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050341839	claims	1. An apparatus for producing a self-colliding charged particle beam comprising: magnetic focusing means for producing a ring-shaped magnetic field symmetrically spaced about an axis such that the magnetic field in a median plane perpendicular to said axis in the region across said ring-shaped field is unidirectional in a first direction parallel to said axis, and wherein said field is nonuniform with axial components in said first direction and radial components directed toward and away from said axis in the region above the below said plane such that the magnetic field strength of said field increases from substantially zero to some maximum value and then decreases to substantially zero as the radial distance between said field and said axis increases, said magnetic focusing means including fist, second, third and fourth concentric coils mounted coaxially with said axis, said first and second coils of substantially equal size spaced equally on either side of said median plane and magnetically polarized in said first direction, said third and fourth coils of substantially equal size to each other and of a smaller diameter than the diameters of said first and second coils, said third and fourth coils equally spaced on either side of said median plane and being magnetically polarized in the opposite direction to the polarization direction of said first and second coils; and particle beam means including a plurality of high-velocity, charged particles having orbits that intersect said plane, said axis and said ring-shaped magnetic field, sand wherein said charged particles travel radially away from said axis in in substantially straight lines, interact with said axial components to curve as they traverse said field, and then travel radially toward said axis in substantially straight lines, and wherein particles traversing said field above and below said plane and having velocities in a predetermined range are focused axially into said median plane by said radial components and particles having velocities outside said range are defocused axially away from said median plane by said radial components. forming a ring-shaped magnetic field symmetrically spaced about an axis with first, second, third and fourth concentric coils mounted coaxially with said axis, said first and second coils of substantially equal size magnetically polarized in a first axial direction and spaced equally on either side of a median plane perpendicular to said axis, said third and fourth coils of substantially equal size and of a smaller diameter than the diameters of said first and second coils, said third and four coils equally spaced on either side of said median plane and being magnetically polarized in the opposite direction to the polarization direction of said first and second coils such that the magnetic field in said median plane is substantially zero in the region of said axis and in the region across said ring-shaped field is unidirectional in said first axial direction, and wherein said field is nonuniform with axial components in said first axial direction and radial components in the region above and below said plane such that the magnetic field strength of said field increases from substantially zero to some maximum value and then decreases to substantially zero as the radial distance between said field and said axis increases; directing a plurality of high-velocity, charged particles in substantially straight, radial lines from said axis toward said ring-shaped field; causing said particles to interact with said axial components and curve as they traverse said field; causing said particles to interact with said radial components such that particles traversing said ring-shaped field above and below said median plane and having velocities in a predetermined range are focused axially toward said median plane by said radial components and particles having velocities outside said range are defocused axially away from said median plane by said radial components; and causing said particles to leave said field and travel radially toward said axis in substantially straight lines. 2. A method of producing a self-colliding charged particle beam comprising:
040594831	abstract	The ability of a nuclear reactor to withstand high seismic loading is enhanced by including, on each fuel assembly, at least one seismic grid which reduces the magnitude of the possible lateral deflection of the individual fuel elements and the entire fuel assembly. The reduction in possible deflection minimizes the possibility of impact of the spacer grids of one fuel assembly on those of an adjacent fuel assembly and reduces the magnitude of forces associated with any such impact thereby minimizing the possibility of fuel assembly damage as a result of high seismic loading. The seismic grid is mounted from the fuel assembly guide tubes, has greater external dimensions when compared to the fuel assembly spacer grids and normally does not support or otherwise contact the fuel elements. The reduction in possible deflection is achieved through reduction of the clearance between adjacent fuel assemblies made possible by the use in the seismic grid of a high strength material characterized by favorable thermal expansion characteristics and minimal irradiation induced expansion.
	summary
	summary
	abstract	The X-ray tube disclosed herein includes an electron emission unit including an electron emission element using a cold cathode; an anode unit disposed opposite to the electron emission unit, with which electrons emitted from the electron emission unit collide; and a focus structure disposed between the electron emission unit and a target unit disposed on a surface of the anode unit that is opposed to the electron emission unit. The electron emission unit is divided into a first region and a second region which can independently be turned ON/OFF. The X-ray tube is focus-designed such that collision regions, at the anode unit, of electron beams emitted from the respective first region and second region substantially coincide with each other.
059129344	claims	1. A system for inspection of fuel cells within a nuclear facility pool, the fuel cells being spaced apart from adjacent fuel cells at a cell spacing, and having a fuel cell surface with a plurality of edges and a plurality of height variations relative to a background area, the system comprising: at least one illumination source for emitting a pulsed light to impinge upon the fuel cell surface at a predetermined location within the nuclear facility pool so that an optically-detectable contrast is created by at least one of the plurality of height variations a camera disposed along an optical path intersecting the fuel cell surface for receiving and detecting light reflected from the fuel cell surface and for generating a digital image signal representative of an image of the reflected light; a timing controller for generating a time signal for synchronizing the activation of each illumination source with detection of reflected light by the camera; a processor for generating a processed image connected to the camera and the at least one illumination source, the processor including an image processor for receiving and processing the digital image signal to enhance the optically-detectable contrast; and a storage medium in communication with the processor for storing the processed image and a designation corresponding to the predetermined location within the nuclear facility pool. a plurality of light sources for emitting an illuminating light to impinge upon the of the plurality of surfaces, each light source comprising a strobe light disposed at an angle with respect to an optical path that is substantially perpendicular to the plurality of surfaces so that a transition between a raised area and a lowered area of the plurality of raised and lowered areas is visually enhanced; a timing controller for selectively activating the light sources one at a time; a camera disposed along the optical path above the plurality of surfaces for receiving and detecting an image of the transition when illuminated by each light source and for generating a digital image signal representative of the image; a processor including an image processor for receiving the digital image signal corresponding to illumination by each light source and for combining and processing the digital image signals to generate a processed image of the topography; and a storage medium for storing the processed image and a corresponding set of data identifying the selected location within the pool. positioning a camera over the fuel cells at a known distance; illuminating the upper surface of the fuel cells using a plurality of light sources, each light source being positioned to emit light at an angle relative to the upper surface so that shadows are formed by the plurality of height differentials, and each light source being separated form other light sources so that a different set of shadows is formed corresponding to each light source; capturing an image of the shadows generated by each light source; and combining and image processing the images of the shadows from the plurality of light sources to generate an enhanced image. illuminating the plurality of surfaces using a plurality of light sources so that a transition between a raised area and a lowered area of the plurality of raised and lowered areas is visually enhanced, each light source comprising a strobe light; one at a time, selectively activating a timing controller for the light sources; positioning a camera having a field of view of the surface; detecting an image of the transition when illuminated by each light source; generating a digital image signal representative of the transition image; combining the digital image signal corresponding to illumination by each light source and processing the combined digital image signals to generate a processed image of the topography; and storing in a storage medium the processed image and a corresponding set of data identifying the selected location within the pool. processing the combined digital image signals to identify the alphanumeric characters; and storing in the storage medium the identities of the alphanumeric characters. 2. The system of claim 1, wherein a portion of the plurality of height variations comprises a plurality of alphanumeric characters imprinted on the fuel cell surface and wherein the processor further includes a character recognition algorithm for operating on the processed image to determine an identity for each of the plurality of alphanumeric characters. 3. The system of claim 1, wherein the at least one illumination source comprises a plurality of light sources, each light source being disposed at an angle with respect to the optical path so that the illuminating light impinges upon the fuel cell surface at an angle to create a shadow corresponding to the plurality of height variations in the fuel cell surface. 4. The system of claim 3, wherein the plurality of light sources is disposed within a circle perpendicular to and centered on the optical path, each light source being separated from an adjacent light source by an angle within the range of 90.degree. and 120.degree.. 5. The system of claim 3, wherein the each light source is a flash lamp for producing pulses of light. 6. The system of claim 5, wherein the flash lamp is formed using quartz tubing. 7. The system of claim 1 wherein the timing signal activates each light source at different times from other light sources of the plurality. 8. The system of claim 7, wherein each light source is a strobe light and the timing signal activates each strobe light in an alternating sequence. 9. The system of claim 8, wherein the camera generates a snap shot of the fuel cell surface corresponding to an activation of each strobe light. 10. The system of claim 9, wherein the image processor combines a plurality of snap shots from a plurality of activations of each strobe light to generate a filtered image signal. 11. The system of claim 10, wherein the image processor combines the filtered image signals corresponding to each strobe light to generate the processed image. 12. The system of claim 1, wherein each of the at least one illumination source and the camera are constructed using radiation tolerant materials. 13. The system of claim 1, further comprising a location detection means for generating a location signal indicative of the predetermined location within the nuclear facility pool, the location detection means connected to the processor for providing the location signal for storage of the designation corresponding to the predetermined location in the storage medium. 14. A system of claim 1, wherein a transition between the edges of two adjacent fuel cells and the cell spacing therebetween comprise a second portion of the plurality of height variations within the fuel cell surface, and wherein the camera is positioned at a known height above the fuel cell surface and the processor further includes calibration means for measuring the cell spacing from the processed image. 15. A system of claim 1, wherein a plurality of edges of a single fuel cell comprise a third portion of the plurality of height variations in the fuel cell surface, wherein the camera is positioned at a known height above the fuel cell surface and the processor further includes calibration means for measuring a cell size. 16. A system for inspection of a plurality of surfaces submerged at a predetermined location within a nuclear facility pool, the plurality of surfaces having a topography comprising a plurality of raised areas and lowered areas, the system comprising: 17. The system of claim 16, wherein plurality of raised areas and lowered areas are alphanumeric characters imprinted into a background area, the processor further including a character recognition algorithm for operating on the processed image to identify the alphanumeric characters, and wherein the storage medium is further used for storage of the identities of the alphanumeric characters. 18. The system of claim 16, wherein the plurality of raised areas and lowered areas comprises a plurality of fuel cells and a plurality of gaps therebetween, respectively, the processor further including calibration means for measuring a gap size within the plurality of gaps from the processed image. 19. The system of claim 16, wherein the plurality of raised areas and lowered areas comprises an upper surface of a fuel cell, the raised areas corresponding to an upwardly tilt of the upper surface and the lower areas corresponding to a downward tilt of the upper surface. 20. The system of claim 16, wherein the plurality of light sources is disposed within a plane perpendicular to and centered on the optical path, each light source being separated from an adjacent light source by an angle within the range of 90.degree. and 120.degree.. 21. The system of claim 16, further comprising a timing controller for selectively activating the strobe lights one at a time. 22. The system of claim 21, wherein the strobe light includes a quartz flash lamp. 23. The system of claim 21, wherein the camera generates a snap shot corresponding to each activation of the strobe lamp. 24. The system of claim 23, wherein the processor combines a plurality of snap shots for generating the processed image. 25. A method for inspecting fuel cells submerged within a containment pool, the fuel cells being spaced apart from adjacent fuel cells at a cell spacing, and having a fuel cell surface with a plurality of edges, each fuel cell having a plurality of height differentials in the fuel cell surface relative to a background area, the method comprising: 26. The method of claim 25, wherein each light source is a strobe light, and the plurality of light sources is activated one-at-a-time in a sequence so that each activation results in a snap shot of the shadows. 27. The method of claim 25, wherein the step of positioning a camera over the fuel cells comprises attaching the camera to a fuel cell handling system for positioning the fuel cells within the containment pool. 28. The method of claim 25, wherein a portion of the plurality of height differentials corresponds to a plurality of alphanumeric characters of a serial number imprinted on the fuel cell surface, and further comprising the step of performing an optical character recognition algorithm on the enhanced image to determine the serial number of each said fuel cell. 29. The method of claim 25, wherein the plurality of height differentials includes a second portion of height differentials corresponding to a transition between the edges of two adjacent fuel cells and the cell spacing therebetween, the method further comprising determining a calibration factor for measuring a distance of the cell spacing from the enhanced image. 30. The method of claim 25, wherein the plurality of height differentials includes a third portion of height differentials corresponding to a tilted fuel cell wherein one edge of the tilted fuel cell is higher than another edge, the method further comprising a means for determining that the one edge of the tilted fuel cell is closer to the camera. 31. The method of claim 25, wherein the steps of illuminating the upper surface, capturing an image and combining and image processing the images are performed during a fuel cell loading procedure in the containment pool. 32. The method of claim 25, further comprising storing the enhanced images in a computer memory along with fuel cell identification and location data. 33. The system of claim 1, further comprising a fuel cell handling mechanism, wherein at least one element of said system for inspection of fuel cells is attached to said fuel cell handling mechanism, and wherein said fuel cell handling mechanism is adapted to provide input data to the storage medium with the designation corresponding to the predetermined location within the nuclear facility pool. 34. The system of claim 16, wherein the camera and the plurality of light sources are mounted on a fuel cell handling machine. 35. A method for inspection of a plurality of surfaces submerged within a nuclear facility pool at a predetermined location, the plurality of surfaces having a topography comprising a plurality of raised areas and lowered areas, the method comprising: 36. The method of claim 35, wherein the plurality of raised areas and lowered areas are alphanumeric characters imprinted into a background area, and further comprising the steps of: 37. The method of claim 35, wherein the plurality of light sources and the camera are mounted on a fuel cell handling machine.
	abstract	An emergency core cooling system directly injects emergency core cooling water, which is supplied from a high-pressure safety injection pump or a safety injection tank for a pressurized light water reactor, into a reactor vessel downcomer. A pipe connector is completely removed from between each direct vessel injection nozzle and each injection extension duct installed on an outer surface of the core barrel, which are opposite to each other. An emergency core cooling water intake port, through which the water is injected from each direct vessel injection nozzle, is formed on the surface of each injection extension duct facing an axis of each direct vessel injection nozzle. Thereby, a structure in which a jet of the emergency core cooling water flows into the injection extension ducts is adopted in a hydraulic connection fashion.
	summary
	claims	1. A method for evaluating an axial void fraction distribution of a fuel irradiated in a nuclear reactor, the method comprising:measuring a first intensity Az of radioactivity originated from a first group at an axial position of the fuel, the first group consisting of neutron-emitting nuclides generated by a neutron capture reaction of a heavy nuclide;measuring a second intensity Bz of radioactivity originated from a second group at the axial position of the fuel, the second group consisting of gamma-emitting radioactive fission product nuclides that is not generated by a neutron capture reaction and of which generated amount is substantially proportional to an exposure of the fuel in a certain range of the exposure;measuring a first reference intensity A0 of radioactivity of the first group at an axial reference position of the fuel at which a void fraction of the fuel can be evaluated;measuring a second reference intensity B0 of radioactivity rays of the second group at the axial reference position;calculating an exponent constant a used in an expression of Az=az×Eα and A0=a0×Eα where E is an exposure of the fuel, az and a0 are proportionality constants;evaluating a value of (az/a0) by an equation of (az/a0)=(Az/A0)(B0/Bz)α(bz/−b0)α, where bz is a value used in an expression of Bz=bz×E as a proportionality constant, b0 is a value used in an expression of B0=b0×E as a proportionality constant;evaluating a correlation curve of (az/a0) and a void fraction; andevaluating the axial void fraction distribution based on the value of (az/a0) and the correlation curve. 2. The method of claim 1, wherein the first intensity and the first reference intensity are neutron emission rates, and the second intensity and the second reference intensity are gamma ray intensities of a fission product that is proportional to the exposure within a certain exposure range. 3. The method of claim 2, wherein the neutron emission rates include rates of neutron emission of nuclides except curium 242, and the gamma ray intensities include a gamma ray intensity of cesium 137. 4. The method of claim 2, wherein the neutron emission rates include rates of neutrons emission of curium 244, and the gamma ray intensities include a gamma ray intensity of cesium 137 or cerium 144. 5. The method of claim 2, wherein the neutron emission rates include rates of neutrons emission of curium 242, and the gamma ray intensities include a gamma ray intensity of cesium 137 or cerium 144. 6. The method of claim 2, wherein the neutron emission rates are neutron emission rates except for curium 242 and curium 244, and the gamma ray intensities include a gamma ray intensity of cesium 137 or cerium 144. 7. The method of claim 2, further comprising:evaluating a neutron multiplication factor kz at a plurality of axial position of the fuel assembly, and a neutron multiplication factor k0 at the reference position;wherein the step of evaluating a value of (az/a0) includes:measuring a neutron flux or a neutron counting rate φz;measuring a neutron flux or a neutron counting rate φ0 at the reference position;calculating a ratio of the neutron emission rate to that at the reference position, Sz/S0, by an equation of Sz/S0=(φz/φ0)/(1−kz)/(1−k0); andassuming Az/A0 equal to Sz/S0. 8. The method of claim 2, further comprising:calculating an axial exposure distribution based on an assembly-averaged fuel exposure and an axial distribution of an intensity of gamma ray, by assuming the axial exposure distribution equal to the axial distribution of an intensity of gamma ray; and,evaluating a conversion factor from an infinite multiplication factor to a neutron multiplication factor by a neutron transport diffusion calculation according to a condition of the measurement;wherein the step of evaluating the axial void fraction distribution includes evaluating the axial void fraction distribution by repeating following (i) to (vii) steps until a convergence of the axial void fraction distribution, assuming an infinite multiplication factor can be expressed as a quadratic function of the exposure where parameters of the function depends on the void fraction:(i) assuming an axial void fraction distribution as appropriate one for an initial calculation, or an axial void fraction distribution calculated at the step of (vii) of a previous loop for a calculation except the initial calculation;(ii) calculating an infinite multiplication factor based on the assumed axial void fraction distribution;(iii) calculating an neutron multiplication factor kz and k0 by using the conversion factor and the multiplication factor;(iv) calculating a ratio of neutron emission rate to that at the reference position Sz/S0 based on the neutron multiplication factor kz and k0;(v) calculating az/a0 based on Sz/S0;(vi) evaluating the axial void fraction distribution based on az/a0 and the correlation curve; and(vii) terminating if a convergence of the axial void fraction distribution is achieved, or returning to the step of (i), if not. 9. The method of claim 8, further comprising:evaluating an exposure at the reference position by a neutron emission rate technique. 10. The method of claim 1, wherein the first group includes gamma ray emitting nuclides transmuted by neutron capture reactions after generation by fission, and the second group includes gamma ray emitting nuclides that is not subjected to a neutron reaction and emit gamma ray proportional to the exposure within a definite range. 11. The method of claim 10, wherein the first group includes cesium 134 or europium 154, and the second group includes cesium 137 or cerium 144.
044407157	summary	BACKGROUND OF THE INVENTION The present invention relates to a method of controlling a nuclear power plant. More particularly, the invention concerns a controlling method which permits the output of a reactor to be controlled to a permissible maximum level while maintaining a water level within the reactor in a safe range, even when partial shutdown occurs in a feed water pump system. In the first place, description will be made briefly of a nuclear power plant which the present invention concerns. As is well known, steam produced from a nuclear reactor is supplied to a main turbine by way of a main steam conduit and a regulating valve. Thereafter, the steam is condensed to water through a condenser. The condensate or water is then fed back to the reactor as feed water through a feed water pipe line under action of a feed water pump system. In the nuclear power plant of this type, the feed water flow pumped by the pump system is controlled by using at least one of signals representing the water level within the reactor, the primary steam flow rate flowing through the main steam conduit and the feed water flow, respectively, so that the water level within the reactor remains constant. The feed water pump system may be composed of four individual pumps, that is, two main pumps each of a 55%-capacity and two auxiliary pumps each of a 27.5%-capacity, where "55%" and "27.5%" indicate the ratios to the amount of the feed water flow required when the plant operates at its rating capacity. When the nuclear reactor is operated with the output in the vicinity of the rated power, the two main pumps are driven for supply feed water. The auxiliary pumps are used for backing up the main feed water pump or for starting the nuclear power plant. With the arrangement of the feed water control system for the nuclear reactor described above, the feed water is supplied to the reactor by the two main feed water pumps of 55%-capacity in the normal operating condition. When the main feed water pump is tripped or shut down for some reasons, then the two auxiliary pumps each of 27.5%-capacity are started to assure the feed water flow to the reactor. By the way, since the main and auxiliary pumps differ from each other in respect of the capacity as mentioned above, there is necessarily brought about a difference between the primary steam flow rate and the feed water flow rate when both the main feed water pumps are tripped or shut down with the auxiliary pump failed to be started. Under such circumstance, the water level within the reactor may be lowered to an emergency level at which the reactor has to be shut down or scrammed. However, the shutdown of the nuclear reactor (which means nothing but the shutdown of the nuclear power plant) due to abnormal condition of the feed water pumps must be avoided by all means. Once the scram occurs, an enormous time is required for restoring the plant to the normal operating condition. In this connection, it is theoretically known that the water level within the reactor is determined as a function of difference between the primary steam flow and the feed water flow. Accordingly, it will be possible to prevent the water level from being lowered to the emergency level by decreasing the primary steam flow rate, which means of course a corresponding decrease in the output power of the reactor (primary steam flow rate). Thus, it is not adequate to merely prevent the scram but desirable to be able to increase the power of the reactor as early as possible so that the nuclear power plant can be operated in a continuous manner even when the trip or partial shutdown occurs in the feed water pump system, because the reactor itself is in the normal condition. SUMMARY OF THE INVENTION An object of the invention is to provide a method of controlling a nuclear reactor so that the water level within the reactor can be held in a safe region while permitting the reactor to be operated at a permissible maximum power, even when one or two of the main feed water pumps are tripped or shutdown. In view of the above and other objects which will become more apparent as description proceeds, it is proposed according to an aspect of the invention that, in a nuclear power plant provided with a plurality of feed water pumps and recirculated flow control means for controlling adjustably the reactor power, the condition of the feed water pumps is detected to thereby determine the recirculated water flow in dependence on the detected pump condition, whereby the recirculated water flow is automatically controlled in accordance with the available pump capacity.
	abstract	A nuclear power reactor may include a plurality of power modules, each including a nuclear fuel and a power conversion system configured to convert heat generated from the nuclear fuel to electricity, where the nuclear fuel of the plurality of power modules collectively forms a reactor core. The nuclear power reactor may also include a sleeve being disposed between the plurality of power modules, where the sleeve has a first end and a second end opposite to the first end. The nuclear power reactor may further include a reactivity booster having a neutron source and a reactivity quencher having a neutron absorber. The reactivity booster may be movable between a first location adjacent the first end of the sleeve and a second location adjacent the reactor core, and the reactivity quencher may be movable between a third location adjacent the second end and the second location adjacent the reactor core.
046637745	abstract	An X-ray diagnostics installation has an X-ray tube for directing an X-ray beam at an examination subject disposed within the beam path on a support plate, the support plate having a secondary radiation grid on one side thereof, and having a receptacle therein for an X-ray film cassette. The support plate is mounted so as to be rotatable about an axis which is perpendicular to a central ray of the X-ray beam such that the secondary radiation grid can be optionally disposed in front of or behind the X-ray film cassette as viewed in the radiation direction, while the film focus distance remains the same.
	abstract	A phosphor powder for producing a high resolution phosphor screen and a phosphor screen are provided. The phosphor screen comprises a substrate, an infrared-absorbing layer, and a phosphor layer coated on the infrared-absorbing layer, wherein the phosphor layer comprises a phosphor powder in which a substantial amount of particles in the powder have a particle size as measured in the longest dimension of greater than 0 and less than about 5 microns. The phosphor screen may also comprise a black, infrared-absorbing substrate with the phosphor layer coated on the substrate. The phosphor powder is formed by preparing a phosphor composition, sintering the composition to form ingots, grinding the ingots to form a powder wherein a substantial amount of particles in the powder have a particle size of greater than 0 and less than about 5 microns, reactivating the powder in an inert atmosphere to a temperature of from about 500xc2x0 C. to about 550xc2x0 C., preparing a suspension of the powder and a nonreactive organic solvent, mixing the suspension to separate particles of the powder which partially fuse during reactivation and to break up large agglomerates of the powder in the suspension, decanting the suspension to at least partially separate the nonreactive inorganic solvent and the powder, and drying the powder to form the phosphor powder.
048287822	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the inspection of fuel assemblies in nuclear power plants and, more specifically, to an arrangement for inspecting fuel rods of boiled water reactors using an ultrasonic transducer which does not require removal of the fuel channel slip encasing the fuel rods. 2. Description of the Related Art Two types of nuclear fuel reactors are primarily used in the nuclear power industry: pressurized water reactors (PWR) and boiling water reactors (BWR). The nuclear fuel for either of these type of reactors is housed in fuel assemblies containing an array of fuel rods, each fuel rod in turn containing the fissionable material used to power the reactor. The fuel assemblies are submerged in circulating coolant water during operation, and are designed to efficiently transfer the generated fission heat to the coolant water. The conventional technique for testing leakage from the fuel rods of both PWR and BWR reactors is to sip the water surrounding the fuel rods, and then test the water for radioactivity. Three different types of sipping methods exist: wet sipping, dry sipping and a more recent technique of vacuum sipping, to which U.S. Pat. No. 4,034,599 assigned to General Electric Co., is directed. In a typical vacuum sipping cycle, the fuel assembly is inserted in an isolation chamber, which is then sealed and flushed with either demineralized or reclaimed water. In preparation for testing, a gas space is established above the fuel assembly by releasing a small quantity of water from the chamber while injecting air at the top at a slightly higher than pool pressure. This gas is then evacuated by a vacuum pump and passed through an in-line beta scintillation detector. Although vacuum sipping is faster than either wet or dry sipping because sample acquisition and laboratory analysis steps are eliminated, all three sipping techniques merely detect the presence of fissionable material which has leaked into the coolant water. None of the sipping techniques isolate a problem to a particular fuel rod, and thus all require subsequent removal and testing of all fuel rods in the assembly. In view of the above drawback, a subsidiary related to the assignee of the present application developed a failed fuel rod detection system for PWR reactors employing an ultrasonic transducer in the form of a two-fingered probe which traverses horizontally through each row of the rod array. The probe transmits an ultrasonic pulse and can detect and localize the presence of minor quantities of water inside each individual fuel rod from the received signal. The presence of water implies a through-wall defect in the particular fuel rod, indicating a fuel rod failure. This ultrasonic apparatus and technique, covered by U.S. Pat. No. 4,193,843 to Womack et al., issued Mar. 18, 1980, is much more accurate than sipping, and provides to the user a real time plot of the PWR fuel assembly showing the actual location and analysis of each of the rods. Thus, no fuel assembly dismantling is necessary for inspection/detection. Although the above-described failed fuel rod detection system could conceivably be used to inspect all types of light water reactor assemblies (both PWR and BWR), the rod assemblies of BWR reactors are typically encased with a fuel channel slip, and thus inaccessible to a horizontal probe. It is possible to remove the fuel channel slip of BWR reactors to test the fuel rods, but such disassembly is highly undesirable. As shown in FIG. 1, the only entrance aperture for testing the fuel rods in BWR assemblies without removing the slip is at the bottom, through the nosepiece 2. However, the nosepiece has only about a 31/2 inch diameter opening, thus partially obscuring direct access to the outer fuel rods from beneath. An inverted tripod extending across this opening in the nosepiece makes access to the individual fuel rods even more difficult. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide an arrangement for the inspection of BWR fuel rod assemblies in which the fuel rods are accessed through the nosepiece of the assembly. Another object of the invention is to provide an arrangement for inspecting BWR fuel rods which isolates a leakage problem to a particular fuel rod. A further object of the invention is to provide an arrangement for inspecting BWR fuel assemblies which is highly accurate and requires minimal re-inspection. These and other objects are achieved by the present invention, comprising an inspection arrangement for BWR fuel assemblies in which an ultrasonic transducer probe is inserted through the nosepiece in the shape of an inverted tripod located at the bottom of each fuel assembly. The fuel rods are supported in each assembly by upper and lower tie plates, in the form of grids. The grids have first apertures for supporting the ends of the fuel rods, and second apertures to allow coolant water to flow between the fuel rods during operation. The probe is inserted up through the second apertures of the lower tie plate grid. The probe may be successively inserted in each second apertures or, preferably, the probe may be inserted only in one of every four apertures and rotated to inspect each of the four surrounding fuel rods for the presence of water. In order to access the second apertures in the fuel assembly covered by the nosepiece, the probe is pivoted about a single pivot point, the pivot point being located in the trident of the nosepiece which is non-symmetric with respect to the array of fuel rods. The pivot point is preferably positioned centrally in the non-symmetric trident. However, the pivot point can also be positioned in any trident or below the tripod. Moreover, multiple single pivot points, one in each trident, can be provided to allow multiple inspections in parallel. In the preferred embodiment of the invention, the probe passes through a ball joint located at the single pivot point. In this arrangement, the lower end of the probe may be moved at its lower end by an x-y scanning bridge to cause the probe to pivot at the ball joint. In a second embodiment, the probe is seated in a two-stage goniometric cradle which, when rotated, cause the probe to pivot about the single pivot point. In a third embodiment, the probe is seated in a goniometric cradle disposed on a rotational table which, when rotated, causes the probe to pivot about the single pivot point. Other features and advantages of the invention are described below, with reference to the accompanying drawings.
047042486	abstract	A nuclear fuel element comprising an elongate block of refractory material having a generally regular polygonal cross section. The block includes parallel, spaced, first and second end surfaces. The first end surface has a peripheral sealing flange formed thereon while the second end surface has a peripheral sealing recess sized to receive the flange. A plurality of longitudinal first coolant passages are positioned inwardly of the flange and recess. Elongate fuel holes are separate from the coolant passages and disposed inwardly of the flange and the recess. The block is further provided with a plurality of peripheral second coolant passages in general alignment with the flange and the recess for flowing coolant. The block also includes two bypasses for each second passage. One bypass intersects the second passage adjacent to but spaced from the first end surface and intersects a first passage, while the other bypass intersects the second passage adjacent to but spaced from the second end surface and intersects a first passage so that coolant flowing through the second passages enters and exits the block through the associated first passages.
	claims	1. An inspection apparatus for inspecting weld zones in a reactor pressure vessel, the inspection apparatus inspecting a weld zone of a control rod drive housing and an area in proximity to the weld zone, the control rod drive housing being placed from the bottom of the reactor pressure vessel to an inside thereof, said inspection apparatus comprising:a probe configured to emit an ultrasonic wave toward the weld zone;a probe holding unit configured to hold the probe such that an ultrasonic wave transmitting surface of the probe is kept in direct contact with or at a constant distance from an outer surface of the reactor pressure vessel;a pressing unit configured to press the probe holding unit parallel to a central axis of the control rod drive housing against the reactor pressure vessel;a rotator configured to rotate the probe holding unit and the pressing unit around the central axis of the control rod drive housing;an elevator configured to move up and down the probe, the probe holding unit, the pressing unit, and the rotator along the control rod drive housing;wherein a size of the ultrasonic wave transmitting surface of the probe is set to a value selected from 35 mm to 120 mm in order that an effective focus area within a range of −6 dB from a maximum sound pressure height reaches an area to be inspected ranging from 150 mm to 200 mm. 2. The inspection apparatus according to claim 1, wherein:a longer side of the ultrasonic wave transmitting surface of the probe if said surface is rectangular-shaped, or a major axis of said surface if said surface is oval-shaped, is set to a value selected from 35 mm to 120 mm. 3. The inspection apparatus according to claim 1, wherein:the probe is configured to inspect a weld zone between the reactor pressure vessel and a structure inside a nuclear reactor, a built-up weld, and an inner-surface cladding portion of the reactor pressure vessel.
	description	This application claims the benefit of U.S. Provisional Patent Application Ser. No. 61/632,016, filed Jan. 17, 2012, entitled “Multi-fractional coal sorter” which is hereby incorporated by reference in its entirety. Be it known that we, Charles E. Roos, a citizen of the United States, residing at 2507 Ridgewood Drive, Nashville, Tenn. 37215, Edward J. Sommer, Jr., a citizen of the United States, residing at 5329 General Forrest Court, Nashville, TN 37215, and Charles D. Roos, a citizen of the United States, residing at 306 West 100th Street #31, New York, N.Y. 10025, have invented new and useful “Multi-Fractional Coal Sorter and Method of Use Thereof.” Run of mine (ROM) coal is used for energy production after it is processed. Wet processing can reduce the ash and sulfur content of the coal, but it wets the processed coal. There are many drawbacks to wet processing. For example, the liquid media requires treatment in a wastewater treatment facility. Coal fines and water produce sludge with environmental problems. Some processes use acids to remove contaminants and pollute water. The latent heat of water in wet coal reduces the recoverable energy from the combustion of coal by one to two percent. This reduction in useful energy increases the carbon footprint to produce electrical power. Some 95% of coal processing currently uses wet methods. A preferred method for processing ROM coal is dry processing. Generally, ROM coal includes a number of rocks of different sizes and compositions (fractions). Their sizes can range from small pebbles (for example 2 mm) to larger rocks (for example 70 mm and even greater) and their composition variable from silicates to iron pyrites. These rocks have a higher density than the coal. Dry processing does not create ecologically undesirable waste or require complex permits. The coal fines, which can be a substantial proportion of the fuel value in the coal, are not affected by dry processing. A commercially useful coal sorter must provide high throughput while it removes rocks and has minimal loss of fuel value. Accordingly, enhancing currently existing dry processing methods and devices is needed. The present invention discloses methods of sorting materials into multiple fractions. Multi-fractional coal sorting devices are also disclosed. The devices and disclosed methods use x-rays for sorting. In certain embodiments, a method of sorting materials into multiple fractions, includes providing a sample, receiving a collimated x-ray beam from an x-ray tube by a detector, determining measurements of x-ray absorption of pieces of the sample, identifying small and large pieces in the sample having a higher density than coal, sorting the large pieces from the sample by use of an air blast, and sorting the small pieces from the sample by use of a smaller air blast. Another embodiment of the invention further includes separating the sorted large pieces from ejected coal by use of a screen. Still another embodiment of the invention further includes separating the sorted small pieces from ejected coal by use of a screen. Another embodiment of the invention includes receiving the collimated x-ray beam by the detector is receiving a plurality of collimated x-ray beams. Yet another embodiment of the invention includes receiving the collimated x-ray beam by the detector which further includes receiving a first collimated x-ray beam from the x-ray tube by a first detector, and receiving a second collimated x-ray beam from the x-ray tube by a second detector. Yet another embodiment of the invention further includes determining identifying characteristics of the sample by use of an infrared 3D imager. Another embodiment of the invention is a multi-fractional coal sorting device, including an x-ray tube, wherein the x-ray tube is in a fixed position, a first collimator attached to the x-ray tube, a second collimator attached to the x-ray tube, a first x-ray detector, wherein the first x-ray detector is in a fixed position to receive x-rays collimated by the first collimator, a second x-ray detector, wherein the second x-ray detector is in a fixed position to receive x-rays collimated by the second collimator, a first microprocessor operationally connected to the first x-ray detector, a first sized ejector operationally connected to the first microprocessor, a second microprocessor operationally connected to the second x-ray detector, a second sized ejector operationally connected to the second microprocessor. Another embodiment of the invention includes the first x-ray detector and the second x-ray detector being under a sample stream, wherein the x-ray tube, the first collimator and the second collimator are positioned above the sample stream, so that collimated x-rays are received by the x-ray detectors. Yet another embodiment of the invention includes the first x-ray detector and the second x-ray detector being above a sample stream, wherein the x-ray tube, the first collimator and the second collimator are positioned under the sample stream, so that collimated x-rays are received by the x-ray detectors. Still another embodiment of the invention is one in which the x-ray tube, the first collimator and the second collimator are positioned above a sample stream. Yet another embodiment of the invention further includes a first collection bin positioned to receive sample ejected by the first sized ejector, a second collection bin positioned to receive sample ejected by the second sized ejector, a first screen attached within the first collection bin, wherein the first screen defines a plurality of openings so that smaller sized objects may pass through the openings, a second screen attached within the second collection bin, wherein the second screen defines a plurality of openings so that smaller sized objects may pass through the openings. In another embodiment of the invention the first collection bin and the second collection bin are combined. Still another embodiment of the invention further includes a third x-ray detector, wherein the third x-ray detector is in a fixed position to receive x-rays collimated by the third collimator, a third microprocessor operationally connected to the third x-ray detector. Certain other embodiments of the invention further include an infrared 3D imager positioned above a conveyor so that identifying characteristics of pieces of the sample on the conveyor are determined. Another embodiment of the invention is a method of sorting materials, measuring sorting efficiency and product purity including providing a sample, receiving a first collimated x-ray beam from an x-ray tube by a first detector, determining measurements of x-ray absorption of pieces of the sample, identifying the sizes of pieces in the sample having a higher density than coal, recording the size distribution from the first detector in a microprocessor, sorting the large pieces from the sample into a reject bin by use of an ejector, receiving a second collimated x-ray beam from the x-ray tube by a second detector located at an entrance to product bin, determining measurements of x-ray absorption of the remaining pieces of the sample, identifying the sizes of pieces in the sample having a higher density than coal, recording the size distribution of absorbing pieces detected by the second detector in the microprocessor, and determining ejections efficiency by comparing the two size distributions. Yet another embodiment of the invention further includes determining the number and sizes of absorbing pieces in the reject bin. Still another embodiment of the invention further includes positioning the first detector and the second detector under the sample stream. Yet another embodiment of the invention further includes protecting the first detector and the second detector with deflection plates. In another embodiment of the invention the sorting device further includes a deflection plate positioned so that the deflection plate defines a lower edge of the sample stream in order to protect each x-ray detector from direct contact with the sample stream. Yet another embodiment of the invention is a method of measuring sorting efficiency and product purity including providing a sample, receiving a collimated x-ray beam from an x-ray tube by a detector, determining measurements of x-ray absorption of pieces of the sample, identifying the sizes of pieces in the sample having a higher density than coal, recording the size distribution and number of items detected by the detector. Still another embodiment of the invention is a method of protecting x-ray detectors while processing coal and other minerals including providing an x-ray detector, providing a diamond coated deflection plate, and positioning the diamond coated deflection plate to protect the x-ray detector in order to deflect any sample items that may contact the x-ray detector during processing of coal and other minerals. Accordingly, one provision of the invention is to provide a method of sorting coal ore from contaminants into multiple fractions. Still another provision of the invention is to provide methods of using x-ray energies for multi-fractional sorting of materials. Yet another provision of the invention is to provide a multi-fractional coal sorting device. The present invention discloses a device 100 for and methods of sorting contaminants away from coal. The device 100 and methods disclose the use of specific x-ray energies to detect contaminants of different sizes so that such large and small materials may be effectively sorted away from other pieces of coal not having contaminants. Briefly, the device 100 disclosed herein includes an x-ray source 102 and collimators so that x-rays are collimated into a narrow fan which is directed at x-ray detectors. This narrow fan uses only a small portion of the total x-rays which are produced over a wide range of angles from the x-ray target. The present invention uses three different x-ray fans from the same tube. Each collimated x-ray beam hits a separate detector which can control separate air jets. This permits a strong air blast from a large air ejector to remove large rock and a much smaller air blast from a small ejector to remove small rock impurities. The methods disclosed herein include the steps of receiving collimated x-ray beams in order to determine x-ray absorption measurements of a sample, identifying small and large pieces of contaminant in the sample, and removing those small and large objects by the use of ejectors having different and appropriate force. The methods are performed at a very fast rate. The methods disclosed herein may be used to “clean” coal so that sulfur, mercury, and the like, are reduced when the coal is used at a coal burning power plant. There are several benefits from the use of methods of removing contaminants from coal in order to provide a cost effective dry method to significantly reduce the amount of contaminants (for example, sulfur) below the levels available with current washing techniques. For example, cleaner coal improves boiler performance by reducing slag and corrosion problems. Also the herein disclosed dry processing method reduces the amount of water used in processing coal for washing reducing requirements for waste water treatment. Further, the “clean” coal's higher heating value increases boiler capacity. Also, the total amount of ash is reduced and less sensible heat is lost to moisture and the bottom ash. Accordingly, the methods disclosed herein provide cost effective methods to remove contaminants from coal in a size fractionated manner. The present invention takes into account that larger rocks require stronger air blasts than needed for smaller rocks. Sizing the air blast to the size of the rock saves compressor power and reduces the loss of coal adjacent to the rock. By way of background, x-ray absorption in a material is a function of the density and atomic number of the material and it is also a function of the energy of the incident x-rays. A given piece of material will absorb x-rays to differing degrees depending upon the energy of the incident x-rays. Materials of differing atomic numbers will absorb x-rays differently. For example, materials having a higher atomic number will absorb x-rays much more readily than will materials having a lower atomic number. Also, the absorption profile of a given material over a range of x-ray energies will be different than the absorption profile of another material over that same range of energies. X-ray transmission through a material is given by the equation N(t)=N0e−ηρt, where N(t) is the number of photons remaining from an initial N0 photons after traveling through thickness t in a material of density ρ. The mass attenuation coefficient η n is a property of the given material and has a dependence upon photon energy. The value ηρ is referred to as the linear absorption coefficient (μ) for a given material. Values of the coefficient μ have been established by researchers to high accuracy for most materials and these values are dependent upon the energy of incident x-ray photons. Values of μ/ρ (=η) for most elements can be found at the National Institute of Standards and Technology (NIST) internet website. The lists of values are extensive covering all stable elements for various values of photon energy (for example, a kilo electron volt, abbreviated as KeV). The value of ρ for a given material is simply its density in gram/cm3 and can be found in many textbooks and also at the NIST website. The ratio N(t)/N0 is the transmittance of photons through a thickness t of material and is often given as a percentage, i.e. the percentage of photons transmitted through the material. In certain embodiments of the present invention, the device 100 for and method of sorting materials includes providing a sample, setting the detection thresholds, and sorting the sample according to the sorting parameters. Disclosed herein are the various embodiments for practicing the methods disclosed. By way of background, U.S. Patents for various x-ray measuring systems include 8,144,831 issued to Sommer, et al. on Mar. 27, 2012; 7,848,484 issued to Sommer, et al. on Dec. 7, 2010; 7,564,943 issued to Sommer, et al. on Jul. 21, 2009; 7,099,433 issued to Sommer, et al. on Aug. 29, 2006; RE36537 issued to Sommer et al. on Feb. 1, 2000; 5,738,224 issued to Sommer et al. on Apr. 14, 1998; 7,664,225 issued to Klein on Feb. 16, 2010; 6,338,305 issued to McHenry, et al. on Jan. 15, 2002; 7,542,873 issued to Vince, et al. on Jun. 2, 2009; 7,200,200 issued to Laurila, et al. on Apr. 3, 2007; 5,818,899 issued to Connolly, et al. on Oct. 6, 1998; 4,486,894 issued to Page, et al. on Dec. 4, 1984; 4,090,074 issued to Watt, et al. on May 16, 1978; 4,377,392 issued to Massey, et al. on Mar. 22, 1983, and U.S. Patent Application Publication No. 2010/0219109 to Roos, et al. on Sep. 2, 2010, each of which is hereby incorporated by reference in its entirety. As an initial matter, in certain embodiments, providing the sample may include providing run of mine ore from a coal mine. In other embodiments, the sample may be coal that has already been subjected to some cleaning method or procedure. In still other embodiments, the sample to be subjected to the methods disclosed herein may be any ore material containing a contaminant. By way of illustration, but not limitation, examples of contaminants include sulfur, mercury, silicates, carbonates, iron, calcium, aluminum, and the like. Further, mining ores are often silicates with metallic inclusions. The metallic inclusions have higher linear x-ray absorption coefficients. Accordingly, if gold ore is crushed, then the small gold inclusions could be detected and ejected by use of the present methods. Referring now to FIG. 1, there is shown an embodiment of the present invention. When referred to herein, x-ray source 102, means a source of x-rays, such as an x-ray tube, or the like, as known to those in the industry. The embodiment includes an x-ray source 102, first collimator 104 and second collimator 106 located under a sample stream 108 flying off conveyor 110 and clears splitter plates 112. The sample stream 108 may, in certain embodiments, also be referred to as a mineral or coal stream. Said collimators produce x-ray fans 114 and 116 that strike a first x-ray detector 118 and a second x-ray detector 120, respectively, which measure the absorption by the sample stream 108. Each x-ray detector 118 and 120 send signals to the first microprocessor 122 and second microprocessor 124, respectively, which communicate with and control the first sized ejector 126 and second sized ejector 128, respectively, that deflect selected objects from the sample stream 108 into bins 130 and 132. A structural support 148 is used to mount detectors 118 and 120, ejectors 126 and 128, microprocessors 122 and 124, as well as other equipment, such as a computer 146, as needed in a given embodiment. Those of ordinary skill in the art are familiar with suitable materials, and positioning of the same, as disclosed herein. Also shown are communication connections 150, such as data cables, and the like, as known to those of ordinary skill in the art, for the necessary electrical, data and information transfer between the various components. Throughout this application, it is understood that the necessary electrical, data and information transfer connections are in place between the various components whether or not such operational connections are shown in the figures. Further, given the schematic nature of the figures, such operational connections are understood to be represented. The first microprocessor 122 selects for ejection large dense rock in the sample stream 108 and the first sized ejector 126 deflects the selected item into bin 130 where it strikes a first screen 136 having openings 140 that allow undersized items to pass into bin 142 where they can be returned to the conveyor 110. It is understood that the first sized ejector 126 will deflect a mixture of rock and coal fines. The impacts of the rock on these screens cause vibrations which facilitate the separation of the coal fines from the ejected rock. The second collimated x-ray beam 116 strikes an x-ray detector 120 which detects x-ray absorption by the smaller rocks in the sample stream 108 and the second microprocessor 124 sends a signal to the second sized ejector 128 in order to deflect the smaller rocks into bin 132 and on to a screen 138, which has openings 140 sized to recover fine coal. The fine coal is collected in bin 144 and can be transferred to sorted coal bin 134. The data from the microprocessors 122 and 124 is analyzed by the computer 146 and is used to adjust and measure the performance of the device 100. Still referring to FIG. 1, as the sample stream 108 passes between the x-ray source 102 and the x-ray detector 118 the sample is irradiated. The x-ray detector 118 is operationally connected by connection 150 to a first microprocessor 122, also called a computer, which directs first sized ejector 126 to send contaminated coal to bin 130. As used herein, microprocessor is referring to a computer, or the like to serve the stated function. Coal that is not ejected is collected in bin 134. As previously disclosed herein, the microprocessor has software, or other means in order to perform the steps indicated herein. In certain embodiments, the determination may be as simple as material having an atomic number of greater than 6. Regarding the manufacture and use of collimators, methods are well known in the industry for making a suitable collimators as described herein. An example of a material which a collimator is made of is steel, having a thickness of 5 mm with an opening of a quarter inch through which the x-rays pass. In other embodiments, collimators may be manufactured of lead or brass and sized as needed. One of ordinary skill in the art is familiar with such collimators Use of a collimator with x-rays is beneficial because they reduce scattered x-rays. In the embodiments shown herein, the collimators are attached to the x-ray source 102 by bolts and also attach to framework or supports (not shown) of the collection bins. Alternatively, the collimators may be attached to x-ray tube 102 housing by means as known to those of skill in the art. A meaningful difference in force produced exists between the first sized ejector 126 and the second sized ejector 128. The air ejector force required to deflect a large rock is much greater than the air blast needed to remove the smaller objects. The air blast that can remove a large rock will also remove a portion of the coal that surrounds it. If the same air blast is used for all detected rock sizes there is too much loss of fuel product. Accordingly, smaller sized air ejectors in an air ejector array are spaced closer together than are larger ejectors and have an air blast profile that is smaller in area and smaller in force than the air blast profile from the larger ejector. Larger sized contaminates are ejected using the larger ejectors and smaller sized contaminates are ejected using the smaller ejectors. If only large ejectors spaced for example at 1 inch spacing are used and if the rock size selected for removal is limited to rocks greater than one inch, it can provide a 70% or greater removal of rock. This has substantial benefits in reducing the transportation costs to send ROM coal from the mine to a typical washing system located near a water source. The removal of large rock still gives a coal product with substantial impurities, which needs further processing. By way of background, coal typically has a specific gravity of 1.2 while the rock and heavier minerals have average values of 2.5. Still referring to the methods disclosed herein, after a decision is made that a contaminant is present and should be ejected, then next determination regards what amount of area needs to be ejected with the appropriate large or small ejectors selected to eject the contaminant. Some x-ray sensing devices have a capacity of 32 linear pixels per inch. Other x-ray sensing devices have a capacity of 64 linear pixels per inch. The ejection area size may be set based upon a required number of pixels detecting a contaminant. For example, if a device having 32 linear pixels per inch is in use which are read 32 times as the sample travels one inch in passing between the x-ray source 102 and the detector and it is desired to eject areas of one square inch, then it could be required that up to 1024 contiguous pixels would need to detect a contaminant in order for the air ejector to be triggered to take action. The number of contiguous pixel readings having reduced x-ray transmissions required to initiate a blast of air for ejection determines the minimum size of the ejected contaminant. The required pixel number is an adjustable perimeter within the method. With the example above, one of ordinary skill in the art may adjust the perimeter to their specific needs. Accordingly, if economic value is provided by removing smaller contaminant inclusions, then the methods disclosed herein may be used. In still other embodiments, the percentage transmission information is saved by the machine and used to normalize the voltage output of each pixel in the x-ray detector array. The pixel number and the percentage of the threshold are adjustable perimeters that can be set manually, or automatically in the x-ray measuring device. The smaller impurities often include iron pyrite which contains some 80% of the toxic mercury and other heavy metal sulfides. Removal of insoluble mercuric sulfite at the mine has substantial economic and environmental benefits. The present invention may provide the benefits of washed coal at higher speeds without water. It can permit the production of clean coal directly from ROM coal at the mine. Rock impurities are not contaminated by the heavy media and the dry process can free operators from the requirement of trucking rock which is expensive. Referring now to FIG. 2, there is shown an embodiment of the device 100 having an x-ray source 102 and collimator 104 located under a sample stream 108 that clears splitter plate 112. The collimated x-ray fan 114 strikes the first x-ray detectors 118 which detect the absorption by the larger rocks in the sample stream 108. The first x-ray detector 118 sends a signal to the first sized ejectors 126 that deflect the rocks into bin 130 and on to a screen 136 (shown in phantom lines) with openings 140 sized to recover small sizes. FIG. 3 shows an embodiment of the present invention in which the x-ray source 102, first collimator 104 and second collimator 106 are located above the sample stream 108 that flies off a conveyor 110. The first collimated x-ray fan 114 strikes the first x-ray detector 118 which detects the absorption by the larger rocks in the sample stream 108. The first x-ray detector 118 sends a signal to the first sized ejector 126 for large rocks that deflects the rocks into bin 130 and on to a first screen 136 having openings 140 sized to recover small sizes. The second collimated x-ray fan 116 strikes a second x-ray detector 120 which detects the absorption by the smaller rocks in the sample stream 108 and sends a signal to the second sized ejector 128 for small rocks that deflects said rocks into bin 132 and on to a second screen 138 with openings 140 sized to recover fine coal. The coal in the sample stream 108 passes over the splitter plate 112 and is collected in bin 134. Also shown is computer 146 which is operationally connected to the microprocessors 122 and 124 (not shown) for recording data and performing other functions as disclosed herein. For the understanding of the Examiner, Applicants note that the first microprocessor 122 of the first x-ray detector 118 is located at the same location that the detector 118 is shown. Also, the second microprocessor 124 of the second x-ray detector 120 is located at the location that the detector 120 is shown. Electrical and data communication connections which exist between the computer 146, detectors 118 and 120 and ejectors 126 and 128 are not shown. Still referring to FIG. 3, there is shown an embodiment of a device 100 for practicing the methods disclosed herein further including a deflection plate 152. This plate 152 is shown in detail in FIG. 11. In certain embodiments of the invention, the diamond coated deflection plate 152 includes a bar, also called a body, that is bolted to the frame of the device 100 and a diamond coated plastic film or diamond coated metal foil that lies over an x-ray detector window 158 shown in FIG. 11. The body of the deflection plate 152 acts to prevent an item in the sample stream 108 from contacting a detector, such as detector 118. In certain embodiments, a flat body shape in alignment with the flow of the sample stream 108 is desired. The deflection plate 152 functions to protect the x-ray detectors 118 and 120, electronics, or other equipment that may be positioned under, or beneath, the sample stream 108. That is, the deflection plate 152 establishes the lower boundary of the sample stream 108, so that items under that lower boundary are not inadvertently struck by something in the sample stream 108. In certain embodiments, the deflection plate 152 is a steel plate with a heat treated diamond coating. The deflection plate 152 protects the detector window 158 (shown in detail in FIG. 11) of the x-ray detector 118 from the coal stream 108. In another embodiment the deflection plate 152 includes a diamond coated steel bar and a diamond coated plastic window. The deflection plate 152 defines the lower portion of the coal stream 108. It protects the window 158 while allowing x-rays 114 to pass through the diamond coated plastic window 154 portion of the deflection plate 152 and window 158 to interact with the detector 118. The use of vapor deposited carbon to provide a diamond coating is well known to those skilled in the art and such services are readily commercially available. It is even used to coat skillets and other cookware. This coating is abrasion resistant and has x-ray absorption comparable to coal. It has better abrasion resistance than tool steel. In other embodiments, the deflection plate 152 may be constructed of material suitable for the function disclosed herein, such as tool steel or ceramics. The deflection plate 152 permits the sensitive portions of the x-ray detector 118 to withstand the bombardment by portions of the sample stream 108. It also will allow placement of the detector 118 on the edge of the coal stream 108 for a reduction in required x-ray power and an increase in the signal to noise levels. Several different embodiments of the present invention are shown. Two different orientations of the x-ray beams are shown in FIGS. 1 and 3. The larger rock sizes are first removed from the sample stream 108 and then a second set of x-ray detectors 120 control the second sized ejectors 128 to remove the smaller impurities. FIG. 2 shows the x-ray fan from an x-ray tube 102 mounted under the sample stream 108 from a high speed conveyor 110. FIG. 3 shows the x-ray tube 102 mounted above the sample stream 108. Referring now to FIG. 4, there is shown an embodiment of the present invention in which an x-ray source 102 and a single collimator 104 produce the collimated x-ray fans 114, 116, 117 which are located over a sample stream 108 flying off conveyor 110. In certain embodiments, the collimator 104 is a single closed structure, except for the openings which allow passage of x-ray beams. With such a construction, other items, such as a computer 146, may be placed inside of the structure. The collimated x-ray fans 114, 116, 117 strike the x-ray detectors 118, 120, 121, respectively, which detect the absorption by items in the sample stream 108. The x-ray detectors 118, 120, 121 send measurements to microprocessors 122, 124, 125. Those microprocessors 122, 124, 125 are at the same location in schematic FIG. 4 as their corresponding detectors 118, 120, 121, respectively. In alternate embodiments, a system of processors (not shown) may be used. The microprocessors, or system of processors process the measurements and send signals to the first sized ejector 126, and second sized ejector 128, respectively, that deflect appropriately sized items into bins 130, 132, and 134. The computer 146 uses the detected signals to measure the number and size of each detected particle in the three bins 130, 132 and 134. In the same manner that the screens are used above, within bin 130 is a first screen 136, and within bin 132 is a second screen 138, each having openings 140 sized to recover small sizes. Still referring to FIG. 4, there are three openings in the collimator 104 housing near the x-ray tube 102 and a second set of three openings in the collimator 104 housing at the base which holds the ejectors 126 and 128. These openings are all parts of the collimator 104 and x-ray beams 114, 116, 117 are shown to pass through them. FIG. 5A-5B shows an embodiment of the first screen 136 and the second screen 138. In FIG. 5A the screen 136 with the smaller openings 140 recovers the fine coal ejected with the smaller rocks and reduces product loss. In FIG. 5B the recovery screen 138 with openings 140 sized to recover smaller rocks and coal ejected with the large rocks. The openings 140 allow smaller rock and coal to pass therethrough, which can be returned to the sample stream 108 for further separation. Referring now to FIG. 6, there is shown an embodiment of the present invention in which a single x-ray detector 118 operates both a first sized ejector 126 and a second sized ejector 128. Shown therein is an x-ray source 102 and first collimator 104 located under a sample stream 108 flying off the conveyor 110 and clearing splitter plates 112. The first collimator 104 produces an x-ray fan 114 that strikes a first x-ray detector 118 which measures the absorption by the sample stream 108. The first x-ray detector 118 sends signals to the first microprocessor 122, which is at the same location in the Figure as the detector 118, which controls the first sized ejectors 126 and the second sized ejectors 128 that deflect selected objects in the sample stream 108 into bins 130 and 132, respectively. Also shown are communication connections 150 and a support structure 148 to which detectors 118, ejectors 126, 128, or the like, may be attached. Specifically, the first microprocessor 122 selects for ejection of large dense rock in the sample stream 108 and the first sized ejector 126 deflects the selected items into bin 130 where they strikes the first screen 136 with openings 140 that pass undersized items into bin 142 where they can be returned to the conveyor 110. Also, the first microprocessor 122 selects smaller rocks for ejection and sends a signal to the second sized ejector 128 that deflects said rocks into bin 132 and on to a second screen 138 with openings 140 sized to recover fine coal. The fine coal is collected in bin 144 and can be transferred to the sorted coal bin 134. An alternate to the embodiment shown in FIG. 6 could combine the multiple x-ray beams 114, 116 and 117 of FIG. 4 into a single x-ray beam so that only a single x-ray beam 114 exists. Similarly, an alternate embodiment could combine the multiple detectors 118, 120 and 121 into a single detector array—say 118, all connected through a single processor to ejectors 126 and 128 (which remain separated as shown in FIG. 4) to eject rocks of various and selected sizes into separated bins 130, 132 or 134 as appropriate. FIG. 7 shows an embodiment of the present invention for inspection of a sample. Shown therein is an x-ray source 102 and first collimator 104 are located under a sample stream 108 flying off conveyor 110. The first collimator 104 produces x-ray fans 114 that strike a first x-ray detector 118 which measure the absorption by the sample stream 108. The first x-ray detector 118 sends signals to the first microprocessor 122 which transmit to computer 146 the sizes and number of items detected by the detector 118. The coal and rock are collected in bin 134. As used herein, in certain embodiments, a bin 130, 132, 134 (FIG. 6) may also allow for placement on a conveyor belt. In this embodiment, it may be desired to collect the sample stream 108 on another conveyor. Shown in FIG. 9 are the linear absorption coefficients from the National Institute of Standards and Technology (NIST) mass absorption coefficients (μ) for iron pyrite (FeS), coal, and silicon dioxide (SiO2) over a range of x-ray energies. Also shown are their densities. Note that coal is a mixture of carbon and hydrocarbons and there is no NIST “standard” for coal. Accordingly, the x-ray absorption coefficients of coal are the NIST data for graphite corrected for coal density of 1.2 grams per cubic centimeter (g/cc). As shown elsewhere herein, the absorption by coal is much less than the absorption of pyrite in silicates for 8 to 20 kilo electron volts (KeV) x-rays. Using the information in FIG. 9 illustrates how a contaminant can be differentiated from coal. For example, it is calculated that use of x-ray energy at a level of 15 KeV results in a 56.6% transmission through coal having a thickness of 1 cm, while contaminants having a thickness of only 1 mm have reduced transmission percentages of 0% (for FeS), and 20.5% (for SiO2). By way of a second example, it is calculated that use of x-rays at an energy level of 20 KeV for which coal having a thickness of 1 cm has a transmission percentage of 73.2%, as compared to contaminants such as FeS and SiO2 which have transmission percentages of 0% and 50%, respectively. In certain embodiments of the present invention, the range of x-ray energies used is dependent upon the thickness of the sample stream 108. In certain embodiments, the range of x-ray energies may be from about 6 KeV to about 100 KeV. In other embodiments, the x-ray energies may be in the range of from about 8 KeV to about 20 KeV. In still other embodiments, the range of x-ray energies may be from about 50 KeV to about 100 KeV. In still other embodiments, the range of x-ray energies is above the absorption edge of the ejected element. Various devices may be appropriate to supply the x-ray energies and x-ray detectors used in the methods disclosed herein. In certain embodiments of the present invention, such a device may be the TruSort machine, second generation, commercially available from National Recovery Technologies, LLC of Nashville, Tenn. In other embodiments, an appropriate x-ray device is available from Commodas Mining GmbH at Feldstrasse 128, 22880 Wedel, Hamburg, Germany, and is called the CommodasUltrasort. It uses dual-energy detection algorithms similar to airport baggage scanners. In still other embodiments, an appropriate x-ray sensing device may be model no. DXRT which is commercially available from National Recovery Technologies, LLC of Nashville, Tenn. The x-ray sensing machine may be a dual energy device. In other embodiments of the present invention, the x-ray device may be a broadband x-ray device such as the vinyl cycle model, which is commercially available from National Recovery Technologies, LLC of Nashville, Tenn. FIG. 10 shows the schematic algorithm in one embodiment of the present invention which uses the number of detector pixels which detect absorption of x-rays above and below the preset threshold. This algorithm uses the ratio of total number of pixels during the recycle time of the x-ray detector array and the number of pixels reporting x-ray intensities below said threshold. Accordingly, in a certain embodiment, the pixel density is 32/inch and there are 1024 pixels readings while the coal passes one inch over the detector. The coal leaves the conveyor belt at 120 inches/sec and the x-ray detector (118, 120 or 121) array is read and reset 32 times during one inch of travel of the coal stream. If the air ejectors (126 or 128) are one inch apart each jet is controlled by the adjacent detectors providing 32 intensity measurements to the computer 146 each time it reads and resets the detector array. The operator can set the detector to only report readings that are less than a preset amount and indicate the presence of objects which absorb more than coal. The computer 146 can collect and analyze the data it collects and adjust the amount of ejected air. Larger items require more air than smaller items. The number of pixels reporting higher absorption is a measure of the size of the object and the amount of air required to eject it. Regarding FIG. 10, the operator inputs items 1000 and 1001. When in use, the x-ray detector 1002 reads sample and asks is the pixel signal below threshold 1003. If yes, then the computer 146 records information 1004. If a large object 1006 is detected, then a large air blast is provided by firing ejector 1005 and further information 1007 is stored. If a small object 1008 is detected, then a small air blast is provided by firing ejector 1009 and further information 1007 is stored. FIG. 11 shows the deflection plate 152 referenced in FIG. 3 in larger detail. It shows that the deflection plate 152 is composed of a mechanical bar that prevents larger items in the coal stream 108 from striking the detector 118 and a diamond coated cover 154, also called a window, for the x-ray window 158 of the detector 118. In certain embodiments, the deflection plate 152 and the diamond coated cover 154 are attached to the x-ray detector 118 and the diamond coated plastic window 154 portion of the deflection plate 152 covers the x-ray window 158 of the detector 118. In other embodiments, the deflection plate 152 is attached to a support structure 148. FIG. 12 shows the data flow from x-ray detectors 118, 120 and 121 to the computer 146 shown in FIG. 4. The computer 146 collects, processes and stores the x-ray data. Detector 120 triggers ejector 128 and detector 118 triggers ejector 126 to fire on selected signals. The computer 146 measures the size distribution and amount of the material ejected into bin 130 using the difference in the signals from detectors 118 and 120. The computer 146 also detects the ejection efficiency with the ratio of ejected items and the number of the triggers to ejector 126. The computer 146 also uses the difference between detectors 121 and 120 to determine the sizes and number of ejected items by ejector 128. The computer 146 provides the size and number of detected particles in each bin 130, 132 and 134. This data allows the threshold settings of the detectors to be set for higher purity and lower product loss and it provides a measure of the sorting efficiency of the device 100. In certain embodiments of the method, the detector threshold can be defined as a percent (for example 80%) of the signal voltage from the thickest regions of the sample of coal, without any inclusions of contaminants. The ejection threshold is then set as a percentage of pixel readings during the measurement cycle that have signals less than the detector threshold. The number of pixel signals with levels less than the threshold sets the minimum size of the ejected contaminate. A detector with 25 pixels/cm can detect 0.4 mm objects. Ejecting on a single low pixel reading could reduce contaminates to 100 ppm, but the product loss makes this impurity level impractical for coal. While ejection on a pixel may be useful for extracting gold from base rock, a more typical requirement for coal could be 250 pixels with low signals out of the typical 625 pixel signals per square cm of the sample. In certain embodiments, the use of dual energy detectors permits determination of relative composition independent of coal thickness. In certain embodiments of the present invention, a complex pattern of matching size measurements of the coal sample is not needed, although it is preferred that the pieces of the sample have sizes less than the average bed depth of the coal sample. Stated another way, the methods disclosed herein operate to identify materials by differences in x-ray absorption and reliably provide signals to rapid ejection mechanisms. With regard to determining an ejection threshold, applicants note that ejection is just one of several appropriate methods of physically separating pieces of the sample. In certain embodiments of the present invention, separation may occur by use of an array of air ejectors, as further described herein. In still other embodiments of the present invention, separation may occur by pushing, moving, or otherwise, thrusting a piece of sample which has reached an ejection threshold so that it is physically separated from a piece of sample which has not reached the ejection threshold. Such pushing or moving may occur by use of fast acting pistons, mechanical levers, or flippers. One of ordinary skill in the art is familiar with various arms, hydraulics, or the like which may be used to physically move a piece of sample which has reached the ejection threshold. The present invention, as described herein, has recordable devices, such as microprocessors, controllers, computers, or the like, in order to allow the machines to make determinations and perform functions. One of ordinary skill in the art is familiar with adjusting, manipulating, or programming such devices in order to achieve the methods set forth herein. By way of example, the DXRT model commercially available from National Recovery Technologies, LLC of Nashville, Tenn., is programmable such that ejection thresholds may be set. In this example, the DXRT machine calculates position and timing information for arrival of the piece of sample at the air ejection array needed to accurately energize downstream ejector mechanisms in the air ejection array and issues the necessary commands at the right time to energize the appropriate ejectors to eject the piece of sample having a contaminant from the flow of other pieces of sample which do not have a contaminant. Accordingly, pieces of sample having sufficiently high percent transmissions are not ejected by the air ejection array. In alternate embodiments, the machine may be set such that the opposite is true. That is, sample containing no contaminants are ejected and pieces of sample containing contaminants are not ejected. Those of ordinary skill in the art recognize that such alterations to the methods disclosed herein may be performed. One of ordinary skill in the art is familiar with the manner of operationally connecting components in detection systems as disclosed herein. All such wires, cables, and the like, needed for such operational connectivity are well known in the art and generally commercially available. Regarding each component of the present invention disclosed herein, operational connectivity includes any connections necessary for power, data or information transfer, or the like, for the operation of the specific device. One of ordinary skill in the art is familiar with such types of connections. Note that U.S. Pat. No. 6,610,981 issued to Sommer, Jr. on Aug. 26, 2003 is hereby incorporated by reference in its entirety. Use of Infrared 3D Imaging with a Coal Sorter The present invention includes an embodiment in which infrared 3D imaging enhances the efficiency of the coal sorter. By way of introduction, adding infrared 3D imaging to electromagnetic radiation material separation can greatly improve the separation efficiency and the throughput of the separation process. An embodiment of the present invention includes an infrared 3D imager 156 to track the position of each discrete piece of material being separated from the time it is identified using an electromagnetic radiation source 102 and detector 118 to the time it has arrived on the correct chute or conveyor. By including the 3D imager 156 with the coal sorter disclosed herein, the invention can verify correct separation of pieces, which depend upon the pieces maintaining predictable vectors of motion. Such invention can also measure the thickness of every piece being separated. This allows accurate separation decisions on a wide range of materials using measurements of single-energy x-rays, materials which before would have required the more costly and complicated measurement of x-rays of multiple energies. Referring now to FIG. 8, there is shown an embodiment of the device 100 of the present invention including an infrared 3D imager 156. As shown therein, the infrared 3D imager 156 is positioned above the end of the conveyor 110 on which the sample stream 108 travels. The infrared 3D imager 156 is operationally connected to computer 146 by communication connections 150 so that information generated by the infrared 3D imager 156 is provided to the computer 146. That information includes geometry and motion information, such as position, velocity, direction of travel, acceleration, rotation, thickness, size, shape and orientation of pieces both before and after ejection. The computer 146 then controls the ejectors 126, as disclosed above, so that such additional information (i.e., shape, thickness, rotation, acceleration, velocity, direction of travel, etc.) is used to more efficiently separate the sample. The computer 146 is also receiving information and data from x-ray detector 118 through microprocessor 122 so that the x-ray detection information and 3D imager information are used in combination. In certain embodiments, the infrared 3D imager 156 is used with known electromagnetic radiation sorters. In other embodiments, the infrared 3D imager 156 is used with non-collimated x-ray beams. Infrared 3D Imagers 156 are known in the art and readily commercially available. For example, an infrared 3D imager 156 may be purchased from Primesense in Tel Aviv, Israel. By way of background, an infrared 3D imager 156 illuminates the pieces of sample with continuously-projected, infrared structured light. By reading the infrared reflections with a CMOS sensor and calculating a dynamic, 3D representation of the material from the CMOS data using parallel computational logic, it may report the position, velocity, direction of travel, acceleration, rotation, size, shape, orientation and thickness of the pieces of material in the 3D representation or any combination of these parameters, as well as the results of calculations based on those parameters. Such information can improve the throughput and/or improve the separation efficiency and/or lower the operating cost of the separation process. It is an unexpected benefit to have this further information. These position, shape, and size measurements mean the present invention equipment can operate at a higher capacity as the sample pieces can be in motion on vectors distinct from the motion vector of the conveyor. Further, the conveyor 110 density can be higher than normal as it is not as necessary to avoid collisions between the sample pieces. The size and shape measurements mean the power requirements of the separation process can be less as the intensity of the physical separation can be varied according to the size and shape of the sample piece. The thickness measurements mean the invention can report the thickness of the sample pieces at the point of identification to allow x-rays of a single energy to provide more information than is currently possible by simply measuring the x-ray absorption alone. In sum, all of these measurements can be made from before or at the point the sample pieces are examined by electromagnetic radiation to the time the sample pieces have definitely and finally passed through the sorter system and are in the collection bins, or on the chutes or conveyors for the rejected fraction or the collection bins, chutes or conveyors for the accepted fraction. In certain embodiments of the present invention, when the infrared 3D imager 156 is tracking individual sample pieces, the following algorithm may be put in use: 1. At the point the sorter system makes the determination to keep or reject, with reject defined as the decision to employ the physical separation technology, e.g. air ejectors or other means, the shape, size and position of every sample piece designated for rejection is recorded. 2. The present invention tracks, in real time, sampling as often as is practical given the speed of the electronics of the day and given the speed of the sample material passing through the system, the position of the sample pieces designated for rejection as they moved towards the physical separation technology. The size, shape and previous position of the sample pieces uniquely identify each sample piece. In an alternate embodiment, the present invention may calculate the speed, direction of travel, acceleration or rotation of the sample pieces. 3. At the moment a sample piece designated for rejection arrives at the physical separation technology, the present invention triggers, or causes to be triggered, the physical separation technology at the optimum position given the position of the sample piece. In alternate embodiments, this decision could also be informed by the shape or size of the sample piece or the motion of the sample piece. 4. Because the size of each sample piece is mapped, the intensity of the physical separation according may vary according to the size, shape or orientation of the sample piece. For example, in the case of pneumatic separation, big sample pieces to be deflected would get more air. 5. The position of each sample piece marked for rejection continues to be tracked until it crosses a threshold marking it as definitely and finally having landed in a collection bin, or on the chute or conveyor carrying the rejected fraction, or another threshold marking it as definitely and finally being mis-classified and landing in a collection bin, or on the chute or conveyor carrying the accepted fraction. Optionally, data on the incidence of misclassification may be recorded and maintained. Also, optionally, such data may be used to vary, or cause to be varied, the speed of the conveyor feeding the sorter system or the intensity of the physical separation technology, or other appropriate parameters in an effort to minimize misclassification. 6. In the cases when the sample pieces marked for rejection missed the intended bins or chutes and bounced back into the mixed sample stream, the present invention maintains surveillance of the sample piece from steps 2-5 until the sample piece crosses a threshold marking it as definitely and finally out of the surveillance. In other embodiments of the present invention, such as in the case of estimating the thickness of the individual sample pieces at the point of identification to allow x-rays of a single energy to provide more information, the following algorithm is of use: 1. Prior to operation, the infrared 3D imager 156 is calibrated with objects of known thickness at the point of x-ray identification. This calibration informs the infrared 3D imager 156 with the data required to report an accurate measurement of the object's thickness. 2. At the point the present invention makes the identification with single-energy x-rays, the infrared 3D imager 156 reports the thickness of the object at the point on the object through which the x-rays pass. This thickness datum combined with the datum of the x-ray absorption from the x-ray detector is used to make, or cause to be make, the decision to accept or reject the sample piece. 3. Optionally, as the sample pieces move towards the physical separation technology, the infrared 3D imager 156 re-samples the thickness estimate and revisits the decision to accept or reject. This step corrects for inaccuracy caused by an estimate of thickness from a single angle. In this case, all objects would be tracked through the infrared 3D imager 156, or at least all about which there had been some accept/reject ambiguity, and not just those marked for rejection. The two algorithms are not exclusive. Both can and likely would be simultaneously used in many embodiments of the present invention. A person of ordinary skill in the art would, with software based on the above algorithms, be enabled to make an embodiment of this invention with a 30 Hz sampling rate using Microsoft's Kinect Controller for Xbox, which is readily commercially available. As known to one of ordinary skill in the art, an embodiment of the present invention with higher sampling rates and a higher-resolution pattern of structured infrared light could be made with software based on the above algorithms and Primesense's PS1080 system on a chip, their PrimeSensor Reference Design and their NITE middleware software, all of which are readily commercially available. As shown in FIG. 1, ROM coal is placed on a fast conveyor 110 that gives the coal sufficient velocity to clear the splitter plates 112 and land in bin 134. The first x-ray detector 118 measures the absorption of the ROM coal stream. Again, the detectors may be located at the end of the conveyor 110 as shown in FIG. 1 or located under the belt of the conveyor 110 (not shown). The absorption signals are sent from the first detector 118 to the first microprocessor 122 which selects dense rocks with sizes greater than 0.75 inch and signals first sized ejector 126 to remove the selected rock from the ROM coal stream. The first sized ejector 126 has sufficient force to deflect heavy rocks into bin 130. The forceful ejection will also eject any coal adjacent to the heavy rock and screen 136 has 0.5 inch openings 140 that allow smaller pieces of rock and coal to pass into bin 142. The rock and coal in bin 142 can be returned to conveyor 110. The remaining small rocks and coal pass by the second detector 120 which measures absorption of the x-ray beam 116 by the sample. The second detector 120 is operationally connected to the second microprocessor 124, which may be set to select rock over 0.4 inch and signal the second sized ejector 128 to deflect such rock it into bin 132. The deflection of small sized rock does not require the force needed to deflect the more massive large rock and the use of smaller more numerous ejectors limit the amount of coal inadvertently ejected with the smaller rock. The smaller rock deflected into bin 132 falls on screen 138 which has 0.25 inch openings 140 that allow passage of the smaller sized items into bin 144, where such smaller items can be mixed with the coal product in bin 134. Obviously, sample that is not deflected has the velocity and projection to land in bin 134. While the preferred embodiment uses multiple x-ray beams from the same x-ray tube 102, in alternate embodiments, multiple x-ray beams can also be obtained by the use of two or more x-ray tubes 102. As shown in FIG. 6, the present invention combines certain features of the previously described embodiments herein to take advantage of the ability of a fine pitch x-ray detector to simultaneously measure absorption of x-rays passing through both large rocks and small rocks. That is, a single first collimator 104 provides an x-ray fan 114 to the first x-ray detector 118. The first x-ray detector 118 provides signals to the microprocessor 122 (not shown, but at same location as the detector 118). Importantly, the first sized ejectors 126 and the second sized ejectors 128 are not combined and remain positioned as shown with one downstream from the other. The x-ray fan 114 passes through the mineral or coal stream 108 and is detected and measured by detector 118. The measurements are processed by microprocessor 122 which runs an algorithm (not shown) that analyzes the measurements and distinguishes large rocks from smaller rocks while distinguishing both from the surrounding coal bed. Microprocessor 122 then signals ejectors 126 or 128 as appropriate to eject smaller rocks into bins 132 and larger rocks into bin 130. In effect this embodiment utilizes a single x-ray beam analysis system connected with a dual ejection system. FIG. 4 shows a device 100 with three x-ray beams 114, 116 and 117 from the same x-ray tube 102 that sorts large and small sized impurities into separate collection bins and uses the data from the three x-ray detectors 118, 120 and 121 to measure the number and size of detected objects in the product and reject bins. This embodiment provides the machine operator with running estimate of product loss and impurity. It also measures the ejection efficiency with the ratio of ejector triggers to items removed from the coal stream. The ejection efficiency data allows the machine operator to adjust the air amount to eject the impurity with minimal loss of coal. As shown in FIG. 7, the present invention operates as an inspector without sorting. Specifically, the microprocessor 122 uses the absorption data from first x-ray detector 118 to determine the number and sizes of higher density rock in the coal sample and records this data in computer 146. Such information is useful as the price of coal is often determined by the rock content in the coal. The present invention makes it possible for power utilities to quickly sample larger quantities of coal. All references, publications, and patents disclosed herein are expressly incorporated by reference. Thus, it is seen that the methods and devices of the present invention readily achieve the ends and advantages mentioned as well as those inherent therein. While certain preferred embodiments of the invention have been illustrated and described for purposes of the present disclosure, numerous changes to the devices and methods may be made by those skilled in the art, which changes are encompassed within the scope and spirit of the present invention as defined by the following claims.
047160128	abstract	An upwardly convex resilient annular strainer, fixed to the secondary core support base plate in a PWR, bends to remain in contact with the lower hemispherical head section of the pressure vessel as the gap between the internals and the lower hemispherical head section varies in size with reactor temperature to prevent debris from lodging in the gap where it could lead to reactor damage.
050330730	summary	I. FIELD OF THE INVENTION This invention relates generally to a system and related method for inspecting an object, and more specifically, a system and related method for radiographically inspecting a relatively stationary object positioned at a selected location to generate a radiographic image of the object. II. BACKGROUND INFORMATION Known systems for performing radiographic inspection of an object include systems which utilize X-ray film. Such inspection systems generally produce high resolution images of the object being inspected, but present a disadvantage in that such systems perform radiographic reproduction over a limited dynamic range. The high resolution of the X-ray film results from the fact that individual silver bromide crystals, which make up the emulsion layer of the X-ray film and which are activated by incident X-rays to create a "picture," are of relatively small size. The dynamic range, on the other hand, is determined by the number of silver bromide crystals per unit area in the emulsion layer that can be exposed to the X-ray radiation. The dynamic range of a given film is limited by the thickness of the silver bromide crystal emulsion that, when developed, will not allow detectable light to be transmitted. The use of higher density film having high dynamic range necessitates the use of a very intense light source for reading the X-ray piCture. This requirement for a very intense light for reading the X-ray film presents a significant disadvantage as to the use of such film. Also, the use of X-ray film for radiographic inspection does not permit compensation for the effects of radiation scattered by the object being inspected. To increase the dynamic range for radiographic inspection systems and to permit compensation for radiation scatter, radiation detectors that produce a linear electronic signal response, when exposed to a wide range of X-ray intensities, in conjunction with electronic processing of the signals with wide range linear amplifiers are used. An example of a known radiographic inspection system, which utilizes electronic detectors, is a "linear flying spot scanner" of the type commonly used for inspecting luggage at airports. This linear flying spot scanner includes a radiation source, a stationary shield, a rotating beam-chopper wheel and electronic detectors. The radiation source transmits a beam of radiation along a path toward a selected location at which an object is positioned. The stationary shield, rotating beam-chopper wheel and the electronic detectors are disposed in the radiation path, with the stationary slit and beam-chopper wheel positioned between the radiation source and the electronic detectors. The stationary shield and beam-chopper wheel are aligned to select a portion of the radiation beam from the source. This portion of the radiation beam is transmitted along the radiation path to a single portion of the selected location at which the object is positioned and to a corresponding selected portion or area of the object. Radiation transmitted to the selected area of the object interacts with the selected area. A portion of the radiation interactive with the area of the object is compton scattered in all directions. Another portion is absorbed by the photoelectric interaction in the object and re-emitted in all directions as fluorescent radiation, while yet another portion passes through the object. X-ray sensitive electronic detectors, disposed between the radiation source and the object, generate signals in response to the fluorescent and compton scattered radiation which is scattered in the backward direction by the selected area of the object, and through-radiation detectors, positioned such that the object is between the radiation source and the through-radiation detectors, generate response signals for radiation passing through the selected area of the object. Response signals generated by the electronic detectors are processed to obtain a data representation of the selected area of the object. The signal processing is performed by a data processor. An image of the object area is obtained using the processed signal data, and using a visual display means which is responsive to the processed data, the image of the selected area of the object is displayed. As described above, the stationary shield and beam chopper wheel select a single line portion of the selected location and thereby allow radiation to be transmitted to just a corresponding selected area of the object. By limiting the portion of the object to which radiation is transmitted, the exposure of the object to radiation is limited. The number of detectors required to generate response signals is also minimized by limiting the radiation transmitted to the object and, thus, just a few large area detectors are utilized. The resolution of such a system is determined by the size of the slits in the stationary shield and the rotating chopper wheel. However, because the linear flying spot scanner only operates to select a single line portion of the selected location and a single corresponding object area, the necessity exists to move the object through the selected inspection location in order to radiographically inspect area of the object other than the single line selected area. Such movement places successive line areas of the object in the selected location to which radiation is transmitted and thereby permits a radiographic image of each line area of the object to be obtained. The requirement of moving the object relative to the linear flying spot scanner is a major disadvantage in that, in order to move the object, specialized moving apparatus is needed. Furthermore, space must be provided with the linear flying spot scanner to accommodate the moving apparatus. Alternatively, certain known radiographic inspection systems utilize a collimated linear array of electronic detectors. The electronic detectors are moved relative to the object and radiation source to successively generate response signals for the successive sections or areas of the object. The radiation source of such a system transmits radiation toward the entire object being inspected, permitting the detector array, as moved, to generate response signals for each section of the object. The moving detector system requires a large number of electronic detectors to inspect the object to ensure high resolution. The major disadvantages of such a system are the need to move the detectors relative to the object and the large number of detectors and associated electronics. The detectors must be balanced for uniform response characteristics. SUMMARY OF THE INVENTION The present invention has as an object to provide a radiographic inspection system which does not have the limited dynamic range of X-ray film inspection systems and which may be used to minimize the effects of fluorescent and compton scattered radiation. The present invention also has as an object providing a radiographic inspection system which does not require movement of either the object or the electronic detectors relative to the rest of the inspection system in order to obtain a radiographic representation of the entire object. The present invention has a further object the provision of such a radiographic inspection system having the capability of determining the atomic elements and atomic element distribution within the object. Additional objects and advantages of the invention will be set forth in the description which follows, and in part will be obvious from the description or may be learned by practice of the invention. The objects and advantages of the invention may be realized and obtained by means of the intrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing objects, and in accordance with the purposes of the invention as embodied and broadly described herein, there is provided a system for radiographically inspecting a predetermined area of a relatively stationary object positioned at a selected location, comprising: a source of radiation operative to transmit along a radiation path toward the selected location a first radiation beam having a cross-sectional area at least corresponding to the predetermined inspection area; detection means, having a detection area at least corresponding to the predetermined inspection area, disposed in the radiation path in alignment with the first radiation beam; scanning point selection means, disposed in the radiation path between the radiation source and detection means, for sequentially selecting and transmitting selected portions of the cross-sectional area of the first radiation beam for striking in sequence corresponding portions of the detection area of the detection means, each selected portion of the cross-sectional area of the first radiation beam corresponding to a pencil beam of radiation, the detection means being responsive to each selected portion of the first radiation beam striking a corresponding portion of the detecting area for generating signals corresponding to radiation interactive with a corresponding portion of the predetermined inspection area of an object at the selected location; position encoder means responsive to said scanning point selection means for determining the position of each selected portion of the first radiation beam; data processing means responsive to said position encoder means and said detecting means for processing the signals generated by the detection means; and display means governed by said data processing means for generating a radiographic image of the predetermined area of the object. In another aspect, such a system further comprises means for filtering the first radiation beam disposed in the radiation path between the radiation source and the scanning point selection means for causing the display means to generate a representation of the various atomic element groups in the predetermined area of the object. A related method is also provided. The method, for radiographically inspecting a predetermined area of a stationary object positioned at a selected location, comprises the steps of: transmitting a first radiation beam having a cross-sectional area at least corresponding to the predetermined inspection area along a radiation path in a direction toward the selected location; selecting, sequentially, portions from the cross-sectional area of the first radiation beam and transmitting the selected portions along the radiation path to the selected location, each said selected portion of the cross-sectional area of the first radiation beam comprising a pencil beam of radiation; detecting each of the selected portions of the first radiation beam striking corresponding portions of the predetermined inspection area of an object at the selected locations and generating signals in response to the radiation striking the corresponding portions; determining the position of each selected portion of the first radiation beam; processing the signals generated in response to the radiation striking the corresponding portions; and generating a radiographic image of the predetermined area of the object. In another aspect the related method further comprises the step of filtering each selected portion of the cross-sectional area to determine a corresponding atomic element group.
062158511	claims	1. A proton beam target for generating gamma rays which are generated therefrom in response to an impinging proton beam, the proton beam target comprising: a .sup.13 C gamma reaction layer for generating the gamma rays therefrom; and a stopping layer for mitigating transmission of the proton beam therethrough, the stopping layer being formed of a refractory metal which has a relatively high hydrogen solubility for dissolving implanted hydrogen atoms therewithin as a result of the impingement of the proton beam and which is chemically reactive with the .sup.13 C gamma reaction layer for chemically bonding therewith. (a) forming a stopping layer of a refractory metal for mitigating transmission of the proton beam therethrough, the stopping layer having a relatively high hydrogen solubility for dissolving implanted hydrogen atoms therewithin as a result of the impingement of the proton bean and being chemically reactive with .sup.13 C; and (b) attaching a .sup.13 C gamma reaction layer for generating gamma rays therefrom in response to the impinging proton beam. (c) attaching the stopping layer onto a cooling support for dissipating heat energy away from the stopping layer. (d) cooling the stopping layer to a temperature less than the braze temperature; and (e) reliving any residual thermal stress developed within the stopping layer through the movement of the stopping layer relative to the cooling support. 2. The proton beam target of claim 1 wherein the refractory metal is chosen from the group consisting of Tantalum, Zirconium, Niobium and Hafnium. 3. The proton beam target of claim 1 wherein the stopping layer is formed of a metal foil. 4. The proton beam target of claim 1 wherein the stopping layer has a thickness of approximately between 20 to 130 microns. 5. The proton beam target of claim 1 wherein the stopping layer has a hydrogen solubility greater than that of gold. 6. The proton beam target of claim 1 wherein the stopping layer is chemically reactive with the .sup.13 C gamma reaction layer to form carbide atoms therebetween. 7. The proton beam target of claim 1 wherein the .sup.13 C gamma reaction layer is sputter deposited onto the stopping layer. 8. The proton beam target of claim 1 further comprises a cooling support for dissipating heat energy away from the stopping layer, wherein the stopping layer is attached to the cooling support and the stopping layer is interposed between the .sup.13 C gamma reaction layer and the cooling support. 9. The proton beam target of claim 8 wherein the cooling support is formed of Copper. 10. The proton beam target of claim 8 wherein the cooling support is formed of Beryllium. 11. The proton beam target of claim 8 further comprises a braze layer which is interposed between stopping layer and the cooling support for attaching the stopping layer to the cooling support. 12. The proton beam target of claim 11 wherein the braze layer is formed of Silver based braze alloy. 13. A method of fabricating a proton beam target for generating gamma rays which are reflected therefrom in response to an impinging proton beam, the method comprising the steps of: 14. The method of claim 13 wherein the .sup.13 C gamma reaction layer is attached to the stopping layer via sputter deposition. 15. The method of claim 13 wherein step (a) further comprises chemically reacting the .sup.13 C gamma reaction with the stopping layer to form a carbide phase therebetween. 16. The method of claim 13 further comprising the step of: 17. The method of claim 16 wherein the stopping layer is attached to the cooling support via brazing. 18. The method of claim 17 wherein the brazing is effectuated at a braze temperature greater than 500.degree. C. 19. The method of claim 18 further comprising the steps of:
	summary
045044392	abstract	It is desirable to being able to vary the temperature of a loop exposed to a flow of hot gases in gas cooled nuclear reactor installations without thereby affecting the gas temperature in the other loops. It should be possible in an extreme case to take the affected loop out of service. This is achieved by arranging under the heat absorbing components in the hot gas channel a means for mixing gas connected with a cold gas line.
046738142	abstract	The invention is directed to a container for receiving and safely storing radioactive or other materials damaging to living organisms, especially such materials as vitrified radioactive fission products or irradiated nuclear reactor fuel elements. The container includes a vessel and a sealing cover. The vessel has a circular opening at one of its ends for receiving the materials to be stored therein, and a sealing cover is seated in this opening to tightly seal the container. In order to obtain a container of the kind which has the required high impermeability to gas, the open end of the inner bore of the vessel is widened to define a conical surface. A sealing cover is pressed into the vessel at this conical surfce. The sealing cover has a peripheral surface which converges toward the interior of the vessel and which has a taper corresponding to that of the conical surface of the vessel. The vessel wall and sealing cover are welded together by means of a fused-mass joint extending around the entire periphery of the cover. By means of the conical configuration of the upper end portion of the vessel and of the sealing cover, a considerably improved seal between the vessel wall and the sealing cover is obtained. A method for sealing the container is also disclosed.
	summary
	claims	1. A system for replacing equipment, comprising:a first memory module associated with a first equipment, the first memory module configured to store one or more first characteristics of the first equipment;a second memory module associated with a second equipment, the second memory module configured to store one or more second characteristics of the second equipment;a portable reader configured to read the first and second memory modules and to compare the read one or more second characteristics with the read one or more first characteristics to determine if the second equipment is a suitable replacement for the first equipment; anda measuring device configured to measure at least one of the one or more first characteristics and the one or more second characteristics, wherein the portable reader is associated with the measuring device, and is further configured to store the measured one or more first characteristics on the first memory module, and to store the measured one or more second characteristics on the second memory module. 2. The system of claim 1, wherein the first equipment is a first section of pipe, and the second equipment is a second section of pipe. 3. The system of claim 1, wherein the one or more first characteristics include a first material classification of the first equipment, and the one or more second characteristics include a second material classification of the second equipment. 4. The system of claim 3, wherein the portable reader is configured to determine whether the second material classification matches the first material classification. 5. The system of claim 1, wherein the first memory module is attached to the first equipment. 6. The system of claim 5, wherein the second memory module is attached to the second equipment. 7. The system of claim 5, wherein the second memory module is attached to a container that at least partially contains the second equipment. 8. The system of claim 1, wherein the portable reader is further configured to erase at least one of (a) existing first characteristics before storing the measured one or more first characteristics on the first memory module, and (b) existing second characteristics before storing the measured one or more second characteristics on the second memory module. 9. The system of claim 1, wherein the measured one or more first characteristics include a first material classification for the first equipment, and the measured one or more second characteristics include a second material classification for the second equipment. 10. The system of claim 9, wherein the measuring device is configured to perform an x-ray fluorescence to measure percentage compositions of one or more elements of the first equipment and the second equipment. 11. The system of claim 10, wherein the measuring device is further configured to determine the first and second material classifications based, at least in part, on the measured percentage compositions of one or more elements. 12. A method of replacing equipment, comprising:storing one or more first characteristics of a first equipment on a first memory module associated with the first equipment;transmitting the one or more first characteristics to a portable reader configured to read the first memory module;measuring one or more second characteristics of a second equipment using a measuring device associated with the portable reader; andcomparing, with the portable reader, the measured one or more second characteristics with the read one or more first characteristics to determine if one of the first and second equipment is a suitable replacement for the other equipment, where the one or more first characteristics include a first material classification and the one or more second characteristics include a second material classification, wherein comparing the measured one or more second characteristics with the read one or more first characteristics includes determining whether the second material classification matches the first material classification. 13. A method of replacing equipment, comprising:storing one or more first characteristics of a first equipment on a first memory module associated with the first equipment;transmitting the one or more first characteristics to a portable reader configured to read the first memory module;measuring one or more second characteristics of a second equipment using a measuring device associated with the portable reader; andcomparing, with the portable reader, the measured one or more second characteristics with the read one or more first characteristics to determine if one of the first and second equipment is a suitable replacement for the other equipment, wherein measuring one or more second characteristics of a second equipment includes:measuring percentage compositions of one or more elements of the second equipment; andclassifying material of the second equipment based, at least in part, on the measured percentage compositions of the one or more elements. 14. A method of replacing equipment, comprising:storing one or more first characteristics of a first equipment on a first memory module associated with the first equipment;storing one or more second characteristics of a second equipment on a second memory module associated with the second equipment;transmitting the one or more first characteristics to a portable reader;transmitting the one or more second characteristics to the portable reader, the portable reader being configured to read the first and second memory modules; andcomparing, with the portable reader, the read one or more second characteristics with the read one or more first characteristics to determine if the second equipment is a suitable replacement for the first equipment, where the one or more first characteristics include a first material classification and the one or more second characteristics include a second material classification, wherein comparing the read one or more second characteristics with the read one or more first characteristics includes determining whether the second material classification matches the first material classification.
043409709	description	DETAILED DESCRIPTION OF A PREFERRED EMBODIMENT Referring to FIGS. 1 and 2 there is shown a power drive of the present invention. Several heat expansion valves are installed around a drive shaft, FIGS. 1 and 2 (8-12) which activates a spindle drive FIG. 1 (11) to turn the big wheel around a side gear, FIG. 2 (13). The movement is accomplished through a hollow cylinder filled with expansible liquid, FIG. 1 (12). At the very end of the valve unit is an opening in which a pushrod is inserted, FIG. 3 (23) the pushrod is sealed tight, fitted with a rubber cap inside the cylinder, FIG. 4 (23-33). When the liquid inside the valve starts to expand or contract, it will move the pushrod in or out. The pushrod is joined against a thrust bearing inside a spindle, FIG. 4 (11-23-24). A pinion gear is inserted into the spindle through a stud. FIG. 4 (4) and a key, FIG. 4 (25) so that the spindle can slide up or down on the stud and turn the pinion gear at the same time, FIG. 4 (4-25-35). Inside the pinion gear is a ratched gear installed. FIGS. 6 & 7 (27-29-76). The ratched gear locks into the spindle drive on the down stroke and disengages on the up stroke, FIG. 6 (26-27-30). A pressure spring is installed between the pinion gear and the drive spindle, FIG. 4 (10). To move the spindle drive upwards into the spindle threat cylinder, FIG. 5 (4-34) together with the contraction of the liquid. The pinion gear consists of three major parts, pinion gear, ratched gear and sliding collar, FIG. 6 (7-29-30). The sliding collar locks the pinion gear against the two side gears, FIG. 6 (7-31-32). One side gear is fastened against the stand, so the pinion gear can drive the big wheel around it. Opposite a second side gear called satellite gear is disconnected are connected to the drive shaft depending on the Design, FIG. 2 (32). Its function is to keep the pinion gear against the side gears, FIG. 6 (5-31-32). NUCLEAR HEAT EXPANSION VALVE The difference between the conventional valve and the nuclear, are the nuclear heat elements. The heat elements are made of two parts, one part uranium 235 and one part plutonium, placed around a cadmium plunger, FIG. 3 (18-20). When the valve is in an up position the cadmium stick will fall through its own weight down till the stopper unit on back of the stick hits the flange of the stopper bushing, FIG. 3 (17-37). In this position the nuclear reaction is neutralized, FIG. 3 (18-20). Around the nuclear heat elements is a lead crystal ring, FIG. 3 (21). The function of this is to cover the back surface of the nuclear heat elements against getting hit by neutrons and electrons which are still bouncing within the cylinder. This is made possible by using lead crystal reflector shields, FIG. 3 (21). I discovered that lead crystal glass reflects radioactive rays including neutrons as long as the crystal glass is kept under 1000 Fahrenheit temperature. In this design only a temperature range up to 500.degree. Fahrenheit is introduced and the heat elements are only in introduction of a nuclear reaction. Toluene is not only an excellent heat transmitter but also an expansible liquid and moderator. The moderation can be intensified through graphite. As soon as the cadmium plunger neutralizes the nuclear reaction, some neutrons will still be active from the previous nuclear reaction made possible by reflecting the neutrons from one crystal lead reflector to the other. As soon as the big wheel turns and the nuclear heat expansion valve stands upside down, the weight of the cadmium stick will pull the cadmium plunger out of the nuclear heat elements, FIG. 5 (18-20-37). Now a nuclear reaction is introduced which will immediately take place because of the large amount of neutrons and electrons still in circulation from the last reaction. In this fashion a very efficient and safe nuclear via mechanical conversion is accomplished without any electronic controls or sophisticated cooling system. As soon as the machine is in motion, it will adopt itself to a permanent speed till the nuclear heat elements are exhausted. To stop the machine, lock screws are placed on the outside, FIG. 3 (9), which are screwed inwards. The cadmium stick will automatically move inward and stop all nuclear actions. Such a machine can produce power of sufficient amounts. A wheel of ten feet in diameter with eight nuclear expansion valves, operating at an 80.degree. Fahrenheit temperature difference, can produce approximately 700 kilowatt an hour which is ideal for consumers, such as farmers, homeowners, factories and businesses, also in remote areas as army bases, but more important underwater explorations and mining, regardless of the depth and pressure. The nuclear power wheel can operate publicly to demonstrate how nuclear energy can be converted into a useful power with absolutely no danger factor involved and, which is also important, free of polluting the environment. This knowledge is very important to the public particularly since I discovered that nuclear waste material, meaning high radioactive material, can be converted into a useful energy in a patent application called Weinert Light Electrons Amplifier. I believe that nuclear waste material will in the near future be just as much in demand as nuclear fuel. The advantage of having two different isotopes used as heat elements, is to have more intense radio activity which allows a smaller amount of nuclear mass. In conjunction with the crystal reflectors a much faster nuclear reaction is introduced. The Nuclear cells can also operate only one plutonium which means Nuclear waste can be used as Nuclear Fuel.
048809888	abstract	To simplify separation of infrared (IR) radiation from visible and ultraviolet (UV) radiation emitted from a xenon lamp (2), the lamp is surrounded by a cylindrical filter (1) reflecting IR radiation inwardly of the filter but passing UV and visible spectral radiation therethrough. IR radiation reflected inwardly of the filter is absorbed, in accordance with the invention, by an absorbing surface (5) extending axially longitudinally and diametrically across the cylindrical filter, with the cylinder axis (4) of the filter. An auxiliary surface (9) can be positioned between the radiation source and the adjacent nearer side wall of the filter. Reflected IR radiation (8) can thus be absorbed by the IR absorbing surface, with essentially only UV and visible light radiation passing through the cylindrical filter.
	abstract	Apparatus for X-ray analysis has a combination of a rotating target X-ray tube and a composite monochromator. The composite monochromator has a first and a second elliptic monochromators joined with each other side by side. Each of the elliptic monochromators has a first focal point at which an X-ray focal spot on a target of the X-ray tube is disposed. Each of the elliptic monochromators has a synthetic multilayered thin film whose d-spacing varies continuously along an elliptic-arc. The shortest distance between the X-ray focal spot and the composite monochromator is set to 40 to 100 mm. Under the shortest distance condition, the effective focal spot size on the target is set to 40 to 100 micrometers to obtain the maximum X-ray intensity on a sample to be analyzed.
055330894	abstract	The invention is an instrument comprising a unit having a frame defining the perimeter about an aperture, and a manually adjustable rack connected to the frame for supporting the unit with its aperture disposed between a source of x-rays and the body of a patient. Radio-opaque flexible panels depending from the frame are pivotable relative to the aperture to collimate and limit x-rays issuing from the source through the aperture to a particular area of the body of the patient. Translucent ports along lower edges of the panels serve to provide upon a photographic film below the patient a scale indication of the area of the body of the patient subjected to the x-rays.
	claims	1. A method for correcting distortions in an electron backscatter diffraction pattern image including lines of picture elements, comprising the steps of: extracting values from a calibration image, the values corresponding to corrections in patterns that are distorted by magnetic fields when using electron backscatter diffraction (EBSD) in a scanning electron microscope (SEM); and shifting picture elements in an electron backscatter diffraction (EBSD) image by the extracted values to remove distortion from the EBSD image to create a corrected EBSD image. 2. The method of claim 1 wherein the EBSD image includes a curved band and wherein shifting picture elements in an EBSD image includes shifting picture elements in the EBSD image by the extracted values to straighten the curved band. claim 1 3. The method of claim 1 wherein shifting picture elements in an EBSD image produces an unfilled portion of the corrected EBSD image, and further comprising filling the unfilled portion of the corrected EBSD image with picture elements having a selected intensity. claim 1 4. The method of claim 3 wherein filling the unfilled portion of the corrected EBSD image includes filling the unfilled portion of the corrected EBSD image with picture elements having an intensity approximately equal to an average intensity of the picture elements in the EBSD image. claim 3 5. The method of claim 1 further comprising displaying the corrected EBSD image. claim 1 6. The method of claim 1 further comprising saving the corrected EBSD image. claim 1 7. A method for correcting distortions in an electron backscatter diffraction pattern image, comprising the steps of: (a) calculating correction parameters for electron backscatter diffraction (EBSD) in a scanning electron microscope (SEM), the correction parameters defining an amount by which points along a curved band in a first distorted EBSD image are shifted to straighten the curved band; (b) storing the correction parameters for use in correcting magnetic field distortions in EBSD patterns using a substantially similar SEM geometry; and (c) shifting picture elements in a second distorted EBSD image by selected amounts determined by the correction parameters to remove the distortion from the distorted EBSD image and create a corrected EBSD image. 8. The method of claim 7 wherein calculating the correction parameters comprises calculating a mathematical curve fitting the curved band. claim 7 9. The method of claim 8 wherein calculating a mathematical curve fitting the curved band comprises calculating a cubic spline fitting the curved band. claim 8 10. The method of claim 8 wherein calculating a mathematical curve fitting the curved band comprises: claim 8 (a) defining a plurality of segment endpoints following the curved band in the first distorted EBSD image; and (b) calculating the mathematical curve fitting the segment endpoints. 11. The method of claim 10 wherein defining a plurality of segment endpoints comprises: claim 10 (a) displaying the corrected EBSD image on an operator display; and (b) defining the plurality of segment endpoints following the curved band in the corrected EBSD image using an operator input device. 12. The method of claim 10 wherein calculating the mathematical curve comprises calculating four cubic spline coefficients for each segment endpoint. claim 10 13. The method of claim 7 further comprising shifting picture elements in a plurality of EBSD images for a particular SEM by amounts determined by the stored correction parameters. claim 7 14. The method of claim 7 wherein shifting picture elements in the second distorted EBSD image produces an unfilled portion of the corrected EBSD image, and further comprising filling the unfilled portion of the corrected EBSD image with picture elements having a selected intensity. claim 7 15. The method of claim 14 wherein filling the unfilled portion of the corrected EBSD image includes filling the unfilled portion of the corrected EBSD image with picture elements having an intensity approximately equal to an average intensity of the picture elements in the corrected EBSD image. claim 14 16. The method of claim 7 further comprising displaying the corrected EBSD image. claim 7 17. The method of claim 7 further comprising saving the corrected EBSD image. claim 7 18. A system for obtaining electron backscatter diffraction patterns from a sample and correcting distortions therein, comprising: (a) an electron microscope configured to direct a beam of electrons onto a sample to be backscattered therefrom; (b) an image collection system that provides an electron backscatter diffraction (EBSD) pattern image including lines of picture elements from the electrons backscattered from the sample; (c) means for acquiring correction parameters defining selected amounts to correct a distortion in a calibration EBSD image; and (d) means for shifting picture elements in the EBSD pattern image by selected amounts to remove the distortion from the image to create a corrected EBSD pattern image. 19. The system of claim 18 wherein the electron microscope is a scanning electron microscope. claim 18 20. The system of claim 18 wherein the image collection system includes an image digitizer for providing a digitized electron backscatter diffraction pattern image including lines of picture elements and wherein the means for shifting picture elements in the EBSD pattern image includes a computer programmed to receive the digitized electron backscatter diffraction pattern image from the image collection system and to shift picture elements in the EBSD pattern image by selected amounts to remove the distortion from the image to create a corrected EBSD pattern image. claim 18 21. The system of claim 18 wherein the EBSD pattern image includes a curved band and wherein the means for shifting picture elements in the EBSD pattern image includes means for shifting picture elements in the electron backscatter diffraction pattern image by selected amounts to straighten the curved band. claim 18 22. The system of claim 21 wherein the means for calculating the correction parameters includes means for calculating a mathematical curve fitting the curved band. claim 21 23. The system of claim 22 wherein the means for calculating the mathematical curve includes means for calcuating a cubic spline fitting the curved band. claim 22 24. The system of claim 22 wherein the means for calculating the mathematical curve includes: claim 22 (a) means for defining a plurality of segment endpoints following the curved band in the electron backscatter diffraction pattern image; and (b) means for calculating a mathematical curve fitting the endpoints. 25. The system of claim 24 wherein the means for defining a plurality of segment endpoints includes: claim 24 (a) an operator display; (b) means for displaying the electron backscatter diffraction pattern image on the operator display; and (c) an operator input device for defining the plurality of segment endpoints following the curved band in the electron backscatter diffraction pattern image displayed on the operator display. 26. The system of claim 24 wherein the means for calculating the mathematical curve includes means for calculating four cubic spline coefficients for each segment endpoint. claim 24 27. The system of claim 18 further comprising memory for saving the correction parameters and means for shifting picture elements in a plurality of electron backscatter diffraction pattern images by amounts determined by the saved correction parameters. claim 18 28. The system of claim 18 further comprising means for fill an unfilled portion of the corrected electron backscatter diffraction pattern image with picture elements having a selected intensity. claim 18 29. The system of claim 28 wherein the means for filling the unfilled portion of the corrected EBSD pattern image includes means for filling the unfilled portion of the corrected EBSD pattern image with picture elements having an intensity approximately equal to an average intensity of the picture elements in the EBSD pattern image. claim 28 30. The system of claim 18 further comprising an operator display for displaying the corrected EBSD pattern image. claim 18 31. The system of claim 18 further comprising memory for saving the corrected EBSD pattern image. claim 18
	summary
	description	This application claims a benefit of priority based on Japanese Patent Application No. 2003-110210, filed on Apr. 15, 2003, which is hereby incorporated by reference herein in its entirety as if fully set forth herein. The present invention relates to an exposure apparatus used for a semiconductor manufacturing process, and a projection exposure apparatus that projects and transfers a reticle pattern onto a silicon wafer. The present invention is suitable for an extreme ultraviolet (“EUV”) exposure apparatus that uses EUV light as exposure light with a wavelength of about 13 to 14 nm and a mirror optical system for projection exposure in vacuum. A prior art example will be described with reference to FIGS. 6 and 7. 101 uses a YAG solid laser etc., and serves as an excitation laser for exciting light-source material atoms into plasma for light emissions by irradiating a laser beam onto an emitting point of the light source, at which the light-source material is in a state of gas, liquid or atomized gas. 102 is a light-source emitting part that maintains an internal structure to be vacuum. Here, 102A is a light source A indicative of an actual emitting point of an exposure light source. 103 is a vacuum chamber for accommodating an exposure apparatus entirely, which can maintain the vacuum state using a vacuum pump 104. 105 is an exposure light introducing part for introducing exposure light from the light-source emitting part 102, which includes mirrors A (or 105A) to D (or 105D), and homogenizes and shapes the exposure light. 106 is a reticle stage, and a movable part of the reticle stage is mounted with a reflective original form 106A that forms a pattern to be exposed. 107 is a reduction projection mirror optical system that reduces and projects an exposure pattern reflected from the original form 106A through mirrors A (or 107A) to E (or 107E) sequentially at predefined reduction ratio. 108 is a position-controlled wafer stage for positioning a wafer 108A as a Si substrate onto a predetermined exposure position so that the wafer stage can be driven in six axes directions, i.e., driven in XYZ directions, tilt around the XY axes, and rotated around the Z axis. The pattern on the original form 106A is to be reflectively reduced and projected onto the wafer 108A. 109 is a reticle stage support for supporting the reticle stage 105 on the apparatus installation floor. 110 is a projection optical system body for supporting the reduction projection mirror optical system 107 on the apparatus installation floor. 111 is a wafer stage support for supporting the wafer stage 108 on the apparatus installation floor. Provided between the reticle stage 105 and the reduction projection mirror optical system 107 and between the reduction projection mirror optical system 107 and the wafer stage 108, which are distinctly and independently supported by the reticle stage support 109, the projection optical system body 110 and the wafer stage support 111, are means (not shown) for measuring relative positions to continuously maintain a predetermined arrangement of them. A mount (not shown) for violation isolation from the apparatus installation floor is provided on the reticle stage support 109, the projection system body 110, and the wafer stage 111. 112 is a reticle stocker as a storage container that temporarily stores, in an airtight condition, plural original forms 106A as reticles supplied from the outside of the apparatus and suitable for different exposure conditions and patterns. 113 is a reticle changer for selecting and feeding a reticle from the reticle stocker 112. 114 is a reticle alignment unit that includes a rotatable hand that is movable in the XYZ directions and rotatable around the Z axis. The reticle alignment unit 114 receives the original form 106A from the reticle changer 113, rotates it by 180°, and feeds it to the reticle alignment scope 115 provided at the end of the reticle stage 106 for fine movements of the original form 106A in the XYZ-axes rotating directions and alignment with respect to the alignment mark 115A provided on the reduction projection mirror optical system 107. The aligned original form 106A is chucked on the reticle stage 106. 116 is a wafer stocker as a storage container for temporarily storing plural wafers 108A from the outside to the inside of the apparatus. 117 is a wafer feed robot for selecting a wafer 108A to be exposed, from the wafer stocker 116, and feeds it to a wafer mechanical pre-alignment temperature controller 118 that roughly adjusts feeding of the wafer in the rotational direction and controls the wafer temperature within predetermined controlled temperature in the exposure apparatus. 119 is a wafer feed hand that feeds the wafer that has been aligned and temperature-controlled by the wafer mechanical pre-alignment temperature controller 118 to the wafer stage. 120 and 121 are gate valves that constitute a gate opening/closing mechanism for supplying the reticle and wafer from the outside of the apparatus. 122 is also a gate valve that uses a diaphragm to separate a space of the wafer mechanical pre-alignment temperature controller 118 from an exposure space, and opens and closes only when feeding in and out the wafer. Such a separation using the diaphragm can minimize a capacity to be temporarily released to the air, and form a vacuum equilibrium state. Thus, when the conventionally structured exposure apparatus supports and positions the mirrors A to E relative to the mirror barrel 107F as shown in FIG. 7, fine displacements and inclinations of the rotational axis in the in-plane translation shift direction occur, and the mirror deforms by its own weight. This cannot satisfy extremely strict mirror surface shape precisions below about 1 nm necessary for the projection optical system mirrors, the illumination optical system mirrors, and the light source mirrors. When the mirror's surface precision and thus the optical aberration deteriorate, the projection optical system, in particular, deteriorates imaging performance to the wafer and lowers light intensity. The exposure light introducing part introduces the exposure light from the light-source emitting part in such a conventionally structured exposure apparatus. The reduction projection mirror optical system reduces and projects the exposure pattern reflected from the original form illuminated by the mirrors A to D in the exposure light introducing part. The reduction projection mirror optical system makes a multilayer of Mo—Si on each of the mirrors A to E by vacuum evaporation or sputtering, and reflects the exposure light from the light source on each reflective surface. In this case, the reflectance per surface is about 70%; the rest is absorbed in the mirror base material and converted into heat. The temperature rises by about 10 to 20° C. in the exposure light reflecting area, and the reflective surface deforms by about 50 to 100 nm around the mirror peripheral even when the mirror uses a material having an extremely small coefficient of thermal expansion. As a result, extremely strict mirror surface shape precisions below about 1 nm necessary for the projection optical system mirrors, the illumination optical system mirrors, and the light source mirrors cannot be maintained. When the mirror surface precision, the projection optical system deteriorates imaging performance to the wafer and lowers light intensity. In addition, the illumination optical system lowers the target light intensity, and causes non-uniform light intensity. The light source mirror deteriorates the light intensity, such as insufficient condensing. They result in deteriorated basic performance of the exposure apparatus, such as exposure precision and throughput. Accordingly, it is an exemplary object of the present invention to provide an exposure apparatus that can precisely control wave front aberration of a projection optical system and has high imaging performance. An exposure apparatus of one aspect according to the present invention includes a reflective element for reflecting and introducing light from a light source to a plate, at least one first driver for providing the reflective element with a force and/or a displacement in at least one directions, and at least one second driver for providing the reflective element with a force and/or a displacement in at least one directions, wherein the first and second drives are connected in series to each other. An exposure apparatus of another aspect according to the present invention for introducing light from a light source to a plate includes a barrel, a support member, a reflective element for reflecting light from the light source to the plate, at least one first driver, connected to said barrel and support member, for providing a force and/or a displacement in at least one directions, and at least one second driver, connected to said barrel and reflective element, for providing a force and/or a displacement in at least one directions. A device fabrication method of another aspect according to the present invention includes the steps of exposing a plate using the above exposure apparatus; and developing the plate that has been exposed. A description will be given of an embodiment according to the present invention with reference to FIGS. 1 to 5. 1 uses a YAG solid laser etc., and serves as an excitation laser for exciting light-source material atoms into plasma for light emissions by irradiating a laser beam onto an emitting point of the light source, at which the light-source material is in a state of gas, liquid or atomized gas. 2 is a light-source emitting part that maintains an internal structure to be vacuum. Here, 2A is a light source A indicative of an actual emitting point of an exposure light source. 3 is a vacuum chamber for entirely accommodating an exposure apparatus, which can maintain the vacuum state using a vacuum pump 4. 5 is an exposure light introducing part for introducing exposure light from the light-source emitting part 2, which includes mirrors A (or 5A) to D (or 5D), and homogenizes and shapes the exposure light. Of course, the number of mirrors in the exposure light introducing part is not limited to four, and may be five, six, seven, eight, or another number. 6 is a reticle stage, and a movable part of the reticle stage is mounted with a reflective original form 6A that forms a pattern to be exposed. 7 is a reduction projection mirror optical system that reduces and projects an exposure pattern reflected from the original form through mirrors A (or 7A) to E (or E) sequentially at predefined reduction ratio. 7F is a mirror barrel that holds mirrors A to E. 8 is a position-controlled wafer stage for positioning a wafer 8A as a Si substrate onto a predetermined exposure position so that the wafer stage can be driven in six-axes directions, i.e., driven in XYZ directions, tilt around the XY axes, and rotated around the Z axis. The pattern on the original form is to be reflectively reduced and projected onto the wafer 8A. 9 is a reticle stage support for supporting the reticle stage 5 on the apparatus installation floor. 10 is a projection optical system body for supporting the reduction projection mirror optical system 7 on the apparatus installation floor. 11 is a wafer stage support for supporting the wafer stage 8 on the apparatus installation floor. Provided between the reticle stage 5 and the reduction projection mirror optical system 7 and between the reduction projection mirror optical system 7 and the wafer stage 8, which are distinctly and independently supported by the reticle stage support 9, the projection optical system body 10 and the wafer stage support 11, are means (not shown) for measuring relative positions to continuously maintain a predetermined arrangement of them. A mount (not shown) for violation isolation from the apparatus installation floor is provided on the reticle stage support 9, the projection system body 10, and the wafer stage 11. 12 is a reticle stocker as a storage container that temporarily stores, in an airtight condition, plural original forms as reticles supplied from the outside of the apparatus and suitable for different exposure conditions and patterns. 13 is a reticle changer for selecting and feeding a reticle from the reticle stocker 12. 14 is a reticle alignment unit that includes a rotatable hand that is movable in the XYZ directions and rotatable around the Z axis. The reticle alignment unit 14 receives the original form from the reticle changer 13, rotates it by 180°, and feeds it to the reticle alignment scope 15 provided at the end of the reticle stage 6 for fine movements of the original form 6A in the XYZ-axes rotating directions and alignments with respect to the alignment mark 15A provided on the reduction projection mirror optical system 7. The aligned original form is chucked on the reticle stage 6. 16 is a wafer stocker as a storage container for temporarily storing plural wafers from the outside to the inside of the apparatus. 17 is a wafer feed robot for selecting a wafer to be exposed, from the wafer stocker, and feeds it to a wafer mechanical pre-alignment temperature controller 18 that roughly adjusts feeding of the wafer in the rotational direction and controls the wafer temperature within predetermined controlled temperature in the exposure apparatus. 19 is a wafer feed hand that feeds the wafer that has been aligned and temperature-controlled by the wafer mechanical pre-alignment temperature controller 18 to the wafer stage. 20 and 21 are gate valves that constitute a gate opening/closing mechanism for supplying the reticle and wafer from the outside of the apparatus. 22 is also a gate valve that uses a diaphragm to separate a space of the wafer mechanical pre-alignment temperature controller 18 from an exposure space, and opens and closes only when feeding in and out the wafer. Such a separation using the diaphragm can minimize a capacity to be temporarily released to the air, and form a vacuum equilibrium state. The above structure further includes, as a solution for insufficient positional shape precision problems in the prior art, means for correcting a mirror position, surface precision, and projection optical system's wave front aberration. FIG. 2 exemplarily shows mirrors C (or 7C) and E (or 7E) in the reduction projection mirror optical system. The mirror C is supported in the mirror barrel 7F via plural rough-movement drive means 25B, plural element positioners 25C, mirror holding element 25D, and plural fine-movement drive means 25E in this order from the mirror barrel 7F. In other words, the above rough-movement drive means and fine-movement drive means can drive the mirror C relative to the mirror barrel 7F. There may be only one rough-movement drive means and only one fine-movement drive means, but the exposure apparatus preferably includes three each, and more preferably each being drivable in two directions. The instant embodiment provides three rough-movement drive means between the mirror barrel and the mirror holding element. Control over the rough-movement drive means makes the mirror holding element drivable in six-axes directions or six degrees of freedom relative to the mirror barrel. In addition, three fine-movement drive means are provided between the mirror holding element and the mirror, and control over the fine-movement drive means makes the mirror drivable in six-axes directions or six degrees of freedom relative to the mirror holding element (and the mirror barrel). In other words, the rough and fine movements of the mirror in six-axes directions are available by a series connection of the rough-movement drive means and fine-movement drive means between the mirror barrel and the mirror. The minimum drive amount of the actuator attached to the fine-movement drive means is made larger than the preferably double or more preferably triple of the minimum drive amount of the driven part in the fine-movement drive means. The minimum driving unit (such as a distance and an angle) in the rough-movement drive means is made larger than the minimum driving unit in the fine-movement drive means. The minimum driving unit in the rough-movement drive means is made larger than the preferably double or more preferably decuple of the minimum driving unit in the fine-movement drive means. There are plural mirror rough-movement displacement measuring means 25F between the mirror barrel and the mirror holding element 25D, for measuring a displacement of the mirror holding element 25D driven by the rough-movement drive means 25B. In addition, there are plural mirror fine-movement displacement measuring means 25G between the mirror holding element 25D and the mirror C, for measuring a displacement of the mirror C driven by the fine-movement drive means 25B. A similar structure that includes plural rough-movement drive means 26B, plural element positioning members 26C, mirror holding elements 26D plural fine-movement drive means 26E in this order from the mirror barrel 7E side is provided for the mirror E, like the mirror C. In other words, the above rough-movement drive means and fine-movement drive means can drive the mirror E relative to the mirror barrel 7E. Here, there may be only one rough-movement drive means and only one fine-movement drive means, but the exposure apparatus preferably includes three (or more) each, and more preferably each being drivable in two directions. There are plural mirror rough-movement displacement measuring means 26F between the mirror barrel and the mirror holding element as means for measuring a position of the mirror E relative to the mirror barrel. The mirror rough-movement displacement measuring means 26F measures a displacement of the mirror holding element 26D driven by the rough-movement drive means 26B. There are plural mirror fine-movement displacement measuring means 26G for measuring a displacement of the mirror E driven by the fine-movement drive means 26E, between the mirror holding element 26D and the mirror E. The measuring means for the mirror relative to the mirror barrel and the mirror holding element provides origins of the mirror rough-movement drive means and fine-movement drive means. By this origin setting, the measurements of the positions among mirrors relative to the mirror barrel are available. In addition to measurements of the positions among mirrors relative to the mirror barrel, measuring means, such as a laser interferometer, is provided as means for precisely measuring a mirror position (with precision below 1 nm), so as to narrow down the reflective surface precision of each mirror within a target surface precision from the projection optical precision. A description will now be given of the measurement method using this laser interferometer. FIG. 3 shows a mirror position measured by the laser interferometer. As shown in a view of the mirror E, the measurement reflective surface is provided on the mirror itself, and respective mirror positions and relative positions from the projection optical system body 10 are measured. In order to XY measurements of the mirror holding element 26D, the measurement light of the mirror displacement measuring means 26H by the laser interferometer, etc. is irradiated onto the reflective surface provided on the mirror holding element 26D, and the displacement changes are measured by the laser Doppler displacement measurement method, etc. Similarly, the mirror displacement measuring means 26J measures a displacement in the Z direction of the mirror holding element 26D. In addition, the measurement light of the laser interferometer is irradiated onto the reflective surface provided on the mirror from the mirror displacement measuring means 26K and mirror displacement measuring means 26L for XYZ measurements of the mirror E, and the XYZ displacement changes of the mirror is measured by the laser Doppler displacement measurement method, etc. The above measuring means sets an origin for the mirror E relative to the mirror barrel 7F, and drives the mirror E to the geometrical design center position. The optical-axis adjustment and aberrational corrections for the total reflection mirror in the projection optical system are conducted at the geometrical design center position. By providing means that uses the laser interferometer, etc. to precisely measure the displacement measuring means of the mirror E from the projection optical system body, the mirror is driven and narrowed down to the target optical aberration using the aberrational target value as an origin through the total reflection mirror in the projection optical system. A description will now be given of the measurement method of the aberrational target value through the total reflection mirror in the projection optical system. While the reticle stage slider 6B retracts of the reticle stage 6, as shown in FIG. 5, the measurement light emitted from a measurement light source supply fiber 23A for a wave front measuring unit is emitted from the measurement light source emission opening 23 in the wave front measuring unit that emits the wave front evaluation light source light. The measurement light is reflected on the entire surface of the reflective surface on the mirror in the projection optical system, and the wave front measurement light-receiving sensor 24 installed on the wafer stage movable part 8 measures the optical wave front aberration of the projection optical system on the mirror's entire reflective surface, as illustrated. Next, a wave front measurement value arithmetic circuit calculates the wave front aberration amount based on the wave front measurement value measured by the wave front measurement light-receiving sensor. A mirror correction drive table arithmetic circuit 29 calculates corrective drive directions, drive amounts, and applied power amounts of mirrors A to E based on this wave front measurement operational value, and transmits them as target values to the mirror fine-movement correction drive means 31. Simultaneously, regarding the positional information of the mirrors A to E, the mirror system displacement measurement arithmetic circuit collects signals from the mirror displacement measuring means 26F, 26G, etc. and mirror displacement measuring means (laser interferometers) 26K, 26L, 26H, 26J, etc., and measures the mirror positions relative to the projection optical system body and the mirror barrel and relative positions among mirrors. After the fine-movement drive means 26E and rough-movement drive means 26B drive each mirror to a target position, the wave front measurement is confirmed again. When the wave front aberration meets the predefined value, the correction ends. When the wave front aberration does not meet the specification, the wave front measurement arithmetic circuit calculates the remaining wave front aberration amount again, and the above correction is repeated for narrowing down to the target specification. The target wave front aberration amount is one obtained after the projection optical system solely adjusts a mirror position initially, and narrows down the aberration below the appropriate target amount. This aberration amount is an origin of the target aberration and mirror position shape in the apparatus. It is possible to narrow down the aberration close to the target position by driving the mirror using the rough-movement drive means. While the instant embodiment drives the mirror using two members, i.e., the rough-movement drive means and the fine-movement drive means, the number of members is not limited to two and three or more drive means can be used to drive the mirror relative to the mirror barrel body. While the instant embodiment uses two types of measuring means, i.e., the rough-movement measuring means and the fine-movement measuring means, to measure mirror positions relative to the mirror barrel body, the mirror position relative to the mirror barrel body can be measured directly: The position measuring means provided on the mirror barrel body can be used to measure the mirror position. Alternatively, a position measuring means is provided on a stool in the exposure apparatus so as to measure a mirror position. Since it is conceivable that the wave front aberration changes according to the temperature and other conditions in the exposure space, it is preferable to measure the wave front aberration regularly and drive the mirror based on the measurement result. If necessary, a wafer can be exposed by driving the mirror. For driving of the mirror, the rough-movement drive means and the fine-movement drive means are provided in a direction in which the wave front aberration sensitively changes as the mirror drives. When the wave front aberration changes are insensitive to the driving of the mirror, only the rough-movement drive means can be provided. While the instant embodiment measures the wave front aberration on the exposure apparatus body, the wafer is exposed on the regular basis, and the mirror may be driven based on the exposure result. A predicted value of a change of the wave front aberration is stored as data in advance, and the mirror may be driven based on the stored changes of the wave front aberration. Referring to FIGS. 8 and 9, a description will now be given of an embodiment of a device fabricating method using the above exposure apparatus. FIG. 8 is a flowchart for explaining a fabrication of devices (i.e., semiconductor chips such as IC and LSI, LCDs, CCDs, etc.). Here, a description will be given of a fabrication of a semiconductor chip as an example. Step 1 (circuit design) designs a semiconductor device circuit. Step 2 (mask fabrication) forms a mask having a designed circuit pattern. Step 3 (wafer making) manufactures a wafer using materials such as silicon. Step 4 (wafer process), which is referred to as a pretreatment, forms actual circuitry on the wafer through photolithography using the mask and wafer. Step 5 (assembly), which is also referred to as a post-treatment, forms into a semiconductor chip the wafer formed in Step 4 and includes an assembly step (e.g., dicing, bonding), a packaging step (chip sealing), and the like. Step 6 (inspection) performs various tests for the semiconductor device made in Step 5, such as a validity test and a durability test. Through these steps, a semiconductor device is finished and shipped (Step 7). FIG. 9 is a detailed flowchart of the wafer process in Step 4. Step 11 (oxidation) oxidizes the wafer's surface. Step 12 (CVD) forms an insulating film on the wafer's surface. Step 13 (electrode formation) forms electrodes on the wafer by vapor disposition and the like. Step 14 (ion implantation) implants ion into the wafer. Step 15 (resist process) applies a photosensitive material onto the wafer. Step 16 (exposure) uses the exposure apparatus 200 to expose a circuit pattern on the mask onto the wafer. Step 17 (development) develops the exposed wafer. Step 18 (etching) etches parts other than a developed resist image. Step 19 (resist stripping) removes disused resist after etching. These steps are repeated, and multilayer circuit patterns are formed on the wafer. The device fabrication method of this embodiment may manufacture higher quality devices than the conventional one. Thus, the device fabrication method using the exposure apparatus, and the devices as finished goods also constitute one aspect of the present invention. According to the instant embodiment, the exposure apparatus can correct fine displacements and inclinations of the rotational axis in the in-plane translation shift direction, mirror's deformations due to its own weight, and wave front aberration in the projection optical system mirrors, preventing the mirror surface precision and thus the optical aberration, and deteriorated imaging performance and lowered light intensity in the projection optical system.
	abstract	The invention relates to the use of Dispersion Ceramic Micro-Encapsulated (DCM) nuclear fuel as a meltdown-proof, accident-tolerant fuel to replace uranium dioxide fuel in existing light water reactors (LWRs). The safety qualities of the DCM fuel are obtained by the combination of three strong barriers to fission product release (ceramic coatings around the fuel kernels), highly dense inert ceramic matrix around the coated fuel particles and metallic or ceramic cladding around the fuel pellets.
	description	The present invention relates generally to nuclear power plants and more particularly to nuclear power plants with boiling water reactors (BWRs). Oxidation caused by oxidants such as oxygen and hydrogen peroxide in the cooling water of BWRs can cause stress corrosion cracking (SCC) on reactor components, and cause corrosion of the zirconium alloy cladding of the fuel pins. Known SCC reduction strategies for reactor components such a reactor pressure vessel (RPV) internals of BWRs include noble metals chemistry addition (NMCA) and hydrogen water chemistry (HWC). These SCC reduction strategies can reduce SCC of internals, but may increase corrosion of the fuel pin cladding. NMCA methods are described for example in U.S. Pat. Nos. 5,818,893, 5,904,991 and 6,793,883, and as defined herein NMCA may include both off-line or on-line addition of noble metals. HWC used with NMCA is described in the background section of U.S. Patent Application Publication No. 2005/0018805, which then also describes application of a reductive nitrogen compound such as hydrazine in combination with hydrogen. Application of an alcohol such as methanol with hydrazine is also disclosed. DE 100 30 726 describes application of photocatalytic substances for depositing on reactor internals, and mentions the injection of hydrogen or methanol. Application of methanol to reduce oxidizing effects is described in U.S. Patent Application Publication No. 2005/0135542, as well as in the publication entitled “Study of the Methanol Injection in Reactor Water in Reactor Water of BWR Plants” by Saneshige et al. presented at the Proc. Symp. on Water Chemistry and Corrosion of Nuclear Power Plants in Asia, Oct. 11-13, 2005, Gyeongju, Korea. Shut down and then start-up of BWRs can occur at regular intervals, for example, every one to three years, and present very different water chemistries than during normal operation. As discussed, for example, in the publication “The First Application of Hydrogen Water Chemistry during Start-up for Mitigation of SCC initiation in Tokai-2 BWR” by Takiguchi and Otoha, oxidant concentration in reactor water is high during start-up. The publication describes the addition of HWC during start-up via an additional hydrogen injection system. U.S. Patent Publication No. 2005/0018805 also describes adding hydrogen and a reductive nitrogen compound during start-up and shut-down. While the use of HWS during start up is known, this procedure has several drawbacks. Contact of the hydrogen with air pockets which can be present during start-up needs to be avoided to prevent the possibility of an explosive reaction. In addition, the use of HWC during start-up has several other drawbacks listed in the publication “The First Application of Hydrogen Water Chemistry during Start-up for Mitigation of SCC initiation in Tokai-2 BWR” such as not being able to maintain a proper target concentration during certain start-up operation conditions. The varying conditions during both start-up and shut-down, such as a varying amount of coolant circulating through the reactor and varying reactor temperatures, thus make proper dosing of gaseous hydrogen exceedingly difficult. Furthermore, hydrogen injection may not be possible at other times, for example during repair or malfunction. Noble metals often also begin to have reduced effect as catalysts during certain periods, for example, when operating at temperatures of 250 degrees Celsius or below and when hydrogen injection is not available. During start-up and shut-down temperatures fall below these levels and hydrogen injection is complex. An object of invention is to reduce oxidation and reduce corrosion on reactor core components and fuel claddings during periods when HWC is not available or desirable, for example, during malfunction of the HWC injection system or start-up and shut-down of the reactor. Another alternate or additional object of the present invention is to provide for safer, less expensive and/or less complex oxidation reduction. The present invention provides a nuclear power plant including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system, and an alcohol injection system connected to the reactor cooling system. By providing a separate alcohol injection system in addition to the hydrogen injection system, the use of hydrogen or alcohol for various operating regimes, for example, during start-up and shut-down as opposed to normal operating conditions, advantageously can be controlled. Alcohol as well advantageously can be provided as a back-up injection system at times when the hydrogen injection system is inoperable or otherwise not available. The use of alcohol, which can be provided in liquid form and does not explode when in contact with air pockets, also advantageously improves safety, requires less careful monitoring and requires less complex delivery equipment. Additional components, such as a reductive nitrogen compound injection system, and NMCA injection system advantageously can be provided as well. The present invention also provides a method for operating a nuclear power plant including providing an alcohol to a BWR during start-up or shut-down operation. Alcohol has typically not been considered suitable for start-up or shut-down, since, as described in U.S. Patent Publication No. 2005/0018805, the radiation levels during start-up and shut down were at lower intensities and alcohols were generally considered useful when gamma radiation was present. However, the present invention recognizes that using alcohol instead of HWC during start-up or shut-down advantageously permits some oxidation reduction, while reducing dangers related to explosions and simplifying equipment requirements. The additional alcohol injection system may be added on to existing BWR nuclear power plants or provided to new power plants, and may be connected to the primary coolant in the reactor via, for example, the reactor water clean up system (CUW), emergency core cooling system (ECCS), primary loop recirculation system (PLR), or control rod drive cooling system (CLD). Advantageously, hydrazine or another reductive nitrogen compound can be provided to the BWR primary cooling circuit at the same time as the alcohol and hydrogen to improve the overall oxidation reduction performance of the additives. The present invention also provides a method for operating a nuclear power plant including providing both hydrogen and an alcohol to a BWR primary cooling circuit during normal operation of the power plant. In the past, methanol was seen as an alternative to hydrogen. By providing both hydrogen and methanol, the mixture of hydrogen and methanol can be altered as desired to provide adequate oxidation reduction. Moderate negative electrochemical potential values can be achieved (slightly below −240 mV), while still providing oxygen reducing species of both types. Moreover, back-up protection can thus be provided, for example when the amount of stored hydrogen for HWC is running low, or the hydrogen injection fails, methanol can be dosed with the hydrogen to reduce the amount of hydrogen being used. FIG. 1 shows schematically one preferred embodiment of a BWR nuclear power plant 10 according to the present invention. Nuclear power plant 10 has a BWR 20 having a core 22 with a plurality of fuel rods having, for example, uranium dioxide fuel in zirconium alloy tubes, a downcomer 26 and a riser 28, shown schematically. Downcomer 26 is connected to a primary coolant inlet 24, which receives water from a primary coolant line 30. Core 22 heats water from the primary coolant line 30, which has a hot leg 32 providing steam to a turbine 40 driving a generator 42. A condenser 44 condenses the steam into water, which exits the condenser into a cold leg 34 of coolant line 30, and is pumped back to the inlet 24 and downcomer 26. A control rod drive cooling system (CLD) 50 can be provided to cool a drive for control rods inserted into core 22. An HWC hydrogen injector 60 can inject hydrogen into cold leg 34, and an NMCA injector 62 can provide a noble metal for example, platinum, into cold leg 34 as well, so that the reactor 20 experiences HWC and NMCA during normal periods of operation. An ECCS 70, PLR 80 and CUW 90, all shown schematically, may also be provided. The specific embodiment of FIG. 1 provides a methanol injector 100 and a hydrazine injector 110, which are connected to CUW 90 for example, downstream of a CUW pump. Methanol and hydrazine thus can be injected into water entering cold leg 34 of primary coolant line 30 via CUW 90, and can circulate through reactor 20. Cooled down CUW sample lines can also be used to receive injector 100, 110 fluids. Injectors 100, 110 also could be located at the ECCS 70, PLR 80 or CLD 50. Alternately, an existing injector, such as one used for sodium-24 injection and connected to CUW 90, could also be modified to use as an injector 100 or 110. The preferred location is past any resin beds. The amount of methanol and hydrazine injected can be controlled by a controller 120, which can receive an input from one or more sensors 130, for example, monitoring the concentration of methanol, hydrazine or hydrogen in downcomer 26. Pre-set quantities of methanol can be injected considering for example hydrogen values in downcomer 26. Sensors 130 can also measure an electrochemical potential (ECP) of one or more components of reactor 20 or core 22, for example the cladding of the fuel pins. FIG. 2 shows one embodiment of methanol injector 100 and hydrazine injector 110, which can include tanks 104, 114, respectively, connected to piping of CUW 90. Metering pumps 102, 112 can release, respectively, methanol and/or hydrazine into CUW 90, and can be controlled individually via controller 120. Controller 120 can be the same controller controlling start-up and shut-down operations or a separate controller. Alternatively, in a gravity-assisted injector embodiment, the amount of methanol or hydrazine provided can be controlled by turning on and off the valves, with the tanks suspended above the piping. While methanol is preferred, other alcohols may be delivered, including ethanol and propanol, which are also preferred. However, formic acid, formaldehyde and acetaldehyde are other examples of suitable alcohols. While hydrazine is the preferred reductive nitrogen compound, other compounds such as ammonia (NH3) are also possible. In one preferred method of the present invention, alcohol is injected into the primary coolant to establish an alcohol concentration from 0.1 to 300 μmol/kg (≈0.0032 to 9.6 ppm for methanol) in downcomer 26 during start-up or shut down, or during another time when HWC 60 is not operating, for example when HWC is malfunctioning. Most preferably, the alcohol concentration is less than 10 μmol/kg. The hydrazine also is preferably provided during shut-down and start-up into the primary coolant to establish a maximum hydrazine concentration of 300 μmol/kg in downcomer 26. In another preferred method of the present invention, it is advantageous to use both hydrogen and methanol injection during normal operation. It is also possible to use all of the hydrogen-containing chemical injectors, such as for hydrogen, methanol and hydrazine, at the same time. The various hydrogen containing chemical injectors can be phased in or phased off due to various phases of operation. Preferably, the combination of the hydrogen containing chemicals added should, stoichimetrically equivalent with respect to hydrogen, be maintained around or below the equivalent of 10 μmol/kg methanol. The metering-in of hydrogen or alcohol or hydrazine, after exceeding a certain limit may lead to several disadvantageous results, and thus most preferably the limit of all of the injected hydrogen-containing chemicals does not exceed the equivalent of 300 μmol/kg methanol FIG. 3 presents the quantities of methanol to be injected in feed-water of a 900 MW electric standard BWR during normal operation so that the effect in the reactor core would be the one provided by an injection in feed water of 1.2 ppm hydrogen. It is an illustration of stoichimetric calculations involving only two components: methanol and hydrogen. Diffusion of hydrogen in the downcomer and carry over of hydrogen by steam are considered in the calculations of FIG. 3. Column 1 thus is the feed-water concentration FC of hydrogen provided by HWC 60, column 2 the predicted downcomer concentration RC during normal operation, and column 3 the predicted core concentration CC of hydrogen during normal operation. Column 4 shows the differential hydrogen predicted as missing from the core as the hydrogen concentration in the feed water decreases and column 5 the equivalent methanol which must be provided to compensate. Columns 6, 7 and 8 then show the desired methanol concentrations in the feed water, downcomer and core respectively to stoichimetrically compensate for the reduced hydrogen injection by HWC 60. Similar calculations can be performed when three or more components are involved and one is replacing the other in successive steps (when one decreases another one increases). In addition, methanol is more effective than hydrogen in reduction hydrogen peroxide concentrations and in producing negative ECP values, as discussed in, for example, the publication “Reasons and Criteria for Selection of Methanol as an Alternative to Hydrogen for BWR Plants” presented at the 5th International Workshop on LWR Coolant Water Radiolysis and Electrochemistry, San Francisco, October 2004 by Bernhard Stellwag and Wilfried Ruehle, hereby incorporated by reference herein. Experimental data in radiation field shows that the hydrogen peroxide is about a factor of two lower with methanol than with hydrogen at the same molal oxygen ratios. Also, the methanol is about three times more effective than hydrogen to reach the same ECP as a function of oxygen molal ratio. Sensors 130 could be provided to ensure exact downcomer concentrations RC via monitoring of the hydrogen and the methanol concentrations in the downcomer 26. Examples of preferred mixed use, with three injectors 100, 110, 60 include: A. injection of methanol at a constant level, while increasing hydrazine injection and decreasing hydrogen injection (shut-down); B. methanol and hydrazine injection both increasing while hydrogen injection is absent (beginning of start-up); C. methanol injected at a constant level with hydrazine still increasing and hydrogen injection absent (end of start-up); D. methanol injection constant with hydrazine decreasing and hydrogen increasing (beginning of normal operation of HWC or NMCA); E. methanol injection increasing or decreasing to compensate for loss of hydrogen injection during normal operation; and F. methanol injection increasing or decreasing to compensate for effects on ECP during on-line NMCA injection. In FIG. 4, a large body of research is summarized showing that zirconium hydrogen pick-up (HPU) or hydrogen pick-up fraction (HPUF) and oxidation increases dramatically, if the ECP of the fuel pin becomes cathodic (negative) and goes below −300 mV. ECP levels of −300 mV or higher thus are preferred as far as fuel operation is concerned. The concentration in feed-water of the combinations of all reducing agents (hydrogen+methanol+hydrazine) applied to a BWR can be such to provide a reducing environment characterized by an ECP potential against the Pt (platinum) reference electrode around or higher than −300 mV (e.g. −260 mV). To protect both the BWR reactor internals and also maintain an increase fuel operation margin, ECP potential against the Pt (platinum) reference electrode around or higher than −300 mV (e.g. −260 mV), preferably should be maintained, rather than −500 mV usually imposed by operators to minimize the SCC of internals. In the present invention, it is preferable to use methanol. Alternatively, other suitable alcohols may be used including, for example, ethanol or propanol, or a mixture of any of these. 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.
	abstract	In a tomosynthesis system a static focal spot is moved in a direction opposite to and generally synchronized with the directional movement of an x-ray source and X-ray collimator blades are moved during each exposure in synchronization with the shifting of the static focal spot. The synchronized movement of the static focal spot, x-ray tube and collimator blades helps keep the effective focal spot fixed in space relative to the breast, detector or both during the entire duration of the exposure and keeps the x-ray field on the detector and breast static. The shifting collimator blades follow an oscillating pattern over the multiple x-ray exposures of a tomosynthesis scan.
055531075	abstract	A pressurized water nuclear reactor pressure vessel has an upper core support plate with passageways for the passage of coolant, an upper support plate above the upper core support plate, and peripheral hollow support columns extending upwardly above the passageways in the upper core support plate for supporting the upper support plate above the upper core support plate and for guiding coolant from the passageways into the plenum. Each peripheral support column has a peripherally slotted upper portion for guiding the coolant in the peripheral support column into the plenum and an unslotted lower portion.
	abstract	A radioactive waste (zeolite to which Cs-137 was adsorbed) in a waste tank and a glass raw material (soda lime glass) in a glass raw material tank are supplied into a solidifying vessel. Graphite in a graphite tank is also supplied into the solidifying vessel. The solidifying vessel is filled with a mixture of the radioactive waste, glass raw material, and graphite and is then disposed in an adiabatic vessel. The radioactive waste and glass raw material in the adiabatic vessel are heated by thermal energy generated due to radiation emitted from Cs-137. The heat is transferred to the peripheral portion of the solidifying vessel through the graphite, raising the temperature of the peripheral portion. The glass raw material is melted and enters clearances among the radioactive waste, producing a vitrified radioactive waste. This radioactive waste solidification method can shorten a time taken to produce a vitrified radioactive waste.
	description	In accordance with some embodiments of the present invention to be described below, a suction port is provided so that dust or particles such as organic matters or metal dust, for example, are discharged outwardly from the suction port, while on the other hand a fresh gas is supplied from a supply port, by which adhesion of dust to an exposure region on a mask can be prevented or reduced. Acid is produced or acid and decomposed matters are produced from a chemical amplification resist after the exposure process. Since, however, the exposure time is only about one second, for most of the period the region of the mask to be used for the exposure process is kept opposed to unexposed portions of a workpiece. Therefore, by evacuating through a suction port the portion close to the mask region to be used for the exposure, almost all the dust (acid producer or acid and decomposed matter) produced form the resist can be discharged. Adhesion of dust to the exposure region can thereby be prevented. Evacuation may be made through a suction port provided at the bottom side. This is effective to prevent a pressure difference between the top and bottom faces of a supporting film (thin film) of a mask, and to keep the flatness of the supporting film. Creation of pressure difference can be prevented by supplying a fresh gas from a supply port, while evacuation is made through a suction port. This effectively avoids adverse influence to the gap of a few tens microns between the workpiece (wafer) and the mask. Even in cases where the suction portion is provided at the top face only, the flatness may be held by adjusting the evacuation or discharging quantity so that the pressure difference is kept at 1 Pa or lower. In addition to prevention of adhesion of dust to the mask, adhesion of dust to the X-ray exposure apparatus can be avoided. This assures stability of X-ray exposure. In regard to deposition of organic matters remaining, if any, an air or oxygen may be introduced through a supply port (which may be provided in an X-ray mask structure or a member which supports the mask structure or, alternatively, which may be defined outside it) and ultraviolet rays may be projected to produce O3, with which UV/O3 cleaning may be performed. More specifically, with the irradiation of ultraviolet light of 185 nm or shorter, oxygen may be transformed into O3, and with the irradiation of ultraviolet light of 254 nm or shorter, O3 may be transformed into activated oxygen by which organic matters can be decomposed As an alternative, oxygen may be introduced through a supply port, and oxygen plasma may be produced between an electrode and another electrode which supports the mask structure, whereby oxygen plasma cleaning may be performed. This may further reduce dust deposition. During the exposure procedure, an exposure process for transferring a desired pattern of a mask onto a workpiece and a mask cleaning process for cleaning the mask structure with UV/O3 cleaning may be performed repeatedly. This may further reduce dust deposition or prevent it As regards cleaning to be performed in an exposure chamber, not only contamination of the mask but also contamination inside the exposure apparatus can be cleaned. There may be a cleaning unit for accommodating a mask structure, separately from a mask cassette chamber, which may be provided with a suction port and a supply port. Contamination of the mask may be monitored and, when a certain level is reached, the mask may be moved into the unit. Air or oxygen may be introduced through the supply port, and ultraviolet light may be projected to the mask structure, to perform UV/O3 cleaning. In that occasion, the unit may be controlled separately from the exposure ambience, and decomposed matters may be discharged outwardly from the suction port to prevent re-adhesion. This is effective to increase the efficiency of decomposition of adhered matters. In a similar structure, UV/O3 cleaning may be replaced by oxygen plasma cleaning. In an exposure apparatus having a mask cassette chamber for keeping the mask when an exposure process for workpieces is not performed, the mask cassette chamber may be used in combination with, or in place of, the aforementioned unit for cleaning the mask. This enables mask cleaning during nonuse of the mask, and accomplishes prevention of mask contamination. In accordance with an X-ray exposure method and an X-ray exposure apparatus of the present invention, high-precision and mass-production printing is enabled. The Stray mask structure, the X-ray exposure method and the X-ray exposure apparatus may be either transmission type or reflection type. Preferred embodiments of the present invention will be described below with reference to the accompanying drawings. [Embodiment 1] FIGS. 2(a)-2(d) show an X-ray mask structure and a mask chuck, according to a first embodiment of the present invention. The X-ray mask structure comprises a holding frame 1 made of Si and having a thickness 2 mm, a supporting film 2 of SiC having X-ray transmissivity and being made by CVD with a thickness 2.0 microns, and an X-ray absorptive material 3 of Ta. The holding frame 1 is formed with suction ports 6. In the case where the suction port 6 has a shape such as shown in FIG. 2(a), a mask chuck 7 for holding the X-ray mask structure inside an X-ray exposure apparatus may be provided with suction ports 8 (FIG. 2(d)) being communicated with evacuation means such as a pump. The evacuation capacity may be adjusted to a level not disturbing the flatness of the supporting film 2 or the ambience inside the exposure apparatus. In a modified form, the suction port 6 may be formed by anisotropic etching of Si and it may have taper such as shown in FIG. 2(b). The suction port nay have a shape such as shown in FIG. 2(c), and it may be connected to evacuation means directly. With the provision of suction ports for discharging dust or decomposed matters therethrough, contamination of the mask surface can be prevented or reduced and the times of cleaning operations can be reduced. This enables prolongation of the lifetime of the mask. [Embodiment 2] FIG. 3A is a sectional view of an X-ray mask structure and a mask chuck, according to a second Embodiment of the present invention. The X-ray mask structure is similar to that of the first embodiment, but there are additional suction ports at the bottom face. With the evacuation from the bottom suction ports, there is no pressure difference created between the top surface and bottom surface of the mask supporting film 2 (thin film). Thus, the flatness of the supporting film can be maintained easily, and the evacuation capacity can be enlarged. Use of the mask having suction ports 8 as described enables prevention of contamination of the mask surface and reduction in number of cleaning operations required. This effectively prolongs the lifetime of the mask. Thus, an X-ray mask structure which can meet mass-production is accomplished. [Embodiment 3] FIGS. 3B and 3C show an X-ray mask structure according to a third embodiment of the present invention. The X-ray mask structure comprises a holding frame 1 made of Si and having a thickness of 2 mm, a supporting film 2 of SiC having X-ray transmissivity and being made by CVD with a thickness of 2.0 microns, an X-ray absorptive material 3 of Ta, and a reinforcing member 4 of Pyrex glass being adhered to the holding frame 1 through anodic bonding. The X-ray mask structure is held by a mask chuck 7 inside an X-ray exposure apparatus. There are suction ports 6 and supply ports 11 extending through the holding frame 1 and the reinforcing member 4. The mask chuck 7 for holding the X-ray mask structure inside the exposure apparatus is formed with suction ports 8 and supply ports 12 being communicated with evacuation means such as a pump or an equipment for supplying fresh gases. The evacuation capacity and gas supply quantity may be adjusted to a level not disturbing the flatness of the supporting film 2 or the ambience inside the exposure apparatus. There may be a suction port 6 at the bottom face side as in the second embodiment. The X-ray absorptive material 3 may be formed on the workpiece side as shown in FIG. 3B, or it may be formed at the opposite side as shown in FIG. 3C. In the case of FIG. 3C, production of the mask is relatively difficult, but the cleaning of dust adhered on the mask""s exposure region is relatively easy. With the provision of suction ports for discharging dust or decomposed matters therethrough, contamination of the mask surface can be prevented or reduced and the times of cleaning operations can be reduced. This enables prolongation of the lifetime of the mask. [Embodiment 4] FIG. 4 is a sectional view of an X-ray mask structure and a mask chuck, according to a fourth embodiment of the present invention The X-ray mask structure comprises a holding frame 1 made of Si and having a thickness of 0.625 mm, a supporting film 2 of SiN having X-ray transmissivity and being made by CVD with a thickness of 2.0 microns, an X-ray absorptive material 3 of W formed by sputtering, and a reinforcing member 4 of SiC adhered to the holding frame 1 by an adhesive agent 5. The reinforcing member 4 of the X-ray mask structure is formed with suction ports 6. Also, the mask chuck 7 is formed with suction ports 8. Dust adhesion to, or contamination of, the mask can be prevented with this structure. [Embodiment 5] FIG. 5 is a sectional view of an X-ray mask structure and a mask chuck, according to a fifth embodiment of the present invention. The X-ray mask structure is similar to that of the fourth embodiment, but the peripheral portion of the holding frame 1 is made to slant by polishing. At this slant portion of the holding frame 1, there is a pellicle frame 9a for holding a pellicle 9. Dust adhesion or contamination from the outside can be prevented by suction ports, while the exposure pattern bearing portion can be protected by the pellicle. With the provision of both the pellicle and the suction ports, dust and contamination can be avoided, and dust adhesion to the exposure region can be prevented efficiently. [Embodiment 6] FIG. 6 is a sectional view of an X-ray mask structure and a mask chuck, according to a sixth embodiment of the present invention. The X-ray mask structure is similar to that of the third embodiment, but the suction port 6 extending through the holding frame 1 and the reinforcing member 4 is formed such as illustrated. It may be directly connected to evacuation means. Forming the connection at the reinforcing member may be easier, as compared with a case where it is formed at the holding frame. Independently of the X-ray exposure apparatus, suction can be performed at the mask solely. [Embodiment 7] FIG. 7 is a sectional view of an X-ray mask structure and a mask chuck, according to a seventh embodiment of the present invention. The X-ray mask structure is similar to that of the fourth embodiment, and the mask chuck is formed with suction ports 8. Regardless of the shape of the mask structure, suction may be performed at the X-ray exposure apparatus side. [Embodiment 8] FIGS. 8A-8C are sectional views of an x-ray mask structure and a mask chuck, according to an eighth embodiment of the present invention. The X-ray mask structure is similar to that of the first embodiment, and there are suction ports 8 and supply ports 12 formed in the mask chuck The suction port 8 and the supply port 12 may have shapes such as shown in FIGS. 8A, 8B or 8C, which are effective to perform suction at a location closer to the exposure region. When the suction port 8 or supply port 12 faces to a side opposite to the mask, as in the example of FIG. 8C, the evacuation capacity and gas supply quantity can be enlarged without adverse effect to the flatness of the supporting film. This prevents dust adhesion to the X-ray exposure apparatus more efficiently. [Embodiment 9] FIG. 9 is a sectional view of an X-ray mask structure and a mask chuck, according to a ninth embodiment of the present invention. The X-ray mask structure is similar to that of the fourth embodiment, and there are suction ports 8 formed in the mask chuck. The suction port 8 may have a shape such as shown in FIG. 9, and connection to the evacuation means may be formed at the side face of the mask chuck. [Embodiment 10] FIG. 10 is a sectional view of an X-ray mask structure and a mask chuck, according to a tenth embodiment of the present invention. The X-ray mask structure is similar to that of the first embodiment, and it has suction ports 6. The mask chuck has suction ports 8, which are concentric with the suction ports 6, and other suction ports 10, operable to perform suction independently. [Embodiment 11] Next, an embodiment of an X-ray exposure apparatus, which uses a it mask and a mask chuck according to any one of the first to tenth embodiments described above for manufacture of microdevices, such as semiconductor devices, thin film magnetic heads, or micro-machines, for example, will be explained. FIG. 11 is a schematic view of a main portion of an exposure apparatus according to an eleventh embodiment of the present invention, which uses an X-ray mask structure. Denoted in FIG. 11 at A is an SOR (synchrotron orbital radiation) ring. Synchrotron radiation light B emitted from the ring A has a sheet-like beam shape with its intensity being uniformly expanded in the lateral direction. The radiation light B is reflected by a cylindrical mirror C so that it is expanded in the vertical direction. By this, a beam of substantially quadrilateral shape in section is produced, and a quadrilateral exposure region can be defined. The expanded radiation light B is adjusted by means of a shutter D such that a uniform exposure amount is provided within an irradiation region. The radiation light passed through the shutter D is directed to an X-ray mask E. The X-ray mask E is attracted to a mask chuck G, and it is placed at a position opposed to a wafer F. The X-ray mask E and the mask chuck G are prepared in accordance with any one of the first to tenth embodiments described hereinbefore. Denoted at F is a wafer which is to be exposed. The wafer F is held by a wafer chuck H which is mounted on a wafer stage I. The wafer stage I is moved for positioning the wafer F. Alignment unit J comprises an optical system for detecting alignment marks provided on the mask E and the wafer F for positioning of them, and calculating means for calculating a deviation between them. Use of the X-ray mask E according to the present invention enables high precision alignment. After completion of alignment of the mask E and the wafer F, a pattern formed on the X-ray mask E is transferred and printed on the wafer F in accordance with the step-and-repeat method or scanning method. All dust materials such as organic matters or metals can be discharged through suction ports (not shown) and a fresh gas can be supplied through supply ports (not shown), by which adhesion of dust to the mask or the exposure apparatus can be prevented. Further, since dust matters created from a resist can be evacuated sufficiently without disturbing the flatness of the supporting film of the mask, evacuation and supply of fresh gas may be performed after sufficiently separating the mask from the wafer in response to completion of one-step exposure. This structure enables high precision X-ray exposure that can meet mass-production. [Embodiment 12] An X-ray exposure apparatus according to another embodiment of the present invention will now be described. FIG. 12A is a schematic sectional view of the mask and wafer portion of an X-ray exposure apparatus according to a twelfth embodiment of the present invention. FIG. 12B is a front view of the mask, as viewed from the wafer side. FIGS. 13A and 13B are schematic views for explaining the procedure of a step-and-repeat process upon a wafer. Denoted in these drawings at 10 are suction ports, and the direction backwardly from the sheet of the drawing corresponds to the gravity direction. As shown in FIG. 12B, there are a certain number of suction ports 10 which are disposed at the left-hand and right-hand sides of the peripheral portion of the mask structure and at the portion downwardly along the gravity direction. While in this example the suction ports are formed at the peripheral portion of the mask structure, if an X-ray mask structure according to any one of the preceding embodiments is used, they may be disposed at the same position of the suction ports of the X-ray mask structure. As shown in FIG. 12A, since the X-ray exposure apparatus generally uses synchrotron radiation as an exposure light source, the mask and the wafer are arranged and disposed vertically along the gravity direction. Further, since X-rays to be used for the exposure comprise X-rays of a relatively long wavelength, called soft X-rays, in many cases the exposure apparatus is placed in a reduced pressure ambience of helium, nitrogen or air, for example. Most of the matters, including butene, which are emitted from a resist applied to a wafer into the ambience after exposure may be molecules heavier than helium, nitrogen or air. Therefore, unless the ambience is disturbed, these matters will fall downwardly. In consideration of this, in this embodiment, suction ports 10 are provided at a downstream position along the gravity direction of the exposure mask so as to promptly discharge these materials so that they are not dispersed inside the exposure apparatus. As a result, the ambience inside the exposure apparatus can be held at high purity, and in addition to this, adhesion of emitted matters to the mask is prevented. The exposure apparatus of this embodiment may have an algorithm in which a wafer is exposed in an order from the downstream side of the gravity direction, such as illustrated in FIGS. 13A or 13B. This is effective to prevent contact of matters, emitted from the resist after being exposed, to the mask. In a case where step motion such as shown in FIG. 13A or 13B is to be executed, it may be performed in accordance with an algorithm with which the ambience gas is drawn also from those suction ports, among the suction ports provided at the opposite sides of the mask, which are opposed to a pattern whose exposure has been completed. Namely, if the step motion is to be done as in FIG. 13A, when the wafer is moved stepwise rightwardly, the ambience gas may be drawn through those suction ports at the left-hand side and downstream side of the mask. If the step motion is to be done as in FIG. 13B, the effect of the present invention can be accomplished only by the provision of suction ports at the left-hand and lower sides of the mask. In an X-ray exposure apparatus arranged so that a resist is exposed in atmosphere, depending on the matters emitted from the resist after its exposure, the matters may be heavier than or lighter than the atmosphere. In order to meet both cases, air may be blown to the mask and the wafer from an upstream position with respect to the gravity direction (down flow blowing), and air may be drawn through suction ports at a downstream position of the mask along the gravity direction. Additionally, the suction ports may be provided at an upstream position. [Embodiment 13] FIG. 14 is a front view of a mask portion of an exposure apparatus according to a thirteenth embodiment of the present invention, as viewed from the wafer side. Suction ports 10 and supply ports 12 are formed alternately in a portion of a mask chuck G, around the periphery of a mask E. They may be formed around the whole periphery of the mask as illustrated, or alternatively, they may be provided only in a portion of it. [Embodiment 14] FIG. 15 is a schematic view of a main portion of an X-ray exposure apparatus according to a fourteenth embodiment of the present invention. Like numerals as those of the preceding embodiments are assigned to corresponding elements. The X-ray exposure apparatus has a structure similar to that of the embodiment of FIG. 11, but suction ports 10 are provided not only in a portion close to the exposure region but also in a mask cassette K (suction ports 10c) for accommodating mask structures E. This enables prevention of dust or contamination also during preservation of the mask structures. [Embodiment 15] FIG. 16 is a schematic view of a main portion of an X-ray exposure apparatus according to a fifteenth embodiment of the present invention. Denoted at A is a SR radiation source, and denoted at B is synchrotron radiation light. Denoted at C is a convex surface mirror, and denoted at D is a shutter. Denoted at E is an X-ray mask. Denoted at F is a wafer, and denoted at G is a mask stage. Denoted at H is a wafer chuck, and denoted at J is an alignment unit. Denoted at K is an auxiliary light source inside an exposure chamber O. Denoted at M is an evacuation port, and denoted at Mxe2x80x2 is a supply port. The structure of the X-ray exposure apparatus is similar to that of the FIG. 11 embodiment, but there is an auxiliary light source K so that not only the mask but also the mask stage inside the exposure apparatus, for example, can be illuminated. The auxiliary light source K may comprise a lamp or laser which can emit wavelength not longer than 185 nm. The mask and the mask supporting member may be provided with supply ports and suction ports (not shown) which are communicated with the supply port Mxe2x80x2 and suction port M of the exposure chamber, respectively. Through the supply port Mxe2x80x2, oxygen or air can be supplied into the exposure chamber O, and, in response to irradiation of ultraviolet rays, UV/O3 cleaning can be performed to any organic deposits on the mask or inside the exposure apparatus. With the provision of the suction port M, decomposed matters of the deposited materials can be discharged outwardly such that the re-deposit of intermediate products of decomposed matters can be prevented. In place of using an auxiliary light source of lamp or laser of 185 nm or shorter, O3 gas may be directly introduced through the supply port. By using an X-ray exposure apparatus wherein the mask can be cleaned in accordance with a UV/O3 cleaning procedure, a high precision exposure process that can meet mass production is accomplished. [Embodiment 16] FIG. 17 is a schematic view of a main portion of an X-ray exposure apparatus according to a sixteenth embodiment of the present invention. The structure of the X-ray exposure apparatus is similar to that of the embodiment of FIG. 11, but there is a separate chamber (unit) N in addition to the exposure chamber O and the mask cassette (not shown). In this additional chamber N, there is an auxiliary light source K. While not shown in the drawing, there is also an additional auxiliary light source provided in the exposure chamber O, as in the fifteenth embodiment. As the exposure process is repeated within the exposure chamber O, the X-ray transmission factor or transmission factor to alignment light may be decreased due to adhesion of any matters, the exposure operation may be discontinued and the mask E may be moved into the unit N. Through the supply port Mxe2x80x2, oxygen or airs may be introduced into the unit N and ultraviolet light is irradiated. In response, UV/O3 cleaning can be performed to organic deposition matters on the mask. The mask and the mask supporting member may be provided with supply ports and suction ports being communicated with the supply port Mxe2x80x2 and suction port M of the unit N. In accordance with an X-ray exposure apparatus having a unit N for performing UV/O3 cleaning as described above, high precision X-ray exposure that can meet mass production is accomplished. [Embodiment 17] FIG. 18 is a schematic top view of a main portion of an X-ray exposure apparatus according to a seventeenth embodiment of the present invention. The structure of the X-ray exposure apparatus is similar to that of the embodiment of FIG. 11, but there is an auxiliary light source K disposed in a mask cassette chamber L. For masks under preservation, oxygen or air may be supplied into the mask cassette chamber L through the supply port Mxe2x80x2. In response to irradiation of ultraviolet light, UV/O3 cleaning is performed on organic matters deposited on the mask. The mask and the mask supporting member may be provided with supply ports and suction ports concentric with the supply port and suction port of the mask cassette chamber L, respectively. In accordance with an X-ray exposure apparatus wherein UV/O3 cleaning of a mask can be performed in the mask cassette chamber L, such as described above, high precision X-ray exposure that can meet mass production is accomplished. [Embodiment 18] FIG. 19 is a schematic view of a main portion of an X-ray exposure apparatus according to an eighteenth embodiment of the present invention. The structure of the X-ray exposure apparatus is similar to that of FIG. 11, but, like the sixteenth embodiment, there is a chamber (unit) N in addition to the exposure chamber O and the mask cassette (not shown). Disposed in the unit N are opposed electrodes P, wherein the X-ray mask is disposed at one of them. The electrodes are connected to a high frequency voltage source Q. As the exposure process is repeated within the exposure chamber O, the X-ray transmission factor or transmission factor to alignment light may be decreased due to adhesion of any matters, the exposure operation may be discontinued and the mask E may be moved into the unit N. Through the supply port Mxe2x80x2, oxygen or air may be introduced into the unit N and, with the application of a high frequency voltage, oxygen plasma is produced therein by which organic matters deposited on the mask can be removed by ashing. The mask and the mask supporting member may be provided with supply ports and suction ports that are concentric with the supply port Mxe2x80x2 and suction port M of the unit N, respectively. In accordance with an X-ray exposure apparatus having a unit N for performing oxygen plasma cleaning as described above, high precision X-ray exposure that can meet mass production is accomplished. [Embodiment 19] FIG. 20 is a schematic top view of a main portion of an X-ray exposure apparatus according to a seventeenth embodiment of the present invention. The structure of the X-ray exposure apparatus is similar to that of FIG. 11, but there are opposed electrodes P disposed inside the mask cassette chamber L, wherein the mask is held at the electrode position. These electrodes are connected to a high frequency voltage source Q. For a mask under preservation, oxygen or air may be supplied into the mask cassette chamber L through the supply port Mxe2x80x2. In response to application of a high frequency voltage, oxygen plasma is produced so organic matters deposited on the mask can be removed by ashing. The mask and the mask supporting member may be provided with supply ports and suction ports that are concentric with the supply port and suction port of the mask cassette chamber L, respectively. In accordance with an X-ray exposure apparatus wherein oxygen plasma cleaning of a mask can be performed in the mask cassette chamber L, such as described above, high precision X-ray exposure that can meet mass production is accomplished. [Embodiment 20] Next, an embodiment of a semiconductor device manufacturing method which uses an exposure apparatus according to any one of the preceding embodiments, will be explained. FIG. 21 is a flow chart of procedure for manufacture of microdevices such as semiconductor chips (e.g. ICs or LSIs), liquid crystal panels, or CCDs, for example. Step 1 is a design process for designing a circuit of a semiconductor device. Step 2 is a process for making a mask on the basis of the circuit pattern design. Step 3 is a process for preparing a wafer by using a material such as silicon. Step 4 is a wafer process which is called a pre-process wherein, by using the so prepared mask and wafer, circuits are practically formed on the wafer through X-ray lithography. Step 5 is an assembling step, called a post-process, wherein the wafer, having been processed by step 4, is formed into semiconductor chips. This step includes an assembling (dicing and bonding) process and a packaging (chip sealing) process. Step 6 is an inspection step that carries out an operation check, a durability check, and so on for the semiconductor devices provided by step 5. With these processes, semiconductor devices are completed and they are then shipped (step 7). FIG. 22 is a flow chart showing details of the wafer process. Step 11 is an oxidation process for oxidizing the surface of a wafer Step 12 is a CVD process for forming an insulating film on the wafer surface. Step 13 is an electrode forming process for forming electrodes upon the wafer by vapor deposition. Step 14 is an ion implanting process for implanting ions to the wafer. Step 15 is a resist process for applying a chemical amplification type resist to the wafer. Step 16 is an exposure process for printing, by exposure, the circuit pattern of the mask on the wafer through an exposure apparatus according to any one of the eleventh to nineteenth embodiments described above A wafer is loaded and disposed opposed to a mask. Any deviation between them is detected by means of an alignment unit, and then the wafer stage is moved to perform alignment of them. After completion of alignment, exposure process is performed. After the exposure, the wafer is moved stepwise to the position for exposure of a subsequent shot, and the procedure including and following the alignment operation is repeated. Step 17 is a developing process for developing the exposed wafer. Step 18 is an etching process for removing portions other than the developed resist image. Step 19 is a resist separation process for separating the resist material remaining on the wafer after being subjected to the etching process. By repeating these processes, circuit patterns are superposedly formed on the wafer. With these processes, high density microdevices can be manufactured. In accordance with the embodiments of the present invention as described hereinbefore, suction ports are provided, either in the X-ray mask structure at positions other than the portion to be used for the exposure, or in an X-ray exposure apparatus, for outwardly discharging dust matters such as organic matters or metal matters. As a result, adhesion of dust to the mask or inside the exposure apparatus can be prevented effectively. Use of suction ports and supply ports enables increased evacuation capacity while maintaining the flatness of a supporting film. This enhances the evacuation effect. Discharging dust through suction ports effectively prevents contamination of the mask surface, and it leads to decreased cleaning operations and also to prolongation of lifetime of the mask. Further, the exposure stability is improved, and maintenance of exposure apparatus is made easier. While the invention has been described with reference to the structures disclosed herein, it is not confined to the details set forth and this application is intended to cover such modifications or changes as may come within the purposes of the improvements or the scope of the following claims.
	claims	1. An X-ray optical transmission grating of a focus-detector arrangement of an X-ray apparatus for generating at least one of projective and tomographic phase contrast recordings of a subject, comprising:at least one wafer including a multiplicity of grating bars and grating gaps, the grating bars and grating gaps being arranged periodically on at least one surface of the at least one wafer, wherein the X-ray optical transmission grating includes at least two sub-gratings arranged in direct succession in a beam direction, wherein a beam profile of a transmitted X-radiation is designed to be at least one of fan-shaped and conical, and the sub-gratings arranged successively in the beam direction comprise different grating periods, and wherein a grating period increases from at least one sub-grating to at least one subsequent sub-grating and the sub-gratings are arranged to be mutually aligned, so that the rays of the beam pass through only one of either of the grating gaps or the grating bars. 2. An X-ray optical transmission grating of a focus-detector arrangement of an X-ray apparatus for generating at least one of protective and tomographic phase contrast recordings of a subject, comprising:at least one wafer including a multiplicity of grating bars and grating gaps, the grating bars and grating gaps being arranged periodically on at least one surface of the at least one wafer, wherein the X-ray optical transmission grating includes at least two sub-gratings arranged in direct succession in a beam direction, wherein a filler material with a relatively higher linear attenuation coefficient than the wafer material in the relevant energy range is arranged in the grating gaps of at least one sub-grating, wherein the transmission grating is a phase grating and, for the sum of the sub-gratings, the height of the filler material in the gaps is dimensioned so that the X-radiation with the energy used for measuring the phase shift generates a phase shift of λ/2 in the X-radiation and after the entire grating, at least in relation to the energy used for measuring the phase shift, the attenuation of the X-radiation is the same after passing through the bars and when passing through the filler material. 3. An X-ray optical transmission grating of a focus-detector arrangement of an X-ray apparatus for generating at least one of projective and tomographic phase contrast recordings of a subject, comprising:at least one wafer including a multiplicity of grating bars and grating gaps, the grating bars and grating gaps being arranged periodically on at least one surface of the at least one wafer, wherein the X-ray optical transmission grating includes at least two sub-gratings arranged in direct succession in the beam direction, wherein a filler material with a relatively higher linear attenuation coefficient than the wafer material in the relevant energy range is arranged in the grating gaps of at least one sub-grating, wherein the transmission grating is a phase grating and, for each of the sub-gratings individually, the height of the filler material in the gaps is dimensioned so that the X- radiation with the energy used for measuring the phase shift generates a phase shift in the X-radiation of λ/2 and after each sub-grating, at least in relation to the energy used for measuring the phase shift, the attenuation of the X-radiation is the same when passing through the bars and after passing through the filler material. 4. The transmission grating as claimed in claim 1, wherein at least two sub-gratings arranged on a separate wafers comprise markings by which mutual alignment is made possible. 5. A focus-detector arrangement of an X-ray apparatus for generating at least one of projective and tomographic phase contrast recordings of a subject, wherein at least one of X-ray optical gratings used is designed as the transmission grating of claim 1. 6. The transmission grating as claimed in claim 2, wherein at least two sub-gratings arranged on a separate wafers comprise markings by which mutual alignment is made possible. 7. The transmission grating as claimed in claim 3, wherein at least two sub-gratings arranged on a separate wafers comprise markings by which mutual alignment is made possible. 8. A focus-detector arrangement of an X-ray apparatus for generating at least one of projective and tomographic phase contrast recordings of a subject, wherein at least one of X-ray optical gratings used is designed as the transmission grating of claim 2. 9. A focus-detector arrangement of an X-ray apparatus for generating at least one of projective and tomographic phase contrast recordings of a subject, wherein at least one of X-ray optical gratings used is designed as the transmission grating of claim 3. 10. An X-ray system for generating projective phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 1. 11. An X-ray system for generating projective phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 2. 12. An X-ray system for generating projective phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 3. 13. An X-ray C-arc system for generating projective or tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 1. 14. An X-ray C-arc system for generating projective or tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 2. 15. An X-ray C-arc system for generating projective or tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 3. 16. An X-ray computer tomography system for generating tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 1. 17. An X-ray computer tomography system for generating tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 2. 18. An X-ray computer tomography system for generating tomographic phase contrast recordings, comprising the X-ray optical transmission grating of the focus-detector arrangement of claim 3.
	description	This application is a continuation of U.S. patent application Ser. No. 13/030,740, filed Feb. 18, 2011, now U.S. Pat. No. 8,767,902 which claims priority to U.S. Provisional Patent Application No. 61/306,754, filed Feb. 22, 2010; the content of which is incorporated by reference herein in its entirety. This application incorporates by reference in its entirety U.S. patent application Ser. No. 12/696,851, filed Jan. 29, 2010, now U.S. Pat. No. 8,571,167; the content of which is incorporated by reference herein in its entirety. The Government has certain rights in the invention pursuant to Work for Others Agreement No. 854V0. The present invention relates to nuclear power plants, and, more particularly, to fast neutron spectrum, sodium cooled reactors with metallic fuel. World electricity demand is expected to as much as double by 2030 and quadruple by 2050. The world electricity demand increase is forecasted to come from developed countries but to an even larger extent from developing countries. To meet rapid growth in developing countries, nuclear energy should be packaged in a configuration tailored to meet their specific needs. A fast neutron spectrum, sodium cooled reactor with metallic fuel is described. FIG. 1 illustrates an exemplary Small Modular Reactor (“SMR”) system 501 of the present invention. The SMR system may include a uranium-fueled core 503. The core composition may be enriched (<20%) uranium/zirconium alloy for the initial core and recycled uranium/transuranic zirconium for subsequent cores. Uranium 235/thoruim/zirconium alloys may also be used in some embodiments. The core 503 may be submerged in a tank 505 of ambient pressure liquid sodium 507. The tank 505 may be thin-walled stainless steel, and may be sized for shipment by barge or rail. The tank 505 may be positioned in a guard vessel 517 and a deck 521 of the tank 505 that may be enclosed by a removable dome 519. The guard vessel 517 and dome 521 together may create a containment 523. The SMR system 501 may be encased in a concrete silo 515. The core 503 and its containment 523 may be emplaced in a concrete silo with a concrete cover. The silo and its cover may create a shield structure to protect the reactor system 501 and the containment 523 from external hazards. The shield structure and/or the containment 523 and reactor 503 may be seismically isolated. The SMR system 501 may also include control rods 513. The liquid sodium 507 from the tank 505 may be pumped by one or more pumps 509 through the core 503 to carry heat away from the core 503. The liquid sodium 507 may carry the heat to one or more sodium to sodium heat exchangers 511. The liquid sodium 507 may be heated from about 350° Celsius to about 510° Celsius. FIG. 2 shows the SMR system 501 within a larger energy generation system 601. The heated sodium 507 may pass through the heat exchanger 511 to heat secondary sodium, which in turn passes through a secondary heat exchanger 603 where the secondary sodium heats supercritical (almost liquid) carbon dioxide. The supercritical CO2 is compressed to 21 MPa, just above its critical point at approximately 7 MPa and approximately 31° C. It is then recuperated to ˜350° C. in regenerative heat exchangers 609; then further heated to ˜500° C. in the Na-to-CO2 heat exchanger. The recuperation and compression of a nearly-liquid fluid allow for an approximately 40% energy conversion at a relatively low temperature compared to ideal gas Brayton cycles. The heated supercritical carbon dioxide may then be used to spin a gas turbine 605 to make electricity in an electrical generator 608 in a carbon dioxide Brayton cycle building 607. The turbine 605 and compressor 606 rotating machinery is very compact owing to the high density of the CO2. “Printed circuit” heat exchangers used for recuperations and for sodium to supercritical carbon dioxide heat exchange 603 are of extremely high power density. Altogether the supercritical CO2 Brayton cycle is much more compact than comparable Rankine steam cycle energy converters. The Brayton cycle may provide the SMR a thermal efficiency (heat energy converted to electricity) of approximately 39% to approximately 41% or more, an efficiency much higher than conventional light water reactor (“LWR”) steam driven turbines. Furthermore, in certain embodiments of the present invention waste heat can be used to meet lower-temperature needs, such as space heating, water desalination, industrial process heat, or can be dissipated through cooling towers. Small sodium-cooled fast reactors may demonstrate important inherent safety characteristics. These reactors may operate with simplified, fail-safe controls that may facilitate rapid licensing by regulatory authorities. For example, in response to an accident condition, such as loss of coolant flow, overcooling in the heat exchanger, control rod runout or loss of ability to reject heat, embodiments of the reactor may shut themselves down without human or safety-system intervention. For instance, as the reactor coolant heats up, the core structures may thermally expand causing increased neutron leakage from the core, in turn causing power levels to decrease in a self-correcting fashion. SMR operation requirements may be significantly simpler than conventional nuclear systems due to a characteristic that allows the reactor to innately follow load requirements brought upon by varying levels of electricity demand. Metal alloy fuel is well demonstrated, both from performance and fabrication perspectives, and can straightforwardly meet long refueling time interval requirements. Additionally, a cermet fuel may be used, while the cermet fuel none-the-less retains metallic alloy fuel attributes. The reactor core may have a long life, up to about 20 years or more. The reactor may not have or require permanent onsite refueling equipment or fuel storage capability. Refueling may be done by an outside service provider that brings refueling equipment with a new core, changes the core out, and takes both used core and refueling equipment away when completed. Fuel handling and shipping can commence at a very short time after reactor shutdown owing to the derated specific power (kwt/kg fuel). One or more multi-assembly clusters in a reactor core may have derated specific power (kwt/kg fuel) for enabling long refueling intervals while remaining in the existing fuels database. This may also enable refueling operations very shortly after reactor shutdown. Refueling operations may start within approximately two weeks of overall reactor shutdown, and may finish within approximately 1 month of overall reactor shutdown. The whole reactor core may be replaced at one time, about every 20 years. As such, the reactor system may have no requirement that the operator handle fuel. The overall unit may be sealed, physically and with electronic monitors, so that any intrusion attempt is easily detected. The elimination of any need or the ability to gain direct access to the fuel and use of smart monitoring systems not only reduces operator requirements, but also addresses proliferation concerns. Additionally, the SMR is small enough to be located below ground, which enhances containment and protection from terrorist activities. Finally, embodiments of the system are small enough that they can be shipped by barge, rail, and truck and installed at the site using modular construction techniques: this ability to remotely manufacture and obtain economies of serial production is a desirable benefit. When the fuel cartridges are returned to the manufacturer/designer/fabricator's facility, nearly all of the used nuclear material can be recycled and used as fuel in future cartridges, greatly reducing the volume and radio-toxicity of the final waste to be stored in a geologic repository. Unlike used fuel from conventional light water reactors, material from SMR's need not be stored for tens of thousands of years. Non-recyclable materials from SMR's require only a few hundred years of storage before the waste decays to levels of radiation associated with the original uranium ore. The reactor concept and its supporting fuel cycle infrastructure may offer a configuration of nuclear energy tailored to meet the needs of emerging electricity markets in developing countries as well as imminent global need for carbon-free non-electric energy sources. This configuration of nuclear energy may rely on the huge energy density of nuclear fuel (>106 times that of fossil fuel) to enable a distributed fleet of small fast reactors of long (20 year) refueling interval, providing local energy services supported by a small number of centralized facilities handling fuel supply and waste management for the entire fleet. The reactors may be sized for local and/or small grids, and are standardized, modularized and pre-licensed for factory fabrication and rapid site assembly. Correspondingly, the centralized fuel cycle infrastructure may be sized for economy of scale to support a large fleet of reactors in the region and may be operated under international safeguards oversight. The configuration is tailored to meet the tenets of sustainable development. FIG. 3 illustrates an exemplary nuclear energy infrastructure in its mature stage. A regional center 701 may supply/ship reactor fuel and/or accept spent fuel returns from sub-regions, such as countries 703. Various regional centers 701 may trade in fissile and fertile material to level out regional surpluses and/or shortages. Reactor Overview Embodiments of the present invention may include an approximately 50 MWe (125 MWt) to approximately 100 MWe (260 MWt) sodium-cooled fast reactor operating on a long (approximately 15 to approximately 20 year) whole core refueling interval. An initial fuel load may be enriched uranium (≦20% enriched) in the form of metal alloy fuel slugs, sodium or helium bonded to ferritic-martinsitic cladding. The reactor may exhibit an internal breeding ratio near unity such that its reactivity burnup swing is small and its core is fissile self-sufficient. A burnup swing of less than approximately 1% Δk/k may facilitate passive safety and passive load follow. Embodiments of the present invention may attain 80 MWtd/kg or more fuel average burnup, and upon pyrometallurgical recycle at completion of its 20 year burn cycle, depleted uranium makeup feedstock may be all that is required for the reload core. Upon multiple recycles, the core composition may gradually shift to an equilibrium transuranic fuel composition, which is also fissile self sufficient, and thus requiring only U238 makeup upon recycle. A forced circulation heat source reactor may deliver heat at ˜500° C. through a sodium intermediate loop that drives a supercritical CO2 (S—CO2) Brayton Cycle power converter attaining ˜40% conversion efficiency and may be capable of incorporating bottoming cycles for desalination, district heat, etc. Other embodiments might drive a Rankine steam cycle. Embodiments of the present invention may employ passive decay heat removal; achieve passive safety response to Anticipated Transients Without Scram (ATWS); and employ passive load follow. The balance of plant may have no nuclear safety function. The plant may be sized to permit factory fabrication of rail and barge shippable modules for rapid assembly at the site. Embodiments of the present invention may have features targeted to meet infrastructure and institutional needs of rapidly growing cities in the developing world as well as non-electric industrial and/or municipal niche applications in all nations. Targeting Emerging Markets Nuclear energy is a well-established industrial business that, over the past 35 years, has attained 13,000 reactor years of operating experience and 16% market share of world electricity supply. Nuclear energy is being deployed primarily in the form of large size (greater than or approximately equal to 1200 MWe) plants in industrialized nations. There are currently 436 reactors deployed in 30 countries. Future growth in nuclear deployments is projected to be as much as 66% or even 100% additional capacity by 2030. The majority of the growth is projected to take place in developing countries where institutional and infrastructure conditions often differ from those that, in the past, favored large scale plants and a once through fuel cycle. Developing nations often have small, local grids of under a few tens of GW, which are unable to accommodate a 1.2 to 1.5 GWe sized plant. Embodiments of the present invention operating at 100 MWe, are not only compatible with smaller grid size but additionally, the smaller capital outlay required for its installation is compatible with a developing country's necessity for sharing limited financing across multiple development projects during the early decades of its rapid economic growth. A twenty year refueling interval with fuel supply, recycle, and waste management services outsourced to a regional center enable a nation to attain unprecedented energy security even absent a need to first emplace a complete indigenous fuel cycle/waste management infrastructure. Moreover, centralization of fuel cycle facilities for economy of scale in technical and institutional safeguards operations may facilitate an international nonproliferation regime even for widespread worldwide deployment of nuclear-based energy supply. The energy supply growth rate in industrialized countries is projected to be slower than in developing countries. Nonetheless, new nuclear plants are needed for replacements of coal and nuclear plants as they are decommissioned at end of life. The large capacity interconnected grids in industrialized nations are compatible with large power rating plants. Niche markets, however, are expected to rapidly emerge in both developed and developing nations for non-electric and/or cogeneration applications of carbon-emission-free nuclear energy. Among these markets may be water desalination, oil sands/oil shale recovery and upgrading, and coal or bio to liquids synthetic fuel production. Passive safety posture precludes any safety function being assigned to the balance of plant and along with the reactor's reduced source term favor siting adjacent to industrial and municipal installations. Features of the Fuel Cycle First, the core power density (kwt/liter) and fuel specific power kwt/kg fuel may be derated so as to achieve a 20 year refueling interval while remaining within the bounds of the established metallic alloy fuels experimental database. This may provide a client long term energy security and a high level of reliable availability. Second, the once in 20 year whole-core refueling may be conducted by factory personnel who bring the refueling equipment and fresh fuel from offsite, conduct the refueling operations, and then return the used core and the refueling equipment to the factory. This may provide the client a way to attain energy security absent a prior need to emplace indigenous facilities for enrichment, fuel fabrication, reprocessing, and waste repositories. Third, the refueling operations may be done on the basis of a fuel handing assembly that may include multiple sub-components. Various numbers of sub-components may be included and may or may not be clustered. As an example, see an exemplary core made of seven fuel assembly clusters 801 in FIG. 4. FIG. 4 shows an exemplary arrangement of core components. For example, an outer layer of shield assemblies 803 may cover a layer of reflector 805, which may cover a layer of outer core 807. Middle core 809 of a lower enrichment may generally surround inner core 811 of still lower enrichment with primary control 813 and secondary control 815 assemblies placed within the core 801. As shown, the fuel, reflector, shield and control rod assemblies are grouped into seven-assembly clusters to speed the rate of core refueling. During operations, the seven-assembly cluster may be transferred after a very short cooling period following reactor shutdown so as to minimally interrupt energy supply availability. The short cooling period and seven-assembly cluster features may be possible due to the derated fuel specific heat (kwt/kg fuel). Fourth, the first fuel loading may be enriched uranium (enrichment<20%) and the core may be fissile self-sufficient such that at the end of the 20 year operation interval, the core contains as much bred-in fissionable content as has been burned out. Upon pyrometallurgical recycle of the used core, only U238 feedstock and fresh cladding may be required for refabrication of a replacement core. Fifth, over multiple recycles, the composition of the core may gradually transition from a U235-rich composition towards an equilibrium transuranic-rich composition that is also fissile self sufficient. The fuel cycle waste stream may exclusively include fission products, which require only 200 to 300 years of sequestration before decaying to background levels of radioactivity, whereas all transuranics may be returned to the reactor as fuel where they are converted to fission products. Sixth, after the first core loading, all subsequent cores may require only U238 as feedstock. This may extend the world's ore resource potential to nearly 100% productive use, and yielding at least a millennium of energy supply. Capability to use thorium-based metallic alloy fuel extends the world's resource base to multi millennia. Seventh, the fuel fabrication technology may offer the option of incorporating LWR used fuel crushed oxide particles onto a metallic alloy to form a cermet. This option, when combined with an added (oxide reduction) step in the pyrometallurgical recycle process may offer a route to cost effective management of LWR used fuel by subsuming it into the fast reactor closed fuel cycle. Features of a Heat Source Reactor First, a core layout may include assembly clusters of individually ducted and orifaced fuel assemblies. As described above, see FIG. 4 for exemplary seven-assembly clusters in a core layout. In other embodiments, other numbers and arrangements may be contemplated. The assemblies may be grouped into clusters for fuel handling while preserving individual fuel assemblies so as to retain the orificing and the limited free bow reactivity feedback characteristics. Replaceable reflector and shield assemblies may be grouped into 3 or 4 assembly clusters. Second, a “limited free bow” core clamping approach may be used. The clamping approach may utilize a removable and vertically adjustable horizontal wedge 901 located in a central assembly position of a core layout of ducted fuel assemblies 913 at an elevation approximately at above-core load pads 903, as shown in FIG. 5A. Note that radial displacement as shown in FIGS. 5A and 5B is exaggerated. The wedge 901 may be attached to a driveline 905 coupled to a vertical positioning mechanism 907 on a reactor deck 909. Preferably, the wedge 901 is at a lower end of the driveline 905, where the driveline 905 is in a vertical orientation. The wedge 901 can be removed/withdrawn to loosen the core for fuel handling, as shown in FIG. 5B. The wedge 901 can be re-inserted to clamp the core 915 and top load pads 917 outward against a core former ring 911 at a top load pad elevation once refueling is completed. Preferably the top load pads 917 may surround one or more ducted fuel assemblies 913 at approximately a top end of the ducted fuel assemblies 913. The above-core load pads 903 may surround one or more ducted fuel assemblies 913 above a fuel elevation, but below the top load pads 917. A grid plate elevation may approximately correspond with a bottom end of the ducted fuel assemblies 913. Third, a core may retain performance parameters, both operational and safety, even as the fuel composition evolves over the 20 year burn cycle and further evolves from one recycle loading to another. Fourth, embodiments of the present invention may include a strategy to monitor reactivity feedbacks throughout core life and to fine-tune their values using the vertical position adjustment of the wedge, should they drift as the core ages over its 20 year burn cycle. The integral reactivity feedbacks may be measured in situ by non-intrusive small adjustments of coolant flow rate, inlet coolant temperature, and control rod position. The rest position of the core clamping wedge 901 may be used to adjust the value of a core radial expansion component of the inherently negative power coefficient of reactivity, as shown in FIGS. 6A-6C. Note that radial displacement as shown in FIGS. 6A-6C is exaggerated. As shown in FIG. 6A, increasing power may increase outward (towards the right in FIGS. 6A-6C) bowing 951 of fuel assemblies 913. Unrestrained flowering upon an increase in core power may result from an increase a radial thermal gradient on the ducted fuel assemblies 913. Inboard ducted fuel assemblies 913 may push outward, as shown in FIG. 6B. Limited free bow core restraint may enhance radial dilation at fuel zone elevation of ducted fuel assemblies 913. As shown in FIG. 6C, an increase in coolant outlet temperature may bathe the wedge driveline 905 with increased temperature such that the driveline's thermal expansion may drive the wedge 901 downward/deeper. This may in turn amplify the radially outward bowing of core fuel assemblies 913 at a fuel zone elevation, which then may increase axial leakage and reduce reactivity. By adjusting a rest position of the wedge 901 at full power and full flow, the amplitude of the bowing enhancement can be fine tuned. Fifth, a passive safety response may be provided for loss of flow, loss of heat sink, chilled coolant inlet temperature and single rod runout transient overpower (ATWS) transient initiators without scram. The innate reactivity feedbacks with respect to power and fuel and coolant temperatures, when combined with a nearly zero reactivity burnup swing and with natural circulation capability at decay heat levels, may take the reactor to an undamaged safe state for all ATWS initiators, i.e., no damage may be incurred and a stable state may be reached for these initiators even if the rods fail to scram. Sixth, a passive decay heat removal channel may be provided to the ambient atmosphere ultimate heat sink always operating as a backup to active decay heat removal channels. The passive channel may always be operating at less than or approximately equal to 1% full power and can be confirmed to be functioning at all stages of core life by in situ non-intrusive measurements. The heat capacity of the core and internal structure is sufficient to safely absorb the initial transient of decay heat in excess of the passive channels' capacity. Features of a Power Plant First, a heat source reactor driving a S—CO2 Brayton cycle energy converter may attain nearly 40% or more heat to electricity conversion efficiency while operating in the working fluid range of ˜500° C., 21 MPa to 31° C., ˜7 MPa. This converter may use rotating machinery of extraordinarily high power density and recuperative heat exchangers of exceptionally high power density. Second, a heat source reactor may passively load follow the energy converter demand for heat. The reactor may sense the balance of plant demand communicated via flow rate and return temperature of the intermediate heat transport loop. The reactor's innate reactivity feedbacks may maintain heat production in balance with heat removal through the intermediate loop within tens of seconds and without need for active adjustments of control rods. Third, a Balance of Plant (BOP) may be provided that carries no nuclear safety function and can be built, operated and maintained to normal industrial standards. The reactor can passively accommodate all physically attainable combinations of flow rate and return temperature returning from the BOP through the intermediate heat transport loop. The passive decay heat removal channel may have no dependence on the BOP, and the nearly zero burnup control swing makes a rod runout TOP resulting from a control system error a no damage event. So the BOP need not carry any nuclear safety function. Fourth, embodiments of the present invention may include a potential to tie a broad diversity of BOP configurations to a standard, pre-licensed heat source reactor since the BOP carries no nuclear safety function. The S—CO2 Brayton cycle may reject ˜60% of supplied heat and may do so between ˜100° C. and 31° C. Many cogeneration options may exist for such a temperature range, including multi-effect distillation desalinization; district heat; district chilled water; ice production and others. Alternately, diverse non-electric industrial processes may be co-sited closely with the heat source reactor, given its self-protection features, small source term, passive load following feature, and high level of availability. Although the foregoing description is directed to the preferred embodiments of the invention, it is noted that other variations and modifications will be apparent to those skilled in the art, and may be made without departing from the spirit or scope of the invention. Moreover, features described in connection with one embodiment of the invention may be used in conjunction with other embodiments, even if not explicitly stated above.
047972475	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The apparatus shown in the drawings is a reactor 11 including a pressure vessel having a body 13 and a head 15. The body 13 is generally circularly cylindrical terminating at the bottom in a spherical bowl. The head 15 is dome-shaped and is provided at its base with an annular flange 17. The head 15 is secured to the body 13 by studs 19 which compress sealing O-rings (not shown) between the closure-head flange 17 and the top of the body 13. The control rods, instrument parts and the like (not shown in detail) extend from the top of the head 15. The body 13 and head 15 are enclosed in thermal insulating shields 21, 23 and 24. The shields 21, 23 and 24 could be composed of a material such as Ricardo Neutron Shielding sold by Richardson, Battery Parts Division, Melrose Park, Ill. 60160. In this case, the shields would reflect neutrons as well as serving as a thermal shield. The shielding can also be purely neutron reflecting. The shield 24 covers the top 26 of the head and need not be removed to afford access to remove the head. The shield 23 embodies a unique feature of this invention. The thermal shield 23 includes a frame 25 of thermal insulating material. Near or at the top of the frame 25, a plurality of panels 27 of thermal insulating material are mounted pivotal each on a hinge 29. The frame 25 is of polygonal transverse cross-section with a panel 27a pivotally mounted on hinges 29 near or at the top of each side 25a of the frame. Typically, the frame 25 may be of octagonal transverse cross-section. Each unit including a side 25a and a panel 27a may be separate from the other units 25a-27a or several or all sides 25a of the frame 25 may be joined. The frame 25 and the panels 27 may be composed of insulating board as shown. In this case, the frame 25 must have sufficient strength and rigidity to support the panels 27 notwithstanding the high temperature of the head 15 during normal operation. The frame 25 and/or the panels 27 may also be formed of opposite metallic plates defining between them a pocket within which thermal insulating material is disposed. The frame 25 and/or the panels 27 may be laminated structures formed of a plurality of overlaid insulating board. Instead of being of polygonal section, the frame 25 may be of circular transverse section with the panels 27 suitably shaped, i.e., curved, at their outer ends so that they are adapted to be pivotally mounted on the frame. At its lower end, the frame 25 is supported on a flange 31 extending from the body near its top. The panels 27 are pivotal from a generally horizontal position (FIGS. 1, 3, 4, 6) to a retracted position (FIGS. 3, 5, 7) in which they are generally vertical or at a relatively small angle to the vertical. Each panel 27a is of generally trapezoidal shape so that when the panels are in a horizontal position, they mate to form a substantially closed thermal shield over the portion of the head over which they extend (FIGS. 2, 6). In the horizontal position, the inner ends of the panels 27 are supported on a bracket 33 mounted on the head. For raising or lowering the thermal shield 23, a plurality of mechanisms or linkages 41 (FIGS. 3, 4, 5) are provided. The mechanisms 41 are disposed so that a single mechanism operates on adjacent sides 27a of the shield 23. Each mechanism 41 (FIGS. 3, 4) includes a rod 43 having a hook 45 at its upper end. A finger 47 projects from the rod 43 at its lower end. Below the hook 45, a circular plate 49 is mounted on the rod 43 and below the plate 49 a sleeve 51 is secured to the bar. The sleeve 51 and bar 43 are penetrated by cross rod 53. The rod 43 is passed between the adjacent panels 27a with the plate 49 above the panels and the sleeve 51 and cross rod 53 below the panels. The rod 43 is moveable in U-shaped guides 55 and 57 suspended from a bar 59 connected to the adjacent sides 25a at their joint. Guide 55 is shorter than guide 57 to accommodate the bar 43 which, when panels 27a are in a horizontal position, is at an angle to the vertical. In moving the panels 27a from the horizontal to the retracted position, the rod 43 slides along the walls of the guides 55 and 57 which function as cam surfaces. The cross bar 53 is positioned to engage a pair of parallel U-shaped members 61 suspended side-by-side each from one of the adjacent panels 27a (FIG. 4). The members 61 function as a cam with the cross rod 53 as a cam follower. The apparatus 63 (FIGS. 6 and 7) for raising or lowering the thermal shield 23 includes a plurality of cable 65 which extend from a rig (not shown) on the seismic support platform 67. The rig is driven by a motor (not shown) in turn pulling the cables 65 upwardly or downwardly. The motor is energized when receiver 80 on the platform 67 is enabled. The platform 67 is supported on columns 69, each provided with a sleeve 71 slideable along the column. Each sleeve 71 has a pulley 73 through which a corresponding cable 65 passes. A cantilever rod 75 extends from each sleeve 71. Each cantilever rod 75 has a hole 77 (FIG. 7) near its outer end through which the cable 65 passes. Each cable 65 has a loop 79 (FIG. 4) at its lower end which engages the hook 45 extending from a bar 43. In normal operation, the thermal shield 23 is in the shielding position (FIG. 6). The sleeve 71 is in the lowermost position with the cantilever rod 75 above the hook 45. When the head 15 is to be removed the motor (not shown) is energized by a remotely-actuable switch (not shown). The actuation is carried out by personnel 82, remote from the reactor who enables a transmitter 84 which communicates with receiver 80 on the platform 67. The cable 65, sleeve 71, cantilever bar 75, linkage bar 43, and cross bar 53 (FIG. 4) of each pair of sides 25a and panels 27a which are to be raised are pulled upwardly. Each cross bar 53 engages the lower surfaces of its associated panels 27a and pivots these panels to the retracted position as shown in dash-dot lines in FIG. 4. As each bar 43 is raised, its finger 47 engages the bottom of guide 55. As each bar continues to rise, it carries the associated frame sides 25a with it. Ultimately, the portion of the shield 23 which was raised is in the position shown in FIG. 7. The thermal shield 23 may be raised in separate sections and the tension studs 19 removed as indicated in FIG. 7. Alternatively, the thermal shield as a whole may be raised and then the studs may be removed. In either event, the head 15 is ultimately removed and the reactor 11 processed as required. When the shield 23 is to be replaced after the head 15 has been secured to the body 13, the motor (not shown) is energized in the reverse direction. Each linkage 41 is then lowered. Initially, each cross bar 53 is in the position shown in dash-dot lines in FIG. 4 As cables 65 are lowered, the shields 23 are lowered until the lower edges of sides 25 engages flange 31. Once this downward movement is completed, each cross bar 53 engages the webs 83 of the members 61 pivoting the panels 27a to the shielding position. While preferred embodiments of this invention have been disclosed herein, many modifications thereof are feasible. This invention is not to be restricted except insofar as is necessitated by the spirit of the prior art.
	claims	1. A plasma welding apparatus for a guide thimble and guide thimble end plug of a nuclear fuel assembly, comprising:a welding chamber including:an end-plug inserting part into which the end plug is inserted and fixed,a guide-thimble inserting part which is provided on the same axis as the end-plug inserting part and into which the guide thimble is inserted and fixed,a torch assembling part to which a plasma welding torch is assembled so as to make a right angle with the end-plug inserting part and the guide-thimble inserting part, andargon inflow and outflow ports through which argon is supplied or discharged;an end-plug transfer unit supplying the end plug to the end-plug inserting part; anda guide-thimble transfer unit transferring the guide thimble to the guide-thimble inserting part,wherein the end-plug transfer unit and the guide-thimble transfer unit are respectively disposed at opposite sides of the welding chamber, andwherein the plasma welding torch welds the end plug to the guide thimble such that only one end of the guide thimble is closed, while another end of the guide thimble is maintained open. 2. The plasma welding apparatus set forth in claim 1, wherein the end-plug inserting part and the guide-thimble inserting part are provided so as to be able to rotate relative to the welding chamber, and are rotated by a driving motor. 3. The plasma welding apparatus set forth in claim 2, wherein the end-plug inserting part and the guide-thimble inserting part are assembled to the welding chamber via at least one bearing. 4. The plasma welding apparatus set forth in claim 1, wherein the argon inflow and outflow ports are configured so that argon is introduced at a lower portion of the welding chamber and is discharged to an upper portion of the welding chamber. 5. The plasma welding apparatus set forth in claim 1, wherein the guide-thimble inserting part includes a collet into which the guide thimble is inserted, a jig that horizontally moves along the collet in a forward/backward direction and provides the collet with a fastening force for fixing the guide thimble, and a lever operating the jig. 6. The plasma welding apparatus set for the in claim 1, further comprising:a sprocket installed on outer circumference of the end-plug inserting part; anda chain connected to the sprocket; anda driving motor which rotates the end-plug inserting part, the driving motor generating a driving force transferred via the chain and the sprocket to the end-plug inserting part. 7. The plasma welding apparatus set for the in claim 1, wherein the guide-thimble transfer unit includes:a guide-thimble transfer driver that loads and unloads the guide thimble toward and out of the welding chamber;a loading rail via which the guide thimble is loaded toward the welding chamber; andan unloading rail via which the guide thimble is unloaded out of the welding chamber, the unloading rail being separated from the loading rail. 8. The plasma welding apparatus set for the in claim 7, wherein the loading rail and the unloading rail are provided to have a slope in a direction perpendicular to a transferring direction of the guide thimble. 9. The plasma welding apparatus set for the in claim 1, wherein the plasma welding torch is movable upward from the argon inflow port.
	abstract	A circulation pipe of a chemical decontamination apparatus including a malonic acid injection apparatus and an oxalic acid injection apparatus is connected to a purification system pipe, which is made of carbon steel, of a boiling water nuclear power plant. A malonic acid aqueous solution is injected from the malonic acid injection apparatus into the circulation pipe. An oxalic acid aqueous solution is injected from the oxalic acid injection apparatus into the circulation pipe. A reduction decontaminating solution including a malonic acid of 5200 ppm and an oxalic acid within a range of 50 to 400 ppm is supplied into the purification system pipe through the circulation pipe. Reduction decontamination for an inner surface of the purification system pipe is executed. After the reduction decontamination for the purification system pipe finishes, the malonic acid and oxalic acid included in the solution are decomposed and furthermore, the solution is purified.
	claims	1. A system for storing radioactive materials, comprising: a canister (4) containing radioactive waste; a container (C) provided with a casing (1), a base (2), and a cover (3), said container (C) delimiting a cavity for housing the canister (4); and a convection-based ventilation system provided with lower air inlets (5), an area (6) of upward air circulation defined between the canister (4) and the inner surface of the storage container, and upper air outlets (7); characterized in that the ventilation system consists of a passive helicoidal convection-based ventilation system, in which the inlets (5) and outlets (7) of the ventilation system have an outer mouth (51, 72) and an inner mouth (52, 71); in that at least the inlets have a decreasing variation in section in the direction of air circulation inside them, producing an increase in velocity, and a decrease in the pressure and temperature of air circulating therethrough; and in that said inlets (5) and outlets (7) are curved and facing an oblique direction with respect to the radial direction of the container, so that the air describes between the inlets (5) and the outlets (7) of an upward helicoidal path around the canister (4). 2. System according to claim 1, characterized in that the outer mouth (51, 72) and the inner mouth (52, 71) of the same air inlet (5) or outlet (7) are laterally out of alignment. 3. System according to claim 1, characterized in that the outer mouth (51, 72) and the inner mouth (52, 71) of the same air inlet (5) or outlet (7) are arranged geometrically at different elevation or height. 4. System according to claim 1, characterized in that the casing (1) of the container (C) comprises at least:an outer shell (11) of steel;a high-density concrete layer (14), specialised in shielding gamma radiation and partially reducing the energy of a neutron radiation, and which includes in its dosing high-density aggregates and;a concrete layer (15), which reduces the energy of a neutron radiation (thermalisation) and absorbs much of it; and which includes in its dosage aggregates with high hydrogen content and aggregates with a neutron absorber. 5. System according to claim 4, characterized in that the casing (1) of the container (C) comprises several shells: an outer one (11), an inner one (12) and an intermediate one (13), of steel, which are concentric and form a permanent formwork for the concrete layers (15). 6. System according to claim 4, characterized in that the base (2) has a peripheral portion with a structure analogous to the wall of the casing (1) and has a closing plate (21) forming a suitable surface for the support of the canister (4). 7. System according to claim 4, characterized in that the cover (3) is made of a metal structure containing the air outlets (7) and forms a permanent formwork for the phased differentiated concreting thereof, with concretes specialised against radiations.
050341847	claims	1. In a control assembly for a nuclear reactor having a piston within a cylinder controlling the longitudinal movement of a control element during insertion into a reactor core, the improvement comprising: an accelerating fluid of predetermined volume and pressure propelling said piston, causing rapid insertion of said control element, a decelerating fluid of decreasing volume and pressure decelerating the movement of said piston, and absorbing means arresting the longitudinal movement of said control element at the point of full insertion of said control element into said reactor core, wherein said decelerating fluid is exhausted through an orifice which is adjustable to control the rate of exhaust, and wherein said orifice is adjusted by means of an end cap of varying diameter which is operatively connected with said absorbing means and is movable longitudinally in said cylinder. 2. The control assembly improvement defined in claim 1, wherein the longitudinal movement of said absorbing means and end cap causes said orifice to open or close, increasing or decreasing, respectively, the rate of exhaust of said decelerating fluid. 3. The control assembly improvement defined in claim 2, wherein said decelerating fluid is contained within said cylinder tube surrounding said piston and within a compression tube in fluid communication with the interior of said cylinder tube. 4. The control assembly improvement defined in claim 3, where said accelerating fluid is air. 5. The control assembly improvement defined in claim 3, wherein said decelerating fluid is air. 6. The control assembly improvement defined in claim 3, wherein said accelerating fluid is a compressible gas. 7. The control assembly improvement defined in claim 3, wherein said decelerating fluid is a compressible gas.
	abstract	In a radiation image storage panel having an energy storable phosphor layer formed by a gas phase-accumulation method, the energy storable phosphor layer gives off an emission having a luminescence width in terms of d in the range of 150 to 395 μm when it is exposed to radiation and then excited with a stimulating light of 50 μm half-width.
	abstract	An X-ray filter comprises an array of filter elements (5) an control circuit, the control circuit comprising an array of switching devices (33) provided on a common substrate (52), a switching device (33) being provided for each filter element for switching a control signal to the respective filter element. An output terminal of each switching device is provided with an external connection portion (54) located at the respective switching device. An array of external connection portions (54) is thus provided over the array of switching devices (33). The connection portions are then bonded to a connection block of the array of filter elements. This avoids the need to use edge connections of the substrate (52), and provides a secure mechanical and electrical connection between the control circuit and the filter array.
	summary
	summary
	summary
	abstract	A method and apparatus for controlling beam emittance by placing a lens array in a drift space of an illumination system component. The illumination system component may be an electron gun or a liner tube or drift tube, attachable to an electron gun. The lens array may be one or more mesh grids or a combination of grids and continuous foils. The lens array forms a multitude of microlenses resembling an optical “fly's eye” lens. The lens array splits an incoming solid electron beam into a multitude of subbeams, such that the outgoing beam emittance is different from the incoming beam emittance, while beam total current remains unchanged. The method and apparatus permit independent control of beam current and beam emittance, which is beneficial in a SCALPEL illumination system.
061335770	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The detailed description set forth below in connection with the appended drawings is intended as description of the presently preferred embodiment of the invention and is not intended to represent the only form in which the present invention may be constructed or utilized. The description sets forth the functions and the sequence of steps for constructing and operating the invention in connection with the illustrated embodiment. It is to be understood, however, that the same or equivalent functions and sequences may be accomplished by different embodiments that are also intended to be encompassed within the spirit and scope of the invention. The extreme ultra-violet photolithography system for facilitating production of semi-conductor components having geometries of 10 nm and smaller of the present invention is illustrated in FIGS. 1-11, which depict a presently preferred embodiment thereof. Referring now to FIG. 1, the extreme ultra-violet photolithography system generally comprises a converging-diverging nozzle 10 from which gas 11 flows, at a supersonic velocity, toward diffuser 12 which captures a substantial portion of the flowing gas 11. The converging-diverging nozzle 10 and the diffuser 12, as well as the collecting and focusing optics 29 and the work piece, i.e., integrated circuit chip(s) being fabricated, are all preferably disposed within a common vacuum chamber 40, so as to facilitate integrated circuit fabrication utilizing photolithography. As described in detail below, the diffuser 12 reduces the velocity of the flowing gas 11, while simultaneously increasing the pressure thereof. Gas flows from the diffuser 12 via conduit 13 to compressor 14, which preferably comprises a 0.71 hp compressor. The compressor 14 compresses, i.e., increases the pressure of, the gas 11 such that it may be recycled to the converging-diverging nozzle 10 and thus used repeatedly to produce extreme ultra-violet light. Gas flows from the compressor 14 to heat exchanger 16, preferably a 64.1 btu/min heat exchanger for removing heat from the compressed gas. According to the preferred embodiment of the present invention, the temperature of the gas entering the heat exchanger 16 is approximately 610.degree. K and the temperature of the gas exiting the heat exchanger 16 is approximately 300.degree. K. The gas exiting the heat exchanger 16 is communicated via conduit 17 to the converging-diverging nozzle 10 where a stagnation pressure of 6,079 torr is developed. Stagnation pressure is defined herein as that gas pressure when no flow occurs. Referring now to FIG. 2 also, the converging-diverging nozzle 10 more particularly comprises a pressure plenum 18 into which the compressed gas from the heat exchanger 16 flows. The converging-diverging nozzle 10 further comprises a converging portion 20 and a diverging portion 22. The converging-diverging nozzle 10 is configured so as to accelerate the gas flowing therethrough to a supersonic velocity, preferably above Mach 2, preferably approximately Mach 3. The diverging portion 22 preferably has a generally rectangular cross-section and is preferably configured such that the length, Dimension L, is substantially greater than the width, Dimension W, thereof. This configuration provides a high aspect ratio which facilitates the exposure of a substantial portion of the flowing gas to the radiated energy beam and which provides a short path for extreme ultra-violet light stimulated thereby through the flowing gas. Referring now to FIGS. 1 and 3, the diffuser 12 generally comprises an opening which corresponds generally in size and configuration to that of the widest portion of the diverging portion of the converging-diverging nozzle 10. Thus the opening of the diffuser has a length which is preferably slightly longer than the length of the converging-diverging nozzle 10 and has a width which is preferably slightly longer than the width of the converging-diverging nozzle, so as to capture a substantial portion of the gas flowing from the converging-diverging nozzle 10. Those skilled in the art will appreciate that various different configurations of the diffuser 12 are suitable. The diffuser decreases in cross-sectional area from the opening 30 thereof to the coupling end 32 thereof, at which the fluid conduit 13 attaches. As discussed in detail below, the cross-sectional area of the diffuser 12 optionally increases again, from the narrowest portion thereof, so as to define a throat. Such tapering or narrowing of the cross-sectional area of the diffuser 12 provides a gradual slowing of the gasses captured thereby, while minimizing the occurrence of undesirable regurgitation which might otherwise occur. Optionally, one or more knife edges are formed in or proximate the diffuser 12, so as to aid in the deceleration of the gasses entering the opening 30. According to the preferred embodiment of the present invention, the periphery of the opening 30 of the diffuser 12 is formed as a first knife edge 31. Additional concentric generally rectangular knife edges 33 and 35 are disposed within the opening 30 of the diffuser 12 and mounted thereto via any suitable means. Knife edge struts may optionally be utilized to mount the second 33 and third 35 concentric rectangular knife edges in place within the opening 30 of the diffuser 12. Those skilled in the art will appreciate that various different numbers and configurations of such knife edges may be utilized to effect generation of shocks which tend to decrease the velocity of the supersonic gas while simultaneously increasing the pressure thereof within the diffuser 12. Isobaric pressure profiles of the gas flowing from the converging-diverging nozzle 10 are provided in FIG. 1. As shown, the radiated energy beam, an electron beam according to the preferred embodiment of the present invention, is directed into that portion of the flowing gas 11 proximate the converging-diverging nozzle 10, so as to enhance the efficiency of the present invention. This is better shown in FIG. 4 which illustrates the relative positions of the electron beam 23 and the flowing gas 11 in perspective. A portion of the extreme ultra-violet light 27 whose emission is stimulated from the flowing gas 11 by the radiated energy beam 23 is collected and focused by collecting and focusing optics 29, which direct the extreme ultra-violet light onto a work piece, i.e., an integrated circuit component being fabricated, as desired. According to the preferred embodiment of the present invention, a vacuum pump, preferably that vacuum pump 36 utilized to evacuate the vacuum chamber 40 within which the gas 11 flows and within which the photolithographic process is performed, evacuates a substantial portion of the gas 11 which is not captured by the diffuser 12 and provides that gas 11 back to the converging-diverging nozzle 10, preferably via the compressor 14 and heat exchanger 16, so as to facilitate recycling thereof. Referring now to FIG. 4, in operation a gas, preferably a noble gas such as argon, helium, or xenon, or a combination thereof, flows at a supersonic velocity from the converging-diverging nozzle 18 when a pressurized supply thereof is provided to the converging-diverging nozzle 18 via gas conduit 17. Sufficient pressure is provided by compressor 14 to achieve the desired gas flow speed. A radiated energy beam, preferably an electron beam, is directed through the supersonic gas flow 11 at a position which minimizes the transmission of the resulting extreme ultra-violet light through the gas 11, thereby mitigating undesirable absorption thereof. A substantial portion of the flowing gas 11 is captured by the diffuser 12 and recycled. A substantial portion of the gas not captured by the diffuser 12 is evacuated from the vacuum chamber 40 via vacuum pump 36 and recycled. At least a portion of the extreme ultra-violet light 27 emitted due to the interaction of the radiated energy beam 23 with the supersonic gas 11 is collected and focused by collecting and focusing optics 29 so as to facilitate photolithography therewith. Thus, according to the present invention, contamination of the collecting and focusing optics 29, as well as any other sensitive surfaces within the vacuum chamber 40, is mitigated. Such contamination is mitigated since supersonic flow of the gas 11 tends to force all of the gas particles, i.e., molecules, atoms, ions, electrons, etc., into the diffuser 12, thereby substantially mitigating the amount of such particles floating freely within the vacuum chamber 40 and capable of coming into contact with such sensitive items. The present invention takes advantage of the gas dynamic properties of the supersonic jet to direct any debris generated during the plasma formation into the diffuser, and thus away from the collection and focusing optics 29, as well as the rest of the photolithography system. The efficiency of the present invention is enhanced by minimizing the amount of gas 11 through which the generated extreme ultra-violet light 27 must pass. As those skilled in the art will appreciate, extreme ultra-violet light is readily absorbed (and thus attenuated) by the noble gasses from which its emission is stimulated. Thus, it is very desirable to minimize the distance through which the extreme ultra-violet light 27 must travel through such gas. This is accomplished by positioning the radiated energy beam 23 close to the surface of the flowing gas 11, preferably by positioning the radiated energy beam 23 proximate the converging-diverging nozzle 10 where the gas flow has a comparatively narrow cross-sectional area and comparatively high density. Thus, according to the present invention, the high density gas region is confined to nearly the same volume as that occupied by the radiated energy beam. Thus, extreme ultra-violet light generated thereby is not required to travel through a substantial portion of the high density gas after leaving the area where stimulated emission occurs. The high aspect ratio configuration of the converging-diverging nozzle tends to maximize the volume of flowing gas available for interaction with the radiated energy beam, while simultaneously minimizing the volume of flowing gas which attenuates the stimulated extreme ultra-violet light. As those skilled in the art will appreciate, the higher the velocity of the flowing gas 11, the smaller the mass flow thereof which will diverge or turn away from the gas flow, i.e., jet, when surrounded by the very low pressure of the vacuum chamber. Any such flow which diverges from the gas jet into the high vacuum surrounding the gas jet must ultimately be pumped out against a very high adverse pressure ratio, which adds substantially to the cost of manufacturing and maintaining the system. Even more important, the gas that diverges from the gas jet becomes a potential contaminant for the collecting and focusing optics and also becomes an undesirable attenuating mass for the extreme ultra-violet light which is produced by the interaction of the radiated energy beam and the gas flow. Further, by converting a significant portion of the kinetic energy of the flowing gas 11 into pressure, the need to increase the pressure of the gas via the compressor 14 is reduced, thereby facilitating operation with a smaller capacity and less expensive compressor 14. Referring now to FIGS. 5 and 6, the generally rectangular concentric knife edges 33, 35 of FIG. 3 are shown in further detail. Each generally concentric knife edge 33, 35 preferably comprises a body 37 and a bevel 39. As those skilled in the art will appreciate, it is the purpose of each knife edge 31, 33, and 35 to produce a shock wave, similar in nature to the sonic boom shock wave associated with supersonic aircraft, which defines a region of increased pressure within the diffuser 12, and thus facilitates reduction of the speed of the flowing gas 11 and simultaneously facilitates an increase in the pressure thereof. Referring now to FIG. 8, the converging-diverging nozzle is optionally configured as a cap 10a which is specifically sized and shaped to fit a standard pulse generator. Thus, the cap 10a comprises a body 50 which is sized to be received within the exit orifice of a pulse generator and a flange 52 which functions as a stop to limit insertion of the body 50 into the exit orifice. A rectangular boss 54 has a rectangular opening 56 formed therein. The converging-diverging bore 58 of the nozzle is formed in a continuous or co-extensive manner in the body 50, flange 52, and boss 54. Such construction facilitates easy removal and replacement of the converging-diverging nozzle 10a, particularly when a standard pulse generator is utilized. Referring now to FIG. 9, a preferred cross-sectional profile of a nozzle orifice is shown. The nozzle comprises a converging region 60 which decreases to form a neck 62 and then increases in cross-sectional area to form the diverging region 64 thereof. The exit plane 66 is that plane of the nozzle flush with the end thereof, i.e., the outer opening thereof. Referring now to FIG. 10, the cross-sectional profile of the diffuser is shown. According to the present invention, the diffuser tapers or converges from the entry plane 70 to define a converging portion 72 thereof. At the end of the converging portion 72 a neck 74 is formed and the diffuser may then optionally diverge or increase in cross-sectional area so as to form a diverging portion 76. As those skilled in the art will appreciate, the velocity of the flowing gas 11 decreases within the converging portion 72, while the pressure thereof simultaneously increases. Referring now to FIG. 11, the calculated density field for a xenon extreme ultra-violet light source jet and diffuser is shown. Gas 11a from within the converging-diverging nozzle exits therefrom at the exit plane 66 to form gas jet 11b. The gas jet 11b enters the diffuser at the entry plane 70 thereof. Within the diffuser 12 first oblique shocks 80 are formed due to the knife edge(s) 31 defined by the opening 30 of the diffuser 12. The oblique shocks 80 interact to form perpendicular shock 82 downstream therefrom. Second oblique shocks 84 are formed as the flowing gas interacts with the internal walls of the diffuser. The second oblique shocks 84 interact with one another so as to form perpendicular shock 86. Third oblique shocks 88 are formed in a similar manner downstream from the second oblique shocks 84. As those skilled in the art will appreciate, each shock defines a high pressure region within which the flowing gas slows. In this manner a plurality of knife edges may be utilized to form shocks so as to effect slowing of the gas flow and increasing the pressure thereof. It is understood that the exemplary method and apparatus for producing extreme ultra-violet light described herein and shown in the drawings represents only a presently preferred embodiment of the invention. Indeed, various modifications and additions may be made to such embodiment without departing from the spirit and scope of the invention. For example, various sizes, shapes, crosssectional configurations, etc. of the nozzle and diffuser are contemplated. It must further be appreciated that various different configurations of the radiated energy beam, other than circular as shown, may be utilized. For example, the radiated energy beam 23 may alternatively be elliptical, square, rectangular, triangular, etc. It is generally desirable that the radiated energy beam 23 be comparable in cross-sectional area to that portion of the flowing gas 11 proximate the converging-diverging nozzle 10, so as to minimize the amount of gas 11 through which stimulated extreme ultra-violet light 27 must flow. Further, it must be appreciated that the method and apparatus for producing extreme ultra-violet light according to the present invention may be utilized in a variety of different applications, and is not limited to use in photolithographic applications. Further, it must also be appreciated that the general method and apparatus of the present invention may alternatively be utilized to produce wavelengths of electromagnetic radiation other than extreme ultra-violet, and thus is not limited to the production of extreme ultra-violet light. Thus, these and other modifications and additions may be obvious to those skilled in the art and may be implemented to adapt the present invention for use in a variety of different applications.
054229200	description	BEST MODE FOR CARRYING OUT THE INVENTION An example of the present invention will be explained in detail on the basis of the drawings. 16 types of UO.sub.2 powders having different average particle sizes were weighed by about 10 mg respectively using a chemical balance. The weighed UO.sub.2 powder was individually placed into a quartz boat of a thermogravimetric analysis apparatus (TBA-50 model, made by Shimadzu Corporation), and after inputting the weighed value into this apparatus, heating was performed under the following heating condition, and the weight change ratio with respect to the heating temperature was determined. The heating and measurement were performed systematically under automatic control, and 16 types of oxidation curves were obtained. One example thereof is shown in FIG. 2. Air flow amount at the inlet of the apparatus: 10 mL/minute Moisture of dry air (dew point): not more than -70.degree. C. Temperature of initiation of temperature raising: room temperature Temperature raising speed: 2.degree. C./minute According to inflection points P initially appearing in the oxidation curves, temperatures of the arrival of the powders to the U.sub.3 O.sub.7 phase were determined respectively. In this example, the arrival temperatures of the 16 types of the UO.sub.2 powders to the U.sub.3 O.sub.7 phase were distributed in a wide range from about 170.degree. C. to about 250.degree. C. From the above-mentioned 16 types of the UO.sub.2 powders were produced 16 individuals of UO.sub.2 sintered pellets by means of the following method. At first, 0.2% of zinc stearate was added to and mixed with the UO.sub.2 powder as a lubricant, and this mixture was placed in a mold to form a green pellet having a diameter of 10 mm and a height of 15 mm under a pressure of about 2 t/cm.sup.2. Next, this formed article was sintered in a hydrogen gas flow at 1750.degree. C. for 5 hours to obtain a UO.sub.2 sintered pellet. Next, the obtained UO.sub.2 sintered pellet was placed in a cylindrical container, an acrylic ester resin was poured around it to embed the pellet in the resin, and thereafter bubbles in the resin were removed using a reduced-pressure pump. Subsequently the pellet embedded in the resin was cut. Its cut face was polished. and then the polished face is etched with a hydrofluoric acid solution to expose crystal grain boundaries. The exposed crystal grain boundaries were observed with an optical microscope, and a negative film of the crystal grain boundaries was manufactured. Using this negative film, according to the cross-sectional method in accordance with ASTM E-112, crystal grain sizes were measured from 16 individuals of the UO.sub.2 sintered pellets respectively. A relation between the arrival temperatures to the U.sub.3 O.sub.7 phase of the above-mentioned 16 types of the UO.sub.2 powders and the crystal grain sizes of the UO.sub.2 sintered pellets manufactured from each of the UO.sub.2 powders was plotted, and a curve R showing the correlation between the U.sub.3 O.sub.7 phase arrival temperature and the crystal grain size shown in FIG. 3 was obtained. Subsequently, about 10 mg of a UO.sub.2 powder of a test sample was weighed, and the thermogravimetric analysis was performed by means of the same operation as described above to obtain an oxidation curve. As shown in FIG. 2, the temperature of arrival of this powder to the U.sub.3 O.sub.7 phase was 210.degree. C. according to an inflection point P initially appeared on the oxidation curve. This temperature of 210.degree. C. was allowed to correspond to the axis of abscissa in FIG. 3, and according to an intersection (a) with the curve R, it was estimated that a UO.sub.2 sintered pellet to be manufactured from the UO.sub.2 powder of the test sample was estimated to have a crystal grain size of about 18 .mu.m. In order to confirm reliability of the present invention, when a UO.sub.2 sintered pellet was manufactured in the same manner as described above using the same UO.sub.2 powder as the above-mentioned test sample, and the crystal grain size was directly measured by means of the above-mentioned cross-sectional method, then the crystal grain size was about 19 .mu.m. Thereby it was found that the crystal grain size of the UO.sub.2 sintered pellet estimated by the present invention is approximately coincident with the crystal grain size measured directly, and the present invention is a reasonable estimating method. As described above, according to the present invention, when the crystal grain size of the UO.sub.2 sintered pellet manufactured from the UO.sub.2 powder is determined, the crystal grain size can be estimated from the oxidation behavior of the UO.sub.2 powder without actually manufacturing a sintered pellet. Thereby the measurement cost for the crystal grain size can be reduced. Especially, when a UO.sub.2 sintered pellet having a large crystal grain size is produced in order to realize a high degree of combustion of nuclear fuel, the judgment of suitability of a raw material powder for the sintered pellet can be performed rapidly and economically, and the ratio of occurrence of an unexpected sintered pellet having a small crystal grain size can be suppressed to be low. INDUSTRIAL APPLICABILITY The method of the present invention is useful for rapidly and economically performing the judgment of suitability of raw material powders for sintered pellets having large crystal grain sizes for realizing a high degree of combustion of nuclear fuel.
	abstract	Method and arrangement for implementing a system (10) for providing multi-angular SPECT radiation sampling utilizing slant-angle collimation. The system (10) includes a collimator (13) positioned between a radiating mass (19) within a patient (60) and a radiation detector (21). The collimator (13) is spaced apart from a translational path (25) of the radiating mass (19) at a predefined distance (24). A plurality of apertures (27) extend through the collimator (13) and each forms a passageway (28) for radiation rays (20) emanating from the radiating mass (19) in a direction substantially aligned with a longitudinal axis (29) of the respective passageway (29) and in this manner enables the aligned radiation rays (20) to strike the radiation detector (21). The plurality of passageways (28) include a first group (30) of passageways adjacently aligned in a first row (32) and arranged so that the longitudinal axes (29) of the first group (30) of passageways (32) are substantially contained in a first plane (34) oriented substantially perpendicularly to a central plane (16) of the collimator (13). Each of the parallel longitudinal axes (29) of the first row (32) of passageways are obliquely oriented with respect to the central plane (16) of the collimator (13) with an included angle (36) therebetween. Each of the included angles, when measured clockwise from the central plane (16) or face of the collimator (13) to a respective longitudinal axis, is an acute angle.
	summary
050733052	claims	1. A method of evacuating a container to a vacuum for use in treating radioactive wastes by placing the waste into the container, and evacuating, sealing off and thereafter compressing the container, the method comprising the steps of: placing the waste into the container, forming over the waste a filter layer formed substantially entirely of particulate material fulfilling one of the following requirements: (1) a layer having a thickness of at least 5 mm and formed of a particulate material not smaller than 40 .mu.m to less than 105 .mu.m in mean particle size; (2) a layer having a thickness of D in mm and formed of a particulate material not smaller than 105 .mu.m to not greater than 210 .mu.m in mean particle size d in .mu.m, the thickness D and the mean particle size d having the relationship represented by: EQU D.gtoreq.(20/105).times.d-15; thereafter aspirating a gas through the filter layer from thereabove to evacuate the container and sealing off and compressing the container. 2. A method as defined in claim 1 wherein a closure provided with an evacuating pipe is welded to the container after the filter layer has been formed, and the gas is aspirated through the evacuating pipe. 3. A method as defined in claim 2 wherein the container is sealed off by collapsing the evacuating pipe. 4. A method as defined in any one of claims 1 to 3 wherein the radioactive waste comprises a hull. 5. A method as defined in any one of claims 1 to 3 wherein the radioactive waste comprises a hull and a block of waste to be treated. 6. A method as defined in any one of claims 1 to 3 wherein the container is a stretchable or contractable container of the bellows type. 7. A method as defined in any one of claims 1 to 3 wherein the particulate material is a metal powder. 8. A method as defined in any one of claims 1 to 3 wherein the particulate material is a stainless steel powder. 9. A method as defined in any one of claims 1 to 3 wherein the particulate material is a ceramic powder.
052221130	abstract	The invention is directed to an X-ray microscope having a pulsed X-ray radiation source which supplies an intensive line radiation such as a plasma focus source. The microscope includes a reflecting condenser which focusses the radiation of the radiation source on the specimen to be investigated and an X-ray optic configured as a zone plate. With the zone plate, the specimen is imaged on an X-ray detector with a high resolution. The above combination of elements makes it possible to free an adequately high amount of X-ray energy at the location of the specimen while providing a high resolution free of image errors so that the required short exposure times are provided for the investigation of living cells.
052672760	abstract	Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.
	description	This application claims the benefit of U.S. Provisional Application No. 61/625,378, filed Apr. 17, 2012. U.S. Provisional Application No. 61/625,378, filed Apr. 17, 2012 is hereby incorporated by reference in its entirety into the specification of this application. The following relates to the nuclear reactor arts, nuclear power generation arts, nuclear reactor instrumentation arts, nuclear vessel feedthrough arts, and related arts. In nuclear reactor designs of the integral pressurized water reactor (integral PWR) type, a nuclear reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel. In a typical design, the primary coolant is maintained in a subcooled liquid phase in a cylindrical pressure vessel that is mounted generally upright (that is, with its cylinder axis oriented vertically). A hollow cylindrical central riser is disposed concentrically inside the pressure vessel. Primary coolant flows upward through the reactor core where it is heated and rises through the central riser, discharges from the top of the central riser and reverses direction to flow downward back toward the reactor core through a downcomer annulus defined between the pressure vessel and the central riser. In the integral PWR design, the at least one steam generator is located inside the pressure vessel, typically in the downcomer annulus. Some illustrative integral PWR designs are described in Thome et al., “Integral Helical Coil Pressurized Water Nuclear Reactor”, U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. Other light water nuclear reactor designs such as PWR designs with external steam generators, boiling water reactors (BWRs) or so forth, vary the arrangement of the steam generator and other components, but usually locate the radioactive core at or near the bottom of a cylindrical pressure vessel in order to increase the likelihood that the reactor core will remain submerged in coolant in a loss of coolant accident (LOCA). The nuclear reactor core is built up from multiple fuel assemblies. Each fuel assembly includes a number of fuel rods. Spaced vertically along the length of the fuel assembly are grid assemblies which provide structural support to the fuel rods. At the top and bottom of the fuel assembly are an upper end fitting and a lower end fitting, respectively, providing structural support. The fuel assembly also includes guide tubes interspersed among the fuel rods. The guide tubes are welded to the grid assemblies as well as the upper and lower end fittings to form the structural support for the fuel assembly. Multiple fuel assemblies are welded or otherwise attached to each other to form a core, which is contained in a core former. The entire core is supported in a core basket, which may be suspended from the reactor lower vessel flange. Control rods comprising neutron absorbing material are inserted into and lifted out of the guide tubes of the fuel assembly to control core reactivity. Instruments that monitor core conditions (e.g. reactor power, temperature, pressure, flow, neutron flux, etc.) and their accompanying cabling may also be inserted into some of the guide tubes. Generally, a guide tube contains either a control rod or an instrument but not both, due to space limitations. The instruments and associated cabling are called incores because they are located in the core. While it is preferred that the incores be located in a guide tube in the center of the fuel assembly, they may also be located at the edge of a fuel assembly due to the arrangement of other core components. Generally, not all fuel assemblies contain an incore, and the fuel assemblies that do contain an incore only contain one. Locating these instruments in the reactor core, or anywhere in the vessel, is a challenge because the reactor vessel contains high temperature and high pressure water and the core produces radiation, an inhospitable environment for electronics. One approach for instrumentation in nuclear reactors uses thimble tubes contained in conduits, as disclosed in, for example, U.S. Pat. No. 5,120,491 to Brown et al., filed Sep. 17, 1991 and U.S. Pat. No. 4,983,351 to Tower et al., filed May 1, 1989. In this approach, thimble tubes housed in conduits run from a seal table through the vessel and into the core. The thimble tubes enter the reactor vessel at either the vessel head or the bottom of the vessel to provide a straight line to the core. Multiple thimble tubes are terminated at the seal table located outside of the reactor, for example in a dedicated compartment separate from the reactor compartment. A problem with this approach is that a failure of a conduit or a seal table connection can cause an unisolable primary leak. Although the resulting leak would be small and within the capacity of the coolant charging pumps, it may still necessitate reactor shutdown. Another problem with the thimble and conduit design is that, if routed from the bottom of the vessel, a leak from around the vessel penetration would, in the absence of any action, cause the core to become uncovered. In the case of control rods, the usual approach is to employ control rod drive mechanisms (CRDMs) with motors that are mounted externally at the top of the reactor vessel. There is currently interest in relocating the CRDM motors inside the pressure vessel, so as to facilitate compact small modular reactor (SMR) designs and to eliminate the CRDM as a potential LOCA source. Electrically driven CRDMs in which the motor is located in the vessel are called internal CRDMs. A disadvantage of internal CRDMs is that they necessitate a large number of electrical penetrations through the pressure vessel for the power, control, and signal (e.g., position indicator sensor) lines required for operating the CRDMs. These lines would add further thimble tube/conduit lines, and would further burden the (typically already heavily loaded) seal table. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In one aspect of the disclosure, an electrical feedthrough including: a cylinder having an outside end and an inside end and a lumen passing between the outside end and the inside end; a flange connected to and sealed with the inside end of the cylinder but not sealing the lumen at the inside end of the cylinder; an electrical termination disposed in the cylinder and sealing the lumen of the cylinder; an outside electrical connector disposed at the outside end of the cylinder; and at least one electrical conductor passing through at least a portion of the lumen and having an end terminating at the electrical termination. The electrical feedthrough is configured to mount at a vessel penetration of a pressure vessel wall with the flange sealed against an inside surface of the pressure vessel wall, the cylinder passing through the vessel penetration of pressure vessel wall, and the outside electrical connector accessible from outside the pressure vessel. In some embodiments, the electrical feedthrough includes an end plate is disposed at the outside end of the cylinder and at least one pressure fitting/connector sealed with and passing through the end plate, the end plate and the at least one pressure fitting/connector sealing the lumen of the cylinder at the outside end of the cylinder. In these embodiments, the pressure fitting/connector includes the electrical connector disposed at the outside end of the cylinder. In some embodiments, the electrical termination of the electrical feedthrough comprises a hermetically sealed connector disposed at the inside end of the cylinder and sealing the lumen of the cylinder at the inside end of the cylinder, and the at least one electrical conductor passes through the cylinder to electrically connect the hermetically sealed connector at the inside end of the cylinder and the outside electrical connector at the outside end of the cylinder. In another aspect of the disclosure, a nuclear reactor comprises a feedthrough and a pressure vessel as set forth in the immediately preceding paragraph, and a nuclear reactor core comprising fissile material disposed inside the pressure vessel, the pressure vessel further containing primary coolant water at an operating pressure, wherein the apparatus comprises a nuclear reactor. In another aspect of the disclosure, an apparatus comprises: a nuclear reactor including a nuclear reactor core comprising fissile material and a pressure vessel containing the nuclear reactor immersed in primary coolant water at an operating pressure, the pressure vessel having a vessel penetration passing through a pressure vessel wall; and an electrical feedthrough sealing the vessel penetration and having an outside electrical connector mounted at the pressure vessel, the outside electrical connector being at atmospheric pressure. FIG. 1 illustrates an integral reactor pressure vessel 6 including an upper vessel 8 and a lower vessel 10. The vessel portions 8, 10 are joined at a mid-flange 14 during operation, but FIG. 1 shows the upper vessel 8 lifted off the lower vessel/midflange 10, 14 (e.g., for refueling). which are shown separated in FIG. 1. The lower vessel 10 includes illustrative feedthroughs 4. In an integral reactor, the vessel houses the reactor core, reactor internals, and steam generator in a common pressure boundary, i.e. in the pressure vessel 6. The integral reactor of FIG. 1 is designed so that the upper vessel 8 contains the steam generators and may be removed as a unit (including the steam generators) for refueling. A common design goal for any reactor is to minimize the number of reactor vessel penetrations, but penetrations cannot be completely avoided as core components must have electrical, hydraulic, and instrumentation lines enter the reactor. Examples of core components that need electrical power or instrumentation lines are control rod drive mechanisms and incore instruments. Regarding the incore instruments, in some previous reactor designs, thimble guide tubes have provided the connections for the incores (instruments located in the reactor core), entering at either the top or bottom of the reactor vessel. Either location allows the tubes to run in a straight route: from the top of the vessel to the top of the core or, alternatively, from the bottom of the vessel to the bottom of the core. While having the thimble tubes enter the core from the bottom is desirable because it avoids interference with the control rods at the top of the core, penetrations in the bottom of the vessel are undesirable because they require draining the vessel for repair and, in an extreme Loss of Coolant Accident (LOCA), could inhibit the core from being submerged in coolant. Placing the incore penetrations at the top of the reactor in an integral reactor is also undesirable due to length and difficulty of routing around upper reactor components such as the control rod assemblies. If the reactor includes an integral pressurizer at the top of the reactor vessel, this further complicates routing of electrical cables from above the reactor. Cabling that runs from above the reactor also typically must be removed to provide access to the reactor core during refueling. Thus, routing the cables from the top of the vessel is undesirable. Routing from the bottom is also undesirable because it requires vessel penetrations below the core region. For these reasons, it is desirable to route cables from the side of the reactor vessel in an integral pressurized water reactor. Note that the term “guide tube” as used in the art may refer not only to the guide tube within the reactor core for guiding a control rod or incore, but also to the tube or conduit that houses a thimble tube outside of the core. To avoid confusion, the tube that houses the incore cabling outside of the core is referred to herein as a “routing tube.” The term “guide tube” is used herein to refer to the tube that is part of the reactor core. Also note that, in some embodiments disclosed herein, mineral insulated (MI) cabling replaces the thimble tubes, eliminating the need for a sealing table. In such embodiments, routing tubes provide support for the MI cables from the guide tube to the vessel wall, preventing flow induced vibration from damaging the cables. Because the routing tube does not exit the reactor, it need not be pressure tight. FIG. 2 shows the components contained in the illustrative lower vessel 10: upper internals 16 which are suspended from the mid-flange 14, an illustrative fuel assembly 18 (the reactor core comprises an array of such fuel assemblies), a core former 20, and core basket 22. The upper internals 16 may include components such as control rod guide frames that guide the control rod assemblies, internal CRDMs, internal reactor coolant pumps (RCPs), and so forth. (In FIG. 2, the upper internals 16 are illustrated showing only the structural frame including hanger plates and tie rods, and a set of RCPs at the mid-flange 14; however, in some embodiments the internal RCPs may be omitted in favor of externally mounted RCPs). In operation, the reactor pressure vessel contains primary coolant water at an operating temperature and pressure controlled by reactivity control of the core (e.g., using control rods, optional soluble neutron poison, or so forth) and pressure control provided by an internal or external pressurizer. The illustrative reactor is a PWR in which the primary coolant is maintained in a subcooled state, and includes an internal pressurizer comprising a steam bubble at the top of the upper vessel 8 and pressure control elements such as resistive heaters, spargers, or so forth (details not shown). The illustrative reactor is suitably operated at a pressure of about 2000 psia (14 MPa), and the electrical feedthroughs 4 are rated for an operating pressure of at least 2000 psia (14 MPa). In other reactor designs the design pressure may be higher or lower. For example, in a reactor with less stringent pressure requirements, the operating pressure may be around 1000 psia (7 MPa) and the electrical feedthroughs 4 are pressure tight at least for pressures up to 1000 psia (7 MPa). The exterior of the pressure vessel is at atmospheric pressure. As used herein, the term “atmospheric pressure” simply indicates the pressure of the atmosphere surrounding the pressure vessel. Typically, this is true atmospheric pressure, i.e. corresponding to the pressure of the atmosphere at the location of the nuclear reactor. However, it is contemplated for the pressure vessel to be disposed in a controlled atmosphere whose pressure is not precisely at true atmospheric pressure. For example, the pressure vessel may be disposed in a containment that is at a slight negative pressure to reduce likelihood that any airborne contaminant escapes the containment. In such an embodiment, “atmospheric pressure” suitably denotes this pressurized or negative pressure atmosphere surrounding the pressure vessel. Core parameters such as pressure, temperature, flow, and neutron flux may be monitored by various incores (that is, incore instruments and associated cables). Each incore is mounted in a guide tube in the core, and its cabling runs from the instrument, down the guide tube, and out of the core. Outside of the core, the cabling is protected from flow induced vibration by a routing tube or conduit which houses the cabling run to the vessel wall. By way of illustrative example, in the case of a neutron flux detector, there may be several Rhodium emitter type sensors mounted in a cable that is inserted into one of the guide tubes and connected to a material insulated cable. In one embodiment, five instruments are place along the height of the core in a guide tube. An Inconel sheath may protect the instruments. With returning reference to FIG. 1 and with further reference to FIGS. 3-6, the incore cabling preferably exits the reactor vessel at feedthrough 4 without creating a weak point in the reactor vessel. With reference to FIG. 3, a suitable embodiment of the feedthrough 4 includes a pressure retaining flange 26 connected to an elongated tube or cylinder 28 which provides the passageway for electrical conductors. (Although the illustrative tube or cylinder 28 has a circular cross-section, other cross sectional shapes such as a square or hexagonal cross-section are also contemplated). FIG. 4 is an end view of the feedthrough 4 viewed along the direction V indicated in FIG. 3. FIG. 5 shows a sectional view of a vessel penetration P through a wall W of the pressure vessel. The wall W has an outside surface Soutside and an inside surface Sinside. The interior volume of the pressure vessel is denoted by reference number 30 in FIG. 5. The pressure retaining flange 26 is located at an inside end of the cylinder 28 and is mounted to the inside surface Sinside of the pressure vessel, and may be secured and sealed by an annular weld 32 as diagrammatically indicated in FIGS. 4 and 5. In this arrangement, the pressure of the primary coolant water contained in the interior 30 of the pressure vessel presses the flange 26 against the inside surface Sinside of the pressure vessel wall W, which promotes pressure retention. The seal provided by the weld 32 seals the vessel penetration P. (In an alternative embodiment, the flange 26 is secured to the inside surface Sinside of the pressure vessel wall W by bolts threading into blind holes in the wall W, and the seal is provided by a gasket or o-ring that is compressed by the flange 26 against the wall W, so that the gasket or o-ring substitutes for the weld 32). As seen in FIG. 6, the elongated cylinder 28 is hollow and provides a passageway or lumen for guide tubes 34 sheathing electrical cables. The lumen of the elongated cylinder 28 is not sealed by the weld 32, and the flange 26 includes openings 33 through which the guide tubes 34 pass (see FIGS. 5 and 6). The openings 33 may be unadorned holes in the flange 26, or alternatively collars or the like may be disposed in the holes in the flange 26 to support the guide tubes 34 (as illustrated). In either case, the interior of the elongated cylinder 28 is in fluid communication with the interior 30 of the pressure vessel and is considered to be at interior pressure. An end plate 36 closes off the cylinder 28 at the end opposite from the flange 26. The end plate 36 is sealed to the elongated cylinder 28 to provide a pressure-tight termination of the cylinder 28. Each electrical cable sheathed by a guide tube 34 terminates at the end plate 36 via a pressure fitting/connector 38. In illustrative FIGS. 3-6 there are four cables sheathed by guide tubes 34, and hence there are four pressure fitting/connectors 38; however, the number of cables and pressure fitting/connectors can be 1, 2, 3, 4, 5, 6, or higher. Thus, the combination of the end plate 36 and pressure fitting/connectors 38 define an electrical termination disposed in the cylinder 28, namely at an outside end of the cylinder 28 in this embodiment, and the electrical termination 36, 38 seals the lumen of the cylinder 28 at the outside end of the cylinder 28. In this embodiment the pressure fitting/connectors 38 also provide external connectors at atmospheric pressure for electrical connection to the electrical feedthrough from outside the pressure vessel. Thus, the feedthrough 4 includes two seals. The first seal is provided by the flange 26 and the weld 32, and seals the vessel penetration P—but the elongated cylinder 28 passing through or at least partway into the vessel penetration P is at interior pressure (that is, at the operating pressure of the pressure vessel). The second seal is provided by the end plate 36 and the pressure fitting/connectors 38, and seals the outside end of the cylinder 28, that is, the end distal from the flange 26. The lumen of the elongated cylinder 28 can take various forms. In one approach, the elongated cylinder 28 has a single lumen (i.e., channel) and all cables pass through that single lumen. In this embodiment the cylinder 28 is substantially hollow and the lumen is suitably coaxial with the cylinder 28. This embodiment has advantages including manufacturing simplicity and the ability to accommodate different numbers of cables by modifying the end plate 36 and the flange 26. (Indeed, in this embodiment the openings 33 in the flange 26 optionally are replaced by a single opening large enough to pass all electrical cables, thus making the flange 26 also capable of accommodating different numbers of cables). Alternatively, the cylinder 28 can be substantially solid with multiple lumens (i.e., channels) passing through, with one lumen for each electrical cable. For example, if there are four electrical cables this embodiment would have four lumens disposed inside the cylinder 28. This embodiment has advantages including a higher overall flow resistance in the event of a LOCA occurring at the feedthrough 4 (assuming the LOCA occurs at only one of the four lumens). The guide tubes 34 protect the electrical cabling from exposure to the pressure and causticity of the environment inside the pressure vessel. In some embodiments, it is contemplated to employ mineral insulated cables (MI cables) with mineral insulation and metal sheathing that are capable of standing up to the environment inside the pressure vessel. In such embodiments, the guide tubes 34 are optionally omitted and the MI cables routed into the cylinder 28 without guide tubes. Although the illustrative flange 36 is circular, other shapes are contemplated. Whatever shape is used, the entire perimeter should be sealed by the weld 32 to create the pressure seal for vessel penetration P of the reactor vessel. In the case of a cylindrical pressure vessel or other pressure vessel configuration in which the inside surface Sinside has substantial curvature, the inside surface Sinside of the wall W of the pressure vessel is suitably machined flush or built up with filler to accept the flat flange. Alternatively, the flange could be machined to match the curvature of the inside surface Sinside. As previously mentioned, instead of welding the pressure retaining flange 26 can be secured by threaded fasteners (e.g. bolts as already mentioned, or studs welded to the interior of the pressure vessel) with a gasket providing the pressure seal. Welding is expected to provide a more reliable seal, but an arrangement secured by removable fasteners may be preferable if the feedthrough 4 is expected to be removed on occasion. In the case of the weld 32, the flange 26 may optionally have an annular J-groove (not shown) to accommodate the weld filler material. Each electrical cable is terminated at its pressure fitting/connector 38 at the exterior of the vessel. In some suitable embodiments, the pressure fitting 38 may be a compressed gland or metal compression seal. The lumen or lumens (i.e., channel or channels) of the cylinder 28 contain primary pressure up to the pressure fitting 38. If the guide tubes 34 are used in combination with MI cables, then the MI cable is housed inside (i.e., sheathed by) the guide tube 34. The length of the elongated cylinder 28 may be chosen to be long enough so that the end plate 36 protrudes from the vessel wall (i.e., extends outside of the outer surface Soutside), or the cylinder length may be chosen so that the end plate 36 is flush with the vessel wall, or the cylinder length may be chosen to be short enough so that the end plate 36 is inset into the wall (as diagrammatically shown in FIG. 5). The illustrated inset configuration advantageously protects the fittings 38 from inadvertent contact, but may make it more difficult to reach the fittings 38 for installation, removal, or maintenance. The optional use of a standard commercially available pressure fitting simplifies servicing of the incores and may reduce component cost, and additionally a standard fitting can be serviced with standard tools. With reference to FIGS. 7-10, another feedthrough embodiment is described. The feedthrough 40 of FIGS. 7-10 is well suited for use in feeding CRDM electrical connections or connections of other types in which there is a large number of electrical cables. The feedthrough 40 again includes a tube or cylinder 42 attached (e.g. welded or forged in one piece) to a flange 44. The flange 44 is again welded to the inside of the pressure vessel by an annular weld at the circumference 46, optionally having a j-groove for weld filler material. Multiple MI cables 48 from the CRDMs attach at the flange 44. The MI cables 48 may include CRDM power cables, CRDM instrumentation cables (e.g., rod bottom and rod height indicator cables), or so forth. Moreover, while an illustrative internal CRDM feedthrough application is described in conjunction with the feedthrough 40, it will be understood that the feedthrough 40 can be used in conjunction with electrical cables for any purpose. With reference to FIG. 8, in this embodiment the flange 44 contains an electrical termination in the form of a hermetically sealed connector 50, which may be a sealed glass connector, a sealed ceramic connector, a sealed glass ceramic connector, or so forth. The MI cables 48 plug into this connector 50. On the opposite side of the sealed connector 50, conductors (for illustration, only one conductor 52 is shown) inside the sealed tube 42 run from the hermetically sealed connector 50 to an exterior (i.e., outside) connector 54. In this embodiment the outside electrical connector 54 is at the outside end of the cylinder 42 and the electrical termination 50 is at the inside end of the cylinder 42. The tube or cylinder 42 in this embodiment does not need to be pressure tight as the hermetic seal 50 is the pressure boundary. Indeed, the lumen of the cylinder 42 can be at atmospheric pressure up to the seal provided by the hermetically sealed connector 50. The outside connector 54 is also not under pressure and may therefore be a conventional heavy duty, multi-pin connector. As with the embodiment of FIGS. 3-6, the connector 54 may protrude from the outer vessel wall, be flush with the vessel wall, or may be inset into the wall. In the embodiment of FIGS. 7-10, the MI cables 48 do not pass through the feedthrough 40, but instead terminate at (plug into) the sealed connector 50. The feedthrough 40 has connectors at either end: one connector for the terminated MI cable and one exterior to the vessel to connect to the instrumentation system. Conductors housed in the insulated tube 42 connect the two connectors. The tube itself may optionally also include mineral insulation. The pressure retaining flange 44 is permanently attached to the insulated tube 42 and is pressure tight, so no coolant enters the tube 42. The pressure retaining flange 44 also ensures that coolant does not leak between the sleeve and the vessel (i.e., thorough the vessel penetration around the outside of the tube 42). For connecting internal CRDMs, the feedthroughs 40 suitably pass through the mid-flange 14. This is advantageous because it allows the cables connected to the CRDMs to be removed with the mid-flange and upper-internals (see FIG. 2) during refueling. In a suitable configuration the feedthroughs 40 are oriented horizontally, and the bend in the MI cables 48 as they extend away from the flange 44 can be made gradually (to accommodate the limited bend radius of typical MI cables) until the MI cables are oriented vertically. However, it is alternatively contemplated that the CRDM feedthroughs pass through the upper or lower vessel. The MI cables may be covered with a seal weld at the pressure flange 44 to protect the electrical connectors. The disclosed electrical feedthroughs are described with illustrative applications to a nuclear reactor. However, it will be appreciated that the disclosed electrical feedthroughs may find application in other environments that employ a pressure vessel containing a working fluid at an operating pressure that is elevated as compared with atmospheric pressure (i.e., the pressure of the atmosphere surrounding the pressure vessel). The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	claims	1. A diagnostic device for a communication network including network equipments coupled to a management system, said device including:diagnostic means adapted to determine the cause of problems occurring in said network by means of one or more diagnostic models and processing means adapted to adapt said diagnostic model as a function of data provided by said management system that is representative of the modification of said network equipment; andsaid processing means including analysis means adapted to analyze the data, which is received from said management system, to determine actions to be undertaken to adapt said diagnostic model taking account of the network equipment modification or modifications that they represent and to generate instructions representative of said actions and adaptation means adapted to adapt a diagnostic model as a function of an instruction or instructions received from said analysis means includes adding at least one new node to the communication network. 2. The device claimed in claim 1 wherein said processing means include a first memory coupled to said analysis means and adapted to store one or more reference diagnostic models and said analysis means are adapted to determine each adaptation action to be undertaken by comparing data received to one or more reference diagnostic models stored in said first memory. 3. The device claimed in claim 2, wherein said analysis means are adapted to determine adaptation actions selected from a group including an action representative of a new model to be constructed, an action representative of a modification to be made to an existing diagnostic model, and an action representative of reconfiguration of an existing diagnostic model. 4. The device claimed in claim 3 further comprising a second memory coupled to said processing means and to said diagnostic means and adapted to store each diagnostic model transmitted by said processing means and wherein said adaptation means are adapted, on receiving an instruction requiring the generation of a new diagnostic model because of specific network equipment modifications, to extract from said first memory a reference diagnostic model adapted to the composition of said network taking account of said specific modifications, in order to generate a new diagnostic model and then to store it in said second memory. 5. The device claimed in claim 3 further comprising a second memory coupled to said processing means and to said diagnostic means and adapted to store each diagnostic model transmitted by said processing means and wherein said adaptation means are adapted, on receiving an instruction requiring the modification of an existing diagnostic model stored in said second memory because of specific network equipment modifications, to extract said existing diagnostic model from said second memory in order to modify it as a function of said specific modifications and then to store it when modified in said second memory. 6. The device claimed in claim 3 further comprising a second memory coupled to said processing means and to said diagnostic means and adapted to store each diagnostic model transmitted by said processing means and wherein said adaptation means are adapted, on receiving an instruction requiring the reconfiguration of an existing diagnostic model stored in said second memory because of specific network equipment modifications, to extract said existing diagnostic model from said second memory in order to reconfigure it as a function of said specific modifications and then to store it when reconfigured in said second memory. 7. The device claimed in claim 2 wherein at least some of said reference diagnostic models take the form of one or more Bayesian networks. 8. The device claimed in claim 1 further comprising a memory coupled to said processing means and to said diagnostic means and adapted to store each diagnostic model transmitted by said processing means. 9. The device claimed in claim 8 wherein said diagnostic means are adapted, on receiving a request for a selected diagnosis, to extract said selected diagnosis from said memory and then to execute the extracted diagnostic model in order to deliver a diagnosis at an output. 10. The device claimed in claim 1 wherein said data is transmitted by said management system in the form of notifications. 11. The device claimed in claim 1 wherein at least some of said diagnostic models take the form of one or more Bayesian networks. 12. A management system comprising a managed communication network and one or more diagnostic devices as claimed in claim 1. 13. The management system claimed in claim 12, constituting a network management system. 14. The management system claimed in claim 12, constituting a service management system. 15. The management system of claim 1, wherein said adaptation means adds at least one of (a) a new node and (b) a new Bayesian network structure. 16. A method of determining the cause of problems in a communication network including network equipment coupled to a management system having a computer readable medium configured to store instructions corresponding to steps of the method, said method including:determining the cause of problems occurring in said network by means of one or more diagnostic models; andadapting said diagnostic model as a function of data provided by said management system that is representative of the modification of said network equipment,wherein said adapting includes analyzing the data, which is received from said management system, to determine actions to be undertaken to adapt said one or more diagnostic models, taking account of the network equipment modification or modifications that they represent, and to generate instructions representative of said actions, andadapting a diagnostic model as a function of an instruction or instructions generated by the analyzing includes adding at least one new node to the communication network. 17. The method of claim 16, further comprising a memory coupled to said processing means and to said diagnostic means, and adapted to store each diagnostic model transmitted during said adapting. 18. The method claimed in claim 16 wherein said data is transmitted by said management system in the form of notifications. 19. The method of claim 16, wherein said adaptation comprises adding at least one of (a) a new node and (b) a new Bayesian network structure.
	description	1. Field of the Invention The present invention relates to a multi-core fuel rod for research reactor and a manufacturing method thereof and, more particularly, to a multi-core fuel rod for research reactor in which monolithic fuel cores made of uranium-molybdenum alloy are disposed in an aluminum (Al) matrix in the form of multi-core, and a manufacturing method thereof. 2. Description of the Prior Art A large amount of radiations and heat are generated when uranium undergoes nuclear fission. A power reactor utilizes heat generated from the nuclear fission and a research reactor utilizes radiations generated from the nuclear fission. A nuclear fuel is a material used for the nuclear fission. Generally, the research reactor has used highly enriched uranium alloy with uranium content above 90% as a nuclear fuel to obtain high neutron flux for an effective research. However the highly enriched uranium may increase a danger of nuclear proliferation, and thereby low enriched uranium alloy as a nuclear fuel has been developed since 1978 to replace the highly enriched uranium under the leading role of the United States. The main purpose of the development is to solve the problem by lowering the enrichment through the development of high-density nuclear fuel enabling high loading of uranium. Metal matrix dispersion fuel has been developed by dispersing uranium silicide (U3Si or U3Si2) in an Al matrix. The uranium silicide is a uranium alloy, which has relatively high uranium density and excellent stability in nuclear irradiation. The dispersion fuel is manufactured with fuel material of a uranium alloy type in a powder form mixed with heat carrier such as aluminum having high thermal conductivity, which keeps the fuel rod at low temperature. Since the late 1980's, highly enriched UAlx fuel has been replaced with low enriched fuel of uranium silicide, and the dispersion fuel using U3Si2 dispersed in Al matrix, as a nuclear fuel, has enabled successful development of a new research reactor which is allowed to increase fuel loading up to 4.8 gU/cc. A high performance research reactor requires high-density fuel, and researches for the high performance fuel had been carried out continuously. However researchers faced the problems that high-density fuel could not be manufactured satisfactorily and fuel reprocessing was not easy. Accordingly, another research was started to survey a material having higher density of uranium than uranium silicide fuel and enabling easier reprocessing. Since the late 1990's, U-Mo alloy fuel has been intensively developed among various applicable fuel materials, because the U-Mo alloy fuel can be manufactured in high density and has excellent stability in the nuclear reactor. A stepwise irradiation test has been carried out to evaluate the performance of U-Mo fuel. A good result is obtained when the irradiation test is performed in a low power operation, however a problem of fuel damages arises when the irradiation test is performed in a high power operation. In the case of high power operation, the temperature of fuel goes up high by a rapidly increased reaction between aluminum and uranium, and pores and intermetallic compound of UAlx are formed. The pores and low-density UAlx increase the volume of fuel, and cause swelling of fuel. The pores and UAlx having lower thermal conductivity further accelerate the temperature rise and swelling of fuel. Excessive swelling of fuel directly causes fuel damages. The reaction between aluminum and uranium is accelerated as the reaction surface area is being increased. FIG. 1 is a photo of U-Mo dispersion fuel taken after irradiation test, which shows U-Mo fuel particles, UAlx reaction layers, and aluminum matrix. Regardless of the fuel particle size, the thickness of each UAlx layer is almost identical, and the volume of UAlx is increasing as the reaction surface area is increasing. Therefore the reaction surface area should be reduced, because the increase of UAlx causes high temperature and swelling of fuel. Generally, monolithic fuel is more interested than a dispersion fuel, because the monolithic fuel has a smaller reaction surface area than the dispersion fuel. Nuclear fuel for research reactor may be classified into plate type and rod type, and an irradiation test for U-Mo monolithic fuel of the plate type has been carried out by U.S. ANL with a good result. Development of rod type monolithic fuel is proceeding in Korea, and the monolithic fuel rods of ring or tube type have been suggested. The above fuel has a disadvantage that reaction surface temperature is high in the case of increased uranium loading. The fuel has an advantage that uranium content can be controlled due to discontinuity in the longitudinal direction, however the fuel has a difficulty in manufacturing. A multi-core fuel rod for research reactor in accordance with the present invention has been developed by locating monolithic fuel with uranium-molybdenum alloy into a fuel rod in a multi-core form, in order to overcome the disadvantages in the conventional fuel rod. The present invention provides high stability of nuclear fuel by significantly reducing the formation of intermetallic compound between uranium alloy and aluminum metal. An object of the present invention is to provide a multi-core fuel rod for research reactor with high stability by significantly reducing the formation of intermetallic compound between uranium compound and aluminum, which causes performance drop of nuclear fuel. The formation of intermetallic compound may be reduced by minimizing the reaction surface area between uranium compound fuel and aluminum, and lowering the surface temperature. In order to achieve the above object, the present invention provides a multi-core fuel rod for research reactor by disposing monolithic fuel cores made of uranium-molybdenum alloy in an aluminum matrix in a multi-core form. Additionally, the present invention provides a manufacturing method of cylindrical monolithic fuel core by melting U-Mo alloy, and injecting the molten U-Mo alloy into a quartz tube having a specific diameter in a vacuum state. Further, the present invention provides a manufacturing method of multi-core fuel rod for research reactor having the monolithic fuel cores loaded in Al matrix in a multi-core form, and Al cladding surrounding the circumference of the Al matrix. Hereinafter, the present invention will be described in more detail. The present invention, an embodiment of which is shown in FIG. 4, relates to a multi-core fuel rod 10 for research reactor. Monolithic fuel cores 12 with a specific diameter are loaded into an Al matrix 14 in a multi-core form, along the positions located at the same distance from the center of Al matrix 14. An Al cladding 16 surrounds the circumference of the Al matrix 14. A multi-core fuel rod 10 is a fuel rod including a plurality of fuel cores 12. The aluminum matrix 14 surrounds the fuel cores 12 and acts as a heat carrier transferring heat effectively. Heat generated from each fuel core 12 is transferred to the heat carrier having high thermal conductivity and easily dissipated to the outside of the fuel rod. Accordingly, the aluminum enables a fuel rod to maintain fuel cores at low surface temperature. In the multi-core fuel rod for research reactor in accordance with the present invention, geometric form of fuel core is cylindrical as that of fuel rod. The number of cylindrical fuel cores is preferably controlled in the range of 2-10 and, more preferably in the range of 3-6. The diameter of fuel core is preferably controlled in the range of 1.0-3.0 mm according to the number of loaded fuel cores. In the case that the number of fuel cores is less than two or the diameter of fuel core is larger than 3.0 mm, it is not suitable for fuel rod because the maximum temperature of fuel core is increased. In the case that the number of fuel cores exceeds 10, it is undesirable because the structure of fuel rod becomes similar to a ring type, which is not helpful for improvement of thermal conductivity. Additionally, in the case of the diameter of fuel core is less than 1.0 mm, it is undesirable because an effect of utilizing multiple fuel cores may not be obtained. In the multi-core fuel rod for research reactor in accordance with the present invention, the monolithic fuel core may be pure uranium free from molybdenum. The fuel core having uranium density higher than 10 gU/cc may be applicable to the fuel for high-performance research reactor, which requires high uranium density. It is important to avoid formation of intermetallic reaction layer having low-density UAlx, when increasing the uranium density. The intermetallic reaction layer having low thermal conductivity decreases the heat transfer between fuel particles and Al matrix, and thereby causes a high temperature in the center of fuel particle as the burnup is proceeding. Additionally, a serious problem influencing the stability and performance of the fuel is caused by low-density reaction layer, which expands the volume of fuel core material and results in damages of cladding. The present invention solves the above problems by inserting fuel cores into Al matrix in a multi-core form, which reduces the reaction surface area significantly. In the multi-core fuel rod for research reactor in accordance with the present invention, the alloy fuel contains uranium and molybdenum in a specific ratio, where the molybdenum is added to the alloy fuel for the gamma-phase stability of uranium in the fuel core, preferably with the content of 6-10 wt % (weight percent). In the case that the molybdenum content is lower than 6 wt %, it is undesirable because gamma-phase stability of uranium is declined. In the case that the molybdenum content is higher than 10 wt %, it is also undesirable because relative content of uranium is decreased. In the conventional aluminum matrix of ring or tube form, and a U-Mo dispersion fuel, UAl2 and UAl3 formed by irradiation in the reactor further form UAl4 by a reaction with Al, generating heat. The reaction of UAl2 is faster than that of UAl3 and generates more heat. The reaction is accelerated according to the temperature rise, and increased sharply at the temperature above 525° C., resulting in formation of low-density UAlx reaction layers. Pores and low-density UAlx formed during the reaction increase the volume of fuel, causing swelling of the fuel and low thermal conductivity. According to the present invention, a plurality of fuel cores are disposed in a multi-core form at the same distances from the center of Al matrix, and significantly reduces the reaction surface area between U-Mo alloy and Al. Therefore, the present invention avoids the disadvantages due to formation of UAlx layer in the prior art, by substantially restraining the formation of intermetallic compound between uranium alloy and aluminum, which causes performance drop of nuclear fuel. Additionally, the present invention provides a manufacturing method of cylindrical monolithic fuel core loaded into an Al matrix. Although the manufacturing method of monolithic fuel core for research reactor may adopt various processes and are not specifically limited as long as a cylindrical alloy is manufactured by mixing uranium and molybdenum in a specific ratio, the fuel core may preferably be manufactured by a casting process. Firstly, an U-Mo alloy is melted in a furnace, and then an open-end of hollow quartz tube is immersed into the furnace in a vacuum state. Subsequently, the casting apparatus is exposed to the atmosphere and the molten U-Mo alloy is easily sucked into the quartz tube by a pressure difference between the inside of quartz tube and the surrounding atmosphere. Degree of vacuum in the quartz, tube is preferably maintained in the range of 2×10−2-6×10−2 torr. If the degree of vacuum is lower than the above range, the molten U-Mo alloy is not easily filled into the quartz tube. Alternatively, if the degree of vacuum is higher than the above range, a closed end of the quartz tube may break away. After filling and solidifying. the quartz tube with the molten alloy, the manufacturing process of cylindrical monolithic fuel core in accordance with the present invention is completed with breaking off the quartz tube only. Furthermore, a multi-core fuel rod for research reactor in accordance with the present invention may be manufactured by the following method. The fuel rod in accordance with the present invention may be manufactured through the steps comprising: forming circular holes inside the circumference of aluminum (Al) matrix, where a plurality of fuel cores will be located; inserting a plurality of fuel cores into the holes; casting the Al matrix by pilgering; cladding the circumference of the resultant product with Al. Core material of fuel rod comprises a rod-type Al alloy having a specific strength and cylindrical U-Mo fuel cores, and 2-10 holes are formed inside the circumference of rod-type Al so that cylindrical U-Mo fuel cores may be disposed in a multi-core form. Monolithic fuel cores manufactured in accordance with the present invention are inserted into the holes, and Al heat carrier is cast by pilgering to surround the fuel cores. Subsequently, both ends of the fuel rod is assembled with sealing plugs, and then extruded with Al in the form of concentric cladding. Both ends of the extruded fuel rod are processed so that the fuel rod can be assembled with end plates. The multi-core fuel rod for research reactor is obtained by processing the gaps between fuel rod cladding and Al sealing plugs with electron beam welding, after the sealing process. Hereinafter, a more detailed description will follow referring to the accompanying drawings for easier understanding of the present invention. FIG. 1 is a photo of uranium-molybdenum alloy taken after irradiation test of dispersion fuel in accordance with a comparison example. In the dispersion fuel, it is well shown that fuel particles of uranium alloy are dispersed in aluminum matrix and reaction layers are formed on the surface of fuel particles. The thickness of each reaction layer is almost same regardless of the fuel particle size. The reaction rate is increased as the temperature rises, and suddenly increased at the temperature above 525° C., and excessive intermetallic compound is formed, causing cracks due to the expansion of volume. The intermetallic compound has low thermal conductivity, and decreases the heat transfer between fuel particles and Al matrix, which causes the temperature rise in fuel particle center higher and higher as the burnup is proceeding. Additionally, there is a serious problem that the reaction layer having a low density gives a great influence to the stability and performance of fuel by destroying a cladding material, because the reaction layer increases the volume of fuel core. FIG. 2 is a cross-sectional view of fuel rod 10′ Formed of fuel core 12′ and cladding 16′ in the prior art. The fuel core is a fuel material of uranium alloy comprising dispersion fuel mixed with aluminum matrix. FIG. 2 shows a cross-section of fuel rod having dispersion fuel 2 shows a cross-section of fuel rod having dispersion fuel mixed with aluminum matrix and fuel particles of conventional type. FIG. 3 is a cross-sectional view of monolithic fuel rod 10″ having a fuel core 12 in accordance with a comparison example. As shown in FIG. 5, the maximum temperature of monolithic fuel core having a fuel core is 812° C. and the surface temperature of the fuel core is 177° C. The monolithic fuel has lower reaction surface temperature than dispersion fuel, however the maximum temperature is 812° C., higher than 600° C., and has low irradiation stability. Therefore, there is a problem that the maximum temperature of the monolithic fuel has to be lowered. FIG. 4 is a cross-sectional view of monolithic fuel rod 10 having 4 fuel cores 12 in accordance with an example embodiment of the present invention. The multi-core fuel rod for research reactor shown in FIG. 4 is manufactured by the same process as the fuel rods shown in FIG. 8. As shown in FIG. 5, the maximum temperature of 4 fuel cores is 322° C. and the maximum surface temperature is 195° C., indicating that the fuel rod is suitable for research reactor. FIG. 5 is a graph showing an ANSYS model of temperature distribution analysis of fuel rod having 1-6 fuel cores in accordance with an example embodiment of the present invention. As the number of fuel cores increases, the maximum temperatures of fuel core and reaction surface decrease, indicating that the fuel rod is suitable for research reactor. FIG. 6 is a schematic view of casting apparatus to manufacture a fuel core in accordance with an example embodiment of the present invention. FIG. 7 is a photo of fuel core with the diameter of 2 mm and length of 120 mm manufactured by a casting apparatus of FIG. 6, in accordance with another example embodiment of the present invention. FIG. 8 shows photos of multi-core U-Mo fuel rods in a manufacturing process in accordance with example embodiments of the present invention. The upper photo in FIG. 8 shows the fuel cores and Al heat carrier having holes to accommodate the fuel cores. The lower left photo shows the cross-section of aluminum heat carrier processed according to the size of fuel core. The lower right photo shows the cross-section of fuel rod with Al cladding according to a conventional extrusion process. Hereinafter, the present invention will be described in more detail by illustrating example embodiments of the invention are illustrated. It will be appreciated that the invention may be embodied in many different forms and should not be construed as limited to the particular embodiments set forth herein. A multi-core fuel rod for research reactor is manufactured with the same method as Example 1, except that the Al carrier is manufactured to accommodate 6 fuel cores. A type of conventional U-Mo fuel particles dispersed in Al matrix is shown in FIG. 2. A monolithic fuel rod 10″ having a fuel core 12″ in the center of Al matrix 14″ is shown in FIG. 3. Under the condition of 7 wt % Mo in U-Mo fuel, the reaction surface areas of dispersion fuel of comparison example 1 are compared to those of monolithic fuel in accordance with the example embodiment of the present invention, of which diameter is controlled to have the same density as the dispersion fuel. The diameter of dispersion fuel particle is 50 or 100 μm and the monolithic fuel is a fuel rod having 1 or 4 fuel cores. The uranium density of dispersion fuel is 6 gU/cc and the diameter of fuel rod is 6.35 mm. The diameter of monolithic fuel having the same uranium density as the dispersion fuel having the particle diameter 50 or 100 μm corresponds to 1.93 or 3.85 mm respectively. Each reaction surface area is calculated and comparative ratios are listed in Table 1, based on the reaction surface area of dispersion fuel having a particle size of 50 μm. TABLE 1Comparison factorReactionsurfaceareaComparativeFuelDiameterNumbers(cm2)ratioDispersion 50 μm5,620,6771393.01.0fuel 100 μm702,585697.00.5Monolithic1.93 mm42.420.0017fuel3.85 mm11.210.0009 The monolithic fuel having 4 fuel cores has only 1/576 times of the reaction surface area compared to the dispersion fuel having 50 μm fuel particles. Swelling of the monolithic fuel is only 1/576 times compared to the dispersion fuel, under the condition that the reaction layers have the same thickness. Temperature distribution of the monolithic fuel has been calculated with regard to the number of fuel cores. ANSYS code is utilized to calculate the temperature distribution. As shown in FIG. 5, four models of temperature calculation have been established. The temperature of nuclear fuel has a close relationship with linear power (power per unit length). Under the same condition of linear power 120 kW/m, the temperature distribution of monolithic fuel rod having a plurality of fuel cores have been compared to that of dispersion fuel. (1) Temperature Distribution of Dispersion Fuel The maximum reaction surface temperature of dispersion fuel in accordance with Comparison Example 1 is 214° C. (2) Temperature Distribution of Fuel Rod having a Fuel Core The maximum temperature of monolithic fuel having a fuel core is 812° C., and the reaction surface temperature of fuel core is 177° C., in accordance with Comparison Example 2. As described above, it is well shown that the monolithic fuel has a lower reaction surface temperature than the dispersion fuel. The uranium-molybdenum alloy has high irradiation stability below 600° C., however the maximum temperature of monolithic fuel having a fuel core is too high. There is therefore a subject that the temperature has to be lowered. (3) Temperature Distribution of Fuel Rod having 3 Fuel Cores The maximum temperature of monolithic fuel having 3 fuel cores in accordance with Example 1 of the present invention is 375° C. As described above, it is clearly shown that the maximum temperature of fuel is 375° C. and this type is suitable for fuel rod. (4) Temperature Distribution of Fuel Rod having 4 Fuel Cores The maximum temperature of monolithic fuel having 4 fuel cores is 322° C., and the reaction surface temperature of fuel core is 195° C., in accordance with Example 2 of the present invention. As described above, the maximum surface temperature of fuel rod is lower than 214° C., that of dispersion fuel, indicating that the reaction of U-Mo fuel with Al can be reduced. It is well shown that the maximum temperature of fuel is 195° C. and this type is also suitable for fuel rod. (5) Temperature Distribution of Fuel Rod having 6 Fuel Cores The maximum temperature of monolithic fuel having 6 fuel cores is 266° C., in accordance with Example 3 of the present invention. It is well shown that the maximum temperature of fuel is 266° C. and this type is also suitable for fuel rod. The multi-core monolithic fuel in accordance with the present invention has a smaller reaction surface area and lower maximum temperature of reaction surface than those of dispersion fuel having the same uranium density, and thereby may be used as a more stable fuel rod, because the reaction between U-Mo fuel and Al may be restrained. Additionally, the multi-core monolithic fuel in accordance with the present invention is expected to effectively restrain the formation of UAlx reaction layer, which reduces the performance of U-Mo fuel. The multi-core fuel rod for research reactor in accordance with the present invention provides a structure having minimized contact surface area between nuclear fuel and aluminum, and thereby prevents excessive reaction between the fuel and aluminum, and reduces the formation of pores and swelling by restraining formation of reaction layer. Additionally, the multi-core fuel rod in accordance with the present invention has a high thermal conductivity, so that heat inside the fuel rod can be efficiently dissipated. Therefore, stability and performance of nuclear fuel may be improved by avoiding the problem that the stability of fuel rod is lowered due to temperature rise in the conventional fuel rod. Furthermore, the uranium density of multi-core fuel is high compared to the conventional dispersion fuel, and may be applicable to a high-performance research reactor, which requires a high uranium density.
	abstract	A power module assembly includes a reactor core immersed in a coolant and a reactor vessel housing the coolant and the reactor core. An internal dry containment vessel submerged in liquid substantially surrounds the reactor vessel in a gaseous environment. During an over-pressurization event the reactor vessel is configured to release the coolant into the containment vessel and remove a decay heat of the reactor core through condensation of the coolant on an inner surface of the containment vessel.
044329334	description	DETAILED DISCLOSURE OF THE INVENTION Briefly, the invention contemplates the use of small, hollow shells of a material such as glass having the dimensions desired for the final product. These shells are exposed to the hydrogen isotopes in the form of gas such as deuterium-tritium under conditions of pressure and heat such that the gas will permeate through the walls of the hollow shells or microspheres and be entrapped within. The filled microspheres may then be used either with the fuel in gaseous form or the fuel may be deposited on the inner walls of the sphere by subjecting the sphere to a cryogenic temperature. The hollow shells to be used for the fuel configuration and in the process will be referred to as microspheres and may be formed of glass, ceramic, carbon, plastic or metal as a basic initial structure. These hollow spheres are available commercially and are sometimes identified by the trademarks MICROBALLOON, ECCOSPHERES and CARBO-SPHERES. They can be obtained in diameters from 10 micrometers to 1000 to 2000 micrometers. These spheres are presently used commercially in syntactic foams, low density structure materials, dielectrics and thermal insulation. Pertinent patents describing the processes and the product of glass and other microspheres are listed in the following Table I. TABLE I ______________________________________ Patent No. Issued Inventors Title ______________________________________ 2,797,201 6/25/57 Veatch Process of Producing Hollow Particles and Resulting Product 3,138,444 6/23/64 Searight Method and Apparatus for Manufacturing Glass Beads 3,161,468 12/15/64 Walsh Process for Producing Hollow Spheres of Silica 3,365,315 1/23/68 Beck Glass Bubbles Prepared by Reheating Solid Glass Particles 3,441,396 4/29/69 D'Eustachio Process for Making Cellular Materials 3,615,972 10/26/71 Morehouse Expansible Thermo- plastic Polymer Particles ______________________________________ The invention will be described in connection with glass microspheres although it will be appreciated that other materials might be utilized. Basically, the microspheres of glass (which is the principal material proposed) are formed by pulverizing a glass containing volatile compounds and injecting the pulverized glass into a gas stream which passes through a hot zone such as that formed by a torch. While in the hot zone, the volatile compounds are vaporized and expand and this, when coupled to the inherent surface tension of glass, causes the molten glass to take the form of a hollow sphere. The process is controlled by the selection of the glass composition, the hot zone temperature, the velocity at which the gases are passed through the hot zone, and the cooling provided thereafter. The commercially available microspheres have been designed to give a relatively low density and therefore have had walls in the neighborhood of 1 to 2 microns thick. The processes may be used to produce microspheres with much thicker walls if this is desired. The following Table II gives examples of properties of commercially available glass microspheres produced by Emerson and Cuming, Inc. Table IIA gives some properties of various available glass compositions. TABLE II ______________________________________ Properties of Commercially Available Glass Microballoons Grade Property IG 101 SI FTL200 ______________________________________ Composition Sodium Silica >95% SiO.sub.2 Borosilicate Particle Size Range, Microns (% by weight) >175 5 0 2 149-175 10 14 8 125-149 12 10 12 100-125 12 12 13 62-100 44 40 52 44-62 10 15 11 <44 7 9 2 Average Particle 80 80 90 Diameter Microns Average Wall Thick- 2 1.5 1.5 ness Microns Softening Tempera- 900 1800 2000 ture or Melting Temperature (.degree.F.) Compressive Strength Volume % Survivors at Pressure (psi) 500 96.3 1000 66.4 1500 46.2 ______________________________________ TABLE IIA ______________________________________ Properties of Glass Annealing Softening Code Type Composition Point Point ______________________________________ vycor (C) 96% SiO.sub.2 910.degree. C. 1500.degree. C. 3% B.sub.2 O.sub.3 1% Al.sub.2 O.sub.3 silica (G) 100% SiO.sub.2 1180.degree. C. 1730.degree. C. fused silica 7740 (C) sodium 81% SiO.sub.2 565.degree. C. 820.degree. C. borosilicate 13% B.sub.2 O.sub.3 2% Al.sub.2 O.sub.3 4% Na.sub.2 O, K.sub.2 O 650 (L) sodium 90% SiO.sub.2, B.sub.2 O.sub.3 -- -- borosilicate 3% Al.sub.2 O.sub.3 0080 (C) soda lime 72% SiO.sub.2 510.degree. C. 696.degree. C. 1% Al.sub.2 O.sub.3 10% CaO, MgO 17% Na.sub.2 O, K.sub.2 O H phosphate 5% B.sub.2 O.sub.3 -- -- 77% P.sub.2 O.sub.5 11% Al.sub.2 O.sub.3 7% ZnO 1720 (C) 62% SiO.sub.2 715.degree. C. 915.degree. C. aluminosilicate 5% B.sub.2 O.sub.3 18% Al.sub.2 O.sub.3 15% CaO, MgO P x-ray shield glass 31% SiO.sub.2 323.degree. C. 379.degree. C. 8% BaO 61% PbO G lead borate 22% B.sub.2 O.sub.3 78% PbO ______________________________________ It will be seen from the above that it is within the state of the art to independently regulate not only the size of the microsphere but the wall thickness and, of course, the chemical composition of the glass. There are two thermonuclear pellet designs which are of immediate interest with respect to the present disclosure. These are shown in FIGS. 1 and 2 respectively. In FIG. 1, there is shown the structural shell 20 in the form of a cross-section of a sphere and this contains a gaseous thermonuclear fuel 22 which, of course, is invisible in the drawing. The hollow solid shell may be made of glass as indicated above or ceramic, metal, plastic, or carbon and it will contain a gaseous mixture of hydrogen isotopes such as deuterium or a deuterium-tritium mix. In FIG. 2, the structural shell 20 is shown but in this case it has layered inside of it a solid coating of the thermonuclear fuel 24 leaving a vacuum centrally of the sphere. In the forming of the above pellet designs, there are certain steps in the process which are illustrated in FIG. 3 showing a flow chart. Step #1 in the process involves the procurement and selection of the microspheres. As has been indicated above, these microspheres are available commercially and the selection involves a consideration of the chemical composition, the size, the wall thickness, the wall strength and the residual gas content. The preferred structural shell material is glass because of its inherent strength, its availability in desired sizes at nominal costs in large quantities, and its relative uniform sphericity and wall thickness. The selection of a specific glass composition is based upon its permeability to hydrogen isotopes (hereafter including all isotopes of hydrogen individually or in mixtures and in the ortho or para spin states) and its softening temperature. High silica contents favor high permeabilities and high softening temperature. The dilution of silica with other glass formers (B.sub.2 O.sub.3, Na.sub.2 O, K.sub.2 O, Li.sub.2 O, Al.sub.2 O.sub.3, PbO and others) lowers both the permeability and the softening temperature. Any desired glass microsphere size can be sorted out by well-known techniques, wet or dry sieving, cyclone separation, hydraulic elutriation, microscopic hand sorting or micro-radiography to mention a few. The sphericity of the microspheres can be evaluated by micro-radiographic techniques. The wall thickness dimensions and uniformity can be evaluated by quantitative, micro-radiography and by employing mass separation techniques (such as gas suspension on carefully sized microspheres). Compressive and tensile wall strengths can be determined by first applying a hydraulic compressive pressure. Separation is performed by introducing the microspheres into a fluid bath. The sealed microspheres float and the broken parts or leaky microspheres sink and thereby separate. Then tensile load may be applied to the microsphere walls by permeation filling of the microspheres by the procedure described later to a pressure higher than desired ultimately. Survivors can be recovered by the sink-float technique mentioned above. There are residual gases within the selected microspheres which are derived from the blowing agents used in the manufacture. These may be CO.sub.2, H.sub.2 O, or SO.sub.2 or mixtures thereof. In some instances, it may be desirable to evacuate these gases by a diffusion process, but the present process contemplates a selective lamination of these gases to reduce the effect on the fusion process. The first step of the process also involves a proper cleaning of the microspheres to remove the particulate matter which may be present on the exterior surface. A typical cleaning process utilizes hot trichloroethylene for the removal of organic material and a water detergent solution wash in an ultrasonic agitated bath for the removal of salts and other water soluble materials. These steps may be followed by a rinse in pure water, an etch in 10% solution of hydrofluoric acid to strengthen the microsphere, a further rinse in pure water to remove the acid residue, and a drying step in methanol. The second basic step in the process is permeation or diffusion filling of the microspheres. This involves loading the selected microspheres into a commercially available pressure vessel as illustrated in FIG. 4 and pressurizing the vessel with deuterium-tritium gas while heating the pressure vessel to temperatures ranging from 150.degree. C. to 800.degree. C. as limited by the sintering temperature of the specific glass used to prevent adherence of the particles to each other. The pressure vessel in FIG. 4 has a suitable interior chamber surrounded by heater coils connected to a power supply. Suitable temperature indicators and thermocouples are provided. A lock-on cover has a gas pressure inlet and a pressure gauge. The gas filling pressure is predetermined to provide the desired quantity of deuterium-tritium within the sphere. First of all, it must be appreciated that the permeation rates of hydrogen or hydrogen isotopes such as deuterium-tritium and similar materials through glass is known, and such data is available in a book entitled Vacuum Technique by Saul Dushman, published by John Wiley & Sons, Inc., New York 1949. The following Table III indicates permeation data taken from this publication. TABLE III ______________________________________ Diffusion Data Compilation System Permeability Data ______________________________________ SiO.sub.2 --H.sub.2 ##STR1## SiO.sub.2 --H.sub.2 at 700.degree. C. K = 2.1 .times. 10.sup.-9 SiO.sub.2 --D.sub.2 at 700.degree. C. K = 1.7 .times. 10.sup.-9 SiO.sub.2 --H.sub.2 at 900.degree. C. K = 6.4 .times. 10.sup.-9 Pyrex-H.sub.2 at 520.degree. C. K = 2.0 .times. 10.sup.-11 * ______________________________________ K = permeability (cc of gas (STP) per sec per cm.sup.2 area per mm thickness per cm of Hg pressure head) T = temperature (.degree.K.)? *N. W. Taylor and W. Rast, "The Diffusion of Helium and of Hydrogen Through Chemically Resistant Glass," Journal of Chemical Physics, 6, October 1938, p. 619. From the above equation in Table III, it has been determined that with respect to H.sub.2, silica has a permeability K factor of 3.times.10.sup.-9 at 750.degree. C. and a K factor of 2.times.10.sup.-11 at 25.degree. C., while sodium borosilicate has a K factor of 2.times.10.sup.-11 at 520.degree. C. Using the published information, it will be seen that hydrogen permeation through silica (SiO.sub.2) shows permeability at 520.degree. C. to 1.1.times.10.sup.-9. The permeability for Pyrex (typically Corning-code 7740) at the same temperature is 1.97.times.10.sup.-11. Integrating the well-known Fick's law equation for diffusion under the proper conditions for filling the hollow spheres, the following equation was derived: ##EQU1## where t.sub.0.99 =time for the inside hydrogen pressure to reach 99% of the external hydrogen pressure (sec.) x=thickness of glass wall (mm) PA1 r=radius of small hollow sphere (cm) PA1 T=temperature of filling (.degree.K.) PA1 K=permeability (cc of gas (STP) per sec. per cm.sup.2 area per mm thickness per cm of Hg pressure head. Using Equation (1) and the gas permeabilities given above, the specific time required to raise the internal hydrogen pressure to 99% of the external pressure is between 50 and 2600 sec. for a temperature of 520.degree. C. The external hydrogen pressure and, therefore, the internal hydrogen pressure attained are not limited by the permeation requirements. The strength of the glass wall limits the pressure gradient across the glass wall during filling and later during storage where the external pressure is reduced to 1 atmosphere. The ultimate pressure contained in a hollow glass microsphere is limited by the bursting strength of the shell. Using the familiar hoop stress formula (Equation 2), the pressure across the wall (.DELTA.P) can be related to the tensile strength (T.S.) of the microsphere wall. ##EQU2## The tensile strength of glasses has received a great deal of attention in the literature. It has been established that surface imperfections greatly affect the tensile strength. Normally, a strength of 100 psi is used for structural design. However, it is also known that glass fibers having 1-2 micron diameters exhibit tensile strengths between 700,000 and 2,000,000 psi. From all published data, it is very likely that the strength of the glass used for glass microspheres has the higher strengths. Using Equation 2, it has been calculated that internal pressures of 15,000 psi or roughly 1000 atmospheres could be contained in glass microspheres. Current requirements for thermonuclear fuel pellets indicate 100 atmospheres is adequate. The pressure and temperature of the deuterium-tritium gas surrounding the microspheres is maintained for a period sufficient to cause the pressure inside the microsphere to rise to 99% of that of the external pressure. This period has been precisely established by a series of experiments in which the particular microspheres are treated at various temperatures for various periods. After these tests, and to confirm the hydrogen isotope pressure inside the microspheres after the diffusion filling, the microspheres are broken and the quantities of released gases are measured by (1) mass spectrometer gas analysis, and (2) actual size of the gas bubble in a fluid. As one example, a batch of microspheres was treated at 300.degree. C. under 100 atmospheres of pressure for a period of 36 hours to obtain an internal pressure of approximately 99 atmospheres and a contained gas density of 10.sup.-2 grams per cubic centimeter. After holding the temperature and pressure for the desired time period, the temperature of the pressure vessel is then lowered to room temperature and finally the gas pressure external to the microsphere is relieved. Inasmuch as the permeation rate of the gas at room temperatures is typically 10.sup.4 times less than that at elevated temperatures, the contained gases will remain trapped in the microspheres during subsequent handling operations. Using the SiO.sub.2 -H.sub.2 equation (Table III) for permeability, the permeability diminishes to 1.41.times.10.sup.-13 at room temperature and that for Pyrex at least an order of magnitude less. The leakage rate can be predicted by the same formulae (Equation 1) developed for the filling prediction. Using Equation 1, the time required to reduce the internal hydrogen pressure to 1% of the original in a quartz microsphere is over 100 hours, and in a Pyrex microsphere over 1000 hours. Even further reductions in differential gas pressure and permeability can be realized by storing the spheres at lower temperatures (dry ice or liquid nitrogen temperatures). After the pressure filling step, it may be desirable to separate microspheres which have been broken in the process or by mechanical handling from those that are intact after the filling operation. This can again be easily accomplished by utilizing a flotation process. The completion of Step #2, that is, the filling with the hydrogen isotopes, will provide a thermonuclear fuel pellet which may be utilized in a fusion process as described in the aforesaid referenced Brueckner applications. A third step in the process which may be in some cases optional involves a coating of the filled microspheres. There may be a variety of reasons for providing this coating. In the first place, it may serve to provide a better seal for the microsphere as a diffusion barrier to improve storage. For example, a lead or bismuth glass, a soda-lime glass, copper, or aluminum coating may serve this function. Secondly, a coating may be applied to improve the surface properties with respect to increasing the laser light absorption. Thirdly, a coating may thicken the structural wall to improve the pellet implosion properties; and, fourthly, a coating may serve as an energy channel to provide a more uniform fuel pellet illumination. These layers may be applied by vacuum vapor deposition, ion sputtering, chemical sprays, electrolytic plating, or fluidized bed techniques. The fourth and final step in the processing of the fuel pellets involves subjecting the pellets to a cryogenic temperature. There are two purposes for this procedure. First, any impurity gases within the pellet will be selectively frozen out on to the inside surface of the structural shell by selecting a temperature intermediate between the freeze-out temperature of the impurity gas and the freeze-out temperature of the hydrogen isotopes. For example, at liquid nitrogen temperatures CO.sub.2 and H.sub.2 O will freeze out leaving a purified hydrogen isotope gas in the pellet core. This step alone will provide a pellet which may be utilized in fusion operations if suitable laser input power is available. The second reason for the use of the cryogenic temperatures would be to freeze out the hydrogen isotope gas as a solid layer on the inside of the microsphere. There are basic and important reasons from the point of view of the fusion process for having the nuclear fuel in this particular configuration. Freezing of the gases within the microspheres may be accomplished in a number of different ways. It is known that liquid hydrogen, deuterium and tritium will wet a glass surface which insures that condensation on the glass will form a continuous uniform layer as the temperature decreases. The rate of temperature decrease is also important to insure an even layer of fuel. The freezing rate can either be such that any liquid deposit cannot conglomerate unevenly in the sphere, or it is also possible to hold the microspheres in suspension while rotating or tumbling them during the cooling operation. This can be done, for example, by suspending the microsphere in a glass capillary tube while flowing cold helium at <13.degree. K. upwardly through the tube. The suspended microsphere will rotate or tumble, thus providing uniform cooling on all sides so as to achieve even distribution of the freezing gas within. There is considerable information in the literature relative to the freezing of hydrogen isotopes. The following Table IV summarizes the critical points regarding the liquification and solidification of hydrogen isotopes and their mixtures. TABLE IV __________________________________________________________________________ Critical, Boiling Point, and Triple Point Constants Predicted Graphically for the Hydrogen Isotopes Comparisons of Predicted Values with Available Experimental Values Boiling Point Critical Constants Triple Point T.sub.b (.degree.K.) T.sub.c (.degree.K.) P.sub.c (atmos) V.sub.c (cc/mole) T.sub.t (.degree.K.) P.sub.t (atmos) expt graph expt graph expt graph expt graph expt graph expt graph __________________________________________________________________________ H.sub.2 20.39 20.4 33.24 33.2 12.797 12.8 66.95 67.0 13.96 14.0 0.071 0.071 HD 22.13 22.4 35.91 36.4 14.64 15.1 62.8 62.8 16.60 16.8 0.122 0.130 38.35 16.432 D.sub.2 23.57 23.6 38.26 38.3 16.28 16.4 60.3 60.3 18.72 18.5 0.169 0.164 41.1 18.9 56.7 HT 23.6 38.3 16.4 60.3 18.5 0.164 DT 24.3 39.5 17.3 58.5 19.7 0.188 T.sub.2 24.92 24.9 40.6 18.1 57.1 20.27 20.5 0.207 0.207 43.7 20.8 53.7 21.65 0.248 __________________________________________________________________________ This table is taken from an article by A. S. Friedman, D. White, and H. L. Johnson, entitled "Critical Constants, Boiling Points, Triple Point Constants and Vapor Pressure on Fixed Isotopic Hydrogen Molecules Based on Simple Mass Relationship," Journal of Chemical Physics, Vol. 19, No. 1, January 1951. It will be seen from this table that while the various isotopes of hydrogen vary in detail from one another, in general, their critical constants and triple point are about the same and therefore the discussion of one isotope would approximate the other isotopes. Additional information on the characteristics of deuterium is shown in an article by R. Prydz, K. D. Timmerhaus, and R. B. Stewart entitled "The Thermal Dynamic Properties of Deuterium," published in Cryogenic Engineering, Vol. 13 (1968), pages 384 to 396. This article includes temperature-enthalpy diagrams. These diagrams are used to predict the behavior of the deuterium during cooling. They indicate that cooling under equilibrium conditions and starting with D.sub.2 gas pressures between 10 to 100 atmospheres in a sealed microsphere, the gas will condense to liquid deuterium, then to solid ice. Standard calculations on condensing liquids indicate that it would take greater than 0.3 of a second for appreciable flow to occur. This flow must be avoided to insure a uniform coating. Best results are obtained by freezing the hydrogen isotope to solid ice in a fraction of a second to prevent liquid flow within the microsphere. This is accomplished by placing the microsphere in a liquid helium pool so that the cooling is the result of thermal conductivity through the silica. Standard heat transfer calculations indicate that it takes 8.times.10.sup.-5 seconds to go through the freezing process. This freezing time is very favorable from the point of view of retaining a uniform liquid layer on the inside of the microsphere during the cooling period. It is highly probable that intermediate cooling mechanisms providing a slightly slower freezing time might be quite adequate to retain the liquid hydrogen in place while solidification is completed. The presence of tritium gas in the microsphere will contribute a small amount of energy by radioisotopic heating. However, this has been evaluated and would have a minor effect on the calculations above discussed. Filling of microspheres with H.sub.2, D.sub.2, and T.sub.2 by permeation has been accomplished under the above-suggested conditions of heat and pressure. Cleaned and sized microspheres were placed in a pressure vessel illustrated in FIG. 4 filled with hydrogen at a pressure of 100 atmospheres (1500 psi), heated to 300.degree. C., and held for 96 hours. The pressure at this temperature was 2800 psi. After treatment or permeation filling, the pressure in the surviving individual microspheres was determined by breaking a filled microsphere while submerged in mineral oil, and observing with a microscope the size of the resulting gas bubble. The results indicated 40 to 90 atmospheres of included gas depending on the wall thickness of the microspheres as compared with 1/3 atmosphere prior to permeation filling. After 30 days at ambient temperatures, there was no detectable change in the determined internal pressure. The above process has also been accomplished using deuterium-tritium gas. It will thus be seen that the disclosed process may be utilized to produce enclosed minute and predetermined quantities of fusion fuel in gaseous form, and also the same fuel may be provided in the form of a hollow spherical shell of solid material within the glass microsphere held at cryogenic temperatures.
	description	This application claims priority to Chinese Patent Application No. 201710432819.9 filed on Jun. 9, 2017, the contents of which are hereby incorporated by reference in their entirety. The present invention relates to nuclear reactor apparatus technology, and more particularly to a nuclear reactor cooling system. Nuclear reactor is an apparatus used for starting, controlling and maintaining the nuclear fission or fusion chain reaction. The reaction rate of the nuclear reactor can be controlled precisely so as to release the energy at a slow speed for the use of people. There are various uses of the nuclear reactor, the most important one is to replace other fuel to generate heat as the steam electric power or the power for driving large apparatuses such as aircraft carrier. Nuclear power generated from the nuclear reactor is an important power resource that any possible fault occurs during the operation of the nuclear reactor should not be overlooked. Containment Spray System (EAS) is a protection system used to lower the pressure and the temperature in the containment when coolant accident happens in the primary circuit. In the existing art, as a protection system capable of running in a condition of power outage, passive containment spray system guarantees the safety of the nuclear reactor. Therefore, further optimizing the structure of the passive containment spay system to improve the reliability in cooling containment is an important research direction for skilled person in the art. Therefore, the present invention provides a nuclear reactor cooling system with higher reliability in cooling containment. According to an embodiment of the present invention, the nuclear reactor cooling system comprises: a cooling water tank disposed above a containment; a spray header connected to the cooling water tank through a first communicating pipe; and an air tank disposed inside the containment; wherein the spray header is disposed outside the containment and used for spraying cooling water to an outer wall of the containment; the cooling water tank is a closed container, and the air tank is connected to a top portion of the cooling water tank through a second communicating pipe. Specifically, the cooling water tank is used to store cooling water which is used as containment coolant. Disposing the cooling water tank above the containment makes the cooling water in the cooling water tank flow into the spray header through the first communicating pipe under gravity and then spray to the outer wall of the containment via the spray header, cooling the containment. Meanwhile, the air tank used to store gas is disposed inside the containment with its internal pressure affected by the temperature in the containment, therefore once the cooling water leak happens in the primary circuit, the temperature in the containment increases and the gas in the air tank expands, increasing the pressure in the cooling water tank. The increased pressure allows the cooling water flow into the spray header more quickly, increasing the flow rate of the cooling water. The faster the temperature in the containment increases, the larger the flow rate becomes, enhancing the cooling ability to the containment, bringing higher reliability of the cooling system. In another embodiment according to the previous embodiment, in order to get a full contact between the outer wall of the containment and the cooling water so as to improve the cooling effect for the containment and in order to collect the cold fluid after the heat exchange between the containment and the cooling water, the nuclear reactor cooling system further comprises a bell shaped shield, wherein the shield is used for covering the containment and setting the containment in its interior, the spray header is disposed in a space between an inner wall of the shield and the outer wall the containment, and the cooling water tank is disposed on a top portion of the shield. The space between the inner wall of the shield and the outer wall the containment is a flow passage for the cold fluid, wherein the cold fluid includes liquid cooling water and gaseous steam formed in the heat exchange between the containment and the cooling water. When the cooling water falls onto the outer wall of the containment, some of it is heated into gaseous steam, and the rest remains liquid. In another embodiment according to the previous embodiment, the nuclear reactor cooling system further comprises an exhaust hole disposed on the top portion of the shield, connecting the space between the shield and the containment with the outer space of the shield so as to stop the temperature of the cold fluid outside the containment from increasing to a high level to affect the heat exchange between the cold fluid and the containment. In another embodiment according to the previous embodiment, the cooling water tank has a shape of annulus and its axis is collinear with that of the shield, bringing stronger stability of the nuclear reactor cooling system. In another embodiment according to the previous embodiment, the nuclear reactor cooling system further comprises a cooling water outlet disposed on a bottom portion of the shield to discharge the cold fluid in the space between the shield and the containment timely, bringing high-efficiency cooling for the containment. In another embodiment according to the previous embodiment, for there are many communicating pipes and nozzles disposed on the spray header, the spray header is axisymmetrically disposed above the containment, cooling the containment uniformly and enabling stronger stability of the nuclear reactor cooling system. In another embodiment according to the previous embodiment, the nuclear reactor cooling system further comprises a rupture disk disposed in the air tank and/or in the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the air tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. During the normal operation of the reactor, the temperature in the containment fluctuates in a small range resulting in small change of the pressure in the containment, so that the cooling water stored in the cooling water tank can spray out in normal condition, in order to avoid the aforementioned situation, the rupture disk is disposed in the air tank and/or in the second communicating pipe. The rupture disk is preferably fixed in a pipe which is removable from the second communicating pipe and located outside the containment, so as to replace the rupture disk or the assembly of rupture disk and the pipe after the rupture of the repture disk. In this way, when the pressure in the space between the air tank and the rupture disk increases to a certain pressure, the rupture disk ruptures and the cooling water in the cooling water tank flows into the spay header and is spayed to the outer wall of the containment, cooling the containment. Compared to the existing art, the present invention has advantages and beneficial effects as follows: The cooling water tank is used to store cooling water which is used as containment coolant. Disposing the cooling water tank above the containment makes the cooling water flow into the spray header from the cooling water tank through the first communicating pipe under gravity and then spray to the outer wall of the containment via the spray header, cooling the containment. Meanwhile, the air tank used to store gas is disposed inside the containment with its internal pressure affected by the temperature in the containment, therefore once the cooling water leak happens in the primary circuit, the temperature in the containment increases and the gas in the air tank expands, increasing the pressure in the cooling water tank. The increased pressure allows the cooling water flow into the spray header more quickly, increasing the flow rate of the cooling water. The faster the temperature in the containment increases, the larger the flow rate becomes, enhancing the cooling ability to the containment, bringing higher reliability of the cooling system. For making the above and other purposes, features and benefits become more readily apparent to those ordinarily skilled in the art, the preferred embodiments and the detailed descriptions with accompanying drawings will be put forward in the following descriptions. The present invention will now be described more specifically with reference to the following embodiments. It is to be noted that the following descriptions of preferred embodiments of this invention are presented herein for purpose of illustration and description only. It is not intended to be exhaustive or to be limited to the precise form disclosed. As is shown in the FIGURE, a nuclear reactor cooling system comprises a cooling water tank 3 disposed above a containment 1; a spray header 4 connected to the cooling water tank 3 through a first communicating pipe 5; and an air tank disposed inside the containment; wherein the spray header 4 is disposed outside the containment 1 and used for spraying cooling water to an outer wall of the containment; the cooling water tank is a closed container, and the air tank is connected to a top portion of the cooling water tank through a second communicating pipe 6. Specifically, the cooling water tank 3 is used to store the cooling water which is used as containment coolant. Disposing the cooling water tank 3 above the containment 1 makes the cooling water in the cooling water tank 3 flow into the spray header 4 through the first communicating pipe 5 under gravity and spray to the outer wall of the containment via the spray header by the spray header 4, cooling the containment 1. Meanwhile, the air tank 8 used to store gas is disposed inside the containment 1 with its internal pressure affected by the temperature in the containment 1, therefore once the cooling water leak happens in the primary circuit, the temperature in the containment 1 increases and the gas in the air tank 8 expands, increasing the pressure in the cooling water tank 3. The increased pressure allows the cooling water flow into the spray header 4 more quickly, increasing the flow rate of the cooling water. The faster the temperature in the containment 1 increases, the larger the flow rate is, enhancing the cooling ability to the containment 1, bringing higher reliability of the cooling system. As is shown in the FIGURE, in this embodiment according to embodiment 1, in order to get a full contact between the outer wall of the containment 1 and the cooling water so as to improve the cooling effect for the containment 1 and in order to collect the cold fluid after the heat exchange between the containment and the cooling water, the nuclear reactor cooling system further comprises a bell shaped shield 2, wherein the shield 2 is used for covering the containment 1 and setting the containment 1 in its interior, a space 9 between an inner wall of the shield 2 and the outer wall the containment 1 is formed, the spray header 4 is disposed in the space 9 which is used as a flow passage for the cold fluid, wherein the cold fluid is formed by the heat exchange between the cooling water and the containment 1, including liquid cooling water and gaseous steam. The cooling water tank 3 is disposed on a top portion of the shield 2. Further, the nuclear reactor cooling system further comprises an exhaust hole 11 disposed on the top portion of the shield 2 connecting the space 9 between the shield 2 and the containment 1 with the outer space of the shield so as to stop the temperature of the cold fluid outside the containment 1 from increasing to a high level to affect the heat exchange between the cold fluid and the containment 1. Further, the cooling water tank 3 has a shape of annulus and its axis is collinear with the axis of the shield 2, providing strong ability for supporting, bringing stronger stability of the nuclear reactor cooling system. Further, the nuclear reactor cooling system further comprises a cooling water outlet 10 disposed on a bottom portion of the shield 2 to discharge the cold fluid in the space between the shield 2 and the containment 1 timely, bringing high-efficiency cooling for the containment 1. Further, for there are many communicating pipes and nozzles disposed on the spray header 4, axisymmetrically disposed above the containment, the spray header 4 cools the containment uniformly from side and have a strong structural stability, enabling stronger stability of the nuclear reactor cooling system. As is shown in the FIGURE, in this embodiment according to embodiment 1 or embodiment 2, the nuclear reactor cooling system further comprises a rupture disk 7. During the normal operation of the reactor, the temperature in the containment 1 fluctuates in a small range resulting in small change of the pressure in the containment 1, so that the cooling water stored in the cooling water tank 3 can spray out in normal condition, in order to avoid the aforementioned situation, a rupture disk is disposed in the air tank 8 and/or in the second communicating pipe 6, wherein the rupture disk 7 is ruptured during an increase of pressure in the air tank 8, and the rupture disk 7 in an intact state is capable of isolating a space on both sides thereof. The rupture disk 7 is preferably fixed in a pipe which is removable from the second communicating pipe 6 and located outside the containment 1, so as to replace the rupture disk 7 or the assembly of rupture disk 7 and the pipe after the rupture of the repture disk. In this way, when the pressure in the space between the air tank 8 and the rupture disk 7 increases to a certain pressure, the rupture disk 7 ruptures and the cooling water in the cooling water tank 3 flows into the spay header 4 and is spayed to the outer wall of the containment 1, cooling the containment 1. While the invention has been described in terms of what is presently considered to be the most practical and preferred embodiments, it is to be understood that the invention needs not be limited to the disclosed embodiments. On the contrary, it is intended to cover various modifications and similar arrangements included within the spirit and scope of the appended claims which are to be accorded with the broadest interpretation so as to encompass all such modifications and similar structures.
052532780	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A fuel assembly in accordance with an embodiment of the present invention applied to a boiling water reactor will be explained with reference to FIGS. 1 and 2. The fuel assembly 1 of this embodiment includes an upper tie plate 2, a lower tie plate 3, a large number of fuel rods 4 whose both ends are supported by the upper and lower tie plates 2, 3 and fuel spacers 5 for keeping the mutual gaps between the fuel rods 4. The fuel rods 4 are disposed in grid with a predetermined pitch. A channel box 6 which is fitted to the upper tie plate 2 encompasses the bundle of fuel rods and extends downward. The lower end portion of the channel box 6 extends below the upper surface of the lower tie plate 3 and encompasses this lower tie plate 3. The upper tie plate 2 is equipped on its upper surface with a handle 7 and with four posts 8. Furthermore, the upper tie plate 2 has a large number of bosses 9 disposed in grid as shown in FIG. 3. Adjacent bosses 9 are coupled with one another by ribs 10. Those spaces 11 which are encompassed by a plurality of bosses 9 and a plurality of ribs 10 define coolant passages and penetrate through the upper tie plate 2. An upper end plug of each fuel rod 4 is inserted into the hole portion bored in each boss 9. A boss 9A at the center of the upper tie plate 2 supports the upper end plug of a water rod 12. Each fuel spacer 5 has a plurality of cylinder members 13, a plurality of loop-like springs 14 and a band 15 for encompassing the outer portion of each cylinder member 13 at the outermost periphery as shown in FIGS. 2A and 2B of U.S. Pat. No. 4,508,679 and in FIG. 2 of U.S. Pat. No. 4,686,079. Each fuel rod 4 is inserted into each cylinder member 13. Seven fuel spacers 5 are disposed with predetermined gaps between them in the axial direction of the fuel assembly 1. The water rod 12 is disposed at the center of the cross-section of the fuel assembly 1. As depicted in FIG. 3 of U.S. Pat. No. 4,686,079, the water rod 12 is supported by the fuel spacers 5 in the horizontal direction and supports the fuel spacers 5 in the axial direction, on the contrary. The outer diameter of the water rod 12 occupies the area in which four fuel rods 4 can be disposed. The channel box 6 is a cylindrical member whose cross-section is square. A clip 16 which is placed on the upper surface of the post 8 is disposed at each corner of the channel box 6. Two kinds of tabs are disposed on the four side surfaces of the channel box 6, respectively. These tabs are defined by projecting part of the side walls of the channel box 6. Accordingly, recesses 17A, 18A are defined on the outer surfaces of the channel box 6. The tabs of one of the kinds are reinforcing tabs 17 and the other kind of tabs are support tabs such as tabs 18 for supporting the spacers and tabs 19 for supporting the upper tie plate. Each reinforcing tab 17 is a projecting portion that extends thinly in a transverse direction, and its width c is considerably greater than its height a in the axial direction. The reinforcing tabs 17 are disposed at positions between the position of the second fuel spacer 5 from the bottom (the second fuel spacer) and the lower end of the channel box 6. The second fuel spacer 5 is placed at a position which is 1/4 of the full length distant from the lower end of the channel box 6. These reinforcing tabs 17 function as creep deformation inhibition means of the channel box 6. A corrugated portion which repeats concavoconvexities in the axial direction of the channel box 6 is defined by a plurality of reinforcing tabs 17. The corrugated portion which exhibits resistance to creep deformation is a creep deformation inhibition portion. The fuel spacers 5 are referred to as the "first fuel spacer", "second fuel spacer", . . . , "seventh fuel spacer" from below in order named. The support tabs 18 and 19 are projecting portions that extend thinly in the axial direction of the fuel assembly 1. The thinly extending direction of these tabs 18 and 19 crosses the direction of the reinforcing tabs 17. Each support tab 18 is disposed in such a manner as to correspond to the level of each fuel spacer 5. Two support tabs 18 are disposed for each fuel spacer 5 on each side surface of the channel box 6. As shown in FIGS. 2 and 4, each support tab 18 comes into contact with the band 15 of the fuel spacer 5 and supports the fuel spacer 5 in the horizontal direction. The support tab 19 comes into contact with the outer side surface of the upper tie plate 2 and supports the upper tie plate 2 in the horizontal direction. These support tabs 18 and 19 are disposed at the positions where they face the second fuel rods 4 from the fuel rods 4 positioned at the corners among the fuel rods disposed at the outermost periphery. Unlike the prior art technology, those tabs which come into contact with the inner surface of the channel box are not formed on the band 15 of the fuel spacer 5. Accordingly, the band 15 can be shaped by merely bending a flat sheet and the production becomes easier. The height h.sub.2 (see FIG. 2) of the support tab 18 is greater than the height h.sub.1 of the fuel spacer 5 so that the support tabs 18 can support the fuel spacers 5 even when the fuel rods 4 extend in the axial direction upon thermal expansion during the operation of the unclear reactor and a difference of the moving distance occurs between the fuel spacers 5 and the channel box 6. The coolant passages 20 are defined between the inner surface of the channel box 6 and the outer surface of the bands 15 of the fuel spacers 5 except for the portions of the support tabs 18. The coolant passages 21 are defined between the inner surface of the channel box 6 and the outer side surface of the upper tie plate 2 except for the portions of the support tabs 19, as well. Incidentally, the height e of the support tabs 18 and 19 from the inner surface of the channel box 6 must be greater than height d of the reinforcing tabs 17 so that the channel box 6 can be fitted to the fuel bundle. The term "fuel bundle" used hereby means the remaining portions of the fuel assembly 1 other than the channel box 6. The support tabs 18 for supporting the first fuel spacer 5 are positioned between the reinforcing tabs 17 in the axial direction. The gap b between the adjacent reinforcing tabs 17 is equal to the height a of the reinforcing tabs 17 in the axial direction. Hole portions into which lower end plugs of the fuel rods 4 and water rod 12 fitted are bored in the lower tie plate 3 in the same way as in the upper tie plate 2. These hole portions exist at the fuel holding portion 3A of the lower tie plate 3. The insertion portion 3B which is disposed at the lower part of the lower tie plate 3 is inserted into a fuel support fixture (not shown in the drawing) inside the core. The center 0 of the insertion portion 3B exists on the axis of the fuel assembly 1. The lower tie plate 3 is equipped with a groove 22 extending from its upper surface to its lower surface on each sidewall of its four side surfaces. A recess which extends downward from the upper surface of the tie plate 3 and is closed at the lower end may be disposed on each outer sidewall in place of the groove 22 described above. The reinforcing tabs 17 disposed at the lower end of the channel box 6 are fitted into the grooves 22. Therefore, the reinforcing tabs 17 are disposed near to the lower end of the channel box 6. Moreover, the width of the water gaps defined between the fuel assemblies loaded into the core are not reduced by the reinforcing tabs 17. In other words, loadability of the control rods between the fuel assemblies is not impeded by these reinforcing tabs 17. The width W.sub.1 between the outer surfaces of the opposed sidewalls of the band 15 of the fuel spacer 5 (FIG. 4) and the width W.sub.2 between the portion of the outer side surfaces of the upper tie plate 2 that come into contact with the tabs 19 must be smaller than the width W.sub.3 between the bottom surfaces 22A of the mutually opposed grooves 22 of the lower tie plate 3. In this embodiment, these widths W.sub.1 and W.sub.2 are smaller than the width W.sub.3. In this embodiment the support portions of the fuel spacers 5 in the horizontal direction are formed in the channel box 6 and for this reason, the height d (FIG. 2) of the reinforcing tubs 17 can be made higher. Accordingly, creep deformation of the channel box can be reduced remarkably. Incidentally, the height d of the reinforcing tabs 17 can be increased maximum to the height e of the support tabs 18. FIG. 6 shows the relation between the ratio of the height d of the reinforcing tabs 17 to the thickness t of the channel box 6 (d/t) and a equivalent thickness. If the height d is 0.8 t, for example, it is possible to obtain the strength equivalent to the state where the thickness of the channel box 6 is about 1.3 t. FIG. 6 shows the characteristics when a=b and the effect of increasing the strength by the reinforcing tabs 17 attains the maximum when a=b. Since the reinforcing tabs 17 are formed between the lower end of the channel box 6 and the position equal to 1/4 of the full length of the channel box 6, creep deformation of the lower part of the channel box 6 can be inhibited drastically. Particularly because the reinforcing tabs 17 are formed at the portion of the channel box 6 opposing the outer side surface of the lower tie plate 3, too, creep deformation in the proximity of the lower tie plate 3 can be inhibited remarkably. Furthermore, the width c of the reinforcing tubs 17 is set preferably to be at least 80% of the width D (FIG. 3) of the portions of the sidewall on one outer surface of the channel box 6 in the horizontal direction other than the curved portion at the corner of the sidewall. Particularly in this embodiment, the width c of the reinforcing tabs 17 can be made greater than the gap W.sub.0 between the adjacent support tabs 18 in the horizontal direction (FIG. 4) without being limited by this gap W.sub.0. The increase in the height d and width c described above improves remarkably the strength of the channel box 6; hence, creep deformation of the channel box can be inhibited remarkably. Accordingly, the residence time of the channel box 6 inside the core can be extended remarkably and the channel box 6 can cope with high burnup of the fuel assembly, and the like. Since the height d is below the height e in this embodiment, the reinforcing tabs 17 are not caught by the fuel spacers 5 and by the upper tie plate 2 at the time of mounting of the channel box 6. For, the channel box 6 is fitted to the fuel bundle from the side of the upper tie plate 2. Particularly, the present embodiment does not need to pass the reinforcing tabs 17 between a plurality of tabs dispose in the fuel spacers at the time of fitting of the channel box as has been necessary in the prior art technology of Japanese Patent Laid-Open No. 13894/1979. Accordingly, the channel box 6 can be fitted to the fuel bundle within a short period of time and its fitting is easy. In this embodiment, creep deformation at the lower part of the channel box 6 can be reduced remarkably. Therefore, the margin of the water gap width between the fuel assemblies 1 with respect to the insertion property of the control rods becomes greater. Part of this creep deformation inhibition effect is helpful to reduce the thickness of the channel box 6 directed to reduce the neutron absorption quantity under the state where an excellent insertion property of the control rods is maintained. This reduction of thickness improves the neutron utilization ratio of the fuel assembly. The decrease in the creep deformation inhibition effect due to the reduction of thickness does not become the problem to the insertion property of the control rods and the problem to the application of the channel box 6 to the fuel assembly having high burnup (at least 38 GWd/t) because the reinforcing tabs 17 and the support tabs 18 are disposed. Creep deformation at the time of high burnup is small. The channel box having a reduced thickness can be applied particularly to the fuel assembly of at least 50 GWd/t, too. The increase in the pressure loss can be checked because the support tabs 18 support the fuel spacers 5. For, the coolant passages 20 are defined. When the reinforcing tabs 17 support the fuel spacers 5, the area of the flow path of the coolant passage 20 at the level of the fuel spacers 5 becomes remarkably small and the pressure loss increases drastically. Support of the upper tie plate 2 by the support tabs 19 prevents the increase of the pressure loss and improves rigidity of the fuel assembly. The width between the inner surfaces of the opposed sidewalls of the channel box 6 is substantially the width W.sub.5 (FIG. 2) between the inner surfaces of the opposed reinforcing tabs 17, at the portions below the second fuel spacer 5. Above this second fuel spacer 5, on the other hand, the width between the inner surfaces of the opposed sidewalls of the channel box 6 is substantially the width W.sub.4. Since W.sub.4 >W.sub.5 ' the pressure loss is high at the lower part of the fuel assembly 1 and is small at its upper part. Therefore, channel stability of the fuel assembly 1 can be improved. This embodiment can increase the strength of the lower pate of the channel box (since the height d is great) and can improve the critical power in the same way as in the fuel assembly 1A shown in FIG. 9 due to the increase in the height e. The characteristics of later-appearing FIG. 10 remains substantially unchanged in this embodiment, too. The characteristics of G.sub.2 /G.sub.1 .apprxeq.1.0 are for the conventional fuel assembly equipped with the channel box whose sidewalls are straight. The present boiling water reactors can be classified broadly into a reactor into which fuel assemblies each having a channel box having the width W.sub.4 of 132.46 mm between the inner surfaces of the opposed sidewalls are loaded and a reactor into which fuel assemblies each having a channel box having the diameter W.sub.4 of 134.06 mm are loaded. The fuel rod pitches are different between these reactors. The pitch of the latter is greater than that of the former. These channel boxes have those sidewalls which do not have each of the tabs described already but are straight in the axial direction. The application of this embodiment to these fuel assemblies having different shapes provides the fuel bundle having a shape that can be used in common for both of them. Accordingly, only one kind of fuel bundle may be prepared for a plurality of kinds of cores and the production of the fuel bundles can be simplified. The support tabs 18 splash cooling water that rises along the inner surface of the channel box 6 toward the fuel rods 4. Due to this function of the support tabs 18, cooling water (liquid) on the inner surface of the channel box 6 that does not originally contribute to cooling of the fuel rods is helpful to cool the fuel rods 4 at the upper part of the fuel assembly 1 where gas-liquid two flows exist. As a result, the cooling effect of the fuel rods near the corners can be improved. The support tabs 18 are preferably positioned between the fuel rods at the corners and the fuel rods 4 adjacent to the former. The critical power of the several fuel rods 4 positioned at the corners can be further improved. The shape of the support tabs 18 may be the same as that of the spacer support tabs 18A and 18B such as shown in FIGS. 17 and 18. A production method of the channel box 6 used for the fuel assembly 1 of this embodiment will be explained with reference to FIG. 7. After a sheet of a zirconium alloy (such as zircaloy-4) is cut in a predetermined shape, the sheet is bent in a U-shape by press machining (step 30). FIG. 8(a) shows the sheet bent in the U-shape. Next, two U-shaped zirconium alloy members are butt-welded on their side surfaces as shown in FIG. 8(b) to produce a cylinder member 35 (step 31). Reference numeral 48 represents a weld portion. The weld bead is flattened by roller machining (step 32) to obtain a cylinder member 35A shown in FIG. 8(c). The tabs 17, 18 and 19 are defined on the sidewalls of the cylinder member 35A by press machining (step 33). As a result, a cylinder member 35B shown in FIG. 8(d) is obtained. Press machining of the step 33 is carried out as shown in FIG. 8(e). A die 36 having a plurality of recesses 37 formed on its surface so as to correspond to the tabs 17-19 is inserted into the cylinder member 35A while penetrating through the cylinder member 35A in its axial direction. The cylinder member 35A is placed on the bed (not shown) of the press. Both end portions of the die 36 are supported by two arms that are fitted to the driving unit of the press outside the cylinder member 35A. One of the arms comes off from the die 36 when the die 36 is inserted into the cylinder member 35A (or when it is removed from the latter, on the contrary). Another die 38 having a plurality of projections 39 is positioned outside the cylinder member 35A. This die 38 is fitted to a rod of the press that moves up and down, though it is not shown in the drawing. Each projection 39 is fitted into the corresponding recess 37 of the die 36 when the die 38 lowers. Accordingly, the tabs 17, 18 and 19 are formed on one sidewall of the cylinder member 35A. Similarly, the tabs are formed on the other sidewalls. Since the formation of each tab is made after the formation of the cylinder member, the tabs can be formed easily even on the sidewall on which the weld portion 48 exists. The flattening work of the weld bead can be made easily, as well. After the step 33 is completed, expanded pipe formation and annealing are carried out for the cylinder member 35B (step 34). Expanded pipe formation and annealing are carried out simultaneously. Their definite method is described in Japanese Patent Laid-Open No. 131354/1982, page 2, lower left column, line 17 to page 33, lower right column, line 9 and in FIGS. 5 and 6. The step 34 is carried out particularly preferably in accordance with the method described on page 3, upper right column, line 16 to lower right column, line 9 and in FIG. 6 of this reference. When the method of the reference described above is employed, it is necessary to form a recess for inserting the tabs 17 or the like on the outer surface of "tyuseishi Jigu" shown in FIG. 6. Since expanded pipe formation and annealing are carried out after the tab formation, strain that occurs at the time of the tab formation can be eliminated and a channel box 6 having high dimensional accuracy can be obtained. The step 33 may be carried out after the step 34 in order to improve dimensional accuracy of each tab. However, dimensional accuracy drops at the time of the expanded pipe formation and strain remains due to the press work and for this reason, machining of the step 34 must be conducted once again. The channel box 6 produced in the manner described above is fitted to the upper tie plate 2 of the fuel bundle, and there is thus obtained the fuel assembly 1. The channel box 6 has two support tabs 18 disposed at the same level on each sidewall but three or more support tabs 18 may be disposed, as well. If the number of support tabs 18 at the same level on one sidewall is greater, the surface pressure per tab can be reduced more greatly and deformation of the fuel spacer 5 can be restricted more effectively. If the number of support tabs 18 is too great, however, the pressure loss of the fuel assembly becomes great. Accordingly, the number of the support tabs 18 at the same level on one sidewall is preferably 2 or 3. FIG. 9 shows the fuel assembly 1A in accordance with another embodiment of the present invention which is obtained by applying the channel box 6 to the fuel assembly described in Japanese Patent Laid-Open No. 311195/1988, page 5, lower right column, line 14 to page 7, lower right column, line 11 and in FIGS. 1-5. The fuel assembly 1A comprises the fuel rods 4A that are arranged in 9 rows by 9 columns. Two water rods 12A having an outer diameter which is greater than the pitch of the fuel rods are disposed at the center of the cross-section of the fuel assembly 1A. The rest of the construction other than the structure described above are the same as those of the fuel assembly 1. The two water rods 12A occupy the area in which seven fuel rods 4A can be disposed. The fuel spacer 5A is substantially the same as the fuel spacer shown in FIGS. 4 and 5 of Japanese Patent Laid-Open No. 311195/1988. The support tabs 18 support the fuel spacers 5A in the horizontal direction. This fuel assembly 1A provides the same effect as that of the fuel assembly 1. FIG. 10 shows quantitatively the increase in the critical power obtained in this embodiment. Symbol G.sub.1 represents the gap width between the adjacent fuel rods 4A as shown in FIG. 9 and G.sub.2 is the gap width between the inner surface of the channel box 6 and the fuel rods 4A that are adjacent to this inner surface with the shortest distance. The conventional fuel assembly equipped with the channel box whose sidewalls are straight, shown in FIG. 3 of Japanese Patent Laid-Open No. 311195/1988, has the characteristics of G.sub.2 /G.sub.1 .apprxeq. 1.0. The height d of the reinforcing tabs 17 can be increased by increasing the height e of the support tabs 18 and the strength at the lower part of the channel box 6 can be further increased. At the same time, since the gap width G.sub.2 is great and the gap width G.sub.1 becomes small, the critical power increases much more than when G.sub.2 /G.sub.1 .apprxeq. 1.0. (Refer to the characteristics when G.sub.2 /G.sub.1 .apprxeq.1.2, 1.4 and 1.7, respectively.) This embodiment can improve the critical power by 10% in comparison with the prior art technology and can increase also the operation margin of the reactor. The local power peaking factor of the fuel assembly is greater than 1.0 and during the period of use of the fuel assembly, the local power peaking factor is generally within the range of from about 1.1 to about 1.25. An embodiment of the core of a boiling water reactor when the fuel assembly shown in FIG. 1 is applied thereto will be explained with reference to FIGS. 11 to 13. FIG. 11 shows the conventional core 43 into which the conventional fuel assemblies 41 equipped with the channel box having the straight sidewalls are loaded and this is part of the core shown in FIG. 2 of Japanese Patent Publication No. 37911/1988. One cell includes four fuel assemblies 41 that are adjacent to the control rods 42 and encompass the control rods 42. The upper end portion of each fuel assembly 41 is rushed to the upper core grid plate 40 by a channel fastener (not shown). In the fuel assembly 41, the width W.sub.4 ' (FIG. 13) between the inner surfaces of the opposed sidewalls of the channel box is 134.06 mm. In the core 43, the gap width W.sub.6 of the water gap into which the control rods 41 are inserted is equal to the water gap W.sub.7 below the upper core grid plate 40 into which the control rods 42 are not inserted. In FIG. 11, symbol C.sub.1 represents the fuel assembly 41 which is subjected to the operation of the first fuel cycle; C.sub.2 is the fuel assembly 41 subjected to the operation of the second fuel cycle; and C.sub.3 is the fuel assembly 41 subjected to the operation of the third fuel cycle. A fuel assembly C.sub.1 is a novel fuel assembly (burnup 0 GWd/t) at the start of the operation of a certain fuel cycle. When the operation of one fuel cycle is completed, the fuel assembly C.sub.3 is taken out from the core as the used fuel assembly and the fuel assembly 1 having burnup of 0 GWd/t is loaded instead. Such a fuel exchange constitutes the core 43A (FIG. 12) of the core of this embodiment which is subjected to the operation of the next fuel cycle. In the core 43A, the fuel assemblies C.sub.2 and C.sub.3 are the fuel assembly 41 and the fuel assembly C.sub.1 is the fuel assembly 1. FIG. 13 shows in enlargement the X.sub.1 portion of FIG. 12. The fuel assembly 41 is equipped with the channel box 6A whose sidewalls are straight. The fuel spacer 5B is the fuel spacer 5 equipped with the tabs 18A on its band. The tabs 18A come into contact with the inner surface of the channel box 6A and support the fuel spacr 5B in the horizontal direction. Since the channel box 6 is equipped with the support tabs 18 and the reinforcing tubs 17, the strength of the lower part of the channel box 6 is higher than of the channel box 6A and the neutron absorption quantity of the channel box 6 is smaller than that of the channnel 6A. The thickness t.sub.2 of the channel box 6 is smaller than the thickness t.sub.1 of the channel box 6A. For example, t.sub.1 is 2.54 mm and t.sub.2 is 2.00 mm. The fuel assembly 1 will be compared with the fuel assembly 41. The width W.sub.4 between the inner surfaces of the opposed sidewalls of the channel box 6 is equal to the width W.sub.4 ' of the channel box 6A and is 134.06 mm. If the width W.sub.4 and the width W.sub.4 ' are different within the range of tolerance (e.g..+-.0.25 mm), the width W.sub.4 and W.sub.4 ' are substantially equal to each other. the width W.sub.8 between the outer surfaces of the opposed sidewalls of the channel box 6 is smaller than the width W.sub.8 ' of the channel box 6A because the thickness t.sub.2 is smaller. Accordingly, the channel spacer 44 that comes into contact with the core upper grid plate 40 is disposed at the portion of the channel box 6 facing the core upper grid plate 40. The width W.sub.9 between the outer surfaces of the opposed sidewalls of the band 15 of the fuel spacer 5 is smaller than the corresponding width W.sub.9 ' of the band of the fuel spacer 5B. The thickness of the band 15 of each fuel spacer is equal. As to the fuel assembly 1 loaded into the core 43A, the area of the coolant passage inside the channel box 6 is substantially equal to that inside the channel box 6A of the fuel assembly 41. Since the width W.sub.9 is smaller than the width W.sub.9 ', G.sub.2 /G.sub.1 of the fuel assembly 1 is greater than that of the fuel assembly 41 whose G.sub.1 /G.sub.2 value is about 1.0. Accordingly, the critical power of the fuel assembly 1 is greater than that of the fuel assembly 41. This embodiment utilizes the decrement of the thickness obtained by reducing the thickness of the channel box 6 for increasing the water gap width W.sub.6 and W.sub.7. Accordingly, the moderation effect of the neutron can be improved around the fuel assembly 1 and reactivity increases. When the operation of one fuel cycle is completed in the core 43A, the fuel assembly C.sub.3 of the core 43A is withdrawn as the used fuel and the new fuel assembly of burnup of 0 GWd/t is loaded. Another embodiment of the core of the boliling water reactor to which the fuel assembly 1 is applied will be explained with reference to FIGS. 14 and 15. Reference numeral 40A represents the upper core grid plate and 41A does the conventional fuel assembly. The width of the grid of the upper core grid plate 40A is bigger than that of the upper core grid plate 40. The fuel assembly 41A has the channel box 6B whose sidewalls are straight. The width W.sub.4 ' of the channel box 6B is 132.46 mm. The width W.sub.9 ' of the fuel assembly 41A is smaller than that of the fuel assembly 41. The core construction in the fuel cycle ahead of the fuel cycle carried out by the core 43B of this embodiment is such that only the fuel assemblies 41A having different core residence time are disposed as the fuel assemblies in the same way as in FIG. 11. The used fuel assembly C.sub.3 is withdrawn from this core and the new fuel assembly 1 having burnup of 0 GWd/t is loaded. This fuel exchange provides a core 43B. Here, the fuel assembly C.sub.1 is the fuel assembly 1 and the fuel assemblies C.sub.2 and C.sub.3 are the fuel assemblies 41A. FIG. 15 shows in enlargement the X.sub.2 portion in FIG. 14. The fuel assembly 41A has the same construction as the fuel assembly 41 except only that the width W.sub.4 ' of the channel box is smaller than that of the fuel assembly 41. In comparison with the channel box 6B, the channel box 6 used in the fuel assembly 1 of this embodiment increases the strength at the lower part and can reduce the neutron absorption quantity. The thickness t.sub.2 of the channel box 6 is smaller than the thickness t.sub.1 of the channel box 6B. For example, t.sub.1 is 2.54 mm and t.sub.2 is 2.00 mm. The fuel assembly 1 and the fuel assembly 41A used for the core 43B will be compared. The width W.sub.8 of the channel box 6 is substantially equal to the width W.sub.8 ' of the channel box 6B. They are substantially equal within the range of tolerance. Accordingly, the width W.sub.4 of the channel box 6 is greater than the width W.sub.4 ' of the channel box 6B. The width W.sub.9 of the fuel spacer 5 is equal to the width W.sub.9 ' of the fuel spacer 5B. In the core 43B, the water gap width W.sub.6 is equal to the water gap width W.sub.7. These water gap widths are greater than the water gap widths W.sub.6 and W.sub.7 of the core 43. The core 43B has originally a high moderation effect of the neutron in the water gap. Therefore, this embodiment utilizes the decrement of the thickness obtained by reducing the thickness of the channel box 6 for increasing the area of the coolant passage inside the channel box 6. Accordingly, the coolant passage area inside the fuel assembly 1 is greater than that in the fuel assembly 41A. The pressure loss of the fuel assembly 1 becomes smaller than that of the fuel assembly 41A and thermal hydraulic stability inside the channel can be improved. The critical power of the fuel assembly 1 where G.sub.2 /G.sub.1 becomes great can be made higher than that of the fuel assembly 41A. Another embodiment of the core of the boiling water reactor using the fuel assembly 1 will be explained with reference to FIG. 16. The core 43C of this embodiment can be obtained by withdrawing the fuel assembly 41 as the fuel assembly C.sub.3 and loading the fuel assembly 1 having burnup of 0 GWd/t in the same way as when the core 43A (FIG. 12) is obtained from the core 43 (FIG. 11). In the core 43C, the fuel assemblies C.sub.2 and C.sub.3 are the fuel assemblies 41 equipped with the channel box 6A and the fuel assembly C.sub.1 is the fuel assembly 1. The cross-section of each fuel assembly 1, 41 has the structure shown in FIG. 13. The difference between the core 43C in which all the fuel assemblies are the fuel assemblies 41 and the core 43 shown in FIG. 11 is that the water gap width W.sub.6 of the core of the former is greater than the water gap W.sub.7. In the fuel assembly 1 used in this embodiment, the center O (FIG. 5) of the insertion portion 3A (FIG. 2) is deviated by .sqroot.2(t.sub.2 -t.sub.1) from the axis of the fuel supporting portion 3B (FIG. 2) on the diagonal line of the lower tie plate 3 extending to the control rods 42 in this cell. The axis of the fuel supporting portion 3B is in agreement with the axis of the fuel assembly 1. Insertion of the insertion portion 3A of such a fuel assembly 1 into a fuel supporting fixture (not shown) makes greater the water gap width W.sub.6 between the fuel assembly 1 and the fuel assembly 41A than the width W.sub.6 between the fuel assemblies 41A. However, the water gap width W.sub.7 between the fuel assembly 1 and the fuel assembly 41A is equal to the width W.sub.7 between the fuel assemblies 41A. The neutron utilization ratio of the core 43C can be improved because the neutron absorption quantity of the channel box is reduced. Moreover, since the narrow gap width W.sub.6 can be enlarged by loading of the fuel assembly 1 into the core 43C, reactivity of the fuel assembly on the narrow water gap side can be improved, so that the difference of reactivity can be reduced between the portion of the fuel assembly facing the wide water gap and the portion facing the narrow water gap. The fuel assemblies 41 and 41A that have been loaded conventionally into different types of cores of the boiling water reactor have different fuel bundles and different sizes of channel boxes. However, when the fuel assembly 1 shown in FIG. 1 is used, the fuel bundle can be used in common for the cores 43A-43C. This can be accomplished by the employment of the channel box having the support tabs 18. In the fuel assembly 1 that is loaded so as th correspond to the cores 43A, 43b and 43C, the size of the fuel bundle or in other words, the width of the upper tie plate 2, the outer diameter and pitch of the fuel rods 4 and the width of the fuel spacer 5 are substantially the same. Therefore, it is no longer necessary to produce separately the fuel bundles for the cores 43A-43c, respectively, and production and management of the fuel bundle become extremely easy. It is also possible to load the fuel assembly 1A shown in FIG. 9 in place of the fuel assembly 1 and to constitute the core 43A-43C. FIG. 19 shows the fuel assembly 1B in accordance with still another embodinent of the present invention. The fuel assembly 1B has the same structure as the fuel assembly 1 shown in FIG. 1 except for the structure of the lower end portion of the channel box and sidewalls of the lower tie plate. In the channel box 6B.sub.1, the reinforcing tabs 17 are disposed at the portion ranging from the upper surface of the lower tie plate 3C to the second fuel spacer. Each sidewall is straight from below the upper surface of the lower tie plate 3C in the axial direction in the channel box 6B.sub.1. The lower tie plate 3C does not have the grooves 22 on the outer side surface of its sidewalls. The fuel assembly 1B provides the same effect as that of the fuel assembly 1 shown in FIG. 1. when the fuel rods 4 gets elongated in the axial direction and the upper tie plate 2 is lifted up during the operation of the reactor, the lower end of the channel box 6B.sub.1 moves up, as well. In this case, the lowermost reinforcing tabs 17 leave the upper surface of the lower tie plate 3C. However, outward creep deformation of the straight portion of the channel box 6B.sub.1 that is positioned below the lowermost reinforcing tabs 17 is by far smaller than in the prior art due to the functions of the reinforcing tabs 17. This creep deformation is greater than that of the channel box 6 shown is FIG. 1. FIG. 20 shows the fuel assembly 1C in accordance with another embodiment of the present invention. The fuel assembly 1C is equipped with the channel box The channel box 6C has the tabs 17, 18 and 19 in the same way as the channel box 6. The reinforcing tabs 17 are formed by projecting outward the sidewalls of the channel box. In order not to impede insertion of the control rods, the width W.sub.11 between the inner surfaces of the opposed reinforcing tubs 17 is equal to the width W.sub.4 of the channel box 6. Therefore, the width W.sub.10 between the opposed inner surfaces of the channel box 6C is smaller than the width W.sub.4 but is equal to the width W.sub.5 (FIG. 2). Under this relationship the width L.sub.2 between the outer side surfaces of the opposed sidewalls of the lower tie plate 3D is smaller than the width L.sub.1 of the lower tie plate 3C described above. The height e.sub.1 of the tabs 18, 19 from the inner surface of the channel box 6C in this embodiment is smaller than the height e of the channel box 6. This embodiment provides the same effect as that of the fuel assembly 1 shown in FIG. 1. However, in this embodiment the gap between the fuel rods 4 positioned at the outermost periphery and the inner surface of the channel box 6C is small throughout the full length in the axial direction. Therefore, the pressure loss becomes greater than in the fuel assembly 1 and channel stability becomes lower than in the prior art fuel assembly. When these fuel assemblies 1C are positioned into the core, however, the water gap width between the adjacent fuel assemblies becomes greater at the upper part of the fuel assemblies. As a result, core stability of the core using the fuel assemblies 1C can be improved. The cores 43A-43C can be constituted by using the fuel assemblies 1B or 1C in place of the fuel assemblies 1. Other examples of the channel boxes 6 to be fitted to the fuel assembly 1 are shown in FIGS. 21-25. These channel boxes can be applied to each of the fuel assemblies 1A-1C The channel box 6D shown in FIG. 21 has the spacer support tabs 18B formed by connecting each tab 18 positioned above the spacer support tubs 18 for supporting the second fuel spacer in the channel box 6. The spacer support tab 18 improves rigidity of the channel box 6D to the bending stress in the axial direction. The channel box exhibits the same function as the channel box 6. The channel box 6E shown in FIG. 22 has the spacer support tabs 18c formed by extending the spacer support tabs 18B described above from the upper end to lower end of the channel box. Besides the function obtained by the channel box 6, this channel box 6E has the function of improving the strength to the bending stress in the axial direction and buckling. In the channel box 6F shown in FIG. 23, the spacer support tabs 18D for supporting the first and second fuel spacers are transversely thin in the same way as the reinforcing tabs 17. The spacer support tabs 18 for supporting the fuel spaces at a higher level than the second fuel spacer are thin in the axial direction. The resistance of the channel box 6F to creep deformation is increased by the spacer support tabs 18D. However, cooling water hardly flows between the fuel spacer 5 and the channel box 6F at the level of disposition of the spacer support tabs 18D. The pressure loss of the fuel assembly equipped with this channel box 6F becomes somewhat greater than that of the fuel assembly equipped with the channel box 6. The pressure loss at the lower part of the fuel assembly (particularly at the portion below the second fuel spacer) is smaller than that at the gas-liquid two phase portion at the upper part of the fuel assembly. Accordingly, when the pressure loss of the fuel assembly is considered as a whole, disposition of the spacer support tabs 18D does not provide extremely large fluid resistance. The channel box 6G shown in FIG. 24 has groove-like reinforcing tabs 17A that continue throughout the entire periphery. These reinforcing tabs 17A improves remarkably the inhibition effect of outward creep deformation of the channel box 6G. The channel box 6G has the function of the channel box 6, too. The channel box 6H shown in FIG. 25 is produced by disposing the tabs 17, 18 and 19 to the channel box having the thick corner portions 46 of Japanese Patent Publication No. 13075/1989. This channel box 6H has the functions of both of the channel box 6 and channel box of Japanese Patent Publication No. 13075/1989. The strength of the channel box 6H at the reduced thickness portion at the center of its sidewalls can be increased remarkably by the reinforcing tabs 17 disposed at that portion, and the thickness of the reduced thickness portion can be further reduced. Since the thickness of the corner portions is thick in this channel box 6H, however, the neutron absorption quantity is greater as much than is the channel box 6 shown in FIG. 1. In the channel box 6H, the width W.sub.8 (FIG. 13) of the channel box 1 is the gap W.sub.12 between the outer side surfaces of the thick corner portions 6H. The reinforcing tabs 17 disposed on the channel box may be replaced by the solid reinforcing tabs 17B as shown in FIG. 26. The shape of the reinforcing tabs 17B defines the corrugated portion on the sidewalls of the channel box. The channel box having the reinforcing tabs 17B has the same function as that of the channel box 6. However, the channel box of the former has a greater neutron absorption quantity than that of the latter. Since the reinforcing tabs 17B exist below the second fuel spacer, the neutron absorption quantity at the lower part of the fuel assembly increases and the power distribution of the fuel assembly in the axial direction can be made flat. The support tabs 18 and 19 may be made solid in the same way as the reinforcing tabs 17B. However, the same problem with the reinforcing tabs 17B occurs, as well. FIG. 27 shows the fuel assembly 1D in accordance with still another embodiment of the present invention. The fuel assembly ID has the channel boxes 6I and the fuel spacer 5C. The structure of this fuel assembly other than these two structures is the same as that of the corresponding portion of the fuel assembly 1. The channel box 6I is equipped with the reinforcing tabs 17 and the support tabs 19 but is not equipped with the support tabs 18. The fuel spacer 5C is formed by disposing support springs 47 to the band 15 of the fuel spacer 5. The support spring 47 comes into contact with the inner surface of the channel box 6I and limits the movement of the fuel spacer 5C in the transverse direction. Whenever the reinforcing tabs 17 pass through the fuel spacer 5C at the time of fitting of the channel box 6I, the support spring 47 is pushed to the band of the fuel spacer 5C. The support spring 47 does not raise any problem at the time of fitting of the channel box 6I and the channel box 6I can be fitted easily to the fuel bundle. The fuel assembly 1D provides the same effect as the fuel assembly 1. However, since the fuel assembly 1D is equipped with the support spring 47, the oscillation of the fuel spacer in the transverse direction becomes somewhat greater than in the fuel assembly 1 at the time of earthquake. Other examples of the fuel assembly are shown in FIGS. 28 and 29. These drawings show the structure of the fuel assembly 1E of this embodiment near the lower tie plate. The rest of the structure are the same as those of the fuel Assembly 1A. The fuel assembly 1E is produced by adding the function of the jet stream described in Japanese Laid-Open No. 212391/1989 to the fuel assembly 1A. The lower tie plate 3E includes a fuel support portion 3A, a cylindrical sidewall portion 51 which has inner side surfaces 53 for defining four outer side surfaces 52 as the side surfaces of the lower tie plate 3E and inner space 54, has a square cross-section and continues the fuel support portion 3A, and a nozzle portion 55 which continues the cylindrical sidewall portion 51 and guides the coolant into the inner space 54. The grooves 22 shown on FIG. 1 are disposed on the outer side surfaces 52 side of the cylindrical sidewall portion 51. Several reinforcing tabs 17 are inserted into the grooves 22. As shown in FIG. 29. the fuel supports portion 3A includes fuel rod fitting holes 56 into which the lower end of the fuel rods 4 is fitted, connecting fuel rod fitting holes 57 and water rod fitting holes 58 into which the lower end of each water rod 12A is fitted. Furthermore, the fuel support portion 3A includes coolant supply holes 60, 61, 62, 63, 64 for guiding cooling water supplied into the inner space 54 above the fuel support portion 3A or in other words, for guiding cooling water into the coolant passage 59 inside the channel box 6, which are disposed between the fitting holes 56, 57, 58. These cooling water supply holes have mutually different cross-sectional areas. The cooling water supply hole 60 at the outermost periphery among these cooling water supply holes forms flow path means for generating the flow of a coolant that inhibits leakage of cooling water from the gap (coolant passage 65) defined between the channel box 6 and the lower tie plate 3E when the fuel assembly 1E is loaded into the core. Hereinafter, the coolant flow that flows out from the cooling water supply hole 60 will be referred to as the "jet flow". The cooling water supply hole 60 has an inside portion 60a which penetrates through the portion of the fuel support portion 3A which is positioned more inward than the inner side surface 53 of the cylindrical sidewall portion 51 and an outside portion 60b which is positioned more outward than the inner side surface 53, comes into the cylindrical sidewall portion 51 and is open to the inner side surface 53. Accordingly, the inner side surface 53 can be seen through the cooling water supply hole 60 as shown in FIG. 29. The cooling water supply hole 60 is constituted in the following way. In FIG. 28, the distance t.sub.3 between the point P which is positioned on the side of the outer side surface 52 of the lower tie plate 3E on the inner peripheral surface of the outlet of the cooling water supply hole 60 (the portion of the inner peripheral surface which is the nearest to the outer side surface 52) and the bottom surface 22A of the groove 22 (the outer side surface 52 of the lower tie plate when the groove 22 does no exist such as the lower tie plate 3C shown in FIG. 19 and the lower tie plate 3D shown in FIG. 20) is smaller than the thickness t.sub.4 of the cylindrical side wall portion 51 of the lower tie plate 3E (the dimension between the bottom surface 22A and the inner side surface 53 when the groove 22 is disposed in the lower tie plate and the dimension between the outer side surface 52 and the inner side surface 53 when the groove 22 does not exist in the lower tie plate). In other words, the point P is positioned closer to the outer side surface 52 than the inner side surface 53 of the cylindrical sidewall portion 51. It can be said that the distance L.sub.3 between the point P of the outlet of the cooling water supply hole 60 and the center axis of the lower tie plate 3E is greater than the distance L.sub.4 between the inner side surface 53 of the lower tie Plate 3E and the center axis of the lower tie plate 3E. Furthermore, it can be said that part of the cooling water supply hole 60 is positioned more outward (closer to the outer side surface 52) than the fuel rod insertion holes (the fitting holes 56 and 57) into which the lower end of each fuel rod 4 positioned at the outermost periphery among the fuel rods 4 is inserted. The cooling water supply holes 60 may be positioned as a whole more outward than the fuel rod insertion holes described above. In order to improve further the leak inhibition function of cooling water and the seal effect by the jet flow, it is very important to increase maximum the total flow passage area of the cooling water supply holes 61 existing at the outer peripheral portion and to contract the total flow passage sectional area of the inner supply holes 62, 63, 64. (For detail, refer to the description on page 20, lines 11-23 of the specification of U.S. patent application Ser. No. 464,151 (filed on Jan. 12, 1990) and page 18, line 9-page 21, line 23 of the specification of European Patent Application No. 90300272.3)). During the operation of the boiling water reactor, cooling water is guided into the channel box 6 of the fuel assembly 1E through the cooling water supply holes 60-64. Most of cooling water rises in the coolant passage 59 inside the channel box 6. Part of cooling water leaks outside the fuel assembly 1E through the coolant passage 65. The leaking quantity of this cooling water is limited by the jet flow which is jetted from the outer peripheral portion of the lower tie plate 3E to the inner surface of the channel box 6. The reason will be explained next. Since the cooling water supply holes 60 are disposed, the flow rate of cooling water increases near the channel box 6 much more in this embodiment than in the conventional apparatus. In this embodiment having such a flow rate distribution of cooling water, a low pressure region is formed by the action of the jet flow near the inner surface of the channel box 6 below the jet flow flowing out from the cooling water supply holes 60 of the outermost periphery, that is, near the inlet portion of the coolant passage 65. Accordingly, leak of cooling water inside the channel box 6 to the outside through the coolant passage 65 can be inhibited. The effect of the jet flow can be made further effective by bringing the outlet of the cooling water supply holes 60 of the outermost periphery close to the bottom surface 22A of the groove 22 of the lower tie plate 3E (or to the outer side surface 52) and point P is close to the outer side surface 52) and distributing each flow path sectional area of the cooling water supply holes 61-64 so that the total flow path sectional area of the outer peripheral portion is greater than the total flow path sectional area at the center. The effect of inhibiting the leakage of cooling water or the seal effect by the jet flow can be further improved by increasing the flow path sectional area of the supply holes at the outer peripheral portion of the lower tie plate while decreasing the flow path sectional area of the supply holes at the center. Furthermore, such a definition of the flow path sectional area of the supply holes provides the effect of making flat the flow velocity distribution inside the fuel assembly 1E after leaving the lower tie plate 3E much more than in the prior art example as shown in FIG. 30. In this embodiment, in particular, the width of the portion positioned outside the fuel rod insertion holes 56 at the outermost periphery (the outside portion 60b) in the direction in parallel with the bottom surface 22A of the groove 22 (or with the outer side surface 52) is greater than the width of the portion (the inside portion 60a) positioned in the region between the fuel rod insertion holes positioned at the outermost periphery in the same direction. Therefore, the jet flow can be supplied substantially uniformly throughout the entire periphery of the inner surface of the channel box 6 except for its corner portions. This is effective for reducing the flow rate of leaking cooling water. This embodiment can limit the flow rate of leaking cooling water from the coolant passages 65 by the jet flow and can obtain the same effect as that of the fuel assembly 1. Particularly because the channel box having the reinforcing tabs 17 at its lower end portion and the jet flow are used in combination, the flow rate of leaking cooling water can be limited extremely. Accordingly, the finger springs that have been disposed conventionally in the coolant passage 65 for limiting the flow rate of leaking cooling water become unnecessary. The fuel assembly 1E can be loaded into any of the cores 43A-43C in place of the fuel assembly 1. The cooling water supply hole 60 of the fuel assembly 1E can be applied to the lower tie plate of each of the fuel assemblies 1B, 1C and 1D. The present invention can increase remarkably the strength of the channel box in comparison with the prior art and can remarkably inhibit creep deformation of the channel box. In other words, the present invention can prolong the residence time of the channel box inside the core as much. The present invention makes it possible to use in common the structures of the fuel assemblies to be loaded into different types of cores other than the channel box for different types of cores.
	summary
	abstract	A debris filter bottom nozzle for a nuclear fuel assembly having a support skirt and a top plate. The top nozzle plate has defined therethrough a plurality of flow holes. The axial bore through each flow hole is contoured as a venture. Preferably, the venturi is formed with a double chamfered inlet.
	summary
	claims	1. A nondestructive inspection system comprising:a plate that holds a sample;a plurality of radiation detectors positioned in predetermined directions relative to the sample held by the plate;a photon beam irradiation part that irradiates the sample held by the plate with a plurality of types of quasi-monochromatic photon beams having differing energies and whose planes of polarization are each controlled, the quasi-monochromatic photon beams being made to be coaxial;a drive part that relatively moves the plate and the photon beam irradiation part;a control processing part that controls the drive part and to which detection signals from the plurality of radiation detectors are inputted; anda display part;wherein the plurality of radiation detectors are positioned in directions that allow for detection of nuclear resonance fluorescence (NRF) gamma rays emitted from an isotope within the sample in emission directions that are dependent on the planes of polarization of the plurality of types of quasi-monochromatic photon beams with which the sample is irradiated,wherein the control processing part identifies an isotope of interest that is present in a quasi-monochromatic photon beam irradiated region in the sample based on the detection signals of the plurality of radiation detectors, andwherein a spatial distribution of the isotope is visualized and displayed on the display part. 2. The nondestructive inspection system according to claim 1, wherein the photon beam irradiation part comprises:an electron beam accelerator that generates an electron beam of a predetermined energy;a first laser light source that generates laser light of a first wavelength and having a first polarization;a second laser light source that generates laser light of a second wavelength and having a second polarization;an optical system that causes the laser light of the first wavelength and the laser light of the second wavelength to collide, as coaxial light, with the electron beam at a first angle; anda collimator that transmits a first polarized photon beam generated in a second angular direction relative to the electron beam due to the collision between the electron beam and the laser light of the first wavelength, and a second polarized photon beam generated in the second angular direction due to the collision between the electron beam and the laser light of the second wavelength. 3. The nondestructive inspection system according to claim 2, wherein the control processing part controls the first laser light source and the second laser light source to generate pulsed light, each with a different time structure, to be generated as the laser light of the first wavelength and the laser light of the second wavelength. 4. The nondestructive inspection system according to claim 1, wherein the plurality of types of quasi-monochromatic photon beams have their energies and planes of polarization defined in such a manner that a plurality of NRF gamma rays are emitted in different directions from a single or a plurality of nuclear levels of a single or a plurality of isotopes in the sample due to irradiation by the photon beams. 5. The nondestructive inspection system according to claim 1, wherein the plurality of radiation detectors are positioned, relative to the sample, in directions parallel to and/or directions perpendicular to the planes of polarization of the plurality of types of quasi-monochromatic photon beams with which the sample is irradiated, or at arbitrary angles. 6. The nondestructive inspection system according to claim 1, further comprising a photon intensity monitor that detects an intensity of the plurality of types of quasi-monochromatic photon beams with which the sample is irradiated. 7. The nondestructive inspection system according to claim 1, wherein the isotope of interest comprises an isotope of an element that forms an explosive, including carbon-12, nitrogen-14, and oxygen-16. 8. The nondestructive inspection system according to claim 1, wherein the energies of the plurality of types of quasi-monochromatic photon beams emitted from the photon beam irradiation part are equal to or less than a neutron emission energy. 9. The nondestructive inspection system according to claim 1, wherein the drive part drives the plate in such a manner that the sample held by the plate is scanned by the plurality of types of quasi-monochromatic photon beams emitted from the photon beam irradiation part. 10. The nondestructive inspection system according to claim 1, wherein the plurality of types of quasi-monochromatic photon beams are laser-Compton scattering (LCS) photon beams generated using a polarization of laser light, and the planes of polarization thereof are selected in accordance with a constituent element of an object to be inspected or with the type of isotope to be inspected. 11. The nondestructive inspection system according to claim 1, wherein the planes of polarization and energies of the plurality of types of quasi-monochromatic photon beams are controlled in such a manner that the NRF gamma rays are emitted in desired directions in accordance with a transition model that is determined by the nuclear structure of an isotope in the sample. 12. The nondestructive inspection system according to claim 1, wherein the plurality of radiation detectors positioned in pre-defined directions to detect the NRF gamma rays utilizing anisotropy of the emission directions of the NRF gamma rays. 13. The nondestructive inspection system according to claim 1, wherein the count rate per radiation detector is lowered by dispersing the emission directions of the NRF gamma rays emitted from a nuclear level of an isotope in the sample in a manner dependent on the planes of polarization of the plurality of types of quasi-monochromatic photon beams with which the sample is irradiated. 14. The nondestructive inspection system according to claim 1, wherein the plurality of radiation detectors positioned in different directions relative to the sample detects the plurality of NRF gamma rays emitted from nuclear levels of the isotopes with close energies. 15. The nondestructive inspection system according to claim 1, wherein the plurality of radiation detectors comprise scintillator detectors. 16. The nondestructive inspection system according to claim 1, wherein the plurality of radiation detectors are structured to detect the NRF gamma rays of or below a neutron generation threshold. 17. The nondestructive inspection system according to claim 1, wherein the photon beam irradiation part comprises:an electron beam accelerator that generates an electron beam of a predetermined energy;a plurality of laser light sources; anda collimator extracting a plurality of polarized LCS photon beams generated, due to the collision, in a second angular direction, whereintwo or more types of polarized laser light with differing wavelengths generated from the plurality of laser light sources are made to collide, as coaxial light, with the electron beam at a first angle. 18. The nondestructive inspection system according to claim 17, wherein the plurality of laser light sources are configured to provide adjustment of the energies of the polarized LCS photon beams. 19. The nondestructive inspection system according to claim 17, wherein the collimator is configured to be adjustable in position, thus enabling position adjustment of the collimator before starting measurement.
	abstract	An improved HE LINAC-based ion implantation system is disclosed utilizing direct digital synthesis (DDS) techniques to obtain precise frequency and phase control and automated electrode voltage phase calibration. The DDS controller may be used on a multi-stage linear accelerator based implanter to digitally synchronize the frequency and phase of the electric fields to each electrode within each stage of the accelerator. The DDS controller includes digital phase synthesis (DPS) circuits for modulating the phase of the electric field to the electrodes, and a master oscillator that uses digital frequency synthesis or DFS to digitally synthesize a master frequency and phase applied to each of the DPS circuits. Also disclosed are methods for automatically phase and amplitude calibrating the RF electrode voltages of the stages.
051223333	abstract	The apparatus comprises a conduit connected to a containment vessel to supply air to a water bath contained in a basin. The conduit is connected to a plurality of air nozzles in the water bath, each of which forms a unit together with a perforate baffle plate disposed above the nozzle to intensively mix the emerging air with water. A set of static mixer elements enclosed by a jacket is disposed in the water bath above the baffle plate.
	abstract	An assembly is provided that includes a device for locking tubes in position relative to one another, with at least a first arm, a locking axle having a plurality of bearing surfaces, the locking axle being movable between a position locking the tubes each between one of said bearing surfaces and the first arm, and a released position, in which the tube segments are free.
048204757	abstract	A burnable absorber cluster assembly includes a support plate, a plurality of burnable absorber rods, and an improved attachment joint for detachably connecting an upper end of each of the absorber rods to the support plate. The attachment joint includes a plug insert defined on the end of each of the rods, a hollow annular wall extending outwardly from each plug insert, holes defined through the support plate for receiving the plug inserts, and recesses formed in the support plate and connected with the holes therein. Each recess is conical-shaped and extends in flared fashion outwardly from one hole to a maximum diameter greater than the diameter of the hole. The annular wall on each plug insert is deformed into conformity with the shape of one of the recesses for providing an interengaging connection between the insert plug and the support plate in which the plug insert is disposed within the support plate hole. The annular wall is also redeformable upon application of a predetermined axially-directed force on the plug insert so as to remove the plug insert from the support plate hole and thereby detach the rod from the support plate. Also, each recess and annular wall have respective interengaging eccentric portions which prevents the plug insert on each rod end from rotating in its corresponding hole.
	abstract	A halide scintillator material is disclosed. The material is single-crystalline and has a composition of the formula A3MBr6(1-x)Cl6x (such as Cs3CeBr6(1-x)Cl6x) or AM2Br7(1-x)Cl7x (such as CsCe2Br7(1-x)Cl7x), 0≦x≦1, wherein A consists essentially of Li, Na K, Rb, Cs or any combination thereof, and M consists essentially of Ce, Sc, Y, La, Lu, Gd, Pr, Tb, Yb, Nd or any combination thereof. Furthermore, a method of making halide scintillator materials of the above-mentioned compositions is disclosed. In one example, high-purity starting halides (such as CsBr, CeBr3, CsCl and CeCl3) are mixed and melted to synthesize a compound of the desired composition of the scintillator material. A single crystal of the scintillator material is then grown from the synthesized compound by the Bridgman method. The disclosed scintillator materials are suitable for making scintillation detectors used in applications such as medical imaging and homeland security.
	description	The present invention is directed to a system and method for harvesting of activated irradiation targets from a nuclear reactor, and a radionuclide generation system configured to produce radionuclides from irradiation targets inserted into an instrumentation finger in a commercial nuclear reactor core. Radionuclides are used in various fields of technology and science, as well as for medical purposes. Usually, radionuclides are produced in research reactors or cyclotrons. Since the number of facilities for commercial production of radionuclides is limited and expected to decrease, it is desired to provide alternative production sites. EP 2 093 773 A2 suggests that existing instrumentation tube systems conventionally used for housing neutron detectors may be used to generate radionuclides during normal operation of a commercial nuclear reactor. In particular, spherical irradiation targets are linearly pushed into and removed from instrumentation fingers extending into the reactor core. Based on the axial neutron flux profile of the reactor core, the optimum position and exposure time of the targets in the reactor core are determined. A driving gear system is used for moving and holding the irradiation targets in the instrumentation tube system. US 2013/0177118 A1 discloses a system that allows irradiation targets to be irradiated in a nuclear reactor and deposited in a harvestable configuration without direct human interaction or discontinuation of power-producing activities. The system includes accessible end-points that store desired produced isotopes for handling and/or shipping. A cask tube is configured for use with multiple storage casks based on target properties. One or more stops may be inserted into the cask tube at desired positions to separate out a certain population of irradiation targets. A specific embodiment shows two stops provided in a cask tube at a predetermined distance that corresponds to a length of leader spheres. The leader spheres may then be emptied out of the cask tube via gravity or pneumatic force into a separate cask or other facility. US 2013/0170927 A1 discloses apparatuses and methods for producing radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets are inserted and removed from instrumentation tubes during operation and converted to radioisotopes. A harvesting station is provided outside of an access-restricted area to permit access for refilling or harvesting the targets during plant operation. A harvesting cask and/or cask tube may be equipped with a target counter or activity detector that counts or measures properties of irradiation targets moving into the harvesting cask. US 2013/0315361 A1 relates to apparatuses and methods for producing radioisotopes in multiple instrumentation tubes of operating commercial nuclear reactors. Irradiation targets are inserted and removed from multiple instrumentation tubes and converted to radioisotopes during operation of the nuclear reactor. Positioning irradiation targets are used to properly position other irradiation targets at appropriate positions within or near the nuclear core. The positioning targets may be sorted out from the harvesting cask due to their markings or physical properties, or other discriminators may selectively divert the positioning targets to alternate termination points. Conventional aero-ball measuring systems are known in the art and disclosed, for example, in GB 1 324 380 A and U.S. Pat. No. 3,263,081 A. Due to the high activity of the activated irradiation targets retrieved from the instrumentation tube system, and since space within the reactor containment is limited, the activated targets are difficult to handle. In particular, the activated targets including the radionuclides must be filled into and stored in containers provided with heavy radiation shielding. The chambers for the Traversing Incore Probe (TIP) system and/or aero-ball measuring system do not have any structures for packaging and transporting those heavy containers. Provision of additional water locks in the reactor containment for handling of the activated targets and shielded containers would also be too expensive. It is an object of the invention to provide a harvesting system for activated irradiation targets from an instrumentation tube system of a commercial nuclear reactor, such as the instrumentation finger of an aero-ball measuring system or a Traversing Incore Probe (TIP) system extending into the core of a nuclear reactor, which allows for an easy handling of shielded storage containers and harvesting of the activated targets during reactor operation, and which can be installed subsequently into existing structures within accessible areas of the reactor containment. The above objects are solved by a radionuclide harvesting system according to claim 1. Advantageous and expedient embodiments of the invention are indicated in the dependent claims which can be combined with each other independently. According to a first aspect, the invention provides an irradiation target harvesting system comprising at least one storage container for receiving activated irradiation targets from an instrumentation tube system of a nuclear reactor; a discharge tube having an exit port configured to be coupled to the storage container; and a lock element provided in the discharge tube for blocking movement of the activated irradiation targets into the storage container; wherein the discharge tube comprises a first discharge tube section, a second discharge tube section and an apex formed at a conjunction of the first and second discharge tube section, wherein the first and second discharge tube sections are directed downwardly from the apex, wherein the exit port is arranged at an end of the first discharge tube section and wherein the second discharge tube section is coupled to the instrumentation tube system. According to a further aspect, the invention provides a radionuclide generation system comprising: an instrumentation tube system of a nuclear reactor including at least one instrumentation finger extending into a core of the nuclear reactor wherein the instrumentation tube system is configured to permit insertion and removal of irradiation targets into the instrumentation finger; a target drive system configured to insert the irradiation targets into the instrumentation finger in a predetermined linear order and to remove the irradiation targets from the instrumentation finger; a core monitoring system and an instrumentation and control unit linked to each other and configured to calculate an optimum axial irradiation position and time for the irradiation targets based on an actual state of the nuclear reactor as provided by the core monitoring system; and a target harvesting system comprising at least one storage container for receiving the activated irradiation targets from the instrumentation tube system of the nuclear reactor, a discharge tube having an exit port configured to be coupled to the storage container; and a lock element provided in the discharge tube for blocking movement of the activated irradiation targets into the storage container; wherein the discharge tube comprises a first discharge tube section, a second discharge tube section and an apex formed at a conjunction of the first and second discharge tube section, wherein the first and second discharge tube section are directed downwardly from the apex, wherein the exit port is arranged at an end of the first discharge tube section and wherein the second discharge tube section is coupled to the instrumentation tube system. In another aspect, the invention is a method for harvesting activated irradiation targets from an instrumentation tube system of a nuclear reactor, wherein the method comprises the steps of: Coupling the instrumentation tube system to a discharge tube having an apex, an exit port and a lock element between the apex and the exit port; Passing the activated irradiation targets from the instrumentation tube system into the discharge tube and blocking movement of the activated irradiation target out of the exit tube by means of the lock element; Separating a predefined quantity of the activated irradiated targets from another quantity of the activated irradiated targets in the discharge tube; Coupling the exit port to a storage container and releasing the lock element to pass the predefined quantity of the activated irradiated targets under action of gravity into the storage container; wherein said separating step comprises passing the predefined quantity of the activated irradiation targets over the apex and keeping the other quantity of activated irradiation targets in the discharge tube or the instrumentation tube system by means of the apex. According to the invention, the activated irradiation targets including the radionuclides are removed from the instrumentation finger and passed into the discharge tube, wherein a predefined quantity of the irradiated targets is separated from the remaining irradiated targets in the discharge tube, and the predefined quantity of the irradiated targets is released from the discharge tube into a shielded storage container under action of gravity. Since only a predefined quantity of the activated targets having a predetermined lower activity is harvested from the system and stored in the shielded container, much smaller containers can be used which are easy to be transported either manually or by means of existing handling structures within the reactor containment. The discharge tube of the target harvesting system is configured to receive the activated irradiation targets from the instrumentation tube system of the nuclear reactor. The discharge tube comprises an apex dividing the discharge tube into a first section proximate an exit port and a second section coupled to the instrumentation tube system. A lock element is provided in the first discharge tube section near the exit port to block movement of the activated irradiation targets out of the discharge tube. Activated irradiation targets passing from the instrumentation tube system into the discharge tube are held in the tube by the lock element. One of a number of shielded storage containers is coupled to the exit port. When the lock element is opened, the predefined quantity of the activated targets leaves the discharge tube driven by gravity, and is collected in the shielded storage container. The quantity of activated targets leaving the exit tube is determined by the length of the target column in the first discharge tube section. Thereafter, the lock element can be closed and the activated targets remaining in the discharge tube can be driven back into the instrumentation tube system using pressurized air or any other suitable gas such as nitrogen. The shielded storage container can be closed and transported out of the reactor containment. The invention provides a simple and cost effective system for portioning and harvesting of activated irradiation targets. The targets can be released from the discharge tube under the action of gravity using a simple lock element, rather than using pressurized air or gas. Thus, generating of aerosols is minimized during release of the targets. Portioning of the targets is possible by defining the length of the first discharge tube section between the lock element and the apex which corresponds to the length of the target column and thus the amount of the targets released from the first discharge tube section. No sorters or target counters are necessary. If desired, the activated targets retained in the second discharge tube section can be driven back into a holding position in the instrumentation tube system, or even back into the instrumentation finger. This will further minimize an operator's exposure to radioactivity. According to a preferred embodiment of the irradiation target harvesting system, the exit port comprises a valve element for pressure-tight sealing the discharge tube. Sealing the exit port facilitates a safe removal of the storage container filled with activated irradiation targets. No aerosols are released into the containment during target transport inside the instrumentation tubes. Preferably, the exit port is located within an accessible area of the reactor containment. Preferably, the discharge tube comprises a T-junction located between the lock element and the exit port, wherein the T-junction is configured for supplying and discharging pressurized air into and out of the discharge tube. Providing the T-junction for air supply is an easy cost effective measure to control movement of the activated irradiation targets in the discharge tube. Still preferably, the T-junction comprises a three-way valve. More preferably, the first discharge tube section, the second discharge tube section and the apex are shaped in the form of an inverse U. The pre-defined quantity of the activated irradiation targets to be separated from the other targets is then defined by the length of a leg of the inverse U, between the lock element and the apex, corresponding to the length of the target column in this leg. The inverse U-shape provides for a reliable portioning of the activated irradiation targets. In addition, release of the target column from the leg of the inverse U is possible under the action of gravity by simply opening the lock element. Thus damage to the activated irradiation targets can be avoided. According to a further embodiment, the lock element is at a first level, wherein the second discharge tube section has a base point opposite to the apex and the base point is at a second level, wherein the first level is higher than the second level. In other words, a height of the target column in the first discharge tube section corresponding to a distance between the lock element and the apex is lower than a height of the target column in the second discharge tube section or distance between the base point of the second discharge tube section and the apex. Due to the greater height of the target column in the second discharge tube section, the apex reliably retains the other quantity of the targets in the second discharge tube section or instrumentation tube system without any further mechanical means. In a further preferred embodiment, the lock element is a magnetically or mechanically operated restriction element, preferably a pin. These elements are available as standard components at low costs. The irradiation target harvesting system may further comprise one or more magnets arranged at the first discharge tube section between the apex and the lock element. Preferably, the magnets can be used to select and separate activated irradiation targets from dummy targets in the discharge tube. The dummy targets are made from an inert material and may be used to achieve a proper axial positioning of the irradiation targets in the nuclear reactor core. More preferably, one of the irradiation targets and the dummy targets is ferromagnetic whereas the other one of the irradiation targets and dummy targets is non-magnetic or paramagnetic. Preferably, the one or more magnets are selected from a permanent magnet and a solenoid. Still more preferably, the one or more magnets are movable along the first discharge tube section so as to selectively expose ferromagnetic targets to a magnetic field and retain the ferromagnetic targets in the discharge tube. The method for harvesting activated irradiation targets from an instrumentation tube system of a nuclear reactor preferably comprises the step of transferring the other quantity of the activated irradiation targets from the second discharge tube section into a holding position in the instrumentation tube system or back into the instrumentation finger prior to removing the storage container from the exit port. The holding position may be remote from the discharge tube in a shielded or access-restricted area of the reactor. This embodiment will further minimize any exposure to radiation of an operator. More preferably, the steps of the harvesting method are repeated until all activated irradiation targets are removed from the system, and portions of the activated irradiation targets are stored in a number of storage containers. The invention provides a system and method for harvesting of activated irradiation targets from an instrumentation tube system of a commercial nuclear reactor wherein the total amount of activity transferred into a storage container is reduced by portioning the activated irradiation targets. Therefore, smaller storage containers with less shielding can be used which allows for an easy handling of the containers within the reactor containment. The harvesting system makes use of simple mechanical components and can therefore be installed subsequently into existing reactor structures such as an aero-ball measuring system. In a preferred embodiment, dummy targets may be used for positioning the irradiation targets in the instrumentation finger, and the activated irradiation targets are separated from the dummy targets by means of their different magnetic properties. One or more magnets positioned at the first discharge tube section are used for retaining one of the dummy targets or irradiation targets within the instrumentation tube system and deliver the other one of the dummy targets or the irradiation targets to a storage container or intermediate tank, respectively. Separating the targets by means of different magnetic properties allows for a fast remote processing of the activated irradiation targets and may also avoid the use of additional mechanical systems such as counters, indexers and gateways to determine the exact position of the dummy targets and irradiation targets. The invention contemplates that a commercial nuclear reactor can be used for producing radionuclides. In particular, conventional aero-ball measuring systems or other instrumentation tube systems of the commercial reactor can be modified and/or supplemented to enable an effective and efficient production of radionuclides. Some of the instrumentation tubes for example of a commercial aero-ball measuring system or Traversing Incore Probe (TIP) system are used to guide the irradiation targets into the reactor core and to lead the activated irradiation targets out of the reactor core. The activation of the targets is optimized by positioning the irradiation targets in predetermined areas of the reactor core having a neutron flux sufficient for converting the parent material in the irradiation targets completely into the desired radionuclide. The proper positioning of the irradiation targets may be achieved by means of dummy targets made of an inert material and sequencing the dummy targets and the irradiation targets in the instrumentation tube system so as to form a column of the targets in the instrumentation finger wherein the irradiation targets are at a pre-calculated optimum axial position in the reactor core and the other positions are occupied by the inert dummy targets or remain empty. FIG. 1 illustrates the basic setup of a radionuclide generation system within a commercial nuclear power plant. As opposed to a research reactor, the purpose of a commercial nuclear reactor is the production of electrical power. Commercial nuclear reactors typically have a power rating of 100+ Megawatt electric. The basis of the radionuclide generation system described in the example embodiments is derived from a conventional aero-ball measuring system used to measure the neutron flux density in the core of the nuclear reactor. A plurality of aero-balls are arranged in a linear order thereby forming an aero-ball column. The aero-balls are substantially spherical or round probes but can have other forms such as ellipsoids or cylinders, as long as they are capable of moving through the conduits of the instrumentation tube system. The aero-ball measuring system includes a pneumatically operated drive system configured to insert the aero-balls into an instrumentation finger extending into and passing the core through its entire axial length, and to remove the aero-balls from the instrumentation finger after activation. For easier reference, hereafter the radionuclide generation system based on a commercial aero-ball measuring system will be also referred to as MAS or Medical Aeroball System. The irradiation target harvesting system of the present invention can also be built as a stand-alone system that is fixed to an unmodified nuclear instrumentation tube of a regular aero-ball measuring system. The inventive target harvesting system is then configured to be movable and only temporarily attached to the instrumentation tube system during target discharge. Thus, no permanent modification of the aero-ball measuring system is required. In the following, the major components of the MAS, which are provided in addition to those of the commercial aero-ball measurement system, or which are supplemented or modified, will be further described. Referring to FIG. 1, a commercial nuclear reactor comprises an instrumentation tube system 12 including at least one instrumentation finger 14 passing through a core 10 of the nuclear reactor. The instrumentation tube system 12 is configured to permit insertion and removal of irradiation targets 16 and optionally dummy targets 18 (cf. FIG. 2) into the instrumentation finger 14. The aero-ball measuring system of the commercial nuclear reactor is adapted to handle irradiation targets 16 having a round, cylindrical, elliptical or spherical shape and having a diameter corresponding to the clearance of the instrumentation finger of the aero-ball measuring system. Preferably, the diameter of the targets 16, 18 is in the range of between 1 to 3 mm, preferably about 1.7 mm. Conduits of the instrumentation tube system 12 penetrate an access barrier 11 of the reactor and are coupled to one or more instrumentation fingers 14. Preferably, the instrumentation finger penetrates the pressure vessel cover of the nuclear reactor, with the instrumentation finger 14 extending from the top to the bottom over substantially the entire axial length of the reactor core 10. An end of the instrumentation finger 14 at the bottom of the reactor core 10 is closed and/or provided with a stop so that the irradiation targets 16 inserted into the instrumentation finger form a column wherein each target 16 is at a predefined axial position. Preferably, one or more humidity sensors (not shown) may be provided in the instrumentation tube system 12 to detect any ingress of primary coolant (or any other liquid) into the MAS. It is understood that the instrumentation fingers 14 used for the MAS are in direct contact with the primary coolant surrounding fuel elements in the core of the nuclear reactor. The humidity sensors may be based on spark plugs which are modified for measuring electrical resistance. More preferably, the instrumentation tube system 12 comprises further sensors (not shown) for monitoring the presence and runtime of the irradiation targets 16 and optionally the dummy targets 18 passing through the instrumentation tube system 12, in particular into and out of the instrumentation finger 14. These sensors are preferably arranged at the conduits penetrating the reactor core 10. The measuring principle may be based on the detection of a variation of the magnetic flux as the irradiation targets 16 and/or the dummy targets 18 pass by the sensors for measuring transport time and completeness. Preferably, the sensors are used to monitor that all irradiation targets 16 have left the instrumentation finger 14 during the removal process while the targets pass the sensors. In addition, or as an alternative, activity sensors can be used for detecting the radiation of the irradiation targets 16 and/or the dummy targets 18. According to a preferred embodiment, the commercial nuclear reactor is a pressurized water reactor. More preferably, the instrumentation tube system is derived from a conventional aero-ball measuring system of a pressurized water reactor (PWR) such as an EPR™ or Siemens™ PWR nuclear reactor. The person skilled in the art will however recognize that the invention is not limited to use of an aero-ball measuring system of a PWR reactor. Rather, it is also possible to use the instrumentation tubes of the Traversing Incore Probe (TIP) system of a boiling water reactor (BWR), the view ports of a CANDU reactor and temperature measurement and/or neutron flux channels in a heavy water reactor. As shown in FIG. 1, the instrumentation tube system 12 is connected to a target drive system 20 configured to insert the irradiation targets 16 and optionally dummy targets 18 into the instrumentation finger 14 in a predetermined linear order and to force the irradiation targets 16 and dummy targets 18 out of the instrumentation finger 14 thereby retaining the linear order of the targets. Preferably, the target drive system 20 is pneumatically operated allowing for a fast processing of the irradiation targets 16 and optionally the dummy targets 18 using pressurized gas such as nitrogen or air. More preferably, the target drive system 20 comprises one or more pneumatically operated valve batteries (not shown) for separate control of the insertion and transport of the irradiation targets 16 and optionally dummy targets 18 in the instrumentation tube system. Based on this separation of controls, the regular aero-ball measuring system for determining the neutron flux in the core and the radionuclide generation system can be operated independently from each other. The valve batteries of the target drive system 20 may be implemented as a further subsystem in addition to the valve batteries of the conventional aero-ball measuring system, or a separate target drive system is installed. Moreover, the target drive system 20 may further comprise a gate system (not shown) including several mechanical and/or electro-mechanical devices configured to guide the irradiation targets 16 and optionally dummy targets 18 into selected conduits of the instrumentation tube system 12 and instrumentation fingers 14 in the reactor core 10. The target drive system 20 cooperates with a target harvesting system 22 configured to receive activated irradiation targets 16 and optionally dummy targets 18 from the instrumentation tube system 12 and pass a pre-defined quantity of the activated irradiation targets into a shielded storage container. The target harvesting system will be described in greater detail below, with reference to FIG. 3. The drive system 20 may also comprise a gate system (not shown) to guide the targets 16, 18 from the reactor core to the target harvesting system 22. With further reference to FIG. 1, an instrumentation and control unit (ICU) 24 is connected to the target drive system 20 and the target harvesting system 22 as well as an online core monitoring system 26 for controlling activation of the irradiation targets 16. Preferably, the ICU 24 is also connected to a fault monitoring system 28 of the aero-ball measuring system for reporting any errors in the MAS. The fault monitoring system 28 may also be designed without connection to the existing aero-ball measuring system, but be connected directly to the main control room. According to a preferred embodiment, the core monitoring system 26 and the instrumentation and control unit 24 are configured such that the activation process for converting the irradiation targets 16 to the desired radionuclide is optimized by considering the actual state of the reactor, especially the current neutron flux, fuel burn-up, reactor power and/or loading. Thus, an optimum axial irradiation position and irradiation time can be calculated for optimum results. It is however not important whether the actual calculation is performed in the ICU 24 or by the core monitoring system 26 of the aero-ball measuring system. The ICU 24 is connected with the online core monitoring system 26 software via an interface. The software is configured to calculate the required irradiation time for the targets online according to the actual neutron flux. The MAS is operated by the ICU 24. Start/Stop signals for target activation are exchanged between both systems. The ICU 24 is further connected to the mechanical components of the MAS, including the sensors. The online core monitoring system 26 of the conventional aero-ball measuring system, such as the POWERTRAX/S™ core monitoring software system available from Areva™, is able to provide substantially all relevant input data for the calculation of optimum activation conditions necessary for an efficient generation of radionuclides. Preferably, the information provided by the core monitoring system 26 to the instrumentation and control unit 24 includes at least one of the following: neutron flux (from ex- or in-core detectors), activation values from an existing aero-ball measuring system, burn-up, reactor power, loading, rod position(s), flow rate, inlet-temperature, pressure, and time synchronization. The more information about the reactor is considered as input data, the more accurate will be the results of the calculation of the optimum axial irradiation position and irradiation time. The before mentioned parameters may include real-time values and any derivatives, like developments over a pre-defined period of time. The information obtained from the core monitoring system 26 may also be used in the ICU 24 to calculate other parameters such as the amount of irradiation targets 16 in a specific instrumentation finger 14 defining the actual length of the respective target column, and the positions of the individual irradiation targets 16 and optionally dummy targets 18 within the target column. Based on the results of the calculations, the ICU 24 and/or an operator will operate the mechanical MAS components. In a preferred embodiment, the ICU 24 is configured such that operation of the valves of the target drive system 20 is at least partly automated to achieve a safe and reliable operation of the target drive system. More preferably, the instrumentation and control unit 24 may be configured to automatically control the pressure in the instrumentation tube system 12, in particular after each insertion of irradiation targets 16 and/or dummy targets 18 by the target drive system 20. Operation of the radionuclide generation system is preferably monitored and controlled at an operator station via a processing unit. The processing unit can be installed in a separate control cabinet in a control cabinet room (not shown). The processing unit is equipped with a display and, inter alia, allows for controlling specific parameters of the valve batteries of the target drive system 20 and target harvesting system 22. At the operator station the condition of the irradiation targets 16 during irradiation and the remaining irradiation time can be monitored. When the irradiation time of a set of targets 16 in an instrumentation finger 14 reaches the calculated time, a message prompts the operator to start the removal and harvesting process with respect to this instrumentation finger 14. The operation of the various valves of the target drive system 20 is partly automated so that repeated actions are performed safely and reliably. After each insertion of irradiation targets 16 and dummy targets 18 into the instrumentation tube system 12, the pressure in the tube system is checked and regulated in a fully automated manner. The ICU 24 also collects further digital signals representative of certain system conditions. Especially, the signals of the humidity sensors allow for a leakage monitoring such as ingress of primary coolant into the instrumentation tube system 12. A load cabinet of the aero-ball measuring system may also provide the electric power for the components of the radionuclide generation system, including the valve batteries and the processing unit in the control cabinet. An additional power inverter having appropriate fuses may be installed in the load cabinet. It is also possible to use an additional 24 volt supply provided in the control cabinet room. With reference to FIG. 2, an instrumentation finger 14 or other aero-ball conduit of the instrumentation tube system 12 penetrates a cover of the reactor pressure vessel. The instrumentation finger 14 extends from the top to the bottom over substantially the entire axial length of the reactor core 10. Irradiation targets 16 and optionally dummy targets 18 are inserted into the instrumentation finger 14 in a linear order to form a target column wherein each target 16, 18 is at a predefined axial position. The instrumentation finger 14 comprises a gas inlet port 30 at a top of the finger which is coupled to the target drive system 20. Alternatively, it is also possible to insert the targets into the instrumentation finger 14 from a drywell at the bottom of the reactor core 10, for example if the TIP system of a boiling water reactor is used as the instrumentation tube system 12. In this case, additional means for retaining the irradiation targets 16 and dummy targets 18 in the instrumentation finger 14 are provided. The irradiation targets 16 are made of non-fissile material and comprise a suitable precursor material for generating radionuclides which are to be used for medical and/or other purposes. More preferably, the irradiation targets consist of the precursor material which converts to a desired radionuclide upon activating by exposure to neutron flux present in the core of an operating commercial nuclear reactor. Useful precursor materials are Mo-98, Yb-176 and Lu-176 which are converted to Mo-99 and Lu-177, respectively. It is understood, however, that the invention is not limited to the use of a specific precursor material. The dummy targets 18 are made of an inert material which is not substantially activated under the conditions in the core 10 of an operating nuclear reactor. Preferably, the dummy targets can be made of inexpensive inert materials and can be re-used after a short decay time so that the amount of radioactive waste is further reduced. For use in a conventional aero-ball measuring system, the irradiation targets 16 and the dummy targets 18 have a round shape, preferably a spherical or cylindrical shape, so that the targets may slide smoothly through and can be easily guided in the instrumentation tube system 12 of the aero-ball measuring system by pressurized gas, such as air or nitrogen, and/or under the action of gravity. According to a preferred embodiment, the dummy targets 18 and the irradiation targets 16 have different magnetic properties. Preferably, either the dummy targets 18 or the irradiation targets 16 are magnetically attractable. More preferably, one of the irradiation targets 16 and the dummy targets 18 is ferromagnetic whereas the other one of the irradiation targets 16 and dummy targets 18 is non-magnetic or paramagnetic. Still more preferably, the dummy targets 18 are made from a ferromagnetic material such as iron or iron alloys, including ferritic stainless steel, or ferrite. With the aid of the online core monitoring system 26 it is possible to determine sections 32, 36 of the instrumentation finger 14 in which the neutron flux is too low for producing radionuclides, and sections 34 where the neutron flux is above the required irradiation target demand and thus sufficient for producing the desired radionuclides. In order to eliminate any waste of expensive irradiation targets 16, dummy targets 18 are provided and positioned preferably in the lower end section 32 of the irradiation finger 14 having a too low neutron flux density for radionuclide generation. According to the embodiment shown in FIG. 2, the irradiation targets 16 are positioned above and are held in place by the dummy targets 18 in a center section 34 of the irradiation finger 14 where the neutron flux is sufficient to completely convert the irradiation targets into the desired radionuclides, as determined by the ICU 24 and/or the online core monitoring system 26. In accordance with a preferred embodiment of the invention, an upper section 36 of the instrumentation finger 14 is kept empty. In a commercial pressurized water reactor, the center section 34 of the instrumentation finger useful for radionuclide generation generally extends over about 3-4 meters, and the end sections 32, 36 extend over 0.5 to 1 meter. These values may vary according to the reactor type and the actual operation status of the reactor, and will be different for boiling water reactors, heavy water reactors and CANDU reactors, respectively. In another embodiment (not shown), one or more of the irradiation targets 16 in the instrumentation finger 14 may be separated from each other by one or more dummy targets 18 thereby defining irradiation target subsections. The irradiation targets in the irradiation target subsections preferably have the same or different material properties. More preferably, the irradiation targets in adjacent subsections differ with respect to the precursor material used for radionuclide generation. This embodiment will enable the production of different radionuclides in a one-step operation. The optimum positioning of the irradiation targets 16 by means of inexpensive dummy targets 18 in an instrumentation tube system 12 of a commercial nuclear reactor provides an effective and economical production of radionuclides during reactor operation, and also avoids production of nuclear waste due to incomplete target activation. The irradiation target harvesting system 22 of the present invention is schematically shown in FIG. 3. A discharge tube 38 is connected to the instrumentation finger 14 through aero-ball conduits of the instrumentation tube system 12. The discharge tube 38 is configured to receive the irradiation targets 16 driven out of the instrumentation finger after activation is completed. The linear order of the irradiation targets 16 and/or the dummy targets 18 is retained in the discharge tube. Preferably, the discharge tube 38 is located outside the reactor core 10, but within accessible areas inside the reactor containment. The discharge tube 38 has an exit port 40 which can be coupled to at least one storage container 42, 42′ for receiving the activated irradiation targets 16 from the instrumentation finger 14. The storage container 42, 42′ preferably has a shielding to minimize an operator's exposure to radiation from the activated irradiation targets 16. A lock element 44 is provided in the discharge tube 38 for blocking movement of the activated irradiation targets to the storage container. The lock element 44 can be a magnetically or mechanically operated restriction element, preferably a pin crossing the discharge tube 38. Referring to FIG. 3, the discharge tube 38 comprises a first discharge tube section 46, a second discharge tube section 48 and an apex 50 formed at a conjunction of the first and second discharge tube section 46, 48. The apex 50 is the highest point of the discharge tube 38. The first and second discharge tube sections 46, 48 are directed downwardly from the apex 50. The exit port 40 is arranged at a free end of the first discharge tube section 46, opposed to the apex, and the second discharge tube section 48 is coupled to the instrumentation tube system 12. The exit port 40 comprises a valve element 52 for pressure-tight sealing the discharge tube. A T-junction 54 is located between the lock element 44 and the exit port 40, wherein the T-junction 54 has a gas inlet port 56 for supplying pressurized gas into the discharge tube 38, and also a gas outlet for discharging gas out of the system. Preferably, the T-junction is configured as a three-way valve. The pressurized gas can be supplied from the target drive system 20, or from an external gas bottle, in particular when the harvesting system is operated as a stand-alone system with no modification of the regular aero-ball measurement system. In the embodiment shown in FIG. 3, the first discharge tube section 46, the second discharge tube section 48 and the apex 50 are shaped in the form of an inverse U. Other profiles of the discharge tube 38 are possible as long as the apex 50 formed between the first and second discharge tube section 46, 48 has a radius which is sufficiently small to effectively separate the target columns in the first and second tube sections 46, 48 from each other. Moreover, as shown in FIG. 3, the lock element 44 is at a first level 58, and the second discharge tube section 48 has a base point 60 opposite to the apex at a second level 62. The first level 58 is higher than the second level 62. Accordingly a distance d1 between the lock element 44 and the apex 50 corresponding to a height of the target column in the first discharge tube section 46 is lower than the distance d2 between the base point 60 of the second discharge tube section 48 and the apex 50 corresponding to a height of the target column in the second discharge tube section 48. The targets in the second discharge tube section 48 may therefore have a higher mass than the total of the target column in the first discharge tube section 46. The height difference assists the effect of the apex 50 and facilitates separation of the targets without using any further mechanical means. The operation of the of the irradiation target harvesting system of the invention is now described in greater detail below. Irradiation targets 16 activated in the instrumentation finger 14 for a period of time sufficient to convert the targets into the desired radionuclide are driven out of the instrumentation finger 14 into the instrumentation tube system 12 using pressurized gas such as air or nitrogen supplied from the target drive system 20. The discharge tube 38 is coupled to conduits of the instrumentation tube system 12 for receiving the irradiation targets 16. A gate system such as a three-way valve can be used to guide the irradiation targets 16 into the discharge tube 38 of the target harvesting system 22. The linear order of the irradiation targets 16 in the instrumentation finger 14 is preserved in the discharge tube 38. At this time, access to the exit port 40 of the discharge tube 38 is blocked by the lock element 44 providing a stop for the activated irradiation targets 16 and preventing the targets 16 from leaving the discharge tube 38. The activated irradiation targets 16 entering the discharge tube 38 are passed over the apex 50 formed at a conjunction between the first and second discharge tube section 46, 48. A pre-defined quantity (16′) of the activated irradiation targets 16 is located proximate to the exit port 40 in the first discharge tube section 46. The quantity (16′) of the irradiation targets 16 in the first discharge tube section 46 directly corresponds to the length of the first discharge tube section 46 between the lock element 44 and the apex 50. The apex 50 is located at the highest point of the target column in the discharge tube 38. The valves of the drive system 20 are then closed and the pressure in the instrumentation tube system 12 and the discharge tube 38 is relieved. The exit port 40 of the discharge tube is coupled to a shielded storage container 42, and the lock element 44 is opened so as to release the pre-defined quantity (16′) of the activated irradiated targets 16 located on one side of the apex 50 in the first discharge tube section 46 and pass the targets 16 into the storage container 42 under action of gravity. The other quantity (16″) of activated irradiation targets 16 located on the other side of the apex 50 in the second discharge tube section 48 is retained in the discharge tube 38, or may flow back into the instrumentation tube system 12, also under action of gravity. The valve element 52 at the exit port 40 is closed for providing a pressure-tight sealing of the exit port 40 and the discharge tube 38, and the shielded storage container 42 is then removed either manually or by means of an automated handling device. In a preferred embodiment, the quantity (16″) of the activated irradiation targets 16 kept in the second discharge tube section 48 by means of the apex 50 is actively transferred from the discharge tube 38 into a holding position (not shown) in the instrumentation tube system 12 prior to removing the storage container 42 from the exit port 40. The holding position may be located within an access-restricted area of the reactor containment. After the portion or quantity (16′) of the activated irradiation targets 16 in the first discharge tube section 46 is transferred into the storage container 42, the valve element 52 at the exit port 40 is closed and pressurized gas from the target drive system 20 is blown into the gas inlet port 56 at the T-junction 54 near the exit port 40. The pressurized gas forces the other quantity (16″) of the activated irradiation targets 16 out of the second discharge tube section 48 back into the holding position in the instrumentation tube system 12 or into the instrumentation finger 14. The valves of the target drive system 20 are then closed and the flow of pressurized air is stopped. The storage container 42 is removed from the exit port 40 and transported to a packaging facility or shipped to the desired application site. Since the discharge tube 38 is completely free of activated irradiation targets 16 the radiation exposure to operating personnel is further minimized. The above process steps can then be repeated for portioning and harvesting a further quantity of activated irradiation targets 16 until all of the activated irradiation targets 16 have been removed from the instrumentation tube system 12. The system is then ready for starting a new radionuclide generation cycle. According to the invention, a portion of the activated irradiation targets 16 in the first discharge tube section 46 is reliably and reproducibly separated from the other activated irradiation targets 16 which are kept in the discharge tube 38 simply by action of gravity without using additional mechanical elements and/or target counters. Since the length of the target column in the first discharge tube section 46, between the lock element 44 and the apex 50 also limits the amount of the targets and thus the total activity harvested in the storage container 42, less shielding of the containers will be necessary, and smaller containers can be used which require less space and are easier to handle. The simpler set-up of the target harvesting system 42 facilitates integration into existing structures within the reactor confinement. A further embodiment of the irradiation target harvesting system 22 is schematically shown in FIG. 4. Components of the harvesting system having the same function as in the previous embodiment are indicated with the same reference numerals. Referring to FIG. 4, the discharge tube 38 is configured to receive the activated irradiation targets 16 and the dummy targets 18 driven out of the instrumentation finger after activation is completed. The linear order of the irradiation targets 16 and the dummy targets 18 is preserved in the discharge tube 38. The irradiation targets 16 and the dummy targets 18 have different magnetic properties. Preferably, the dummy targets 18 are ferromagnetic whereas the irradiation targets 16 are non-magnetic or paramagnetic, or vice versa. One or more magnets 64 are arranged at the first discharge tube section 46 between the apex and the lock element. Preferably, the one or more magnets 64 are selected from a permanent magnet and a solenoid. The lock element 44 is actuated magnetically and comprises a pin 66 crossing the first discharge tube section 46 and an electromagnetic coil 68 for operating the pin 66. The one or more magnets 64 are surrounding the first discharge tube section 46. Preferably, the magnets 64 are movably arranged along a longitudinal axis of the first discharge tube section 46. The one or more magnets 64 and/or the lock element 44 are remotely controlled by the ICU 24 so that an automatic and fast processing of the irradiation targets 16 and the dummy targets 18 is achieved. The T-junction 54 near the exit port in the first discharge tube section comprises a gas inlet 56 linked to the target drive system 20 to blow pressurized gas such as air or nitrogen into the discharge tube 38 and/or instrumentation tube system 12 for driving the irradiation targets 16 and/or dummy targets 18 out of the first and/or second discharge tube section to a holding position in the instrumentation tube system 12 or the instrumentation finger 14. The exit port 40 of the first discharge tube section can be coupled to one or more storage containers 42, 42′ and/or an intermediate tank 70 to selectively receive the irradiation targets 16 or dummy targets 18 released from the first discharge tube section 46. The storage containers 42, 42′ and/or the intermediate tank 70 may be movable to match with the exit port 40, or the exit port 40 is supported on a pivotable joint 72 to be movable between storage containers 42, 42′ and the intermediate tank 70. Alternatively, the exit port 40 may include a gate to selectively direct the irradiation targets 16 or dummy targets 18 to the appropriate containers 42, 42′ and tank 70. The embodiment of the target harvesting system shown in FIG. 4 is useful for portioning the activated irradiation targets 16 in the discharge tube, as described above, and also for separating the activated irradiation targets 16 from dummy targets 18 in the first discharge tube section 46 due to their different magnetic properties. In particular, the dummy targets 18 and/or the irradiation targets 16 can be exposed to a magnetic field to retain either the dummy targets 18 or the irradiation targets 16 in the first discharge tube section 46 and release the other one of the irradiation targets 16 or the dummy targets 18 from the first discharge tube section 46. For operating the radionuclide generation and target harvesting system, the irradiation targets 16 and the dummy targets 18 are inserted into the instrumentation finger 14 using the target drive system 20, and the irradiation targets 16 are activated by exposure to neutron flux in the nuclear reactor core when in power generating operation to substantially convert the precursor material of the irradiation target into the desired radionuclide. The dummy targets 18 and the irradiation targets 16 are arranged in the instrumentation finger 14 in a pre-calculated linear order such that the dummy targets 18 hold the irradiation targets 16 at a predetermined axial position in the reactor core. The optimum axial position of the irradiation targets 16 is calculated by the ICU 24 and/or the online core monitoring system 26 and corresponds to a neutron flux density sufficient for completely converting the irradiation targets 16 to the radionuclide during a predetermined period of time. The remaining positions in the instrumentation finger 14 are occupied by the dummy targets 18 which hold the irradiation targets 16 in place. The irradiation targets 16 are preferably positioned in the center section 34 of the instrumentation finger 14 in the reactor core 10, and the dummy targets 18 are preferably positioned in the end section 32 of the instrumentation finger 14, i. e., in the lower part of the reactor core 10 where the neutron flux density is insufficient for completely activating the irradiation targets 16 (cf. FIG. 3). As shown in FIG. 2, the space above the irradiation targets 16 in the upper section 36 can remain empty. The irradiation targets 16 are activated in the instrumentation finger 14 for a period of time sufficient for complete conversion of the precursor material of the irradiation targets to the desired radionuclide, as determined by the online core monitoring system 26 and instrumentation and control system 24. The time to achieve complete conversion of the precursor material will depend on the reactor type and status, neutron flux conditions, type of precursor material and various other parameters known to a person skilled in the art, and may range from several hours to days, or until saturation of activity. Complete conversion means a conversion rate of the precursor material providing a radionuclide content suitable for medical or industrial application of the irradiation targets 16. After activation of the irradiation targets 16 and conversion to the desired radionuclide is completed, the target drive system 20 is operated to force the dummy targets 18 and the activated irradiation targets 16 out of the instrumentation finger 14 into the discharge tube 38 using pressurized gas such as nitrogen or air. The linear order of the dummy targets 18 and the irradiation targets 16 in the instrumentation finger 14 is preserved in the discharge tube 38 so that the irradiation targets 16 are located proximate to the exit port of the discharge tube 38. The dummy targets are at a distal end of the target column in the discharge tube 38 or instrumentation tube system 12. At this time, the discharge tube 38 is blocked by the lock element 44 providing a stop for the targets 16, 18 and to prevent the activated irradiation targets 16 and dummy targets 18 from leaving the discharge tube. The valves of the drive system 20 are then closed and the pressure in the instrumentation tube system 12 is relieved. The irradiation targets 16 entering the discharge tube 38 are passed over the apex 50 formed at the conjunction between the first and second discharge tube section 46, 48. A pre-defined quantity of the activated irradiation targets 16 is located proximate to the exit port 40 in the first discharge tube section 46. The quantity of the irradiation targets 16 in the first discharge tube section 46 directly corresponds to the length of the first discharge tube section 46 between the lock element 44 and the apex 50. The exit port 40 of the discharge tube is coupled to a shielded storage container 42, and the lock element 44 is opened so as to release the pre-defined quantity of the activated irradiated targets 16 located on one side of the apex 50 in the first discharge tube section 46 and pass the targets 16 into the storage container 42 under action of gravity. The other quantity of activated irradiation targets 16 and dummy targets 18 located on the other side of the apex 50 in the second discharge tube section 48 is kept in the discharge tube 38, or may flow back into the instrumentation tube system 12, also under action of gravity. The valve element 52 at the exit port is then closed for providing a pressure-tight sealing of the exit port 40 and the discharge tube 38, and the shielded storage container 42 is removed either manually or by means of an automated handling device. In a preferred embodiment, the quantity of the activated irradiation targets 16 and dummy targets 18 kept in the second discharge tube section 48 by means of the apex 50 is transferred from the discharge tube 38 into a holding position in the instrumentation tube system 12 prior to removing the storage container 52 from the exit port, as described above with reference to FIG. 3. The above process steps can then be repeated for portioning and harvesting further quantities of activated irradiation targets 16 from the discharge tube 38. When sensors at the instrumentation tube system 12 indicate the presence of dummy targets 18 and irradiation targets 16 in the first discharge tube section 46, as shown in FIG. 4, the dummy targets 18 and/or the activated irradiation targets 16 are exposed to a magnetic field to retain either the dummy targets 18 or the activated irradiation targets 16 in the first discharge tube section 46 and release the other one of the activated irradiation targets 16 or the dummy targets 18 from the first discharge tube section 46 through the exit port 40 into a storage container 42, 42′ or intermediate tank 70. For separating the irradiation targets 16 from the dummy targets 18 and selectively removing the irradiation targets 16 from the first discharge tube section 46, the solenoids 64 are moved along the longitudinal axis of the first discharge tube section 46 and arranged adjacent to the ferromagnetic dummy targets 18 so that one or more dummy targets 18 proximate to the exit port 40 are associated to and exposed to the magnetic field of a solenoid 64. The lock element 44 is then opened, and the non-magnetic irradiation targets 16 are released from the first discharge tube section 46 under the action of gravity and passed into the storage container 42, 42′ for further processing and/or shipping to an application site. The magnetic dummy targets 18 are kept in the first discharge tube section 46 by the action of the magnetic field generated by the solenoids 64. After the irradiation targets 16 are separated from the dummy targets 18 and are harvested in the storage container 42, 42′, the exit port is coupled to the intermediate tank 70, the magnetic field is switched off and the dummy targets 18 are transferred to the intermediate tank 70 under the action of gravity for further use after a short decay period. The dummy targets 18 kept in the second discharge tube section 48 can be forced out of the discharge tube into the intermediate tank 70 using pressurized gas from the target drive system 20. Alternatively, some or all of the dummy 18 targets can be driven back into the instrumentation finger 14 by closing the valve element 52 at the exit port 40 and blowing pressurized gas from the target drive system 20 into the gas inlet port 56 at the T-junction 54. A new radionuclide generation cycle can then be started by inserting fresh irradiation targets 16 and/or dummy targets 18 into the instrumentation tube system 12. According to another embodiment, the irradiation targets 16 in the instrumentation finger 14 may be separated by one or more dummy targets 18 thereby defining irradiation target subsections (not shown). The irradiation targets 16 in the irradiation target subsections can have the same or different material properties. Also in this embodiment, the activated irradiation targets 16 and dummy targets 18 are driven out of the instrumentation finger 14 into the discharge tube 38 preserving the linear order of the dummy targets 18 and the irradiation targets 16. The first discharge tube section 46 is blocked by the lock element 44 so as to prevent the irradiation targets 16 and dummy targets 18 from leaving the discharge tube 38. If one or more ferromagnetic dummy targets 18 are now positioned ahead of the irradiation targets 16 proximate the exit port 40, the solenoids 64 are arranged adjacent these dummy targets 18. Pressurized gas from the target drive system 20 is blown into the gas inlet port 56 of the T-junction 54 and used to drive the non-magnetic irradiation targets 16 over the apex 50 back into the second discharge tube section 48 or a holding position in the instrumentation tube system 12. The magnetic dummy targets 18 are retained in the first discharge tube section 46 by means of the activated magnetic field. After the valves of the target drive system 20 are closed, the magnetic field is switched off, the lock element 44 is opened, and the dummy targets 18 are released from the discharge tube 38 under the action of gravity and transferred into the intermediate tank 70 coupled to the exit port 40. In the next step, the lock element 44 is closed and the irradiation targets 16 and optionally remaining dummy targets 18 are driven out of the instrumentation finger 14 or holding position using pressurized gas from the target drive system 20. The activated irradiation targets 16 are now positioned proximate the lock element 44 and the exit port 40 and can be portioned and/or separated from the dummy targets 18 as described above. If necessary, the portioning and separation steps can be repeated until all irradiation targets 16 from the various target subsections are selectively harvested from the instrumentation tube system 12 in the multiple storage containers 42, 42′. It is understood that the target separation and harvesting method of the invention will also be applicable vice-versa if the dummy targets 18 are non-magnetic and the irradiation targets 16 are ferromagnetic. The radionuclide generation and target harvesting systems according to the invention can also be installed in a nuclear power plant having no conventional aero-ball measuring system. The aero-ball measuring system as described above only provides a basis to facilitate an installation of the radionuclide generation system since no additional instrumentation tubes, fingers and the like need to be installed only for the radionuclide generation. Possible reactor types for such an application include boiling water reactors, heavy water reactors and CANDU (CANada Deuterium Uranium) reactors.
	description	Applicant claims priority under 35 U.S.C. §119 of German Application No. 199 27 694.3 filed Jun. 17, 1999. Applicant also claims priority under 35 U.S.C. §365 of PCT/DE00/01934 filed Jun. 13, 2000. The international application under PCT article 21(2) was not published in English. 1. Field of the Invention The invention is a semiconductor sensor with pixel structure, containing a capacity for each pixel which stores charge and converts it into voltage that can be read out. The pixel structure is in the main completely covered with a conductive layer. Furthermore the invention is about the sensor's use integrated in a vacuum system (IC) with photo cathode. The semiconductor is ideal for image processing in opto-electronic applications. 2. Prior Art Opto-electronic image converters are known in semiconductor technology. With the so-called CCD-device electrons are removed from the lattice structure by the incidence of photons on the sensor surface, which are then stored in so-called pixel cells at the conversion place and can be read out afterwards in different ways. Moreover active-pixel-sensors (APS) are known, where each pixel is connected with at least one transistor that converts or decouples the pixel information for direct read out at inquiry. A disadvantage, however, is that the known image converters react sensible against electrons that striking the sensor surface. The kinetic energy of the electrons influences the crystalline semiconductor structure, so that the system is forced open under operation with the result of pixel defects, which show for example in dark currents. Therefore it does not make sense to integrate such a semiconductor element into an image intensifier tube, to, for example, convert the information directly into a video signal. A pixel structure covered by a conductive layer is known from the EP 0 588 397 A2 in the form of an x-ray image detector. The conductive layer serves as a pre-voltage electrode, in order to lead charges that are generated in a photo conductor layer, under influence of a by this electrode produced electric field, to each individual gathering electrode of the pixels. Similar semiconductor image sensors are known for example from the EP 0 444 702 A1, the patent abstract of Japan JP 06310699 A, 1994, JPO, the U.S. Pat. No. 4,789,888 and the U.S. Pat. No. 5,311,038 in the form of light- and/or x-ray sensible sensors. From the two last-mentioned files it is known in addition to cover gaps, in order to avoid the unwelcome influence of incident light. From the U.S. Pat. No. 3,902,095 an electron beam semiconductor intensifier with shielded diode junctions is known, which is designed as electron beam semiconductor tube, for example as transmitter amplifier, switching tube or as driving circuit for other high-voltage intensifier tubes, like for example traveling-wave power tubes for radar transmitters. The semiconductor is partially covered with a conductive coating, the shielding having openings above a segment of the diode junction. Here the electrons are to strike well-aimed, the screening preventing unwelcome electron bombardment in the periphery areas of the device (diode). By this unwelcome surface- and bulk charging effects are eliminated. Therefore the maximum reverse bias voltage is increased and hence the maximum possible output power of tubes employing such diodes is increased, too. The interconnection of a plurality of such diodes causes an increase of the intensifier's efficiency, the capacitance being reduced in proportion to the exposed surface. Nevertheless the semiconductor tube is neither designed nor suited for an image transmitting detection. Moreover from the patent abstracts of Japan: JP 2-94566 A E-944, 1990, Vol. 14/No. 292 a semiconductor with pixel electrodes with a further, at least gap-covering layer is known. But this is an optical semiconductor sensor, at which multiple reflections on the inside of the optical covering (layer are reduced or eliminated by a reflection avoiding film. Besides, from the DE 42 23 693 A1 an x-ray image intensifier together with a vacuum housing, an input illumination screen, an electron optic and an image sensor, which is located on the side of the vacuum housing that is opposite to the input illumination screen, is known, where the from a photo cathode resulting electrons are converted directly into a video signal in a CCD-image converter. At this, the image sensor contains a layer system, which has at least one semiconductor layer that initiates an electron conversion and is of amorphous semiconductor material. This publication therefore shows a concrete different solution of the following problem. Based on the EP 0 588 397 A2 it is the task of the invention to describe a semiconductor sensor with a pixel structure that detects electrons and whose semiconductor structure is protected from electron bombardment. This task is solved by a semiconductor sensor according to an embodiment of the invention. The covering of the pixel surface with a conductive layer permits the capture of the incident electrons, that are stored as charge and are converted into voltage and read out in a for semiconductor image sensors conventional way. The conductive cover acts as electrode for the incident electrons, the electrode being part of a capacity and therefore serving for charge storage. The essential is that the electron-sensible pixel surface and the semiconductor structures which lie beneath are screened by the conductive layer. The incident electrons by this are processed separately as measuring signal for the particular struck pixel. The electrons that strike the coating are therefore caught pixel orientated and prevented from penetrating into the semiconductor structure and from there causing dark currents, faulty pixels, failure of lines and columns, or the like defects. As the surface of gaps between the pixels is covered with a second conductive layer, the second conductive layer being insulated from the pixel surface coatings, the gaps between the pixels are also protected against the unwelcome entry of electrons into the semiconductor layer. This from the pixel surface coatings insulated second layer can by application of a potential be used for field modification, for example as acceleration potential. The layers preferentially consist of a conductive, light-impervious material, for example metal. Incident electrons are absorbed by the conductive material and led to the semiconductor structure. The light-impervious material prevents hazardous photons from penetrating the sensor surface, and undesirable signals are thus avoided. The layers preferentially consist of aluminum, as this material is easy to apply, and has a good conductivity and light-imperviousness. Depending on the design of the semiconductor, the second conductive layer that covers the gaps, can be used as capacitor electrode. Moreover a potential for the acceleration of the incident electron current can be applied to the second conductive layer. If the sensor's detection surface is provided with a electron-intensifying material and the channels to the pixel surfaces are given, an electron multiplication right in front of the sensor surfaces is achieved. The openings in the coating allow the electrons to pass unhindered to the sensor surface and electrons that strike the sides of the openings only lead to an electron multiplication and a signal intensification. Due to the fact that the electron-intensifying coating on the upper- and lower side is each covered with a conductive thin layer to which an electric potential is applied, the desired electron flow is guaranteed and an acceleration voltage is applied by the layer thickness of the electron-intensifying coating. If neighboring pixel surfaces have different potential, a charge binning of the electrons that strike the sensor surface on particular, pre-selected pixels can be generated. Such a charge binning, the so-called ‘Binning’, can for example be used for a preliminarily reduced resolution with a higher detection ratio. In the semiconductor used according to an embodiment of the invention, an image sensor is given with which photons are at first converted into electrons in a vacuum system and these electrons are than directly converted into an image-giving voltage signal (video signal). The semiconductor does not show pixel defects at the desired bombardment of electrons, therefore a totally image-oriented conversion of the at the beginning striking optical signal into a electric signal, for example for the display on a video monitor, takes place. It is best to generate the intensification in the image intensifier tube with one or more multi-channel plates. The multi-channel plates are responsible for the intensification of the image information. In FIG. 1 a part of a semiconductor sensor with two pixels is shown. On the sensor's surface that shows into the direction of the electron flow which is to be detected, in FIG. 1 above, conductive layers, which are assigned to the pixel structure 1, are assembled as active pixel surfaces 11, that preferably consist of aluminum. This conductive layer 11 is through-connected to a semiconductor element 3. The semiconductor element 3 also has a pixel structure 1, which reproduces an assigned circuit for each pixel surface 11 that consists of conductive material. Besides the not shown semiconductor element 3 can be designed as active-pixel-structure. Between the conductive layer 11 of each pixel and the semiconductor element 3 an insulation 13 is planned with the exception of contact 12. The insulation 13 separates the sensor surface 11 which is contacted 12 to the semiconductor element 1 from a second conductive layer 21 which screens the gaps between the pixel surfaces 11 in such a way that here striking electrons are prevented from reaching the semiconductor structure below and from causing mistakes. The second conductive layer 21 in the example is applied to a passivation layer 2 which is applied to the upper side of the semiconductor element 3 together with recesses for the insulated connections 12, 13. In the following the operation of the semiconductor sensor is explained. The electron flow E, that is to be detected, strikes the conductive layer 11. Each pixel-orientated registered electron is by some sort of capacitor electrode led from the active pixel surfaces 11 to the semiconductor element 3, in order to convert charge into voltage. Due to the semiconductor's design it is possible to use the second conductive layer 21 as capacitor electrode, too. Moreover it is also possible to here apply a voltage to accelerate the incident electron flow E. The conductive layer 11 of each pixel is designed as electrode to a ‘floating diffusion’ 31 or ‘floating gate’ for charge conversion. FIG. 2 shows a preferred combination of the semiconductor sensor in detail similar to the design according to FIG. 1 in a vacuum system 4 in schematic sectional view. The vacuum system 4 has an input window 41 with a photo cathode 42. In the vacuum area 46 of the device 4 at least on multi-channel-plate 43 with a variety of channels 44 is planned. Unlike conventional image converter—respectively image intensifier tubes no illumination screen for the conversion of the electron flow E into visible light is necessary at the output window 45, but directly arranged within the vacuum system of the semiconductor sensor according to the design of FIG. 1. Here the direction of the pixel structure 1 causes that the signal information from the channels 44 of the multi-channel-plate 43 exactly strikes the conductive layers 11 of the pixels. Therefore an image-orientated processing of the image information which was taken at the photo cathode 42 is possible. The output window 45 is arranged as termination element for the vacuum area 46 on the backside of the semiconductor element 3. In this termination element 45 the connection of the semiconductor structure can be led outside. In the following the way of operation of the image processing lay out will be explained: Optical image information in the shape of a photon current Ph reach the photo cathode 42 of the vacuum system (VA-MOS) 4 through the input window 41. There electrons are emitted from the photo cathode material by the incident photons. In order to intensify the from the photo cathode emitted electron flow E the electrons are accelerated by a electric field at the multi-channel-plate. The resulting electron flow reaches the pixel surfaces 11 of the semiconductor sensor with its pixel structure 1 through the channels 44 of the multi-channel-plate. Below the semiconductor the vacuum area 46 of the system 4 is closed with a termination element 45 which for example is designed as ceramic plane with through-connections for the semiconductor sensor. The operation of the semiconductor sensor corresponds to that described in the assembly example according to FIG. 1. There the electron flow E is directly converted from the semiconductor sensor into an electric signal. By the help of this structure a received image can therefore be converted in a post signal processing on the semiconductor element or extern, for example in a video signal. In FIG. 3 a semiconductor sensor according to FIG. 1 is shown, on whose sensor surface an electron-intensifying coating 5 is planned. The coating in the main consists of an electron-intensifying material 51, for example alpha-silicon. To achieve a corresponding electron intensification in this material the coating 5 has a upper and a lower thin layer 52,53 which consist of conductive material. For the contact surface between the lower thin layer 53 and the pixel surface 11 a insulation is planned, for example the insulation 13. To lead the electrons to the pixel surfaces 11 transit channels 54 in the coating 5 are planned. These transit channels 54 are preferably pixel-orientated assembled. It must be emphasized that the invented semiconductor sensor is able to convert an electron flow E direct, without preliminary conversion into light signals, and pixel-orientated into an electric signal. Therefore there is no danger of pixel damage like in conventional image sensors. 1 pixel structure 11 pixel surface, conductive layer 12 connection 13 insulation 2 passivation layer 21 second conductive layer 22 gap 3 semiconductor 31 floating diffusion 4 vacuum tube 41 input window 42 photo cathode 43 multi-channel-plate (MCP) 44 channel 45 output window or termination element 46 vacuum area 5 electron-intensifying coating (secondary electron emitting coating) 51 electron-intensifying material (alpha-silicon) 52 upper thin layer 53 lower thin layer 54 transit channel E electrons Ph photons
	summary
050680820	summary	BACKGROUND OF THE INVENTION This invention relates to a fuel assembly for a nuclear reactor and more particularly to a fuel assembly for a boiling water reactor capable of ensuring a long operation cycle and maintaining a high shut-down margin. A fuel assembly for a boiling water reactor (BWR) is constructed by a square channel box in which a number of fuel rods each comprising a metallic clad in which nuclear fuel material is packed are regularly arranged. The reactor core of the BWR includes a plurality of cells each comprising a cruciform control blade and four fuel assemblies surrounding the control blade and these cells are arranged in a regulated manner in the core. Namely, each fuel assembly and control blade have axes perpendicular and parallel to each other, and a coolant operated as a moderator flows from the lower portion towards the upper portion of the reactor core. Concerning the BWR, steam void is not formed in a portion near the lower end of the core effective portion, i.e., the lower end of a heat generating portion at which an exothermic reaction is performed, but a lot of voids is generated at the area above the central portion of the reactor core, and the generated voids move up towards the upper portion of the reactor core. Accordingly, the void fraction in the BWR becomes high towards the upper portion of the reactor core, and as a result, the moderation characteristics to neutrons are lowered and hence the output power is also lowered. In order to obviate these defective matters, in a conventional technique, it has been performed to increase the enrichment of the fissile nuclide to be contained in the fuel at a portion of high void fraction or it has been performed to mix a burnable poison with the fuel element to suppress the increasing of the power output at a portion of low void fraction. For the reasons described above, in the BWR, the burn-up at the upper portion of the core is liable to be delayed, and hence the concentration of U-235 becomes relatively higher than that of the other portion of the core. In addition, since a fissile nuclide such as Pu-239 is generated by the voids, it is difficult to maintain the shut-down margin of the reactor core at the upper portion thereof. Moreover, recently, many efforts have been made for elongating the reactor operation cycle of the reactor and improving the degree of burn-up of the fuel in order to satisfy the economical requirement. In these cases, however, the enrichment of the fuel is necessarily increased, so that the maintenance of the shut-down margin of the reactor is made further difficult. The fuel assembly which has been conventionally used and a fuel assembly which is expected to be used in the near future for the boiling water reactor (BWR) will be described hereunder by way of typical examples with reference to the drawings. FIG. 73A is a perspective view of a fuel assembly of conventional type and FIG. 73B is a schematic vertical sectional view of a fuel rod consisting of the fuel assembly. Referring to FIG. 73A, the fuel assembly comprises water rods, not shown, and fuel rods 2 secured by an upper tie plate 4, a spacer 5 and a lower tie plate 6, and a channel box 1 surrounding the outer periphery of the thus secured water rods and fuel rods 2. Each of the fuel rods 2, as shown in FIG. 73B, comprises a clad or sheath 7, a plurality of fuel pellets 8 arranged in the clad 7, a spring 9 located in a gas plenum disposed above the pellets 8 in the clad 7, an upper plug 10 for closing the upper opening of the clad 7, and a lower plug 11 for closing the lower opening of the clad 7. FIG. 74 is a cross sectional view of the conventional fuel assembly shown in FIG. 73A, in which sixty-two fuel rods 2 and two water rods 3 are arranged in the channel box 1 to constitute the fuel assembly. The water rods 3 serve to suppress the shortage of the water acting as the moderator in the interior of the fuel assembly, but the water rods 3 are axially uniformly arranged, so that there may arise such problems as excessive water condition at the lower portion of the reactor core or water shortage condition at the upper portion thereof. FIG. 75 also shows a cross sectional view of a fuel assembly which has been developed for improving the characteristics of the fuel assembly shown in FIG. 74, and the fuel assembly shown in FIG. 75 includes one water rod 12 having a diameter larger than that of the water rod 3 to pass non-boiling water therethrough. However, even in this example, there arises a problem of the excessive water condition at the lower portion of the reactor core and the water shortage condition at the upper portion thereof as described with respect to the former example shown in FIG. 74. FIG. 76 shows a cross sectional view of a further example of a conventional fuel assembly developed for improving the fuel assembly of the type shown in FIG. 74, and the fuel assembly of FIG. 76 comprises four square channel boxes 13 each containing sixteen fuel rods 2 which are arranged to constitute a water area of a boiling moderator material and a cross-shaped space 14 defined by the respective channel boxes 13 constitutes a water area of a non-boiling moderator material to thereby aim the uniform distribution of the output power in the horizontal direction. With the fuel assembly of this character, however, there also arise problems of the excessive water condition at the lower portion of the reactor core and the water shortage condition at the upper portion thereof. FIG. 77 shows a cross section of a still further example of the conventional fuel assembly of the type improving that shown in FIG. 75. The fuel assembly of FIG. 77 is constructed by nine sub-bundles 15 each comprising nine fuel rods 2, and relatively wide gaps 16 are defined between the respective sub-bundles 15. With the fuel assembly of this example, the problems of the excessive water condition and the water shortage condition at the lower and upper portions of the reactor core have not been solved. As described hereinabove, concerning the BWR, steam voids are formed in the area of the location of the fuel assemblies except the lowest portions thereof and the voids move up towards the upper portion of the reactor core, and accordingly, the void fraction in the BWR becomes high towards the upper portion of the reactor core. As a result, the moderation characteristics to neutrons are lowered and hence the fission rate is also lowered. In other words, the burning progresses at the lower portion of the reactor core and the burning is delayed at the upper portion thereof. In order to obviate this phenomenon; that is, in order to suppress the lowering of the output power at the upper portion of the reactor core, it has been performed to increase the enrichment of the fissile nuclide to be contained in the fuel disposed at the upper portion of the reactor core. However, the increasing in the void fraction at the upper portion of the reactor core and the increasing in the enrichment of the fissile nuclide of the upper portion of the reactor core will result in the difficulty for maintaining the shut-down margin at the upper portion of the reactor core in the shut-down period of the BWR. On the other hand, in order to elongate the reactor operation cycle to meet the economical requirement, it will be desired to further increase the enrichment of the fuel. However, these facts result in the further reduction of the subcriticality at the upper portion of the reactor core, and finally, there may arise a case where the reactor is not shut-down. Because of this problem, in the conventional technique, it is considerably difficult to elongate the operation cycle of the reactor. SUMMARY OF THE INVENTION An object of this invention is to substantially eliminate the drawbacks and defects encountered in the conventional technique described above and to provide an improved fuel assembly particularly constituting a reactor core of a water boiling reactor (BWR) capable of ensuring the maintenance of reactor shut-down margin even in the increasing in the enrichment of a fuel, and improving an axial output power distribution. This and other objects can be achieved in one aspect according to this invention by providing a fuel assembly of the type in which a number of fuel rods each constructed by filling a fuel material in a clad are arranged, the fuel assembly comprising at least one first fuel rod having a partial effective fuel area filled with a fuel material and having a portion in which enrichment of a fissile nuclide is significantly reduced in a clad of the fuel rod or the fissile nuclide does not exist at all at an axial level including a portion (called shutdown zone) at which subcriticality is made small at a period in which maintenance of reactor shut-down margin is made difficult during a reactor operation period, and a second fuel rod having a total effective fuel area filled with a fuel material throughout an entire axial length of the clad of the fuel rod. According to the preferred embodiment of this invention, the first fuel rod having a partially effective fuel area may be constructed as a fuel rod provided with a partially disposed interposed zone in which the enrichment of the fissile nuclide is significantly reduced in the clad or the fissile nuclide does not exist at all. The first fuel rod may be further constructed according to the preferred embodiment of this invention a short fuel rod having an axial length shorter than that of the second fuel rod having a total effective fuel area. Further according to the preferred embodiment of this invention, tube means through which a moderator passes are arranged symmetrically with respect to at least one diagonal line, in cross section, of the fuel assembly and the first fuel rod having the partial effective fuel area is disposed in an area defined between these tube means. According to this invention, there is also provided, in another aspect, a fuel assembly of the type in which a number of fuel rods each constructed by filling a fuel material in a clad are arranged, the fuel rod comprising at least one first fuel rod having a partial effective fuel area filled with a fuel material and having a portion in which enrichment of a fissile nuclide is significantly reduced in a clad of the fuel rod or the fissile nuclide does not exist at all at an axial level including a first portion at which subcriticality is made small at a period in which maintenance of reactor shutdown margin is made difficult during a reactor operation period and a second portion located between the first portion and a lower end of the effective fuel area, and a second fuel rod having a partial interposed zone in which enrichment of a fissile nuclide is significantly reduced in a clad of the fuel rod or the fissile nuclide does not exist at all at the second portion mentioned hereinabove. According to the fuel assembly of the characters described above, in regions or zones axially located adjacent to an interposed zone of a fuel rod in which the enrichment of the fissile nuclide is significantly reduced or the fissile nuclide does not exist at all, the neutron interaction (binding effect) is weakened at a reactor cold period and is increased in a reactor high temperature operation period, particularly, during the occurrence of voids. This phenomenon will be explained with reference to the action of the thermal neutrons each having a short diffusion length. Namely, since the density of water is large (i.e., about 1.0 g/cm.sup.3) in the reactor cold period, the diffusion length of the thermal neutron becomes small and the interaction of the thermal neutrons in the zones adjacent to the interposed zone of the fuel rod is reduced, and as a result, the neutron multiplication characteristics are lowered. With the boiling water reactor, in the reactor high temperature operation period, the temperature of the water is about 286.degree. C. (reference value) and the density thereof is about 0.74 g/cm.sup.3 even when no void is generated, and the migration length of the neutron in the water is increased to about 1/0.74 (i.e. 1.35) times to that in the reactor cold period. Moreover, the density of the steam-water mixture in the occurrence of the voids is lowered to an extent of about 0.3 g/cm.sup.3, and as a result, the thermal neutron diffusion length in the gas-water mixture is increased to 1/0.3 (.apprxeq.3) times. Consequently, the neutron mutual interaction in the fuel areas adjacent to the interposed zone is increased and hence the neutron multiplication characteristics are also increased. According to the functions described above, by introducing the interposed zone into the fuel rod, the effective multiplication factor K.sub.eff is lowered in the reactor cold period, that is, the reactor shut-down margin (subcriticality) is made large, and on the other hand, in the reactor high temperature operation period, the effective multiplication factor K.sub.eff can be prevented from becoming lower even if the fuel amount is reduced by the introduction of the interposed zone, and in a certain case, it may even become possible to increase the effective multiplication factor by suitably designing the fuel rod inclusive of the interposed zone in comparison with that in the case of the fuel rod with no interposed zone. In addition, during the operation period of the BWR, since the void ratio is high at the upper portion of the reactor core, the moderator is insufficient, but according to this invention, the amount of the fuel at the upper portion of the reactor core is reduced, so that the water-to-fuel volume ratio is increased, thus resolving the insufficiency of the moderator. The output power is therefore increased and the power axial distribution can be improved. Moreover, since the water much exists in the upper portion of the reactor core, the void factor (i.e. large load) can be alleviated. The coolant flows upwardly from the lower portion of the reactor core, and the voids are not generated in the lower portion thereof but are generated in the other portion, and particularly, in the upper portion of the reactor core, the void factor is increased. This fact means that the flow rate of the coolant in steam-water mixture state increases largely. Since the pressure drop is usually in proportion to about the square of the coolant flow rate, the pressure loss in the upper portion of the fuel assembly is made large. The pressure loss is varied in accordance with the wetted areas of the fuel rods or channel box and the structure or numbers of the spacers. In this connection, according to this invention, the fuel rod is partially removed at an area upper than a portion having a length of two-third (2/3) to five-sixth (5/6) of the effective fuel length measured from the lower end of the effective portion of the fuel to thereby locate the vanishing zone, so that the number of the fuel rods in the upper portion of the fuel assembly where the pressure loss is made large can be reduced. Accordingly, the wetted area is reduced, thus effectively reducing the pressure loss, and as a result of this, the reduction of the driving power of a recirculation pump is made possible. The stability of the channel box is improved and the flow of the coolant in the channel is also made stable, which results in the reduced variation of the void factor and the improvement of the reactor core stability. Since the thermal neutron flux is increased around the tube through which the moderator passes, the thermal neutron flux increasing region can be widened by arranging a plurality of such tubes in the fuel assembly. The reactivity of the fuel is also increased in the thermal neutron flux increasing region, thus increasing the effective multiplication factor K.sub.eff. The tubes may be preferably arranged symmetrically with respect to the diagonal line of the fuel assembly in cross section to make easy the design thereof. Preferred embodiments of this invention will be further described hereunder in detail with reference to the accompanying drawings.
054901863	claims	1. A shipping container for a nuclear fuel assembly including a plurality of fuel rods and a bottom nozzle; the bottom nozzle including a longitudinally extending recess; said shipping container comprising: support means for supporting the plurality of fuel rods and the bottom nozzle; housing means for housing said support means and said nuclear fuel assembly; and bottom nozzle holding means secured to said support means for holding the bottom nozzle of said nuclear fuel assembly, said bottom nozzle holding means including recess holding means for holding the bottom nozzle within the longitudinally extending recess. screw means for rotating the cam means, said screw means having a head and a shaft, the head abutting the surface of the end support means, the shaft having a non-threaded portion and a threaded portion, the non-threaded portion adjacent the head and passing through the hole of the end support means and also passing through the hole of the base means, the threaded portion adjacent the non-threaded portion and threaded through the threads of the hole of the cam means; a collar fixedly attached to the threaded portion and separated from the cam means; and spring means biased between the cam means and said collar for providing a pre-load force for said screw means. gripper means for gripping the shoulder within the bottom nozzle; and engaging means for engaging said gripper means against the shoulder. 2. The shipping container as recited in claim 1 wherein the recess holding means includes wedge means for wedging against the bottom nozzle within the longitudinally extending recess and moving means for moving the wedge means within the longitudinally extending recess. 3. The shipping container as recited in claim 1 wherein the bottom nozzle further includes an end and a tapered bore which tapers toward the end, and wherein the recess holding means includes gripper means for gripping the tapered bore within the bottom nozzle and moving means for moving the gripper means against the tapered bore. 4. The shipping container as recited in claim 3 wherein the bottom nozzle includes a first barrel, a taper, and a second barrel having a diameter smaller than the first barrel; the taper interconnecting the first barrel and the second barrel which forms the end of the bottom nozzle; and wherein said bottom nozzle holding means further includes spacer means having a hole for inserting the second barrel therein and a tapered surface for abutting the taper in order to space the end of the bottom nozzle from said support means. 5. The shipping container as recited in claim 1 wherein the bottom nozzle further includes a shoulder within the longitudinally extending recess, and wherein the recess holding means includes gripper means for gripping the shoulder within the bottom nozzle and engaging means for engaging the gripper means against the shoulder. 6. The shipping container as recited in claim 5 wherein the bottom nozzle further includes an end; wherein said support means includes end support means for supporting the end; wherein the gripper means includes a plurality of grippers for gripping the shoulder within the bottom nozzle, each of the grippers having a gripping end and a pivot end; and wherein the engaging means includes base means for pivotally mounting the pivot end of each of the grippers and moving means for moving the gripping end of each of the grippers, the moving means including operating means for engaging the moving means which engages each of the gripping ends in order to move the gripping ends toward the shoulder within the bottom nozzle, the operating means also for disengaging the moving means in order to move the gripping ends away from the shoulder within the bottom nozzle, the base means being inserted adjacent the end support means and within the end of the bottom nozzle. 7. The shipping container as recited in claim 6 wherein the moving means includes cam means having a plurality of cam surfaces for camming a corresponding one of the gripping ends of the plurality of grippers. 8. The shipping container as recited in claim 7 wherein each of the plurality of grippers corresponds to one of the plurality of cam surfaces, and wherein adjacent ones of the plurality of grippers include a spring means for forcing each of the adjacent ones of the plurality of grippers against a corresponding one of the plurality of cam surfaces. 9. The shipping container as recited in claim 8 wherein said nuclear fuel assembly has a central longitudinal axis; wherein each of the end support means, the base means and the cam means has a hole which is positioned on the central longitudinal axis, the end support means having a surface, the hole of the cam means having threads; and wherein the operating means includes: 10. The shipping container as recited in claim 9 wherein the cam means further includes blocking means for blocking rotation of the cam means, the blocking means including a plurality of blocking surfaces, each of the blocking surfaces being between adjacent ones of the plurality of cam surfaces in order that each one of the blocking surfaces abuts the corresponding one of the gripping ends of the grippers whenever the cam means is fully disengaged. 11. The shipping container as recited in claim 9 wherein the cam means further includes blocking means for blocking rotation of the cam means, the blocking means including a plurality of blocking tabs, each of the blocking tabs being attached to a corresponding one of the cam surfaces in order that each one of the blocking means abuts the corresponding one of the gripping ends of the grippers whenever the cam means is fully engaged. 12. The shipping container as recited in claim 1 wherein the bottom nozzle has an end, wherein said support means includes end support means for supporting the end, and wherein said bottom nozzle holding means further includes spacer means for spacing the end of the bottom nozzle from the end support means. 13. The shipping container as recited in claim 12 wherein the spacer means is made of ASTM 240, type 304 stainless steel in order to preclude contamination of the bottom nozzle by the end support means. 14. A bottom nozzle holding apparatus for use with a shipping container for a nuclear fuel assembly including a plurality of fuel rods and a bottom nozzle; the bottom nozzle including a longitudinally extending recess, and a shoulder within the longitudinally extending recess; said bottom nozzle holding apparatus comprising: 15. The bottom nozzle holding apparatus as recited in claim 14 wherein said shipping container has a support mounted therein; wherein said gripper means includes a plurality of grippers for gripping the shoulder within the bottom nozzle, each of the grippers having a gripping end and a pivot end; and wherein said engaging means includes base means for pivotally mounting the pivot end of each of the grippers, moving means for moving the gripping end of each of the grippers, the moving means including operating means for engaging the moving means which engages each of the gripping ends in order to move the gripping ends radially and angularly outward toward the shoulder within the bottom nozzle, the operating means also for disengaging the moving means in order to move the gripping ends radially and angularly inward away from the shoulder within the bottom nozzle, the base means being inserted adjacent said support.
	summary
	abstract	A system and method for storing multiple canisters containing high level waste below grade that afford adequate ventilation of the spent fuel storage cavity. In one aspect, the invention is a ventilated system for storing high level waste emitting heat, the system comprising: an air-intake shell forming an air-intake cavity; a plurality of storage shells, each storage shell forming a storage cavity; a lid positioned atop each of the storage shells; an outlet vent forming a passageway between an ambient environment and a top portion of each of the storage cavities; and a network of pipes forming hermetically sealed passageways between a bottom portion of the air-intake cavity and at least two different openings at a bottom portion of each of the storage cavities such that blockage of a first one of the openings does not prohibit air from flowing from the air-intake cavity into the storage cavity via a second one of the openings.
063339617	description	DETAILED DESCRIPTION The invention is described below in the context of specific embodiments. However, it will be understood that the invention is not limited to the specific embodiments. A reflection mask according to a first representative embodiment is shown, in section, in FIG. 1. The reflection mask comprises a multilayer mirror 2 formed on a surface of a substrate 1. An absorber layer 3, defining a prescribed circuit-pattern, is formed on the multilayer mirror 2. As indicated in its name, the multilayer mirror 2 is a reflective laminate of multiple thin layers. The thin layers normally are configured as alternating superposed layers of a first material and a second material. The "thickness period" is the combined thickness, in the thickness dimension of the multilayer mirror, of a layer of the first material and an adjacent layer of the second material. The multilayer mirror 2 does not have a constant thickness period through the thickness dimension of the multilayer mirror 2. Rather, the thickness period changes at least once through the thickness dimension of the multilayer mirror 2. FIGS. 2(b)-2(d) show various exemplary schemes in which the thickness period is changed according to the invention. For comparison purposes, FIG. 2(a) depicts a conventional multilayer mirror having a constant thickness period "d" from the substrate 1 to the surface of the multilayer mirror 2. As noted above, in a reflection mask according to the invention, the thickness period changes at least once with distance through the thickness dimension of the multilayer mirror 2. For example, in FIG. 2(b), the multilayer mirror has two thickness-period structures (each termed a "block"), wherein the laminations of "block 1" have a thickness period of d.sub.1, and the laminations of "block 2" have a thickness period of d.sub.2. In the example shown in FIG. 2(c), the first pair of laminations (i.e., the first layer of the first material and the first layer of the second material) is formed, on the substrate side, at a d.sub.1 thickness period. With formation of each subsequent pair of laminations (up to a total of N pairs of laminations), the thickness period is changed in a progressive manner such that the thickness period on the surface of the multilayer mirror is d.sub.N. In FIG. 2(d), the multilayer mirror has three thickness-period structures (block 1, block 2, and block 3), wherein the laminations in block 1 have a thickness period of d.sub.1, the laminations in block 2 have a thickness period of d.sub.2, and the laminations in block 3 have a thickness period of d.sub.3. The number of blocks can be greater than three; as the number of blocks is increased, the lamination profile more nearly approaches the profile shown in FIG. 2(c). The reflection mask having a three-block profile (FIG. 2(d) exhibits a reflectance spectrum in which the FWHM is intermediate the FWHM of light reflected from the two-block mask (FIG. 2(b)) and the FWHM of light reflected from the mask in which the thickness period changes progressively (FIG. 2(c)). Changing the thickness period, through the thickness dimension, of the multilayer mirror of a reflection mask allows the FWHM of the reflectance spectrum of the mask to be manipulated. Note that, in FIGS. 2(b)-2(d), the thickness period of laminations at the surface of the multilayer mirror 2 is larger than the thickness period of laminations adjacent the substrate 1. Alternatively, the thickness period of laminations nearer the surface side may be smaller than of laminations nearer the substrate side. In a reflection mask according to this embodiment, the greater the difference in thickness period through the thickness dimension of the multilayer mirror 2, the greater the increase in the FWHM of the reflectance spectrum of the multilayer mirror 2; however, reflectance of EUV light from the multilayer mirror 2 generally decreases with such changes. The actual amount of EUV light reaching the wafer is determined by the amount of EUV light, of a predetermined wavelength range, actually transmitted through the optical system including the reflection mask. The amount of EUV light, of the predetermined wavelength range, actually transmitted through the optical system is a function of the amount of EUV light reflected from each mirror of the optical system and from the reflection mask. Hence, according to the invention, within a certain range of wavelengths, the acceptable tolerance range of shifts in center wavelength reflecting from the mask and passing through the optical system can be expanded without causing an excessive decrease in wafer illumination. A representative embodiment of a microlithography system according to the invention is depicted schematically in FIG. 3. The depicted embodiment is a projection-exposure apparatus employing light in the soft X-ray range (i.e., EUV light) as the exposure-illumination light. Projection-imaging is performed using an imaging-optical system 19, which forms a "reduced" (demagnified) image of the pattern defined by the mask 18 on the wafer 10. In FIG. 3, the optical axis of the imaging-optical system 19 extends in the Z-direction, the plane of the page extends in the X- and Z-directions, and the Y-direction is perpendicular to the plane of the page. As noted above, the pattern to be transferred onto the wafer 10 is defined by the reflection-type mask 18, which is mounted on a mask stage 21. The wafer 10 is mounted on a wafer stage 9. Typically, exposure is performed in a step-and-scan manner, wherein the mask pattern is projected in successive portions ("shot regions") while synchronously moving the mask stage 21 and wafer stage 9 relative to each other as exposure progresses. Scanning of the mask 18 and wafer 10 typically is performed in a single dimension (here, the X-direction) relative to the imaging-optical system 19. Upon exposing all the shot regions on the mask 18 onto respective regions of the wafer surface, exposure of the pattern onto a die of the wafer 10 is complete. Exposure then can progress stepwise to the next die on the wafer 10. The soft X-ray beam (i.e., the EUV beam) used as the illumination light for exposure has low transmittance through the atmosphere. Hence, the optical path through which the EUV beam passes desirably is enclosed in a vacuum chamber 20. The vacuum chamber 20 is evacuated using a suitable vacuum pump 101. The EUV beam desirably is produced by a laser-plasma X-ray source comprising a xenon target gas. The laser-plasma X-ray source comprises a laser source 90 (serving as an excitation-light source) and a xenon gas supply 91. The laser-plasma X-ray source is enclosed by a vacuum chamber 50a. The EUV beam produced by the laser-plasma X-ray source passes through a window 50b in the vacuum chamber 50a. The vacuum chamber 50a is separate from the vacuum chamber 20 because debris tends to be generated by a nozzle 12 that discharges the xenon gas. The laser source 90 is configured to generate laser light having a wavelength that can be within the range from infrared to ultraviolet. For example, a YAG laser or excimer laser can be used. The laser light from the laser source 90 is condensed and irradiated onto the stream of xenon gas (supplied from a gas supply 91) discharged from the nozzle 12. Such irradiation of the stream of xenon gas causes heating of the xenon gas sufficiently to form a plasma. Photons of EUV light are emitted as the laser-excited molecules of xenon gas drop to a lower energy state. A parabolic mirror 14 is situated in the vicinity of xenon-gas discharge. The parabolic mirror 14 collects and condenses the EUV light produced by the plasma. The parabolic mirror 14 constitutes herein the condenser optical system, and the parabolic mirror 14 is situated such that its focal point is nearly at the locus of discharge of the xenon gas from the nozzle 12. The parabolic mirror 14 comprises a multilayer film ("multilayer") suitable for reflecting the EUV light. The multilayer typically is provided on the concave surface of the parabolic mirror 14. EUV light reflected from the multilayer passes through the window 50b of the vacuum chamber 50a to a condenser mirror 15. The condenser mirror 15 condenses and reflects the EUV light to the reflection-type mask 18. To such end, the condenser mirror 15 also comprises a surficial multilayer that is reflective to EUV light. EUV light reflected from the condenser mirror 15 illuminates the prescribed shot region on the reflection-type mask 18. As referred to herein, the parabolic mirror 14 and condenser mirror 15 collectively comprise the "illumination system" of the FIG.-3 apparatus. The reflection-type mask 18 is configured with a multilayer EUV-reflective mirror as described above, so further description of the mask 18 is omitted here. As the EUV light reflects from the mask 18, the EUV light becomes "patterned" with pattern data from the mask 18. The patterned EUV light passes through the projection system 19 to the wafer 10. In this embodiment, the imaging-optical system 19 comprises four EUV-reflective mirrors: a concave first mirror 19a, a convex second mirror 19b, a convex third mirror 19c, and a concave fourth mirror 19d. Each of the mirrors 19a-19d comprises a respective multilayer (reflective to EUV light) applied to a backing material (substrate). The mirrors 19a-19b in this embodiment are arranged so that their respective optical axes are coaxial with each other. To prevent obstructing the optical path defined by the respective mirrors 19a-19d, appropriate cutouts are provided in the first mirror 19a, the second mirror 19b, and the fourth mirror 19d. (In FIG. 3, the dashed-line portions of the mirrors indicate the respective cutouts.) An aperture stop (not shown) is provided at the position of the third mirror 19c. The EUV light reflected by the reflection-type mask 18 is reflected sequentially by the first mirror 19a through the fourth mirror 19d to form a reduced image of the mask pattern, based on a prescribed demagnification ratio .beta. (for example .vertline..beta..vertline.=1/4, 1/5, or 1/6) within the respective shot region on the wafer 10. The projection system 19 is configured so as to be telecentric on its image side (wafer side). The reflection-type mask 18 is supported, at least in the X-Y plane, by the movable reticle stage 21. The wafer 10 is supported, desirably in each of the X-, Y-, and Z-directions by the movable wafer stage 9. During exposure of a die on the wafer 10, while EUV light is irradiated to each shot region on the mask 18 by the illumination system, the mask 18 and wafer 10 are moved in a coordinated manner relative to the imaging-optical system 19 at a prescribed velocity according to the demagnification ratio of the imaging-optical system 19. Thus, the mask pattern is scanned progressively and exposed within a prescribed shot range (for a die) on the wafer 10. During exposure, to prevent gases generated from the resist on the wafer 10 from depositing on and adversely affecting the mirrors 19a-19d of the imaging-optical system 19, the wafer 10 desirably is situated behind a partition 11. The partition 11 defines an aperture 11a through which the EUV light can pass from the mirror 19d to the wafer 10. The space defined by the partition 11 is evacuated by a separate vacuum pump 110. Thus, gaseous contaminants produced by irradiation of the resist are prevented from depositing on the mirrors 19a-19d or on the mask 18, thereby preventing deterioration of optical performance of these components. In this embodiment, the parabolic mirror 14, the condenser mirror 15, and the mirrors 19a-19d of the imaging-optical system 19 are each configured with a respective multilayer comprising alternating thin-film layers of molybdenum and silicon. The multilayers, used to reflect EUV light having a nominal wavelength of 13.4 nm, have a thickness period of 6.84 nm. Each multilayer has 50 pairs of laminates each comprising a 2.24-nm thick molybdenum layer and a 4.6-nm thick silicon layer. At 13.4 nm nominal wavelength of EUV light, high reflectance is achieved only within a narrow wavelength range (the full-width at half-maximum, or "FWHM," is 0.55 nm). The nominal peak reflectance of 13.4-nm EUV light from each multilayer is 73.8 percent; however, due to imperfections (e.g., surficial diffusion) arising during manufacture of the multilayers, the actually achieved reflectance is lower, approximately 67 percent. Thus, the exposure apparatus of FIG. 3 comprises six mirrors each including a Mo/Si multilayer. A typical reflection spectrum from one of such mirrors is shown in FIG. 4, showing a nominal reflectance of 73.8% and a FWHM of 0.55 nm. Specifically, the proportion of incident light passing through such an optical system is (0.738).sup.6 =16.2%, as shown in the transmission spectrum of FIG. 5. In FIG. 5, the FWHM of the spectrum is 0.35 nm. If a difference should exist between the center wavelength of reflected light from the reflection mask and the center wavelength of light passing through the 6-mirror optical system, then the quantity of light (percentage of incident light entering the optical system) actually reaching the wafer 10 is even smaller than 11.9%=(0.738).sup.7, including mask reflectance. The larger the number of reflection mirrors used in the optical system, the larger the value of a ratio of the percent of incident light passing through the optical system to the magnitude of the difference in center wavelengths. In view of the above, whenever a reflection-type mask 18 is used with an optical system comprising multiple reflection mirrors each including a multilayer, it is highly desirable to reduce as much as possible any shifts in center wavelengths of reflected light from all portions of the multilayer mirror of the mask 18. If there are any non-uniformities in the plane at the thickness period of the multilayer mirror of the reflection mask 18, then illumination non-uniformities will occur at the wafer 10. In an illumination non-uniformity, the quantity of light reaching the wafer 10 (i.e., "exposure dose") varies with location on the exposed portion of the wafer. The linewidth of a circuit pattern transferred onto a particular locus of the wafer surface is a function of the exposure dose at that locus (i.e., the total amount of light energy incident on the locus). I.e., as the exposure dose changes, the linewidth changes. To prevent linewidth changes arising from changes in exposure dose, it is desirable that illumination non-uniformities on the wafer be limited to .+-.1 to .+-.2 percent or less. FIG. 6 depicts a representative change in illumination on the wafer under a condition in which there has been a shift in the center wavelength of light passing through the optical system (having the transmittance spectrum shown in FIG. 5) relative to the center wavelength of light reflected from a reflection mask exhibiting the reflection spectrum shown in FIG. 4. In FIG. 6, if the shift in center wavelength is zero (i.e., no shift in center wavelength), then the wafer receives an amount of illumination that is 100% of nominal. FIG. 6 shows two curves. The curve representing a plot of square points corresponds to wavelength shifts to longer wavelengths, and the curve representing a plot of "x" points corresponds to wavelength shifts to shorter wavelengths. As shown in FIG. 6, in order to limit changes in wafer illumination to within 1 percent (100%-99%), it is desirable that wavelength shifts be no greater than 0.033 nm, and that the ratio of the magnitude of center-wavelength shift (0.033 nm) to the center wavelength (13.4 nm) be no greater than 0.25 percent. Similarly, in order to limit changes in wafer illumination to within 2 percent, it is desirable that wavelength shifts be no greater than 0.046 nm, and that the ratio of the magnitude of center-wavelength shift (0.046 nm) to the center wavelength (13.4 nm) be no greater than 0.34 percent. Therefore, it is especially desirable that the shift in center wavelength over the surface of the reflection mask be no greater than 0.34 percent at the wavelength used by the microlithography apparatus to make an exposure. As noted above, less than the entire mask (i.e., less than an entire die on the wafer) normally is exposed at any one instant. Typically, the area of the wafer that is exposed at any one instant (i.e., the exposure-field size) is approximately 26.times.33 mm at a demagnification ratio .beta.=1/4. Under such conditions, the size of the exposure-field on the mask is approximately 104.times.132 mm. Over such an exposure field, it is normally very difficult to have the multilayer mirror on the mask exhibit variations in thickness period within a range of 0.34 to 0.25 percent. To counteract this problem, according to the present embodiment, the FWHM of the reflectance spectrum of the reflection mask 18 is increased. This reduces the ratio of the change in illumination (on the wafer) to the shift in center reflection wavelength (at respective positions on the reflection mask). Hence, even if the magnitude of shift of center wavelength of reflected light is increased, the ratio of the change in reflectance to the change in wavelength remains small, thereby reducing the resulting change in illumination on the wafer. According to the invention, an increase in FWHM of the reflectance spectrum of the reflection mask can be achieved by forming the multilayer mirror of the mask using alternating layers of molybdenum and silicon configured as, for example, two blocks (see FIG. 2(b)). An example reflectance spectrum of such a reflection mask is shown in FIG. 7. In this reflection mask, the difference in thickness period between the two blocks is 0.4 nm. In FIG. 7, even though the peak reflectance is slightly lower (67 percent) than the peak reflectance (74%) shown in FIG. 4, the FWHM in the FIG.-7 spectrum is substantially larger (0.77 nm compared to 0.55 nm in FIG. 4). The beneficial effect of using the FIG.-7 mask is illustrated in FIG. 8. Specifically, FIG. 8 depicts a representative change in illumination on the wafer under a condition in which there has been a shift in the center wavelength of light passing through the 6-mirror optical system (having the transmittance spectrum shown in FIG. 7) from the two-block reflection mask. In FIG. 8, if the shift in center wavelength is zero, then the wafer receives an amount of illumination that is 100% of nominal. FIG. 8 shows two curves. The curve representing a plot of square points corresponds to wavelength shifts to longer wavelengths, and the curve representing a plot of "x" points corresponds to wavelength shifts to shorter wavelengths. As shown in FIG. 8, limiting changes in wafer illumination to within 1 percent can be obtained with wavelength shifts within a tolerance of no greater than 0.057 nm (compare to 0.034 nm in FIG. 6). Also, the ratio of the magnitude of center-wavelength shift (0.057 nm) to the center wavelength (13.7 nm) is within 0.43 percent (compare to 0.25 percent in FIG. 6), representing an approximately 70 percent improvement. The tolerance with respect to shifts of the thickness period in the direction within the plane of the reflection mask is correspondingly larger. Similarly, limiting changes in wafer illumination to within 2 percent can be obtained with wavelength shifts within a tolerance of no greater than 0.082 nm; the ratio of the magnitude of center-wavelength shift (within the plane of the reflection mask) to the center wavelength for which the mask is designed is now within 0.61 percent (compare to 0.34 percent in FIG. 6), also representing a 70-percent increase. Therefore, whenever a reflection mask including a multilayer mirror divided into two blocks with different respective thickness periods is used as described above, and the reflection mask has experienced a drop in peak reflectance, the resulting drop in illumination on the wafer is only approximately 10 percent compared to a much higher drop when a reflection mask is used having a multilayer mirror with only one thickness period. (I.e., total reflectance determined from FIGS. 4 and 5, compared to FIGS. 5 and 7, yields a 10% difference.) FIG. 9 shows a reflection spectrum obtained with a reflection mask comprising a multilayer mirror made of alternating layers of molybdenum and silicon, in which the thickness period progressively changes through the thickness dimension (see FIG. 2(c)). In the mask used to produce FIG. 9, the difference in the thickness period between the substrate side and the surface side is 0.4 nm. With this mask, the peak reflectance is 72.3 percent (FIG. 9), and the FWHM is 0.65 nm (compare to 0.55 nm in FIG. 4). The beneficial effect of using the FIG.-9 mask is illustrated in FIG. 10. Specifically, FIG. 10 depicts a representative change in illumination on the wafer under a condition in which there has been a shift in the center wavelength of light passing through the 6-mirror optical system (having the transmittance spectrum shown in FIG. 9) from the reflection mask. In FIG. 10, if the shift in center wavelength is zero, then the wafer receives an amount of illumination that is 100% of nominal. FIG. 10 shows two curves. The curve representing a plot of square points corresponds to wavelength shifts to longer wavelengths, and the curve representing a plot of "x" points corresponds to wavelength shifts to shorter wavelengths. As shown in FIG. 10, limiting changes in wafer illumination to within 1 percent can be obtained with wavelength shifts within a tolerance of no greater than 0.044 nm (compare to 0.034 nm in FIG. 6). Also, the ratio of the magnitude of center-wavelength shift to the center wavelength is within 0.33 percent (compare to 0.25 percent in FIG. 6), representing a 30-percent increase. Similarly, limiting changes in wafer illumination to within 2 percent can be obtained with wavelength shifts within a tolerance of no greater than 0.062 nm; the ratio of the magnitude of center-wavelength shift (within the plane of the reflection mask) to the center wavelength for which the mask is designed is now within 0.46 percent (compare to 0.34 percent in FIG. 6), again representing an increase of approximately 30 percent. Furthermore, drops in illumination on the wafer are hardly apparent at all using this mask. FIG. 11 depicts certain relationships of the increase in the tolerance of shifts in center wavelength and the reduction in illumination on the wafer as obtained with two types of reflection mask according to the invention: (a) a mask in which the multilayer mirror is formed by a two-block laminated structure with different thickness periods, and (b) a mask in which the multilayer mirror is formed by a laminated structure in which the thickness period changes in a progressive manner through the thickness dimension. In FIG. 11, the reflection mask in which the multilayer mirror is formed by laminated molybdenum/silicon layers having fixed thickness periods is regarded as providing a wafer illumination of 100 percent. The two curves shown in FIG. 11 represent the relationship between the tolerance of shifts in center wavelength at any location on the reflection mask and the corresponding illumination on the wafer whenever the tolerances of illumination non-uniformity are .+-.1 percent and .+-.2 percent, respectively. The upper right position in this graph is desirable because, the farther one moves toward the upper right of this graph, the more the illumination on the wafer increases and the greater the increase in the tolerance of wavelength shifts. With the respective curves A.sub.1 and A.sub.2 as starting points, regions in the direction of arrow 100a or arrow 100c reflect the relationship in respective cases in which the thickness periods of the multilayer mirror on the mask have been changed in a progressive manner. (In the figure, these are the portions denoted by "G".) Situations are shown in which the difference in the thickness periods on the substrate side and on the surface side become larger with movement along the curve from A.sub.1 or A.sub.2 in the direction of the arrow 100a or 100c, respectively. On the other hand, with the respective curves A.sub.1 and A.sub.2 as starting points, regions in the direction of arrow 100b or arrow 100d reflect the relationship in respective cases between illumination onto the wafer and the tolerance of shifts in center wavelength whenever the mask has a multilayer mirror having a two-block configuration with different respective thickness periods. (In the figure, these are the portions denoted by "D".) As one moves farther on the curve from A.sub.1 or A.sub.2 in the direction of the arrow 100b or 100d, respectively, the difference in the thickness periods of the two blocks becomes larger. The locations of the respective curves A.sub.1 or A.sub.2 pertain to a mask having a multilayer mirror in which the thickness period is fixed. A mask having a two-block multilayer mirror with two different thickness periods has a different behavior than a mask having a multilayer mirror in which the thickness period changes in a progressive manner through the thickness dimension. In situations in which the tolerance of the shift in center-wavelength of reflected light from various locations on the reflection mask is relatively low, using a reflection mask having a multilayer mirror in which the thickness period progressively changes would exhibit almost no drop in reflectance. It is thus possible to increase the FWHM of the reflectance spectrum, with substantial effect. In situations in which the tolerance of the shift in center-wavelength of reflected light from various locations on the reflection mask is relatively large, it may be more efficient to use a reflection mask having a multilayer mirror having a two-block structure, wherein each block has a different thickness period. With FIG. 11, it is possible to consider the reduction in wafer illumination with an increase in tolerance of wavelength shifts, and to select optimal parameters. A microlithographic exposure apparatus that uses a reflection mask according to the invention exhibits substantial advantages when used to manufacture integrated circuits and the like. In this regard, an exemplary semiconductor integrated-circuit manufacturing process is summarized below that uses such a microlithographic exposure apparatus. The process comprises a wafer-preparation step, a mask-preparation step, at least one wafer-processing step, a chip-assembly step, and a chip-inspection step. The wafer-processing step comprises a thin-film-formation step, an oxidation step, a resist-coating step, an exposure step, an etching step, an ion-implantation step, a resist-stripping step, and a washing step. Whenever semiconductor-device manufacturing is performed using a microlithographic exposure apparatus, according to the invention, to perform the exposure step, it is possible to transfer linewidths onto the wafer with high accuracy and to increase the yield of acceptable product. The invention is further described in the context of examples. It will be understood that the examples are not limiting in any way. EXAMPLE 1 To manufacture a reflection mask, a silicon substrate was used having a diameter of 8 inches and a thickness of 10 mm. The surface of the silicon substrate was polished to a planarity within 0.2 nm (surface roughness RMS value). Ion-beam sputtering was used to laminate thin films of molybdenum (Mo) and silicon (Si) alternately on the polished surface of the substrate, thereby forming a surficial multilayer mirror. The thickness period of the first 25 layer pairs was 6.5 nm. A second group of 25 layer pairs had a thickness period of 6.9 nm. To improve the planar uniformity of the film-thickness distribution for the layers, layer formation was performed while using a film-thickness-compensation plate immediately in front of the substrate. See, Murakami et al., Applied Optics 32:7057-7061, 1993, especially page 7059 and FIG. 3. This reference is incorporated herein by reference. In this way, alternating layers of silicon and molybdenum (50 layer pairs in total) were formed in which the difference in thickness period between the first block and the second block was only 0.4 nm. The uppermost layer of the multilayer mirror was a molybdenum layer. Measurements revealed that this multilayer mirror exhibited a reflection characteristic nearly identical to that shown in FIG. 7. At this stage, the in-plane uniformity of the center wavelength of light reflected from the multilayer mirror was evaluated. To produce an illumination uniformity on the wafer of no more than .+-.1 percent, as shown in FIG. 8, the tolerance of the center-wavelength distribution was 13.4 nm .+-.0.057 nm (i.e., within .+-.0.43 percent). By way of comparison, a conventional multilayer mirror comprising alternating layers of molybdenum and silicon having a fixed thickness period exhibited a center-wavelength distribution in which the tolerance was 13.4 nm .+-.0.033 nm (i.e., within .+-.0.25 percent). Hence, a multilayer mirror according to this example exhibited a tolerance of center-wavelength distribution that was approximately 70 percent greater than a conventional multilayer mirror. As a result of the achieved wider tolerance, the yield of acceptable product was improved substantially over conventional methods. Formation of an EUV-absorbing layer was performed using an electrolytic plating technique, as disclosed in Murakami et al., Jpn. J. Appl. Phys. 34:6696-6700, 1995. After applying a layer of photoresist onto the multilayer mirror, the desired circuit pattern was drawn on the resist using an electron beam. After development of the resist, and using remaining resist as a mask, a 100-nm-thick EUV absorber layer of nickel was applied by electrolytic plating. The remaining resist was stripped to complete manufacture of a reflection mask such as that shown in FIG. 1. EXAMPLE 2 The surface of a 230-mm square glass substrate (ULE low-thermal-expansion glass made by Corning, 20-mm thick) was polished to within 0.2-nm planarity (surface roughness RMS value). High-frequency magnetron sputtering was used to apply alternating layers of molybdenum (Mo) and silicon (Si) on the polished surface, thereby forming a multilayer mirror comprising alternating layers of molybdenum and silicon. Fifty pairs of layers were applied. The thickness period of the first pair of layers (one layer of Mo and one layer of Si) was 6.68 nm. The thickness period was increased in a progressive manner with each subsequent pair of layers, wherein the thickness period of the uppermost pair of layers 7.08 nm. Thus, the increase in thickness period from the first to the 50th layer pair was 0.4 nm. To improve the uniformity of the layer-thickness distribution over the surface of the substrate, film formation was performed while using a film-thickness-compensation plate situated immediately in front of the substrate. The uppermost layer was a silicon layer. Measurements revealed that this reflective member exhibited a reflection characteristic that was nearly the same as shown in FIG. 9. The in-plane uniformity of the reflection center wavelength of the multilayer mirror was evaluated. To ensure an illumination non-uniformity on the wafer of .+-.2 percent, as shown in FIG. 10, the tolerance of the center-wavelength distribution was within 13.4 nm .+-.0.062 nm (i.e., within .+-.0.46 percent). Hence, compared to a conventional molybdenum/silicon multilayer having a fixed thickness period, the center wavelength distribution of this example exhibited a tolerance of 13.4 nm .+-.0.046 nm (within .+-.0.34 percent). Hence, the tolerance of center-wavelength distribution was increased by approximately 30 percent relative to the conventional reflective member. Such an increase in the tolerance produces a corresponding increase in the yield of the manufacturing process for making such reflective members. Application of an EUV-absorbing layer to the multilayer mirror of this example was performed by dry etching as described in Oshino et al., Japanese Kokai Patent Application No. Hei 52-185739. Specifically, high-frequency magnetron sputtering was used to form a 10-nm thick intermediate carbon layer on the multilayer of molybdenum and silicon. Then, a 100-nm thick layer of tungsten (W) was formed. The intermediate layer of carbon prevents damage to the multilayer during a subsequent etching of the absorber (W) layer. A photoresist was applied to the W layer and patterned using electron-beam drawing. The resist was developed. Dry etching of the W layer was performed using an RIE (Reactive Ion Etching) apparatus and a reaction gas mainly comprising CF.sub.4. Etching of the C layer was performed using oxygen (O.sub.2) gas. Lastly, the photoresist was stripped to complete formation of a reflection mask such as that shown in FIG. 1. Therefore, according to the present invention, the tolerance of the in-plane distribution of center wavelength of reflected EUV light from a multilayer used in the reflection mask is increased sufficiently to substantially improve the manufacturing yield and reduce the manufacturing cost of reflection masks as used in microlithography. Whereas the invention has been described in connection with representative embodiments and examples, it will be understood that the invention is not limited to those embodiments and examples. On the contrary, the invention is intended to encompass all alternatives, modifications, and equivalents as may be included within the spirit and scope of the invention, as defined by the appended claims.
050705199	description	DETAILED DESCRIPTION In a preferred but non-limiting example of the invention, the system is set up for an x-ray exposure by sliding an x-ray source/modulator assembly 10 vertically (in the y-direction) along column 12 to align it with the desired patient field. When energized, assembly 10 generates a fan shaped beam 14 which propagates in the z-direction and is thin in the vertical (y) direction and wide in the horizontal (x) direction and sweeps this beam 14 vertically (either up or down) across an object position 16. Also prior to the x-ray procedure, a detector/film assembly 18 is vertically aligned with source/modulator assembly 10 by sliding it along column 20. As fan shaped beam 14 sweeps vertically across object 16, assembly 18 receives an object-attenuated beam both at a detector 52 and at an x-ray film cassette 54 (FIG. 4). As seen in FIG. 1A, beam 14 can be thought of as being a beam 14a when it emerges from x-ray tube 22, then as a beam 14b after being shaped by collimators 28 and 30 and, if desired, by modulators 32 as well, and as a beam 14c after it is attenuated by an object (if any) at object position 16. In a single-scan mode, detector 52 (FIG. 4) generates a feedback signal to control modulator pins 32 (FIGS. 2 and 3a-3c) which modulate respective sectors of beam 14a in order to reduce the local variation in the exposure which object-attenuated beam 14c delivers at the image plane of film cassette 54. In a dual-scan mode, beam 14a in the first scan is at low intensity and preferably is not modulated, and detector 52 generates a two-dimensional map of the exposure delivered by object-attenuated beam 14c at the detector plane but the x-ray film at cassette 54 does not receive significant exposure. In the second scan, beam 14a is at higher intensity and is locally modulated on the basis of the 2D map derived in the first scan to equalize the exposure which object-attenuated beam 14c delivers at the image plane everywhere except at a selected field. As an alternative, the exposure at the selected field can also be equalized but in a different way from that in the other fields of the image, for example by compressing the exposure range less in the selected field. As best seen in FIGS. 2 and 3a-3c, source/modulator assembly 10 comprises an x-ray tube 22 rotated in the vertical plane by source rotation drive 24 in order to sweep object 16 with fan shaped 14b, using the focal spot of tube 22 as the center of rotation. Affixed to tube 22 to pivot therewith is a modulator assembly 26. In order to shape the radiation from tube 22 into the desired fan shaped 14b, assembly 26 has a fixed collimator 28 defining the maximum horizontal extent of the fan shaped and a patient field collimator 29 which can be adjusted to define the size of the irradiated, typically rectangular area at the image plane, and further has an adjustable collimator 30 forming a collimator slit that determines the vertical dimension of fan shaped beam 14b. For a given vertical sweep of fan shaped beam 14c across detector/film assembly 18, the setting of patient field collimator 29 typically is fixed so that fan shaped beam 14c at the film plane irradiates only a field of a desired size and shape, e.g., standard x-ray film. The slit aperture of collimator 30 is set depending on factors such as the overall size and expected attenuation properties of the object to deliver the desired overall intensity to the object. For example, the slit aperture can be set to a vertical dimension in the range of 0" to 0.5" at the aperture plane, which corresponds to about 0" to 3.5" at the image plane. Beam 14c exiting the object impinges on detector/film assembly 18 which, as seen in FIG. 4, comprises the following components arranged in the propagation (z) direction of the x-rays: an anti-scatter grid 50, a detector 52, and a film cassette 54. These elements are mounted on a supporting arm 56 slidably mounted on column 20. Grid 50 can comprise a 12:1 scatter rejection grid for reducing the amount of scattered radiation reaching detector 52. Detector 52 can comprise a flat plate gas ionization detector having an active volume of, e.g., 17.times.17.times.0.25" filled with an ionizing gas such as Xenon. The electrodes on one side are 70 vertically extending strips 0.23" high and 17" long, separated horizontally by insulating spaces of 0.02". Preferably detector 52 attenuates fan shaped beam 14c as little as possible, e.g., at about 12% attenuation. Film cassette 54 can be a standard 14.times.17" cassette mountable in either orientation. In order to modulate horizontally spaced sectors of fan shaped beam 14, modulator assembly 26 uses 35 modulator pins 32 which are in generally horizontal inner row 34 and outer row 36 (FIG. 3b) along respective arcs which are centered at the focal spot of tube 22 and are in a plane that includes the focal spot. Pins 32 slide individually and selectively into fan shaped beam 14a in the vertical, y-direction (vertically along the plane of the paper in FIG. 3a and normal to the paper in FIGS. 3b and 3c) such that the pin section in a horizontal plane within fan shaped beam 14a is generally triangular, as illustrated in FIGS. 3b and 3c. This triangular area and the attenuation of a respective beam sector increases as a pin moves further into fan shaped beam 14a. In the dual-scan mode, where the purpose of the first scan is to derive information for controlling the second scan, beam 14b for the first scan is collimated into a fan shaped beam of the same general shape but preferably is not modulated, and is kept at a low intensity which is sufficient for adequate signal-to-noise ratios at detector 52 but insufficient to cause significant exposure of the film in cassette 54. The desired low intensity can be selected by controlling x-ray tube 22, e.g., by controlling the power supplied to tune 22. It is possible to have some intensity variations in the fan shaped beam during the first scan, so long as they are known or can be measured so that they can be taken into account in deriving the information for controlling the second scan. To derive the information for controlling the second scan, the system uses the signal from detector 52 to generate a two-dimensional map of the transmission of the object at position 16, and from that derives control parameters such as when and how to move each modulator pin 32 during the second scan to achieve the desired selective equalization in which the image is equalized everywhere except at a selected field or is equalized differently at the selected field. The two-dimensional transmission map from the first scan is determined as follows. Each of the 70 horizontal detector strips "j" of detector 52 outputs signals "d.sub.ij " during the first scan, where "i" identifies the position of a scan line (i.e., a particular footprint of the wide fan shaped beam 14c on detector 52 over a particular time interval during the first scan) and "j" identifies a particular detector strip of detector 52. If "S" is the x-ray intensity incident on the slit aperture defining beam 14 (and on any modulator pins 32 which are in beam 14), "m.sub.ik " is the transmission of the k-th modulator pin 32 (a value between 0 and 1), and "b.sub.ijk " is the amount of signal in the j-th detector strip due to the k-th modulator pin 32 when the pin is fully away from fan shaped beam 14 (a value between 0 and 1), the transmission "t.sub.ij " of the patient body at object position 16 as measured at the i-th scan line and the j-th detector can be represented as: ##EQU1## The values of "d.sub.ij " used in expression 1 are those after the detector strip outputs are processed for gain and offset variations and, if desired, filtered for noise. The gain correction is derived through calibration runs with the modulator pins 32 fully retracted (out of beam 14), and the offset correction is derived by measuring the outputs of the detector strips with x-ray tube 22 off. The values of "m.sub.ik " are known by keeping track of the position to which the k-th modulator pin is driven for the i-th scan line and from pre-stored information of the value "m.sub.ik " for each position of each pin 32. The "b.sub.ijk " values are derived in a calibration run in which the modulator pins 32 are fully retracted out of beam 14 one at a time and beam 14 is scanned and the detector strip outputs "d.sub.ij " are stored. The system therefore can calculate the 2D map "t.sub.ij " of the patient's transmission function immediately after the first scan in the dual-scan mode or even in real-time while the first scan is in progress. The reason for the 2D map "t.sub.ij " and the use to which the map is put can be explained by reference to FIGS. 5-9. FIG. 5 illustrates a typical Gamma function of x-ray film, showing the non-linear relationship between exposure level "e" delivered to the x-ray film and optical density "OD(e)" of the x-ray image in a typical film/screen combination. An average level "e.sub.av " of exposure delivered to the x-ray film produces a desirable optical density OD(e) of, e.g., approximately 1.7 for the lung field. Optical densities OD(e) of 2.2 and 0.7 translate into desired maximum and minimum exposure levels "e.sub.max " and "e.sub.min " and can be considered the limits of the exposure delivered to the x-ray film at which images are diagnostically useful. Different film/screen combinations imply different sets of values for the exposure levels e.sub.av, e.sub.max and e.sub.min but the same general principles apply. In non-equalized radiography, when the exposure is set correctly for the lung field other areas of the chest would not be imaged in a diagnostically useful way if they are outside the e.sub.max and e.sub.min limits. The relationship between film/screen exposure and patient body transmission is illustrated in FIG. 6, where if the transmission of the lung field is between t.sub.1 and t.sub.2 with an average of t.sub.av, the required corresponding exposures e.sub.1 and e.sub.2 and e.sub.av can all be within the limits of e.sub.min and e.sub.max of FIG. 5. However, the minimum and maximum of the patient body transmission, t.sub.min and t.sub.max in this case translate to exposures e.sub.min ' and e.sub.max ' delivered to the film, which are outside the limits e.sub.min and e.sub.max of FIG. 5 and therefore are outside the optical density range OD(e) for diagnostically useful images. By locally varying the exposure delivered to the object to increase it at thick parts of the object to raise e.sub.min ' and to reduce it at thin parts of the object to lower e.sub.max ', equilization can in effect compress the range e.sub.min ' to e.sub.max ' of the exposure delivered to the film to make it fit within the limits e.sub.min and e.sub.max, as illustrated in FIG. 7. While such equalization could be applied throughout the range of e.sub.min ' to e.sub.max ', as in the prior art, the invention takes a different approach in an effort to provide further improvement: it equalizes some but not all areas of the image. As an alternative, it equalizes different fields of the image differently. As seen in FIG. 8, an equalization function in accordance with an example of the invention shows no equalization in the patient transmission range from t.sub.1 and t.sub.2 which corresponds to the lung field. FIG. 8 shows equalization only outside the lung field. According to FIG. 8, the exposures delivered to the patient and to the film/screen exposure are forced up in the range between t.sub.min and t.sub.1 and down in the range between t.sub.2 and t.sub.max. The equalization function M(t) of FIG. 8 can be normalized to that illustrated in FIG. 9 to make it corresponds more directly to modulation positions of modulator pins 32. The points t.sub.min, t.sub.1, t.sub.av, t.sub.2 and t.sub.max shown in FIG. 9 are all derived from the 2D map t.sub.ij discussed above. Known types of smoothing and boundary identification are used to define a coarse mapping of the patient including the lung areas and the mediastinum. From the transmission values t.sub.ij in the so-identified lung areas, the upper and lower transmission limits t.sub.1 and t.sub.2 are identified by a known type of a search process and the average value t.sub.av is calculated from the so-identified limits t.sub.1 and t.sub.2. The largest value of transmission in the so-identified mediastinum is t.sub.max. To find t.sub.min, a histogram of the values t.sub.ij is constructed which shows a clear demarkation between air and soft tissue and therefore a suitable t.sub.min. For the second scan, the system selects an exposure level "S" which is the exposure level at modulator assembly 16 as determined by factors such as the anode current of x-ray tube 22, the fan shaped beam collimation and the exposure time. For the minimum patient body transmission t.sub.min, "S" is defined by the ratio (e.sub.min /t.sub.min) when all of modulator pins 32 are fully retracted. "S" is calibrated by relying on the relationship e=S when modulator pins 32 are fully retracted (i.e., when m=1). "S" is a term in the approximation e=M(t)t=Sm(t)t, where "e" is the exposure at the image plane, "S" is the entrance exposure at modulator assembly 16, "m(t)" is the transmission of a modulator pin (where M(t) ranges from m.sub.min and 1, and m.sub.min is the smallest transmission value attainable by modulator assembly 16), "t" is the patient's body transmission, and "" denotes multiplication. When derived in accordance with said approximation, "S" represents the maximum exposure level at modulator assembly 16 which is required to create an equalized exposure e.sub.min at the film point imaging the part of the body having the minimum transmission t.sub.min, for example the densest part of the mediastinum. In accordance with the example of the invention discussed here, the exposure should be substantially constant over the lung field; hence, the equalization should have a constant value "m.sub.av " for patient body transmission values corresponding to the lung field. The desired average exposure at the lung field is "e.sub.av " which corresponds to the desired optical density OD(e.sub.av) of 1.7 for the exemplary screen/film combination discussed here. The modulator transmission "m.sub.av " required to reduce the exposure S leaving modulator assembly 16 to the level "e.sub.av " is m.sub.av =e.sub.av /(t.sub.av S). If "m.sub.av " is greater than the minimum transmission possible with the modulator, "m.sub.min ", then the modulator has sufficient dynamic range to satisfactorily equalize the patient's attenuation, including actually reducing exposure in areas where the film would be more exposed than at the lung field. For the typical case, the dynamic range of 5:1 (which pins 32 have) is expected to be more than enough to equalize the patient transmission field. If "m.sub.av " is less than "m.sub.min ", then the modulator does not have enough dynamic range for full equalization. In that case, the factor "S" is determined by a different relationship, namely, S=e.sub.av /(t.sub.av m.sub.min). This relationship forces the exposure level at the lung field to be at the desired level when the modulator is at the minimum transmission m.sub.min, thus using all of the available dynamic range to equalize the mediastinum. The modulation function m(t) for the second scan in a preferred embodiment of the invention is defined by the following expression, which corresponds to the curve illustrated in FIG. 9: EQU m(t)=(a/t)+b (2) where ##EQU2## The shape of the modulation function can be modified in specified regions by adding a gain parameter "q" such that m(t)=(a/t.sup.q +b). This modulation function m(t) is specified for each modulator pin 32 for each scan line (where the term scan line denotes the position of fan shaped beam 14c over a short interval of time during the second scan). The system keeps track of which transmission values are associated with which modulator pin 32 because the incidence of the beam sectors defined by the respective modulator pins are specified by the mapping "b.sub.ijk " discussed earlier. This information has been previously processed by function fitting to find the actual position of the peak of each sector of beam 14c at detector 52 as a function of scan line "i", and this may be represented as an array "p" of elements "p.sub.ik " which contains the beam sector positions at the detector plane as a function of the scan line "i" and modulator pin "k" following the derivation of the mapping "b.sub.ijk " discussed earlier. The actual position of the peak of each sector "p.sub.ik " for pin "k" is used to interpolate the vector of transmission values for scan line "i" of the second scan determined using expression (1) set forth above, to calculate the transmission value "t" associated with pin "k". Each modulator pin 32 (pin "k") is then set to a new position corresponding to the appropriate modulation value "m.sub.ik ", as determined in accordance with the modulation function as in expression (2) set forth above. FIG. 10 illustrates certain main components of a exemplary system embodying the invention. Components 10, 16 and 18 are as described above. Component 58 is a control circuit Which operates as described above to cause the system to carry out a first scan and derive the parameters for controlling the second scan in the manner described above, and to cause the system to carry the second scan modulating everywhere except over the lung field in the manner described above. The derivation of relevant parameters discussed above is carried out by computer circuits programmed to carry out the calculations described in detail above. The input signals for the calculations are derived using known types of transducers and are converted to digital form using known techniques. The results of the calculations are converted to suitable control signals using known techniques. As one alternative to the currently preferred embodiment, some equalization can be allowed in the selected field, different from the equalization carried out elsewhere in the same image. For example, the modulation function in the lung field can be less steep than for other fields in the image. Another alternative is to use a pre-scan flash exposure and a two-dimensional detector array, which is capable of giving a 2D map of transmission without scanning, in place of the strips of detector 52, to derive the initial patient's transmission function "t.sub.ij " from this initial flash exposure rather than from a first scan, and to use the results to control the modulator pins in a scan corresponding to the second scan as discussed in detail above. In the last alternative, the cumulative exposure of the initial flash and the scan in the lung field should add up so a sufficient exposure level to image the lung field well but the relative contribution of the initial flash and the scan in the lung field can be divided as desired.
046637745	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to X-ray diagnostics installations, and in particular to a support plate for an examination subject for such an installation. 2. Description of the Prior Art X-ray diagnostic installations are known for the production of mammography exposures having a support plate for supporting the patient's breast, the support plate having a receptacle chamber for receiving an X-ray film cassette, and a secondary radiation grid overlying the receptacle such that the grid is disposed between the examination subject and the cassette. A compression means may also be provided above the support plate. In such conventional installations, the secondary radiation grid attenuates the secondary radiation emerging from the examination subject, however, has the disadvantage of increasing the distance between the subject and the film in the cassette. Additionally, in some instances it is preferable to make exposures without the use of a secondary radiation grid. SUMMARY OF THE INVENTION It is an object of the present invention to provide an X-ray diagnostic installation having a support plate with a secondary radiation grid which can be optionally positioned with the secondary radiation grid disposed in front of or behind the X-ray film cassette contained in the support plate. The above object is inventively achieved in a support plate for an X-ray diagnostics installation which is mounted so as to be rotatable about an axis which is perpendicular to a central ray of the X-ray beam. The secondary radiation grid is disposed on one side or surface of the support plate, and the X-ray cassette is contained in a chamber with the secondary radiation grid overlying the chamber on one side. The support plate can be rotated so as to dispose the secondary radiation grid over or in front of the X-ray cassette in the direction of beam propogation, such that the secondary radiation grid is between the examination subject and the cassette, and can be rotated such that the secondary radiation grid is beneath or behind the cassette in the direction of beam propogation. X-ray exposures without the use of the secondary radiation grid can thus be produced in a simple manner by pivoting or rotating the support plate. The secondary radiation grid is relatively mechanically sensitive, and need not be removed from the support plate, as in conventional installations, in order to make this type of exposure. In a further embodiment of the invention, a radiation detector for providing a signal to an automatic exposure control unit may be disposed at the side of the support plate facing away from the radiation grid. The radiation detector thus receiving the radiation dose following the X-ray film in the direction of beam propogation. When an X-ray film cassette is inserted into the chamber for exposure without the secondary radiation grid, in which case the support plate is rotated such that the secondary radiation grid is disposed behind the X-ray film cassette in the direction of beam propogation, the exposure time can also be controlled by the automatic exposure unit. If a specific side of the radiation detector must be exposed to the X-ray beam, the radiation detector may be mounted so as to be rotatable about an axis which is substantially parallel to the axis of revolution of the support plate, or is coincident therewith, so that the necessary side of the detector can be rotated so as to have the X-ray beam incident thereon.
	summary
	claims	1. A molybdenum-100 target assembly comprising:a sintered molybdenum-100 disc;a target holder provided with a recess having a flat surface for receiving therein the sintered molybdenum-100 disc, the target holder comprising a dispersion-strengthened copper composite; andan intermediate layer comprising a brazing alloy of copper and phosphorus therebetween,wherein the intermediate layer is engagingly brazed in between the sintered molybdenum-100 disc and the flat surface of the recess in the target holder. 2. A method of making a molybdenum-100 target assembly, comprising:preparing a pressed molybdenum-100 disc;sintering the pressed molybdenum-100 disc;brazing the sintered molybdenum-100 disc into a recess provided in a target holder,thereby producing the molybdenum-100 target assembly of claim 1. 3. The method of claim 2, wherein the step of preparing the pressed molybdenum-100 disc comprises:placing a selected amount of a molybdenum-100 powder into a cylindrical tool and die set, andapplying a selected pressure thereto for at least 30 sec. 4. The method of claim 3, wherein the selected amount of molybdenum-100 powder is selected from a range of 0.3 g to 3 g. 5. The method of claim 3, wherein the selected amount of molybdenum-100 powder is 1.6 g. 6. The method of claim 3, wherein the selected pressure is selected from a range of 2,000 lbs to 100,000 lbs. 7. The method of claim 3, wherein the selected pressure is 30,000 lbs. 8. The method of claim 2, wherein the step of sintering the pressed molybdenum-100 disc comprises:increasing the temperature from ambient to 1,300° C. at a rate of 5° C./min;increasing the temperature from 1,300° C. to 1,700° C. at a rate of 2° C./min;maintaining the temperature at 1,700° C. for 5 h;decreasing the temperature from 1,700° C. to 1,300° C. at a rate of 2° C./min; anddecreasing the temperature from 1,300° C. to ambient at a rate of 5° C. 9. The molybdenum-100 target assembly of claim 2, wherein the dispersion-strengthened copper composite comprises aluminum oxide ceramic particles. 10. The molybdenum-100 target assembly of claim 3, wherein the brazing alloy comprises a range of 2-18 wt % silver, a range of between 75-92 wt % copper, and a range of 5-7.25 wt % phosphorus.
053032721	abstract	The present invention relates to an apparatus for manipulating the springs which hold the fuel rods firmly inside the grid cells of the grids in a nuclear fuel assembly. Conventionally fuel rods are inserted into the grids while the grid springs are in the active position, thus exposing the rods to possibilities of surface scratches and potential problems. The invented automatic apparatus avoids the formation of such surface damage on the fuel rods, by inserting the keys to inactivate the grid springs before inserting the fuel rods into the grids and removing the keys after all the rods are in place. The apparatus thus contributes to efficient manufacturing of the fuel assembly by simplifying the key quantity management operations and other tasks associated with the assembling operations.
	description	The present invention relates to a device and a method for leakage testing by sipping of a nuclear fuel assembly. A nuclear fuel reactor comprises a vessel in which a plurality of nuclear fuel assemblies are arranged together forming the core of the nuclear reactor. The vessel has a removable cover that must be removed to load and unload the nuclear fuel assemblies. The nuclear reactor is generally arranged in a reactor well filled with water when the reactor is stopped, the well communicating with at least one pool that is also filled with water, making it possible to perform the maintenance operations. Each nuclear fuel assembly contains nuclear fuel. More specifically, each nuclear fuel assembly comprises a bundle of nuclear fuel rods, each nuclear fuel rod comprising an elongate tubular cladding closed at both of its ends and containing nuclear fuel. During operation, a coolant circulates through the core, along the nuclear fuel assemblies, and in particular along their nuclear fuel rods. The coolant keeps the core at an operating temperature and also serves as moderator for the nuclear reaction. During its utilization in a nuclear reactor, a nuclear fuel rod of a nuclear fuel assembly can lose its integrity and have a leakage for example by piercing or cracking of the cladding of the rod. Such a leakage allows fission products resulting from the nuclear reaction to escape, in particular fission gases (xenon, krypton) and iodine components. These fission products become mixed with the coolant and can become deposited on the elements making up the nuclear reactor, thus increasing the level of radioactivity of the facility. The claddings of the rods of the nuclear fuel assemblies make up the first confinement “barrier” of the fission products. During utilization, the monitoring of the tightness of the rods of the nuclear fuel assemblies loaded in a nuclear reactor is done by regular measurements of the radioactivity present in the coolant. The measurements of activities in gas and iodine components make it possible to detect a leakage and are used to estimate the number of affected rods, their burn-up, their location in the core and the size of the flaw(s). However, these measurements do not make it possible to determine which nuclear fuel assembly contains a fuel rod having a leakage. This determination takes place through individual inspection of the potentially affected spent nuclear fuel assemblies when the reactor is stopped, and to that end, the cover of the nuclear reactor vessel is removed. To detect a potential leakage of a spent nuclear fuel assembly, it is possible to perform a leakage testing by sipping. A leakage testing by sipping consists of causing a relative increase in the internal pressure of the nuclear fuel rods of a nuclear fuel assembly relative to the outside pressure. Since the pressures inside and outside the fuel rod tend to balance each other out naturally, a transfer of the fission products that are essentially in gaseous form inside the rod occurs toward the outside. The relative increase of the internal pressure of the fuel rods is for example obtained by causing an increase in the temperature of the nuclear fuel assembly or a decrease in the outside pressure. It is possible to perform a leak testing by in-core sipping, i.e. in the nuclear reactor. The leakage testing by in-core sipping is done on the nuclear fuel assemblies in position in the core of the nuclear reactor. Examples of leakage testing devices by in-core sipping are disclosed in documents U.S. Pat. Nos. 3,856,620A, 4,082,607A, 4,248,666A, EP1183692B1 and EP1810297B1. However, leakage testing by in-core sipping on a nuclear fuel assembly can be disrupted or interfered with by the presence of potential fission products trapped in the coolant and/or coming from adjacent assemblies. Furthermore, this leakage testing is only made possible for nuclear fuel assemblies including a housing surrounding the bundle of nuclear fuel rods and acting as sipping cell, as is the case for the nuclear fuel rods of boiling water reactors (BWR). Thus, such sipping devices have only a very limited effectiveness for nuclear fuel assemblies not including a housing, in particular nuclear fuel assemblies for pressurized water reactors (PWR). It is possible to perform a leakage testing by sipping during the unloading of a nuclear fuel assembly outside the core, using a loading machine equipped with a sipping apparatus, as described in EP1183692B1. Indeed, due to its unloading, the nuclear fuel assembly is generally raised to a depth lower than that of the nuclear reactor, such that the pressure outside the nuclear fuel assembly decreases, which causes a release of fission products outside a nuclear fuel rod having a leakage. In particular released are gaseous fission products and water containing dissolved fission products and/or those in the form of suspended particles, thus making it possible to perform a leakage testing by sipping. However, in some cases, in particular when the sipping apparatus is contaminated by fission products released by a previously verified nuclear fuel assembly, this loading machine does not make it possible to determine with certainty whether a nuclear fuel assembly has a leakage, and the nuclear fuel assembly can only be classified as “doubtful”. It is then appropriate to perform a leakage testing by sipping in a dedicated sipping cell to confirm the leakage. A leakage testing by sipping can be done in a sipping cell dedicated to sipping and arranged in a pool in communication with the reactor well in which the nuclear reactor is located. The sipping cell comprises one or several accommodations for receiving the nuclear fuel assembly/ies to be verified. Such sipping cells are disclosed in documents U.S. Pat. Nos. 4,072,559A, 4,039,376A and JP5726688B2. It is possible to provide fixed sipping cells. However, in the old nuclear power plants, these sipping cells are also old, and their maintenance and update are costly. It is possible to provide mobile sipping cells that can be installed specially in the pool during maintenance operations of the nuclear reactor and uninstalled once the maintenance operations are completed. However, such sipping cells are generally cumbersome, heavy, and time-consuming to install, decontaminate and uninstall. In all cases, the leakage testing by sipping of each potentially affected nuclear fuel assembly requires handling the nuclear fuel assembly up to a sipping cell, performing the leakage testing by sipping, handling nuclear fuel assembly again outside the sipping cell, rinsing and optionally decontaminating the sipping cell before inserting a following nuclear fuel assembly. At the end of the leakage testing by sipping, the sipping cell must further be decontaminated. This therefore requires many handling operations of the nuclear fuel assemblies, which are time-consuming to perform and increase the risk of damaging a nuclear fuel assembly. Furthermore, the sipping cells are placed in the pool. They must be compatible with their environment (anchor points, etc.) and must be designed for normal operation and in case of incident, in particular in case of earthquake. Such sipping cells represent a costly investment with respect to their episodic use. The present disclosure provides a sipping device making it possible to perform leakage testing by sipping on nuclear fuel assemblies simply, quickly, inexpensively, while reducing or even eliminating the additional handling operations of these assemblies and limiting the interference effects that can distort the result of the leakage testing due to the presence of adjacent assemblies. To that end, a leakage testing device is provided for testing leakage of a nuclear fuel assembly by sipping, comprising a collection assembly that is configured to close an upper end of a cell of a storage rack for storing nuclear fuel assembly/ies discharged from a nuclear reactor, so as to prevent water contained in the cell from escaping via the upper end of the cell, and to collect products containing possible fission products released by a nuclear fuel assembly contained in the cell. The device may also include one or more of the following features, considered alone or according to all technically possible combinations: the collected products are gas and/or water contained in the cell, the potential fission products contained in the collected products are in gaseous form and/or dissolved in the water contained in the cell and/or suspended in the water contained in the cell, the collection assembly comprises a tube extender configured to be arranged at the upper end of the cell while extending the cell upward, a bell-shaped cover configured to cap the extender, a lower edge of the cover surrounds the extender while being located at a lower level than that of an upper edge of the extender, the extender and the cover are secured to one another so as to be able to be handled jointly, the cover has a pyramidal or conical shape converging upward, the cover comprises at least one tap to feed a pressurized gas below the cover, the cover comprises at least one tap to collect any fission products, the cover comprises a tap to perform both the feeding of the pressurized gas and the collection of any gaseous fission products, a control assembly separate from the collection assembly and connected to the collection assembly, the control assembly being configured for the analysis of products collected by the collection assembly to detect the possible presence of fission products, a heating unit configured to be arranged in the bottom of a cell and to support a nuclear fuel assembly received in the cell, a bubbling device configured to loosen fission product bubbles from the nuclear fuel assembly. A method for testing leakage of a nuclear fuel assembly by sipping is also provided, the leakage testing being done while the nuclear fuel assembly is stored in a cell of a rack for storing nuclear fuel assemblies discharged from a nuclear reactor. The method may also include one or more of the following features, considered alone or according to all technically possible combinations: the collection of products using a collection assembly configured to close an upper end of the cell so as to prevent water contained in the cell from escaping via the upper end of the cell, and the analysis of collected products to detect possible fission products released by the nuclear fuel assembly contained in the cell, the arrangement at the upper end of the cell of a tubular extender of the collection assembly to extend the cell upward, the extender being capped by a bell-shaped cover of the collection assembly, the leakage testing of several nuclear fuel assemblies arranged in the respective cells of the same storage rack, by moving the collection assembly successively from one of the cells to the other in order to test leakage of the nuclear fuel assemblies, the heating of a nuclear fuel assembly received in a cell, the heating is done using a heating unit attached at the bottom of the cell. The nuclear power plant 2 illustrated in FIG. 1 comprises a nuclear reactor 4 arranged in a reactor well 6, and a storage pool 8 for storing nuclear fuel assemblies 18 discharged from a nuclear reactor 4. The storage pool 8 is in communication with the reactor well 6. The reactor well 6 and the storage pool 8 are filled with water to provide radiological protection for the assembly. The water level is provided so that the handling necessary during maintenance operations, and in particular the handling of the nuclear fuel assemblies 18, are done underwater. The nuclear reactor 4 comprises a vessel 10 and a cover 14 mounted removably on the vessel 10. The vessel 10 is connected to a primary circuit by tubings 16 for the circulation of a coolant through the vessel 10. The nuclear reactor 4 is loaded with a set of nuclear fuel assemblies 18 arranged side by side and together forming the core 20 of the nuclear reactor 4. For the loading and unloading of the nuclear fuel assemblies 18, the cover 14 of the vessel 10 is removed, in order to free the access to the inside of the vessel 10. The nuclear power plant 2 comprises one or several storage racks 22 arranged in the storage pool 8 and in particular configured to store nuclear fuel assemblies 18 discharged from the nuclear reactor 4, in particular during maintenance operations of the nuclear reactor 4. A storage rack 22 comprises a plurality of individual tubular cells 24, each cell 24 being configured to receive a respective nuclear fuel assembly 18. Each cell 24 is vertically elongated. Each cell 24 has an upper end 24A and a lower end 24B. Each cell 24 has an upper opening 26 located at the upper end 24A of the cell 24 and a lower opening 28 located at the lower end 24B of the cell 24. The upper opening 26 is sized for the insertion and removal of a nuclear fuel assembly 18 vertically. The lower opening 28 is arranged in a bottom configured to support a nuclear fuel assembly 18 received in the cell 24. The lower opening 28 and the upper opening 26 allow the circulation of water from the bottom toward the top through the cell 24 by convection. When a nuclear fuel assembly 18 is received in the cell 24, this circulation allows a regulation of the temperature of the nuclear fuel assembly 18. Each cell 24 has a sufficient height for a nuclear fuel assembly 18 inserted in a cell 24 to be fully contained in the cell 24. During operations for maintenance and reloading with fuel of the nuclear reactor 4, all of the nuclear fuel assemblies 18 of the nuclear reactor 4 are discharged. When the presence of fission products has been detected beforehand in the coolant of the nuclear reactor 4, an exam is done to test leakage of the nuclear fuel assemblies 18 that are potentially affected. Indeed, for safety and radiological cleanliness reasons of the nuclear reactor 4, a nuclear fuel assembly 18 having a leakage cannot be reloaded in the nuclear reactor 4. The nuclear power plant 2 is equipped with a leakage testing device 30 configured to perform leakage testing by sipping on the nuclear fuel assemblies 18. As illustrated in FIGS. 1 and 2, the leakage testing device 30 is configured to perform leakage testings by sipping on nuclear fuel assemblies 18 stored in cells 24 of the storage rack 22. The leakage testing device 30 comprises a collection assembly 32 configured to be arranged at the upper end 24A of a cell 24 of the storage rack 22, the collection assembly 32 being configured to close the upper end 24A of the cell 24 so as to prevent water from coming out through the upper end 24A of the cell 24 and to collect any gaseous fission products released by a nuclear fuel assembly 18 received in the cell 24. The closing of the upper end 24A of the cell 24 makes it possible to cause an increase in the temperature of a spent nuclear fuel assembly 18 stored in the cell 24, due to its residual power. This power is produced primarily by the radioactivity of the nuclear fuel contained in the nuclear fuel assembly 18 as well as by the residual fissions. A sipping verification is more effective when it is done immediately after transfer of the nuclear fuel assembly 18 from the nuclear reactor 4 to the cell 24, the residual power of the nuclear fuel assembly 18 then being higher. The collection assembly 32 comprises a cover 34 to close the upper end 24A of the cell 24 and to collect any gaseous fission products released by a nuclear fuel assembly 18 received in the cell 24. The cover 34 assumes the form of a bell. The cover 34 delimits an inner volume. The cover 34 has a lower edge 34A and an apex 34B. Preferably, the cover 34 has a convergent shape over all or part of its height. The upper part of the cover 34 here has a pyramidal or conical shape converging toward the apex 34B. The cover 34 comprises at least one tap to collect gases present below the cover 34. The branch to collect gases is preferably located near the apex 34B. The cover 34 comprises at least one tap located on the cover 34 and making it possible to supply the cover 34 with pressurized gas, for example pressurized air. The cover 34 filled with pressurized air and located above the cell 24 makes it possible to further limit the circulation of water in the cell 24, even inasmuch as the water cannot circulate in the cover 34. Conversely, gaseous fission products can escape from the water and be collected by the cover 34. In the illustrated example, the cover 34 comprises a tap 36 to perform both the collection of gases trapped by the cover 34 and the bringing of pressurized gas below the cover 34. In a variant, the cover 34 comprises separate taps for gas collection and for the injection of pressurized gas. Optionally, the cover 34 bears a temperature probe 38 configured to measure the temperature of the water contained in the cell 24 on which the sipping device is arranged. The temperature probe 38 is for example configured to be submerged in the water contained in the cell 24 when the collection assembly 32 is arranged on the upper end 24A of the cell 24. Optionally, the cover 34 comprises a tap configured to withdraw water in the cell 24 on which the collection assembly 32 is arranged. The cover 34 is for example formed by a mechanically welded assembly made from stainless steel, in particular AISI 304, 304L, 316 or 316L steel. In the illustrated example, the collection assembly 32 comprises a tubular extender 40 configured to be arranged at the upper end 24A of the cell 24 so as to extend it upward. The extender 40 makes it possible to extend the cell 24 past the upper edge of adjacent cells 24. In the operating position, the extender 40 protrudes relative to the adjacent cells 24. In particular, the extender 40 has an upper edge 40A located at a level higher than that of the upper ends 24A of the adjacent cells 24. In the illustrated example, the extender 40 is configured to sink partially into the cell 24. The extender 40 has a lower portion 42 configured to be fitted into the cell 24, and an upper portion 44 provided to protrude upward from the cell 24 when the extender 40 is fitted into the cell 24. The extender 40 has an outer shoulder 46 at the junction between the lower portion 42 and the upper portion 44. The outer shoulder 46 forms a stop limiting the pushing of the lower end 40B of the extender 40 into the cell 24. The fitting of the extender 40 into the cell 24 ensures correct horizontal positioning of the extender 40 relative to the cell 24. Furthermore, the outer shoulder 46 ensures correct vertical positioning. The extender 40 is for example formed by a mechanically welded assembly made from stainless steel, in particular AISI 304, 304L, 316 or 316L steel. The cover 34 is configured to cap the extender 40. The cover 34 and the extender 40 thus cooperate effectively to close the upper end 24A of the cell 24 and collect the gases coming from the cell 24. In particular, the extender 40 is partially engaged inside the cover 34. The lower edge 34A of the cover 34 is located at a level lower than that of the upper edge 40A of the extender 40. With this configuration, when the collection assembly 32 is installed at the upper end 24A of a cell 24 and the cover 34 is filled with pressurized gas, the volume of pressurized gas can drop to a level lower than that of the upper edge 40A of the extender 40. The water contained in the cell 24 is thus prevented from coming out through the top of the cell 24, and therefore from passing through the cell 24. This is obtained without the lower edge 34A of the cover 34 interfering with the upper ends 24A of the adjacent cells 24. As illustrated by the arrows F, the water contained in the cell 24 optionally circulates in a closed loop in the cell 24. The lower edge 34A of the cover 34 has dimensions larger than those of the upper edge 40A of the extender 40 such that the extender 40 can engage inside the cover 34. Preferably, the lower edge 34A of the cover 34 is located at a higher altitude than the upper end of the nuclear fuel assembly 18. This arrangement makes it possible to guarantee that the nuclear fuel assembly 18 always remains surrounded by water and therefore to satisfy the nuclear safety criterion requiring the presence of water to avoid any uncontrolled heating of the nuclear fuel assembly 18. In one specific embodiment, the cover 34 and the extender 40 are secured to one another so as to be able to be handled jointly, as a single unit. The collection assembly 32 is therefore unitary. The cover 34 and the extender 40 here are connected by a horizontal rod 47 passing through the cover 34 and the extender 40. Optionally, the cover 34 and the extender 40 can form a single and same mechanically welded or bolted assembly. The collection assembly 32 comprises a handling system 48 for handling the collection assembly 32. The handling system 48 here is positioned on the cover 34. The handling system 48 is for example configured to be able to be grasped using handling tools provided to handle nuclear fuel assemblies. Thus, the handling tools already provided to handle nuclear fuel assemblies make it possible to handle the collection assembly 32 without having to provide specific handling tools. Optionally, the collection assembly 32 comprises a maintaining device 50 to keep the collection assembly 32 in position installed on a cell 24. The maintaining device 50 for example comprises a ballast. The ballast exerts a continuous vertical force retaining the collection assembly 32 on the cell 24. As a variant or addition, the maintaining device comprises a mechanical locking system 51. Such a locking system 51 for example comprises one or several hook(s) or latch(es) provided to engage with the storage rack 22. Such a locking system 51 is for example able to be actuated remotely using poles. The leakage testing device 30 comprises a control assembly 52 that is separate and remote from the collection assembly 32. As illustrated in FIG. 2, the control assembly 52 is provided to be placed on the edge of the storage pool 8 and is connected to the collection assembly 32, for example by pipes 56 and/or cables. The control assembly 52 comprises a pressurized gas source 54, for example pressurized air, connected to the tap 36 of the cover 34 by a pipe 56, to supply the cover 34 with pressurized gas. The control assembly 52 comprises a measuring device 58 connected to a tap 36 of the cover 34 by a pipe 56 in order to collect gases trapped by the cover 34, and configured to perform measurements on gases collected by the cover 34. The measuring device 58 is for example configured to measure rays emitted by the collected gases, for example gamma rays and/or beta rays. The measuring device 58 is for example configured to take ray counting measurements. The control assembly 52 comprises a computer 60 configured to analyze the measurement signals provided by the measuring device 58. The computer 60 is preferably configured to determine the possible presence of gaseous fission products in the collected gases based on measuring signals provided by the measuring device 58. The control assembly 52 comprises a man-machine interface device 62 configured to restitute the result of analyses provided by the computer 60 for a user. The man-machine interface device 62 for example comprises a display screen that may or may not be touch-sensitive, a keyboard, a pointing device, a touch-sensitive slab and/or a printer. Optionally, the computer 60 is configured to account for a temperature measuring signal provided by a temperature probe 38 equipping the collection assembly 32 and/or to restitute the temperature for the user via the man-machine interface device 62. Knowing the temperature makes it possible to monitor that the sipping is done under safe conditions, without boiling of the water contained in the cell 24. Optionally, the measuring device 58 is configured to detect solid and/or dissolved gaseous fission products and/or fission products suspended in water sampled in the cell 24 using the collection assembly 32. The detection of the dissolved or suspended fission products is done for example by gamma ray counting measurements using a spectrometer. The detection of fission products dissolved and/or in suspension in the water makes it possible to improve the effectiveness of the detection. A leakage testing method by sipping implemented using the leakage testing device 30 is described below. The spent nuclear fuel assemblies 18 are discharged from the nuclear reactor 4 and each inserted into a cell 24. In parallel and over the course of the discharge of the nuclear reactor 4, the collection assembly 32 is installed at the upper end 24A of a first cell 24. To that end, the lower portion 42 of the extender 40 is engaged in the cell 24. If applicable, a locking system 51 is activated to keep the collection assembly 32 secured to the cell 24. The collection assembly 32 is then in the position of FIG. 2. The extender 40 is arranged at the upper end 24A of the cell 24 so as to extend it vertically upward and the cover 34 caps the extender 40. Pressurized gas is sent into the cover 34. The pressurized gas chases out the water present in the cover 34. A gas pocket is imprisoned under the cover 34 and prevents the water contained in the cell 24 from leaving through the upper opening 26 thereof. The pressurized gas is injected below the cover 34 using the pressurized gas source 54 connected to the cover 34 by the pipe 56. The water present in the cell 24 gradually heats up due to the residual power of the nuclear fuel assembly 18. The presence of the cover 34 prevents the water from leaving the cell 24 as it would do by convection without the cover 34. Thus, the nuclear fuel assembly 18 is cooled less and its temperature increases. The water present in the cell 24 possibly circulates in a closed loop in the cell 24 due to the convection. It heats up while circulating upward along the nuclear fuel assembly 18 and descends again along the walls of the cell 24. Due to the temperature increase, the pressure inside the nuclear fuel rods of the nuclear fuel assembly 18 increases. If a nuclear fuel rod has a leakage, gaseous fission products escape from said rod of the nuclear fuel assembly 18, rise in the cell 24 and are collected by the cover 34. The gases trapped below the cover 34 are led to the measuring device 58, measurements are done on these gases by the measuring device 58, and an analysis of the measuring signals is done by the computer 60. The analysis results are restituted to a user by means of the man-machine interface device 62. The leakage testing device 30 is next moved toward a cell 24 containing the next nuclear fuel assembly 18 to be verified. The method preferably comprises the implementation of leakage tests successively on several nuclear fuel assemblies 18 located in respective cells 24 of the or several storage racks 22, by moving the leakage testing device 30 from one cell 24 to another in order to perform the following leakage testing. Thus, several successive leakage tests can be done on several nuclear fuel assemblies 18 received in respective cells 24 without having to handle the nuclear fuel assemblies 18. Furthermore, the nuclear fuel assemblies 18 discharged from a nuclear reactor 4 are generally stored in storage racks 22. Thus, the leakage testing of the nuclear fuel assemblies 18 can be done without additional handling of the nuclear fuel assemblies 18 relative to the normal manipulations, which represents a considerable time savings. Optionally, the leakage testing can be done at the end of the discharge of the nuclear fuel assemblies 18 from the nuclear reactor 4. The extender 40 makes it possible to extend the cell 24 above the adjacent cells 24, so as to be able to position the cover 34 effectively to trap possible gaseous fission products emitted by the nuclear fuel assembly 18 present in the cell 24. In particular, the cover 34 can cap the extender 40 with the lower edge 34A of the cover 34 located at a level lower than that of the upper edge 40A of the extender 40, which prevents the gas from leaving the cover 34, at least as long as the pressure of the gases under the cover 34 is lower than the pressure of the water at the lower edge 34A of the cover 34. Some “cold” nuclear fuel assemblies 18 have a lower residual power than others. If the residual power is too low, simply closing the upper end 24A of a cell 24 in which the nuclear fuel assembly 18 is stored due to the installation of the collection assembly 32 at the upper end 24A of the cell 24 can be insufficient for satisfactory sipping. Optionally, as illustrated in FIG. 3, the leakage testing device 30 comprises a heating unit 64 configured to be inserted at the bottom of a cell 24 and to heat the water contained in the cell 24. The heating unit 64 is configured to support a nuclear fuel assembly 18 stored in the cell 24. The heating unit 64 is thus sized to support the weight of a nuclear fuel assembly 18 stored in the cell 24. The heating unit 64 is provided with a height that is as small as possible to prevent the nuclear fuel assembly 18 from protruding outside the cell 24. The heating unit 64 for example comprises a parallelepiped box 66. The box 66 is for example formed by a mechanically welded assembly made from stainless steel, for example AISI 304, 304L, 316 or 316L steel. The box 66 has water circulation passages to allow the circulation of water in the cell 24 vertically from the bottom toward the top when no sipping cycle is implemented in the cell 24. The heating unit 64 comprises a heat source 68 to heat the water contained in the cell 24 during a sipping cycle. The heat source 68 here is a heating electrical resistance contained in the box 66. The heat source 68 is located in the box 66. The heating unit 64 comprises an electrical power supply 70 to supply electricity to the heat source 68. The electrical power source 70 comprises an electrical battery housed in the heating unit 64 and/or an electrical power cable to connect the heating unit 64 to a remote electrical power source. The electrical power source 70 here comprises a power cable 72 that is provided to come out through the lower opening 28 of the cell 24 when the heating unit 64 is inserted in the cell 24 through the upper opening 26 of the cell 24. The power cable 72 is provided with an electrical connector 74 at its end opposite the heating unit 64. Preferably, the heating unit 64 comprises a cable handling chain 76 carrying the power cable 72. Due to its weight and its flexibility, the cable handling chain 76 facilitates the passage of the cable through the lower opening 28 of the cell 24 during the insertion of the heating unit 64 into the cell 24. In one specific embodiment, the cable handling chain 76 is configured such that once the heating unit 64 is inserted into the bottom of a cell 24, the electrical connector 74 rests on the bottom of the storage pool 8. Preferably, the cable handling chain 76 comprises segments 78 articulated relative to one another such that the electrical connector 74 becomes offset laterally on a determined side relative to the heating unit 64 when the heating unit 64 is brought closer to the ground. Thus, when the heating unit 64 is lowered into a cell 24, the cable handling chain 76 passes through the lower opening 28 of the cell 24, then the electrical connector 74 rests on the bottom of the storage pool 8, then shifts laterally when one continues to lower the heating unit 64. Preferably, during operation, the heating unit 64 is inserted into a peripheral cell 24 of the storage rack 22 such that the electrical connector 74 shifts on the side of the storage rack 22 and is accessible on the side of the storage rack 22, as illustrated in FIG. 3. The side on which the electrical connector 74 shifts depends on the orientation of the heating unit 64. Advantageously, the latter comprises a mistake-proofing mark making it possible to orient it correctly during its insertion into the peripheral cell 24, so as to ensure that the electrical connector 74 will come out on the side of the storage rack 22 and will be accessible from the edge of the storage pool 8. For the connection to a remote electricity source, the leakage testing device 30 for example comprises a connecting pole 80 bearing an electrical connecting cable 82 provided at its lower end with an electrical connector 84 complementary to the electrical connector 74 of the power cable 72. The operation of the sipping device is similar to that of the sipping device of FIG. 2, with the exception that the heat for the heating of the nuclear fuel assembly 18 is provided by the heating unit 64. For the installation, the heating unit 64 is first installed at the bottom of the cell 24, then the nuclear fuel assembly 18 is inserted in the cell 24 so as to rest on the heating unit 64, then the collection assembly 32 is placed at the top of the cell 24. The heating unit 64 is possibly connected to an electricity supply source by means of its power cable 72 using a connecting pole 80. In the variant illustrated in FIG. 4, the heating unit 64 further comprises a bubbling device 86 for generating gas bubbles, for example air bubbles, in order to improve the collection of fission products potentially glued on the tubular sheath of the nuclear fuel rods. The bubbling device 86 can be implemented independently of the heating unit 64, in particular in the case of the leakage testing for nuclear fuel assemblies 18 having a sufficient residual power. During operation, the bubbles generated by the bubbling device 86 rise along the nuclear fuel assembly 18 and drive bubbles from gaseous fission products that are glued on the nuclear fuel assembly 18, in particular on nuclear fuel rods or grates of the nuclear fuel assembly 18. In other words, the bubbling device 86 is configured to loosen fission product bubbles from the nuclear fuel assembly 18. This therefore improves the effectiveness of the detection of leakages. The bubbling device 86 for example is in the form of a trellis or ring(s), or toroid(s) with a short height making it possible to generate bubbles over all or part of the surface of the bubbling device 86 while arranging recesses for the circulation of water in the cell 24 when the cell 24 is not closed by a collection assembly 32. To generate bubbles, the bubbling device 86 is for example connected to a gas source. It is possible to use the pressurized gas source 54 provided to fill the cover 34. For this connection, the heating unit 64 for example comprises a supply pipe carried by the cable handling chain 76, the electrical connector 74 further being configured for a fluid connection of this supply pipe to a connecting pipe carried by the connecting pole 80 and connected to the gas source. Owing to the leakage testing device and method, it is possible to test leakage of nuclear fuel assemblies 18 discharged from a nuclear reactor 4 by sipping simply, reliably and quickly. The leakage testing is done directly in the storage rack 22 in which the nuclear fuel assemblies 18 discharged from the nuclear reactor 4 are stored. It is not necessary to have a dedicated fixed or mobile sipping cell separate from the storage rack 22, or to handle the stored nuclear fuel assemblies 18 in order to verify nuclear fuel assemblies 18 having a sufficient residual power. The leakage testing is further done outside the nuclear reactor 4, which limits the handling operations above the nuclear reactor 4 itself and therefore limits the risks inherent to such procedures. The leakage testing device and method further make it possible to perform leakage testing of “cold” nuclear fuel assemblies 18 in a cell 24 equipped with a heating unit 64. The leakage testing device 30 is light, compact and easy to install and uninstall. It makes it possible to shorten the procedure times and reduce the risks of damaging the nuclear fuel assemblies 18, since it makes it possible to reduce the handling of the nuclear fuel assemblies 18, in particular when they have a sufficient residual power. The leakage testing with the leakage testing device 30 only requires moving the leakage testing device 30 from one cell 24 to another, thus allowing non-negligible time savings during the successive testing of several nuclear fuel assemblies 18. The leakage testing device 30 uses the storage rack 22 already present in the storage pool 8. Thus, it is not necessary to provide additional anchor points in the storage pool 8, as is for example the case for a dedicated fixed or mobile sipping cell. It is also no longer necessary to provide cumbersome and costly studies to justify its behavior in case of accident to the nuclear safety authorities. The contamination risks of the leakage testing device 30 by direct contact with a nuclear fuel assembly 18 are eliminated. The leakage testing device 30 makes it possible to perform several successive leakage tests using a same collection assembly 32 without having to perform rinsing or decontamination between two successive leakage tests. The leakage testing device and method is not limited to the example embodiments described above. Alternatives can be considered. The leakage testing device 30 here comprises a single collection assembly 32. In a variant, the leakage testing device 30 comprises several collection assemblies 32. This makes it possible to perform several leakage tests on several respective nuclear fuel assemblies 18 in parallel or to perform a leakage test on a nuclear fuel assembly 18 using a collection assembly 32 during the preparation of another leakage test on another nuclear fuel assembly 18 using another collection assembly 32. When the leakage testing device 30 comprises several collection assemblies 32, it is possible to provide a control assembly 52 shared by several collection assemblies 32 and able to be connected to one of these collection assemblies 32 at a time. The shared control assembly 52 is then successively connected to the different collection assemblies 32. In the illustrated examples, the extender 40 bears on the cell 24 on which the collection assembly 32 is installed. Optionally or in a variant, the extender 40 is configured to bear on one or several cells 24 adjacent to the cell 24 on which the collection assembly 32 is installed. This makes it possible to distribute the weight of the collection assembly 32 over several cells 24. In the illustrated examples, among the cover 34 and the extender 40, only the latter bears on the cell 24. In a variant or optionally, the cover 34 bears on the cell 24 on which the collection assembly 32 is installed and/or on one or several cells 24 adjacent to the cell 24 on which the collection assembly 32 is installed.
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	claims	1. A nuclear fuel assembly for a pressurized water nuclear reactor, the nuclear fuel assembly having a zoned configuration representing all nuclear fuel rods of the fuel assembly, the zoned configuration consisting of:a central zone comprising nuclear fuel rods which, before irradiation, contain uranium oxide but do not contain any plutonium oxide, wherein the central zone does not comprise, before irradiation, any fuel rod based on uranium and plutonium mixed oxide, anda peripheral zone extending along outer faces of the nuclear fuel assembly, the peripheral zone only comprising before irradiation nuclear fuel rods exclusively based on uranium and plutonium mixed oxide, the peripheral zone consisting of a single outermost layer of nuclear fuel rods,such that there are no nuclear fuel rods other than the single outermost layer of nuclear fuel rods that, before irradiation, are exclusively based on uranium and plutonium mixed oxide. 2. The assembly according to claim 1, wherein at least some of the nuclear fuel rods of the central zone are poisoned nuclear fuel rods which contain before irradiation a consumable neutron poison. 3. The assembly according to claim 1, the assembly comprising, in addition to the nuclear fuel rods, guide tubes for receiving rods of a control cluster and an instrumentation tube, the nuclear fuel rods, the guide tubes and the instrumentation tube occupying all the nodes of a regular network. 4. The assembly according to claim 1, wherein the assembly does not comprise any outer casing. 5. The assembly according to claim 1, wherein the nuclear fuel is formed as solid pellets contained in the nuclear fuel rods. 6. A nuclear fuel assembly for a pressurized water nuclear reactor, the nuclear fuel assembly having a zoned configuration consisting of:a central zone comprising nuclear fuel rods which, before irradiation, contain uranium oxide but do not contain any plutonium oxide, wherein the central zone does not comprise, before irradiation, any fuel rod based on uranium and plutonium mixed oxide, anda peripheral zone extending along outer faces of the nuclear fuel assembly, the peripheral zone only comprising before irradiation nuclear fuel rods exclusively based on uranium and plutonium mixed oxide, the peripheral zone consisting of a single outermost layer of nuclear fuel rods,the assembly comprising, in addition to the nuclear fuel rods, guide tubes for receiving rods of a control cluster, the nuclear fuel rods and the guide tubes occupying all the nodes of a regular network. 7. The assembly according to claim 6, wherein at least some of the nuclear fuel rods of the central zone are poisoned nuclear fuel rods which contain before irradiation a consumable neutron poison. 8. The assembly according to claim 6, wherein the assembly does not comprise any outer casing. 9. The assembly according to claim 6, wherein the nuclear fuel is formed as solid pellets contained in the nuclear fuel rods.
061852791	description	DETAILED DESCRIPTION OF THE INVENTION FIG. 1 represents the X-ray emission part of an X-ray source of the prior art as well as the part of the diaphragm from this source. Conventionally, the X-ray source comprises a cathode 10 and a rotating anode 11 which are contained in a casing 12 that is transparent to X-rays. The assembly is itself closed in a casing 13 provided with a skirt 14 made of material that is opaque to X-rays, for example made of lead, with the exception of a region located facing the anode 11, which consists of a window 15 made of a material that is transparent to X-rays. The gap between the casing 12 and the protective casing 13 and the window 15 is filled with oil which ensures cooling and high-voltage electrical insulation of the X-ray source. Consequently, the casing 13 must provide sealing with respect to the oil contained in the gap between this casing 13 and the casing 12 and also ensure that the X-radiation beam will only be emitted through the window 15. As is well known, the cathode 10 emits an electron beam which strikes the focal track of the rotating anode 11 which in turn emits an X-ray beam towards the window 15. As shown by FIG. 1, conventionally, the X-ray beam emitted by the anode 11 is shaped by means of a diaphragm 30 made of a material that is opaque to X-rays, the opening 31 of which is dimensioned in order to obtain an X-radiation beam output by the focus of the anode and to intercept the X-radiation emitted by the source outside the focus, or extrafocal radiation. The parasitic extrafocal radiation should be eliminated because it impairs the quality of the image subsequently obtained. It is furthermore desirable to be capable of varying the dimension of the focal surface of the anode and consequently the geometry of the X-ray beam. Therefore, in order to eliminate the extrafocal radiation, it is necessary to modify the dimension of the opening of the diaphragm 30 as a function of the dimension of the focal surface as well as the collimation of the X-ray beam. Conventionally, in order to adapt the opening of the diaphragm 30 to the X-ray beam and to be able to eliminate the extrafocal radiation, moving fingers 32, made of a material that is opaque to X-rays, are provided and are manipulated by means of a complex rod system 33 in order to dimension the opening 31 as a function of the focal area and the desired area for the final image. This arrangement of the prior art, which has just been described, has several drawbacks. Firstly, since the diaphragm 30 is a separate part from the window 15, it must be placed externally to this window and it is therefore not possible to eliminate the extrafocal radiation as close as possible to the focal surface for emitting the X-ray beam. With regard to the quality of the final image obtained, it is particularly beneficial to be able to eliminate the extrafocal radiation as close as possible to the focal surface of the source. Furthermore, the use of opaque fingers 32 requires a complex mechanical rod system 33 for dimensioning the opening 31 of the X-ray beam diaphragm. In an embodiment of the invention, these drawbacks are overcome by providing a window which is also used as a diaphragm for eliminating the extrafocal radiation. By virtue of the fact that the window itself fulfills the function of a diaphragm, it is possible to place this diaphragm closer to the focal surface of the X-ray beam. Furthermore, the window of one embodiment of the invention permits simple dimensioning of the passage opening for the X-ray beam without requiring a complex mechanical rod system. FIG. 2 represents a window 15 in a casing for an X-ray source in an embodiment of the invention. This window 15 has the general shape of a frustoconical dome and has an end wall 17 and a sidewall 18 ending in a flange 19 for mounting it in the protective casing 13. This window 15 consists of a material that is transparent to X-rays. In an embodiment of the invention, this window 15 comprises an internal chamber designed to permit the displacement of an opaque material which is introduced from outside the chamber in such a way that the area of the X-ray beam passage region varies as a function of the volume of opaque material in the chamber, in order thus to eliminate the extrafocal parasitic radiation from the beam. In a recommended embodiment, the chamber comprises a first channel wound on itself in the shape of a spiral. Preferably, the chamber comprises a second channel in the shape of a spiral, juxtaposed with the first channel in a direction perpendicular to the planes on which the openings of the first and second channels are located, this first and this second channel being offset relative to each other in such a way that the turns of one spiral cover the space between the turns of the other spiral, so as to ensure complete opacity to radiation in the part of the spirals through which the opaque material passes. Preferably, the electromagnetic radiation beam is an X-ray beam and the source is an X-ray source. In the embodiment represented, this chamber comprises an annular cavity 20 arranged in the sidewall 18 and connected at its lower end to a means 25 for controlling the displacement of the material that is opaque to X-rays. This annular chamber 20 is connected at its upper end to the outermost turn of each of two spiral channels 22, 23 arranged in the end wall 17 of the window by means of tubes 21. The spiral channels 22 and 23 are arranged in the end wall 17 in juxtaposed planes parallel to the end wall 17. As represented, the turns of the two channels 22, 23 are offset relative to one another in such a way as jointly to cover virtually the entire area of the end wall 17. Clearly, a single spiral channel could be used, preferably a spiral channel with adjoining turns, as represented in FIG. 3. Use could also be made of more than two superposed spiral channels. The innermost turn of each of these spiral channels 22 and 23 is also connected, via a conduit 24 arranged in the window, to the device for controlling the opaque material. Clearly, it is, if so desired, possible to do without the annular cavity 20 and replace it by two conduits. The material that is opaque to X-rays, for example mercury, is introduced by means of the control device 25 into the annular chamber 20 then into the outer turns of the spiral channels 22 and 23 towards the center of these spiral channels, so as to leave a central passage of suitable dimension for the X-ray beam. The turns of the spiral channels 22 and 23, forming the central passage for the X-ray beam, are filled, also by means of the control device 25, within a liquid that is transparent to X-rays, for example alcohol. The turns of the spiral channels 22 and 23, which are filled with opaque material and surround the central turns filled with material that is transparent to X-rays, therefore form a diaphragm that is opaque to X-rays, making it possible to eliminate the extrafocal radiation. Depending on the quantity of opaque material introduced into the turns of the channels 22 and 23, it is therefore possible to dimension the central passage that is transparent to X-rays and easily to eliminate the extrafocal radiation as a function of the dimension of the focal surface of the X-ray source. Furthermore, since the diaphragm function is fulfilled in the window of the outer casing of the X-ray source itself, it is possible to achieve this elimination of the extrafocal radiation very close to the focal surface of the source. FIG. 4 schematically represents a spiral channel connected to a device for controlling the displacement of the opaque material 25, which is particularly recommended for the present invention. This control device 25 comprises an enclosure in which a piston 26 is arranged, this piston being movable in translation and dividing the enclosure into a first chamber and a second chamber 28. The first chamber 27 is connected by a conduit 21 to the outer turn, for example, of the spiral channel 22. The second chamber 28 is itself connected by a conduit 24 to the innermost turn, for example, of the spiral channel 22. The chamber 27 is filled with material that is opaque to X-rays, for example mercury, whereas the chamber 28 is filled with a material that is transparent to X-rays, for example alcohol. The piston 26 may, for example, be a piston that is movable by means of a magnet 29. As represented, the piston 26 includes two movable magnetic plates, separated from each other, to define a space that is generally filled with air for the purpose of taking up thermal expansion. As is immediately clear, when the magnet is moved either to the right or to the left, the piston 26 is also moved, either to the right or the left, thus introducing into the spiral channel 22 more or less of the product that is opaque to X-rays. It is thus easily possible to dimension the central passage for the X-ray beam and eliminate the extrafocal parasitic radiation in accordance with the purpose intended for the X-ray source. It is also possible to use, instead of the device described above, a peristaltic pump as the device for controlling displacement of the material that is opaque to X-rays. Various modifications in structure and/or function and/or steps may be made by one skilled in the art without departing from the scope and extent of the invention.
	description	This application is a divisional application of U.S. patent application Ser. No. 15/493,310 filed on Apr. 21, 2017, which is a continuation of U.S. patent application Ser. No. 15/311,052 filed Nov. 14, 2016, which is a 371 application of PCT/US2015/031140 filed May 15, 2015, which claims priority to U.S. Provisional Patent Application No. 61/993,365 filed on May 15, 2014. This invention was made with government support under DE-SC0008947 awarded by the Department of Energy. The government has certain rights in the invention. The present invention relates to labeled radiopharmaceuticals. Radiometals (e.g., 64Cu, 89Zr, 68Ga, 86Y and 99mTc) play a pivotal role in nuclear medicine as therapeutic and imaging agents for radiation therapy and labeling of biologically important macromolecules like proteins, peptides and antibodies. In the recent past, a rapid increase has been noted in both clinical and preclinical studies involving 68Ga-labeled radiopharmaceuticals [Ref. 1-5]. This increase can be attributed to the favorable physical characteristics of 68Ga (Epmax 1.8 MeV, β89%, T1/2=67.7 minutes) for imaging various rapidly changing processes (proliferation, apoptosis, angiogenesis) and targets (growth hormones, myocardial and pulmonary perfusion, inflammation and infection), and to some extent, to newer, more reliable production and labeling methods [Ref. 1-5]. Gallium-68 labeled somatostatin analogs have already shown their superiority over the existing agent 111In-DTPA-octreotide through enhanced sensitivity, specificity, accuracy and cost effectiveness for the diagnosis of patients with neuroendocrine tumors [Ref. 1, 6-9]. The clinical promise of 68Ga-labeled radiopharmaceuticals clearly warrants growth of the supply of 68Ga to meet the increasing demand in various nuclear medicine facilities. Presently, 68Ga can be produced by two different approaches, (1) solid targetry [Ref. 10, 11] and (2) the 68Ge/68Ga generator [Ref. 12]. The former requires high capital cost and expertise and specialized cyclotron facilities that accommodate solid targets, whereas, the latter is more broadly accessible in nuclear medicine facilities not equipped with an on-site cyclotron. The simplicity and lower capital cost of the 68Ge/68Ga generator have made it more popular among the nuclear medicine facilities with relatively lower number of requirements for 68Ga labeled doses [Ref. 1, 12]. However, the breakthrough of trace quantities of the long-lived 68Ge parent isotope (t1/2=271 days) into the eluted 68Ga remains a concern [Ref. 13]. Furthermore, with increasing applicability of 68Ga-labeled radiopharmaceuticals, one can foresee a need for alternative production methods to meet the increasing demand especially for the relatively busy nuclear medicine centers having an on-site cyclotron. There have been previous attempts to produce 68Ga using a cyclotron, initially employing a solid target method using 68Zn electrodeposition on a copper substrate [Ref. 10, 14] and more recently using a solution target containing an enriched 68ZnCl2 solution [Ref. 15]. The solid target methods require a lengthy separation step, which is not optimal for short-lived isotopes like 68Ga, as well as expensive solid target infrastructure. Thus, needed in the art are methods and systems to extend the applicability of the solution target approach to the production of 68Ga and other radiometals using a low energy cyclotron. The present invention overcomes the aforementioned drawbacks by disclosing improved methods and systems for preparing radionuclides (e.g., 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc). The methods of producing radionuclides (e.g., 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc) may comprise irradiating a closed target system with a proton beam. The closed target system may comprise a solution target. This disclosure provides a solution target for production of a radionuclide from a target material. The solution target comprises a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam at the target cavity; a target window foil for covering an opening of the target cavity; and a coolant flow path disposed in the passageway upstream of the target window. In one version of the present disclosure, there is provided a cyclotron solution target with improved heat transfer characteristics for production of radiometals. The target can be used in medium energy cyclotrons (11-20 MeV protons) for the production of radionuclides (e.g., 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc). The target has improved thermal transfer characteristics needed for cooling of the target solution. The target includes a water-cooled insert and a helium cooled region between the cyclotron vacuum and the solution. In one embodiment, a first foil separates the cyclotron vacuum from the helium flow region and a second foil (e.g., 20-30 microns) separates the helium region from the target solution. The geometry of the target insert is designed to maximize the heat transfer from the target solution to the cooling water. In another aspect, the disclosure provides a solution target system for production of a radionuclide from a target solution. The system comprises a target body including a target cavity; a housing defining a passageway for directing a particle beam at the target cavity; a target solution distribution valve in fluid communication with the target cavity; a first source of a first target solution, the first source being in fluid communication with the target solution distribution valve; and a second source of a second target solution, the second source being in fluid communication with the target solution distribution valve, wherein the target solution distribution valve is movable from a first distribution position in which the first target solution flows into the target cavity to a second distribution position in which the second target solution flows into the target cavity. In one form, the system provides a target solution loading and unloading system that can select from up to six different metal solutions as reagents. There may also be up to six different delivery lines to minimize cross-contamination of isotopes. In yet another aspect, the disclosure provides a method for producing a solution including a radionuclide, the method comprising bombarding a target solution with protons to produce a solution including a radionuclide, wherein the radionuclide is selected from 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, and 44Sc. The foregoing and other aspects and advantages of the invention will appear from the following description. In the description, reference is made to the accompanying drawings that form a part hereof, and in which there is shown by way of illustration certain embodiments of the invention. Such embodiments do not necessarily represent the full scope of the invention, however, and reference is made therefore to the claims and herein for interpreting the scope of the invention. Like reference numerals will be used to refer to like parts from Figure to Figure in the following description of the drawings. The invention provides a solution target for production of a radionuclide from a target material. The solution target comprises a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam (e.g., protons) at the target cavity; a target window foil for covering an opening of the target cavity; and a coolant flow path disposed in the passageway upstream of the target window. By “upstream”, we mean situated in the opposite direction from that in which the particle beam flows. The coolant flow path can decrease an energy value of the particle beam to 15 MeV or less. The solution target may include an energy degrading foil disposed within the passageway and upstream of the coolant flow path. The energy degrading foil may comprise a cobalt alloy. The target window foil may comprise an aluminum alloy. The solution target may comprise a target solution port in fluid communication with the target cavity. The solution target may further comprise a pressure source in fluid communication with the target cavity. The pressure source may comprise helium gas. The coolant flow path may comprise helium gas, and the helium gas may be confined in a tubular conduit. In the solution target, the target body may further include a second coolant flow path disposed adjacent to the target cavity. The second coolant flow path is defined by a space between a first target body component and a second target body component. The second coolant flow path may comprises water. The target cavity may defined by a conical wall. The target material may comprise a solution of a metal. Referring to FIG. 1A and FIG. 1B, a solution target 10 according to one non-limiting embodiment of the invention is shown. The solution target 10 includes a target body 14 having a first target body component 16 and a second target body component 18. A target cavity 22 is defined by a wall 24 and is disposed within the first target body component 16. The target cavity 22 is configured to receive a target material (e.g. a target solution). In some forms, the wall 24 of the target cavity 22 may be a conical wall. The wall 24 may comprise tantalum or niobium. The solution target 10 further includes a housing 26 configured to define a passageway 28 for a particle beam (e.g., a proton beam). The housing 26 may include a first housing section 30 having an inlet 32 for the particle beam and a portion of the passageway 28 disposed within. A second housing section 36 is spaced between the first housing section 30 and the first target body component 16 and includes a second portion of the passageway 28. A sealing member 38, for example an O-ring, can be disposed between the first housing section 30 and the section housing section 36 to provide a sealed housing 26. A coolant flow path 40 is disposed within the second portion of the passageway 28. The coolant flow path 40 can be confined within a tubular structure, such that the particle beam flows around the tubular structure. The coolant flow path 40 may be a helium cooling system configured to degrade the energy of the particle beam to 15 MeV or less. The solution target 10 further includes a coolant housing 46 configured to define a second coolant flow path. The coolant housing 46 includes an inlet 48 configured to receive a coolant, for example, water. Additionally, the coolant housing 46 includes two outlets 50 configured to expel the coolant. The solution target 10 further includes a first material housing component 60 and a second material housing component 62. A target material inlet 64 is disposed within the first material housing component 60 and is in fluid communication with a target material passageway 66 and the target body cavity 16. An overpressure source 68 is disposed within the second material housing component 62 and is in fluid communication with the target material passageway 66 and the target body cavity 16. The solution target 10 further includes a target window foil 70 disposed between the second housing section 36 and the target body 14. The target window beam entry foil 70 is configured to cover an opening of the target body cavity 16. The target window foil 70 may include a cobalt alloy, and may be configured to degrade the energy of the particle beam. In one example, the target window foil 70 is Havar®. Havar® is a heat treatable cobalt base alloy that provides very high strength, such as an ultimate tensile strength of 330,000 psi when cold rolled and heat treated. The alloy provides excellent corrosion resistance and is non-magnetic. The nominal composition is: cobalt 42.0%, chromium 19.5%, nickel 12.7%, tungsten 2.7%, molybdenum 2.2%, manganese 1.6%, carbon 0.2%. and balance iron. In some forms, an energy degrading foil 74 may be disposed within the passageway 28 and between the first housing section 30 and a second housing section 36. The energy degrading foil 74 may be a metallic foil, for example an aluminum or aluminum alloy foil. In operation, the particle beam (e.g., proton beam) is introduced into the passageway 28 through the inlet 32. The particle beam follows a particle beam path 80 through the first housing section 32 and the energy degrading foil 74. The partial beam 80 continues through the coolant flow path 40 and through the target window foil 70 into the target cavity 22. The target material (e.g., a solution of a metal) is introduced into the target material inlet 64 and follows a target material path 84 through the target material passageway and into the target cavity 22. The target material disposed within the target cavity 22 is bombarded with the particle beam to create a radionuclide. The bombarded target material may be removed using material path 84. A coolant is introduced from a source of coolant (e.g. water) into the coolant housing 46 via the inlet 48. The coolant follows a coolant path 88 from the inlet 48 into a space between the first target body component 16 and the second target body component 18. The coolant path 88 travels adjacent to the wall 24 of the target cavity 22 and through the coolant housing 46 to the two outlets 50. The invention further provides a solution target system for production of a radionuclide from a target solution. The solution target system includes a target body including a target cavity; a housing defining a passageway for directing a particle beam at the target cavity; a target solution distribution valve in fluid communication with the target cavity; a first source of a first target solution wherein the first source is in fluid communication with the target solution distribution valve; and a second source of a second target solution, wherein the second source is in fluid communication with the target solution distribution valve. The target solution distribution valve is movable from a first distribution position in which the first target solution flows into the target cavity to a second distribution position in which the second target solution flows into the target cavity. The solution target system may further include a target loading/unloading valve in fluid communication with the target solution distribution valve and the target cavity, wherein the target loading/unloading valve is movable from a first loading position in which at least one of the first target solution and the second target solution flows into the target cavity and a second unloading position in which a radionuclide solution flows from the target cavity to a collection vessel. The solution target system may further include an injection valve in fluid communication with the target solution distribution valve and the target cavity, wherein the injection valve holds a sample of at least one of the first target solution and the second target solution before the sample flows into the target cavity. The solution target system may further include a source of pressurized gas in gas communication with the top entry (62) to target cavity. The solution target system may further include a pressurized gas delivery valve in gas communication with the source of pressurized gas and the target cavity, wherein the pressurized gas delivery valve controls delivery of pressurized gas to the target cavity. The solution target system may further include a pressure regulating valve in gas communication with the target cavity, wherein the pressure regulating valve controls entry of pressurized gas into the target cavity. The solution target system may further include a pressure monitor in gas communication with the target cavity, wherein the pressure monitor provides a means to measure pressure within the target and may also be used to control the pressure regulating valve. In the solution target system, a target window foil may cover an opening of the target cavity; and a coolant gas flow path may be disposed in the passageway upstream of the target window. An energy degrading foil may be disposed within the passageway and upstream of the coolant flow path. The first target solution may comprise a first solution of a metal and the second target solution may comprise a second solution of a metal wherein the first solution of a metal is different from the second solution of a metal. The solution target system may further include one or more additional sources of an additional target solution. Each of the additional sources is in fluid communication with the target solution distribution valve. The target solution distribution valve is movable to one or more additional distribution positions in which each of the additional target solutions flows into the target cavity. The invention further provides a method for producing a solution including a radionuclide. The method includes the step of bombarding a target solution with protons to produce a solution including a radionuclide, wherein the radionuclide is selected from 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, and 44Sc. The target solution may be bombarded with protons in a target cavity of a solution target operating as a closed system. The method does not include a target dissolution step. The target cavity may be cooled with a coolant during the bombardment. The target cavity is maintained at a pressure below 150 psi. or below 75 psi, during the bombardment. The solution target may have a volume capacity of two milliliters or less. The target cavity may be defined by a generally conical wall and a target window foil. In one version of the method, the target solution is yttrium nitrate and the radionuclide is 89Zr. In another version of the method, the target solution is 68Zn-enriched zinc nitrate and the radionuclide is 68Ga. In another version of the method, the target solution comprises 63Cu-enriched copper (II) nitrate and the radionuclide is 63Zn. In another version of the method, the target solution comprises 86Sr-enriched strontium nitrate and the radionuclide is 86Y. In another version of the method, the target solution comprises 64Ni-enriched nickel nitrate and the radionuclide is 64Cu. In another version of the method, the target solution comprises scandium nitrate and the radionuclide is 45Ti. In another version of the method, the target solution comprises 52Cr-enriched or natural chromium nitrate and the radionuclide is 52Mn. In another version of the method, the target solution comprises 44Ca-enriched or natural calcium nitrate and the radionuclide is 44Sc. In another version of the method, the target solution comprises 61Ni-enriched or natural nickel nitrate and the radionuclide is 61Cu via deuteron irradiation. The target solution may further comprise a dilute solution of nitric acid. The concentration of nitric acid in the target solution may be 0.001 M to 2.5 M, or 0.2 to 2 M. In another version of the method, the target solution comprises aqueous ethanol and the radionuclide is 13N. The method may further include the steps of passing the solution including the radionuclide through a column including a sorbent to adsorb the radionuclide on the sorbent; and eluting the radionuclide off the sorbent. The term “column”, as used herein, refers to a separation technique in which the stationary bed is within a cartridge. The particles of the solid stationary phase or the support coated with a liquid stationary phase, such as resin or sorbent, may fill the whole inside volume of the cartridge (packed column) or be concentrated on or along the inside cartridge wall leaving an open, unrestricted path for the mobile phase in the middle part of the cartridge (open tubular column). Differences in rates of movement through the medium are calculated to different retention times of the sample. The radionuclidic purity of the radionuclide can be greater than 99%, preferably greater than 99.9%, after eluting the radionuclide off the sorbent. The method may further include the step of passing the eluted radionuclide through a second column including a sorbent. The invention further provides a disposable, single-use kit for isolation of a radionuclide from a solution including the radionuclide wherein the solution is produced by bombarding a target material in a target cavity of a solution target with protons. The kit includes a chromatographic column including a sorbent to adsorb the radionuclide on the sorbent, wherein the radionuclide is 68Ga, 89Zr, 64Cu, 86Y, 63Zn, 61Cu, 99mTc, 45Ti, 52Mn or 44Sc. In one form, the sorbent comprises a hydroxamate resin. The kit may include an eluent for eluting the radionuclide off the sorbent. The eluent may comprise a phosphate. The kit may include a radionuclide product vessel for receiving eluted radionuclide from the column. The kit may include a first fluid conduit for placing the column in fluid communication with the target cavity, and a second fluid conduit for placing the column in fluid communication with the radionuclide product vessel. The kit may include a second chromatographic column including a sorbent to adsorb impurities in an eluent used for eluted radionuclide from the column. The invention further provides a synthesis system that employs the kit of the invention for performing the process of isolation of the radionuclide from the solution. The synthesis system may be automated using a controller. The controller can execute a stored program to: (i) deliver the solution including the radionuclide from the target cavity to a collection vessel, (ii) deliver the solution including the radionuclide from the collection vessel to the column, (iii) thereafter deliver an eluent to the column, and (iv) thereafter deliver the eluted radionuclide to a radionuclide product vessel. The following Examples are provided in order to demonstrate and further illustrate certain embodiments and aspects of the present invention and are not to be construed as limiting the scope of the invention. Example 1 extends the applicability of the solution target approach to the production of 68Ga using a low energy (<20 MeV) cyclotron. Since the developed method does not require solid target infrastructure, it offers a convenient alternative to 68Ge/68Ga generators for the routine production of 68Ga. A new solution target with enhanced heat exchange capacity was designed and utilized in Example 1 with dual foils of aluminum (0.20 mm.) and Havar® cobalt alloy (0.038 mm.) separated by helium cooling to degrade the proton energy to ˜14 MeV. The water-cooled solution target insert was made of tantalum and its solution holding capacity (1.6 mL) was reduced to enhance heat transfer. An isotopically enriched (99.23%) 1.7 M solution of 68Zn nitrate in 0.2 N nitric acid was utilized in a closed target system. After a 30 minute irradiation at 20 μA, the target solution was unloaded to a receiving vessel and the target was rinsed with 1.6 mL water, which was combined with the target solution. An automated module was used to pass the solution through a cation-exchange column (AG-50W-X8, 200-400 mesh, hydrogen form) which efficiently trapped zinc and gallium isotopes. 68Zn was subsequently eluted with 30 mL of 0.5 N HBr formulated in 80% acetone without any measurable loss of 68Ga. 68Ga was eluted with 7 mL of 3 N HCl solution with 92-96% elution efficiency. The radionuclidic purity was determined using an HPGe detector. Additionally, ICP-MS was employed to analyze for non-radioactive metal contaminants. The product yield was 192.5±11.0 MBq/μ·h decay-corrected to EOB with a total processing time of 60-80 minutes. The radionuclidic purity of 68Ga was found to be >99.9%, with the predominant contaminant being 67Ga. The ICP-MS analysis showed small quantities of Ga, Fe, Cu, Ni and Zn in the final product, with 68Ga specific activity of 5.20-6.27 GBq/μg. Depending upon the user requirements, 68Ga production yield can be further enhanced by increasing the 68Zn concentration in the target solution and extending the irradiation time. In summary, a simple and efficient method of 68Ga production was developed using low energy cyclotron and a solution target. The developed methodology offers a cost-effective alternative to the 68Ge/68Ga generators for the production of 68Ga. Our initial attempts to reproduce the solution target method described by Jensen et al. [15] showed that irradiation of aqueous 68ZnCl2 solutions resulted in rapid pressure increase due to radiolysis-mediated release of hydrogen and oxygen. Additionally, we noticed a pinhole in the Havar® cobalt alloy target window foil, which may be related to a reaction of the Havar® cobalt alloy foil and with the ZnCl2 solution. We have recently performed an extensive study on the mechanistic aspects of water radiolysis during production of 89Zr in a solution target [Ref. 16-17]. This study showed that the use of nitrate salts in dilute nitric acid solutions dramatically decreased rates of water radiolysis during radiometal production [Ref. 17]. Eliminating the use of ZnCl2 could also prolong the life of the Havar® cobalt alloy foil. In the present work of Example 1, we extended our solution target approach to the production of 68Ga using a 1.7 M solution of zinc nitrate (isotopically enriched) in 0.2N nitric acid for the production of 68Ga in a closed solution target system using 20 μA beam current over 30 minute proton irradiation. We also report, an automated separation of 68Ga from 68Zn using cation-exchange resin. A BMLT-2 target (FIGS. 1A and 1B) was employed for 89Zr production as described previously. The target included a water-cooled, elongated Ta insert having target volume of 1.6 mL. The dual-foil design used helium cooling on a 0.2 mm aluminum degrader foil and a 0.025 mm Havar® target window foil. The target was installed on a PETtrace cyclotron (GE HealthCare, Waukesha, Wis.) delivering 16.4 MeV protons. The incident proton energy to the target solution was approximately to 14 MeV. A semi-automated system was employed as described previously for target loading and unloading [see Example 2 above]. Yttrium nitrate (Y(NO3)3.6H2O) was purchased from Strem Chemicals (Newburyport, Mass., USA). Oxalic acid dehydrate [TraceSELECT®, ≥99.9999% metals basis] was purchased from Sigma-Aldrich (St. Louis, Mo., USA). Nitric acid (70% purified by redistillation) ≥99.999% trace metals basis as purchased from Fisher Scientific (Suwanee, Ga., USA). Chelex-100 resin (50-100 mesh, sodium form) was purchased from Bio-Rad. Accell Plus CM (300 Å, WAT 010740) weak cation exchange resin (carboxylate resin) was purchased from Waters Inc. (Milford, Mass., USA). The phosphate buffer used for 89Zr elution was 1.2 M K2HPO4/KH2PO4 buffer (pH 3.5) prepared in-house using ≥99.999% trace metals basis K2HPO4 and KH2PO4, both purchased from Sigma-Aldrich (St. Louis, Mo., USA). Na2HPO4 and NaH2PO4≥99.999% trace metals basis were also purchased from Sigma-Aldrich (St. Louis, Mo., USA) For determining radionuclidic purity, a high-purity germanium gamma spectrometer (Canberra, Meriden, Conn., USA) counter running Genie 2000 software was used. Sample radioactivity was measured using a CRC dose calibrator (489 setting, CRC-55tPET, Capintec, Ramsey, N.J., USA). A Perkin Elmer ELAN DRC II ICP mass spectrometer was employed to analyze trace metal contaminants. IR spectra were recorded as KBr pellets using a ThermoNicolet Avatar 370 FT-IR (Waltham, Mass., USA). Specific activity (GBq/μg) of 68Ga was measured by estimating the total Ga metal present in the final product after purification. The other metal contaminants including Zn, Fe, Cu, Ni, and Ga were also analyzed using an ICP-mass spectrometer. For the measurement of radionuclidic purity we used a gamma ray spectrometer (Canberra, Meriden, Conn., USA, DSA1000) equipped with a high-purity germanium (HPGe) detector. Based on the preliminary study by Jensen et al [Ref. 15], our first irradiations used 1.7 M 68ZnCl2 in the solution target system. At a beam current of 20 μA, the target pressure rose rapidly to >150 psi within 5 minutes, and the run had to be aborted. This result was not surprising, as we had observed similar results in our previous studies with yttrium chloride [Ref. 16]. Based on our previous results, we switched to 68Zn(NO3)2 in 0.2 N nitric acid solution, pressurized with oxygen at ˜40 psi. Furthermore, due to the enhanced cooling of the new target design, the target was operated as a closed system. Under these conditions, the target was irradiated for 30 minutes at 20 μA. The in-target pressure was maintained below 100 psi over the run. To accomplish the separation of 68Ga from 68Zn, we modified the cation-exchange method developed by Strelow [Ref. 18] in the early 1980s for the separation of non-radioactive Ga salts from zinc and other metal salts. A column of 1.3 g of AG-50W-X8 resin was used. 68Ga was trapped at more than 99% efficiency. We did not observe loss of 68Ga during the elution of 68Zn with 0.5 N HBr in 80% acetone or during subsequent rinsing of the column with water. Finally, 68Ga was eluted using 3 N HCl solution (7 mL) with 92-96% elution efficiency. The elution efficiency can be increased by increasing the amount of 3 N HCl solution, but at the cost of increasing the volume of acid in the subsequent processing steps. The isotope separation process was automated using an in-house radiochemistry module with actuated valves. Production conditions of the solution target were 1.7 M 68Zn(NO3)2 in 0.2 N HNO3 solution at 20 μA beam current over a 30 minute irradiation. The production yield of 68Ga was found to be 192.5±11.0 MBq/μA·h after isotope separation having specific activity in the range of 5.20-6.27 GBq/μg (see Table 1). These values are decay-corrected to end of bombardment (EOB). TABLE 1Yields of 68Ga in solution target (n = 3)ICP-MS determinedUncorrectedCorrectedYieldSpecificSpecificquantities of Ga and otherYieldYield(MBq/activityactivitytrace metals in product(GBq)(GBq)μA · h)(GBq/μg)(mCi/μg)(μg, n = 2)0.96 ± 0.101.93 ± 0.11192.5 ± 11.05.20-6.27140-169Ga (0.32, 0.39), Fe (17.0,95.1), Cu (51.4, 12.6), Ni(3.5, 33.7), Zn (1.9, 1.9)these amounts were estimated for the total 7 mL of the product solution. At 2 hours after EOB, a purified sample of 68Ga was subjected to HPGe spectrometry and only two peaks were evident, at 511 keV and 1077 keV both corresponding to 68Ga. Therefore, the radionuclidic purity of 68Ga was >99.9% (see FIG. 3). However, when same sample was reanalyzed 36 hours after EOB additional peaks were observed corresponding 67Ga, which on decay correction to EOB, was estimated to be <0.07% of the total 68Ga (see FIG. 4). The Ga-67 emissions were predominated by the Ga-68 emissions at 2 hours after EOB. The presence of 67Ga impurity can be attributed to two possible nuclear reactions: 68Zn(p,2n)67Ga and 67Zn(p,n)yGa [Ref. 19, 20]. We would anticipate ˜0.1% 67Ga impurity in the 68Ga product for a 60 minute irradiation. Radiochemical purity, as measured by cation-exchange HPLC, was >99.9% (see FIG. 5). The saturation yields for 68Ga before and after the separation of isotopes were found to be 0.43±0.01 GBq/μA and 0.36±0.02 GBq/μA, respectively (see Table 2). We attempted to make a comparison with the saturation yield using the solid target method as reported by Sadeghi et al [Ref. 11]. These authors reported production yield as 5.032 GBq/μA·h, presumably without taking the isotope separation into consideration. The calculated saturation yield from their data is 35.3 GBq/μA. Thus, our yields were approximately 83 times less than the solid target yields. However, Sadeghi et al [Ref. 11] did not specify the time of isotope separation so we cannot compare isolated product yields. We would anticipate that the comparison of isolated yields would be somewhat more favorable for the solution target method because it does not require a target dissolution step. TABLE 2Saturation yields of 68GaBeforeBeforeAfterAfterseparationseparationseparationseparation(mCi/μA)(GBq/μA))(mCi/μA)(GBq/μA)11.5 ± 2.80.43 ± 0.019.8 ± 0.570.36-0.02n = 3 and all values decay corrected to EOB. Considering that the primary end-use of the 68Ga would be in for labeling of molecular targeted peptides, we also analyzed for the presence of trace metal contamination in the final product using ICP-MS. Although very low quantities of Ga were found (0.32-0.39 μg), relatively higher amounts of Fe, Cu and Ni were observed (see Table 1). If necessary, these contaminants can be further reduced before labeling using SnCl2/TiCl3 and Amberchrom CG-161m resin as described by Van der Meulen et al [Ref. 21]. Furthermore, in order to achieve the optimum labeling conditions (pH) for various chelators, the final 68Ga solution (7 mL, 3 N HCl) can be concentrated using an anion-exchange column, where [68GaCl4]− will be effectively trapped and can be further eluted with water as previously described [Ref. 1, 22]. The final pH can be readjusted using the buffer required to meet the labeling conditions. We estimate the cost of a single dose (370 MBq) of the 68Ga produced from a cyclotron using the 68Zn(NO3)2 in 0.2 N HNO3 solution target method to be $20-25 assuming 85-90% recovery of 68Zn, which we found in our initial attempts. The unoptimized separation time was in the range of 60-80 minutes, but we would anticipate reduction of this time to ˜45 min with further development. We were able to make ˜25 mCi (0.92 GBq)68Ga at end of separation. By increasing 68Zn nitrate concentration (2×) and beam current (2×), we anticipate that >100 mCi (3.7 GBq) can be practically achieved. This would serve the need for 2-4 subjects, depending on labeling yields. A solution target approach for production and automated separation of 68Ga was successfully developed employing a solution of 68Zn nitrate in 0.2 N nitric acid. The production yield was found to be 192.5±11.0 MBq/μA·h decay-corrected to EOB with a specific activity in the range 5.20-6.27 GBq/μg. Radiochemical and radionuclidic purities were both >99.9%. Increasing the target solution concentration of 68Zn and irradiation time may further increase the production yield. The isotope separation method employed AG 50W-X8 resin eluted with a solution of 0.5 N HBr in 80% acetone to remove the zinc isotopes, followed by elution of 68Ga in 3 N HCl. The new target design with reduced target volume (1.6 mL) and enhanced heat transfer allowed irradiation as a closed system. Gallium-68 can, therefore, be produced on a low energy cyclotron in sufficient quantities to provide a viable alternative to the 68Ge/68Ga generator for those facilities that have an on-site cyclotron. The existing solid target production method of radiometals requires high capital and operational expenditures, which limit the production of radiometals to the small fraction of cyclotron facilities that are equipped with solid target systems. In Example 2, we develop a robust solution target method, which can be applicable to a wide array of radiometals and would be simply and easily adopted by existing cyclotron facilities for the routine production of radiometals. We have developed a simplified, solution target approach for production of 89Zr using a niobium target by 14 MeV energy proton bombardment of aqueous solutions of yttrium salts via the 89Y(p,n)89Zr nuclear reaction. The production conditions were developed, following a detailed mechanistic study of the gas evolution. Although the solution target approach avoided the expense and complication of solid target processing, rapid radiolytic formation of gases in the target represents a major impediment in the success of solution target. To address this challenge we performed a systematic mechanistic study of gas evolution. Gas evolution was found to be predominantly due to decomposition of water to molecular hydrogen and oxygen. The rate of gas evolutions varied >40-fold depending on solution composition even under the same irradiation condition. With chloride salts, the rate of gas evolution increased in the order rank Na<Ca<Y. However, the trend was reversed with the corresponding nitrate salts, and further addition of nitric acid to the irradiating solution minimized gas evolution. At developed condition, 89Zr was produced in moderate yield (4.36±0.48 MBq/μA·h) and high effective specific activity (464±215 MBq/μg) using the solution target approach (2.75 M yttrium nitrate, 1.5 N HNO3, 2 hours irradiation at 20 μA). The novel findings of Example 2 on substrate dependent, radiation-induced water decomposition provide fundamental data for the development and optimization of conditions for solution targets. The developed methodology of irradiation of nitrate salts in dilute nitric acid solutions can be translated to the production of a wide array of radiometals like 64Cu, 68Ga and 86Y, and is well suited for short-lived isotopes. The PET radioisotope 89Zr has recently received growing interest due to its good imaging properties (Epmax 0.9 MeV), and its ease of conjugation to proteins and antibodies using the chelator desferrioxamine [Ref. 23-25]. The close match between the physical half-life and the clearance kinetics of antibodies has created a niche for this isotope [Ref. 26, 27]. 89Zr is commonly produced via the 89Y (p,n)89Zr nuclear reaction wherein Y foils [Ref. 28, 29], pellets [Ref. 30], sputtered materials [Ref. 31] or depositions [Ref. 32, 33] are irradiated with 12.5-15 MeV protons. The solid target approach entails high capital and operational costs for target processing, limiting the production of this isotope to the small fraction of cyclotron facilities that are equipped with solid target systems. A potential way to avoid use of solid target systems is to develop a solution target method for 89Zr production. Solution targets are easily filled and unloaded to hot cells via tubing in the same way that 18F targets are operated. Solution targets have been recently reported for the production of 86Y, 68Ga and 94mTc [Ref. 34, 35]. The major impediment in the development of solution targets is radiolytic formation of gases and subsequent generation of extremely high-pressures in closed target systems during irradiation, a fact previously reported by our group and others [Ref. 34-38]. In most cases, the irradiation times have been limited by this rise in target pressure. We have employed the use of a backpressure regulator to slowly bleed gas from the target system [Ref. 37, 38], but this has its drawbacks of potential loss of solution or water vapor from the target, and subverting the recombination of molecular hydrogen and oxygen formed by radiolysis. In Example 2, we investigated the mechanistic basis of the gas formation and how it could be minimized or even overcome to develop a robust solution target for the radiometal production. For target development, we selected 89Zr as target isotope based on its level of interest in the imaging community and the affordability of its yttrium reagent materials. We realized that 89Zr would represent a worst-case scenario for solution target applications because its longer half-life (T1/2=78.4 hours) would require long irradiations to produce significant amounts of radiometal and would not compare favorably to solid target production methods. Even so, the low cost for development efforts and the long irradiation requirement for 89Zr production would provide an appropriate situation to test a solution target for longer irradiations of other interesting isotopes. We anticipated that a production methodology developed for 89Zr could be translated to other shorter-lived isotopes such as 68Ga, 64Cu and 86Y. In Example 2, we investigated the in target chemistry associated with proton bombardment of aqueous solutions of yttrium salts and developed conditions for 89Zr production using Y(NO3)3 solutions. To maintain a constant target pressure, the gas was expelled through a backpressure regulator (60 PSI, Optimize Technologies, Oregon City, Oregon, USA) and evolved gases were collected in a volumetric syringe placed at the end of the line in a shielded hot shell. To avoid inaccuracies caused by the variable period of beam tuning on the gas evolution, we started the recording of gas volumes after one minute into the irradiation. Rate of gas evolution was averaged over the remainder of the irradiation period. Yttrium hydroxide was synthesized by reacting yttrium chloride (3.3 mmol, 1.0 g) in 3.0 mL of water with sodium hydroxide (11.5 mmol, 0.46 g) also in 3.0 mL of water initially at 0° C. for 15 minutes followed by 30 minutes at room temperature. Obtained precipitate was filtered and washed with 30 mL of deionized water to remove sodium chloride, unreacted sodium hydroxide and yttrium chloride. The desired precipitate was further freeze-dried to remove traces of water. FT-IR spectroscopy was performed using a KBr pellet (Thermo-Nicolet 370 FT-IR Avatar). The specific activity (MBq/μg) of 89Zr was calculated based on measurement of total amount of Zr present in the final 89Zr(HPO4)2 solution using ICP-MS. The sample was assayed for other trace metals including Al, Ga, Fe, Cu, Ni, Zn, Co, Pb, Y, Sc, Rh, Mg, Lu, In, and Ca but only Fe, Y, Al and Zr were found to be present at detectable levels. To determine the concentration of NO2 in the gas evolved from the irradiated target we utilized the “TGS-method” as adopted by the US Environmental Protection Agency (EPA). It is a colorimetric determination assay initially described by Mulik et al. and others [Ref. 40, 42]. The method utilized an NO2 absorption solution that was composed of the following reagents: 2% triethanolamine (w/v); 0.05% guaiacol/o-methoxyphenol (w/v); and 0.25% (w/v) sodium metabisulfite in water. Expelled gas from the solution target was bubbled through 30 mL of the absorption solution kept at 0° C. After completed bubbling and absorption of NO2, the solution was stored at 4° C. overnight for the decay of short lived radioisotopes (11C and 13N). To 1 mL of the solution was added 100 μl of diluted hydrogen peroxide solution (1 part 30% H2O2 to 125 parts water), and the solution was mixed well. To this solution, 500 μl of (2.15% (w/v)) sulfanilamide in 10% HCl was added. After vigorous mixing, 600 μl of the coupling reagent (2% (w/v) ammonium 8-anilino-naphthalene sulfonate in water) was added and the solution was mixed thoroughly. Absorbance (550 nm) was determined after letting the mixture stand at room temperature for 1 minute. The absorption solution was used as a blank. Calibration standards for NO2 concentrations were generated by bubbling specific volumes of commercially obtained 0.981% NO2 (Praxair, Rochester, Minn., USA) to 30 mL of absorption solution. Absorbance measurements for the standards defined a linear calibration curve. To determine the effective volume of NO2 released from the target, a linear regression was performed using the calibration data. The efflux gas was bubbled through three 30 mL vials connected in series containing deionized water at 0° C. The solutions were stored at 4° C. overnight for the decay of the short-lived radioisotopes (11C and 13N). The chlorine concentration (mg/mL) was measured according to the EMD-chlorine kit procedure. Evolved target gas was qualitatively analyzed for H2 using a hydrogen detector (SRI Instruments, Torrence, Calif., USA). A Clark-type oxygen probe detector (YSI Instruments, Yellow Springs, Ohio, USA) was used to quantify oxygen content in the evolved target gas. During an irradiation of a solution of 1.7 M Y(NO3)3 in 1 N HNO3, the evolved target gas was passed through a trap of 1 N NaOH before flowing through a chamber containing the oxygen probe submerged in a minimal volume of water. For the measurement of radionuclidic purity we used a high-purity Germanium (HPGe) radiation detection spectrometer (Canberra, Meriden, Conn., USA). All values are given as mean±standard deviation using Microsoft office excel program. During initial proton irradiations of aqueous Y(NO3)3 solutions (0.85-1.7 M), prolonged evolution of gases (>5 ml/min) and in-target precipitation of salts limited irradiations to less than 5 minutes at 25 μA. A backpressure regulator (0.41 MPa) was installed at the headspace of the target to allow release of gases while maintaining target pressure. The infra-red spectrum of the target precipitate was obtained and compared with that of synthesized Y(OH)3. The Y(OH)3 standard sample showed a broad peak at 3445 cm−1 that matched with the broad peak at 3495 cm−1 from the target precipitate. This confirmed the target precipitate as yttrium hydroxide. The infra-red spectra of Y(NO3)3 and commercially obtained Y2O3 were also compared with obtained precipitate but did not match (no OH stretching peak), ruling out the possibility of Y(NO3) and Y2O3. We embarked on a mechanistic study of the gas evolution and in-target precipitation that might inform a more optimal production strategy. Irradiation of yttrium nitrate solutions resulted in the evolution of NO2 as predicted from literature studies [Ref. 43, 44] (see Table 3). However, only small quantities (<0.1%) of NO2 were found. Irradiation of yttrium chloride solutions resulted in evolution of relatively small (<2.5%) fractions of Cl2 (see Table 4), however total gas evolution dramatically increased (>39 mL/min, see Table 5). Using a hydrogen detector and Clark-type oxygen probe detector, the predominant evolved gases were characterized as H2 and O2, respectively. The hydrogen concentration was not quantified. During an irradiation of 1.7 M Y(NO3)3 in 1 N HNO3, the evolved target gas was passed through a trap of 1 N NaOH before flowing through a chamber containing the oxygen probe submerged in a minimal volume of water. The oxygen concentration was found to be 33.3%±0.04% as consistent with the stoichiometric decomposition of water to molecular H2 and O2. TABLE 3Efflux of NO2 gas at different time intervals during 120 min proton irradiation of 2.75 M Y(NO3)3 solution at 20 μA beam currentAmount of NO2Amount of NO2Amount of NO2Amount of NO2Total amountRate of NO2Average ± sdevolved 0-30evolved 30-60evolved 60-9090-120of NO2 evolvedevolvedrate of NO2 evolvedRunmin (mL)min (mL)min (mL)min (mL)0-120 min (mL)(mL/min)(mL/min)10.523.404.905.4214.260.1180.093 ± 0.0422.013.863.863.8213.570.10930.130.190.684.795.790.048 TABLE 4Estimated chlorine (Cl2) evolution during 5 min YCl3 irradiation.Chlorine Chlorine Chlorine Total Total Rate of Average contentcontentcontentChlorineChlorineChlorineRate of ChlorineRunain Vial 1 (mg)in Vial 2 (mg)in Vial 3 (mg)content (mg)content (mL)content (mL/min)Evolution (mL/min)11.23.61.26.01.90.380.33 ± 0.0821.21.21.23.61.10.2331.23.61.26.01.90.38aIrradiation conditions: 1.7M YCl3, 1N HNO3, 5 min. 20 μA beam current. The rate of gas evolution depended on both cation and anion solutes and their concentrations (see Tables 5-6). A solution of 1.7M YCl3 produced 8-fold higher gas as compared to 1.7M Y(NO3)3. In contrast, NaCl produced one-third the rate of gas evolution compared to NaNO3 (see Table 5). The rate of gas evolution for calcium salts lay between Y and Na salts (see Table 5). Thus, a gradual increase in the rate of gas evolution was observed for chloride salts upon descending diagonally from Na to Y in periodic table (see FIG. 7). This trend was reversed for nitrate salts and led to decreases in the rate of gas evolution from Na to Y (see Table 5). The reversed trend strongly indicates disparate interactions of metal cations (Na+, Ca2+ and Y3+) and anions (Cl− and NO3−) in the solution target. The effect of changing salt concentration yielded mixed results. Increasing YCl3 concentration from 0.57 M to 1.7 M resulted in a 47% increase in gas evolution, while no effect was observed for the same increase in concentration of Y(NO3)3 (see Table 5). TABLE 5Rate of gas evolution of different saltsolutions at variable concentrationsMolar concentrationRate of gas evolution (mL/min)Yttrium (M3+)Y(NO3)3YCl30.57M6.22 ± 0.9626.66 ± 0.01 a 1.7M5.29 ± 0.07 39.22 ± 2.26 a, bCalcium (M2+)Ca(NO3)2CaCl20.85M8.75 ± 0.5010.67 ± 1.87c 1.7M 8.89 ± 0.69d19.16 ± 3.58 e, fSodium (M+)NaNo3NaCl1.7M18.78 ± 0.69g 4.89 ± 0.51 h.i25 μA beam current, 5 min irradiation time and n = 3a p value < 0.002 versus equimolar Y(NO3)3.b p value < 0.02 versus 0.57M YCl3.cp value < 0.05 versus 0.57M YCl3.dp value < 0.02 versus 1.7M Y(NO3)3.e p value < 0.05 versus 1.7M Ca(NO3)2.f value < 0.003 versus 1.7M YCl3.gp value < 0.001 versus 1.7M Y(NO3)3 and 1.7M Ca(NO3)2.h p value < 0.001 versus 1.7M NaNO3.i p value < 0.02 versus 1.7M CaCl2. In an attempt to reduce the rate of gas evolution via the radiolysis mechanism, the free radical scavengers nitric acid and ethanol were employed as additives (see Tables 6, 7, 8). Addition of 1 M nitric acid to the irradiating solutions resulted in 40%, 50% and 75% reductions in the rate of gas evolution for 1.7 M solutions of Y(NO3)3, YCl3 (see Table 6) and NaNO3, respectively (see Table 7). Furthermore, 48% and 50% reductions in rate of gas evolution were observed for 2.55M solutions of Ca(NO3)2 and CaCl2), respectively, on addition of 1 M nitric acid (see Table 6). Exceptionally, a 20% increase in gas evolution was observed on addition of 1 M of nitric acid to 5.1 M NaCl (see Table 6). This increase in gas evolution can be attributed to the chemical reaction of concentrated NaCl and nitric acid, which leads to the formation of HCl gas and NaNO3 [Ref. 45]. Overall, addition of 1M nitric acid resulted in significant reductions of gas evolution. When Y(NO3)3 concentration was kept at 1.7 M increasing the nitric acid concentration from 1 M to 2 M gave a 67% reduction in rate of gas evolution (see Table 8), but no significant difference was observed above 2 M. Furthermore, the addition of nitric acid eliminated the precipitation of salts within the target. This may be attributed to an immediate resolubilization of Y(OH)3 in nitric acid to Y(NO3)3, if indeed the former is produced in the radiation spur. Alternatively, the nitric acid may effectively scavenge the hydroxyl radicals before they interact with yttrium. The effect of nitric acid addition suggests radiolysis to be a dominant mechanism for gas evolution in the solution target. Addition of ethanol to both YCl3 and Y(NO3)3 resulted in rapid formation of precipitates in the target (see Table 7), precluding meaningful interpretation of the gas evolution rates. TABLE 6Effect of cation and nitric acid on the rate of gas evolution at same anion concentrations.Molar concentration of the target solutionRate of gas evolution (mL/min)aNo Acid1M HNO3No Acid1M HNO3Yttrium (M3+)5.29 ± 0.07 Y(NO3)33.17 ± 0.43v39.22 ± 2.26YCl318.72 ± 2.69v 1.7M18.72 ± .69v Ca(NO3)22.97 ± 0.29v20.00 ± 1.25CaCl210.00 ± 2.05vCalcium (M2+)5.75 ± 0.79 NaNO3 3.75 ± 0.25v10.08 ± 0.58NaCl12.75 ± 1.56 2.55M10.00 ± 2.05v Sodium(M+)14.25 ± 0.43 5.1Ma25 μA beam current, 5 min irradiation time and n + 3.vp value < 0.05 versus same salt without HNO3. TABLE 7Effect of ethanol on the rate of gas evolutionRate of gasRate of gaseffluxaeffluxaTarget solution(mL/min)Target solution(mL/min)1.7M YCl339.22 ± 2.261.7M YCl3 +41.0 (ppt)(n = 1)50 mM EtOH1.7M Y(NO3)3 5.29 ± 0.071.7M Y(NO3)3 +4.25 ± 0.63 (ppt)50 mM EtOH1.7M NaNO318.78 ± 0.691.7M NaNO3 +4.44 ± 0.53ν 1M HNO3H2O (no salt)No observed gas1M HNO35.83 ± 0.29ε a25 μA beam current, 5 min irradiation time and n = 3, ppt - observed precipitate.νp value < 0.05 versus 1.7M NaNO3.εp value < 0.001 versus pure water. TABLE 8Effect of nitric acid and yttrium nitrate concentrationon the rate of gas production.Rate of gasRate of gaseffluxeffluxTarget solution(mL/min)Target solution(mL/min)1.7M Y(NO3)3 +3.17 ± 0.430.7M Y(NO3)3 +3.63 ± 0.221.0M HNO31.5M HNO31.7M Y(NO3)3 + 1.75 ± 0.14a1.53M Y(NO3)3 + 2.11 ± 0.25b2.0M HNO31.5M HNO31.7M Y(NO3)3 +1.72 ± 0.352.75M Y(NO3)3 +1.54 ± 0.613.0M HNO31.5M HNO325 μA beam current, 5 min irradiation time and n = 3.ap value < 0.05 versus 1.7M Y(NO3)3 + 1N HNO3.bp value < 0.05 versus 0.7M Y(NO3)3 + 1.5M HNO3. The colligative properties of the solutions were examined by keeping the anion concentration the same while changing the cation. When chloride is kept constant at 5.1 M, NaCl (5.1 M), CaCl2) (2.55 M), and YCl3 (1.7 M) showed a pattern of increasing gas evolution (Na→Y) in the absence of acid (see Table 6). This pattern of gas evolution was not observed when the nitrate ion concentrations were kept the same (see Table 6). While no difference was observed between Y and Ca, the rate of gas evolution was paradoxically higher with Na. We do not have a clear explanation of this finding, but there appears to be an interaction between the various cations and anions. Since radiolysis-induced gas evolution may be at a minimum rate with nitrate salt solutions, the effect of cation may be accentuated. The radiolysis of water due to ionizing radiation is well documented in literature [Ref. 46, 47]. Mechanistic studies of water radiolysis have used kinetics, linear energy transfer characteristics and identification of free radicals [Ref. 48-66]. The primary radiation induced decomposition products of water are hydrogen and hydroxyl radicals:H2OH.+HO. (1)Eqs. (2a-d) describe the putative reactions for the formation oxygen and hydrogen gas from hydrogen and hydroxyl radicals:H.+H.→H2 (2a)H.+H2O→H2+HO. (2b)HO+2H2O2→2H2O+O2+HO. (2c)2H2O2→2H2O+O2 (2d)Irradiation of pure water with protons in a cyclotron target produced minimal quantities of gas, as evidenced by a small and stable increment in target pressure and no measurable gas efflux (see Table 7). As noted previously [Ref. 60, 64-66] in this situation the steady-state rate of formation of radicals must equal the rate of recombination of radicals (Eq. (1)). The presence of metal salt in the water may alter the balance of formation and recombination of radicals and thereby promote radiolysis. Thus, the formation of any chemical species utilizing either H. or OH. radicals will reduce their rate of recombination. Recent studies conducted by the Mostafavi group [Ref. 52-56] using picosecond pulse radiolysis on 2 M sodium chloride/bromide solutions showed the formation of Cl/BrOH.− intermediates via Eqs. (3) and (4) [Ref. 62, 63]. At 5-6 M halide concentrations, the predominant free radical formed is X2.− (X═Cl, Br) [Ref. 52, 53]. The high yield of XOH.− (X═Cl, Br) at 2 M concentrations suggests consumption of hydroxyl radicals by halide ions. We could not measure free radical intermediates in our studies, but our observations of high gas evolution rates with chloride salts, particularly of Ca and Y are consistent with this mechanism. In particular, the increase of chloride concentration as the solutes were changed from NaCl→CaCl2→YCl3 resulted in dramatic increases in gas evolution (see Table 5), putatively due to increase in the rate of hydroxyl radical scavenging to form the ClOH.− intermediate and thus decrease H. and OH. recombination. Increasing chloride salt concentrations while maintaining the same cation also increased the rate of gas evolution (see Table 5, FIG. 8a).HO.−+X−→XOH.−(X═Cl,Br) (3)HO−+X.→XOH.(x═Cl,Br) (4)Interestingly, when solutions of the chlorides of Na, Ca and Y were irradiated while maintaining the same chloride concentration (see FIG. 8b), differences in gas evolution rates were also observed, demonstrating the role of the metal cations as well. To further support the hydroxyl radical consumption hypothesis, when 1 N HNO3 was added to the solution of YCl3 the rate of gas evolution went down by 50% with no precipitation observed. This can be attributed to the high affinity of the hydroxyl species with the acidic protons (H+), thereby forming H2O.+ (also known as “hole”):HO.+HNO3→H2O.+NO3− (5a)According to the studies of Balcerzyk et al. [Ref. 52, 53], a hole can further interact with a nitrate ion to form a nitrate radical and water molecule (Eq. (5b)).H2O.++NO3−→NO3.+H2O(indirect effect-hole scavenging) (5b)Additionally, nitrate radicals can also be formed directly by irradiation of nitrate (Eq. (6)) or through conversion of nitrate to nitrite and the oxygen radical intermediate (.O—) (Eq. (7a) & (7b)) [Ref. 51, 52]:NO3−→NO3.+e−(direct) (6)NO3−→NO2−+.O.(direct) (7a)NO2−+.O−+H→NO3.+HO.(indirect) (7b)Finally, the nitrate radical can combine with hydrogen radical to form nitric acid, thereby competing with the formation of hydrogen gas:H.+NO3.→HNO3 (8) Thus, the presence of nitric acid effectively converts hydrogen and hydroxyl radicals back to water via the combination of Eqs. (1) and (5-8), although several free radical species (NO3., H2O.+, O—) may be involved as intermediates. Irradiation of nitrate salts in the absence of nitric acid allows greater interaction of hydroxyl radicals with metal cations resulting in increased gas evolution. However, hydroxyl radicals would be predicted to be increased in nitrate solutions according to Eqs. (6a and b) whereas hydroxyl radicals are consumed by chloride according to Eqs. (3 and 4), consistent with the markedly higher gas evolution rates observed with chloride salts compared to the nitrate salts. Nevertheless, the presence of nitrate ion per se appears to decrease water decomposition as shown by the comparison of gas evolution rates for increasing salt concentrations (see FIG. 8c). An increase in gas evolution was seen with increased chloride concentration (holding the cation concentration constant), while the opposite was seen for nitrate salts (see FIG. 8a versus 8c). As noted previously, there exist disparate metal ion (Na+, Ca2+, Y3+) to anion (Cl− and NO3−) interactions in the different solutions (see FIG. 8a-8d). FIG. 8 summarizes the trends found in gas evolution. Further mechanistic study may be required to elucidate the effects of different metal cations with anions in solution in response to proton irradiation. For example, pico-second pulse radiolysis experiments [Ref. 52-56] may be able to determine concentrations of critical intermediates that may provide more precise information about such interactions. Based on these findings, we developed the production conditions with 2.75 M Y(NO3)3 solutions to give 89Zr production yield of 4.36-4.55 MBq/μA·h before isotope separation with specific activity 464±215 MBq/μg at 20 μA beam current over 1-2 hours of irradiation (see Table 9). To avoid precipitation of target salts, the HNO3 concentration was increased from 1 M to 1.5 M as the duration of irradiation was increased from 1 hour to 2 hours. All runs were performed with 2.75M Y(NO3)3 and beam current of 20 μA. Yields were before isotope separation. Measurements of specific activity for seven runs encompassing the conditions are shown in Table 10. After 24 hours post-EOB (end of bombardment), a purified sample of 89Zr was analyzed on an HPGe gamma spectrometer. The analysis showed two photon peaks at 511 keV and 909 keV energies, both corresponding to 89Zr. No other peaks were observed, indicating that the radionuclidic purity of 89Zr was >99% (see FIG. 9). TABLE 9Developed yields of 89Zr in solution target.Irradiation TimeHNO3YieldYield(min)(M)(MBq)(MBq/μA · h)60 (n = 3)1.00 81.03 ± 14.44.55 ± 1.0790 (n = 3)1.25108.04 ± 17.34.47 ± 0.62120 (n = 3) 1.50153.18 ± 15.14.36 ± 0.48 TABLE 10Specific activities of purified 89ZrAverage ± Experimenta1234567sd (n = 7)Specific 192.4358.9418.1518666802.9292.3464 ± 215Activity(MBq/μg)a2.75M Y(NO3)3 + 1.5N HNO3., 120 min irradiation at 20 μA beam current. In summary, a solution target approach for production of 89Zr was successfully developed. The major obstacles in implementing a solution target for 89Zr production were identified as gas evolution due to radiation-induced water decomposition and in-target precipitation of salts. Gas evolution followed a consistent trend across Groups 1-3 metals, however, disparate interactions between the metal cations and anions (chloride versus nitrate) were observed. Gas evolution can be minimized significantly by addition of nitric acid and use of nitrate salts. The present study of Example 2 provides foundational information on the design and development of solution targets, and should be applicable for production of other radiometals such as 68Ga, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc. Applications for use of the radiometals include, without limitation, labeling of diagnostic and therapeutic compounds for medical purposes. Example 3 describes the production of 13N-ammonia with a solution target. For production of 13N-ammonia, the solution target was configured with a standard Havar® cobalt alloy window foil (40 μm thickness) without a degrader foil or helium cooling. The target was loaded with 1.6 mL 50 μM EtOH in deionized water and pressurized to 90 psi. The target was irradiated with 16.5 MeV protons (GE PETtrace) for 15 minutes at 35 μA beam current. Target pressure was stable during the run at ˜133 psi. The target liquid was unloaded to a hot cell, passing through a sterile filter (0.2 m) before collection in a product vial. The amount of 13N-ammonia produced was 513±29 mCi uncorrected for decay. The conventional silver target used for 13N-ammonia production yields approximately 330 mCi under the same irradiation conditions, however, in this case, the target is rinsed once with nonradioactive water to increase the transfer of the 13N-ammonia produced. Undesirable release of volatile 13N-labeled byproduct gases (NOx) were reduced from approximately 15 mCi with the silver target to approximately 3 mCi with the new target. Example 4 improves the cyclotron production yield of 89Zr using a solution target by developing an easy and simple synthesis of the hydroxamate resin used to process the target, and investigating biocompatible media for 89Zr elution from the hydroxamate resin. A new solution target with enhanced heat dissipation capabilities was designed by using helium-cooled dual foils (0.2 mm Al and 25 mm Havar®) and an enhanced water-cooled, elongated solution cavity in the target insert. Irradiations were performed with 14 MeV protons on a 2 M solution of yttrium nitrate in 1.25 M nitric acid at 40-μA beam current for 2 hours in a closed system. Zirconium-89 was separated from Y by use of a hydroxamate resin. A one-pot synthesis of hydroxamate resin was accomplished by activating the carboxylate groups on an Accell™ Plus CM cation exchange resin using methyl chloroformate followed by reaction with hydroxylamine hydrochloride. After trapping of 89Zr on hydroxamate resin and rinsing the resin with HCl and water to release Y, 89Zr was eluted with 1.2 M K2HPO4/KH2PO4 buffer (pH 3.5). ICP-MS was used to measure metal contaminants in the final 89Zr solution. The new target produced 349±49 MBq (9.4±1.2 mCi) of 89Zr at the end of irradiation with a specific activity of 1.18±0.79 GBq/μg. The hydroxamate resin prepared using the new synthesis showed a trapping efficiency of 93% with a 75 mg resin bed and 96-97% with a 100-120 mg resin bed. The elution efficiency of 89Zr with 1.2 M K2HPO4/KH2PO4 solution was found to be 91.7±3.7%, compared to >95% for 1 M oxalic acid. Elution with phosphate buffer gave very small levels of metal contaminants: Al=0.40-0.86 μg (n=2), Fe=1.22±0.71 μg (n=3), Y=0.29 μg (n=1). The new solution target doubled the production quantity of 89Zr from the previously reported 153±15 MBq to 349±49 MBq for the same 2 hour of irradiation and 40 μA beam current. The new one-pot synthesis of hydroxamate resin provides a simpler synthesis method for the 89Zr trapping resin. Finally, phosphate buffer elutes the 89Zr from the hydroxamate resin in high efficiency while at the same time providing a more biocompatible medium for subsequent use of 89Zr. Zirconium-89 has gained considerable attention for labeling antibodies due to its favorable PET imaging characteristics (β+max—0.9 MeV, 22.7%) and a half-life (T1/2—78.4 hours) that matches the biological half-life of antibodies. Several monoclonal antibodies have been labeled with 89Zr using the bifunctional chelator desferrioxamine (DFO) and are currently under clinical investigation for diagnostic imaging and therapeutic monitoring applications. For example, 89Zr-DFO-cetuximab is an anti-epidermal growth factor receptor monoclonal antibody labeled with 89Zr that is being investigated for PET imaging of metastatic colorectal cancer [Ref. 67], while 89Zr-DFO-bevacizumab is an anti-vascular endothelial growth factor-A monoclonal antibody labeled with 89Zr that is being investigated for evaluation of lesions in patients with Von Hippel-Lindau disease [Ref. 68], renal cell carcinoma [Ref. 69], breast cancer [Ref. 70], and neuroendocrine tumors [Ref. 71]. Zirconium-89-DFO-trastuzumab is a radiolabeled anti-HER2 monoclonal antibody currently under investigation for breast cancer imaging [Ref. 72]. Various other monoclonal antibodies are being investigated for PET imaging of pancreatic, ovarian and prostate cancers [Ref. 73]. Zirconium-89 has also been investigated for cell labeling using 89Zr-protamine sulfate [Ref. 74], 89Zr-oxine [Ref. 75, 76], and 89Zr-desferrioxamine-NCS 89ZrDBN [Ref. 77]. The 78.4 hour half-life of 89Zr allows in-vivo tracking of the labeled cells for prolonged periods of time (weeks), which is being instrumental in understanding the cell-based therapies and imaging. Zirconium-89 can be produced using solid and liquid target methods via the 89Y(p,n)89Zr reaction [Ref. 78-80]. Solid target methods require higher infrastructure cost because of the expense of the target handling system and are challenging to implement in PET facilities equipped with self-shielded cyclotrons. However, the solution target approach offers a viable alternative to the solid target system that can meet the modest needs for 89Zr at most institutions. Production of radiometals in a solution target is challenging because of in-target salt precipitation and unstable target pressures caused by gas evolution during irradiation [see Example 2 above], but these problems have been attenuated by using nitrate salts of the target metal in dilute nitric acid solutions [see Example 2 above]. We reported a mechanistic study on the effect of solution composition on in-target chemistry [see Example 2 above], the effect of nitric acid [see Example 2 above], and a second-generation solution target (Brigham Mayo Liquid Target-2 or BMLT-2) design that enhances heat dissipation during production of 68Ga [FIGS. 1A and 1B described above]. In the present study of Example 4, we investigated the use of this BMLT-2 target for the production of 89Zr. A significant improvement in the 89Zr-production yield compared to previous results [see Example 2 above]. We also developed a new method of elution of 89Zr from hydroxamate resin that offers a biocompatible formulation of 89Zr for preclinical and clinical studies. Finally, we also developed a short, simplified and one pot synthesis of the hydroxamate resin used for the separation of 89Zr from the 89Y target material. The hydroxamate resin was synthesized by stirring the Accell™ Plus CM cation exchange carboxylate resin (2.00 g), methyl chloroformate (2.0 mL, 25.8 mmol) and triethylamine (2.0 mL, 14.3 mmol) in anhydrous dichloromethane (30 mL) at 0° C. for 30 minutes and then at room temperature for additional 90 minutes. The temperature of the mixture was further lowered to 0° C. before addition of hydroxylamine hydrochloride (0.6 g, 8.63 mmol) and triethylamine (2.0 mL, 14.3 mmol). The resultant mixture was stirred at room temperature for an additional 15 hours. The solvent was removed under vacuum, and cold water was poured with constant stirring into the flask containing the functionalized resin. The resin was filtered, washed extensively with water, and dried under vacuum. Total amount of hydroxamate resin recovered was 1.9-2.0 g. Hydroxamate resin was packed into an empty cartridge having plastic frit on both the sides. The resin was activated with acetonitrile (8 mL), water (15 mL), and hydrochloric acid (2 mL 2 M) as described previously [Ref. 79, 80] prior to testing the trapping efficiency. Different amounts of hydroxamate resin were evaluated for their efficiency in trapping of 89Zr. The conversion of carboxylate group to a hydroxamate group on the resin was also characterized by infrared (IR) spectrometric analysis. Separation of 89Zr from the irradiated Y(NO3)3/HNO3 solution was achieved by trapping the 89Zr on a custom-made hydroxamate column as described previously [Ref. 80]. The Y(NO3)3 target material was removed from the column by rinsing the resin with 2 M HCl (75 mL) and water (10 mL). Zirconium-89 was eluted from the resin using 1 M oxalic acid or a new elution reagent composed of 1.2 M K2HPO4/KH2PO4 (pH 3.5) (phosphate buffer). For the latter, 89Zr elution was performed using a 2-step procedure whereby after rinsing with 2 M HCl followed by water to remove the Y target material, the 89Zr-containing resin was first wetted with ˜0.25 mL of phosphate buffer and allowed to interact on the column for 30-min before eluting the 89Zr with an additional 1.75 mL of phosphate buffer. The eluted 89Zr is in the form of 89Zr-hydrogen phosphate (89Zr(HPO4)2). The previously described synthesis of the hydroxamate resin is a cumbersome process involving multiple reactions, purifications, and use of hazardous chemicals such as 2,3,5,6-tetrafluorophenol [Ref. 79, 82, 83]. To address the need for a more practical and economical preparation of hydroxamate resin we developed a simplified, one-pot synthesis method using the same starting carboxylate resin. The carboxylate group on the resin was activated by methyl chloroformate and then converted into a hydroxamate functionality using hydroxylamine hydrochloride in presence of triethylamine [Ref. 84]. The synthesis method was optimized for 2 g scale. We characterized the hydroxamate resin by both infrared spectroscopic analysis and 89Zr trapping following the literature approach [Ref. 79]. The asymmetric and symmetric stretching frequencies of the carbonyl of (>C═O) of the carboxylate moiety appeared at 1571.6 and 1403.7 cm−1. The relative intensity of both the peaks was decreased compared to the non-functionalized resin while additional peaks at 1726 cm−1 and 1672 cm−1, characteristic of the asymmetric and symmetric stretching frequencies of hydroxamate (—CONHOH) appeared in the hydroxamate functionalized resin compared to starting carboxylate resin [Ref. 79]. Additionally, a medium to strong band at 1550 cm−1 for —N—H deformation and C—N stretching vibration for amide II bands were reported for hydroxamic acids [Ref. 85]. In our case, we also observed a strong band at 1552 cm−1, confirming the presence of the hydroxamate group. The resin was tested for 89Zr trapping efficiency with different bed loads. A trapping efficiency of 93% was obtained with a 75 mg bed load. Higher amounts of resin (100-120 mg) increased the trapping efficiency to 96-97%, but larger buffer volumes were then needed to elute the 89Zr, which resulted in more dilute final solutions (see below). Thus 75 mg of resin was found to be optimal to obtain a highly concentrated solution of 89Zr-hydrogen phosphate or 89Zr-oxalate. Oxalic acid has been extensively used for elution of 89Zr from the hydroxamate resin [Ref. 79, 82, 83]. However, the toxicity of oxalic acid prevents this solution to be used directly in biological systems [Ref. 79]. Holland et al. [Ref. 79] described a method to convert 89Zr-oxalate to 89ZrCl2 using a QMA cartridge (anion exchange resin) followed by elution with 0.9% saline solution or 1.0 M hydrochloric acid. Hydrochloric acid gave good elution efficiency but required 1100° C. evaporation step prior to the further use of the 89Zr, whereas 0.9% saline solution gave only 22-38% elution efficiency [Ref. 79]. There is, therefore, a need for an improved 89Zr elution/formulation strategy. To address this issue we employed phosphate buffer to elute the 89Zr from the hydroxamate resin. The elution efficiency was optimized by evaluating various concentrations of phosphate buffers (Table 11), and 1.2 M K2HPO4/KH2PO4 at pH 3.5 was found to be optimal, producing 91.7±3.7% elution efficiency using a total of 2 mL of buffer after an initial 30 minutes of equilibration on the column. Elution was performed in four 0.5 mL aliquots to provide an increased concentration of 89Zr in one of the four fractions. For comparison to the conventional approach, a 1 M-oxalate solution (1 mL) was also utilized to elute 89Zr. The metal contaminants were analyzed in both the 89Zr-hydrogen phosphate and 89Zr-oxalate solutions, and found that both solutions contain similar small amounts of Al, Fe and Y (see Table 12). Indeed, we have utilized this formulation strategy to effectively label stem cells without the need to reformulate 89Zr eluted in oxalic acid [Ref. 77]. TABLE 1189Zr elution efficiencies from hydroxamateresin as a function of eluent.EluentElution Efficiency1M H3PO467-84%1M NaH2PO440-90%2M (pH 3.5) NaH2PO473-94%1M (pH 7.0) Na2HPO4/41-93%NaH2PO41.2M (pH 3.5)91.7 ± 3.7% (n = 8)K2HPO4/KH2PO4, TABLE 12Metal impurities in the final 89Zr product as a function of eluent.Metal contaminants inMetal contaminants in89Zr-hydrogen phosphate eluent89Zr-oxalate eluentAl = 0.40-0.86 μg (n = 2),Al = 2.39-5.38 μg (n = 2),Fe = 1.22 ± 0.71 μg (n = 3),Fe = 1.10-1.20 μg (n = 2),Y = 0.29 μg (n = 1)Y = 0.50 μg (n = 1)Total 4 samples were analyzed for a wide range of metal impurities (listed in Example 4 method section) but detectable contaminants were present only in few samples and their numbers are written in the parenthesis.*Total 2 samples were analyzed for a wide range of metal impurities (listed in Example 4 method section) but detectable contaminants were present only in few samples and their numbers are written in the parenthesis. The enhanced heat dissipation capacity of the BMLT-2 target (FIGS. 1A and 1B) resulted in stabilized target pressure and avoidance of in-target precipitation issues during 89Zr production for beam currents up to 40 μA. We used 2 M yttrium nitrate in a 1.25 M nitric acid to minimize in-target precipitation. Initially, we tested different beam currents ranging from 10-40 μA by monitoring the change in-target pressure in a closed target system. Once we established that target pressure reached a steady state by 10-20 minutes of irradiation, we extended the irradiation time 2 hours. The production rate per mA was increased by approximately 12% from 3.82±0.37 MBq/μA·h to 4.36±0.60 MBq/μA·h, but due to the doubling of beam current the overall activity produced was found to be more than double in same 2 hours of irradiation from 153.18±15.1 MBq [see Example 2] to 348.88±48.7 MBq (Table 13). The specific activity of the 89Zr was not changed significantly from 464±215 MBq/μg [see Example 2] to 1186±799 MBq/μg. Analysis of metal contaminants in the 89Zr product solution was carried out using ICP-MS (Table 14), and only small amounts of Al (2.28±1.23 μg), Fe (1.90±1.23 μg), and Y (0.29±0.21 μg) were found. All other metals were below detection limits. TABLE 1389Zr production rates in first and second generation solution targets.MolarDecayconcentrationBeamCorrectedProductionof Y(NO3)3CurrentHNO3EOB activityrate(M)(μA)(M)(MBq)(MBq/μA · h)TS-16502.75201.50153.1 ± 15.13.82 ± 0.37target (4.14 ± 0.4 mCi)(n = 3)BMLT-22.00401.25348.8 ± 48.74.36 ± 0.60target 9.4 ± 1.2 mCi)(n = 7)data from Example 2.All irradiations were performed for 2 hours. Yields were calculated before isotope separation at end of beam. TABLE 14Specific activities of purified 89Zr in firstand second generation solution targets.Specific Activity of 89Zr producedSpecific Activity of 89Zr inusing the TS-1650** solution targeta BMLT-2 (FIGS. 1A and 1B)(n = 7) (MBq/μg)target (n = 3)* (MBq/μg)464 ± 2151186 ± 7992.75M Y(NO3)3 in 1.5N HNO3, 2 hours irradiation at 20 μA beam current.2.0M Y(NO3)3 in 1.25N HNO3, 2 hours irradiation at 40 μA beam current.**A F-18 target designed by Bruce Technologies (see Example 2) A purified sample of 89Zr was analyzed on an HP—Ge gamma spectrometer for radionuclidic purity. The energy spectrum showed two photon peaks at 511 keV and 909 keV energies, both corresponding to 89Zr. No other peaks were observed, indicating that the radionuclidic purity of 89Zr was >99% (see FIG. 10). In summary, the newly designed BMLT-2 target of FIGS. 1A and 1B increased the production rate of 89Zr by approximately 12% from 3.82±0.37 MBq/μA·h to 4.36±0.60 MBq/μA·h; but since the new target increased current capacity 2-fold, the 89Zr yield for the same 2 hour irradiation increased by more than 2-fold from 153.18±15.1 MBq to 348.88±48.7 MBq. But, the specific activity of the 89Zr did not change significantly from 464±215 MBq/μg to 1186±799 MBq/μg. Additionally, a simple one-pot synthesis of the hydroxamate resin and a new method of elution of 89Zr from the hydroxamate resin with a phosphate buffer were developed to improve the practicality of 89Zr production and provide a final formulation in a more biologically acceptable medium than the previously used oxalic acid solution. [1] Velikyan I., Prospective of 68Ga-Radiopharmaceutical Development. Theranostics 2014; 4: 47-80. [2] Banerjee S R, Pomper M G., Clinical applications of Gallium-68. Appl Radiat Isot 2013; 76: 2-13. 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[84] Pandey M K, DeGrado T R, Qian K, Jacobson M S, Clinton H E, Duclos Jr. R I, Gatley S J., Synthesis and Preliminary Evaluation of N-(16-18F-fluorohexadecanoyl)ethanolamide (18F-FHEA) as a PET Probe of N-Acylethanolamine Metabolism in Mouse Brain. ACS Chem Neurosci, 2014; 5(9):793-802. [85] Socrates G., Infrared and Raman characteristic group frequencies: Tables and Charts. 3rd ed. Wiley, 2004. The citation of any document or reference is not to be construed as an admission that it is prior art with respect to the present invention. Thus, the present invention provides methods of producing and isolating 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc and solution targets for use in the methods are disclosed. The methods of producing 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc include irradiating a closed target system with a proton beam. The closed target system can include a solution target. The methods of producing isolated 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc further include isolating 68Ga, 89Zr, 64Cu, 63Zn, 86Y, 61Cu, 99mTc, 45Ti, 13N, 52Mn, or 44Sc by ion exchange chromatography. An example solution target includes a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam at the target cavity; a target window for covering an opening of the target cavity; and a coolant flow path disposed in the passageway upstream of the target window. Although the invention has been described with reference to certain embodiments, one skilled in the art will appreciate that the present invention can be practiced by other than the described embodiments, which have been presented for purposes of illustration and not of limitation. Therefore, the scope of the appended claims should not be limited to the description of the embodiments contained herein.
053234294	claims	1. A method for assessing deterioration of penetrations traversing a vessel wall of a nuclear reactor pressure vessel defining an enclosure for an electrolyte, comprising the steps of: instrumenting at least one of the penetrations by installing a probe, the probe having a plurality of electrodes disposed in an array and positioned adjacent and electrically insulated from metal defining a penetration wall, the probe including conductors coupled individually to respective ones of the electrodes and a conductor coupled to the metal defining the penetration wall such that the penetration wall functions as a working electrode at distinct positions adjacent the respective electrodes; exposing both the electrode and the penetration wall to the electrolyte in the enclosure; monitoring electrical activity between the electrode and the penetration wall to develop a signal representing at least one of electrochemical potential, electrochemical impedance, and current passing through the electrolyte between the electrodes and the penetration wall, said electrical activity varying with corrosion of the penetration wall at said distinct positions; and, assessing deterioration of the penetration wall as a function of the electrical activity. a reactor vessel arranged to enclose a quantity of nuclear fuel and a coolant forming an electrolyte, the reactor vessel having at least one wall traversed by a plurality of penetrations for coupling external devices to internal structures of the reactor vessel, the penetrations defining conduits and being subjected to stress-corrosion damage due to operation of the nuclear reactor; an array comprising a plurality of electrochemical sensors mounted in at least one of the penetrations, the electrochemical sensors each having an electrode insulated from a surface of the at least one of the penetrations and being exposed together with said surface to the electrolyte during operation of the nuclear reactor, said surface and said electrode forming an electrochemical sensing cell wherein said surface functions as a working electrode at a distant area of the surface cooperating with said electrode of the respective electrochemical sensor such that each said electrochemical cell is responsive to localized corrosion of the surface at the distinct area; signal conductors coupled to the electrode and to the surface of said at least one of the penetrations, the signal conductors developing voltage and current signals as a function of electrochemical activity leading to the stress-corrosion damage; a detector circuit coupled to the signal conductors and being operable to encode data representing at least one of electrochemical potential, electrochemical impedance, and current passing through the electrolyte between the electrode and the surface with said localized corrosion of the surface; and, means for reading out said data for assessing deterioration of the penetration wall as a function of the electrochemical activity. 2. The method according to claim 1, wherein said electrical activity is monitored during operation of the nuclear reactor pressure vessel. 3. The method according to claim 1, wherein the reactor pressure vessel has a plurality of said penetrations traversing the vessel wall and further comprising selecting from the plurality of penetrations said at least one penetration that is instrumented and estimating corrosion in at least one other of the penetrations from deterioration of the penetration wall adjacent the electrodes. 4. The method according to claim 2, wherein the penetration traverse the reactor pressure vessel at an angle relative to a plane of the reactor pressure vessel at the penetration, whereby the penetration is vulnerable to stress corrosion cracking at an area adjacent the reactor vessel, and further comprising placing the electrodes at least in said area. 5. The method according to claim 1, wherein the penetration comprises a tube attached to the reactor pressure vessel at a substantially circular opening through the vessel wall, and further comprising placing a plurality of electrodes at various positions adjacent discrete areas on at least at one of an internal surface and an external surface of the penetration tube. 6. The method according to claim 5, wherein the tube protrudes internally from the vessel wall and is attached to the vessel wall by welds, and wherein the electrodes are placed in the tube adjacent the welds. 7. The method according to claim 1, further comprising analyzing the signal for a level of electrochemical noise by monitoring for variations in at least one of electrochemical potential and electrochemical current at a range of frequencies. 8. The method according to claim 7, comprising placing at least two said electrodes as a pair in the penetration wall, one of said pair functioning as a reference electrode, and monitoring for variations in said electrochemical potential and said electrochemical current using said at least two electrodes. 9. The method according to claim 1, comprising instrumenting the at least one penetration with an inserted probe having a plurality of electrode pairs, the electrodes in each pair being positioned adjacent one another and close to a distinct area of the penetration wall, each of said electrodes and the penetration wall forming a cell comprising a reference electrode and two working electrodes, and wherein said monitoring and assessing steps include separately monitoring and assessing corrosion in respective distinct areas for each of the cells. 10. A stress-corrosion monitoring apparatus for a nuclear reactor, comprising: 11. The stress-corrosion monitoring apparatus according to claim 10, wherein the at least one penetration traverses a head structure of the reactor pressure vessel. 12. The stress-corrosion monitoring apparatus according to claim 11, wherein the penetration comprises a tubular conduit traversing a wall defined by the head structure of the reactor pressure vessel, the tubular conduit having a longitudinal axis disposed at an acute angle relative to a plane of the wall of the head structure at the penetration, and is attached to the wall of the head structure adjacent an internal end of the penetration, and wherein the array encompasses an area of the tubular conduit at the head structure. 13. The stress-corrosion monitoring apparatus according to claim 12, wherein the penetration is one of a plurality of penetrations of the head structure of the reactor pressure vessel, the penetrations coupling at least one of a control rod positioning mechanism and a sensor signal cable through the head structure to internal structures of the reactor pressure vessel. 14. The stress-corrosion monitoring apparatus according to claim 10, wherein each of the cells comprises a reference electrode and two working electrodes defined by the penetration surface and two said electrodes, positioned to interact through the electrolyte with each said distinct area. 15. The stress-corrosion monitoring apparatus according to claim 14, wherein the electrodes for the respective distinct areas are spaced circumferentially around a sensor probe inserted into the penetration, whereby the sensor signals represent circumferentially spaced areas of the penetration surface. 16. The stress-corrosion monitoring apparatus according to claim 15, wherein the electrodes for the respective distinct areas also are spaced axially on the sensor probe. 17. The stress-corrosion monitoring apparatus according to claim 14, wherein the detector circuit is operable to develop potential noise data from a potential between the penetration surface and one of the electrodes, and current noise data from a current between the penetration surface and another of the electrodes.
	abstract	A container is designed to safely store radioactive debris. The container has an overpack having an elongated body extending between a top end and a bottom end. A basket is situated inside of the overpack. The basket has elongated canisters. Each of the canisters has an elongated body extending between a top end and a bottom end. At least one of the canisters has an insert with a plurality of elongated perforated tubes that contain radioactive debris. The perforations enable gas flow, primarily air, through the side wall to enable evaporation of liquid, primarily water, from the radioactive debris, by increasing the exposed surface area of the debris.
050135201	claims	1. Apparatus for the insertion of mutually parallel, elongated fuel rods into an elongated can having a rectangular cross section, a longitudinal direction and a lateral transverse slit formed therein, comprising a holder for holding a can, a fuel rod positioning arm having an insertion end for insertion through the slit in the can in an insertion direction at right angles to the longitudinal direction of the can, a support structure attached to said insertion end of said arm for supporting fuel rods, said support structure having a pivot axis at right angles to the insertion direction of said arm and to the longitudinal direction of the can about which said support structure is pivotable back and forth within a given pivot angle, said support structure having a jacket surface with two fuel rod support surfaces being curved outwardly about said pivot axis and offset alongside one another in the direction of said pivot axis, one of said support surfaces merging from a first segment with a relatively shorter radial spacing from said pivot axis than the other of said support surfaces, into a second segment with a relatively greater radial spacing from said pivot axis than the other of said support surfaces, and said radial spacing of one of said two support surfaces from said pivot axis increasing steadily within said given pivot angle, and said radial spacing of the other of said two support surfaces from said pivot axis decreasing steadily, as seen in the same circumferential direction. 2. Apparatus according to claim 1, including a resilient restoring element, said support structure and said holder being displaceable relative to one another in a direction parallel to said insertion direction of said arm counter to said resilient restoring element. 3. Apparatus according to claim 1, wherein said arm with said support structure and said holder are displaceable relative to one another in the longitudinal direction of the can. 4. Apparatus according to claim 1, wherein the can has another lateral transverse slit formed therein, and including a counterpart structure disposed on said holder opposite said support structure for insertion through the other slit in the can, said counterpart structure having toothing defining gaps in the longitudinal direction of the can. 5. Apparatus according to claim 4, wherein said counterpart structure is an elongated, rotatable eccentric shaft disposed transversely to the longitudinal direction of the can, said shaft having an outer jacket surface with grooves formed therein in circumferential direction forming said toothing. 6. Apparatus according to claim 4, wherein said counterpart structure has a support surface for retention of the can. 7. Apparatus for the insertion of mutually parallel, elongated fuel rods into an elongated can having a rectangular cross section, a longitudinal direction and a lateral transverse slit formed therein, comprising a holder for holding a can, a fuel rod positioning arm having an insertion end for insertion through the slit int eh can in an insertion direction at right angles to the longitudinal direction of the can, a support structure attached to said insertion end of said arm for supporting fuel rods, said support structure having a fuel rod support surface disposed thereon, and having a pivot axis at right angles to the insertion direction of said arm and to the longitudinal direction of the can about which said support structure is pivotable back and forth within a given pivot angle.
	claims	1. A compression sleeve useable in a nuclear reactor, comprising:a single-piece tubular body,the tubular body having a first end and a second end,the first end shaped to surround an end of a jet pump sensing line,the second end shaped to surround a replacement line portion for the jet pump sensing line,at least one of the first and second ends being a tapered, malleable end configured to deform to provide a seal between the compression sleeve and the respective jet pump sensing line or replacement line portion,the tapered, malleable end having a first, second, and third section, the second section having an outer surface with a concave, longitudinal cross-sectional profile, the second section being between the first and third sections, the longitudinal cross-sectional profile existing along a longitudinal length of the compression sleeve. 2. The sleeve of claim 1, wherein the jet pump sensing line and the replacement line portion are in a Boiling Water Reactor (BWR). 3. The sleeve of claim 1, wherein the tapered, malleable end includes grooves on an inner surface forming a crenulated pattern configured to form a labyrinth seal with the respective jet pump sensing line or the replacement line portion. 4. The sleeve of claim 1, wherein,thicknesses of each of the first, second, and third section are not equal and are configured to permit deformation of the first, second, and third sections while substantially preventing deformation to portions of the tubular body other than the first, second, and third sections. 5. The sleeve of claim 4, whereinthe first section is closest to the tubular body and tapers continuously from the tubular body at an angle,the third section is farthest from the tubular body and tapers at about the same angle, andthe concave cross-sectional profile of the second section is continuous between the first section and third section. 6. The sleeve of claim 1, wherein the tubular body has an inner diameter of about 0.01 inches larger than an outer diameter of a connected component. 7. The sleeve of claim 1, wherein the tubular body is made of about 99% nickel, and is heat annealed at about two-thousand (2000) degrees Fahrenheit for about 15 minutes. 8. A mechanical coupling assembly for repairing a sensing line of a jet pump in a nuclear reactor, comprising:a compression sleeve including,a single-piece tubular body,the tubular body having a first end and a second end,the tubular body being malleable, heat-annealed, and about 99% nickel,the first end shaped to surround an end of a jet pump sensing line;a replacement tube, the second end shaped to surround the replacement tube,the first and second ends being tapered, malleable ends configured to deform to provide a seal between the compression sleeve and the respective jet pump sensing line and replacement tube,the tapered, malleable ends each having a first, second, and third section, the second section having an outer surface with a concave, longitudinal cross-sectional profile, the second section being between the first and third sections, the longitudinal cross-sectional profile existing along a longitudinal length of the compression sleeve;a collar configured to join the compression sleeve and at least one of the replacement tube and the end of the jet pump sensing line; andat least one compression nut configured to mate with the collar to compress and collapse the concave second section of the compression sleeve against an outer surface of at least one of the replacement tube and the end of the jet pump sensing line. 9. The mechanical coupling assembly of claim 8, wherein the replacement tube replaces a segment of the sensing line. 10. A compression sleeve used in a pipe repair within a nuclear reactor, comprising:a single-piece tubular body having a first end and a second end,wherein the tubular body is made of about 99% pure Nickel,the first end shaped to surround an end of a jet pump sensing line,the second end shaped to surround a replacement line portion for the jet pump sensing line; anda reduced-thickness, circumferential segment having an outer surface with a concave, longitudinal cross-sectional profile at each of the first and second ends to facilitate deformation of the ends when connected to one of the jet pump sensing line and the replacement line portion so as to provide a seal between the connected component and sleeve, the longitudinal cross-sectional profile existing along a longitudinal length of the compression sleeve. 11. The sleeve of claim 10, wherein one or both of the ends includes a plurality of spaced circumferential grooves on an inner surface forming a crenulated pattern configured to form a labyrinth seal with a component that is to be received therein. 12. The sleeve of claim 10, wherein each of the ends is a tapered section that forms a conical, angled seating surface at each end. 13. The sleeve of claim 12, wherein the conical, angled seating surface is at an approximately 10 degree angle with a horizontal axis of rotation of the sleeve. 14. The sleeve of claim 1, wherein the concave cross-sectional profile of the second section is configured to allow a fastening device to fit over and collapse the second section. 15. The sleeve of claim 1, wherein the second section is configured to compress and collapse against an outer surface of at least one of the replacement line portion and the end of the jet pump sensing line when a compression force is applied to the outer surface of the second section, the compression force being in a direction that is about perpendicular to the outer surface of the second section. 16. The sleeve of claim 10, wherein the reduced-thickness, circumferential segment is configured to compress and collapse against an outer surface of at least one of the replacement line portion and the jet pump sensing line.
	claims	1. A low profile transport apparatus for transporting radioactive waste, the apparatus comprising:a cylindrical annular body comprising a vertical axis, an open top end, an open bottom end, and a circumferentially-extending sidewall defining a cavity configured for holding a cask containing radioactive waste;a top plate attached proximate to the top end of the annular body;a bottom plate attached proximate to the bottom end of the annular body;the top and bottom plates each extending perimetrically around the annular body and projecting laterally outwards therefrom;a plurality of support assemblies spaced circumferentially apart on the annular body, the support assemblies configured for supporting the cask in the cavity;a front roller housing extending forward from a front end of the annular body; anda rear roller housing extending rearward from a rear end of the annular body,wherein each of the front and rear roller housings comprise a pair of vertically extensible and retractable roller assemblies configured for rolling engagement with a horizontal transport surface. 2. The apparatus according to claim 1, wherein the top and bottom plates each include a front extension protruding forward from a front end of the annular body and a rear extension protruding rearward from a diametrically opposite rear end of the annular body, the front and rear extensions defining top and bottom walls of the front and rear roller housings. 3. The apparatus according to claim 2, wherein the front and rear extensions protrude outwards farther from the annular body than the laterally projecting portions of the top and bottom plates. 4. The apparatus according to claim 2, wherein each roller housing includes a first and second pair of horizontally spaced apart inner and outer side plates each extending vertically between the top and bottom walls, each pair of inner and outer side plates defining a roller assembly compartment holding one of the roller assemblies. 5. The apparatus according to claim 4, wherein the roller assembly compartments are disposed at opposite lateral side ends of the front and rear roller housings. 6. The apparatus according to claim 5, further comprising a plurality of reinforcement plates extending vertically between the top and bottom walls, the reinforcement plates spaced apart between the roller assembly compartments in each roller housing. 7. The apparatus according to claim 4, wherein each roller assembly includes a jack operably coupled to a roller at a first end and supported at a second end from one of the first or second pairs of inner and outer plates, each jack configured to move its respective roller into and out of one of the roller compartments. 8. The apparatus according to claim 7, wherein each jack is attached to a horizontal support member having ends attached to one of the first or second pairs of inner and outer plates. 9. The apparatus according to claim 4, wherein each roller assembly includes a roller attached to a horizontal base plate and a pair of horizontally spaced apart low-friction plates disposed perpendicularly on the base plate and having a vertical orientation, the low-friction plates formed of a low-friction material. 10. The apparatus according to claim 9, wherein the low-friction plates are movable along and engage the inner and outer plates of each roller assembly compartment when the roller assemblies are retracted or extended, the low-friction plates configured to prevent the rollers from supinating or pronating. 11. The apparatus according to claim 9, wherein each pair of low-friction plates is attached to a vertical beam mounted to the base plate, one of the low-friction plates being attached to each of two opposing sides of the vertical beam. 12. The apparatus according to claim 9, wherein each roller includes a plurality of axially aligned wheels supported between a pair of inner and outer wheel plates configured to prevent the wheels from supinating or pronating. 13. The apparatus according to claim 1, wherein each support assembly comprises a plurality of vertically elongated support plates attached along their inner sides to the annular body and projecting radially outwards therefrom. 14. The apparatus according to claim 13, wherein the support plates extend vertically from the top plate through the bottom plate and form downward projections that extend beneath the bottom plate for engaging the transport surface. 15. The apparatus according to claim 13, further comprising a plurality of cover plates extending vertically between the top and bottom plates along, outer sides of the support plates. 16. A low profile transport system for transporting nuclear waste, the apparatus comprising:a cask containing nuclear waste;a cylindrical annular body comprising a vertical axis, an open top end, an open bottom end, and a circumferentially-extending sidewall defining a cavity;a plurality of support assemblies spaced circumferentially apart on the annular body, the support assemblies configured to provide cantilevered support for the cask in the cavity;a top plate attached proximate to the top end of the annular body and including cantilevered front and rear extensions protruding from the annular body;a bottom plate attached proximate to the bottom end of the annular body and including cantilevered front and rear extensions protruding from the annular body;a front roller assembly housing formed from the front extension of the top plate and the front extension of the bottom plate; anda rear roller assembly housing formed from the rear extension of the top plate and the rear extension of the bottom plate;the roller assembly housings disposed on diametrically opposite front and rear portions of the annular body;each roller assembly housing comprising two roller assemblies, each roller assembly disposed between a pair of vertical inner and outer plates attached to the top and bottom plates at their ends;each roller assembly including a roller attached to a bottom of a base plate and laterally spaced apart low-friction plates extending upwards from the base plate;the roller assemblies being vertically movable between a retracted position in which the rollers are at least partially retracted into their respective roller housing and an extended position in which the rollers are at least partially projected from their respective roller housing; andwherein the low-friction plates slide along the inner and outer plates when the rollers move between the retracted and extended positions and prevent supination or pronation of the rollers assemblies. 17. The system according to claim 16, wherein each roller assembly further comprises a hydraulic jack coupled to the base plate and operably coupled to the roller to move the roller between the retracted and extended positions. 18. The system according to claim 16, wherein bottom ends of the support assemblies contact a support surface when the roller assemblies are in the retracted position, and the bottom ends of the support assemblies do not contact the support surface when the roller assemblies are in the extended position. 19. The system according to claim 16, wherein a lower portion of the cask nests within the cavity of the annular body and an upper portion of the cask that comprises a majority of a height of the cask protrudes upwards from the open top end of the cavity. 20. A low profile transport system for transporting nuclear waste, the apparatus comprising:a cask containing nuclear waste;a cylindrical annular body comprising a vertical axis, an open top end, an open bottom end, and a circumferentially-extending sidewall defining a cavity;a plurality of support assemblies spaced circumferentially apart on the annular body, the support assemblies configured to support the cask in the cavity in a cantilevered manner;a top plate attached proximate to the top end of the annular body and including cantilevered front and rear extensions protruding from the annular body;a bottom plate attached proximate to the bottom end of the annular body and including cantilevered front and rear extensions protruding from the annular body;a pair of front roller assemblies supported between the front extensions of the top and bottom plates, each front roller assembly including a vertically movable roller operably coupled to a hydraulic jack arranged in a compartment formed by a pair of vertical compartment side plates attached to the front extensions; anda pair of rear roller assemblies supported between the rear extensions of the top and bottom plates; each rear roller assembly including a vertically movable roller operably coupled to a hydraulic jack arranged in a compartment formed by a pair of vertical compartment side plates attached to the rear extensions;each roller assembly further including a base plate and a laterally spaced apart pair of low-friction plates extending upwards from the base plate, each low-friction plate disposed between the jack and one of the compartment side plates;each roller assembly being vertically movable between a retracted position in which the rollers are at least partially retracted into their respective compartment, and an extended position in which the rollers are at least partially projected downwards beyond their respective compartment;wherein the low-friction plates slide along the compartment side plates when the rollers move between the retracted and extended positions and prevent supination or pronation of the rollers assemblies.
	claims	1. An X-ray imaging apparatus comprising:an X-ray source configured to generate and irradiate X-rays according to X-ray irradiation conditions including a tube voltage, a tube current, an exposure time, and a filter;a display configured to provide a graphic user interface to receive a selection about the X-ray irradiation conditions; anda controller configured to:calculate a conversion ratio of a dose of x-rays that have not transmitted the filter to a dose of X-rays that have transmitted the filter,calculate, based on the conversion ratio, a parameter that represents a dose of radiation, to which an influence of the filter is reflected, based on a selected X-ray irradiation conditions and control the display to display the parameter,wherein the parameter comprises at least one of a converted amount of tube current corresponding to X-rays that have transmitted the filter, an amount of dose of X-rays that have transmitted the filter, or the conversion ratio,wherein the converted amount of tube current is given by an amount of a selected tube current multiplied by the conversion ratio,wherein the amount of dose X-rays that have transmitted the filter is given by the amount of the selected tube current multiplied by a dose per amount of tube current when the selected filter is used, andwherein the controller is configured to obtain the conversion ratio of the dose of X-rays that have not transmitted the filter to the dose of X-rays that have transmitted the filter, based on the dose per amount of tube current corresponding to the selected X-ray irradiation conditions and the dose per amount of tube current corresponding to an occasion when the filter is not used in the selected X-ray irradiation conditions. 2. The X-ray imaging apparatus of claim 1, wherein the display is configured to display the parameter in a numerical value, or a diagram or image representing the numerical value. 3. The X-ray imaging apparatus of claim 1, further comprising:a storage configured to store relationships between dose per amount of tube current (mAs) and tube voltage by differing types or thickness of the filter. 4. The X-ray imaging apparatus of claim 3, wherein the controller is configured to search the storage for a dose per amount of tube current corresponding to the selected X-ray irradiation conditions, when the selection of the X-ray irradiation conditions is input. 5. The X-ray imaging apparatus of claim 4, wherein the controller is configured to additionally search for a dose per amount of tube current corresponding to the occasion when no filter is used in the selected X-ray irradiation conditions. 6. The X-ray imaging apparatus of claim 1, wherein the controller is configured to when at least one of an imaging protocol or a size of a subject is selected, control the display to display a basic X-ray irradiation condition corresponding to the selected at least one of the imaging protocol or the size of the subject. 7. The X-ray imaging apparatus of claim 6, wherein the controller is configured to obtain a parameter that represents a dose to which an influence of the filter is reflected based on the basic X-ray irradiation condition. 8. The X-ray imaging apparatus of claim 1, wherein the controller is configured to re-obtain a parameter that represents a dose of radiation, to which an influence of the filter is reflected, whenever a selection of the X-ray irradiation condition is changed, and control the display to display the parameter. 9. A control method of an X-ray imaging apparatus, the method comprising:providing a graphic user interface configured to receive a selection of an X-ray irradiation conditions including a tube voltage, a tube current, exposure time, and a filter;calculating a conversion ratio of a dose of x-rays that have not transmitted the filter to a dose of X-rays that have transmitted the filter,calculating, based on the conversion ratio, a parameter that represents a dose to which an influence of the filter is reflected based on a selected X-ray irradiation condition; anddisplaying the parameter on a display,wherein the parameter comprises at least one of a converted amount of tube current corresponding to X-rays that have transmitted the filter, an amount of dose of X-rays that have transmitted the filter, or the conversion ratio,wherein the converted amount of tube current is given by an amount of a selected tube current multiplied by the conversion ratio,wherein the amount of dose X-rays that have transmitted the filter is given by the amount of the selected tube current multiplied by a dose per amount of tube current of when the selected filter is used,wherein the calculating of the parameter comprises:obtaining the conversion ratio of the dose of X-rays that have not transmitted the filter to the dose of X-rays that have transmitted the filter, based on the dose per amount of tube current corresponding to the selected X-ray irradiation conditions and the dose per amount of tube current corresponding to an occasion when the filter is not used in the selected X-ray irradiation conditions. 10. The method of claim 9, wherein the displaying of the parameter on the display comprises:displaying the parameter in a numerical value, or a diagram or image representing the numerical value. 11. The method of claim 9, further comprising: storing relationships between dose per amount of tube current (mAs) and tube voltage by differing types or thickness of the filter in a storage. 12. The method of claim 11, wherein the obtaining of the parameter comprises:searching the storage for a dose per amount of tube current corresponding to the selected X-ray irradiation condition, when the selection of the X-ray irradiation condition is input. 13. The method of claim 12, wherein the obtaining of the parameter comprises:searching for a dose per amount of tube current corresponding to the occasion when no filter is used in the selected X-ray irradiation condition.

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