radioactive waste, hazardous waste, or their mixtures. Since this structure may not necessarily be used to directly contain contaminate material, when contaminated metal is used alone in this embodiment, the contaminated metal should not be diluted or alloyed with virgin filler or the like to the point that the contaminated metal constitutes less than 35 weight % of the structure. In this instance, metal slag residue from a metal melter, for example, can be used to make such a slab walls 2, bricks, panels, blocks, sheets, slabs, grids, floors, liners, impact, limiters or the like, as described in detail previously. This structure could also, for example, have a contaminated or non-contaminated, particulate, plastic or rubber matrix containing radioactive concrete, metal, or slag, or a contaminated metal mat matrix filled with plastic. FIG. 3 shows another embodiment of the invention, where a storage module 10 is a container having walls and lid substantially or completely of plastic. Useful plastic or plastic resins include, for example, rubber compounds, polypropylene, polycarbonate, polyester, polyvinylidene fluoride, polyvinyl acetate, polyvinyl chloride, epoxy resins, phenolic resins and, preferably, polyethylene which contains a large number of hydrogen atoms per cm.sup.3, mixtures and copolymers thereof, and the like, all of which are well known plastics or plastic resins. When the plastic is thermosetting (will degrade rather than melt upon heating) to get good bonding, the plastic should be uncontaminated. This is usually not as important when thermoplastic resins are used since they can be remelted after initial use. Epoxy and phenolic resins are examples of thermosetting resins. When plastic is used as the binder material making up the container walls, hazardous solids in the form of, for example, soil, and "toxic" chemicals such as polychlorinated biphenyls, petroleum hydrocarbons, pentachlorophenols (PCB's, PHC's and PCP's respectively) and the like, can be included. This module can enclose a high integrity container as shown. While not shown in FIG. 3, the plastic container can contain contaminated metal fibers in discrete form distributed throughout the walls, bottom and lid and/or contaminated metal mesh or contaminated metal mat. The contaminated metal fibers that can be used in discrete form will have the same ranges as those set forth later in this specification for concrete containers. The contaminated metal mesh or mat can be from 60% to 95% porous (5% to 40% of theoretical density). When a metal mesh or mat is used in the plastic container it will provide a metal, open matrix, readily completely impregnated by liquid plastic resinous materials. The wall thickness of the plastic container shown in FIG. 3.can range from 0.5 cm (0.2 inch) to about 7.6 cm (3 inches), and the metal mesh or mat can constitute from 10% to 100% of the final wall thickness, that is, if the plastic container wall is 5 cm thick, the central metal mesh or mat skeleton or matrix can be from 0.5 cm to 5 cm thick. Preferably, however, metal fiber will not penetrate through the exterior wall surface. In processing, the mesh or mat could be placed in a mold and flowable, liquid plastic resin could be injected or poured into the mold to completely impregnate the open, porous mesh or mat; and then the plastic/mesh or that combination can be heated under pressure to cure the plastic resin and cause it to fill all voids, and consolidate the container walls. The plastic can comprise two copolymerizable components or a single plastic resin, all with appropriate diluents, hardeners, flow control agents, catalysts, initiators, and other appropriate additives. Referring now to FIG. 4 of the drawings, one embodiment of a unitary structure such as storage module 10 is shown, which includes a container 12 formed of concrete or other suitable material, having a bottom and sidewalls. The container 12 is closed by lid 14 placed atop the upper most edges of the container. The lid 14 is attached to the container by way of ridge 13 on container 12 and a recess 15 on the lid 14. For the sake of facilitating a stacking relationship of the storage modules 10 in adjacent columns, the storage modules 10 can be shaped as hexagonal prisms, as shown in FIG. 4. Each of the sides of the hexagon is illustrated as a substantially flat side 16, and between each of the sides 16 is corner side 18. When the storage modules 10 are stacked, the sides 16 of adjacent modules 10 abut one another, to define a honeycomb-type arrangement when viewed as a plan view from above. Forklift grooves 22 are shown at the bottom of the module. In all instances, the containers described here and previously can be used as multipurpose containers for processing, transport, storage and/or burial, and can be transported from the manufacturing site of the container to the waste site, loaded with contaminated waste, and transported to a burial site for burial or storage. Also, processing such as solidification, dewatering, and the like, can be carried out in the container before it is sealed. Within the container 12 in FIG. 4, the sidewalls 16 and 18 define an interior space 20, which is preferably cylindrical. Within the interior space 20, several stacks of steel drums 26 are illustrated. These drums can contain radioactive or hazardous waste in liquid, solid or compressed form. Here, the module walls do not directly contact the waste. In this invention, the container and lid walls 24 also contain processed radioactive, hazardous, or mixed waste material intermixed as discrete fibers, particles, or the like with concrete and other additives. A void space, created in the cylindrical interior space 20 between the waste containers or packages 26 and sides of the container 12, can be filled with a contaminated or noncontaminated granular fill material 36, such as a cementitious grout. FIG. 5 of the drawings shows one embodiment of a square or round contaminated waste storage container made of concrete or other suitable material having sides 16 and lid 14, with or without a bottom. The sides and top can be set into place by means of lugs 30 and held in place by bolts 32. This module can enclose a high integrity container 34. The container and lid walls 24 in this invention will contain processed radioactive, hazardous or mixed waste material as described previously. A plastic liner is shown as 35. FIG. 6 shows an idealized, enlarged crosssection of an example of the walls 24 of some of the containment systems previously described. As shown, a particulate sand filler 36, shown as medium sized open circles, and cement 38, shown as medium sized shaded circles, form a matrix 42 which would make up most of the body of the wall 24. Interdispersed in the matrix material 24 could be fine flyash particles shown as 40, radioactive metal fibers 44 and/or radioactive, ground, large concrete aggregate or other hazardous material 46 shown as hatched large sized open circles, and virgin, ground, large aggregate shown as large sized open circles 50. While not shown, cut pieces of contaminated plastic material may also be included within the matrix. Unlike the use of fibers or aggregate in a containment structure wall made of totally virgin, uncontaminated materials, it is preferred that the radioactive materials used herein be in a non-agglomerate form and do not clump to create a fiber-aggregate volume 48 of high radioactivity; or at least such agglomerates must be minimized. The radioactive fibers, radioactive concrete, and hazardous solid waste should be uniformly and homogeneously dispersed throughout the matrix 42 so that they are essentially discrete fibers and particles separated and encapsulated by the matrix material. In the case of a plastic sheet or plastic particulate matrix, without use of metal fibers or metal mat, the contaminated particulates 46 or fibers 44 would be dispersed between encapsulating plastic particulates or connected chains or polymers. As can be seen in FIG. 6, the largest particles are aggregate particles 46 and 50 and the next largest particles are cement 38 and sand or other filler 36. The cement particles 38 along with the sand or other filler particles 36 form a matrix containing all the other materials. The flyash particles 40, if used, would be the next smallest, and together with the very small, shaded, silica fume particles 52, if used, would provide a series of different sized particles to provide substantially complete interior void volume filling, providing an essentially void free, low porosity article. As described previously, small metal fiber material such as stainless steel can also be used in the structure. Also, spherical and amorphous particles can be used in place of or with the metal fibers 44 shown. It may also be desirable to include uncontaminated carbon, ceramic, plastic or fiberglass fibers as additional reinforcement. In order to accomplish such a radioactive dispersion in the matrix, the radioactive fibers, concrete particles and hazardous waste solids should meet important size and profile parameters and must be added in a certain sequence to the composition mixture. FIG. 7 shows one embodiment of the method of this invention, which will here be described in substantial detail for use of an uncontaminated sand and cement matrix and radioactive metal and radioactive concrete additives, as an example, although the method of this invention is not at all so limited, and the article can include metal alone, metal with hazardous or mixed waste, or a variety of metal or plastic binders such as steel, polyethylene resin, plastic with a metal mat matrix, and the like, as pointed out in detail previously. FIG. 7 shows two Flow Paths for treatment of radioactive additive. Flow Path I relates to radioactive metal treatment and Flow Path II relates to radioactive concrete treatment. In Flow Path I, radioactive metal, such as stainless steel tubes used in cooling nuclear reactors, piping used on nuclear sites, metal spent fuel channel boxes, centrifuges, compressors and motors used for uranium enrichment, siding used on nuclear reactor site buildings, other stainless steel, nickel, iron, lead, chromium, technetium or other radioactive metal components used in or near nuclear reactors, or the like can be the radioactive starting material. In most cases the metal radioactive starting material will be inspected, and segregated according to metal type, for example, a run of all stainless steel may be made separate from a run of all carbon steel, based on the desired end product to provide a radioactive metal feed (A). All manner of configurations can be used, although in most instances the metal feed will be cut to a convenient size for further treatment, with, for example, shears, acetylene torch, or if necessary an underwater plasma torch, or the like. The radioactive metal feed will then be transported to a metal melter, such as an induction furnace operating at a temperature over about 1,200.degree. C., which will be described later in the Examples. In the metal melting furnace, a purified, lower grade radioactive, all metal bottom phase is formed, and separated from an impure, higher grade radioactive metal slag top phase. Up to 15 weight % uncontaminated metal as additive, preferably only up to 10 weight % uncontaminated metal, such as nickel, chromium, or the like, may, in some cases, be added and reacted in the furnace where it may be necessary to provide a bottom phase metal composition having special strength characteristics necessary to cast fibers or other articles or structures. The all metal bottom phase can be used for further processing into metal fibers, or can be cast directly into a variety of articles, such as blocks, slabs, walls, containers, and the like by means of path 53. Generally, if the slag phase is not separated, any metal fibers produced would not have the physical properties required to provide high tensile strength to the concrete matrix into which they are added, or the metal articles or structures, such as slabs, would not have the required composition for the embodiment needed. However, in rare instances, where very pure radioactive metal feed is used, a top slag phase may not form. Step (B) in FIG. 7 can produce spun metal fibers from the molten metal or cast metal ingots. Such melt spinning of fibers is well known in the art, and further details on such a melt casting process can be found in U.S. Pat. Nos. 4,930,565 (Hackman et al.) and 4,907,641 (Gaspar). While the top phase slag can be poured into molds to make highly radioactive shield block ingots by means of path 53', a portion of the slag could also be quenched by water to form generally spherical particles or agglomerates of amorphous shaped metal which can be used separately as filler via path 54, or added to the sand as shown in FIG. 7. The size and width or diameter of included metal bars, fibers, generally spherical particles and amorphous particles affects tensile strength and compressire strength of the containment system in which they are used. Radioactive or non-radioactive, cast, reinforcing bars, when used, can have substantial lengths, preferably, of from about 25 cm to about 50 cm and diameters of from about 0.10 cm to about 3 cm. The radioactive metal fibers should have lengths of from about 0.5 cm to about 20 cm, preferably from about 1.0 cm to about 3.5 cm and have a length:width (length to width) aspect ratio of between 200:1 and 20:1, preferably between 150:1 and 75:1. Therefore, if the fiber length is 10 cm, the width or diameter can range from 0.05 cm to o.5 cm. Fibers of non-conforming geometry are sent back through the metal melt and fiber casting process. The most preferred fibers are stainless steel fibers having a chromium content of from 15% to 26% and a nickel content of from 8% to 14%. When used in concrete the preferred weight range of the metal fibers is from about 2% to about 55%, most preferably from 2 weight % to 30 weight %. The generally spherical metal particles can have diameters from about 0.001 mm to about 30 mm and the amorphous metal particles can have a thickness of from about 0.01 mm to about 30 mm. If the fibers are below 0.5 cm in length, they would have no advantageous effect on the tensile strength of the concrete. If the fibers are over 20 cm in length, they will clump and deform, and will not have the desired effect on the tensile strength of the concrete. The metal will be free of any oil residue through the melt casting and prudent handling and storage. In the Flow Path Row II concrete processing steps, large sections or slabs of radioactive concrete or large size radioactive gravel are provided in step (A). The concrete may be wall sections, chunks, slabs, and dust, resulting from demolition of or around nuclear reactor structures, or it can be used waste containers which have acquired low level radioactivity, and the like. The concrete can be pulverized to provide generally spherical particles having diameters (rough diameters) from about 0.001 mm to about 30 mm in step (B). Road gravel which has become radioactive over the years from vehicles or the like traveling over nuclear site roads using such gravel can also be used, as well as gravel from filtration ponds or the like. This gravel if of large size can also be ground. Its final size should be in the same range as the pulverized concrete. If the particles are below 0.001 mm diameter, it may be detrimental to the strength characteristics of the concrete. If the particles are over 30 mm diameter, it would be difficult to form containers and the like, particularly those having thin walls. Preferably, the radioactive gravel or concrete particles are not of one size but are distributed substantially equally between 0.001 mm and 30 mm diameter. A variety of particle sizes are required in the embodiment of the method of this invention, the largest being the radioactive concrete aggregate or radioactive road gravel and any virgin aggregate used, the next finest being the filler and virgin sand used with the Portland cement, followed by fine flyash and ultra-fine silica fume, as described later; which all interact to provide high density and low permeability by having small particles within a larger particle matrix and eliminating most void volume. Radioactive concrete dust could be used in association with the uncontaminated sand as shown by path 52 in FIG. 7. Preferably, all the radioactive concrete material will be reduced in size by crushing or grinding to provide a "fresh", wettable surface. As mentioned previously, uncontaminated river "pea" gravel or man-made virgin aggregate can be used in addition to the radioactive concrete particles. The uncontaminated aggregate will have the same rough diameter particle size range of about 0.001 mm to 30 cm as the radioactive concrete particles. Based on the amount of radioactive concrete particles to be used, the amount of virgin gravel can be determined so as to provide a useful, dry concrete mix, approximating weight ratios of (aggregate):(sand):(cement):(water) of (10):(5 to 7):(3 to 4):(1 to 2). The particle size of the sand, which can be river sand or finely ground rock or aggregate and which acts as filler should be in the rough diameter particle size range of from about 0.015 mm to about 10 mm, a range below the aggregate size. Amorphous radioactive metal slag particles resulting from water quenching of slag, as shown in FIG. 6 can be used as filler in substitution for part of the sand as long as it is round if needed to fit into the sand particle size range, that is, a thickness of from about 0.015 mm to about 10 mm. As previously mentioned, any very fine radioactive concrete dust in the sand particle size range can be added with the sand in making the initial cement mixture, but its weight amount will be determined by the limits on weight % radioactive concrete allowed to be added. Portland cement will be used in an amount determined by the above (aggregate):(sand):(cement):(water) ratio. The Portland cement is uncontaminated and is preferably a low heat of hydration cement, that produces a minimum of heat during cure, and which requires less water than standard cement. Such cement is commercially available and generally designated as moderate to low hydration Type IV, or low hydration Type IV. It is preferred to limit the amount of water used in the concrete mixture, supplementing the need for water for workability with plasticizer materials. The water used, can be regular, uncontaminated water, or water that is radioactive. If the water is radioactive it can be processed by filtration to remove organic impurities to provide a stabilized liquid. Plasticizers are used with concrete to increase the plasticity of the concrete mixture for extended periods of time. Useful plasticizers are commercially available under the tradename RHEOBUILD (manufactured by Master Builders Co.). These plasticizers are commonly salts, such as calcium or sodium, of beta-naphthalene sulfonate polymers or other hybrid mixtures in compliance with ASTM C-494. For the purpose of this invention, the plasticizer is added not only to reduce water content but to increase the workability of the concrete mix and its flowability and extend the possible mixing time such that thorough mixing of all components will occur. The plasticizer must be added to the concrete mixture in the final stages after the aggregate, sand, cement, and other chemical additives have been introduced into the "dry" mix and initial slump tests are taken. Use of the plasticizer prior to initial slump testing can lead to erroneous water-cement ratio calculations. The amount of water used in this invention is in accordance with the weight ratio previously described, and such that the consistency of the concrete mixture will have a 3 cm to 7 cm slump after addition of sand and water and prior to addition of plasticizer, where the term "slump" means the amount of contraction of the top of a cone of concrete upon cure and is a term standard in the art, defined in ASTM C-143, and where no subsidence is O slump. Use of minimal water provides a desirable, relatively dry consistency cement mixture. Subsequent addition of plasticizer will increase the slump level and flowability or plasticity of the concrete. Radioactive water, such as that resulting from quenching slag as shown in FIG. 7 can be fed by path 56 to replace some or all of the water used in the concrete mixture. Other components that are added to the concrete mixture, as shown in FIG. 6, are flyash 40 and silica fume 50. Flyash is the very fine ash produced by combustion of powdered coal with forced draft, and often carried off with the fuel gases from such processes. A baghouse filter or electrostatic precipitator is necessary for effective :recovery. Considerable percentages of CaO, MgO, silica and alumina are present in the flyash. The particle size of the flyash is preferably from about 0.001 mm to about 0.01 mm. This provides particles finer than sand and larger particles than silica fume. Silica fume, or fumed silica, is a colloidal form of silica, SiO.sub.2, made by combustion of silicon tetrachloride in hydrogen-oxygen furnaces. It is a fine white powder, and for the purposes of this invention will have a particle size range from about 0.00015 mm to about 0.0015 mm, providing ultra-fine particles which are extremely important in adding strength and increasing the density of the cured containment system, so that it has a low permeability eliminating leakage possibilities. The preferred range of these components is about 0.2 to about 2 parts by weight of silica fume, and about 0.5 to about 4 parts by weight of flyash, based on 100 parts of binder material, where from about 0.1 to about 1 part by weight of chemical plasticizer, based on 100 parts of binder material is added with the processed, radioactive material, based on mix workability requirements. Other additives can also be used, for example air entrainer materials, which, when added in a small effective amount, causes microscopic air bubbles in the cured containment system upon cure. These microscopic air bubbles provide an insulative effect and increase freeze/thaw resistance to cracking. Another useful additive is a hardener which also allows reduction of water content and improves workability and finish. Although virtually any circumstance is possible, clean, uncontaminated cement will be used when it is a binder; and thermoplastic resins, such as epoxy resins will always be clean and uncontaminated when used as a binder, as mentioned previously, to insure good bonding. After the concrete mixture is thoroughly mixed and at a consistency of about a 3 to 7 cm slump, the plasticizer and radioactive metal, are slowly added, preferably, over a 10 minute to 20 minute period, at a stir-mixing rate, preferably, of approximately 30 rpm to 50 rpm, for batches of 900 to 2,700 kg. The resulting form of the cast composition can be any of those shown in the drawings, for example, the container structure of FIGS. 4 and 5, or the wall or barrier structure of FIG. 2, and the like. These structures can have a liner coating, or layer, 35, as shown, for example in FIG. 5, on the inside or outside, such as a plastic resin, water barrier coating, metallized coatings, and the like, to serve a variety of purposes including preventing stirred liquids from leaking out or exterior water or liquids from leaking in. Plastic or plastic resins, described previously, for example polyethylene, polyvinylidene fluoride, polypropylene and polyvinylchloride, are particularly advantageous for inner and/or outer plate or sheet coverings of the exterior of the container or as closely conforming interior liners, having thicknesses of from about 0.2 cm to about 2.5 cm. These liners and exterior covers can add substantially to fracture resistant properties as well as containment and leak proof properties. These plastic resin liners and exterior covers can be conveniently used in concrete fabrication of containers as inner and outer forms for the concrete. These liners and exterior covers, preferably, are closely attached to the concrete, most preferably by means of anchor means, integral to the liner or cover, which are embedded in the concrete. For example, liner or cover ribs, dovetails or T portions and the like, extending from the liner or cover into the concrete to anchor, connect and interlock the plastic resin and the concrete upon the concrete setting are advantageous, as described in U.S. Ser. No. 07/758,220, filed on Sep. 12, 1991 by the assignee of this invention, entitled "Storage Module For Nuclear Waste With Improved Liner" (Meess W. E. 55,126-C2). The liner or cover can also be simply molded or injected to close fit, and possibly impregnate the surface of the concrete, creating a bond with the concrete container, or can be glued in place by a suitable, high strength, water resistant adhesive. In the case of making a plastic container such as shown in FIG. 3, the following general steps would be taken. First, contaminated plastic drums, bags or the like are cut to an appropriate size so that they can be melted to form a fluid, pourable mass. Then the hazardous fluid mass would be, generally, centrifugally cast, as is well known in the art. This casting can be used in conjunction with the metal fibers or metal mesh or metal mat as previously described. The following examples further illustrate the invention, and should not be considered limiting in any way. EXAMPLE 1 Contaminated waste structures, in the form of blocks, were made utilizing a licensed, pilot, metal melter induction furnace from cooled, cast, melted, radioactive metal tubing waste. The all-metal blocks had a specific activity over about 130 Bq/g and had no support. The blocks were substantially free of slag residue. Blocks of this type would be useful for shielding material. A transportable, multipurpose container similar to that shown in FIG. 1 could also easily be cast from such radioactive metal tubing waste. EXAMPLE 2 Two small box forms were made to provide cast concrete containers having outside dimensions of 406 mm wide.times.406 mm long.times.355 mm high, with an internal right circular cylinder cavity measuring 300 mm in diameter by 300 mm high. The forms were coated with an organic release material to aid in stripping once the concrete was poured. The forms were also coated with sealant at the joints to prevent concrete bleed. These forms were then filled with a radioactive concrete mixture, resulting in containers with 50 mm thick walls and bottom to which a separate lid was formed having a minimum thickness of 50 mm and outside dimensions of 40.6 cm.times.40.6 cm to match the container which was 40.6 cm.times.40.6 cm. The completed modules provided stackable modules providing a minimum of 50 mm of concrete between the internal cavity and the environment. The container is the same as shown in FIG. 8 as 80, with top 81 and cavity 82 for waste material. While this container sacrificed packaging efficiency with thickening of walls at the diagonal corners, its fabrication was considered acceptable for a prototype. The concrete mix for both containers contained highly radioactive metal slag agglomerates, contaminated fibers, contaminated concrete aggregate and contaminated water, where both samples also contained uncontaminated aggregate, sand, cement, and flyash. Radioactive slag was obtained from the Westinghouse Electric Corporation, Scientific Ecology Group (SEG), metal melt facility at Oak Ridge, Tenn. The slag material was analyzed at the SEG laboratory facilities using a Tennelec Model CPVDS30-29195 for a SOLO CUP Analysis. The highest levels of activity were found to be from Cesium 137, Cobalt 60, and Uranium 235 and 238, and the total activity was found to be 7750.8870 disintegrations per minute per gram (dpm/g). A substantial amount of this was to be used in the container. Contaminated clean-up water resulting from daily SEG cask maintenance and cleaning operations was obtained for use as the concrete mix water and contaminating liquid. The water used was analyzed at the SEG laboratory facilities. The analysis of the clean-up water shows an actively level resulting primarily from one radionuclide, Cobalt 60, and the total activity was found to be 0.222 dpm/g. Crushed recycled concrete was obtained for this project by demolishing a portion of the lid of an earlier manufactured concrete container. The concrete was crushed by hand into pellet sizes no larger than 19 mm in diameter. The recycled concrete was washed in the above described contaminated water for fine partial removal and therefore is deemed to have the same radiological properties, resulting primarily from Cobalt 60. Stainless steel reinforcing fibers were obtained. These fibers are commercially available under the trade name MelTEC Stainless Steel Fibers. The stainless steel fibers were rinsed in the above described contaminated water to aid in the adherence of the stainless steel fibers to the cement. Due to the use of contaminated water the stainless steel fibers are considered to have the same radiological characteristics, resulting primarily from Cobalt 60. The fibers were all approximately 15 mm to 16 mm in length. While the water, crushed concrete and steel fibers were not strictly radioactive as defined previously, they did have activity levels above natural activity levels and were certainly tainted with radioactive elements and were contaminated. Clean limestone based aggregate material conforming to AASHTO number 67 stone gradation and properties was obtained. Clean river sand conforming to AASHTO requirements for concrete sand was also obtained. This sand material as well as the clean limestone based aggregate are generally the same as that used in making normal precast concrete. Type 1 Portland Cement was obtained. The cement was as commercially available for sale for all construction activities. Standard concrete quality fly ash was also obtained from the SEG solidification operations. A summary of the materials used in both containers is shown below in Table 1: TABLE 1 ______________________________________ PARTS BY PARTICLE MATERIAL WT WT % SIZE (mm) ______________________________________ Cement 60 14.8 0.0015-0.01 Sand 110 27.1 0.1-1.0 Aggregate 140 34.6 0.1-10 Contaminated 15 2.7 15.87 Stainless Fibers Contaminated 25 6.2 0.1-20 Fractured Concrete Radioactive Slag 25 6.2 1-30 Contaminated Water 30 7.4 ______________________________________ Prior to any mixing or handling of materials, all workers were dressed in protective clothing and had passed the SEG Radiation Workers Safety Course. In addition to written instructions oral directions were also given for the mixing of the concrete batch. The composition steps were as follows, where all components were added by weight measurements in the following order with complete mixing: 1) Aggregate 2) Radioactive Slag 3) Contaminated Fractured Concrete 4) Sand 5) Cement 6) Contaminated Water 7) Contaminated 15.87 mm (5/8 inch) Stainless Steel Fibers 8) Plasticizers (0.2 wt. %) The mixing procedure was to first add the aggregate (including slag and recycled concrete), sand, cement and water, mixing thoroughly as the ingredients were added. The water was added at a rate to maintain an approximate four inch slump according to ASTM standards. Once the concrete mix was completed, the steel fibers were added slowly to the mix ensuring even distribution. Care was taken to add water to maintain a workable slump of at least 4 inches. With the concrete thoroughly mixed and all ingredients added in proper proportions the concrete was ready for pouring. The concrete was thoroughly mixed, ensuring even distribution of all materials including radioactive constituents. The concrete was introduced to the forms by hand and rolled into its final position. The two containers were cured for 2 days in the forms under plastic sheeting. The freshly poured containers were sprayed with a service water mist and wrapped in plastic and allowed to mist cure according to ASTM standard specifications for three days. After three days of curing the containers were unwrapped and the wood and plastic forms removed. All residual material resulting from the forming process was cleaned from the containers using hand tools. The completed containers were inspected and then sprayed with a water mist, repackaged and allowed to cure for the remaining 25 days to reach full design strength. Following the 28 day curing period the containers were removed from their wrapping and cosmetic repairs for surface blemishes and minor honeycombing, caused by form bleed, were completed. The tops were fitted to the containers and they were then surveyed by SEG Health Physics personnel for radiological activity. The results of all the testings on the containers is shown in Table 2 below: TABLE 2 ______________________________________ Sample Containers 1 and 2 Radioactivity Levels ______________________________________ Water 3.76 .times. 10.sup.-7 .mu.Ci/ml* = 0.222 dpm/g Fibers 3.76 .times. 10.sup.-7 .mu.Ci/ml* = 0.222 dpm/g Fractured Concrete 3.76 .times. 10.sup.-7 .mu.Ci/ml* = 0.222 dpm/g Slag 7.75 .times. 10.sup.3 dpm/gr** = 2.5 .times. 10.sup.-3 .mu.Ci/ml ______________________________________ 0.0139 Bq/g *129.5 Bq/g The use of 6.2 wt % highly radioactive slag provided substantial radioactive material content in the containers. Final density of the concrete containers was over 90% of theoretical. Similar results would be achieved with commercial sized containers, for example, containers having internal dimensions of 182 cm length.times.122 cm width and 122 cm height, with a lid weighing 318 kg and a container weight of 1,730 kg; able to contain up to about 1375 kg of contaminated material within the container structure itself and an internal payload of about 2,270 kg of contaminated material. Thus, utilization of this invention could increase total contaminated payload by about 60% compared to a container made of uncontaminated material. The two sample concrete containers were then delivered for use as waste process, transport, storage and disposal containers at the SEG facility. The small size and uniqueness of the prototypes limited the uses of the containers. Sample container 1 was lined with a polypropelene liner similar to material commercially available for material transport buckets. The exterior surfaces of container 1 were coated with a commercially available concrete penetrant and sealer. With the liner tightly in place and the coating dried, radiological waste material from the SEG facility was placed in the container. The lid of the container was sealed and the outside of the container was surveyed for surface activity. A level of activity below that of background was found and recorded. The container was then transported for temporary storage at the SEG facility. Sample container 2 was not lined or coated as was Sample container 1. Instead, Sample container 2 had a standard, radioactive waste container placed inside the body of the sample container. The standard container held waste material deposited and processed to SEG standards. The material in the sample container was then wrapped, sealed and the lid placed on the sample container. The entire sample container was then surveyed for surface activity and was found to be below 0.5 milirem. This sealed sample container was then permanently sealed, packaged and legally transported by SEG from Oak Ridge, Tenn. to a waste disposal site in Barnwell, S.C. As can be seen contaminated materials can be used, or reused, in a wide variety of articles, many of which are, or can be, applied to purposes in which they are again, or further, contaminated.
	description	The present invention relates generally to an apparatus and a method for ultraviolet light and, more specifically, to lamp systems and methods for treating or curing materials on a substrate or product by application of ultraviolet light. Ultraviolet lamp systems, such as those used in the heating or curing of adhesives, sealants, inks or other coatings for example, are designed for coupling microwave energy to an electrodeless lamp, such as an ultraviolet (UV) plasma lamp bulb mounted within a microwave chamber of the lamp system. In ultraviolet lamp heating and curing applications, one or more magnetrons are typically provided in the lamp system to couple microwave radiation to the plasma lamp bulb within the microwave chamber. The magnetrons are coupled to the microwave chamber through waveguides that include output ports connected to an upper end of the chamber. When the plasma lamp bulb is sufficiently excited by the microwave energy, it emits ultraviolet radiation in one direction through an open lamp face of the lamp system to irradiate a substrate which is located generally near the open lamp face. A source of pressurized air is fluidly connected to a housing of the lamp system which contains the magnetrons, the microwave chamber and the plasma lamp bulb. The source of pressurized air is operable to direct cooling air, such as 350 CFM of cooling air for example, through the housing and into the microwave chamber to properly cool the magnetrons and the plasma lamp bulb during irradiation of the substrate by the lamp system. In some UV heating and curing applications, the lamp system includes a mesh screen mounted at the open lamp face which is transmissive to ultraviolet radiation but is opaque to microwaves. The configuration of the mesh screen also permits the significant air flow of cooling air to pass therethrough and toward the substrate. In some applications, however, the substrate may require a clean environment, such as in a curing chamber, so that the substrate will not be contaminated during the heating and curing process by contaminants carried by the cooling air in contact with the substrate. Moreover, the substrate may also be somewhat delicate and therefore susceptible to damage in harsh environments, such as under the influence of the significant air flow of the cooling air which impinges upon and possibly disturbs the substrate. Oftentimes, the substrate may also be adversely affected by excessive heat which may be generated by the plasma lamp bulb during the irradiation process. Still further, the substrate may configured about the perimeter of a material, thus requiring multiple and different application of UV light to irradiate the substrate around the product. Embodiments of the present invention address these and other problems associated with the prior art by providing apparatus for generating ultraviolet light and methods for operating an ultraviolet light source. In one embodiment, the apparatus comprises a chamber enclosing an interior space, a light source including a lamp head coupled to the chamber, and an ultraviolet (UV) transmissive member. The lamp head includes a lamp face configured to transmit the ultraviolet light and cooling air provided to the apparatus. The ultraviolet (UV) transmissive member is positioned above the lamp face and below the interior space to define a plenum therebetween. The UV transmissive member is configured to transmit the ultraviolet light into the interior space and to divert the cooling air from entering the interior space. An exhaust system, which is coupled in fluid communication with the plenum, is configured to exhaust the cooling air from the plenum. In another embodiment, the method comprises emitting ultraviolet light from the lamp head, cooling the lamp head with cooling air, transmitting the ultraviolet light through a UV transmissive member in to an interior space of a chamber, and diverting the cooling air away from the interior space of the chamber. These and other advantages will be apparent in light of the following figures and detailed description. It should be understood that the appended drawings are not necessarily to scale, presenting a somewhat simplified representation of various features illustrative of the basic principles of embodiments of the invention. The specific design features of embodiments of the invention as disclosed herein, including, for example, specific dimensions, orientations, locations, and shapes of various illustrated components, as well as specific sequences of operations (e.g., including concurrent and/or sequential operations), will be determined in part by the particular intended application and use environment. Certain features of the illustrated embodiments may have been enlarged or distorted relative to others to facilitate visualization and clear understanding. Turning now to the drawings, wherein like numbers denote like parts throughout the several views, FIGS. 1-3 show a microwave-excited UV lamp system or light source 10 consistent with embodiments of the invention. Light source 10 includes a pair of microwave generators, illustrated as a pair of magnetrons 14, that are each coupled to a longitudinally extending microwave chamber 16 through a respective waveguide 18. Each waveguide 18 has an outlet port 20 coupled to a lower end of the microwave chamber 16 so that microwave energy generated by the pair of microwave generators 14 are coupled to the microwave chamber 16 in spaced longitudinal relationship adjacent opposite lower ends of the microwave chamber 16. An electrodeless plasma lamp 22, in the form of a sealed, longitudinally extending plasma lamp bulb, is mounted within the microwave chamber 16 and supported adjacent the upper end of the microwave chamber 16 as is well known in the art. The light source 10 includes a housing 24 which is connected in fluid communication with a source of pressurized air (not shown in FIGS. 1-3) in one embodiment through an air inlet duct 26 as is well known in the art. The air inlet duct 26 is located at a lower end of the housing 24 and the upper end of the housing 24 forms a lamp head 28. The source of pressurized air is operable to direct a flow of cooling air, represented diagrammatically by arrows 30 in FIGS. 1 and 2, through the housing 24 and into the microwave chamber 16 to cool the magnetrons 14 and plasma lamp bulb 22 as will be described in greater detail below. The flow of cooling air 30 passes through the microwave chamber 16 and is emitted or transmitted through openings in an open lamp face 32 of the lamp head 28. Light source 10 is designed and constructed to emit ultraviolet light, illustrated diagrammatically by arrows 34 in FIG. 2, through the open lamp face 32 of the light source 10 upon sufficient excitation of the plasma lamp bulb 22 by microwave energy coupled to the microwave chamber 16 from the pair of microwave generators 14. While a pair of magnetrons 14 are illustrated and described herein, it is to be understood that the light source 10 may include only a single magnetron 14 to excite the plasma lamp bulb 22 consistent with embodiments of the invention. As shown in FIG. 3, light source 10 includes a starter bulb 36 and a pair of transformers 38 (one shown) that are each electrically coupled to a respective one of the magnetrons 14 to energize filaments of the magnetrons 14 as understood by those skilled in the art. The magnetrons 14 are mounted to inlet ports 40 of the waveguides 18 so that microwaves generated by the magnetrons 14 are discharged into the microwave chamber 16 through the longitudinally spaced apart outlet ports 20 of the waveguides 18. Preferably, the frequencies of the two magnetrons 14 are split or offset by a small amount to prevent intercoupling between them during operation of the light source 10. In specific embodiments, a first magnetron 14 may produce a signal of about 2.4 GHz, while a second magnetron 14 produces a signal that has a difference up to about 20 MHz from the first magnetron 14. A longitudinally extending lamp reflector 42 is mounted within the microwave chamber 16 for reflecting the ultraviolet light 34 emitted from the plasma lamp bulb 22 toward a substrate (not shown) which is located outside the open lamp face 32 of the lamp head 28. The lamp reflector 42 may have an elliptical configuration in transverse cross-section, although parabolic or other cross-sectional configurations are possible without departing from the spirit and scope of the present invention. Alternatively, the lamp reflector 42 may be configured as a flood reflector as is known in the art (as opposed to a focus reflector, also known in the art). Ultraviolet light, as used herein, is radiation having a wavelength (or frequency) within the ultraviolet band of the electromagnetic spectrum. The lamp reflector 42 is made of coated glass. For example, one side of the lamp reflector 42 (e.g., the plasma lamp bulb side) includes a dichroic coating while the other side of the lamp reflector 42 may be sandblasted. Thus, the lamp reflector 42 is substantially transparent to the microwave energy generated by the pair of magnetrons 12 but substantially opaque to, and substantially reflective of, the ultraviolet light 34 emitted by the plasma lamp bulb 22. Alternatively, the lamp reflector 42 may be made of another material having suitable reflective, refractive, and/or thermal properties, such as polished aluminum, which is also substantially transparent to the microwave energy generated by a magnetron 14 but substantially opaque to, and substantially reflective of, the ultraviolet light 34 emitted by the plasma lamp bulb 22. As shown in FIG. 2, lamp reflector 42 includes a pair of longitudinally extending reflector panels 44 that are mounted in opposing, i.e., mirror facing relationship within the microwave chamber 16 and in spaced relationship to the plasma lamp bulb 22. Each reflector panel 44 is preferably made of coated glass, although other materials having suitable reflective and thermal properties are possible as well. When made of coated glass, for example, each reflector panel 44 is transparent to the microwave energy generated by the pair of magnetrons 14 but opaque to and reflective of the ultraviolet light 34 emitted by the plasma lamp bulb 22. Further referring to FIG. 2, a longitudinally extending intermediate member 46 is mounted within the microwave chamber 16 in spaced relationship to the reflector panels 44, and also in spaced relationship to the plasma lamp bulb 22. The intermediate member 46 may be made of glass, such as PYREX®, and may be uncoated to be non-reflective of the ultraviolet light 34 emitted by the plasma lamp bulb 22. When the pair of reflector panels 44 and the intermediate member 46 are mounted within the microwave chamber 16 to form the lamp reflector 42, a pair of spaced, longitudinally extending slots 48 are formed between the reflector panels 44 and the intermediate member 46. The pair of spaced, longitudinally extending slots 48 are operable to pass cooling air 30 from the pressurized air source toward the plasma lamp bulb 22 so that the cooling air 30 substantially envelops the outer surface of the plasma lamp bulb 22 to cool the bulb 22. Details of the construction of the lamp reflector 42 are fully described in commonly owned and co-pending U.S. Pat. No. 6,696,801, entitled “Microwave Excited Ultraviolet Lamp System With Improved Cooling,” the disclosure of which is hereby incorporated by reference herein in its entirety. Alternatively, other reflector configurations are possible as well as will be readily understood by those of ordinary skill in the art. The cooling air 30 passes through the microwave chamber 16 and is emitted through the open lamp face 32 of the lamp head 28. As illustrated in FIGS. 2 and 3, the light source 10 is further configured with mounting flanges 50 having threaded apertures 52 that may be utilized to mount a curing chamber and/or exhaust system to the light source 10. With respect to the cooling air 30, a sufficient flow rate, such as about 350 CFM for example, is provided within the housing 24 to insure proper operation of the magnetrons 14 and the plasma lamp bulb 22. To insure that a sufficient cooling air flow rate is being provided during operation of the light source 10, a differential pressure transducer 56 is mounted in fluid communication with the lamp head 28 and the housing 24. The differential pressure transducer 56 senses the pressure drop in the housing 24 and provides a signal to a controller of the light source 10 (not shown) to shut down the light source 10, adjust the flow of energy to the magnetrons 14 (thereby adjusting the intensity of the UV light from the plasma lamp bulb 22), and/or adjust the flow of air in the light source 10 when the desired pressure is not sensed. With reference to FIGS. 4-5, a lamp system includes a source for pressurized air 60 and/or a curing chamber 70 are mounted to the light source 10. The pressurized air source 60 may collect ambient air, filter that air, and cool the air. The pressurized air source 60 may then provide that collected, filtered, and cooled air as cooling air 30 to the light source 10 through the air inlet duct 26. The curing chamber 70, in conjunction with the light source 10, is configured to provide UV light to substantially all of the perimeter of a substrate (not shown) as well as isolate the substrate from the cooling air 30 emitted through the open lamp face 32 of the lamp head 28. Referring to FIGS. 4-9, the curing chamber 70 is secured to the light source 10 utilizing a plurality of threaded bolts 71 that are threadingly engaged with respective apertures 52 of the mounting flanges 50. The curing chamber 70 includes an interior space that is at least partially defined by a lower housing 72 and an upper housing 74 connected by a hinge 76. The lower housing 72 is secured to the upper housing 72 by a latch assembly 78 (e.g., a latch 78a on the upper housing 74 that is secured by a securing mechanism 78b on the lower housing 72) when closed, but secured to the light source 10 via the aforementioned threaded bolts 71. The curing chamber 70 includes a top panel 79 but does not include a bottom panel. Rather, the bottom of the curing chamber 70 is open to the cooling air 30 and ultraviolet light 34 emitted by the light source 10. Two light shrouds 80 are configured on the sides of the curing chamber 70. Specifically, the interiors of the light shrouds 80 are in fluid communication with the interior space of the curing chamber 70 through an inlet 81a and an outlet 81b, although the inlet 81a and outlet 81b may be swapped. The light shrouds 80 are configured to prevent substantial amounts of UV light from escaping the curing chamber 70. Each light shroud 80 also includes a lower shroud housing 82 and an upper shroud housing 84 as well as respective end caps 86 (e.g., end cap 86a on the upper shroud housing 84 and end cap 86b on the lower shroud housing 82). The end caps 86 may be removed to allow a substrate (e.g., an object such as a rope, bar, or pipe) carrying the UV curable material with to be introduced or fed into one light shroud 80, through the curing chamber 70, and out another light shroud 80. Alternatively, the end caps 86 may be machined to include an aperture 89 extending through the ends caps 86 that is substantially similar to the cross-section of the substrate (e.g., the aperture may have a similar cross-section corresponding to the aforementioned rope, bar, or pipe) such that the substrate and UV curable material can be introduced or fed through one light shroud 80, through the curing chamber 70, and out another light shroud 80. As illustrated throughout the figures, each light shroud 80 is substantially hollow such that a guide mechanism (not shown) may be positioned in the bottom of the light shroud to support and/or guide a product through the curing chamber 70. As best shown in FIGS. 6 and 7, the curing chamber 70 includes an ultraviolet light (UV) transmissive member 88 held in place by a mounting bracket 90. The UV transmissive member 88 may be comprised of quartz. The mounting bracket 90, in turn, is secured to the front wall 92 and the back wall 94 of the curing chamber 70. The front wall 92 is not directly connected to the back wall 94, but is instead indirectly connected with the back wall 94. A curing chamber reflector 120 is positioned between the front wall 92 and the back wall 94. A generally air-tight seal is formed between the UV transmissive member 88 and the walls of the curing chamber 70. Such an air-tight seal ensures that the interior of the curing chamber 70 that is in fluid communication with the light shrouds 80, and therefore the interior of the curing chamber 70 in which the substrate is located, does not receive any cooling air 30 emitted from open lamp face 32 of the lamp head 28. A sealant or other gasket (not shown) may be in communication with the edges of the UV transmissive member 88, or the UV transmissive member 88 and the mounting bracket 90 or other walls of the curing chamber 70, to form the generally air-tight seal. At least one rubber gasket 96 may also be positioned between the light source 10 and the curing chamber 70 to create a generally air-tight seal therebetween. In some embodiments, the UV transmissive member 88 is a substantially rectangular member with a substantially flat cross-section and a perimeter that generally corresponds to the top-view cross section of the curing chamber 70 without substantially causing convergence or divergence of the ultraviolet light 34 transmitted therethrough. In alternative embodiments, the UV transmissive member 88 may have a cross section such that the ultraviolet light 34 transmitted through the UV transmissive member 88 is diverged or converged. An exhaust system 100 is in fluid communication with the open lamp face 32 of the lamp head 28. Specifically, the exhaust system 100 is configured with an air inlet port 102 to receive the cooling air 30 emitted through the open lamp face 32 into a plenum formed by the walls of the curing chamber 70 and the UV transmissive member 88. The exhaust system 100 receives the cooling air 30 through a hole 103 extending through the front wall 92 of the curing chamber 70 that is positioned below the UV transmissive member 88. An exhaust duct 104 is configured to direct the cooling air 30 to a location remote from the lamp head 28 so that the cooling air 30 does not contact the substrate. In one embodiment, the exhaust system 100 is an exhaust duct that in turn includes a top wall 108, an opposite bottom wall 110, a pair of opposite side walls 112 and an end wall 114 which are configured to form an elongated and enclosed plenum 116. The enclosed plenum 116 is in fluid communication with the open lamp face 32 of the lamp head 28. The exhaust system 100 is attached to the front wall 92 of the curing chamber 70 through at least one screw 118. The curing chamber 70 is further configured with at least one curing chamber reflector 120 configured on interior of the upper housing 74 as well as at least one side reflector 122. In some embodiments, at least one side reflector 122 is configured on each of the front wall 92 and back wall 94. In specific embodiments, and as illustrated in FIG. 6, two side reflectors 122 are configured on each of the front wall 92 and back wall 94: one side reflector 122a above the UV transmissive member 88 attached to the back wall 94, one side reflector 122b below the UV transmissive member 88 attached to the back wall 94, one side reflector 122c above the UV transmissive member 88 attached to the front wall 92, and one side reflector 122d below the UV transmissive member 88 attached to the front wall 92. Curing chamber reflector 120 and side reflectors 122a-d may have an elliptical configuration in transverse cross-section (e.g., as illustrated with side reflectors 122c and 122d), although parabolic or other cross-sectional configurations are possible without departing from embodiments of the invention. Specifically, and as illustrated in FIGS. 4-6, side reflectors 122a and 122b include a flat cross-section and the side reflectors 122c and 122d have a curvature and are not flat so that the side reflectors 122a-d have different cross-sectional profiles. Thus, utilizing the curing chamber reflector 120 and the side reflectors 122, the curing chamber 70 may be configured to provide ultraviolet light 34 to substrate positioned around the perimeter of a product. At least a portion of the lower housing 72 is configured to overlap the upper housing 74. Specifically, as illustrated in FIG. 8, at least a portion of the front wall 92 of the upper housing 74 is removed as at 130 while the lower housing 72 includes an upwardly directed projection 132 that overlaps the removed portion 130. Similarly, as illustrated in FIG. 9, the portion of a light shroud 80 connected to the upper housing 74 is also removed as at 140 while the portion of the light shroud 80 connected to the lower housing 72 also includes an upwardly directed projection 142 that overlaps the removed portion 140. As such, ultraviolet light that may escape from the connection between the lower housing 72 and upper housing 74, as well as light that may escape from the connection where the two halves of the light shrouds 80 meet, is prevented. Thus, embodiments of the invention provide a lamp system that includes a curing chamber 70 attachment for providing an enhanced dual cure. Specifically, the curing chamber 70 is operable to provide a full product cure (sometimes referred to as a “3-Dimensional” or “360° cure”). The curing chamber 70 can also be configured to maximize the intensity of ultraviolet light on the substrate geometry by adjusting the angles of the side reflectors 122 and thus the ultraviolet light reflected onto the substrate. Thus, the use of multiple plasma lamp bulbs 22, multiple lamp heads 28, and/or multiple and different applications of ultraviolet light for a substrate are prevented. Additionally, the curing chamber 70 provides an interior space that is warmed to about 300° F. during operation, and thus not only provides a full product ultraviolet light cure but also provides a thermal cure for an enhanced dual cure. With respect to the curing chamber 70, the overlapping design of the edges of the lower housing 72 and the upper housing 74, as well as the overlapping design of the edges of the light shrouds 80, prevents substantial ultraviolet light loss and transmission to the operating environment. The enhanced dual cure is also provided without contaminating the substrate and/or the product (e.g., a UV curable material) while also providing a stable platform for the product. Specifically, the cooling air 30 from the light source 10 does not contaminate the substrate or product, nor does the cooling air 30 impart unwanted motion to the substrate or product on the substrate. The product, in turn, can be supported by the previously discussed guide mechanisms in the respective light shrouds 80. A person having ordinary skill in the art will recognize that the environments illustrated in FIGS. 1-9 are not intended to limit the scope of embodiments of the invention. In particular, the light source may include fewer or additional components consistent with alternative embodiments of the invention. Indeed, a person having skill in the art will recognize that other alternative hardware environments may be used without departing from the scope of the invention. For example, a person having ordinary skill in the art will appreciate that the light source 10 may be coupled with an actuable shutter assembly (not shown) to ensure that as little ultraviolet light 34 escapes from the light source 10 as possible when that actuable shutter assembly is closed. Such a shutter assembly is disclosed in U.S. Pat. No. 6,933,683 entitled “Microwave Powered Lamphead Having External Shutter,” the disclosure of which is incorporated by reference herein in its entirety. Furthermore, while embodiments of the invention has been illustrated by a description of the various embodiments and the examples, and while these embodiments have been described in considerable detail, it is not the intention of the applicants to restrict or in any way limit the scope of the appended claims to such detail. Additional advantages and modifications will readily appear to those skilled in the art. Thus, the invention in its broader aspects is therefore not limited to the specific details, apparatuses, and/or methods shown and described. Accordingly, departures may be made from such details without departing from the spirit or scope of applicants' general inventive concept.
049873130	abstract	A method of and apparatus for the treatment and storage of nuclear power plant wastes utilizes a jacket heating a cast iron storage container directly to vaporize liquid radioactive waste under suction applied to the container with interposition of a baffle preventing entrainment of droplets from the container.
061880736	claims	1. A radiographic intensifying screen comprising a support, a phosphor layer containing phosphor and a surface protective layer overlaid in order, wherein the surface protective layer shows a scattering length in the range of 5 to 80 .mu.m, said scattering length being measured at main wavelength of light emitted from the phosphor. 2. The radiographic intensifying screen of claim 1, wherein the scattering length is in the range of 10 to 70 .mu.m. 3. The radiographic intensifying screen of claim 1, wherein the surface protective layer contains light-scattering fine particles having a mean grain size of 0.1 to 1 .mu.m and a refractive index of not less than 1.6. 4. The radiographic intensifying screen of claim 1, wherein the surface protective layer contains light-scattering fine particles having a mean grain size of 0.1 to 1 .mu.m and the refractive index of not less than 1.9. 5. The radiographic intensifying screen of claim 1, wherein the surface protective layer contains light-scattering fine particles comprising at least one material selected from the group consisting of zinc oxide, zinc sulfide, titanium dioxide and lead carbonate, and the particles have a mean grain size of 0.1 to 1 .mu.m. 6. The radiographic intensifying screen of claim 1, wherein the surface protective layer comprises a binder containing fluorocarbon resin and light-scattering fine particles dispersed therein. 7. The radiographic intensifying screen of claim 1, wherein the surface protective layer comprises a binder containing polyester resin and light-scattering fine particles dispersed therein. 8. The radiographic intensifying screen of claim 1, wherein the surface protective layer has a thickness of 2 to 12 .mu.m. 9. The radiographic intensifying screen of claim 1, wherein the surface protective layer has a thickness of 3 to 9 .mu.m. 10. The radiographic intensifying screen of claim 1, wherein the phosphor contained in the phosphor layer has the following formula: EQU M.sub.2 O.sub.2 X:Tb 11. The radiographic intensifying screen of claim 1, wherein a light-reflecting layer is provided between the support and the phosphor layer. 12. The radiographic intensifying screen of claim 1, wherein the phosphor layer comprises a binder and the phosphor dispersed therein, and the weight ratio of the binder to the phosphor is in the range of 1/12 to 1/200. 13. The radiographic intensifying screen of claim 1, wherein the phosphor layer comprises a binder and the phosphor dispersed therein, and the weight ratio of the binder to the phosphor is in the range of 1/16 to 1/100.
	summary
047028823	summary	BACKGROUND OF THE INVENTION Cross Reference to Related Applications Reference is hereby made to the following copending applications directed to related subject matter and commonly assigned with the present application: 1. Nuclear Reactor Fuel Assembly With A Removable Top Nozzle, Ser. No. 537,775, filed Sept. 30, 1983; PA0 2. Removable Top Nozzle And Tool For A Nuclear Reactor Fuel Assembly, Ser. No. 564,058, filed Dec. 21, 1983; PA0 3. Improved Method And Apparatus For Removably Mounting A Top Nozzle On A Nuclear Reactor Fuel Assembly, Ser. No. 644,756, filed Aug. 27, 1984; and PA0 4. Nuclear Reactor Fuel Assembly with a Removable Top Nozzle, Ser. No. 644,758, filed Aug. 27, 1984. FIELD OF THE INVENTION The present invention relates generally to a nuclear reactor fuel assembly top nozzle design, and more specifically to a nuclear reactor fuel assembly top nozzle design which simplifies access to the fuel assembly rod bundle and which transmits tensile loads from the top nozzle via grid skirt extensions of a top spacer grid. DESCRIPTION OF THE PRIOR ART In most nuclear reactors the core portion comprises a large number of fuel assemblies which include elongated fuel elements or rods grouped together in bundles and supported by a skeletal framework. The fuel assemblies are generally elongated structures which are supported by and aligned with upper and lower transversely extending core support plates. As a part of the skeletal structure, conventional fuel assemblies generally include a plurality of control rod guide tubes or thimbles held in an organized array by spacer grids axially spaced along and attached to the control rod guide thimbles. The top and bottom nozzles are secured to the control rod guide thimbles on opposite ends thereof, thereby forming an integral fuel assembly. The top and bottom nozzles respectively extend slightly above and below the ends of the fuel rods, capturing the rods therebetween. An example of a reactor core having a fuel assembly of this convention design is described in U.S. Pat. No. 3,235,463 to Sankovich. If, during operation of a fuel assembly in a nuclear reactor, one or more fuel rods develop cracks or other flaws, they must be removed from the reactor. Since, in conventional fuel assemblies, the fuel rods are part of an integral assembly where the guide thimbles are semi-permanently attached to the top and bottom nozzles, it is difficult to detect and remove any failed or defective rods. To gain access to such rods, it is typically necessary to remove the affected fuel assembly from the reactor core and then break the attachments which secure the nozzles to the control rod guide thimbles. In doing so, the fuel assembly is often rendered unfit for further use in a reactor because of the damage done to both the guide thimbles and to the nozzles. In view of the high costs associated with replacing fuel assemblies, efforts have been made to develop reconstitutable fuel assemblies in order to minimize operating and maintenance expenses. Several different types of reconstitutable fuel assemblies have been devised, the majority of which employ threaded arrangements for attaching the top nozzle to the control rod guide thimbles whereby the top nozzle can be removed to gain access to the fuel rods contained in the skeletal assembly. A type of reconstitutable fuel assembly is described by Klaumb et al. in U.S. Pat. Nos. 3,770,583 and 3,814,667 which illustrate a top nozzle of the type having a hold-down device which essentially comprises coil springs coaxially disposed about upwardly extending alignment posts that are threaded through an end plate with fastener nuts located on the underside of the plate. An upper hold-down plate is slidably mounted on the alignment posts and the coil springs are interposed, in compression, between the hold-down plate and the end plate. A radially enlarged shoulder on the upper end of the alignment posts retain the hold-down plate on the posts. In an attempt to improve upon the Klaumb et al. device, Anthony et al. set forth a different threaded joint arrangement as illustrated in U.S. Pat. No. 3,992,259. Another reconstitutable fuel assembly contruction is illustrated and described in U.S. Pat. No. 3,828,868 to Jabsen which employs a threaded arrangement for attaching the top nozzle to the control rod guide thimbles so that the top nozzle can be removed to gain access to the fuel rods. The top nozzle removal method described essentially comprises unscrewing a multitude of threaded connections to release the top nozzle from the guide thimbles and then pulling the top nozzle over the external threads during removal. The sequence is reversed during top nozzle replacement. The reconstitutable fuel assemblies which employ such threaded arrangements not only suffer from high manufacturing costs, complicated designs, difficulty in removing and reattaching the top nozzle, but, in addition, it has been found that after irradiation, the threaded connection may become corroded and inoperable, thus requiring destructive operations to the nozzle and/or the guide thimbles in order to remove the top nozzle which results in costly damage and generally renders the fuel assembly unfit for further use. Further, repeated removal and replacement of the top nozzle from and to the threaded portions of the guide thimbles as well as retightening the threaded connections increases the risk of damaging the threads. Any thread seizure or damage incurred during removal and replacement can ruin the fuel assembly skeleton and require that a new fuel assembly be substituted for it. Further, conventional reconstitutable fuel assembly designs utilize guide thimbles to support both tensive and compressive loads on the fuel assembly as well as for providing guidance for control elements. It is these multiple functions which give rise to arrangements which require complex and intricate components including numerous loose parts designed to permit the top nozzle to be disconnected from and reconnected to the fuel assembly. An additional nonwelded top nozzle attachment arrangement is illustrated and described in U.K. Pat. No. 1,228,610. However, this fuel assembly is equally difficult to disassemble since the upper ends of the guide thimbles are fixed in holes formed in the top nozzle by being expanded after insertion in order to conform to the shape of the holes and annular grooves therein. SUMMARY OF THE INVENTION It is therefore an object of the present invention to provide a nuclear fuel assembly having an easily detachable and replaceable top nozzle. It is a further object of the present invention to provide a top nozzle design which permits quick and simple access to fuel rods of a fuel assembly without introducing loose parts to the fuel assembly. It is a further object of the present invention to provide a fuel assembly design wherein tensive loads on the top nozzle are transmitted to the fuel assembly via a top spacer grid. It is a still further object of the present invention to provide a fuel assembly design wherein no fastening of guide thimbles to the top nozzle is required. In order to accomplish these as well as other objects, the present invention comprises a fuel assembly having a longitudinally extending control rod guide thimble and a transversely extending top spacer grid. The fuel assembly further comprises a top nozzle and an attaching structure for removably mounting the top nozzle on the fuel assembly. The attaching structure comprises a first means for transferring compressive loads on the fuel assembly through the top nozzle and directly to the control rod guide thimble; and a second means for transferring tensive loads on the fuel assembly through the top nozzle and the transversely extending top spacer grid to the guide thimble. Preferably, the top nozzle includes a transversely extending adapter plate having an aperture therein. An upper end of the guide thimble is sized to clearance fit in the adapter plate aperture. A load collar is affixed to the the guide thimble adjacent the upper end thereof for transferring compressive loads through the adapter plate to the guide thimble. Further, the top spacer grid is preferably provided with an upstanding skirt extension and a latch for securing the skirt extension to the top nozzle. The invention is further directed to a method for assembling and disassembling a top nozzle of a fuel assembly having a guide thimble with an upper end, a top spacer grid, and a plurality of upstanding fuel rods. According to the invention, the method comprises the steps of providing means on the guide thimble for supporting compressive loads on the fuel assembly and providing means for resilently latching the top spacer grid to the top nozzle. In order to assemble the top nozzle on the fuel assembly, it is lowered onto the means for supporting compressive loads and then latched to the top spacer grid with the resilent latching means. Preferably, a load collar attached to the upper end of the guide thimble is used to support the compressive load and the top nozzle includes an adapter plate having an aperture sized for clearance fitting the upper end of the guide thimble. The top nozzle is lowered with the aperture aligned with the upper end of the guide thimble until the adapter plate bears on the load collar. Advantageously, the top spacer grid includes a grid skirt extension having an aperture therein and the top nozzle includes a resilently biased locking pin. During latching the locking pin is held in a deflected position until the grid skirt extension aperture passes into alignment with the locking pin. Thereupon, the locking pin is released into the grid skirt extension aperture. In order to remove the top nozzle, the locking pin is withdrawn from the grid skirt extension aperture and the top nozzle is lifted from the fuel assembly to thereby access the plurality of upstanding fuel rods. Additional objects, advantages and novel features of the invention are set forth in the description which follows and will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages may be realized and attained by means of the invention particularly pointed out in the appended claims.
	claims	1. A method for controlling an axial power distribution in a nuclear reactor having a core and a control rod, comprising:(a) calculating an axial offset of a current power distribution (AOP) using Expression (1), an axial offset of a power distribution capable of providing a current xenon concentration distribution under an equilibrium condition (AOX) using Expression (2), and an axial offset of a power distribution capable of providing a current iodine concentration distribution under an equilibrium condition (AOI) using Expression (3),AOP=(PT−PB)/(PT+PB) Expression (1).AOX=(PTX−PBX)/(PTX+PBX) Expression (2)AOI=(PTI−PBI)/(PTI+FBI) Expression (3)WhereinPT is a relative power in an upper half of the core,PB is a relative power in an lower half of the core,PTX is a relative power in the upper half of the core capable of providing a current xenon concentration under an equilibrium condition,PBX is a relative power in the lower half of the core capable of providing a current xenon concentration under an equilibrium condition,PTI is a relative power in the upper half of the core capable of providing a current iodine concentration under an equilibrium condition, andPBI is a relative power in the lower half of the core capable of providing a current iodine concentration under an equilibrium condition;(b) calculating a parameter DAOPX (=AOP−AOX) and a parameter DAOIX (=AOI−AOX);(c) displaying a trajectory having a plot point which plots the parameter DAOPX on X axis and the parameter DAOIX on Y axis, whereby the plot point forms an ellipse having a major axis;(d) determining if the axial offset of the current power distribution (AOP) exceeds an allowable range;(e) setting off alarm when the axial offset of the current power distribution (AOP) exceeds the allowable range and the plot point has deviated from the major axis; and(f) moving the control rod after the alarm is set off so that the plot point is guided back to the major axis, whereby the axial power distribution is controlled, and whereby the xenon oscillation is controlled at the same time. 2. A nontransitory computer readable medium having a computer program, which when executed by a computer, causes the computer to perform the method according to claim 1. 3. The method according to claim 1, further comprising in step (f) calculating a distance on the X axis from the deviated plot point to the major axis, calculating a movement volume of the control rod based on the distance, and giving an operation signal of inserting the control rod by the movement volume to a control rod driving mechanism. 4. The method of claim 1, wherein the ellipse has an origin on the major axis, and wherein plot point is guided to the major axis excluding the origin. 5. A system for controlling an axial power distribution in a nuclear reactor having a core and a control rod, comprising:(a) an axial offset calculation unit that calculates an axial offset of a current power distribution (AOP) using Expression (1), an axial offset of a power distribution capable of providing a current xenon distribution under an equilibrium condition (AOX) using Expression (2), and an axial offset of a power distribution capable of providing a current iodine distribution under an equilibrium condition (AOI) using Expression (3),AOP=(PT−PB)/(PT+PB) Expression (1)AOX=(PTX−PBX)/(PTX+PBX) Expression (2)AOI=(PTI−PBI)/(PTI+PBI) Expression (3)WhereinPT is a relative power in an upper half of the core,PB is a relative power in an lower half of the core,PTX is a relative power in the upper half of the core capable of providing a current xenon concentration under an equilibrium condition,PBX is a relative power in the lower half of the core capable of providing a current xenon concentration under an equilibrium condition,PTI is a relative power in the upper half of the core capable of providing a current iodine concentration under an equilibrium condition, andPBI is a relative power in the lower half of the core capable of providing a current iodine concentration under an equilibrium condition;(b) a parameter calculation unit that calculates a parameter DAOPX (=AOP−AOX) and a parameter DAOIX (=AOI−AOX);(c) a trajectory display unit that displays a trajectory having a plot point, wherein the plot point plots the parameter DAOPX on X axis and the parameter (DAOIX) on the Y axis, and forms an ellipse having a major axis;(d) an allowable range excess determination unit that determines if the axial offset of the current power distribution (AOP) exceeds an allowable range;(e) an alarming unit that sets off alarm when the axial offset of the current power distribution (AOP) exceeds the allowable range and the plot has deviated from the major axis; and(f) a control rod moving unit that moves the control rod after the alarm is set off so that the plot point is guided back to the major axis, whereby the axial power distribution is controlled, and whereby the xenon oscillation is controlled at the same time. 6. The system according to claim 5, characterized in that the control rod moving unit further calculates a distance on the X axis from the deviated plot point to the major axis, calculates a movement volume of the control rod based on the distance, and gives an operation signal of inserting the control rod by the movement volume to a control rod driving mechanism. 7. The system of claim 5, wherein the ellipse has an origin on the major axis, and wherein the plot point is guided to the major axis excluding the origin.
	abstract	Provided is an X-ray output apparatus including an X-ray output unit including a plurality of X-ray sources and configured to output parallel X-ray beams, a shield on which positions that are capable of blocking the output parallel X-ray beams and positions that are capable of transmitting the parallel X-ray beams are variable, and a control unit configured to control the output of the parallel X-ray beams in the X-ray output unit and the positions through which the parallel X-ray beams are transmitted in the shield.
043222676	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows an embodiment of this invention. Steam W.sub.M generated in the reactor 1 is normally fed to a main turbine, not shown, via a main steam conduit C.sub.M and isolation valves IV immediately inside and outside a container vessel PV. When a fault occurs inside or outside the reactor 1 and when the reactor 1 is scrammed by a mechanism, not shown, with the result that the isolation valves IV are closed, the decay heat Q.sub.R generated in the reactor core 25 is taken out or dissipated by an RHR (residual heat removal) system shown in FIG. 1. Schematically, the RHR system has such a constitution as follows: the steam generated in the reactor 1 is led to a heat exchanger 4 and condensed there into water 5 and the condensed water 5 is fed back into the reactor 1 by a pump 10 driven by a turbine 12 actuated by the reactor steam so as to cool the reactor core 25. A control valve 3 is controlled by a condensate flow control apparatus A, which is well known and easily obtainable. The function of this condensate flow control apparatus A is as follows: the reactor steam generated by the decay heat Q.sub.R is led through the control valve 3 to the heat exchanger 4 and condensed into water 5 at a rate of a flow W.sub.H by a cooling pipe assembly 6 in the heat exchanger 4, and the pressure P.sub.H in the heat exchanger 4 is controlled to the value (P.sub.H).sub.ref set by a reference pressure setting means 22 by means of a pressure detector 14, a pressure controller 15 and the regulation valve 3 so that the condensate flow W.sub.H from the heat exchanger 4 may be stabilized. Make-up water flow W.sub.c from a make-up water tank 8 is added to the condensate flow W.sub.H at the point J and the resultant flow, i.e. feedwater flow, W.sub.F is fed to the pump 10. A check valve 9 serves to prevent the return of the supplied make-up water to the make-up water tank 8. The aperture of a control valve 7 is controlled by a make-up water control apparatus C. The feed water W.sub.F is pressurized by the pump 10 and supplied into the reactor 1 through a check valve 11. The pump 10 is driven by the turbine 12 and the flow W.sub.TB of steam to drive the turbine 12 is controlled by a regulation valve 13 near the inlet of the turbine 12, which valve is in turn controlled by a feedwater flow control apparatus B. According to this invention, the problems concerning the lowering of the reactor water level and the temperature fall rates can be solved by the combination of the control apparatuses B and C described above. Namely, the RHR system as an embodiment of this invention operates as follows. In the case of scram, the reactor water level lowers since the steam W.sub.TB to drive the turbine 12 is exhausted out of the system. To compensate the exhausted steam, i.e. reactor water, the make-up water control apparatus C is used to change the pressure at the outlet of the make-up water tank 8 according to the change in the reactor water level so that a required amount of make-up water W.sub.c is supplied into the reactor 1 and also the condensate flow W.sub.H is decreased. On the other hand, the temperature fall rates in the reactor 1 vary owing to the supply of the make-up water and the change in the decay heat with time. Accordingly, the residual heat is removed safely during a shortest period of time by controlling the feedwater flow W.sub.F by the feedwater control apparatus B in such a manner that the maximum one of the temperature fall rates at several points in the reactor 1 always equals the preset desired value. First, the control of the reactor water level by the make-up water control apparatus will be explained. The level L.sub.R of the reactor water is detected by a water level detector 31. A water level controller 32 determines the desired value (P.sub.s).sub.ref for the pressure at the outlet of the make-up water tank 8 in accordance with the difference between the detected level L.sub.R and the preset value (L.sub.R).sub.ref set by a reference water level setting means 33. Then, the pressure P.sub.s at the outlet of the make-up water tank 8 is controlled to the desired value (P.sub.s).sub.ref by the combined operations of a pressure detector 16, a pressure controller 17 and the control valve 7. The relationships between the signals at points in the control apparatus C, the various flows and the level of the reactor water are as shown in FIG. 2. Now, reference should be had to FIG. 2. Suppose that the RHR system is started at an instant and that the level L.sub.R of the reactor water lowers as shown in diagram (a). The references value (L.sub.R).sub.ref set by the reference water level setting means 33 is made approximately equal to the reactor water level L.sub.R assumed before the instant t.sub.0 and the water level controller 32 receives an input (L.sub.R).sub.ref =L.sub.R and in turn delivers a constant desired value, i.e. reference pressure (P.sub.s).sub.ref before the instant t.sub.0. The pressure P.sub.s detected at the joining point J by the pressure detector 16 is slightly higher than a pressure P.sub.c to open the check valve 9 before t.sub.0 since the RHR system is not operating before t.sub.0. The pressure P.sub.s increases rapidly after the start of the operation of the RHR system but remains lower than (P.sub.s).sub.ref. The pressure controller 17 receives an input (P.sub.s).sub.ref -P.sub.s >0 and controls the aperature V of the control valve 7 to the fully opened state. Consequently, the detected pressure P.sub.s is higher than the pressure P.sub.c to open the check valve 9 and it therefore follows that W.sub.c =0 and W.sub.H =W.sub.F. When the level L.sub.R of the reactor water lowers at t.sub.0 and the condition that (L.sub.R).sub.ref -L.sub.R .gtoreq.0 is reached, the pressure setter 32 decreases the value (P.sub.s).sub.ref. When the condition that (P.sub.s).sub.ref -P.sub.s <0 is reached at t.sub.1, the pressure controller 17 starts to close the control valve 7. As the aperture of the valve 7 decreases, the pressure P.sub.s at the point J lowers. At an instant t.sub.2, the pressure P.sub.c to open the check valve 9 exceeds the pressure P.sub.s, that is, P.sub.c >P.sub.s so that make-up water starts to flow at a rate W.sub.c from the tank 8 via the check valve 9. The expression (3) given below shows the relationship between the flow W.sub.c and the pressure P.sub.s. ##EQU3## where k is the coefficient of friction loss of the make-up water pipe, P.sub.o the pressure in the make-up water tank 8, H.sub.c the difference between the potential heads at the surface of the make-up water in the tank 8 and at the position of the feedwater pipe, and .gamma..sub.w the specific weight of the make-up water, all these quantities being constant. Accordingly, W.sub.c varies depending only upon P.sub.s, as shown in FIG. 3. If the value under the radical sign in (3) is negative, as occurs in the case where P.sub.s exceeds P.sub.c, then the check valve 9 is completely closed so that W.sub.c =0. After t.sub.2, it is seen that W.sub.F =W.sub.c +W.sub.H. It should here be noted that W.sub.F remains constant before, at and after t.sub.2 since W.sub.F is controlled to a roughly constant value by the control apparatus B governing the operation of the pump 10, that is, since W.sub.H is decreased by an amount equal to the increment of W.sub.c. As a result of the above control that make-up water more than the exhausted water is supplied into the RHR system, the reactor water recovers its initial level (L.sub.R).sub.ref. The instant t at which the water level L.sub.R returns to the value at t.sub.0, is theoretically equal to the instant, i.e. t.sub.3 in FIG. 2(a), when the exhausted amount ##EQU4## becomes identical with the compensated amount ##EQU5## Until the instant t.sub.3, before which (L.sub.R).sub.ref >L.sub.R, both the preset value (P.sub.s).sub.ref and the aperture V of the valve 7 decrease. But after t.sub.3, the condition that (L.sub.R).sub.ref <L.sub.R is reached so that (P.sub.s).sub.ref begins to increase. The actual change in the pressure P.sub.s lags behind the preset value (P.sub.s).sub.ref and therefore P.sub.s starts to increase in retard of (P.sub.s).sub.ref. At this time, (P.sub.s).sub.ref >P.sub.s so that the aperture V of the valve 7 continues to increase for a certain time after t.sub.3. When P.sub.s increases until P.sub.s >P.sub.c at t.sub.4, the check valve 9 is completely closed and thereafter W.sub.c =0. After t.sub.4, the water level L.sub.R lowers owing to the exhaustion of W.sub.TB and the condition that (L.sub.R).sub.ref >L.sub.R is again reached at t.sub.5. The operation after t.sub.5 is substantially the same as that after t.sub.0 in the initial stage and the description of the further operation will be omitted. The control apparatus C attempts to stabilize the reactor water level by repeating the control operation in the period from t.sub.0 to t.sub.5. In FIG. 2, although the level fluctuation appears considerable, this is for the convenience of explanation and the actual fluctuation can be confined within a very small range. Now, consider the operation of the control apparatus A while the control apparatus A is at operation. In the control apparatus A, the pressure detected by the detector 14 increases upon closure of the valve 7 so that the regulation valve 3 is choked to decrease the internal pressure of the heat exchanger 4. Consequently, the pressure at the inlet of the control valve 3 increases to decrease the flow W.sub.H. The control apparatus C performs the control of reactor water level and the control of pressure in cascade, but only the control of water level is basically necessary so that it is only necessary to control directly the aperture of the control valve 7 by the pressure setter 32. In practice, however, the response to the change in the variable to be controlled is slow if the water level control system alone is employed. Therefore, the provision of the pressure control system as a minor control loop will improve the stability of control. Next, the feedwater flow control apparatus B will be described. When the RHR system is started, the control apparatus B controls the temperature fall rates at various points in the reactor to be smaller than their corresponding, preset values. Namely, since the rate (-.DELTA.Temp/.DELTA.t) of the fall of the reactor water temperature as one of the typical temperature fall rates associated with a reactor is expressed such that ##EQU6## the flow W.sub.F of feed water into the reactor must be varied in accordance with the variations with time of Q.sub.R and W.sub.c in order to keep the temperature fall rate (-.DELTA.Temp/.DELTA.t) of the reactor water constant at a desired value. The symbols appearing in the expression (4) are the same as those used in the expression (2), W.sub.c (i.sub.f -i.sub.F) being an amount of heat required to make the make-up water flow W.sub.c saturated in the reactor and W.sub.c (i.sub.g -i.sub.f) to turn the saturated water W.sub.c into steam. To achieve such a control as described above, temperature detectors 34 and 35 detect temperatures at different points in the reactor 1 and the respective temperature fall rates (-.DELTA.Temp/.DELTA.t) are calculated by temperature fall rate calculator 36 and 37. The most critical one of the temperature fall rates is selected by a maximum temperature fall rate selector 38. On the basis of the difference between the above selected value and the value (-.DELTA.Temp/.DELTA.t).sub.ref set by a reference temperature-fall-rate setting means 39, the set value (W.sub.F).sub.ref for the flow of feed water supplied into the reactor 1 is determined by a function generator 40. The function generator 40 decreases the value for (W.sub.F).sub.ref if the temperature fall rate in question is excessive while it increases (W.sub.F).sub.ref in case of the rate being insufficient. The actual feedwater flow W.sub.F is controlled by controlling the flow W.sub.TB of steam supplied to the turbine 12 on the basis of the value (W.sub.F).sub.ref. Namely, on the basis of the difference between the flow W.sub.F detected by a flow detector 18 and the value (W.sub.F).sub.ref set by the function generator 40, a function generator 19 generates a turbine speed reference signal (N.sub.T).sub.ref. Consequently, the flow W.sub.TB of steam supplied to the turbine 12 is controlled by the combined operation of a turbine speed detector 20, a turbine speed controller 21 and the control valve 13 so that the turbine speed N.sub.T is controlled to make W.sub.F equal to (W.sub.F).sub.ref. As a result, the temperature fall rate (-.DELTA.Temp/.DELTA.t) in (4) is controlled to a desired value. For example, if the temperature fall rate is excessive, the function generator 40 delivers an output for decreasing (W.sub.F).sub.ref so that (N.sub.T).sub.ref decreases to decrease the feedwater flow W.sub.F by choking the valve 13. As a result of this control, the value W.sub.F (i.sub.g -i.sub.F) in (4) decreases so that (-.DELTA.Temp/.DELTA.t) decreases. On the other hand, if the temperature fall rate is insufficient, the function generator 40 delivers an output for increasing (W.sub.F).sub.ref so that (N.sub.T).sub.ref increases to increase the flow W.sub.F by opening the valve 13. Accordingly, W.sub.F (i.sub.g -i.sub.F) in (4) increases to increase (-.DELTA.Temp/.DELTA.t). The control hereafter is continued until the condition that (-.DELTA.Temp/.DELTA.t).apprxeq.(-.DELTA.Temp/.DELTA.t).sub.ref has been reached. In the feedwater control apparatus B, the function generators 40 and 19 may be replaced by blind controllers which receive deviation signals. The control apparatus B is a cascade control system comprising a temperature fall rate control section, a flow control section and a speed control section, but it may be only a temperature control system in which the control valve 13 for the turbine steam flow is directly controlled by the output of the function generator 40. The provision of the minor loops for flow and speed with the temperature fall rate control system will improve the stability and the response speed of the overall control system. As described above with the control apparatuses A, B and C of the RHR system shown in FIG. 1, if each of the fluctuations of the reactor water level and the temperature fall rate occurs separately, the corresponding individual control apparatus operates to compensate the fluctuation. It should here be noted that each of the water level controller 32 and the function generator 40 must have suitable sensitive thresholds and that only when they receive inputs in excess of the thresholds, they control the aperture of the valves 7 and 13. With this constitution, the control of too frequent and negligible fluctuations can be prevented from taking place. Now, the effects of the combination of the control apparatuses B and C will be considered. As seen from FIGS. 2a to 2d, the control apparatus C controls the make-up water flow W.sub.c only during a period ranging from t.sub.2 to t.sub.4 and the control apparatus B operates alone for the time other than the above period. Therefore, there is a possibility only during the period t.sub.2 to t.sub.4, that both the control apparatuses B and C operate jointly. In such a case, the most severe conditions are the following two: (a) the temperature fall rate is excessive and the reactor water level lowers, and (b) the rate is insufficient with elevated level. Consider first the former case (a). According to this invention, since the flow of steam going out of the reactor equals W.sub.H +W.sub.TB and the flow W.sub.F of feed water supplied into the reactor equals W.sub.H +W.sub.c, the following reasoning will hold. Since the temperature fall rate is excessive, the control tends toward decreasing W.sub.F. On the other hand, since the reactor water level L.sub.R lowers, P.sub.s decreases (the aperture of the value lessens) to decrease W.sub.H. The decrease in P.sub.s causes W.sub.c to increase. As a result, W.sub.F (i.sub.g -i.sub.F) in FIG. 4 decreases and W.sub.c (i.sub.f -i.sub.F) in (4) increases to decrease the temperature fall rate. The water level L.sub.R rises since W.sub.c in excess of W.sub.TB is supplied into the closed loop (i.e. feedwater pipe). Thus, the temperature fall rate-.DELTA.Temp/.DELTA.t and the reactor water level L.sub.R can both resume desired values. In the latter case (b) mentioned above, the insufficient rate of the fall of the reactor temperature causes W.sub.F to increase while the elevated water level L.sub.R causes the decrease in W.sub.c. Since in the RHR system having a closed loop, L.sub.R is determined depending only on W.sub.c, the transient rise of the water level returns to its initial level after the condition that W.sub.c =0 has been reached. The temperature fall rate in this case increases with the increase in W.sub.F (i.sub.g -i.sub.F). Also, the control apparatuses B and C can operate jointly in the two cases as follows: (c) insufficient temperature fall rate with lowered water level and (d) excessive temperature fall rate with elevated water level. In either case, the joint control serves to stabilize both the fluctuations. This will be briefly described below. Since the temperature fall rate is determined depending on both W.sub.H and W.sub.c, the control apparatus B is affected by the control apparatus C. However, after the condition that W.sub.c =0 has been reached, the control apparatus B operates alone. In addition to this, the control apparatus operates intermittently so that the above influence takes place only for a short period. This is how (-.DELTA.Temp/.DELTA.t) can be shifted toward stabilization. On the other hand, since the RHR system has a closed loop, the water level L.sub.R is normally determined depending on both W.sub.c and W.sub.TB (whereas the transient fluctuation of L.sub.R is caused by the variation of W.sub.F). When the control apparatus C is operating, W.sub.c >W.sub.TB. It is considered that the control apparatus C is seldom affected by the control apparatus B during this while. Thus, the water level L.sub.R is also stabilized. In conclusion, the simultaneous operation of the control apparatuses B and C also functions to stabilize both L.sub.R and (-.DELTA.Temp/.DELTA.t). FIGS. 4a to 4c show the performance characteristics of an RHR system as an embodiment of this invention. If the above described control is performed with (W.sub.F).sub.ref set at W.sub.F shown in FIG. 4a, the reactor water level can be kept constant while the temperature fall rate is maintained at a desired value. Namely, the reactor water level L.sub.R can be always kept constant by the supply of make-up water (FIG. 4c) and also the reactor temperature Temp can be lowered at the maximum fall rate, i.e. during the shortest period, except in the period of several tens of minutes after scram. Immediately after scram, Q.sub.R becomes very much, as shown in FIG. 2, and in order to attain the temperature fall rate having a usual value, the feedwater flow W.sub.F must be very large so that the turbine and the pump are overloaded. To prevent this overload, W.sub.F must be prevented from becoming too large by setting the maximum limit allowable to the output (W.sub.F).sub.ref of the function generator 40. As described above, according to this invention, the reactor water level can be stably kept constant and moreover the temperature fall rates can be kept at desired values smaller than the corresponding maximum values allowable, so that the reactor can be cooled down at the maximum speed (i.e. double the speed attainable by the conventional control system) without deteriorating the pressure vessel etc. Further, according to this invention, the automated system contributes much to the reduction of human labor. Although in the above embodiment the temperature of the saturated water in the reactor is directly detected, this invention can be equally applied to the system in which the pressure of the saturated steam in the reactor vessel is detected and converted to temperature to determine the temperature of the reactor water. This type of the embodiment will enjoy a higher control capability since the response of pressure to the rapid change in the variables of the reactor is faster than that of temperature. Furthermore, in the above embodiment, the temperature fall rates are obtained at several points in the reactor and the flow of feedwater into the reactor is determined depending on the maximum one of the temperature fall rate. In the case, however, where the reactor vessel, the flange of the reactor vessel, the feedwater nozzle etc. have their individual allowable temperature fall rates different from one another, a system may be employed in which reference values (W.sub.F).sub.ref are set for the respective members in the reactor and the flow of feedwater is determined by choosing the minimum one of the reference values. In this case, the temperature fall rate at the weakest member is controlled so that the reliability can be improved. In addition to the above merits, the following advantages can be obtained according to this invention. (1) Since the reactor water level can be stably kept constant, the thermal impacts which are inherent to the operation of the ECCS can be prevented so that the deterioration of the strength of the reactor structure can be suppressed. (2) Since the temperature fall rates at several points in the reactor can be set smaller than their allowable limiting values, the reactor can be cooled down in the shortest time without degrading the characteristics of instruments. (3) The automated constitution can reduce human labor and improve the operating reliability as well.
	description	The invention relates to the reprocessing of spent nuclear fuel, in particular spent nuclear fuels containing uranium oxide. Conversion of spent uranium oxide nuclear fuel into a plutonium rich molten salt fuel was described in the PCT application with publication number WO/2017/158335. The first step in the process described was electro-reduction of uranium oxide above the melting point of the uranium or uranium alloy produced. The problem with such electro-reduction that had been observed earlier was that the molten metal produced failed to agglomerate into a continuous molten phase. This problem was solved in WO/2017/158335 by reducing the uranium oxide in batches, with a period of reduction continuing without addition of more uranium oxide between batches of uranium oxide. During this period, higher actinides and some lanthanides were reduced to their metals, dissolved in the alloy and residual uranium oxide which was preventing agglomeration of the molten uranium metal was chemically reduced by the dissolved higher actinides and lanthanides. While this method appears effective, it prevents continuous operation of the process, producing a uranium alloy containing higher actinides and certain lanthanides in a batch approach. Batch processes are generally more costly but in the context of plutonium they are also severely limited in batch size, which is limited by the amount of plutonium that can be accumulated in the electrolyte before nuclear criticality concerns become unacceptable. A further problem with the batch process is that the inter-batch reduction process produces a mixed metal phase of actinides and lanthanides which are not mutually soluble. This inter-batch alloy must be mixed with the larger mass of primarily uranium alloy produced during the preceding batch reduction and mixed until the lanthanides dissolve in the larger mass of uranium. This adds substantial complexity and the production of a concentrated plutonium rich alloy at this stage imposes yet more stringent limits on the batch size that can be used without experiencing criticality problems. Finally, the batch process exposes the material containing the uranium alloy to substantially stronger reducing conditions which excludes use of commercial ceramics such as zirconia from the application. In particular, molten lanthanides as a separate metal phase are very aggressive reducers of ceramic oxides. There remains a need therefore for a method of electro-reduction, and conversion of the resulting uranium alloy to molten salt fuel which can be operated continuously but still achieves full agglomeration of the alloy into a single metal phase. According to a first aspect of the invention, there is provided a method of reprocessing spent nuclear fuel. The spent nuclear fuel is added to an electro-reduction cell, wherein the electro-reduction cell comprises a halide salt electrolyte, and anode, and a cathode comprising an alloy of uranium and a first metal forming a low melting point alloy with uranium, the first metal being one or more of: iron; chromium; nickel; manganese; and cobalt. The spent nuclear fuel is electrochemically reduced at a potential sufficient to reduce plutonium and lanthanides in the spent nuclear fuel, in order to form a molten alloy of the first metal, uranium and higher actinides present in the spent nuclear fuel. The alloy is extracted from the electro-reduction cell while uranium oxide is still present in the electro-reduction cell. The spent nuclear fuel comprises uranium oxide and at least 1 mol of lanthanides per tonne of uranium in the spent nuclear fuel, and the electro-reduction cell is operated at a temperature above the melting point of the alloy. According to a second aspect, there is provided apparatus for reprocessing spent nuclear fuel. The apparatus comprises an electro reduction cell, a feed, an alloy removal system, and a controller. The electro-reduction cell comprises a tank, an anode and a cathode, and a heating system. The tank is configured to contain a halide salt electrolyte. The anode and cathode are located within the tank and configured to electrochemically reduce spent nuclear fuel at a potential sufficient to reduce plutonium and lanthanides in the spent nuclear fuel, in order to form an alloy of a first metal, uranium and higher actinides present in the spent nuclear fuel, the cathode comprising an alloy of uranium and the first metal, the first metal being one or more of: iron; chromium; nickel; manganese; and cobalt. The heating system is configured to maintain the tank at a temperature above a melting point of the alloy. The feed is configured to provide spent nuclear fuel to the electro-reduction cell, the spent nuclear fuel comprising uranium oxide and at least 1 mole of lanthanides per tonne of uranium. The alloy removal system is configured to remove the alloy from the electro-reduction cell. The controller is configured to cause the alloy removal system to remove the alloy from the electro-reduction cell while uranium oxide remains in the cell. According to a third aspect, there is provided a method of detecting failure of a ceramic coating of an electrolysis cell. The electrolysis cell comprises an electrically conductive tank connected to electrical ground and having a ceramic coating on its inner surface, and containing an electrolyte, and anode, and a cathode, such that in normal use the electrolyte is not in contact with the electrically conductive material of the tank. Current is monitored between the electrically conductive tank and the electrical ground while the electrolysis cell is operating. A rise in said current is detected. In response to detecting a rise in said current, it is determined that the ceramic coating has failed. According to a fourth aspect, there is providedn apparatus for use as an electrolysis cell. The apparatus comprises an electrically conductive tank, an electrolyte, anode, and cathode, and a controller. The electrically conductive tank is connected to electrical ground and has a ceramic coating on its inner surface. The electrolyte, anode, and cathode are contained within the electrically conductive tank such that in normal use the electrolyte is not in contact with the electrically conductive material of the tank. The controller is configured to: monitor current between the electrically conductive tank and the electrical ground while the electrolysis cell is operating; detect a rise in said current; in response to detecting a rise in said current, determine that the ceramic coating has failed. Further embodiments are set out in claim 2 et seq It has been unexpectedly discovered that the addition of a metal such as iron, or other metal that reduces the melting point of the uranium alloy, to the molten uranium cathode in an electroreduction process (as described as a possible but rather undesirable option in WO/2017/158335) has an effect on the electrochemical behaviour of the reduction cell beyond the simple reduction in operating temperature described in WO/2017/158335. FIG. 1 shows the apparatus of WO/2017/158335. The apparatus comprises an outer structure 101, which encloses an electrolysis tank 102. The electrolysis tank 102 contains an electrolyte 103 in which is immersed an anode 104 (preferably located towards the top of the electrolyte) and a cathode 105. Spent nuclear fuel is added from a feed 106 into the electrolyte, and forms a layer 107. As current is passed between the anode and cathode, the spent fuel is electrochemically reduced to form an alloy 108 at the cathode 105. The alloy sinks to the bottom of the tank 102, and the cathode is preferably located such that the cathode will be immersed in the alloy when it forms. The process is performed at a temperature such that the resulting alloy is molten—the temperature is controlled by means of heaters and cooling ducts 109 located within insulation 110 placed between the outer structure and the tank. The electrolysis product at the anode will generally be a gas (e.g. oxygen where the spent fuel is an oxide fuel, or a halide where the spent fuel is a molten salt fuel), and this is released via an off gas tube 111 to a condenser (not shown). The apparatus also comprises an alloy removal system (not shown) for removing the molten alloy from the electrolysis tank. WO/2017/158335 makes no suggestion that adding iron to the cell affects the requirement set out in that patent that electrolysis must be continued until essentially all the uranium oxide or halide is reduced to metal in order for the molten alloy to successfully agglomerate into a uniform molten metal phase without entrained uranium oxide. It has now been unexpectedly found that provided iron or a similar metal (e.g. chromium, cobalt, manganese, or nickel, or some combination) is added to the molten cathode, electrolysis conditions can be devised for a uranium oxide fuel which result in agglomeration of the uranium alloy cathode into a uniform metal phase even though uranium oxide remains in the electro-reduction cell (i.e. in contact with the cathode). In order to ensure that the alloy forms, the electrochemical cell should be initiated with a molten uranium iron (or uranium plus another metal) alloy at the cathode, rather than solid iron (or other metal). The electrochemical cell is otherwise as illustrated in FIG. 1. The key to the required electrolysis condition is that the current density or cathode overpotential must cause reduction of at least a portion of one of the major lanthanide fission products present in spent nuclear fuel, e.g. cerium or neodymium, to their metal form which then dissolves in the molten uranium alloy cathode. Without wishing to be bound by theory, we believe that the reason for this unexpected result is that molten lanthanide metals such as cerium and neodymium are essentially immiscible in molten uranium but have substantial solubility in a uranium iron alloy (or an alloy of uranium and any combination of the metals noted above). This higher solubility reduces the activity coefficient of the lanthanide metal in the alloy so that significant reduction of the lanthanides to their metal form takes place under conditions which with a pure uranium cathode would not result in significant reduction. The dissolved lanthanide metals then chemically reduce the entrained uranium or higher actinide oxides in the alloy to uranium or higher actinides, producing as by-product lanthanide oxides which rise to the surface of the molten alloy as a result of their much lower density than uranium or higher actinide oxides. This allows the cathode to agglomerate into a uniform metal phase. The inventive process thus requires all of the following conditions. 1) Incorporation into the molten cathode of iron, chromium, cobalt, manganese, and/or nickel which each reduce the melting point of uranium and increases the solubility of lanthanides metals in the alloy 2) Electrolysis at a sufficiently high current density to cause reduction of lanthanide oxides or oxyhalides to metal dissolved in the uranium alloy cathode, even though excess uranium oxide remains in contact with the cathode 3) Presence of lanthanides (e.g, cerium or neodymium) in the spent fuel or, where levels of lanthanides are less than 1 mol/tonne uranium (for example in low burnup fuel) addition of a suitable lanthanide along with the spent fuel. The selection of the optimal current density is a necessarily empirical process. It is well known in the art that very high overpotentials can be applied in molten salt electrolysis without resulting in co-reduction of metals with significantly different reduction potentials. For example, reduction of a mixture of calcium chloride and sodium chloride in a Downs Cell results in production of pure sodium with negligible calcium contamination even at high electrode potentials and current densities. The discovery that uranium co-reduces with higher actinides and even lanthanides at such high overpotentials even when an excess of uranium oxide is present is therefore surprising. Without wishing to be bound by theory, we hypothesise that the reason that co-reduction occurs with uranium oxide as the main reducible material is the solid form and very low solubility of the uranium oxide in the electrolyte. This results in the rate of reduction of the uranium oxide being largely kinetically rather than thermodynamically limited which allows for co-reduction of the higher actinides when there is still unreduced uranium oxide in the cell. This hypothesis is supported by the observation that spent nuclear fuel in pellet form more readily co-reduces with plutonium and higher actinides than is the case with powdered oxide fuel where higher current density and overpotential is required to achieve co-reduction. The precise current density and voltages required will be affected by the geometry of the electro-reduction cell, the composition of the electrolyte, the nature of both the anode and cathode, the particle size and porosity of the feedstock spent uranium oxide fuel and the lanthanide content of the spent fuel. The inventive process thus uses an empirically determined current density leading to a cathode overpotential that is high enough to reduce lanthanides in the spent fuel to metal simultaneously with the reduction of the uranium and higher actinide oxides. For example, the overpotential may reduce at least 5% of the cerium or neodymium in the spent fuel to metal simultaneously with the reduction of the uranium and higher actinide oxides This process permits continuous or semi-continuous addition of spent oxide fuel to the electro-reduction cell with continuous or semi-continuous withdrawal of the resulting molten alloy. By semi-continuous, is meant that aliquots of feedstock or product are added or withdrawn from the apparatus in quantities substantially lower than the quantity in the apparatus. In either the continuous or semi-continuous case, in contrast to the method of WO/2017/158335, alloy is withdrawn while unreduced uranium oxide is still present in the electrolysis cell (in addition to any potential withdrawal after full reduction of the uranium oxide, e.g. when shutting down the continuous process). The alloy will consist primarily of uranium, higher actinides, lanthanides, noble and semi-noble metal fission products and other metals such as iron added along with the uranium oxide feedstock to reduce the melting point of the alloy. As the process continues to operate, fission products accumulate in the electrolyte and the concentrations of plutonium and americium not co-reduced with the uranium and therefore remaining in the electrolyte rise to an equilibrium level where addition of those elements in the spent nuclear fuel feedstock is equal to the rate of reduction of the elements and their removal in the molten alloy. The electrolyte must be replaced either when the heat producing capacity of the accumulated fission products (primarily caesium and strontium) exceeds the cooling capacity of the electroreduction cell or when the accumulation of certain fission products (again primarily caesium and strontium) results in sufficient co-reduction of the fission products with the uranium oxide as to seriously contaminate the molten alloy. Rendering the electrolyte substantially actinide free is a desirable feature of the process described in WO/2017/158335. That benefit is lost in the continuous process where periodic exhaustive reduction of the electrolyte is not carried out. An additional process is therefore required to remove residual actinides from the electrolyte prior to the disposal of the electrolyte as radioactive waste. The spent electrolyte can be cleaned of residual actinides by a number of procedures including exhaustive electrolysis (i.e. by ceasing adding new spent fuel, and continuing electrolysis until all of the actinides are reduced and dissolved into the alloy). The preferred method however is extraction of the actinides by exchange between the electrolyte and a molten metal such as bismuth or cadmium containing a dissolved metal that is more reactive than the actinides. Calcium is the preferred reactive metal although other group 1 & 2 metals such as magnesium or sodium can also be used. The exchange can be carried out in a multi-stage process, conveniently carried out in a column where the metal/salt exchange happens in a countercurrent manner over a number of “plates”. For industrial application, a particularly useful approach to the removal of residual actinides is to withdraw the spent electrolyte continuously or in small batches, recover the actinides into molten cadmium/calcium alloy then return the cadmium plus actinides to the electro-reducer along with replacement fresh electrolyte. The cadmium volatilises at the temperature of the electrolyte and is recovered in the off gas condenser system from the electro-reducer cell along with fission product cadmium from the spent fuel. With this approach, the composition of the electrolyte can be maintained substantially constant over months or years of operation resulting in greater uniformity of the output molten alloy. The molten alloy is converted to molten salt nuclear fuel as described in WO/2017/158335 through contacting the molten alloy with a molten salt mixture containing a salt that will be reduced to its metal by higher actinides or lanthanides in the molten alloy. This can be a batchwise operation or operated continuously in direct linkage to the output of the electro-reduction cell or with intervening storage of the alloy. Where the feedstock spent nuclear fuel is of low burnup and therefore contains lower amounts of lanthanide fission products than plutonium then direct extraction of the combined lanthanide and higher actinide components of the uranium alloy is practical. With higher burnup fuels, where higher concentrations of lanthanides are present however a pre-extraction of the molten alloy may be required whereby the higher proportions of the lanthanides in the spent fuel are reduced to lower concentrations by contacting the alloy with sufficient molten salt, typically UCl3 or FeCl2 together with NaCl, to extract a substantial fraction of the lanthanides into the salt while leaving substantially all of the higher actinides in the metal. The molten alloy output of that pretreatment stage can then be contacted with further molten salt to remove the remainder of the lanthanides and substantially all of the higher actinides from the molten alloy into the molten salt. While direct flow of alloy from the electro-reducing cell to the molten salt contacting apparatus can be industrially efficient, the fact that the electroreduction stage must be carried out in a high radiation facility (a hot cell) while the molten salt contacting stage can be carried out in a lower radioactivity facility means it can be convenient to extract the molten alloy from the electroreduction cell and cast it under an inert gas into solid metal pellets which are easily handled, mixed in bulk and transferred to the molten salt contacting apparatus where the pellets are remelted. A particularly simple procedure for the molten salt extraction of higher actinides from the alloy results from the inclusion of iron in the molten uranium alloy cathode. At the temperatures of molten uranium (1100 C) the separation factor for extracting plutonium and americium from the alloy into an NaCl based salt is relatively poor with 15% of the plutonium and 25% of the americium remaining in the alloy after a single extraction. At the lower temperature of 800 C permitted by inclusion of iron in the alloy, less than 2% of both plutonium and americium remain in the alloy. It is therefore practical to use a simple batchwise extraction process to recover the higher actinides from the uranium alloy. A suitable apparatus for such a process is shown in FIGS. 2A and 2B. The alloy 201 is accumulated, via the dip tube 105 from the electro-reduction cell, in a carbon steel container 202 coated with an yttria wash to protect the steel from corrosion by the molten uranium alloy. The time of contact is very limited so plasma spraying is not required though it is an option. Use of such washes in uranium casting is a well established method. Prior to addition of the alloy to this container, the container is heated empty in an argon atmosphere to 800° C. This causes the carbon in the steel to reduce the oxide layer on the steel surface to metal. This prevents the oxide layer from reacting with plutonium in the alloy forming plutonium oxide which will not extract into the salt. The salt 203 is added to the container and the alloy melted using induction heater/stirrers 204 in standard industrial forms. Efficient stirring of the deep layer of alloy ensures rapid equilibration of the two phases and is conveniently achieved in a non-contact manner by use of induction heaters to melt and stir the alloy. Mechanical stirring of the alloy is also possible as an alternative to the induction heaters. The salt phase does not require stirring as it is very shallow and mixes adequately by convection. When equilibration is complete, the container is allowed to cool until the uranium alloy freezes (725° C.). At that temperature the salt remains liquid and is withdrawn via a suction dip tube 205 after the entire apparatus has been tilted as shown in FIG. 2B. A single extraction is predicted to recover 98% of the Pu and Am into the salt layer. However, the remaining few % and any residual salt not recovered via the dip tube can be recovered by repeating the process with a fresh batch of salt. That second extraction will recover essentially all the Pu/Am in a salt which will be essentially 60% NaCl/40% UCl3. That second salt extract can then be used to carry out the first extraction of the next batch of uranium alloy. Extraction of alloy from the electroreduction cell is conveniently via a dip tube formed from a uranium alloy resistant material such as aluminium nitride or yttria coated steel with the alloy being transferred by differential gas pressure without the need for penetrations of the crucible containing the molten alloy cathode. For a continuous electro-reduction process to be practical, it is desirable that the materials of the electro-reduction cell have long lives under the conditions of operation. This is challenging where uranium alloy is produced because such alloys are known to corrode virtually all metals, including refractory metals such as tantalum. Ceramic containers are therefore desirable, but most ceramics are reduced to their metals by the highly reactive uranium and even more so, higher actinides and lanthanides. The most resistant ceramic oxide is known to be yttria but this has very limited physical strength and resistance to thermal cycling. Coating of metals with yttria by techniques such as plasma spraying have been explored. However, yttria has low but not insignificant solubility in molten salts and such protective coatings must be regularly inspected and repaired. Such a process is very challenging in the high radiation environment of nuclear fuel reprocessing. A novel method has been devised to overcome this difficulty of inspection. Plasma sprayed yttria (yttrium oxide) linings have very low electrical conductivity and there is therefore negligible current leakage from the molten uranium alloy cathode to the structural metal underlying the yttria coating. Even a small failure of the yttria lining however results in a large current leakage from the cathode to the structural metal and hence to earth. Incorporation of a suitable earth leak detector into the apparatus thus provides immediate warning of any failure of the yttria lining before the uranium alloy can significantly corrode the structural metal of the electrolysis cell. In fact, this approach is generalizable to any ceramic-coated electrolysis cell, as shown schematically in FIG. 3. The electrolysis cell comprises a metal tank 301, which is connected to electrical ground 311 and has on its inner surface a non-conductive ceramic coating 302. The electrolysis cell contains an electrolyte 303, in which are immersed an anode 304 and a cathode 305. When the ceramic coating is intact, the metal tank is insulated from the electrolyte, and so no (or only a very small) current flows to ground 311. Where there is a break in the ceramic coating below the level of the electrolyte, a large current will flow to ground via the electrolyte, the break, and the metal tank, and this can be detected by a current detector 312. An alternative to applying the yttria coating directly to the structural material of the electrolysis cell is shown in FIG. 4. The electrolysis cell 400 of FIG. 4 utilises a solid yttria crucible 401 to contain the molten cathode 402. The oxide pellets 403 float above the molten cathode 402 within the crucible. Other features of the cell may be as disclosed in any of the above examples. Such crucible are relatively fragile and have poor resistance to thermal shocks and thermal cycling. In order to provide them with a long operating life in the electro-reduction cell they can be entirely surrounded by the electrolyte 404. In this way the yttria crucible is protected from thermal stress. A further alternative is to use a lower cost and more robust ceramic for the electro reduction cell such as zirconia, magnesia, aluminium nitride, silicon carbide etc and to coat the structural ceramic with yttria, either with the ceramic in the “green” unfired form as a post firing procedure including plasma spraying as described above for metal cell materials In the case of any use of yttria in the electroreduction cell there is a challenge that yttria has significant solubility in molten salts. Yttria is however present in the spent nuclear fuel (as a fission product) and accumulates in the electrolyte and thereby radically reducing solubilisation of the yttria crucible. Optionally, powdered yttria can be added to the electrolyte to fulfil the same function where yttria from the spent fuel is insufficient to provide the necessary protection.
	description	1. Field This invention pertains generally to a method for operating a pressurized water nuclear reactor and more particularly to a method for automatically controlling the average coolant temperature and the axial power distribution of such a reactor. 2. Description of the Related Art The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated and in heat exchange relationship with a secondary circuit for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports a plurality of fuel assemblies containing fissile material. The primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump, and the system of pipes which are connected to the vessel form a loop of the primary side. For the purpose of illustration, FIG. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid reactor coolant, such as water or borated water, is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. An exemplary reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertical, co-extending fuel assemblies 22, for purpose of this description, the other vessel internal structures can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals' function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 2), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel through one or more inlet nozzles 30, flows down through an annulus between the reactor vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies are seated and through and about the assemblies. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding areas 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and friction of forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more coolant nozzles 44. The upper internals 26 can be supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate, primarily by a plurality of support columns 48. A support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plate 40. Rectilinearly moveable control rods 28, which typically include a drive shaft 50 and spider assembly 52 of neutron poison rods, are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and the top of the upper core plate 40. The support column 48 arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability. FIG. 3 is an elevational view, represented in vertically shortened form, of a fuel assembly being generally designated by reference character 22. The fuel assembly 22 is of the type used in a pressured water reactor and has a structural skeleton which, at its lower end, includes a bottom nozzle 58. The bottom nozzle 58 supports the fuel assembly 22 on the lower core plate 36 in the core region of the nuclear reactor. In addition to the bottom nozzle 58, the structural skeleton of the fuel assembly 22 also includes a top nozzle 62 at its upper end and a number of guide tubes or thimbles 84 which align with the guide tubes 54 in the upper internals. The guide tubes or thimbles 84 extend longitudinally between the bottom and top nozzles 58 and 62 and at opposite ends are rigidly attached thereto. The fuel assembly 22 further includes a plurality of transverse grids 64 axially spaced along and mounted to the guide thimbles 84 and an organized array of elongated fuel rods 66 transversely spaced and supported by the grids 64. Also, the fuel assembly 22, as shown in FIG. 3, has an instrumentation tube 68 located in the center thereof that extends between and is captured by the bottom and top nozzles 58 and 62. With such an arrangement of parts, fuel assembly 22 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 66 in the array thereof in the assembly 22 are held in spaced relationship with one another by the grids 64 spaced along the fuel assembly length. Each fuel rod 66 includes a plurality of nuclear fuel pellets 70 and is closed at its opposite ends by upper and lower end plugs 72 and 74. The pellets 70 are maintained in a stack by a plenum spring 76 disposed between the upper end plug 72 and the top of the pellet stack. The fuel pellets 70 composed of fissile material, are responsible for creating the reactive power of the reactor. The cladding which surrounds the pellets functions as a barrier to prevent the fission byproducts from entering the coolant and further contaminating the reactor system. To control the fission process, a number of control rods 78 are reciprocally moveable in the guide thimbles 84 located at predetermined positions in the fuel assemblies 22. A rod cluster control mechanism 80, positioned above the top nozzle 62, supports a plurality of the controls 78. The control mechanism has an internally threaded cylindrical hub member 82 with a plurality of radially extending flukes or arms 52 that form the spider previously noted with regard to FIG. 2. Each arm 52 is interconnected to a control rod 78, such that the control rod mechanism 80 is operable to move the control rods vertically in the guide thimbles 84 to thereby control the fission process in the fuel assembly 22 under the motive power of a control rod drive shaft 50 which is coupled to the control rod hub 80, all in a well known manner. The newer reactors, such as the AP1000 nuclear plant design offered by Westinghouse Electric Company LLC, Cranberry Township, Pennsylvania, employ two different types of control rods, i.e., the traditional control rods (black control rods) and gray control rods, the latter having a reduced reactivity worth, i.e., control rods that absorb fewer neutrons per unit area than the traditional control rods. The gray control rods are employed to implement a MSHIM operation and control strategy which has as an objective constant axial offset control. The term MSHIM is derived from the fact that reactivity control uses the gray control rod banks as a “mechanical shim” rather than the chemical shim, i.e., changes in soluble boron concentration, employed in a number of operating commercial reactors today, in order to provide fine reactivity control. the MSHIM strategy employs two independently controlled control rod groups to provide fine control of both the core reactivity and axial power distribution during a wide range of operational scenarios. In the AP1000 reactor design, the MSHIM operation and control strategy is implemented by a digital rod control system that automatically controls the core reactivity (reactor coolant system temperature) using four banks of gray control rods and two banks of traditional control rods, all moving in a defined overlap. Furthermore, automatic axial power distribution (i.e., the axial offset, also known as the core axial flux difference) control is provided using a single, heavy bank of traditional control rods which move independently of the reactivity control banks. Changes in the concentration of the chemical shim within the reactor coolant is generally limited to only that required to directly compensate for fuel and/or burnable absorber depletion during a given fuel cycle. The digital rod control system that is responsible for implementing the MSHIM operation and control strategy is basically characterized by the use of two separate rod controllers that independently maintain the reactor temperature and core power distribution, respectively within preselected bands. In order to achieve stable reactor control over the range of anticipated operating scenarios, the two rod controllers are interdependent in certain aspects. For instance, there is a prioritization scheme for the two rod controllers in scenarios where both controllers determine that rod motion is demanded. In such a case, the controller responsible for maintaining core power (average core temperature) in a specified band is given priority. However, it has been recognized by the inventors hereof that there are certain circumstances where core operation could be further improved by deviating from this strategy. Accordingly, it is an object of the embodiments hereafter described to provide a new operating strategy that will further enhances core stability and fuel performance. These and other objects are achieved by the inventions hereafter claimed which provide for a method of operating a pressurized water reactor that has a core of a plurality of fuel assemblies and at least a first bank of control rods that are primarily moved into and out of selected fuel assemblies in the core to adjust the axial flux difference to substantially maintain or restore the axial flux difference within a target band. Furthermore, the pressurized water reactor has at least a second bank of control rods that are primarily moved into and out of other selected fuel assemblies in the core to adjust the average temperature of the core to substantially maintain or return the average temperature to within a second target band. The operation of the first bank of control rods and the second bank of control rods is such that the first bank of control rods and the second bank of control rods are not moved together. The method gives the second bank of control rods priority of movement when the first bank of control rods and the second bank of control rods receive a demand signal at the same time to move in different directions. Furthermore, the method gives the first bank of control rods priority of movement when the first bank of control rods and the second bank of control rods receive a demand signal at the same time to move in the same direction. In one embodiment, when the first bank of control rods is moving and the second bank of control rods gets a signal instructing the second bank of control rods to move in a different direction, the first bank of control rods will stop moving and the second bank of control rods will take over movement as it was instructed. There are two aspects of reactor control in reactors that employ the AP1000 design. The M control banks (MA, MB, MC, MD, M1 and M2) automatically regulate the average reactor coolant temperature (Tavg) and the AO bank of rods automatically regulates the core axial flux difference (AFD). A core map which shows the location of each of the banks of control rods is shown in FIG. 4 and Table 1 identifies the types of rods employed by each of the banks, the number of clusters within each bank and their function. TABLE 1Bank IDGroup AssociationCluster Design Type# of ClustersMAMSHIM ControlGray (GRCA)4MB4MC4MD4M1Black (RCCA)4M28AOAxial Offset Control9S1Shutdown8S28S38S48Total69 The Tavg controller moves the M banks into or out of the core during power maneuvers to regulate the coolant temperature and restore it to a +/−1.5° F. deadband around a programmed value which is a function of the turbine load. Similarly, the AFD controller regulates the axial core power distribution and restores it to a +/−1% deadband around a target value. An assumption in the AP1000 reactor design safety analysis requires the Tavg control to have precedence over AFD control. As a result, during a power maneuver, the M banks move first to regulate the Tavg. As they move, they cause changes in AFD. When the coolant reaches its +/−1.5° F. control deadband, the M banks stop and the AO bank begins to regulate the AFD. The AO bank will move until the AFD is within its target deadband. The movement of the AO bank may cause the coolant temperature to exceed its control deadband. If this occurs, the AO bank will stop and the M banks will again move to correct the coolant temperature. When this is completed the AO bank will move again to resume the AFD correction. FIG. 5, which includes the graphs shown in FIGS. 5A, 5B, 5C and 5D, shows the AFD, Tavg, M and AO bank changes during a typical operation transient. Because the M banks have preference, the Tavg transient is well regulated. The AO bank correction near the end of the transient restores the AFD to within 1% of its target. In this example, the maximum deviation of the AFD from its control band is 8%. For more severe transients or under off normal conditions, the AFD deviation could be large enough to compromise peaking factors or pellet clad interaction limits (values as large as 20-30% have been seen in preliminary calculations). A more detailed understanding of the MSHIM operation and control strategy can be found in a paper entitled ROBUSTNESS OF THE MSHIM OPERATION AND CONTROL STRATEGY IN THE AP1000 DESIGN (Paper No. ICONE17-75314) which was given at the Proceedings of the 17th International Conference on Nuclear Engineering, Jul. 12-16, 2009, Brussels, Belgium. The inventors have recognized that allowing the AO bank to regulate the AFD during a plant operational transient, even where the average reactor coolant is outside its deadband, would reduce AFD deviations; but, at first glance, the Tavg control preference requirement imposed by the AP1000 safety analysis would seem to preclude that type of operation. However, close examination of the response characteristics of the M and AO banks does provide an opportunity for AFD correction during a major portion of any power change. Specifically: 1. Moving either the AO or M banks more deeply in to the core will cause a reduction in Tavg and moving either of them further out of the core will cause Tavg to increase. 2. Moving the AO bank (within its allowed operating band) more deeply into the core will cause the AFD to become more negative and moving it further out of the core will cause the AFD to become more positive.Accordingly, the underlying concepts provided for herein are: 1. If the M banks are moving into the core to reduce the Tavg and the AO bank has a demand to make the AFD more negative, allowing the AO bank to move will both reduce the Tavg and correct the AFD. 2. Similarly, if the M banks are moving out of the core to increase the Tavg and the AO bank has a demand to make the AFD more positive allowing the AO bank to move will both increase the Tavg and correct the AFD.Implementation of these concepts can be stated as follows: In the AP1000, when the AO and M banks both have a demand to move in the same direction (both in or both out of the core), disable the M banks and let the AO bank move. This will produce the correct movement of the Tavg and the AFD. The normal observation would be that allowing the AO and M banks to move together (since they both have a demand to move in the same direction) would improve regulation of both Tavg and AFD. This is true for Tavg control. Allowing both banks to move in the same direction would speed up the correction of Tavg. However, the same is not true for AFD control. The reactivity worths and overlaps of the M banks in the AP1000 design are such that as the M banks move in one direction (in or out) they alternately cause the AFD to become more negative and more positive. This is shown in FIG. 6. Hence, allowing both the AO and M banks to move simultaneously to speed up Tavg control is likely to be detrimental to AFD control. In addition, the design and arrangement of the rod control power supplies may preclude simultaneous movement of the AO and M banks. The fundamental principle underlying the inventions claimed hereafter is the fact that natural core feedbacks, i.e., changes in moderator temperature/density, generally result in consistent responses in Tavg and AFD. For instance, when core power is reduced, reactivity increases resulting in an increased Tavg. Coincidentally, AFD also becomes more positive. Both would require rod insertion to compensate. The invention claimed hereafter takes advantage of the fact that the heavy control rods, i.e., the black control rods, used for AFD/A0 control inherently have higher reactivity worth than the gray rods in the M banks normally used for Tavg control; meaning that the AO bank would compensate for both parameters under such conditions. In other words, the inherent, short-term core feedbacks are found to be naturally consistent and the method claimed hereafter leverages that consistency. This is not necessarily the case for the long-term core feedbacks, e.g., xenon transients, but the time spans are much longer for these affects, such that the “independence” of the two controllers is adequate to control for these long-term effects. A control system logic arrangement that will implement the concepts claimed hereafter is shown in FIG. 7. The reactor control system and the AFD control system will generate a demand for M and AO bank movement based on the need for correction of the coolant average temperature (Tavg) or core axial power distribution. The demand to reduce Tavg will move the M banks in except when there is a demand to make the AFD more negative. In this case, the M bank demand will be ignored and the AO bank will move in to reduce the Tavg at the same time make the AFD more negative. The demand to increase Tavg will move the M banks out except when there is a demand to make the AFD more positive. In this case, the M bank demand will be ignored and the AO bank will move out to increase the Tavg and at the same time make the AFD more positive. The demand to make the AFD more negative will move the AO bank in only when there is a corresponding demand to move the M banks in or when the M bank demand is in its control deadband. Similarly, a demand to make the AFD positive will move the AO bank out only when there is a corresponding demand to move the M banks out or when the M bank demand is in its controlled deadband. When the AO bank reaches its deadband and stops movement, the M banks will take over movement if Tavg is not in its deadband. This logic, shown in FIG. 7, demonstrates how Tavg control is given preference over AFD control while allowing AFD control for the majority of the time during an operational transient. FIGS. 8A, 8B, 8C and 8D show the effect of this control strategy on the same transient previously plotted in FIG. 5 for the prior art control strategy. The improvement in AFD control, without compromising Tavg control and while giving Tavg preference, is significant. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	claims	1. A target supply unit mounted to a chamber, comprising:a nozzle unit having a through-hole to allow a target material to be outputted therethrough;a cover provided to define a first space inside the cover and to cover the nozzle unit, the cover having a through-hole to allow the target material to pass therethrough from the first space to a second space inside the chamber; anda device configured to pump out gas inside the first space. 2. The target supply unit according to claim 1, further comprising:an electrode provided to face the nozzle unit; anda voltage generator configured to apply a voltage between the target material and the electrode. 3. The target supply unit according to claim 1, wherein the cover is formed of an electrically conductive material. 4. The target supply unit according to claim 1, whereinthe nozzle unit is disposed in the chamber,the cover is disposed in the changer to form together with the nozzle the first space which is a closed space inside the chamber. 5. A target supply unit mounted to a chamber, comprising:a nozzle unit having a through-hole to allow a target material to be outputted therethrough;a plurality of electrodes provided in a direction in which the target material travels;an electrical insulator for holding the plurality of electrodes;at least one voltage generator configured to apply a voltage between the plurality of electrodes;a cover provided to define a first space inside the cover and to cover the nozzle unit, the plurality of electrodes, and the electrical insulator, the cover having a through-hole to allow the target material to pass therethrough from the first space to a second space inside the chamber; anda device configured to pump out gas in the first space. 6. The target supply unit according to claim 5, whereinthe nozzle unit is disposed in the chamber, andthe cover is disposed in the changer to form the first space in the chamber, the first space is a closed space containing the nozzle unit, the plurality of electrodes, and the electrical insulator. 7. An extreme ultraviolet light generation apparatus, comprising:a chamber provided with a window through which a pulse laser beam travels into the chamber; anda target supply unit mounted to the chamber, the target supply unit comprising:a nozzle unit having a through-hole to allow a target material to be outputted therethrough;a cover provided to define a first space inside the cover and to cover the nozzle unit, the cover having a through-hole to allow the target material to pass therethrough from the first space to a second space inside the chamber; anda device configured to pump out gas inside the first space. 8. The extreme ultraviolet light generation apparatus according to claim 7, wherein the device is configured to pump out gas in the first space so that a gas pressure in the first space be lower than that in the second space. 9. The extreme ultraviolet light generation apparatus according to claim 7, whereinthe nozzle unit is disposed in the chamber,the cover is disposed in the changer to form together with the nozzle the first space which is a closed space inside the chamber. 10. An extreme ultraviolet light generation apparatus, comprising:a chamber provided with a window through which a pulse laser beam travels into the chamber; anda target supply unit mounted to the chamber, the target supply unit comprising:a nozzle unit having a through-hole to allow a target material to be outputted therethrough;a plurality of electrodes provided in a direction in which the target material travels;an electrical insulator for holding the plurality of electrodes;at least one voltage generator configured to apply a voltage between the plurality of electrodes;a cover provided to define a first space inside the cover and to cover the nozzle unit, the plurality of electrodes, and the electrical insulator, the cover having a through-hole to allow the target material to pass therethrough from the first space to a second space inside the chamber; anda device configured to pump out gas in the first space. 11. The extreme ultraviolet light generation apparatus according to claim 10, wherein the device is configured to pump out gas in the first space so that a gas pressure in the first space be lower than that in the second space. 12. The extreme ultraviolet light generation apparatus according to claim 10, whereinthe nozzle unit is disposed in the chamber, andthe cover is disposed in the changer to form the first space in the chamber, the first space is a closed space containing the nozzle unit, the plurality of electrodes, and the electrical insulator.
047160055	claims	1. A method of forming a seal between the confronting planar sealing surfaces on two annular structural members which are drawn together by axially extending bolts comprising the steps of: forming a seal assembly by locking a toroidal, crushable seal member within a substantially flat annular spacer member against an annular shoulder formed on the inner diameter thereof with an annular resilient member seated in a radially extending groove in the inner surface of said substantially flat annular member, said flat annular member being axially thinner than said toroidal, crushable seal member; placing said sealing assembly between said planar sealing surfaces and positively aligning said assembly relative to said axially extending bolts; and tightening said bolts to draw said planar sealing surfaces toward each other and into contact with the flat annular member while crushing said toroidal, crushable seal member to form a seal between said planar sealing surfaces. a toroidal, crushable seal member; a flat annular spacer member surrounding said toroidal, crushable seal member and defining an annular shoulder on the inner diameter thereof on which said toroidal, crushable seal member seats, said flat annular spacer member further defining a radially extending annular recess on its inner surface, and an annular resilient member seated in said annular recess and bearing against said toroidal, crushable seal member to center the toroidal, crushable seal member within the flat annular spacer member and resiliently lock the toroidal, crushable seal member against said shoulder while permitting the seal member to expand radially as it is crushed by the drawing of the planar surfaces on said structural members toward each other by the securing means, the thickness of said flat annular spacer member being less than that of said toroidal, crushable seal member to set the distance that the planar surfaces are drawn toward each other, and therefore provide uniform crushing of the toroidal, crushable seal member. a toroidal, crushable metallic, tubular seal member; a flat, annular spacer member defining a radially extending V-shaped annular recess in its inner surface and axially extending bores aligned with the bores in said pressure vessel and head flanges through which said bolts pass to position the spacer member relative to said flanges; and a toroidal coil spring seated in said V-shaped groove and sized to grip and position said toroidal, crushable metallic tubular seal member inside said flat annular spacer, said flat annular spacer being substantially rigid and axially thinner than said seal member such that as said bolts are tightened to draw said flanges toward each other and down against said spacer member, the seal member is crushed between said planar sealing surfaces to form a seal therebetween. 2. A sealing assembly for providing a seal between confronting planar sealing surfaces on two annular structural members which are drawn toward each other by securing means; said sealing assembly comprising: 3. The sealing assembly of claim 2 wherein said flat annular member defines means which engage the securing means to positively position said flat annular spacer member and therefore also said toroidal, crushable seal member between said planar sealing surfaces. 4. The sealing assembly of claim 3 wherein said securing means comprises elongated fasteners passing through said structural members including said sealing surfaces and wherein said flat annular spacer member defines passages through which said elongated fasteners pass to positively position said spacer member and therefore siad toroidal, crushable seal member relative to said structural members. 5. The seal assembly of claim 4 wherein said radially extending annular recess in said flat annular member is V-shaped. 6. The seal assembly of claim 5 wherein said annular resilient member is a toroidal coil spring. 7. The seal assembly of claim 6 wherein said toroidal, crushable seal member is a metallic tubular member. 8. The seal assembly of claim 7 wherein said structural members define a fluid pressure chamber and wherein said metallic, tubular seal member defines radial bores on its inner surface which communicate with the pressure chamber and through which pressurized fluid passes to pressurize the seal member. 9. The sealing assembly of claim 2 wherein said radially extending annular recess is substantially V-shaped and said annular resilient member is a toroidal coil spring. 10. The sealing assembly of claim 9 wherein said toroidal, crushable seal member is a metallic, tubular member. 11. In a nuclear reactor, the combination of a generally, cylindrical pressure vessel having a flange extending radially around its upper end which defines an axially facing planar sealing surface and axially extending bores passing through said sealing surface, a substantially hemispherical head having a radially extending flange around its lower end which defines an axially facing planar sealing surface complementary to the sealing surface on the pressure vessel flange and which also defines axially extending bores aligned with the bores in the pressure vessel flange, bolts passing through the bores in the pressure vessel and head flanges to draw said flanges toward each other, and a sealing assembly comprising: 12. The combination of claim 11 wherein said pressure vessel is pressurized and wherein said toroidal, crushable metallic tubular seal member defines radially extending bores in its inner surface through which said seal member is pressurized by pressure in the pressure vessel.
	abstract	A coupling structure of a fuel assembly includes a screw coupling means to provide a coupling structure of a fuel assembly which prevents foreign matters from mixing in the fuel and in which no vertical play is generated at a fitting portion between a latch sleeve 10 and a lock key 20. The coupling structure of the fuel assembly includes: the latch sleeve 10 having a projecting portion 11 formed at an upper end portion thereof; an upper nozzle 30 having formed therein a latch sleeve installation hole 31 into which the upper end portion of the latch sleeve 10 is inserted; and the lock key 20 which is installed inside the latch sleeve installation hole 31, and which has an opening portion 21 formed on a bottom surface thereof, the opening portion 21 having a shape corresponding to the projecting portion 11. The screw coupling means screw couples the lock key 20 to the upper nozzle 30 with a male screw thread 23 formed on a side surface of the lock key 20.
	claims	1. An X-ray computer tomography apparatus comprising:an X-ray tube which generates X-rays;a two dimensional array type X-ray detector which detects the X-rays;a rotating mechanism which rotates the X-ray tube and the X-ray detector around a rotation axis;a pair of collimators which form X-rays from the X-ray tube into a cone beam shape;a collimator moving mechanism which separately moves the pair of collimators in a direction substantially parallel to the rotation axis;a reconstruction processing unit which reconstructs image data in a reconstruction range set by an operator based on an output from the X-ray detector; anda collimator control unit which controls a position of each of the collimators,wherein the collimator control unit controls the position of each of the pair of collimators in accordance with a distance between a central plane of the X-rays which corresponds to a cone angle of substantially 0° and an end face of the reconstruction range, andthe collimator moving mechanism moves each of the pair of collimators in a range from an outermost position corresponding to a maximum cone angle corresponding to a width of the X-ray detector, beyond the central plane of the X-rays, at each edge of the reconstruction range, to an innermost position offset from a position corresponding to a cone angle of substantially 0° to an opposite side. 2. The apparatus according to claim 1, wherein the collimator control unit determines the position of each of the pair of collimators based on one of a distance end and a near point on an end face of the reconstruction range when viewed from the X-ray tube. 3. The apparatus according to claim 2, further comprising a bed top configured to have an object placed thereon, and wherein the collimator control unit switches, in a scan start interval, a reference for determination of the position of one of the pair of collimators from the distant end to the near point in accordance with movement of the bed top. 4. The apparatus according to claim 3, wherein the collimator control unit increases a moving speed of one of the pair of collimators in synchronism with switching from the distant end to the near point. 5. The apparatus according to claim 3, wherein the collimator control unit switches, in a scan end interval, a reference for determination of the position of the other of the pair of collimators from the near point to the distant end in accordance with movement of the bed top. 6. The apparatus according to claim 5, wherein the collimator control unit decreases a moving speed of the other of the pair of collimators in synchronism with switching from the near point to the distant end. 7. The apparatus according to claim 5, wherein the collimator control unit fixes both the positions of the pair of collimators at the outermost position corresponding at a maximum cone angle in an intermediate interval between the scan start interval and the scan end interval. 8. The apparatus according to claim 7, wherein the collimator control unit fixes the position of the other of the pair of collimators to the outermost position in the scan start interval. 9. The apparatus according to claim 1, further comprising a bed top configured to have an object placed thereon, and wherein the collimator control unit moves, in a scan start interval, the position of one of the pair of collimators from the innermost position to the outermost position in accordance with movement of the bed top. 10. The apparatus according to claim 9, wherein the collimator control unit moves, in a scan end interval, the position of the other of the pair of collimators from the outermost position to the innermost position in accordance with movement of the bed top on which the object is placed. 11. The apparatus according to claim 10, wherein the collimator control unit fixes the position of one of the pair of collimators to the outermost position in the scan end interval. 12. The apparatus according to claim 10, wherein the collimator control unit fixes both the positions of the pair of collimators at the outermost position corresponding to a maximum cone angle in an intermediate interval between the scan start interval and the scan end interval. 13. The apparatus according to claim 1, wherein the collimator control unit controls the position of each of the collimators in accordance with a distance obtained by subtracting a predetermined distance from a distance between the central plane of the X-rays and the end face of the reconstruction range. 14. An X-ray computer tomography apparatus comprising:an X-ray tube which generates X-rays;a two dimensional array type X-ray detector which detects the X-rays;a rotating mechanism which rotates the X-ray tube and the X-ray detector around a rotation axis;a bed including a movable top on which an object is placed;a pair of collimators which form X-rays from the X-ray tube into a cone beam shape;a collimator moving mechanism which separately moves the pair of collimators in a direction substantially parallel to the rotation axis;a reconstruction processing unit which reconstructs image data in a reconstruction range set by an operator based on an output from the X-ray detector;a collimator control unit which controls the collimator moving mechanism to separately set a position of each of the pair of collimators based on the position of the top; andthe collimator moving mechanism moves each of the pair of collimators in a range from an outermost position corresponding to a maximum cone angle corresponding to a width of the X-ray detector, beyond the central plane of the X-rays, at each edge of the reconstruction range, to an innermost position offset from a position corresponding to a cone angle of substantially 0° to an opposite side. 15. An X-ray computer tomography apparatus comprising:an X-ray tube which generates X-rays;a two dimensional array type X-ray detector which detects the X-rays;a rotating mechanism which rotates the X-ray tube and the X-ray detector around a rotation axis;a bed including a movable top on which an object is placed;a variable aperture collimator mechanism including a pair of collimators which forms X-rays from the X-ray tube into a cone beam shape;a reconstruction processing unit which reconstructs image data in a reconstruction range set by an operator based on an output from the X-ray detector;a collimator control unit which controls the collimator mechanism to set an asymmetrical aperture of the collimator mechanism based on the position of the top; anda collimator moving mechanism that moves each of the pair of collimators in a range from an outermost position corresponding to a maximum cone angle corresponding to a width of the X-ray detector, beyond the central plane of the X-rays, at each edge of the reconstruction range, to an innermost position offset from a position corresponding to a cone angle of substantially 0° to an opposite side. 16. An X-ray computer tomography apparatus comprising:an X-ray tube which generates X-rays;a two dimensional array type X-ray detector which detects the X-rays;a rotating mechanism which rotates the X-ray tube and the X-ray detector around a rotation axis;a reconstruction processing unit which reconstructs image data in a reconstruction range set by an operator based on an output from the X-ray detector;a pair of collimators which form X-rays from the X-ray tube into a cone beam shape; anda collimator moving mechanism which separately moves each of the pair of collimators in a range from an outermost position corresponding to a maximum cone angle corresponding to a width of the X-ray detector, beyond the central plane of the X-rays, at each edge of the reconstruction range, to an innermost position offset from a position corresponding to a cone angle of substantially 0° to an opposite side.
	summary

Subsets and Splits

No community queries yet

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