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047587261	summary	The invention relates to a collimator exchanging system for moving a collimator between a measuring position and a storage position. A collimator exchanging system of this kind is described in Journal of Nuclear Medicine, Vol. 16, December 1975, pp. 1195-1196. As is mentioned therein, the system disclosed is not suitable for comparatively heavy collimators. Moreover, there is no facility for storing the collimator in a storage position. Such a facility is extremely desirable notably for heavy collimators. Using known exchanging systems, the mounting of the collimator in an active positon, for example, as a collimator for a radiation detection apparatus, such as a gamma camera, is a rather cumbersome operation and unambiguous positioning is not fully ensured. On the other hand, the storage of a collimator in a storage position is not ideal because an additional operation is required for securing the collimator. When the latter operation is omitted, which is not uncommon, the collimator could, for example, drop out of the storage position. It is an object of the invention to mitigate these drawbacks. To achieve this, a collimator exchanging system of the kind set forth in accordance with the invention is characterized in that from a position of adjustable height in a cart the collimator can be coupled only in a given orientation with respect to the measuring position, or the storage position, from a fixed position in the cart to either an active position or a fixed storage position by means of a simple operation. The exchanging system in accordance with the invention is constructed so that the collimator can be coupled only from a fixed orientation, so that the occurrence of accidents or errors during the exchange is precluded. Because uncoupling as well as coupling is realized by a simple operation and by means of a single cooperating system of means, a less stable intermediate position is avoided and the collimator cannot be coupled in an incorrect orientation. In a preferred embodiment the active device, the cart, and the storage device include a catching device which cooperates with a locking mechanism of the collimator. The active device notably includes at least two and preferably three locking pins, for example, mushroom-shaped pins, which cooperate with locking sockets in the collimator. The locking sockets may be constructed, for example, to be rotatable and be provided at the top as well as the bottom with locking holes which cooperate by a catching mechanism of the cart with locking pins of an active device or a storage device. A coupling mechanism of a further embodiment includes a ring which is secured to the collimator, and which couples and uncouples the collimator to and from the active device or the storage device by rotation, and hence to and from the cart. Notably two locking holes which are situated one over the other now cooperate so that in a fixed position of the collimator they couple the collimator to the cart in a first rotary position and couple the collimator to one of the devices in a second rotary position. For the fixed mounting of the collimator in an active position, a preferred embodiment includes a resilient clamping device which is activatable by a clamping effect between the collimator on the cart and a relevant part of the active device, and which enables coupling only in the activated condition.
	abstract	There is described a radiological image capturing apparatus, which makes it possible to obtain a good X ray image in which contrast of the peripheral portions are emphasized by employing the Talbot interferometer method and the Talbot-Lau interferometer method. The apparatus is provided with an X-ray tube, a multi-slit member, a first diffraction grating, a second diffraction grating and an X-ray detector. The second diffraction grating contacts the X-ray detector. A distance L between the multi-slit element and the first diffraction grating is set to be not less than 0.5 m, a distance Z1 between the first diffraction grating and the second diffraction grating is set to be not less than 0.05 m, and a slit interval distance d0 of the multi-slit element is set to be not less than 2 μm. With the settings, the above-mentioned good X-ray image can be obtained by using the Talbot-Lau interferometer system.
041728076	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS According to one embodiment of the method of the present invention, particles of a radioactive substance are mixed with particles of a material resistant to leaching by water so as to form a mass which is then enclosed in a capsule and subjected to isostatic pressing at a pressure and temperature required for the formation of a coherent, tight body of the mass. The material resistant to leaching by water may advantageously be composed of oxides of the types normally contained in glasses of various kinds and in rocks, examples of which include SiO.sub.2, B.sub.2 O.sub.3, MgO, alkaline oxides, alkaline earth oxides, TiO.sub.2, ZrO.sub.2, Fe.sub.2 O.sub.3, Fe.sub.3 O.sub.4 and Cr.sub.2 O.sub.3. Further, the material may, among other things, consist of rocks existing in nature which are well-known for their long-term stability, for example, rocks composed of silicates, aluminates, chromates and titanates. Rocks with the ability to retain gases may be of special interest, as well as zeolites with the ability to selectively take up strontium and cesium from a solution. In addition, glasses such as boron silicate glass and phosphate glass may be used. Preferred materials include aluminum oxide, titanium oxide, quartz and rocks existing in nature. The particle size of the radioactive substance and of the leach resistant material is preferably below 325 mesh. Of the total weight of radioactive substances and resistant material in the mixture, the weight of radioactive substances constitutes preferably 15-40 percent and the weight of resistant material preferably 60-85 percent. According to another embodiment of the method of the present invention, a particulate mass containing a material resistant to leaching by water in which the radioactive substances are fixed or a material in which the radioactive substances are fixed and which upon heating forms a material resistant to leaching by water is enclosed in a capsule and then subjected to isostatic pressing at a pressure and temperature sufficient for the formation of a coherent, tight body of the mass. The material resistant to leaching by water in which the radioactive substances are fixed may, among other things, be insoluble salts or other insoluble compounds of the radioactive substances such as titanates, aluminates, phosphates, silicates and oxides. The salts or the other insoluble compounds may, among other things, be precipitated from solutions containing radioactive substances by adding corresponding soluble salts. The particles of the resistant material suitably have a size of below 1 mm. The material in which the radioactive substances are fixed and which upon heating forms a material resistant to leaching by water may, among other things, be ion exchange materials which have taken up the radioactive substances through ion exchange upon contact with a solution containing the radioactive substances. A suitable size of the particles of the ion exchange material is from 0.1 to 1 mm. Examples of ion exchange materials which may be used for taking up radioactive substances are, among other things, zeolites and compounds with the formula M [M'.sub.x O.sub.y H.sub.z ].sub.n, where M is an exchangeable cation of the charge +n and M' may be Ti, Nb, Zr or Ta, for example NaTi.sub.2 O.sub.5 H. Ion exchange materials which have been taken up radioactive substances, upon heating, normally form multi-phase polycrystalline, ceramic materials which are resistant to leaching by water. For example, upon contact with a solution containing radioactive strontium, NaTi.sub.2 O.sub.5 H forms Sr[Ti.sub.2 O.sub.5 H].sub.2 which, when heated, is broken down into SrTiO.sub.3 and TiO.sub.2. According to this second embodiment, one or more materials resistant to leaching by water other than those in which radioactive substances are fixed and those which are formed during the heating previously mentioned, respectively, may be incorporated in the particulate mass. Examples of such other resistant material include oxides of such kinds as are normally comprised in glasses of various kinds and different rocks, such as, for example, SiO.sub.2, B.sub.2 O.sub.3, Al.sub.2 O.sub.3, MgO, alkaline oxides, alkaline earth oxides, TiO.sub.2, ZrO.sub.2, Fe.sub.2 O.sub.3 and Cr.sub.2 O.sub.3, and further rocks existing in nature which are well-known for their long-term stability, such as rocks composed of silicates, aluminates, phosphates and titanates. Preferred materials include aluminum oxide, titanium oxide, quartz and rocks existing in nature. The mentioned resistant materials may also be added to materials which are brought into contact with the radioactive substances and in which the radioactive substances are then fixed. Thus, the resistant materials may be mixed with ion exchange materials before the ion exchange materials are brought into contact with the radioactive substances so as to reduce the management of radioactive substance. A suitable amount of resistant material incorporated in the particle mass may be 1-95 percent of the total weight of the particle mass and the incorporated material. The particles of the incorporated resistant material suitably have a size of less than 1 mm and preferably of less than 0.5 mm. The capsule may, among other things, consist of sheets of tantalum, titanium, zirconium, alloys based on these metals such as, for example, zirconium based alloys sold under the tradename Zircaloy, steel, iron, nickel and further of quartz glass or boron silicate glass. The capsule material should be matched to the resistant material so that it has a sufficiently high melting point for the capsule to fulfil its duties, and substantially the same coefficient of thermal expansion in such cases when the capsule is left to provide a reinforced containment. With quartz or titanium oxide as resistant material, quartz glass is preferred for use in the capsule, and with boron silicate glass as resistant material, a capsule of the same material is preferred. In certain cases it may be suitable to use a capsule of metal which is internally provided with a layer of quartz glass or boron silicate glass. Between the capsule and the mass to be contained it may be suitable to arrange an intermediate layer of a resistant material such as any of the previously exemplified resistant materials. It may be particularly suitable to use a material of the same chemical composition as the mass but without radioactive isotopes as the intermediate layer. The particles in the material in the intermediate layer suitably have a size of less than 1 mm and preferably less than 0.2 mm. The intermediate layer may, for example, be applied as a layer of a thickness of a few mm or cm on the inner wall of the capsule. The pressure during the isostatic pressing amounts to at least 10 MPa and is preferably between 50 and 300 MPa. The temperature is, of course, dependent on what materials are included in the particulate mass but is at least 700.degree. C. A suitable temperature for particulate masses containing titanates, quartz or titanium dioxide is 1200.degree.-1300.degree. and for particulate masses containing aluminates and aluminum oxides is 1250.degree. C.-1350.degree. C. Turning now to FIG. 1 of the drawing, shown is one embodiment of the capsule 11 which may be used in the method of the present invention. The capsule 11 confines particulate mass 10 and includes opening 12 which is closed by fused portion 13. In FIG. 2, displaceable press stand 22 includes wheels 23 running on rails 24 on floor 25. The press stand 22 is of the type which consists of an upper yoke 26, a lower yoke 27 and a pair of spacers 28 which are held together by a prestressed strip sheath 29. The press stand 22 is movable between the position shown in FIG. 2 and a position where the stand surrounds high-pressure chamber 42. The high-pressure chamber 42 is supported by a column 49 and contains a high-pressure cylinder which includes an inner tube 50, a surrounding prestressed strip sheath 51 and end rings 52 which axially hold together the strip sheath and constitute a suspension device by which the high-pressure chamber is attached to column 49. The chamber 42 has a lower end closure 53 projecting into the tube 50 of the high-pressure cylinder. In the end closure there is a slot in which there are arranged a sealing ring 54, a channel 55 for the supply of a pressure medium, preferably argon or helium, and a channel 56 for cables for feeding heating elements 57 for the heating of the chamber 42. The heating elements 57 are supported by a cylinder 58 resting on an insulating bottom 59, which protrudes into an insulating sheath 60. The upper end closure comprises an annular portion 61 with a sealing ring 62 which seals against the tube 50. The sheath 60 is suspended from portion 61 and gas-tightly connected thereto. The end closure also comprises a lid 63 for closing the opening in portion 61, which is usually permanently mounted in the high-pressure cylinder. The lid 63 is provided with a sealing ring 64 sealing against the inner surface of portion 61 and with an insulating lid 65 which, when the high-pressure chamber is closed, projects into cylinder 60 and constitutes part of the insulating shell which surrounds the furnace space 66. The lid 63 is fastened to a bracket 67 which is carried by a raisable, lowerable and rotatable operating rod 68. Yokes 26 and 27 take up the compressive forces acting on end closure 53 and lid 63 when pressure is applied to the furnace space 66. The invention will be explained in greater detail by way of examples with reference to the accompanying drawing. EXAMPLE 1 25 parts by weight of high-level waste from a plant for reprocessing waste from a nuclear reactor, the waste having been converted into oxides in conventional manner and having a particle size of less than 80 mesh, is mixed with 75 parts by weight of quartz powder having a particle size of less than 100 mesh. The quartz was treated in vacuum to remove dissolved gasses prior to mixing. The mixture 10 is placed in a capsule 11 of Vycor glass which, to 96 percent by weight, consists of quartz and which is considered to fall under the concept quartz glass as used in this application. When the mixture 10 is introduced into the capsule 11, the capsule has no indentation 13 as is shown in FIG. 1. The capsule 11 is degassed at room temperature to a pressure of 0.1 Pa with a vacuum pump connected to the opening 12. The capsule 11 is then sealed at this pressure by fusing the capsule at 13. The capsule is placed in the furnace space 66, the lid 63 first having been lifted up and then lowered for closing the furnace space, and then the pressure and the temperature are successively increased to around 200 MPa and about 1200.degree. C., respectively, and are maintained at these values for about two hours, when the desired density and sintering have been obtained. The capsule 11 with the enclosed material is then allowed to cool, whereafter the pressure is reduced to atmospheric pressure and the capsule is removed from the furnace. The capsule 11 is allowed to remain as reinforcement for the material. The capsule 11 may be transported for permanent storage possibly enclosed in a steel container (not shown). EXAMPLE 2 A waste solution from a plant for reprocessing of high-level waste from a nuclear reactor is treated by the method of this invention. The solution consists of a 2-molar nitric acid solution and contains in the form of radioactive substances 7.0 g/l Zr, 6.9 g/l Mo, 8.0 g/l Nd, 4.5 g/l Ru, 5.4 g/l Cs, 4.8 g/l Ce, 3.8 g/l Fe, 3.1 g/l Pd, 3.3 g/l Ba, 1.5 g/l Sr, 2.5 g/l La, 2.3 g/l Pr, 2.3 g/l Am, 12.6 g/l U, 23.8 g/l Gd and various other radioactive substances in lesser amounts. The pH of the solution is adjusted to around 1 by adding ammonia. The solution is then passed through a cylindrical column of titanium containing an ion exchange material consisting of NaTi.sub.2 O.sub.5 H in the form of particles with a size of from 0.1 to 1 mm. The ion exchange material is mixed with the same quantity by weight of particles of TiO.sub.2 having a size of from 0.1 to 0.5 mm. The solution is then passed through a second cylindrical column of titanium containing an ion exchange material consisting of a zeolite of the formula Na.sub.8 Al.sub.8 Si.sub.40 O.sub.96.24H.sub.2 O. This ion exchange material also consists of particles of a size of from 0.1 to 1 mm. The water content of the two columns is removed by heating to around 900.degree. C. under vacuum. The ion exchange materials are then at least partially decomposed which leads to the formation of titanate containing radioactive substances and titanium dioxide in the first column. Each column with its contents is then placed in a cylindrical capsule of low carbon steel having a bottom and is then embedded in titanium dioxide powder having a particle size of less than 0.2 mm so that the spaces between the capsule and the column around the envelope surface of the column as well as above the column top and below the bottom are filled with the titanium dioxide powder. The titanium dioxide powder will also fill any spaces in the column which are accessible to the powder. Each capsule is then provided with a tightly fitting lid with an evacuation opening. After evacuation of each capsule at a pressure of 0.1 Pa and subsequent closing of the evacuation opening, each capsule with its contents is placed in a high-pressure furnace as shown in FIG. 2. When the capsules have been placed in the furnace space and the furnace space has been sealed, the pressure and the temperature in the furnace space are increased to around 100 MPa and about 1300.degree. C., respectively, and are maintained at these values for about two hours, when the desired density and sintering of the formed body is obtained. The capsules with the enclosed material are then allowed to cool, whereafter the pressure is reduced to atmospheric pressure and the capsules removed from the furnace. Each capsule is allowed to remain as reinforcement for the body. EXAMPLE 3 An 0.9 molar nitric acid solution containing in the form of radioactive substances 1.17 g/l (NH.sub.4).sub.6 Mo.sub.7 O.sub.24.4H.sub.2 O, 3.75 g/l Nd(NO.sub.3).sub.3.6H.sub.2 O, 0.59 g/l CsNO.sub.3, 1.23 g/l Ce(NO.sub.3).sub.3.6H.sub.2 O, 2.80 g/l Fe(NO.sub.3).sub.3.9H.sub.2 O, 0.57 g/l UO.sub.2 (NO.sub.3).sub.2.6H.sub.2 O, and 0.63 g/l Ni (NO.sub.3) is treated by the present method. The pH of the solution is adjusted to 1.3 by the addition of NaOH. The solution is then passed through a cylindrical column containing an ion exchange material consisting of NaTi.sub.2 O.sub.5 H in the form of particles having a size of 0.1 to 1 mm. The ion exchange material is mixed with the same amount of a mixture of TiO.sub.2, SiO.sub.2 and Al.sub.2 O.sub.3 having a grain size of 0.05 to 0.5 mm. The capacity of the ion exchange material corresponds approximately to 2.5% of the adsorbed waste calculated as oxide on dried ion exchange material. The ion exchange material is thereafter heated in air at 600.degree. C. and ground into a fine powder. The powder mixture is then packed in a capsule of iron having a tightly-fitting lid with an evacuation opening. After evacuating for 24 hours at a pump pressure of 0.1 Pa and heating to 750.degree. C., the capsule with the pump connected is sealed. The capsule is then placed in a high-pressure furnace as shown in FIG. 2 and the furnace space closed. The pressure is then raised to 150 MPa and the temperature to 1300.degree. C., and these conditions are maintained for two hours. The capsule with the enclosed material is then allowed to cool, whereafter the pressure is reduced to atmospheric pressure and the capsule is removed from the furnace. The contents of the capsule constitute a dense body without pores and voids and contain different crystalline phases, amoung other things TiO.sub.2, NaTiO.sub.3 and Al.sub.2 TiO.sub.5, so that the radioactive substances are fixed in a water-insoluble state. EXAMPLE 4 A waste solution from a plant for reprocessing high-level waste from a nuclear reactor is treated. The solution consists of a 2-molar nitric acid solution and contains in the form of radioactive substances, among other things, 60.5 g/l Nd, 5.9 g/l [PO.sub.4.sup.3- ], 10.6 g/l Cs, 11.5 g/l Mo, 10.5 g/l Sr, 10.0 g/l Zr, 5.1 g/l Fe and 0.3 g/l Ni. 7.2 g/l Ca and 2.2 g/l Al in the form of nitrates as well as 65 g of finely-divided SiO.sub.2 (grain size 100 Angstrom) are added to said solution. The solution is evaporated and then calcined in air for one hour at 500.degree. C. Thereafter, 60 parts by weight of calcine are mixed with 40 parts by weight of .alpha.-Al.sub.2 O.sub.3 by grinding in a ball mill. The mixture is then heated in air for two hours at 900.degree. C., remainders of nitrates and water then being driven off. The mixture is packed in a cylindrical capsule of iron which has a tightly-fitting lid with an evacuation opening. The capsule is evacuated for 24 hours at a pressure of 0.1 Pa at the pump and heated to 750.degree. C. and is thereafter sealed with the pump connected. When the capsule has been placed in a high pressure furnace as shown in FIG. 2 and the furnace space has been sealed, the pressure is raised to 150 MPa and the temperature to 1300.degree. C., and these conditions are maintained for seven hours. The capsule and the enclosed material are then allowed to cool, whereafter the pressure is reduced to atmospheric pressure and the capsule is removed from the furnace. The contents of the capsule constitute a dense body without pores and voids and with a density of 4.82 g/lcm.sup.3. The body contains different crystalline phases, among other things a phase of corundum type, (AlFe).sub.2 O.sub.3, a phase of fluorite type, (Zr, Ca, Nd)O.sub.2, a phase of pollucite type CsAlSi.sub.2 O.sub.6, a phase of scheelite type, (Sr Ca) MoO.sub.4, and a phase of apatite type (Ca Nd).sub.10 (SiO.sub.4, PO.sub.4, AlO.sub.4).sub.6 O.sub.2, in which the radioactive substances are fixed. A SEM-analysis of elements Cs, Sr and Nd shows that these elements are very evenly distributed in the body. EXAMPLE 5 The waste solution described in Example 4 is treated with formic acid at a temperature of 90.degree. C., whereby the nitrates are decomposed in accordance with the formula 2NO.sub.3.sup.- +4HCOOH.fwdarw.N.sub.2 O+4CO.sub.2 +3H.sub.2 O+20H.sup.-, and whereby metal oxides and metal hydroxides are precipitated. After drying, the precipitated substances are mixed with Al.sub.2 O.sub.3, placed in a capsule and then subjected to isostatic pressing in the manner described in the preceding example. The method according to the invention may be utilized not only for treatment of high-level waste from the reprocessing of nuclear reactor fuel, but for other treatments as well. It is also possible to use the method for treatment of high-level waste in connection with fuel reprocessing for the manufacture of plutonium for nuclear weapons, as well as for treatment of other radioactive substances for anchoring the radioactive material in a resistant body. Rock may be defined for the purpose of the present invention as a natural aggregation of mineral matter found in the crust of the earth. While there has been shown and described what is considered to be preferred embodiments of the present invention, it will be obvious to those skilled in the art that various changes and modifications may be made therein without departing from the invention as defined in the appended claims.
	abstract	Disclosed are a method and a device for recovering uranium (U) using a process for chemically treating washing wastewater of a uranium hexafluoride (UF6) cylinder. The method and the device are provided to separate uranium (U) from the wastewater released during a process of washing the uranium hexafluoride (UF6) cylinder and to release a filtrate that satisfies atomic energy licensing standards and environmental regulation standards using evaporation and condensation. Accordingly, an independent technology and process for treating the wastewater released during the process of washing the uranium hexafluoride (UF6) cylinder are ensured, which provides easier maintenance and greatly reduces costs compared to the purchase and operation of apparatuses manufactured by foreign makers.
	claims	1. A laser emitting device applied in a three-dimensional image capturing device, comprising:a plurality of laser devices that radiate pulse modulated laser beams irradiating a measurement subject for a distance measurement, said plurality of laser devices being separated into predetermined groups having plural laser devices; anda laser emitting operating processor that controls said laser devices to radiate laser beams concurrently in each of said predetermined groups, wherein each laser device of said plurality of laser devices in each of said predetermined groups is disposed at predetermined intervals, each illuminating area of a laser beam radiated from said each laser device overlapping each other at a distance of said measurement subject, wherein each of said predetermined groups comprises a pair of laser devices that are disposed opposite each other with a center of a photographing lens in between. 2. The device of claim 1, wherein said each group of laser devices radiates laser beams at a different timing. 3. The device of claim 1, wherein said plurality of laser devices are disposed at regular intervals along a periphery of a photographing lens in a circular arrangement. 4. The device of claim 3, wherein at least one laser device of a predetermined group is disposed between said plural laser devices forming another predetermined group. 5. The device of claim 4, wherein a number of said plurality of laser devices is six. 6. A three-dimensional image capturing device, comprising:a plurality of laser devices that radiate pulse modulated laser beams irradiating a measurement subject for a distance measurement, said plurality of laser devices being separated into predetermined groups having plural laser devices;an imaging device that accumulates a signal charge corresponding to a quantity of light received at said imaging device;a signal charge accumulation control processor that controls an accumulating operation of the signal charge generated in said imaging device due to a reflected light beam of said laser beam, which is reflected by said measurement subject;a signal charge integrating processor that repeatedly drives said signal charge accumulation control processor, so that said signal charge accumulated in said imaging device is integrated; anda laser emitting operating processor that controls said laser devices to radiate laser beams concurrently in each of said predetermined groups, wherein each of said laser devices in each of said predetermined groups is disposed at predetermined intervals, each illuminating area of said laser beam radiated from said each laser device overlapping each other at the distance of said measurement subject, wherein each of said predetermined groups comprises a pair of laser devices that are disposed opposite each other with a center of a photographing lens in between. 7. The device of claim 6, wherein said each group of laser devices radiates laser beams at a different timing. 8. The device of claim 6, wherein said plurality of laser devices is disposed at regular intervals along a periphery of a photographing lens in circular arrangement. 9. The device of claim 8, wherein at least one laser device of a predetermined group is disposed between said plural laser devices forming another predetermined group.
048719110	summary	The invention relates to an electron beam apparatus comprising, arranged inside an evacuatable housing, an electron source which comprises an electron emitter for generating an electron beam having a comparatively high emission current density and an electron-optical lens system. In an electron beam apparatus such as, for example, electron microscopes, electron beam writers and similar apparatus, it is often desirable to have an electron source available which source is capable of supplying an electron beam having a high current density, sufficient stability and attractive properties in relation to the electron-optical system of the apparatus. Electron beam apparatus comprising a conventional thermal cathode or a cathode made of LaB.sub.6, such as described in U.S. Pat. No. 3,631,290, supply an electron beam whose brightness is insufficient for many applications. Maximum current densities which can be achieved in this respect are, for example, approximately 5 a/cm.sup.2 or 30 A/cm.sup.2, respectively. The emissive surface of such electron source which is of relevance for the electron-optical system is too large in relation to the electron-optical properties of the apparatus, so that optimum beam formation, spot formation or imaging cannot be achieved. Electron beam apparatus comprising a field emission source as described in U.S. Pat. No. 3,631,291 are capable of supplying electron beams having a comparatively high current density, but such a source has several drawbacks such as instability of emissive power, positioning and geometry of the emitter. Moreover, such sources are not suitable for the supply of the frequently desired large total beam currents. From an electron-optical point of view, the dimension of the cathode and hence in this case of the virtual object of such a source is too small to obtain sufficiently high current densities. This is mainly due to the necessarily limited geometry of the emitter. Moreover the energy spread of the electrons in an electron beam produced by the described sources is comparatively large so that chromatic errors which are inadmissibly large are liable to occur in the electron-optical image. An electron source as described in U.S. Pat. No. 4,419,561 has succeeded in mitigating some of the drawbacks of the field-emission source, but the dimensions, positioning and the energy spread again render this source less suitable for many applications. The drawbacks of these sources relative to the heating of the emission wire and the positioning thereof have been partly solved by means of an electron source as described in Netherlands Patent Application 8302275, corresponding to U.S. Pat. No 4,591,753. However, the temperature threshold, the comparatively fast evaporation of the cathode wire and the resultant contamination of the apparatus as well as the non-adapted object dimensions and the large energy spread are still drawbacks of this source. It is the object of the present invention to mitigate these drawbacks; and to achieve this, an electron beam apparatus of the kind set forth is characterized in that the electron emitter comprises a semi-conductor element in which there is provided, parallel to an emissive surface, a p-n junction which is to be connected in the reverse direction and whose dimensions define surface dimensions of the emissive surface to electron optical properties of the apparatus with the current density and the current intensity of an electron beam to be emitted being optimized at the same time. Because an electron beam apparatus in accordance with the invention uses a cold cathode as the electron emitter, the known thermal problems are avoided in this apparatus. Using this electron emitter, electron beams having a sufficiently high current density and current intensity can be readily achieved and the dimensions of the emissive surface defined by the transverse dimensions of the p-n junction enable optimum beam and spot shaping and electron-optical imaging. The semiconductor electron emitter is constructed so as to operate with a p-n junction which is connected in the reverse direction. Due to the use of a source comprising a p-n junction connected in the reverse direction, so-called hot electrons are emitted. This means that the electrons to be emitted must overcome a potential gradient when they emerge from the emissive surface. Known drawbacks of the use of an electron emitter having a negative electron affinity (NEA) for generating so-called cold electrons are now avoided. An important one of these drawbacks is the high susceptibility of the emissive surface to external disturbances of the emission. For specific properties of a semiconductor element which emits hot electrons and which is connected in the reverse direction, reference is made to an article by Bartelink et al in Physical Review, Vol. 130. No. 3, May 1963, pp 972-985. In a preferred embodiment of an electron beam apparatus in accordance with the invention, the largest transverse dimension of a consecutive emissive surface is limited to a maximum of, for example approximately 10 .mu.m. Using semiconductor techniques, such as ion implantation, for example it is comparatively easy to realize a p-n junction having such a dimension. This offers major advantages for an electron beam apparatus because the dimensions of the emissive surface is thus adapted to electron-optical requirements to be imposed relative to the electron-optical lens system of the apparatus. Surprisingly, it has been found that by using dimensions of a customary order of magnitude a very effective emitter can indeed be realized. Even with an amply high current density for a source with these dimensions no problems are encountered relative to discharge of non-emitted electrons, so that no disturbing internal heating occurs in the semiconductor element and no potential distribution is generated which would adversely affect the current density distribution of the emitted electron beam. For some applications, for example, for beam writers, it may be attractive to use an elongated emissive surface, for example having a length/width radio of, for example from 5:1 to 10:1. Thus, a beam having an elongated cross-section can be realized and spaced charge problems can be reduced. Moreover, by using a line-shaped beam-splitting system, a row of 10 separate spots can be formed, for example. For optimizing the images use may be made, if desired, of a non-rotationally symmetrical electron-optical lens. Such an elongated emissive surface is preferably rectangular but may also be, for example, substantially elliptical, to realize a particularly narrow spot (see U.S. Pat. No. 3,881,136). A rectangularly emissive surface may notably form a square which may be attractive, for example, for beam writers because this allows for the formation of an electron beam having a rectangular cross-section and hence a similar shaped writing electron spot. In such an apparatus for the manufacture of integrated circuits, it may be advantageous to use a regular polygon in view of the manufacturing properties. Using known techniques, for example as described in U.S. Pat. No. 4,419,561, it is then also possible to apply beam shaping. In apparatus such as electron microscopes, it will usually be advantageous to use a round emissive surface in view of the customary rotationally-symmetrical electron-optical lens system. In all these cases dimensions of the emissive surface can be optimally adapted to the electron-optical system. In a preferred embodiment transverse dimensions of a singular emissive surface are limited to a maximum of 10 .mu.m, for example and usually preferably to a value of between, for example 0.5 .mu.m and 5 .mu.m. It is also comparatively simple to realize composite emissive surfaces, such as a central emissive surfaces, which is surrounded by an annular second surface, or an array or a matrix of several surfaces. When a sufficiently large distance is chosen between the sub-regions of a composite emmisive surface, the emission of each of these surfaces can be simply and independently controlled. The p-n junction in the electron emitter is preferably located at a depth of from approximately 0.01 .mu.m to 0.05 .mu.m below the emissive surface. Because at the surface no difference occurs in the semiconductor material of the emissive surface and that of the surrounding cathode surface, no edge problems will occur. The semiconductor element of a preferred embodiment consists of Si; and very good results have been obtained by means of this material also in relation to service life. However, it is alternatively possible to use, for example SiC, Si--SiO.sub.2 combinations, GaAs or similar group III-V combinations. Actually, the choice of the material is not relevant for the invention since as long as the requirements relation to a high current density together with emissive surface dimensions which are adapted to the apparatus are satisfied and a sufficiently long service life can be achieved. In order to achieve a further increase of the emission density, in a further embodiment the semiconductor element is provided at the area of the emissive surface with a substantially monomolecular layer of an appropriate material such as Cs, Ba, AgO, PtO, MiO, CO.sub.2, C etc.; good results have been obtained notably with Cs and Ba. The method of manufacturing a semiconductor electron emitter makes it comparatively simple to construct the emitter as a matrix or an array of emissive elements, for example, a series of 10 elements or an orthogonal matrix of 10.times.10 elements. Using multi-systems as described in U.S. Pat. No. 3,491,236, U.S. Pat No. 4,524,278 or U.S. Pat. No. 4,568,833, it is comparatively simple to render beams of each of these elements separately controllable because direct cathode control of at least the beam current is now possible, and a large part of the beam splitting system of the apparatus can now be dispensed with. Such a multi-beam apparatus is excellently suitable, for example, for the production of integrated circuits, notably those circuits where the electron beam writes directly in the semiconductor material i.e. without intermediate masks.
	summary
	summary
047724480	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention provides a support pin system which is a non-welded mechanical system. A novel support pin and a nut are provided for fastening a first structural member to a second structural member. A first pin portion passes through a through-bore in the first structural member and has a threaded section which mates with the nut. A second pin portion has a solid body section and a split-base section, which split-base section biasingly engages a bore provided in the second structural member. The solid body section is accommodated by the bore by a close clearance fit. The split-leaf base section has an intermediate section with an outer diameter which is less than the outer diameter of the solid body section. Thus, loads applied transversely to the longitudinal axis of the support pins system are reacted substantially in pure shear through the solid body section of the support pin. By reacting the transverse loads substantially in this manner, rather than through the bending moment of the leaves of the pin, the bending loads on the pin are substantially reduced and the bending stresses are substantially relieved in areas of the pin where stress corrosion cracking has occurred in prior art pins. When the support pin system is used to fasten a nuclear reactor control rod guide tube flange to a nuclear reactor upper core plate, the support pin system preferably includes a locking nut rather than a simple securing nut such as an hexagonal nut or a spline nut. The locking nut crimpingly engages longitudinal recesses provided in the free end of the support pin, thereby positively preventing relative rotation of the nut about the pin. According to the present invention a novel locking nut is provided which has a crimpable cylindrical section integrally connected to an internally threaded section. The integral connection may be achieved by integrally machining the part or by joining the two elements in any of the well known manners, such as by brazing or welding. The nut is coaxially inserted about the support pin and threadedly engaged onto a mating threaded section of the support pin. For nuclear reactor use, the nut is torqued to a predetermined preload and the crimpable section of the nut is crimped to engage the recesses provided in the support pin, whereby relative rotation therebetween is positively prevented. Preferably the recesses are longitudinal recesses. The integral structure of the present locking nut overcomes the prior art requirement for a separate crimped cap and/or dowel pin, and thereby reduces the possible number of small parts which could be dislodged in the event of a failure of one or more components of the support pin system. A support pin system for nuclear reactor use preferably has a support pin fabricated from a nickel based alloy, most preferably strain-hardened 316 stainless steel which is highly resistant to stress corrosion cracking unlike the previously used Inconel-750. Such cold worked 316 staqinless steel has an excellent operating history in reactor internals applications. The preferred material for the novel, integral locking nut according to the present invention is 304 stainless steel. The present invention also provides a locking nut retainer which is a modified bolt locking cap and which may be used to positively retain virtually any nut in position about an elongate threaded element when the structural member having the elongate threaded element disposed therein can be provided with a counter-bore having an annular recess therein for accommodating the present locking nut retainer. Thus, the present invention contemplates a locking nut retainer adapted to be positioned within and accommodated by a structural member having an elongate threaded element disposed therein for positively retaining a nut in position around the elongate threaded element once the two are matingly engaged. A split cylindrical wall portion which is crimpable and has an axial slot defined therein is provided to render the retainer substantially resiliently compressible along the radial axis thereof to allow for insertion of the retainer into the bore and annular recess provided in the structural member. The retainer is provided with tabs extending radially from the cylindrical wall portion along the external surface thereof. The retainer is positionable within the bore and the tabs are accommodated by the annular recess. The crimpable split cylindrical wall portion of the retainer crimpingly engages the external periphery of the nut in use whereby the nut is positively retained around the elongate threaded element and the locking nut retainer is positively retained in the structural member. At least one crimp, preferably at least two crimps, are provided for this purpose. The present invention further provides a locking nut which is adapted to be coaxially positioned in threaded engagement with an elongate threaded element, such as the novel support pin of the present invention, and to crimpingly engage same in use. The locking nut features an internally threaded section which threadedly engages the elongate threaded element in use. Integrally connected to the internally threaded section is a crimpable cylindrical section which extends coaxially therefrom. Thus, when the elongate threaded element is provided with an end section having a plurality of recesses, preferably longitudinal recesses, on the external surface thereof, at least one, but preferably at least two, of the recesses are crimpingly engaged by the crimpable cylindrical section of the locking nut. Such engagement positively prevents relative rotation of the nut about the elongate threaded element and is useful in many applications requiring a mechanical joining system, which can be installed without welding and removed with relative ease, albeit by the probable destruction of the locking nut, such as for the fastening of a nuclear reactor control rod guide tube flange to a nuclear reactor upper core plate. The invention can be better understood by referring to FIG. 1, an elevational side view, partly in cross-section, of a new and improved support pin system constructed in accordance with the present invention, as generally indicated by the reference character 10. The cooperative parts of support pin system 10 are shown, for example, as being used in a nuclear reactor power plant for securing control rod guide tubes to upper core plate 12, the previously referred to second structural member. The support pin system 10 may be installed ab initio or may be retrofitted when an existing conventional guide tube support pin system requires repair or replacement due to failure under, for example, stress corrosion cracking conditions. The vertically positioned control rod guide tubes are secured to the upper core plate 12 through means of annular guide tube flanges 14, the previously referred to first structural member, one flange 14 formed about the lower periphery of each guide tube. In order to interconnect the guide tube flange 14 to the upper core plate 12, a plurality of vertically disposed support pins 16 are utilized, only one such pin 16 being shown substantially in full view. The support pin orientation here is vertical, however, any orientation would be within the scope of the invention. Pins 16 are disposed in a circumferential array within the annular guide tube flange 14, usually in pairs. An upper or first pin portion 18 is disposed within the guide tube flange 14 and a lower or second pin portion 20 is disposed within the upper core plate 12. In order to accommodate the dispostion of the support pins 16, the upper surface of the upper core plate 12 is provided, at each locus of a support pin 16, with a bore 22 within which a lower, split-leaf base section 24 of the support pin is adapted to be frictionally inserted and retained in a biased engagement. The bore 22 is shown as a through-bore however, in some applications it may be a blind bore, whereby the split leaves are retained therewithin should they shear. The guide tube flange 14 is provided, at each locus of a support pin 16, with a throughbore 26 for accommodating an intermediate shank portion 28 of the support pin 16, and it is further seen that the shank portion 28 and base section 24 of support pin 16 are integrally connected by means of an annular shoulder 30. A lower counter-bore 32 (a second counter-bore 32) is defined within the lower surface of the guide tube flange 14, in the surface thereof proximate to the upper core plate 12, so as to be coaxially or concentrically disposed with respect to the guide tube flange through-bore 26, and in this manner, the support pin annular shoulder 30 is appropriately accommodated and seated within the lower surface of the guide tube flange 14. The first pin portion 18 of the support pin 16 has an end section 34 and an externally threaded section 36. The upper end section 34 of the support pin 16 projects vertically upwardly and axially outwardly from the guide tube flange through-bore 26, along longitudinal axis 38 and is adapted to be threadedly mated with an annular, axially elongated, internally threaded, securing nut 40. Internally threaded section 42 of the nut 40 threadedly engages the externally threaded section 36 of the support pin 16. The nut 40 fixedly retains the guide tube flange 14 in its mounted mode upon each support pin 16, and therefore fixedly secures or fastens the flange 14 of the nuclear reactor control rod guide tube upon the upper core plate 12. In order to properly accommodate the securing nut 40 upon the threaded section 36 of the support pin 16, the upper surface of guide tube flange 14, the surface thereof remote from the upper core plate 12, is provided at each location of a support pin 16, with an upper counter-bore 44 (a first counter-bore 44) which is co-axially or concentrically defined with respect to flange through-bore 26 and the lower counter-bore 32. An annular floor surface 46 is thus defined within the guide tube flange 14 upon which the lower face of securing nut 40 is engaged and seated in a manner similar to the engagement and seating of the support pin annular shoulder 30 upon an annular ceiling surface 48 of guide tube flange lower counter-bore 32. In this manner, when the securing nut 40 is threadedly engaged upon the externally threaded section 36 of the support pin 16, the guide tube flange 14 will be securely retained between the securing nut 40 and the support pin 16 by means of the engaged seating of the lower end of the nut 40 upon the annular floor surface 46 of flange 14, as well as by means of the engaged seating of the support pin annular portion 30 upon the annular ceiling surface 48 of the guide tube flange 14. With continuing reference to FIG. 1, extending downwardly from annular shoulder 30 of the first pin portion 18 is a solid body section 50 of the second pin portion 20. The solid body portion 50 has an outer diameter 52 which is accommodated by the core plate bore 22 by a close clearance fit. The split-leaf base section 24 extends from the solid body section 50 and has a split intermediate section 54 which extends from the solid body section 50 and terminates in a split end section 56. The split end section 56 biasingly engages at least a portion of the wall of the core plate bore 22, whereby the support pin 16 is fixedly secured within the upper core plate 12 by a frictional fit. However, the split intermediate section 54 has an outer diameter 58 which is less than the outer diameter 52 of the solid body section 50 so that split intermediate section 54 does not engage the wall of the bore 22. Thus, the split-leaf base section 24 frictionally engages the bore 22 at the split end sections 56, but not along the split intermediate sections 54. This arrangement allows loads applied transversely to the longitudinal axis 38 of the support pin 16 to be reacted substantially in pure shear by the second pin portion 20 substantially through the solid body section 50. Thus, the prior art support pin tendency to undergo stress cracking and shearing in the intermediate shank portion 28 of the support pin 16 is no longer a problem because the present structure minimizes the bending stress on the intermediate shank portion 28 by restricting bending of support pin 16 along its length by means of the close clearance fit provided between the solid body section 50 and the walls of the core plate bore 22. Prior art support pins were frequently susceptible to stress cracking corrosion in the crotch portion where the split intermediate section joined the solid body section. The prior art stress cracking corrosion and shearing tendencies of prior art support pin resulted from structural designs, such as that disclosed in the previously referred to U.S. patent application Ser. No. 576,645 by J. T. Land et al. Such prior art support pins had solid body sections and split intermediate sections having the same outside diameter so that applied loads reacted in bending as well as in shear and machining tolerances could be more relaxed to accommodate misalignments since the prior art configurations did not provide a close clearance fit. The present support pin 16 accommodates misalignments by increasing the length especially of the second pin portion 20. The present support pin 16 is also preferably provided with a solid body section 50 having an outer diameter 52 which is greater than outer diameter 58 of the intermediate shank portion 28 of the upper or first pin portion 18, for additional strength when loads are reacted in shear therethrough. With continuing reference to FIG. 1 and by way of example, when the guide tube flange 14 is approximately 3.81 cm thick and the upper core plate 12 is approximately 5.72 cm thick, the support pin 16 is advantageously about 13.34 cm in length. The bore 22 provided in upper core plate 12 is a through-bore and accommodates the second pin portion 20 which has a length of 5.72 cm. Further, the outer diameter 58 of the intermediate shank portion 28 is approximately 1.80 cm, the outer diameter of the annular shoulder 30 is 3.38 cm, the outer diameter of the solid body section 50 is 2.69 cm, the total outer diameter of the split intermediate section 54 is 2.54 cm and the total outer diameter of the split-leaf section 56 in its free state is 2.72 cm. The diameter of the core plate bore 22 is 2.70 cm such that a close clearance fit of 0.013 cm exists between the core plate bore 22 and the solid body section 50, but a clearance of 0.165 cm exists between the split intermediate section 54 and the core plate bore 22. With continuing reference to FIG. 1, securing or locking nut 40 has an internally threaded section 42 and a crimpable cylindrical section 60 extending from and integrally connected to the internally threaded section 42. The internally threaded section 42 threadedly engages the externally threaded section 36 of the first pin portion 18 of the support pin 16. Threaded sections 42, 36 cooperate to retain the guide tube flange 14 between the locking nut 40 and the annular shoulder 30. In an alternate embodiment without an annular shoulder 30 (not shown), the guide tube flange 14 could be retained between the locking nut 40 and the solid body section 50 of the second pin portion 20 of the support pin 16. Once the locking nut 40 has been threadedly engaged upon the threaded section 36 of the support pin 16, and appropriately torqued to a pre-determined load limit or value, it is of course desirable to insure the fact that the pin and nut assembly 16, 40 remains intact in their assembled state so as to, in turn, insure the fact that the guide tube flange 14, and therefore the nuclear reactor control rod guide tube, remains positionally fixed with respect to the nuclear reactor upper core plate 12. As seen in FIG. 1, the uppermost end section 34 of support pin 16 above the threaded section 36 thereof is provided with a plurality of recesses 64, which are shown in FIG. 1 as longitudinally or axially extending longitudinal recesses 64, which are equiangularly spaced about the support pin 16 at 90.degree. intervals and serve as crimp receiving sections or grooves. For some modifications, the recesses need not be longitudinal, although longitudinal recesses are preferred herein. The upper end section 34 of the support pin 16 passes through the coaxially aligned crimpable cylindrical section 60 of the nut 40, and once the nut and pin assembly 16, 40 is fully threadedly engaged and the predetermined torque load value or limit has been attained, then crimps 62 are made by pressing upon and deforming portions of crimpable cylindrical section 60 into crimp receiving sections or longitudinal recesses 64 of the support pin 16. The thickness of the crimpable cylindrical section 60 is approximately 0.051 cm, for example. Although FIG. 1 shows crimp 62 opposed by an uncrimped but crimpable section 60, in practice, diametrically opposed crimps 62, 62 are formed by simultaneously pressing from opposite directions, so as to operatively engage two diametrically opposed longitudinal recesses 64 formed upon the upper end section 34 of support pin 16. Four longitudinal recesses 64 are indicated in FIG. 1, therefore two diametrically opposed crimps can be formed. This crimping operation must of course be performed in situ at the location sites of the support pins 16. In view of this requirement, these crimping operations will be performed by means of suitable, remotely controlled tools, not shown, whereby such crimping operations may be performed in an irradiated underwater environment without exposing maintenance personnel to the irradiated environment when support pin system 10 is retrofitted in an operating nuclear reactor. Spacial restrictions frequently do not permit use of an hexagonal nut since an hexagonal torque wrench cannot be positioned about the nut structure in an annular, 360.degree. mode so as to impart the necessary torque to the securing nut. Accordingly, pursuant to the present invention, the locking nut 40 is shown in FIGS. 1 and 2 as a spline nut and is provided with ten, vertically oriented splines 66 defined within the external surface thereof in a circumferential array and alternatingly associated with adjacent spline grooves 68. In this manner, a suitable splined torque tool, not shown, can axially engage the locked nut splines 66, and once engaged, rotational torque applied thereto. Due to the aforenoted spacial constraints and restrictions of the use environment, the splined torquing tool need not engage the locking nut 40 in a complete 360.degree. annular relationship as would be true of a conventional external hexagonal torque wrench. In particular, the external splined torque wrench may engage the locking nut splines 66 over a circumferentially extending arcuate area of less than 180.degree., and the splined torque wrench stroke may be, for example, 36.degree.. Also shown in FIG. 1 is a locking nut retainer shown generally at 70 and having a split cylindrical wall portion 72 which is crimpable and has an axial slot 74 defined therein, as shown in FIG. 3, a plan top view. Tabs 76 extend radially from the wall portion 72 and are positioned along the external surface thereof as shown in FIGS. 1 and 3. The locking nut retainer 70 is positioned around the locking nut 40 and extends upwardly along at least a portion of nut 40 as shown in FIG. 1. The locking nut retainer is accommodated by and positively retained in use by guide tube flange 14. The top surface of guide tube flange 14, remote from the upper core plate 12, is provided with a retainer counter-bore 78, which counter-bore 78 has an annular recess 80 radially defined in the wall thereof. Counter-bore 78 may be the same as upper counter-bore 44 or may have a slightly greater diameter as shown in FIG. 1. The axial slot 74 of the locking nut retainer 70 renders same substantially resiliently compressible along the radial axis thereof to facilitate insertion of the retainer 70 during use. Locking nut retainer 70 is compressed to reduce the effective diameter thereof and is positioned within counter-bore 78. The tabs 76 are accommodated by annular recess 80, and when the compressive force on the locking nut retainer 70 is released, retainer 70 resiliently returns substantially to its original size, or may be so adjusted, and is retained positively within guide tube flange 14. After the locking nut 40 has been torqued, the split cylindrical wall portion 72, which is crimpable, is crimpingly connected to the locking nut 40. The crimp receiving sections of locking nut 40 are the spline grooves 68. At least one crimp and preferably at least two crimps are formed by pressing sections of the split cylindrical wall portion 72 into respective adjacent spline grooves, whereby the locking nut retainer 70 is positively retained in position around support pin 16. The thickness of the split cylindrical wall portion is approximately 0.051 cm. Two crimps 82 are shown engaging spline grooves 68 in FIG. 2 and are spaced 144.degree. apart. The axial slot 74 and tabs 76 suggest this preferred spacing, however, any number of crimps 82 could be formed with any degree of angular separation depending on the degree of fastening desired and the type of locking nut 40 used for the particular application. Should locking nut 40 disengage itself from a support pin 16, such as in the event of a failure of all of the crimps 62, or should the support pin 16 shear under stress, locking nut retainer 70 retains these dislodged components in place and the nuclear reactor system is protected thereby. Such a damaged support pin system 10 clearly needs to be replaced, however, and repair and replacement are facilitated by the non-welded system according to the present invention. Further, such locking nut retainers 70 need not be used in combination with a locking nut 40, but may themselves be retrofitted onto any of the conventional prior art support pin systems, such as those using a nut with a separate dual crimp cap. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims.
056573600	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS A first embodiment of a reactor container provided with a dry well cooling system according to the present invention will be described hereunder with reference to FIG. 1. In FIG. 1, denoted by reference numeral 1 is a reactor container which is divided to define a dry well 2 and a wet well therein. An in-dry-well heat exchanger 3 is disposed in the dry well 2, and an in-dry-well blower 4 is provided on and connected to the primary side of the in-dry-well heat exchanger 3. The wet well is not shown in FIGS. 1-13 in connection with first to thirteenth embodiments of the present invention, but it will be easily understood by persons skilled in the art that the wet well is disposed below the dry well 2 in the reactor container 1 such as located in a reactor container of FIGS. 14 to 17. A circulation pipe 9 is connected to the secondary side, i.e., a heat transfer pipe, of the in-dry-well heat exchanger 3, and a normal cooling system 6 is connected to the circulation pipe 9. The normal cooling system 6 includes an equipment cooling pump 7, an equipment cooling heat exchanger 8, and a seawater pump 13. A standby cooling system 14 is branched from the circulating pipe 9 at an intermediate portion between the reactor container 1 and the normal cooling system 6. The standby cooling system 14 is intended to flow cooling water into the heat transfer pipe of the in-dry-well heat exchanger 3 for cooling the same. The cooling water is introduced by a standby cooling pump 15 to a standby cooling heat exchanger 16 where the water is cooled. Then, the cooling water is conveyed to the heat transfer pipe of the in-dry-well heat exchanger 3. In other words, the standby cooling system 14 comprises the standby cooling pump 15, the standby cooling heat exchanger 16, and a standby seawater pump 17 for supplying seawater 12 from the sea 5 as cooling water to the secondary side of the standby cooling heat exchanger 16. A second embodiment of the present invention shown in FIG. 2 is basically similar to the above first embodiment except that the standby cooling pump 15 and the standby cooling heat exchanger 16 are omitted. The standby cooling system 14 is constituted as a direct seawater cooling system which acts to directly supply the seawater 12 to the in-dry-well heat exchanger 3 by using only the seawater pump 17. More specifically, a standby cooling discharge pipe 14a and a supply pipe 14b are connected, as seawater circulation line in communication with the sea 5, to respective lines of the circulation pipe 9, and the standby seawater pump 17 is connected to the supply pipe 14b. With the first and second embodiments explained above, even when the normal cooling system 6 is brought into an outage due to failure or inspection, the in-dry-well heat exchanger 3 can be cooled by operating the standby cooling system 14 and the interior of the dry well 2 can be cooled by the in-dry-well blower 4. As a result, the reliability of the dry well cooling system can be improved. With the following embodiments, the present invention will be described with reference to structure different from that of the first or second embodiment by adding new reference numerals and like reference numerals to portions or members corresponding to those shown in FIGS. 1 and 2 and the explanations thereof are omitted herein. A third embodiment of the present invention will be described hereunder with reference to FIG. 3. In the third embodiment, the standby cooling system 14 is likewise connected to the circulation pipe 9, but it comprises a standby cooling pump 14c and an air cooler 14d connected to each other. With this third embodiment, therefore, the standby cooling system can be operated with no need of operating the seawater pump. Accordingly, even during the periodic inspection of the seawater cooling system, the dry well cooling system can be operated by using the standby cooling system. A fourth embodiment of the present invention will be described hereunder with reference to FIG. 4. The fourth embodiment is different from the above first embodiment in that electric power necessary for operating various equipment can be supplied from not only a normal power supply 18 under ordinary condition, but also an emergency power supply 19 at need. With the fourth embodiment, therefore, even when the normal power supply 18 is disabled due to the outage of the external power source, the dry well cooling system can be operated by the electric power supplied from the emergency power supply 19 to remove the heat from the reactor container. A fifth embodiment of the present invention will be described hereunder with reference to FIG. 5. In the fifth embodiment, the normal cooling system 6 in the above first embodiment is disconnected from the in-dry-well heat exchanger 3 and constituted as a separate system, whereas a dedicated cooling system 20 is provided on and connected to the secondary side of the in-dry-well heat exchanger 3. The dedicated cooling system 20 comprises a dedicated cooling pump 21, a dedicated cooling heat exchanger 22, and a dedicated cooling seawater pump 23. With this embodiment, even when the RHR system is failed and its cooling function is disabled, the cooling system in the dry well can be operated so that the function of the dry well cooler for removing the heat from the reactor container is ensured. A sixth embodiment of the present invention will now be described with reference to FIG. 6. In the sixth embodiment, the in-dry-well blower 4 and heat exchanger 3 disposed in the dry well 2 inside the reactor container 1 are associated with an environmental condition resistance maintaining equipment 24 to maintain their performance so that the blower 4 and the heat exchanger 3 can endure against serious environmental conditions, i.e., high-temperature, high-pressure, high-humidity and aqueous atmosphere, in an anticipation of a severe accident, thereby enabling the dry well cooling system to operate. With this sixth embodiment, even if the atmosphere in the dry well 2 should exceed conventional design conditions, i.e., a temperature of 200.degree. C., pressure of 10 Kg/cm.sup.2 and humidity of 100%, in the event of a severe accident, the performance of the dry well cooling system can be maintained to ensure sufficient heat removal from the reactor container. Also, even when an atmosphere of sprayed water droplets is produced in the dry well 2 by a reactor container spray, not shown in FIG. 6, the dry well cooling system can be operated to effect sufficient heat removal. Therefore, even if the heat removal function of the RHR system should be disabled in the event of a severe accident, the dry well cooling system can remove the heat from the reactor container with the sufficient reliability. A seventh embodiment of the present invention will be described hereunder with reference to FIG. 7. In the seventh embodiment, an extra-dry-well blower 4a is mounted outside the reactor container 1 through a pipe 4b communicating with the dry well 2 of the reactor container 1, and the delivery side of the extra-dry-well blower 4a is connected to an inlet pipe 3b on the primary side of an extra-dry-well heat exchanger 3a. A return pipe 3c on the primary side of the extra-dry-well heat exchanger 3a is connected to the reactor container 1 to be communicated with the dry well 2. An emergency dry well cooling system equipment cooling system 25 is connected to a circulating pipe 3d on the secondary side of the extra-dry-well heat exchanger 3a. The emergency dry well cooling system equipment cooling system 25 comprises an emergency dry well cooling system equipment cooling pump 26, an emergency dry well cooling system equipment cooling heat exchanger 27, and an emergency dry well cooling system equipment cooling seawater pump 28, which are operatively connected to each other. Thus, the seventh embodiment of the present invention shown in FIG. 7 includes a system dedicated for secondary cooling of the extra-dry-well heat exchanger 3a, i.e., the emergency dry well cooling system equipment cooling system 25. In this embodiment, the emergency dry well cooling system equipment cooling pump 26 conveys cooling water from the extra-dry-well heat exchanger 3a to the emergency dry well cooling system equipment cooling heat exchanger 27, and the emergency dry well cooling system equipment cooling seawater pump 28 conveys the seawater 12 to the emergency dry well cooling system equipment cooling heat exchanger 27, thereby cooling the heat in the dry well 2. An eighth embodiment of the present invention will be described hereunder with reference to FIG. 8. This eighth embodiment is of the same arrangements as the above seventh embodiment except that an emergency dry well cooling system equipment air cooling system 26 is provided in place of the emergency dry well cooling system equipment cooling system 25. The eighth embodiment of the present invention shown in FIG. 8 includes an emergency dry well cooling system equipment air cooling pump 26 and an air cooler 31 for cooling the secondary side of the extra-dry-well heat exchanger 3a. With this embodiment, the emergency dry well cooling system equipment cooling system can be operated with no need of operating the seawater pump. As a result, even during inspection of the seawater cooling system, the emergency dry well cooling system can be operated. A ninth embodiment of the present invention will be described hereunder with reference to FIG. 9. The ninth embodiment is different from the seventh embodiment in that not only the normal power supply 18 but also the emergency power supply 19 are electrically connected to the extra-dry-well blower 4a, the emergency dry well cooling system equipment cooling pump 26, and the emergency dry well cooling system equipment cooling seawater pump 28. In the ninth embodiment of the present invention shown in FIG. 9, electric power can be supplied from an emergency diesel generator (EDG), i.e., the emergency power supply 19, to the emergency dry well cooling system and the emergency dry well cooling system equipment cooling system. With this embodiment, therefore, even when the normal power supply is disabled due to outage of the external power source, the emergency dry well cooling system can be operated by electric power supplied from the emergency power supply 19 to remove the heat from the reactor container. A tenth embodiment of the present invention will be described hereunder with reference to FIG. 10. The tenth embodiment is different from the seventh embodiment in that a pressure sensor 32 and a temperature sensor 33 are disposed in the reactor container 1, and an emergency dry well cooling system automatic start-up circuit 34 is installed which detects an abnormal temperature or pressure increase in the dry well 2 by receiving an output signal from the sensor 32 or 33, also detects a function outage of a residual heat removing (RHR) system 35 by receiving a function outage signal output from the same, and then automatically starts up the emergency dry well cooling system. The emergency dry well cooling system automatic start-up circuit 34 is electrically connected to the extra-dry-well blower 4a, the emergency dry well cooling system equipment cooling pump 26, and the emergency dry well cooling system equipment cooling seawater pump 28 via signal lines 37 for delivering output signals over the respective signal lines. With this embodiment, even in a case where the residual heat removing system 35 should fail to start up operation and the operator should miss the failure of the RHR system 35, the emergency dry well cooling system is automatically started up on condition that the temperature or pressure in the dry well 2 exceeds a threshold, or the function of the RHR system is disabled, making it possible to ensure integrity of the reactor container 1. An eleventh embodiment of the present invention will be described hereunder with reference to FIG. 11. The eleventh embodiment is different from the seventh embodiment in that the extra-dry-well blower 4a is connected to the return pipe 3c on the primary side of the extra-dry-well heat exchanger 3a. Stated otherwise, in the eleventh embodiment, the extra-dry-well blower 4a is disposed downstream of the extra-dry-well heat exchanger 3a. With this embodiment, since the extra-dry-well blower 4a is not exposed to the atmosphere having the temperature raised in the dry well 2, the function of the extra-dry-well blower 4a will not be deteriorated and the soundness of the reactor container can be maintained more surely. A twelfth embodiment of the present invention will be described hereunder with reference to FIG. 12. The twelfth embodiment is different from the seventh embodiment in that a header 38 is installed in the dry well 2 and the return pipe 3c on the primary side of the extra-dry-well blower 4a is connected to the header 38. With this embodiment, since the emergency dry well cooling system is provided with the header 38 at its delivery port, the atmosphere in the dry well 2 can be cooled with higher efficiency. A thirteenth embodiment of the present invention will be described hereunder with reference to FIG. 13. The thirteenth embodiment is different from the seventh embodiment in that a branch pipe 39 is connected to the pipe 4b interconnecting the reactor container 1 and the extra-dry-well heat exchanger 3a, a safety valve 40 is connected to the branch pipe 39, a discharge pipe 41 is connected to the safety valve 40, and a vent line 42 is connected to the discharge pipe 41. With this embodiment, in the event of a severe accident such as anticipated transient without scram (ATWS) sequence, if the pressure in the dry well is abruptly raised in excess of the design pressure of the reactor container, the safety valve 40 is opened at a preset pressure and, therefore, the atmosphere in the dry well can be introduced to the vent line 42 to keep soundness of the reactor container 1. According to the present invention, as described above, the heat removal from the dry well in the event of a severe accident can be performed by the normal dry well cooling system and the emergency dry well cooling system with high reliability and, at the same time, the pressure in the reactor container can also be reduced. As a result, it is possible to prevent breakage of the reactor container in the event of a severe accident. A reactor container vent system which has been contemplated to be installed in the past is problematic in that because the atmosphere in the reactor container is directly discharged to open air for pressure reduction, fission products, though a very small amount, may be discharged to the environment and heat removal from the reactor container cannot be expected. By contrast, according to the present invention, it is possible to safely settle down a severe accident without discharging fission products. According to the present invention, there are further provided the following embodiments which are represented by FIGS. 14 to 17, respectively, as fourteenth, fifteenth, sixteenth and seventeenth embodiments, in which like reference numerals are added to common members or equipments. The fourteenth embodiment will be described hereunder with reference to FIGS. 14. Referring to FIG. 14, a reference numeral 101 denotes a reactor container the inside of which is divided into an upper dry well 102a and a lower wet well 102b, and a pipe line 105 is communicated with the dry well 102a and the wet well 102b, respectively. An extra-dry-well heat exchanger 103 is also connected to the pipe line 105 and an extra-dry-well blower 104 is connected to the downstream side of the heat exchanger 103. A primary cooling circulation pipe 106 is connected to a secondary side of the heat exchanger 103 and a primary cooling system is connected to the primary cooling circulation pipe 106. The primary cooling system comprises a primary cooling pump 107, a cooling system heat exchanger 108, a secondary cooling seawater circulation pipe 109 and a secondary cooling seawater pump 110, which are operatively associated with each other. FIG. 15 represents the fifteenth embodiment of the present invention, which is similar to the fourteenth embodiment but different in an arrangement that the cooling system is composed of a direct seawater cooling system capable of supplying the seawater 111 directly to the heat exchanger 103 only through a seawater circulation pipe 112 and a seawater pump 113. According to the fourteenth and fifteenth embodiments, even in a time when an equipment is damaged through a failure of the RHR system at an accident, the atmosphere in the reactor container 101 is guided to the heat exchanger through the operation of the blower 104 to thereby perform the cooling function, thus surely maintaining the soundness of the reactor container. FIG. 16 represents the sixteenth embodiment of the present invention, in which there is located a cooling pool 118 opened to the atmosphere as a heat sink and a heat exchanger 117 is located in the cooling pool 118 filled up with water. The heat exchanger 117 is connected to the pipe line 105 and the blower 104. According to this embodiment, there is no need of any dynamic equipment such as pump for the operation of the cooling system, and because of this reason, the reactor container can be cooled with high reliability at a time of an occurrence of a severe accident. FIG. 17 represents the seventeenth embodiment of the present invention, which is different from the fourteenth embodiment in a location of a blower duct 125 in the reactor container 101. The blower duct 125 is connected to the pipe line 105. According to this embodiment, since the blower duct 125 is connected to a discharge side of the emergency dry well cooling system, the atmosphere in the reactor container 101 can be effectively cooled. As can be seen from the above disclosure and drawings of FIGS. 14-17, according to the fourteenth to seventeenth embodiments, the pipe line 105 is connected to both the dry well and wet will, which is different from the former embodiments in which the pipe line is connected only to the dry well. In a case where an accident occurs in a normal operation condition, the reactor scrams, and thereafter, vapor in the RPV is flown out in the vent tube or into the suppression pool through a safety relief valve. At this time, the vapor is condensed by the water in the suppression pool and a fission product contained in the vapor is captured in the suppression pool water through a scrubbing effect and transferred to the wet well gas phase. According to these embodiments of the present invention, the atmosphere in the wet well is circulated to the suppression pool liquid phase through the cooling system, the dry well and the vent tube, thus the fission product in the suppression pool gas phase being scrubbed more effectively. When the accident occurs during the normal operation, the reactor is shut-down and the core is cooled. Thereafter, when the reactor container is cooled, by the RHR system through one of the suppression pool cooling mode and the dry well spray mode. In the prior art technology, the suppression pool cooling means is not provided, but according to the present invention, the safeness of the reactor can be realized with high performance. Further, it will be easily understood that the first to thirteenth embodiments may be selectively applicable to the fourteenth to seventeenth embodiments of the present invention by persons skilled in the art without specifically describing herein through drawings. It is also to be noted that the present invention will be described hereinbefore with reference to the preferred embodiments but it is not limited to them and many changes and modifications may be made without departing from the subjects and scopes of the appended claims.
048083690	summary	BACKGROUND OF THE INVENTION This application is a continuation of application Ser. No. 026,842, filed Mar. 17, 1987. This invention relates to an emergency core cooling apparatus, and more particularly to an emergency core cooling apparatus having high-pressure systems and low-pressure systems. A boiling water nuclear power plant is provided with an emergency core cooling apparatus for dealing with a loss of coolant accident. An example of a conventional emergency core cooling apparatus for a boiling water reactor is discussed under "Current Status of Advanced Boiling Water Reactor (ABWR)" in the "Hitachi Review" (December 1984 issue, Vol. 33 - No. 6), pages 299-306. As will be described in detail later, this conventional emergency core cooling apparatus has two systems of high-pressure core spray apparatus (which will hereinafter be referred to as HPCS), one system of high-pressure coolant injection apparatus (which will hereinafter be referred to as HPCI) and three systems of low-pressure flooding apparatus (which will hereinafter be referred to as LPFL). The HPCS's and HPCI are high-pressure emergency core cooling apparatuses, and the LPFL's low-pressure emergency core cooling apparatuses. SUMMARY OF THE INVENTION An object of the present invention is to provide an emergency core cooling apparatus capable of preventing the effective heat generating portion of a core from being exposed no matter what kind of breakage occurs in a pipe for an emergency core cooling apparatus. Another object of the present invention is to provide an emergency core cooling apparatus capable of cooling a core when the nuclear reactor is stopped under normal conditions, by using compact coolers. The first characteristics of the present invention reside in that the elevation of the coolant discharge ports, which are in a reactor vessel, of a plurality of high-pressure emergency core cooling systems, which are adapted to supply a coolant to the interior of a core-surrounding shroud in the reactor vessel, is set higher than that of the coolant discharge ports, which are in the shroud, of a plurality of low-pressure emergency core cooling systems, which are adapted to supply a coolant to the interior of the reactor vessel. Since the positions of the coolant discharge ports, which are in the reactor vessel, of the high-pressure emergency core cooling systems are set higher than those of the coolant discharge ports, which are in the shroud, of the coolant discharge pipes for the lower-pressure emergency core cooling systems, a core cooling operation can be carried out effectively by the high-pressure emergency core cooling systems from a point in time earlier than the time of depressurization of the nuclear reactor in the case where the breakage of a pipe (breakage of a pipe in a low-pressure emergency core cooling system) occurs in a low position, at which the securing of a predetermined water level in the nuclear reactor must be done under severe conditions at the time of occurrence of the breakage of a pipe, in the nuclear reactor. When the breakage of a pipe (breakage of a pipe in a high-pressure emergency core cooling system) occurs in a high position, a large quantity of water resides in the nuclear reactor, and, moreover, the water level in the nuclear reactor rapidly reaches the elevation of the coolant discharge ports of the high-pressure emergency core cooling systems, so that the depressurization rate in the nuclear reactor due to the steam discharge becomes high. Consequently, the occurrence of an increase in the coolant injection rate of the remaining high-pressure core cooling systems and the early starting of the injection of a coolant by the low-pressure core cooling systems can be expected. As a result, a decrease in the water level in the nuclear reactor is suppressed. The second characteristics of the present invention reside in that the requirements, which are other than the requirements constituting the first characteristics mentioned above, are added thereto, the additional requirements consisting of providing the coolant discharge ports, which are in the shroud in the low-pressure emergency core cooling systems, at the portions of the interior of the shroud which are above the core, and providing the cooling means in the low-pressure emergency core cooling systems with a means for supplying the coolant, which is in the portion of the interior of the reactor pressure vessel which is below the core or between the shroud and reactor vessel, to the core when the nuclear reactor is stopped under normal conditions. According to such second characteristics, the low-pressure emergency core cooling systems are furnished with the functions of supplying a coolant to the core, and the coolant in the reactor vessel is supplied from the portion of the interior of the reactor vessel which is on the outer side of the shroud or the lower portion of the interior thereof to the cooling means in the low-pressure emergency core cooling systems. Accordingly, when the nuclear reactor is stopped under normal conditions, the coolant cooled by the cooling means passes through the core necessarily, so that the core can be cooled reliably. Since the high-temperature coolant, which has passed through the core, is necessarily supplied to the cooling means in the low-pressure emergency core cooling systems, the operation efficiency of these cooling means can be improved, and the capacity of these cooling means can be reduced.
	summary
	claims	1. A method for supporting maintenance for a molding system in which apparatuses perform operational steps during operation cycles of the molding systems and for determining or anticipating which of the apparatuses must be maintained before operation of the molding system is caused to be interrupted by an apparatus malfunction, including the steps of measuring, at short intervals, cycle times that correspond to a time for completion of each operation cycle of the molding system, and operation times of the operational steps performed by the apparatuses, wherein the operation times measured are those affecting each cycle time, and storing data on the measured cycle times and the measured operation times in a database, retrieving from the database data on the measured cycle times and specifying any measured cycle time that is longer than a normal cycle time, and determining which operational step, among the operational steps performed by the apparatuses that affect the specified measured cycle time, may cause an interruption of the operation of the molding system, based on a sum of ones of the measured operation times for each operational step that exceed a predetermined time, or a sum of the number of measured operation times for each operational step that exceed the predetermined time. 2. A method for supporting maintenance for a molding system in which apparatuses perform operational steps during operation cycles of the molding system and for determining which of the apparatuses must be maintained before operation of the molding system is caused to be interrupted by an apparatus malfunction, including the steps of measuring, at short intervals, cycle times that each correspond to a time for completion of each operation cycle of the molding system, and operation times of the steps performed by the apparatuses, wherein the operation times measured are those affecting the cycle times, and storing data on the measured cycle times and the measured operation times in a database, retrieving data on the measured cycle times from the database and specifying any measured cycle time that is longer than a normal cycle time, determining which operational step, among the operational steps performed by the apparatuses that affect cycle times, may cause an interruption of the operation of the molding system, based on a sum of ones of the measured operation times for each operational step that exceed a predetermined time, or a sum of the number of measured operation times for each operational step that exceed the predetermined time, and monitoring changes between subsequently measured operation times of the determined operational step. 3. A method for supporting maintenance for a molding system in which apparatuses perform operational steps during operation cycles of the molding system and for anticipating which of the apparatuses must be maintained before operation of the molding system is caused to be interrupted by an apparatus malfunction, including the steps of measuring, at short intervals, data on cycle times each corresponding to a time for completion of each operation cycle of the molding system, and operation times of operational steps performed by the apparatuses, wherein the operation times measured are those affecting the cycle times, and storing data on the measured cycle times and the measured operation times in a database, retrieving measured cycle times from the database and specifying any measured cycle time that is longer than a normal cycle time, determining which operational step, among the operational steps of the apparatuses that affect cycle times, may cause an interruption of the molding system, based on a sum of ones of the measured operation times for each operational step that exceed a predetermined time, or a sum of the number of measured operation times for each operational step that exceed the predetermined time, and monitoring transitions of subsequently measured operation times of the determined operational step. 4. A method for supporting maintenance of a molding system of the type in which a plurality of apparatuses are operative during operation cycles of the molding system, and wherein the plurality of apparatuses perform operational steps during the operation cycles, the method including the steps of determining or anticipating which of the plurality of apparatuses must be maintained before operation of the apparatuses degrade to a condition where operation of the molding system is interrupted, comprising the steps of: measuring, at short intervals, cycle times of the operation cycles of the molding system and operation times of the apparatuses, wherein each measured cycle time corresponds to a time for completion of one of the operation cycles of the molding system, and each measured operation time corresponds to a time for completion of one of the operation steps performed by the apparatuses and which operational step is capable of affecting the measured cycle times; storing the measured cycle times and the measured operation times so that a plurality of measured cycle times are stored, and so that for each of the operational steps a plurality of measured operation times are stored; identifying any measured cycle time that is longer than a normal cycle time for the corresponding operation cycle; and evaluation for each optional step performed during and capable of affecting the operation cycle corresponding to the identified measured cycle time, a sum of corresponding measured operation times that exceed a predetermined time, or a sum of the number of the corresponding operation times that exceed the predetermined time, to identify which operational step and associated apparatuses may cause an interruption of the molding system, wherein the corresponding measured operation times are obtained during a plurality of repetitions of the operation cycles. 5. The method of claim 4 , further including the step of: claim 4 monitoring changes in the operation times of the identified operational step for changes which exceed predetermined criteria to identify apparatuses which should be repaired. 6. The method of claim 4 , further including the step of: claim 4 monitoring transitions between the operation times of the identified operational step, for changes which exceed predetermined criteria, to identify apparatuses which should be repaired.
062087045	description	DETAILED DESCRIPTION OF THE INVENTION The features and other details of the apparatus and method of the invention will now be more particularly described with reference to the accompanying drawings and pointed out in the claims. The same number present in different figures represents the same item. It will be understood that the particular embodiments of the invention are shown by way of illustration and not as limitations of the invention. The principle features of this invention can be employed in various embodiments without departing from the scope of the present invention. The specific activity of a radioisotope, within a volume of target material, is the number of radioactive disintegrations per second of includes of the radioisotope (in curies (Ci)) measured per gram of the radioisotope's element, including all isotopes of the element, within the volume of target material. Specific activity provides an indication of the concentration of the radioisotope within the volume of target material. Typically, the specific activity is not uniform across a volume of target material, but is averaged across the volume of target material. The level of specific activity, which constitutes a high specific activity, is dependent upon the radioisotope and its use. For example, wherein the radioisotope is molybdenum-99 (Mo.sup.99), which subsequently decays to the daughter product technetium-99 (Tc.sup.99), a high specific activity for Mo.sup.99 is typically an average specific activity of about 0.5 Ci/gram of molybdenum, or more. Preferably, the specific activity of Mo.sup.99 is about 1.0 Ci/gm, or more. More preferably, a high specific activity of Mo.sup.99 is about 5 Ci/gram, or more. Even more preferably, a high specific activity of Mo.sup.99 is about 10 Ci/gram, or more. A radioisotope can be generated in a target material using high energy photons from a photon beam in at least one isotopic conversion reaction. A target material is a material which consists of or contains a targeted isotope, which when exposed to high energy photons, forms the radioisotope as a product. Typically, a targeted isotope has a high atomic number (Z), for example, a Z of about 30 or more. A radioisotope product can be a final product, such as Cadmium-115 or Tantalum-179. Alternatively, a radioisotope product, such as Cadmium-109 or Osmium-191, can be an intermediate which subsequently decays to form a desired daughter product. Preferably, a radioisotope product is longer-lived. A longer-lived radioisotope, as defined herein, is a radioisotope with a half-life suitable to allow shipping and the subsequent use of the radioisotope, or a daughter product, after generating the radioisotope. Typically, a longer-lived isotope has a half-life of about 12 hours or more. Preferably, the half-life is about 48 hours or more. More preferably, the half-life is about 60 hours or more. Most preferably, the radioisotope product is Mo.sup.99. Suitable isotopic conversion reactions include, for example, (.gamma.,n), (.gamma.,2n), (.gamma.,p) and (.gamma.,pn) reactions. An energy level, suitable for a high energy photon, is an energy level which is at least equal to the threshold (minimum) energy level, of the Giant Resonance region of the cross-section versus energy curve for the desired isotopic conversion reaction, required to produce the reaction between a photon and the targeted isotope. The specific activity of a photon-beam generated radioisotope, within a volume of a target material, depends upon several variables, including the intensity (photon energy per unit area per unit time) of the high energy photons in the photon beam and the thickness of the target material. As shown in FIG. 1, the peak specific activity level, for a photon beam of any intensity, is at the target material surface irradiated by the photon beam. A photon beam with a higher intensity of high energy photons, irradiating the same target material, typically generates a higher peak specific activity than does a photon beam with a lower intensity of high energy photons. A high intensity of high energy photons is an intensity sufficient to generate a high specific activity of a radioisotope. Typically, a suitable intensity of high energy photons is that derived from an electron beam of at least 50 microamps/cm.sup.2 (.mu.a/cm.sup.2). Preferably, the intensity of high energy electrons is at least 500 .mu.a/cm.sup.2. More preferably, the intensity of high energy electrons is at least 1,000 .mu.a/cm.sup.2. In addition, as also shown in FIG. 1, specific activity levels within the target material decrease exponentially with increasing depth along the thickness of the target material. The thickness of the target material is the distance from the irradiated side of he target material to the opposite face. Thus, the average specific activity of a radioisotope within a volume of target material increases with decreasing target material thickness. The maximum specific activity (saturation activity) achievable by isotopic conversion in a volume of target material varies linearly with the production rate of the radioisotope. Typically, saturation activity is achieved only following irradiation periods that are significantly longer than the half-life of the radioisotope. Saturation activity (S) is calculated by the following equation: EQU S=1.62.times.10.sup.13 f.multidot.R/A wherein f is the fraction of isotope of the targeted element which is targeted isotope and A is the atomic weight of the targeted element. R, which is indicative of the intensity of high energy photons, is the photon path length per unit volume and per unit energy (".phi.(E)") weighted by the photon cross-section (".sigma.(E)"), in targeted over all photon energy levels. The specific formula for calculating the value of R is as follows: EQU R=.intg..sigma.(E).multidot..phi.(E).multidot.dE. The photon energy levels included in the calculation of R may be limited to those in the Giant Resonance range as lower energy photons are not effective. Specifically, lower energy photons do not result in photonuclear conversion of Mo.sup.100 to Mo.sup.99. One embodiment, of the apparatus for producing a high specific activity of a product radioisotope in a volume of target material, is illustrated in FIG. 2. Apparatus 10 includes target material 12, convertor 14 and electron accelerator 16. Target material 12 contains a loading of a targeted isotope which can be established based upon the intended isotopic conversion reaction and the concentration of product radioisotope desired. The specific isotopic conversion reactions occurring within target material 12 typically depend upon the desired product isotope and the availability of nuclei of the targeted isotope within target material 12. In one embodiment, the loading of a targeted isotope in target material 12 is at naturally occurring levels. Preferably, target material 12 contains enriched levels of the targeted isotope. The targeted isotope can be in elemental form, in at least one compound (e.g., a salt or oxide), and/or complexed. The targeted isotope within the target material can be in any physical state, for example, a particulate, a liquid, in solution, in a suspension, in a slurry, or a in a larger solid mass. Examples of other components optionally contained in target material 12 include materials in which the targeted isotope is retained, such as a metallic or ceramic material, or materials in which the targeted isotope is dispersed such as in a liquid (e.g., water or oils) or in particulates. Apparatus 10 further includes electron beam 18 and photon beam 20. Electron beam 18 is generated by electron accelerator 16 and is directed into convertor 14, wherein photon beam 20, which includes high energy photons, is generated. Photon beam 20 radiates from convertor 14 into target material 12. Typically, photon beam 20 is a substantially collimated high energy photon beam. A suitable convertor contains at least one high Z material, for example tungsten or platinum, which is refractory under the conditions of the method of invention. A high Z material is used to improve the efficiency of the conversion within convertor 14 of high energy electrons from electron beam 18 into high energy photons to form photon beam 20. The total extent of convertor 14 in the direction of the trajectory of electron beam 18 should be sufficient to absorb a significant portion of the energy of electron beam 18 while transmitting photon radiation in an energy range suitable for the desired isotopic conversion reaction. Concurrent with transforming the energy of electron beam 18 into high energy photons in photon beam 20, convertor 14 also shields target material 12 from any significant residual electron beam. If convertor 14 is too thick, photons emitted from convertor 14 will be degraded in energy due to passing through the material of convertor 14. If convertor 14 is too thin, significant levels of electrons will pass through convertor 14 and impinge upon target material 12. The preferred thickness of convertor 14, for obtaining optimum product isotope yield, depends on electron beam energy, the composition of convertor 14, and the Giant-Resonance region threshold energy of the targeted isotope. An example of an optimal convertor is a convertor containing approximately six plates of tungsten alloy of aggregate thickness 5 mm separated by cooling ducts for water cooling. The intensity of high energy photons generated in convertor 14 is proportional to the power density (PD) of electron beam 18 in convertor 14. Thus, the specific activity of a radioisotope within a volume of target material 12 is also proportional to the power density. Power density within convertor 14 is calculatable by the following equation: EQU PD=E.times.i/V wherein E is the energy of electron beam 14, i is the current of electron beam 18 and V is the volume of convertor 14 through which electron beam 18 passes. The power density used in this invention is limited by the heat removal capacity of convertor 14. In another embodiment illustrated in FIG. 3, convertor 14 is composed of two or more plates 22 of high Z material, such as tungsten, instead of a single solid convertor to allow better heat removal from convertor 14 and thus, higher power densities of electron beam 18 therein. Plates 22 can be fabricated from the same or different material. The plates are typically enclosed by external shell 24, which maintains the geometry of convertor 14 and also retains any optional coolant within convertor 14. In a preferred embodiment, plates 22 do not have equal thicknesses. The thicknesses of the plates is varied to equalize the heat loads on the plates. The heat load on each plate is derived from the energy transferred to the plate by electron beam 18 and by generated photons passing through each plate. Typically, the heat loads on plates distal to electron accelerator 16 are greater than the heat loads on proximal plates as electron beam 18 deposits energy in a plate after the electrons are slowed by previous plates. In addition, photons generated in the proximal plates can also deposit energy in subsequent, distal plates. Thus, in a more preferred embodiment, plates 22 proximal to electron accelerator 16 are thicker than plates 22 which are distal to the electron accelerator 16 to better equalize the heat generation in each plate 22. Plates 22 and cooling channels 26 in convertor 14 do not need to be perpendicular to the direction of electron beam 18. Preferably, the cross-sectional areas of convertor 14, or plates 22, are perpendicular to the path of electron beam 18. Optionally, means are provided for removing heat from at least a portion of convertor 14. Heat removal is provided by typical means, such as by radiation, conduction and/or convection. Heat removal means are disposed around and/or through convertor 14. Examples of suitable heat removal means include coolant channels 26 which are disposed within the material forming convertor 14 (e.g., wherein the convertor material is a honeycomb), etched along the surface of convertor 14, etched along the surface of plates 22 and/or are disposed between plates 22. Alternatively, convertor 14 includes porous material in the form of frit wherein coolant flows through the interstices within the frit for heat removal. Heat removal means also include convertor inlet 28 and convertor outlet 30, which are disposed at shell 24 of convertor 14. Preferably, heat generated within convertor 14, or within each plate 22 of convertor 14, is removed by fluid coolant flow into convertor 14 through convertor inlet 28, through coolant channels 26 and out of convertor 14 through convertor outlet 30. Suitable means of fluid coolant flow include, for example, single-pass fluid flow, natural circulation and forced recirculation. Typically, outside of convertor 14, the coolant is then cooled, such as by being directed through heat exchanger 32A. Suitable fluid coolants include liquids, such as water or liquid gallium and gases, such as helium. For very high power densities within convertor 14, such as greater than about 3 thousand watts/cm.sup.3 or more, it is preferred that convertor 14 be a porous metallic frit which is cooled by fluid coolant flowing at high pressure through the pores, or interstices, within the frit. In the embodiment wherein convertor 14 is tungsten and the targeted isotope is Mo.sup.100 the optimum yield of a Mo.sup.99 product isotope yield is when plates 22 of convertor 14 have a combined thickness slightly less than the stopping distance for an electron in electron beam 18. When plates 22 have a combined thickness less than the electron stopping distance, backing 34 is disposed between convertor 14 and target material 12 to capture electrons without significantly degrading the energy of the photon beam. Suitable materials for backing 34 include lower Z metals such as aluminum. Typically, the high energy photon beam is directed through backing 34 at or near the center of backing 34. Further, the cross-sectional area of backing 34 is preferably equal to or larger than the width of high energy photon beam 18. Optionally, backing 34 can be cooled by means for removing heat, not shown, such as heat transfer to a cooling medium (e.g., water). In yet another embodiment illustrated in FIG. 4, convertor 14 consists of molten or liquified high Z material 33, which is recirculated from convertor inlet 28, through convertor 14, out of convertor outlet 30, through heat exchanger 32B, and subsequently back into convertor inlet 28. Heat generated in convertor material 33 within convertor 14 by the electron beam then dissipates, or is removed by suitable means, such as heat exchanger 32B, while the convertor material is outside of the convertor. FIG. 5 illustrates an alternative embodiment of the apparatus of this invention wherein separate, or separable, increments of target material 12 are irradiated in series thereby producing a high specific activity of radioisotope in the first increment and pre-irradiating the second increment to commence building up the concentration of the radioisotope within the increment. Apparatus 100 includes target assembly 36, convertor 14 and electron accelerator 16. Electron beam 18 is generated by electron accelerator 16 and is directed into convertor 14, wherein photon beam 20, which includes high energy photons, is generated. Photon beam 20 extends from convertor 14 into target assembly 36. Target assembly 36 includes a target material which is separated or separable into at least two increments, with first target material increment 38 located proximal to convertor 14 and second target material increment 40 located adjacent to first target material increment 38 and distal to converter 14. Additional targets material increments 42 are disposed, in series, behind second target material increment 40. An increment of a target material is an amount of target material which is separate or separable from the target material contained within target assembly 12. Each increment of target material, such as first target material increment 38, second target material increment 40 and additional target material increments 42, contains a loading of a targeted isotope within the target material of the target. Typically, wherein the targeted isotope is contained within a larger solid mass, first target material increment 38 and second target material increment 40 consist of separate sections of the target material. Target assembly 36 also includes inlet 44A and outlet 46A. Inlet 44A is disposed at or near the end of target assembly 36 distal to convertor 14. Inlet 44A is provided as a means for directing additional targets 21 into target assembly 36 on the distal side of second target material increment 40. Outlet 46A is disposed at or near the end of target assembly 36 that is proximal to convertor 14. Outlet 46A is provided as a means for separating a distal target material increment from its adjacent target material increment (e.g., separating first target material increment 38 from second target material increment 40) by directing the distal target material increment out of target assembly 36 through outlet 46A. Preferably, target assembly 36 also includes means, such as pushrod 48, for conveying increments of target material through target assembly 36 toward convertor 14, and then out of target assembly 36. Alternatively, other known means for non-destructively conveying target material can also be used to convey targets or target material through target assembly 36. Examples of other suitable conveying means include, for instance, conveyor belts, screws, pistons and pumps. The target assembly 36 may further include photon reflector 50. Photon reflector 50 is disposed around at least a portion of target assembly 36. Photon reflector 50 is typically composed of high Z metals (e.g., a Z of about 30 or more), such as molybdenum-98, uranium, tantalum, tungsten, lead and other heavy metals. Photon reflector 50 reflects at least a portion of the high energy photons impinging the reflector material (e.g., from the incoming photon beam or scattered from the in-series target material increments) into the target material within target assembly 36. Optionally, target assembly 36 includes neutron shielding 52 which is disposed at least partially around photon reflector 38. Suitable types of neutron shielding include shielding with a high hydrogen content, such as a plastic or water, which thermalizes and/or captures at least a portion of the neutrons emitted during an isotopic conversion reaction. The depth of target material 12 through which photon beam 20 passes within the aggregate of in-series target material increments, disposed within target assembly 36 is determined based upon the loading of targeted isotopes within each increment, the desired concentration of product isotopes within each increment, the energy level of photon beam 20 and the period of irradiation. Preferably, the target material, contained in the in-series target material increments, has an aggregate thickness that results in the capture of all but an insignificant amount of high energy photons in photon beam 20 which impinge the target material and do not scatter outside of the target material. For example, wherein the targeted isotope is Mo.sup.100 and the desired product is Mo.sup.99, the aggregate thickness of the targets is typically between about 6 cm to about 10 cm for a photon beam produced by a tungsten convertor exposed to a 30-40 Mev electron beam. The cross-sectional area of target material 12 within target assembly 36 perpendicular to photon beam 20 can be varied depending upon the focal area of photon beam 20 on target material increment 38 and the expected spread of the photon beam 20 along the path of photon beam 20 through target material 12. The cross-sectional area of target material 12 is usually about equal to, or larger than, the focal area of photon beam 20. In an alternative embodiment illustrated in FIG. 6, target material 12 is in a particulate, liquid, slurry or any other physical form wherein an increment of target material 12 is not contained in a single solid mass. Thus, increments of target material 12 are not separate but are separable. Target assembly 36 includes means for containing target material 12 within target assembly 36, such as cylinder 54 which is disposed within target assembly 36. Suitable containing means, include containers for solids and/or liquids, which are refractory, such as titanium. The material composition and structural design of the container should not result in a significant reduction in the energy of photon beam 20 or a significant increase in the scatter of photons from photon beam 20. Cylinder 54 includes baffles 55 which control the flow in cylinder 54 to assure generally uniform irradiation. Target assembly 36 also includes means for directing increments of target material 12 through cylinder 54. This directing means includes inlet 44B and outlet 46B. Inlet 44B is disposed at or near the end of cylinder 54 distal to convertor 14. Outlet 46B is disposed at or near the end of cylinder 54 that is proximal to convertor 14. In this embodiment, target material 12, which is typically in liquid, slurry or particulate form, is directed into cylinder 54 through inlet 44B, moves towards and the proximal end of cylinder 54, and then comes out of cylinder 54 through outlet 46B. The movement (e.g., flow) of target material 12 through cylinder 54 can be continuous of intermittent. Suitable means to direct flow of target material 12 include, for example, pumps, pistons and gravity feeding. The flow of target material 12 through cylinder 54 can be controlled, for instance, by a valve or clamp located in a position suitable to stop flow (e.g., at inlet 44B or outlet 46B) and/or by controlling the flow directing means (e.g., starting and stopping a pump). In another embodiment illustrated in FIG. 7, wherein the increments of target material 12 are separate, but not solid masses, target assembly 36 further includes means for separately containing each increment of target material 12. Typically, target material 12 is in a particulate, liquid or slurry form. Suitable containing means, such as container 56, include containers which can contain a solid and/or liquid, wherein the container is refractory under the method of this invention. The material composition and structural design of the container should not result in a significant reduction in the energy of photon beam 20 or a significant increase in the scatter of photons from photon beam 20. An example of a suitable container material is titanium. In this embodiment, containers 56 enter the distal end of target assembly 36 through inlet 44B, are directed toward the proximal end of target assembly 36 while concurrently being irradiated by photon beam 20, and then leave target assembly 36 through outlet 46B. Operation of the embodiment of FIG. 2 for producing a high specific activity of a radioisotope will now be described. Electron accelerator 16 generates electron beam 18 which is directed into convertor 14. At least a portion of the electrons of electron beam 18 are captured in an (electron,.gamma.) reaction by the high Z material of convertor 14 to generate photons, including high energy photons in photon beam 20. Typically, most electrons are captured and most photons pass through convertor 14. Typically, electron accelerator 16 generates an electron beam 18 with an average energy level of about 25 MeV or more, preferably between about 30 MeV and about 50 MeV. The total power of electron beam 18 is limited by the design of electron accelerator 16 and by the design, thickness and heat removal capability of convertor 14. If the beam energy is too low, there will not be sufficient photons in the Giant Resonance region to produce a high specific activity of the radioisotope and the electron range in convertor 14 will be so short as to make heat removal from convertor 14 very difficult. If the beam energy is too high, many photons will have energies above the optimal range, direct electron heating of target material 12 will be a problem and electron accelerator 16 will be relatively expensive. In addition, increased production of impurities, such as niobium, can result for other isotopic conversion reactions. Photon beam 20 is directed from convertor 14 and focused onto target material 12. Target material 12 is typically placed in close proximity to convertor 14 and in alignment with the exit of photon beam 20 from convertor 14. Sufficient distance between convertor 14 and target material 12 may be left to interpose material to attenuate electromagnetic fields to deflect electron beam 18 or to interpose material to modify the photon spectrum of photon beam 20, but this distance is minimized in order to use the photon beam at high intensity. If no attenuation is required, target material 12 may be in contact with convertor 14. Within target material 12, at least a portion of the high energy photons of photon beam 20, react with the targeted isotope to form a concentration of a radioisotope within the target material by an isotopic conversion reaction, such as by (.gamma.,n), (.gamma.,2n), (.gamma.,p) or (.gamma.,pn) reaction. Preferably, a significant number of the photons of photon beam 20 are high energy photons which have an energy level falling within the range of energy levels included in the Giant Resonance region of the cross-section versus energy curve for the desired isotopic conversion reaction. More preferably, a significant portion of the photons of photon beam 20 have energy levels about equal to the peak energy level of the Giant Resonance region. For heavier materials, the energy levels corresponding to the Giant Resonance region are relatively lower while for lighter materials the energy levels are relatively higher. Preferably, the energy of electron beam 18 should be about 2 to about 3 times the energy level of the peak of the Giant Resonance region of the targeted isotope. For example, in the (.gamma.,n) isotopic conversion of Mo.sup.100 to Mo.sup.99 it is preferred that at least a significant portion of photons in photon beam 20 have energy levels falling within the Giant Resonance region for this reaction, specifically between the threshold energy level of about 10 MeV and the high energy limit of about 19 MeV. More preferably, photon energy levels are about 15 MeV, which is the peak of the Giant Resonance region. The electron beam energy for this isotopic conversion is typically between about 25 Mev to about 50 Mev, with a preferred energy range of about 35 Mev to about 40 MeV. The energy level of a generated photon is directly dependent upon the energy level of electron beam 18, with the peak energy level of generated photons being equal to about the energy level of electron beam 18. Typically, most generated photons have energy levels at less than half the peak energy. Therefore, the energy level of at least a portion of the electrons in electron beam 18 at a minimum must be equal to the threshold (minimum) energy level required to produce the desired isotopic conversion reaction between a generated photon and the targeted isotope. Preferably, the energy level of electron beam 18 is within or above the Giant Resonance region of the desired isotopic conversion reaction. In a preferred embodiment, wherein the targeted isotope is molybdenum-100 (Mo.sup.100) which is isotopically converted to molybdenum-99 (Mo.sup.99), which then decays to the desired daughter product technetium-99 (Tc.sup.99), the photon beam produced includes .gamma. radiation at an energy level of about 8 Mev or more. More preferably, a substantial amount of the .gamma. radiation produced is at energy levels between about 8 Mev and about 16 MeV. Achievement of an average specific activity of Mo.sup.99 of about 1.0 Ci/gm of Mo in solid molybdenum requires a relatively high power density in convertor 14. Specifically, in the saturation activity equation, the product of f.multidot.R must have a value greater than about 2.2.times.10.sup.-8 sec.sup.-1. This value of R is difficult to achieve because of technical limitations on electron beam power density and convertor heat removal. Therefore, the volume in which the average specific activity of 1.0 Ci/gm can be maintained is typically limited to target material volumes having relatively small thicknesses. In determining the maximum volume of target material, the cross-sectional area of the target material usually must be equal to or less than the focal area of photon beam 20. Thus, target material volume is often limited to a few cubic centimeters or less. For example, for a natural molybdenum target, containing approximately 10% Mo.sup.100, a 35 Mev electron beam of 1.0 milliampere current focused onto a 1.0 cm radius target disk yields, with an optimal convertor, an average specific activity of about 1.0 Ci/gm for a target material thickness of about 0.5 cm. The power density in the active regions of the convertor would be about 35,000 watts/cm.sup.3. Higher specific activities can be achieved by isotopic enrichment of the target material. A target material enriched to 100% Mo.sup.100 would yield a specific activity in excess of 10 Ci/gm up to a target material thickness of about 0.5 cm for the same conditions. Molybdenum target thicknesses greater than 0.5 cm, having an average specific activity of at least 1.0 Ci/gm, can be obtained by varying the isotopic enrichment of Mo.sup.100 in the target material and/or by varying the energy levels of the photons in the photon beam, providing the value of the product f.multidot.R is at least 2.2.times.10.sup.-8 sec.sup.-1. For a thick target, the activity produced in the first 0.5 cm depth of the target is only 28% of the total generated in the target. However, the other 72% of the desired product isotope is so diluted with unconverted target material as to be below commercial interest. On the other hand, to irradiate a single target of 0.5 cm thickness or less results in lost photon energy. The portion of the thick target with less than threshold activity represents a potentially valuable resource, unusable if unimproved. Accordingly, by providing an incremental target as in FIG. 5, only that portion of the target which has been irradiated to an average specific activity above a given threshold value is removed for processing. Additional portions of the target, irradiated to less than the threshold value, can be sequentially irradiated to the threshold value in such fashion as to optimize the combination of specific activity of individual target elements and total radioisotope production rate. Preferably, each target increment is 0.5 cm thick or less. Within, at least, first target material increment 38 and second target material increment 40, a portion of the high energy photons of photon beam 20, react with the targeted isotope to form a high specific activity in first target material increment 38 and pre-irradiate second target material increment 40, and possibly additional target material increments 42, to commence building up the specific activity of the radioisotope within these increments. This method also includes moving first target material increment 38 and second target material increment 40 toward outlet 46A, and closer to convertor 14, by the action of push rod 48 applying force to the distal side of second target material increment 40 through additional target material increments 42. Alternately, the targets can be moved by any suitable automated or non-automated means. Further, the movement of targets can be continuous, concurrent, sequential or stepwise. Ultimately, first target material increment 38 is pushed through outlet 46A and is removed from target assembly 36. Further second target material increment 40 is pushed to the original position of first target material increment 38 whereupon photon beam 20 then focuses upon second target material increment 40 to complete producing a high specific activity therein. Additional target material increments 42 can be added in-series behind second target material increment 40 through inlet 44A. In this method, the ratio of the specific activity of the product radioisotope in each increment to the amount of product isotope removed per unit time can be optimized depending upon the need for a high discharge rate of product radioisotope or a high specific activity of product radioisotope. The concentration of the product radioisotope generated by the isotopic conversion reaction is dependent upon the intensity of the high energy photons in photon beam 20, upon the volume of target material 12 irradiated, upon the radioactive half-life of the product isotope, and upon the amount of target material 12 which is irradiated. The intensity of photons is approximately dependent linearly upon the current level of electron beam 18 for the same focal area, with higher currents generating more high energy photons per unit time, which then are directed into the target material to react with more targeted isotope per unit time. The volume of target material 12 irradiated by photon beam 20 depends upon the focal area of photon beam 20 upon target material 12 and the amount of photon scatter within the target material. Typically, the focal area of photon beam 20 is a function of the angle of emission of high energy photons from convertor 14. Most higher energy photons, having an energy level which falls within the Giant Resonance region for the desired isotopic conversion reaction, are emitted in a narrow cone whose axis is aligned along the direction of an extended axis of electron beam 18. The intensity of higher energy photons, which are emitted at an angle to the axis of the cone, rapidly decreases as the angle from the cone increases. For instance, at an angle of about 5 degrees from the axis of the cone, the intensity of peak photons is about one fifth of the intensity of peak photons emitted about the center of the cone. In addition, the intensity of higher energy photons, having approximately one-half peak photon energy, is lower by about two orders of magnitude at an angle of 25 degrees from the axis of the cone than the intensity along the axis of the cone. Thus, photon beam 20 is strongly peaked in the forward direction along an extended axis of electron beam 18. Therefore, the focal area of photon beam 20 is determined by the focal area of electron beam 18 on convertor 14. With increasing electron beam energies, the focal area of photon beam 20 becomes smaller with a minimum area being the size of the focal area of electron beam 18 on convertor 14. Thus, with increasing photon beam energies, the cross-sectional area of target material 12 is further limited. To optimize the specific activity of product radioisotope in each target material increment, when removed from the target assembly, the focal width of photon beam 20 is minimized to produce a higher concentration of product radioisotope near the center of first target material increment 38 with lower concentrations near the edges of the target. As photon beam 20 travels through the target material and spreads, such as from scattering, the concentration is reduced near the center of the target material and is increased nearer to the edges of the target material 12. Thus, after passing through first target material increment 38, photon beam 20 will pre-irradiate second target material increment 40 and additional target material increments 42 to produce lower levels of product isotope throughout these incremental targets (e.g., near the centers and at the edges). Preferably, the focal area of electron beam 18 is minimized to attain greater concentrations of product isotope near the centers of the targets. The lower limit on focal area of electron beam 18 on convertor 14 is dependent upon the heat dissipation capability of convertor 14. The focal area of electron beam 18 should not be so small as to create a high power density in the affected potion of convertor 14 which leads to localized melting, destruction and/or loss of function of the convertor material. The amount of time a target is irradiated can depend upon the movement rate of the in-series target material increments, while in photon beam 20, toward outlet 46. Target material increments are introduced, moved and discharged at rate such that the combination of segment thickness and discharge rate yields a product of the desired specific activity of product isotope. A high discharge rate of targets will result in the recovery of a larger fraction of the generated radioisotope but the specific activity of the discharged material will not be as high as that which would result, all other factors remaining unchanged, from a low target material increment discharge rate. FIG. 8 further illustrates the calculated effect on production rate and specific-activity of product of varying the flow rate of target material within the photon beam. FIG. 8 is based upon an electron beam energy of 35 MeV, an electron beam current of 1.10 ma, and cylindrical Mo.sup.100 target segments which are 2.0 cm in radius and 0.5 cm thick. The method of this invention can also be employed to produce concentrations of stable isotopes. The invention will now be further and specifically described by the following examples. EXAMPLE 1 Mo.sup.99 Production by Photonuclear Transmutation of Mo.sup.100 A cylinder of molybdenum (4 inches diameter), having a natural isotopic abundance, was sliced in planes perpendicular to the length of the cylinder into separate foils and slabs of molybdenum. Each foil was followed by a separate slab. Each foil had a thickness of about 0.01 inch (0.25 mm), while each slab had a thickness between about 0.75 inches and about 1.5 inches. The foils were used to determine the specific activity of Mo.sup.99 at different points within the aggregate thickness of the foils and slabs. In the target, the six foil/slab units were situated in series, with the slabs closer to the .gamma. beam source having the narrower widths. Each foil or slab was touching the adjacent slab or foil. A 2 inch diameter, 4.3 mm thick tungsten slab, used as a convertor plate, was located between the .gamma. beam source and the target. The convertor was also touching the first foil of the target. A 28 MeV electron beam, having a current of 1.84 microamperes (.mu.a) and a beam width of 1.5 cm, was directed substantially perpendicularly into the side of the convertor proximal to the electron beam source. A .gamma. beam was generated, substantially perpendicular to the distal side of the convertor. The .gamma. beam was directed into the target. The target was exposed for 4.6 hours to the generated .gamma. beam generated. Twenty-six hours after irradiation, the total activity of technetium-99 (Tc.sup.99), and the Giant-Resonance beam half-width, were then measured for each foil using a calibrated intrinsic-germanium crystal, by measuring the amount of .gamma.s having an energy specific to Tc.sup.99 decay (i.e., 140.1 keV) which were emitted at the center point of each foil, and by measuring the radial distance from the center of the foil over which the activity is reduced by one half to show beam spread. The results of center point activity measurements for the six sequential foils are provided in FIG. 9. As shown therein, the activity of Tc.sup.99 measured at the center point of the first foil, located at surface of the target (depth=0), was 30.3 microcurie (.mu.Ci). The center point activities for foils deeper in the target declined non-linearly as a function of their relative depths within the target. This demonstrates that the intensity of the photon flux in the Giant-Resonance energy range falls off quickly with distance in the target material. The half-width measurements for the six sequential foils are also provided in FIG. 9. The half-width of the first foil (depth=0) was 1.5 cm. The half-widths measured for foils deeper in the target showed some increase with depth, for example the half-width for a foil at a depth of about 6 cm was about 3.3 cm. These half-width measurements demonstrate that the .gamma. radiation beam, though spreading from scatter of .gamma.s within the target, remained sufficiently collimated to support the production of Mo.sup.99 throughout a cross-section of the target without a significant loss of .gamma. radiation energy from the target material. While this invention has been particularly shown and described with references to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the scope of the invention encompassed by the appended claims. For example, a target of palladium-104 was irradiated using the method described to make quantities of palladium-103, an isotope used in brachytherapy for prostate cancer. A target containing radium-226 was irradiated using the method described to produce radium-225, the parent isotope of actinium-225 and bismuth-213. Actinium-225 and bismuth-213 are medical isotopes used in clinical and pre-clinical trials for forms of leukemia, myeloma and solid mass tumors.
	claims	1. An ion implanter which emits an ion beam from an ion source and in which an ion implanting is performed to a substrate disposed in a process chamber, wherein the ion beam has a positive charge and a substantially rectangular cross section or a substantially long ellipsoidal cross section having a long side direction and a short side direction, the ion implanter comprising:a beam current measuring device that measures a beam current density distribution of the ion beam in the long side direction;a deflecting electrode that deflects at least a part of the ion beam in the long side direction toward the short side direction, based on a result measured by the beam current measuring device; anda shield member that partially shields the ion beam deflected by the deflecting electrode,wherein the deflecting electrode includes a plate electrode and an electrode group including a plurality of electrodes, the electrode group being disposed to face the plate electrode so as to interpose the ion beam between the plate electrode and the electrode group, the plate electrode being positioned along a long side direction that defines the cross section of the ion beam, and the electrode group being positioned along the long side direction of the ion beam, wherein the plurality of electrodes of the electrode group are commonly positioned on a same long side direction of the ion beam, opposite the long side direction where the plate electrode is positioned,the plate electrode is electrically grounded,the plurality of electrodes are electrically independent from each other, andeach of the plurality of electrodes is connected to an independent power source from other power sources to perform a potential setting. 2. The ion implanter according to claim 1, whereinthe ion beam in the long side direction has a longer size than the substrate, andthe substrate is transferred along the short side direction of the ion beam when the ion implanting is performed to the substrate. 3. The ion implanter according to claim 2, whereinwhen the result measured by the beam current measuring device does not satisfy with a desired value, the plurality of electrodes is set so that all electrodes has a negative potential. 4. The ion implanter according to claim 2, whereinwhen the result measured by the beam current measuring device is not a desired value, the plurality of electrodes configuring the electrode group is set so that some electrodes are at a negative potential and the remaining electrodes are at ground potential. 5. The ion implanter according to claim 2, whereinthe plurality of power sources are connected to a bias power source to collectively set potential of the plurality of power sources based on ground potential. 6. The ion implanter according to claim 5, whereinthe bias power source collectively sets the potential of the plurality of power sources at a negative potential. 7. The ion implanter according to claim 1, whereinwhen the result measured by the beam current measuring device does not satisfy with a desired value, the plurality of electrodes is set so that all electrodes has a negative potential. 8. The ion implanter according to claim 1, whereinwhen the result measured by the beam current measuring device is not a desired value, the plurality of electrodes configuring the electrode group is set so that some electrodes are at a negative potential and the remaining electrodes are at ground potential. 9. The ion implanter according to claim 1, whereinthe plurality of power sources are connected to a bias power source to collectively set potential of the plurality of power sources based on ground potential. 10. The ion implanter according to claim 9, whereinthe bias power source collectively sets the potential of the plurality of power sources at a negative potential.
055419690	description	DETAILED DESCRIPTION OF THE INVENTION FIG. 1 is a partial cross-sectional view of the midloop water level monitor 10 connected to the main pipe 12 of the hot water leg of a nuclear power plant. Main pipe 12 is the hot water leg. Chamber 18 is placed at substantially the same elevation as the main pipe 12. An upper connecting pipe 14 forms a fluid connection from the top region of pipe 12 to the upper region of tank 18. A lower connecting pipe 16 fluidly connects the bottom region of pipe 12 to the lower region of tank 18. A fluid 20, such as water, is shown partially filling the main pipe 12 and the chamber 18. A fluid level indicator 22 is located in chamber 18. In the preferred embodiment of the invention, the water level indicator 22 is comprised of closely spaced heated junction thermocouples (HJTCs) that indicate by temperature gradient the level of water in the chamber 18 and therefore in the pipe 12. The signal from the HJTCs is transmitted from the chamber to a remote location along line 24. HJTCs are preferred because they are fully qualified for operation at system pressures and temperatures during plant power production. In the preferred embodiment of the invention, the water level monitor 10 further comprises two isolation valves 26, 28 that can be used to selectively isolate the chamber 18 from the pipe 12. These selective isolation valves 26, 28 allow maintenance to be performed on chamber 18 while the power plant is running and pressurized hot water is running through pipe 12. It should be recognized that a plurality of valves may be used to isolate the chamber 18 from the pipe 12. It is recognized that chamber 18 may be located at any remote location from the pipe, as long as that location is on substantially the same elevation as the hot water pipe 12. Therefore, the chamber 18 can even be located in a completely separate room from the hot water pipe. The preferable location of the connecting upper pipe 14 and the lower connecting pipe 16 is onto the main pipe 12 at substantially the longitudinal middle of the hot water pipe. This allows the best indication of the water level in the pipe and will allow the monitoring of the water level so that work areas in the steam generator are not flooded during maintenance, and that loss of cooling at the core does not occur. FIG. 2 is a partial cross section of a variation of the water level monitor of FIG. 1 wherein the water level monitor is not directly connected to the hot water pipe 12, but is instead connected into secondary pipes 30 and 32 that connect to the hot water pipe 12. The lower connecting pipe 16 may be directly connected to any lower secondary pipe coming from the bottom of pipe 12, for example, the shut down cooling line. Similarly, the upper connecting pipe may be connected to any secondary upper pipe 32 that connects to the top of the main pipe 12, for example, the surge line. The secondary and connecting pipes are of sufficient size to ensure equalization of the air pressure between the chamber and the hot leg pipe, and therefore the water levels in the chamber and the hot leg pipe. Again, it is recognized that tank 18 must be at substantially the same elevation as pipe 12. The result is that a fluid 20, such as water, in the hot water main pipe 12 will be at the same fluid level in the pipe and in the chamber. Isolation valves 26 and 28 may isolate the chamber from the rest of the hot water system comprising the main pipe and any secondary or connecting pipes, such as pipe 32. FIG. 3 is a partial view of a nuclear power plant. The midloop water level monitor 10 comprising the tank 18 and upper and lower connecting pipes 14, 16 is connected substantially midway on the hot water pipe 12. The hot water pipe 12 connects the reactor vessel 40 to steam generator 42. Water is heated in the core within vessel 40, travels through the hot pipe 12 to the steam generator 42 where steam is generated in the secondary water loop (not shown). The water, thus cooled, then leaves the steam generator 42 by line 44, travels through a pump (not shown) and returns to the reactor core 40 by inlet 46. This path of the water comprises a hot water loop. It is necessary during down times of the plant operation to perform maintenance inside the steam generator 42, particularly in the lower head 45. Because of the orientation of the reactor 40 and the steam generator 42, the hot water pipe 12 is in a substantially horizontal position at an elevation such that the upper region of pipe 12 aligns with a portion of head 45. The result is that even small changes in the level of water in the pipe 12 can result in the flooding of the bottom of the heat exchanger 42 if the level is too high, or insufficient water in the core to perform critical cooling functions if the water level is too low. In addition, if the level of water in the hot leg falls too low, the shut down cooling suction pipe (not shown) will form a vortex of cooling water while drawing the water from the hot leg pipe. This vortex reduces cooling flow of the water. As a result, the core appears to be adequately covered with water, but the reduced flow because of the vortex allows a core overheating condition to occur. Because of this critical cooling function, it is important that operators in the control room 48 be able to monitor the level of water in the pipe even during shut down periods. The signal indicative of the water level from the midloop water level monitor is transmitted along line 24 to control panel 50 where the information is displayed. This allows an operator in the control room 48 to completely monitor all information about the hot water loop system even during a shut down period. The signal can also be transmitted to a computer for integration into a plant safety monitoring and display system.
	description	This application is a national stage application under 35 U.S.C. 371 of PCT Application No. PCT/CN2013/071485 filed Feb. 7, 2013, which claims the benefit of China Application No. 201210108701.8 filed Feb. 7, 2013, the entire disclosure of which is incorporated herein by reference. The present invention relates to a high polymeric fiber. China is a big power but not a strong power in respect of chemical fiber. According to experts' introduction and statistics of component authorities, China's output of high-performance special fibers is only one percent of the world's output. Currently, three major high-performance fibers in the world are carbon fiber, aramid fiber and high molecular weight polymeric fiber. Carbon fiber is still in an experimental and initial production phase, and a product thereof can only be applied to a field such as wearable fillers. Over 70 percent of high molecular weight polymeric fiber in European and American developed countries is applied to military fields such as body armor, bullet-proof helmet, anti-bullet armor of military facilities and equipment, and aerospace. Development of high-performance fiber exhibits comprehensive power of a country, and high-performance fiber is an important material basis for building a modernized strong country. To this end, it is particularly urgent and desirable to expedite production and application of high-performance special fibers in China from the perspective of state interests. An object of the present invention is to provide boron carbide high polymeric fiber as a novel dedicated material. A new material boron carbide high polymeric fiber is inventive after long-term development and repeated experiments. The material is fabricated from the following raw materials: boron carbide, high polymeric ethylene emulsion, hydrochloric acid, antioxidant and catalyst. The boron carbide high polymeric fiber according to the present invention may be used in fields such as firearms manufacture, maritime rescue, fire protection and fire fight, anti-bullet and anti-explosion armor, biochemical nuclear industry treatment, and may be extensively applied to civil field, aerospace, military fairs and national defense. The present invention exhibits performances such as extremely good resistance against high temperature and low temperature, super anti-acid and anti-base performance, excellent extensibility, wear resistance and anti-impact capability and the like, and may be used in a long period of time at a range of temperature −170-2100° C. under a standard atmospheric pressure. The material has a super anti-acid and anti-base performance, and the material per se will not substantively change if it is immersed consecutive 30 days in caustic soda solution at a normal temperature and under a normal pressure or immersed consecutive 30 days in hydrochloric acid solution; the material has excellent performances such as extensibility, wear resistance, anti-impact capability and resistance against ultraviolet ray, and its fiber strength is over ten times that of steel wire with the same cross section. The fiber has a low density and can float on water; the material has a low breaking elongation and has a very strong energy-absorbing capability, and therefore has a prominent anti-impact capability. The material can resist against ultraviolet radiation and prevent neutron and γ rays; the material has a low dielectric constant and high electromagnetic wave transmissivity. Therefore, the present invention may be extensively applied to civil use, nuclear industry, aerospace, military affairs and national defense. The resultant boron carbide high polymeric fiber fills in a gap of the same class of products in China. The material is recyclable and pollution-free. The present invention provides a novel material boron carbide high polymeric fiber fabricated from the following parts of raw materials by weight: 50-60 parts of boron carbide, 150-193 parts of high polymeric ethylene emulsion with a mass concentration 40%-50%, 116 parts of hydrochloric acid with a mass concentration 37%, 3-5 parts of antioxidant, and 7 parts of catalyst. The boron carbide is one boron carbide selected from a group consisting of w14, w20, w28 and w40; The high polymeric ethylene emulsion is selected from polyethylene emulsion HA-soft80, GFN and 85-CDIT; The antioxidant is any one selected from compounds among antioxidants such as diphenylamine, p-diphenylamine and dihydroquinoline, and derivatives or polymers thereof. The catalyst is one selected from trifluoromethanesulfonic anhydrides as catalysts. The present invention further provides a method of fabricating a novel material boron carbide high polymeric fiber from the above parts of raw materials by weight, comprising the following steps: (1) immersing boron carbide in hydrochloric acid for 90-120 minutes to remove impurities; (2) placing the cleaned and dried boron carbide in a 2600° C. high-temperature high-pressure furnace for melting; placing the resultant molten boron carbide in a high temperature-resistant spinning furnace to produce boron carbide precursors; (3) adding a catalyst in a high polymeric ethylene emulsion containing an antioxidant, placing the boron carbide precursors therein, and producing inorganic high molecular polymer, namely, a preliminary product of boron carbide high polymeric fiber, under 15-30 atmospheric pressures. The preliminary product is subjected to surface treatment through calcium hydroxide solution, and then cleaned and dried to finally obtain the boron carbide high polymeric fiber according to the present invention.
050900374	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, a gantry 20, such as may be used in a "third generation" computed tomography (CT) scanner, includes an x-ray source 10 collimated by collimator 38 to project a fan beam of x-rays 22 through imaged object 12 to detector array 14. The x-ray source 10 and detector array 14 rotate on the gantry 20 about center of rotation 13. The rotation of the gantry 20, as indicated by arrow 28 is within a gantry plane 60, aligned with the x-y plane of a Cartesian coordinate system. The imaged object 12 rests on table 17 which is radio-translucent so as not to interfere with the imaging process. Table 17 may be controlled so that its upper surface translates along the z axis perpendicular to the x-y imaging plane, moving the imaged object 12 across the gantry plane 60. The detector array 14 is comprised of a number of detector elements 16, organized within the gantry plane 60, which together detect the projected image produced by the attenuated transmission of x-rays through the imaged object 12. The fan beam 22 emanates from a focal point 26 in the x-ray source 10 and is directed along a fan beam axis 23 centered within the fan beam 22. The fan beam angle, measured along the broad face of the fan beam, is larger than the angle subtended by the imaged object 12 so that two peripheral beams 24 of the fan beam 22 are transmitted past the body without substantial attenuation. These peripheral beams 24 are received by peripheral detector elements 18 within the detector array 14. Referring to FIG. 6, the x-ray source 10 includes an anode 29 position within an evacuated glass envelope and rotated about anode shaft 25 for heat dispersion. A stream of electrons from a cathode (not shown) is accelerated against the face of the anode 29 to produce the x-ray beam 19. The face of the anode 29 is beveled with respect to the fan beam axis 23 so that radial displacement of the electron beam by focussing plates, (not shown) as is known in the art, will produce a z-axis displacement of the focal point 26. The amount of this displacement may be controlled by x-ray controller 62. Referring to FIG. 2, the angular position .theta. of the gantry 20 along the z-axis with respect to the imaged object 12 is shown by arrows 11. The z-axis position of the imaged object 12 with respect to the imaging plane 60 changes constantly during the acquisition of each tomographic projection set. Accordingly, arrows 11 are shifted along a helix within the imaged object 12 along the z-axis. The pitch of the helix will be referred to as the scanning pitch. The z-axis distance from the center 9 of the slice being acquired to the volume elements 7 intercepting the fan beam 22 is termed the "helix offset" of that volume element. In the present invention the fan beam axis 23 may be shifted along the z-axis during the helical scan to reduce the helix offset as will be described. Referring to FIG. 3, uncollimated x-rays 19 radiating from the focal point 26 in the x-ray source 10 (not shown in FIG. 3) are formed into a coarse fan beam 21 by primary aperture 40. As is understood in the art, the uncollimated x-rays 19 are produced by a high voltage x-ray tube typically including a rotating anode (not shown) receiving a high energy beam of electrons and re-emitting x-ray radiation. The coarse fan beam 21 is collimated into fan beam 22 by means of collimator 38. Referring generally to FIGS. 3, 4(a) and 4(b), collimator 38 is comprised of a cylindrical x-ray absorbing molybdenum mandrel 39 held within the coarse fan beam 21 on bearings 42 allowing the mandrel 39 to rotate along its axis. A plurality of tapered slots 41 are cut through the mandrel's diameter and extend along the length of the mandrel 39. The slots 41 are cut at varying angles about the mandrel's axis to permit rotation of the mandrel 39 to bring one such slot 41 into alignment with the coarse fan beam 21 so as to permit the passage of some rays of the coarse fan beam 21 through the slot 41 to form fan beam 22. Referring to FIG. 4(a) and 4(b), the tapered slots 41 are of varying width and hence the rotation of the mandrel 39 allows the width of the fan beam 22 to be varied between narrow (1 mm) as shown in FIG. 4(b) and wide (10 mm) as shown in FIG. 4(b). The slots 41 ensure dimensional accuracy and repeatability of the fan beam 22. The slots 41 are tapered so that the entrance aperture 43 of each slot 41, when orientated with respect to the coarse fan beam 21, is wider than the exit aperture 45. The exit aperture 45 defines the width of the fan beam 22 and the extra width of the entrance aperture 43 prevents either edge of the entrance aperture 43 from blocking the coarse fan beam 21 during small angular rotation of the mandrel 39. Such small rotations of the mandrel 39 are used to provide adjustment of the z-axis position of the fan beam 22 as will be discused in detail below. Referring again to FIG. 3, a positioning motor 48 is connected to one end of the mandrel 39 by flexible coupling 50. The other end of the mandrel 39 is attached to a position encoder 46 which allows accurate positioning of the mandrel by motor 48. Fan beam angle shutters 44 at either ends of the mandrel 39 control the fan beam angle. Referring now to FIG. 5, the control system of a CT scanner, suitable for use with the present invention, has gantry associated control modules 60 which include: x-ray controller 62 which provides power and timing signals to the x-ray source 10, and which in certain embodiments of the invention, controls the position of the focal point 26; collimator controller 64 which controls the rotation of the collimator 38; gantry motor controller 66 which controls the rotational speed and position of the gantry 20; and the data acquisition system 68 which receives projection data from the detector array 14 and converts the data to digital words for later computer processing. The gantry associated control modules 60 communicate with the x-ray tube 10, collimator 38 and detector 14 via slip rings 61. It will be recognized that direct cabling using a take up reel may be substituted for the slip rings 61 for a limited gantry rotation system. The x-ray controller 62, the collimator controller 64 and the gantry motor 66 controller are connected to a computer 70. The computer 70 is a general purpose minicomputer such as the Data General Eclipse MV/7800C and may be programmed to synchronize the rotation of the gantry 20 with the position of the fan beam 22 per the present invention as will be described in detail below. The data acquisition system 68 is connected to image reconstructor 72 which receives sampled and digitized signals from the detector array 14 via the data acquisition system 68 to perform high speed image reconstruction according to methods known in the art. The image reconstructor 72 may be an array processor such as is manufactured by Star Technologies of Virginia. The speed and position of table 17 along the z-axis is communicated to and controlled by computer 70 through of table motor controller 74. The computer 70 receives commands and scanning parameters via operator console 76 which is generally a CRT display and keyboard which allows an operator to enter parameters for the scan and to display the reconstructed image and other information from the computer 70. A mass storage device 78 provides a means for storing operating programs for the CT imaging system, as well as image data for future reference by the operator. Referring now to FIG. 6, the z-axis position of the exit aperture 45 of the collimator 38 may be adjusted so that the fan beam 22, as indicated by fan beam axis 23, diverges from the gantry plane 60 in the z-axis dimension during the acquisition of the first projection of a projection set. The amount of divergence of the fan beam axis 23 from the gantry plane 60 is such that a volume element 7 at position 80 within a slice and moving toward the gantry axis 60 with motion of table 17, is intersected by the fan beam axis 23. The position of the table 17 during the acquisition of the projection set is determined by the table motor controller 74. The collimator 38 as controlled by the collimator controller 64 is coordinated by computer 70 with the position of table 17 so that during the movement of the table 17 and imaged object 12, the fan beam axis 23 is swept as to constantly intercept volume element 7. As the projections of each projection set are acquired, during a period T.sub.1, the imaged object 12 is translated along the z-axis with respect to the imaging plane 60 so that volume element ultimately moves to position 82 at the last projection of the projection set. Typically, the amount of translation will be equal to the slice thickness w. At the completion of the acquisition of the projection set, the exit aperture 45 of the collimator 38 is returned to the position it had at the start of the projection set, moving in the opposite direction, during a period T.sub.2, so that the fan beam axis 23 intercepts a new volume element in a new slice. The new volume element has the same relative position 80 with respect to the gantry plane 60 as did volume element 7 at the start of the acquisition of the previous projection set. Preferably, positions 80 and 82 are located symmetrically about the gantry plane 60 so as to reduce the maximum deviation of the fan beam axis 23 from the gantry plane 60 during any acquisition. At the halfway point in the acquisition of the projection set, the focal point 26, the center line of the exit aperture 45 of the collimator 38, fan beam axis 23 and the center of illumination of the detector array 14 will be perfectly aligned with the gantry plane 60. At all other times, these various points may deviate from the gantry plane 60. The measures of the deviation of the center line of the exit aperture 45 of the collimator 38, the volume element intersected by fan beam axis 23, and the center of illumination of the detector array 14 from the gantry plane will be termed C.sub.z, V.sub.z, and D.sub.z respectively. For the first described embodiment shown in FIG. 6, F.sub.z, the position of the focal point 26 with respect to the gantry plane 60 is constant and zero. Referring to FIG. 10, during the first period T.sub.1 of the acquisition of a projection set, the displacement of the collimator C.sub.z will increase so that the fan beam axis 23 tracks the movement of the volume element 7. For large values of 1.sub.2 and 1.sub.3 and small values of slice thickness w, the relationship between the collimator displacement C.sub.z and the displacement V.sub.z of the fan beam axis 23 with axis of translation 84 of the volume element 7 is: ##EQU1## where l.sub.1 is the distance between the focal point 26 and the exit aperture 45 of the collimator 38 and l.sub.2 is the distance between the exit aperture 45 and the translation axis 84 of the volume element 7. Accordingly, during the first period T.sub.1, the position of table 17 as determined via the table motor controller 74, determines the position of the exit aperture 45 after suitable scaling by computer 70 as given in equation (1) above. During a second time period T.sub.2, immediately after the first time period T.sub.1, the exit aperture 45 is returned to the position it had at the start of that acquisition of projections to prepare for acquisition of a second projection set. Preferably this period T.sub.2 is made a short as possible by moving the collimator 38 at its maximum speed. During this return period T.sub.z, no projection data is taken and the x-ray fan beam 22 may be decreased in intensity according to any of several methods known in the art such as decreasing current flow to the x-ray tube or shuttering the x-ray beam 19. It will be noted that the displacement D.sub.z of the fan beam axis 23 with respect to the surface of the detector array 14 will be larger than the displacement V.sub.z according to the following ratio: ##EQU2## where l.sub.3 is the distance between the axis of translation 84 of the volume element 7 and the exposed surface of detector array 14. Generally, the detector elements 16 of detector array 14 exhibit a change of sensitivity as a function of the z-axis position of their illumination. Hence a variation in D.sub.z will introduce some variation into the projections measurements. This variation may be corrected by using the peripheral beams 24 and peripheral detector elements 18 to provide a reference for correcting sensitivity variation according to compensation methods understood in the art. One such method is given in U.S. Pat. No. 4,559,639 hereby incorporated by reference. In a second embodiment shown in FIG. 7 and 11, both the x-ray focal point 26 and the exit aperture 45 of the collimator 38 are moved. Movement of the x-ray focal point 26 is accomplished by refocussing the electron beam on the anode 29 as has been previously described or by physical translation of the x-ray source 10 under the control of servo motors. The measure of the deviation of the focal point 26 from the gantry plane 60 will be termed: F.sub.z. Referring to FIG. 11, in this second embodiment, the intersection D.sub.z of the fan beam axis 23 on the detector array 14 is maintained constant (at zero displacement) by controlling the displacement F.sub.z of the focal point and the displacement C.sub.z of the exit aperture 45 with respect to the displacement V.sub.z of the volume element as follows: ##EQU3## Referring to FIG. 14, the acquisition volume 86 within the imaged object 12 over which projection data is acquired in a non-helical scan will be approximately one half of acquisition volume of a helical scan: assuming that the scanning pitch times the rotation for one projection set is approximately equal to the slice thickness w. The present invention, as described in the above two embodiments, enlarges the acquisition volume over the non-helical acquisition volume 86 by flanking volumes 88 which are outwardly conically concave. This increase in acquisition volume represented by volumes 88 increases the helix offset of the projection data slightly but much less than that produced by helical scanning which adds areas 90 to effectively double the acquisition volume. In general, the greater the distance l.sub.1 +l.sub.2 in comparison to the radius of the image object 12 about the translation axis 84, the less the flanking volume 88 and thus the less the helix offset of the data. Referring to FIGS. 14 and 8, a third embodiment of the invention eliminates the flanking volumes 88 and produces an acquisition volume 86 identical to that of non-helical scanning. Referring to FIG. 12, the displacement D.sub.z of the collimator 38 and F.sub.z of the focal point 26 are set equal to the displacement V.sub.z of the volume element 7. The fan beam axis 23 is thus maintained parallel to the gantry plane 60 at all times. In a fourth embodiment, shown in FIGS. 9 and 13, the displacement C.sub.z of the exit aperture 45 of the collimator 38 is fixed (and equal to zero) and the displacement F.sub.z of the focal point 26 is adjusted according to the following relationship: ##EQU4## The acquisition volume (not shown) for this method and the amount of displacement D.sub.z of the fan beam axis 23 on the detector array 14 will be greater than the comparable quantities for the previously described method, for CT systems of similar dimensions as a result of the greater angular divergence of the fan beam axis 23 from the gantry plane 60 necessary to track a given volume element 7 without movement of the collimator 38. For each of the above embodiments, the projection data for volume elements near volume element 7 on the translation axis 84, there will be little helix offset. To the contrary, the volume elements removed from volume element 7 and the translation axis 84 will have increasing amounts of helix offset for greater values of x and y as dictated by the angle of the fan beam axis 23 with respect to the gantry plane 60. For this reason, it may be desirable to position the volume element 7 and the translation axis 84 near internal structures of interest within imaged object 12. The translation axis 84 will normally intersect the center of rotation 13 of the gantry 20. The center of rotation 13 and the translation axis 84 may both be moved within the imaged object simply by adjusting the height of table 17. Alternatively, the translation axis 84 may be moved independently from the center of rotation 13 by adjusting the fan beam angle 23 as a function of gantry rotation 28. This is most easily accomplished by modifying the apparent value of l.sub.2 and l.sub.3 used by computer 70 in the above embodiments as a function of gantry angle .theta. as follows: EQU l.sub.2 '=l.sub.2 -cos(.theta.+.alpha.) (.DELTA.) (6) EQU l.sub.3 '=l.sub.3 +cos(.theta.+.alpha.) (.DELTA.) (7) where .alpha. is the angle with respect to the center of rotation 13 between the volume of interest and gantry angle .theta.=O, .DELTA. is the distance between the volume of interest and the center of gantry rotation 13, and l.sub.2 ' and l.sub.3 ' are substituted into the above equations in place of l.sub.2 and l.sub.3 respectively. For the embodiments shown in FIGS. 6, 7, and 9, it will be understood that the amount of helix offset, reduced as it is, also varies as a function of the order of the projection within the projection set. For example, when the starting and ending positions 80 and 82 of the volume of interest 7 are symmetrically displaced about the gantry plane 60, the centermost projections will have no helix offset and the starting and ending projections will have the most helix offset. For this reason, it is desirable to weight the projections so as to deemphasize the starting and ending projections and to emphasize the centermost projections of the projection set. Such weighting systems are disclosed in co-pending application 07/440,531 entitled:"Method for Reducing Patient Translation Artifacts in Tomographic Imaging" filed Nov. 22, 1989. Finally, for the first, second, and forth embodiments, where the center of illumination of the detector 14 changes during the acquisition of projections, it is important that the detector 14 be sufficiently wide so as to always receive the entire fan beam 22. Many modifications and variations of the preferred embodiment which will still be within the spirit and scope of the invention will be apparent to those with ordinary skill in the art. For example, the collimator may be of a conventional bladed design. Further it will be apparent that this method is applicable to so called "forth generation" CT machines where the detector array 14 is stationary and may surround the imaged object 12. Clearly the the x-ray tube and collimator may be also mechanically translated and tipped as a single unit. Finally, the table motion need not be constant during the acquisition of successive projection sets but may be slowed, for example, during the period T.sub.2 when the fan beam 22 repositions itself at a starting position. In order to apprise the public of the various embodiments that may fall within the scope of the invention, the following claims are made:
	abstract	A sealing mechanism for a reactor vessel (RV) cable penetration tube improves the functional and structural integrity of a cable inserted in an RV through a penetration tube due to use of a precise thimble. The sealing mechanism includes a penetration tube configured to penetrate an RV from an outside to an inside thereof and having a penetration hole for communication with the inside of the RV, a cable configured to be inserted in the RV through the penetration hole of the penetration tube, and a thimble placed between the cable and the penetration tube, wherein a dimple groove portion is provided on the thimble in a direction from an outer surface of the penetration hole toward the cable.
050248022	claims	1. In a nuclear power plant, a method of controlling the steam generator water level, wherein the steam generator has an upper level tap corresponding to an upper level, a lower level tap corresponding to a lower level, a riser positioned between said lower and upper taps, and level sensor means for indicating water level between a first range limit and a second range limit, said sensor means being connected to at least said lower tap, comprising: calculating a measure of velocity head at about the lower level tap; calculating a measure of full water level as the upper level less said measure of velocity head; calibrating said level sensor means to provide an output at said first limit corresponding to an input thereto representative of said measure of full level; calculating a high level setpoint equal to the level of said riser less a bias amount which is a function of the position of said riser relative to the span between said taps; and controlling said water level when said sensor means indicates that said high level setpoint has been reached. determining a first measure of velocity head corresponding to water at said upper level; determining a second measure of velocity head corresponding to water level at said threshold level; calibrating said level sensing means to provide a maximum indication of water level at said upper level adjusted by said first measure, and to provide a minimum indication of water level corresponding to water level at said lower level; and determining said high level control threshold corresponding to water level at said threshold level less said second measure. positioning an upper tap on said steam generator above said riser and obtaining a reference level pressure corresponding to the water read at said upper tap; positioning a lower tap below said riser and in a high velocity region of said steam generator and obtaining a lower tap pressure therefrom, said riser thereby being positioned at a given percentage of the span between said lower and upper taps; providing a differential pressure sensor having an output range between lower and upper limits; calculating a first differential pressure between said reference level pressure and said lower tap pressure when steam water is at said lower tap, and calibrating said sensor to indicate said upper limit when said first difference pressure is placed thereacross; calculating a second differential pressure between said reference level pressure and a net pressure calculated for water at said upper tap less velocity head at maximum power plant power, and calibrating said sensor to indicate said lower limit when said second differential pressure is placed thereacross; calculating a differential high level setpoint pressure as the difference between reference pressure and the net pressure calculated as pressure at the lower tap corresponding to water level at said riser less an amount equal to the velocity head at maximum velocity times span percentage; and controlling water level as a function of the sensed differential pressure compared to high level setpoint differential pressure. 2. The method of claim 1, wherein said sensor means senses differential pressure and has inputs connected to said lower and upper taps respectively, and comprising calculating the differential pressure for water level at said lower tap and calibrating said level sensor means to provide an output at said second limit corresponding to water level at said lower tap. 3. The method of claim 2, wherein said measure of velocity head is velocity head at maximum power. 4. The method of claim 3, comprising calculating said bias amount as maximum velocity head less riser span percentage times velocity head at full power. 5. A method of monitoring water level relative to a high level control threshold in a steam generator, said steam generator being operated with a water level between a lower level and an upper level, said threshold being between said lower and upper levels, and having pressure sensing means subject to water velocity error for measuring said water level, comprising 6. A method for monitoring the water level in a steam generator of a nuclear power plant so as to be able to control said water level relative to the steam generator riser, comprising the following steps:
	description	The present invention relates to a nuclear power plant having highest ranked safety features, which can be decommissioned with certainty, easily and in a very inexpensive way at the end of its service life. More in detail, the invention concerns a plan of a nuclear power system, with its radioactive components and relevant facilities for the treatment and disposal of radioactive wastes that are located in a safe manner altogether underground, being covered by an adequate thickness of rocks. This plan permits an absolute environment and population protection together with a safe and easy decommissioning. Conventional power plants for the production of electric energy from controlled-fission nuclear reactions are today designed and constructed in such a way that all the main works which said power plants are comprised of, in particular the nuclear isle, the stream generators, the turbine/electric generator isle and the safeguard buildings, are located above the soil surface. Safety and security of these power plants that are considered as very dangerous targets, particularly after the Chernobyl power-plant disaster and the terrorist attacks to the Twin Towers of New York, are based on various systems. In particular, their safety is obtained by using redundant structural solutions (for both plant engineering and building works), separating circuits for heat exchange and cooling, constructing with pre-assembling piping, adopting high constructive standards, protecting and multiplying safeguard buildings and applying extremely rigorous procedures for the management of plants and personnel. However it is evident that such solutions are absolutely inefficient to protect a nuclear power plant against the most probable attacks, i.e. a launch of large aircraft filled with fuel or explosive, or even of rockets, towards the nuclear isle or pools containing spent nuclear fuel. These are scenarios that have become very realistic in the last years. It is evident that events of this kind would generate what military experts cell “dirty bombs”, whose effects, even if not so destroying like those of a nuclear bomb, would be anyway so dangerous to pollute with following fall out large areas, up to thousands of square kilometers, with a result of making these areas inhabitable for centuries. Several solutions have been proposed by skilled in the art from the 60's. Even if such solutions were not specifically designed to withstand extreme terrorist attacks that were not foreseeable in the past, they aimed to reach high safety levels against accidental releases of radioactive materials, by locating a nuclear reactor in underground caverns that were excavated for this purpose. However, the proposed solutions do not allow sufficiently advantageous effects to be achieved so that their adoption can be suggested, on the one hand for the safety, on the other for the overall economic cost. A solution for an underground nuclear power plant is disclosed for example in the patent RU No. 2.273.901. The solution proposed by the present invention fits in this context, which provides an embodiment of a nuclear power plant able to assure: a strong reduction of the cost needed for decommissioning the nuclear power plant by applying innovative procedures, the decommissioning being certain at the end of its life service; an absolute protection of population and environment outside the nuclear power plant from radioactive releases due to either accidental causes or any terrorist attacks or catastrophic natural events; a final supersafe storage of low-intermediate level radioactive wastes that are produced during the operation of the nuclear power plant, by avoiding the handing in the soil surface of such radioactive wastes towards main treatment/conditioning/disposal centres; an interim, insuperably safe storage of spent fuel, of high-level long-life radioactive materials and, if necessary, of spare rods for refuelling the nuclear reactor; a consequent better relationship with the population, and then a higher certainty of the initial investments. These and other results are obtained according to this invention by: a) applying techniques known in mining that allow underground caverns having considerable sizes to be excavated, and assure both the nuclear isle and safeguard buildings and control buildings of a nuclear power plant to be housed into said caverns, by exploiting their absolute capability to prevent radiation and radioactive releases of any kind toward the external environment due to both plant malfunctions or damages caused by terrorist attacks, in virtue of underground rocks having suitable thickness (hundreds of meters); b) constructing specific facilities for characterising, treating and disposing the radioactive wastes beneath the surface of the ground; c) adopting suitable procedures for a facilitated decommissioning of the nuclear isle at the end of the service life of the nuclear power plant. Therefore, a specific object of the present invention is an underground nuclear power system in which the nuclear isle of one or more nuclear power plants are installed in caverns, and further, side by side with them, a centre for characterising, treating and conditioning radioactive wastes and two repositories are installed in suitable caverns, with a final repository being adapted to store low-intermediate level nuclear wastes and a temporary repository being adapted to store spent fuel, high-level long-life radioactive materials and, in case, spare nuclear rods for reactor refueling. Preferably, according to the invention, a facilitated decommissioning occurs at the end of the service life of the nuclear power plant by sealing radioactive components of the nuclear isle (after closing mechanical openings, disconnecting commands/control systems and after removing nuclear fuel, liquid of primary circuit, and, if necessary, those plant parts that are contaminated by high activity radionuclides), by filling up voids of the cavern hosting the reactor and steam generators with concrete, also of a lightened type, and by closing the entrance of the cavern with metallic doors of adequate thickness and any interposition of walls made of injected concrete between said doors. Further, according to the invention, an entrance to underground facilities works is made absolutely secure from both terrestrial terrorist attacks and other attacks performed by rockets, aircraft and similar weapons. Furthermore, according to the invention, the entrance is made so that the underground facilities works cannot be flooded by extreme natural events. Yet, according to the invention, the radiation containment systems and the safeguard buildings protection are realised by the rocks of the caverns appositely modelled during the excavation in order to host said components. Moreover, according to the invention, any kind of radioactive wastes of low-intermediate level produced during the service life of the nuclear power plant is stored in a final way inside the underground plant in safe conditions, reducing/avoiding the transport of such wastes outside. Moreover, all high level nuclear material is temporarily stored in the same site, and if the site is proved as suitable, it will become a final repository also for the high level nuclear material. Further according to the invention, nuclear fuel supplies can be constituted, thereby reducing the total number of transports for the nuclear power reactor refuelling. Moreover, according to the invention, the occupation of an external area is extremely limited. Yet, according to the invention, masses of natural water can be used for cooling. Further, according to the invention, an access to the nuclear power plant is preferably of sub-horizontal type. Further, according to the invention, a system for characterising, conditioning and disposing radioactive wastes is provided. Always according to the invention, commercial, high power PWR reactors can be hosted in said nuclear power plant. Yet, according to the invention, the underground caverns housing the nuclear isle and the underground caverns storing the radioactive wastes and materials have a roof, an inverted arch and side-walls made impermeable, and will be provided with systems for collecting natural or accidentally released fluids. Finally, according to the invention, if the underground nuclear plant is realised under a pre-existing conventional nuclear power-station (of superficial type) to be dismantled, the pre-existing conventional nuclear power-station can be decommissioned by transferring the radioactive wastes so produced into underground repositories of the new underground power plant, thereby avoiding any danger of nuclear pollution to the surrounding environment. In FIG. 1 of the enclosed drawings, a layout of a modern EPR nuclear power plant according to the known technique is shown, which comprises a reactor building 1, a fuel building 2, safeguard buildings 3, diesel emergency generator buildings 4, an auxiliary nuclear building 5, a nuclear waste building 6, and a turbine building 7. Turning now to FIG. 2, in a plant according to the invention, all the radioactive components, the relevant safeguard buildings and the emergency generators are provided inside underground caverns excavated for this purpose, having adequate dimensions and depth, and being connected to the surface by means of equipped access inclines and/or vertical shafts. Thus all the nuclear components can be confined with respect to the external world. In this manner, the following components of a PWR nuclear power-plant are situated inside the underground caverns: reactor/steam generator/pressurizer unit; safeguard buildings; control room; emergency generators. Substantially a very reduced volume of buildings remains outside. Further, with the same principle, at marginal costs, it is possible to construct caverns usable, depending on the necessities and the situation of a Country, as a final repository of radioactive wastes of low-intermediate level, as a temporary repository of spent fuel and for storing any spare nuclear fuel; as a final repository of high level (and/or long life) radioactive wastes. Every underground room will be isolated and equipped with suitable control systems for maintaining pressure differential needed by the activities to be performed therein. Air extracted for maintaining said pressure differential will be treated by suitable filtering systems. An optimal diagram is that one shown on the left part in FIG. 2, with the entrance being made in a hill slope so that steam generator and turbines can be located at the same level. In any case, similar results can be obtained (see right part of FIG. 2), with the reactor being placed at a lower level with respect to the entrance; in this case mayor adjustments should be made to a secondary circuit, in particular if the turbines were located at a higher level than the steam generator. This solution could be adopted for all those cases in which a site hosting an old nuclear power plant to be decommissioned should be re-used for a new nuclear power plant. In this case, the above solution would allow the external area “to be cleaned up” from the radioactive materials by simply moving beneath the surface of the ground, into repository caverns, the radioactive wastes that are produced by decommissioning the old external nuclear power plant. A comparative analysis, which is made from a new French-German nuclear power plant (1,600 equivalent MW EPR), shows that moving beneath the surface of the ground the nuclear isle of a conventional nuclear power plant according to the solution proposed by the invention, does not imply additional costs with respect to a conventional nuclear plant fully constructed superficially. In fact, the excavation cost (cavern and access tunnels), including material and personnel handling systems is about 150 million of Euro. The figure is very close to, if not even lower than, the overall amount requested, in the case of the new EPR, for: a reactor basement necessary to contain materials melted by a hypothetical fusion of a reactor core; two (inner and outer) shelters for containing the radiation, both being 1.3 m thick; part of the external safeguard buildings, and in particular of the necessary civil works, as in the EPR configuration, four safeguard buildings are located around the reactor in a star-shaped configuration so to reduce the risk to be altogether destroyed by a mono-directional aerial incursion. Tasks today assigned to these components in a classic EPR would be very well accomplished super-safely by the rock cover above the underground caverns where the excavations designed to host the radioactive parts of the nuclear power plant and the safeguard buildings according to the present invention should be conducted. As far as the facilitated decommissioning of the nuclear isle, it will be obtained: i.1) by constructing the roof, the inverted arch base and the side-walls of the cavern for reactor and steam generator in such a way to make them impermeable and installing an appropriate system for collecting fluids (which can be both natural or coming from possible accidental releases); i.2) by applying (after removing, at the end of the service life of the nuclear power plant, the spent fuel, the circulation liquid of the primary circuit and, if necessary, the parts contaminated with high-activity long-life radionuclides, and after sealing the mechanical openings and disconnecting the command and operating control systems) a spritz beton lining (or a universal primer for metal/mortars of cement) on the components to be sealed; i.3) by installing, in the relevant points of the cavern and the components of reactor and steam generator, sensors for monitoring temperature, humidity, and radioactivity; i.4) by filling up voids of the cavern for reactor and steam generator with concrete injection (possibly of expanded/lightened type); i.5) by hermetically closing the entrance to the cavern for reactor and steam generator; i.6) by continuously monitoring the system by the sensors in i.4). The safety against external attacks and natural catastrophic events will be obtained, as already said, by moving to a place beneath the surface of the ground the nuclear isle, the security systems and the emergency generators and, moreover, by: ii.1) anti-intrusion devices located at the entrance of the underground nuclear power plant; ii.2) anti-flooding devices; ii.3) moving to a place beneath the surface of the ground, where possible, also partially, the building hosting the turbines/electrical generators; ii.4) using, where possible, natural water masses (sea, river . . . ) for the cooling, instead of air towers; ii.5) devices for maintaining rooms and caverns depressurised and for treating/filtering the air so extracted. The characterisation-localisation-licensing studies will be developed with the following progression: iii.1) initial localisation, according to the existing rules, of a site for hosting a nuclear power plant, a final low-intermediate level radioactive waste repository (300 years lasting) and a temporary repository (50 years lasting) for spent fuel, high level wastes and any spare rods for reactor refueling; iii.2) studies and tests, to be carried out during a part of the 50-60 years of service life of the nuclear power plant, in order to verify the suitability of the temporary underground cavern to house, in a final manner, high-level long-life nuclear wastes (for 50,000 years). A nuclear power plant so designed can use, with relatively simple modifications feasible in short time, the reactors already licensed and now existing in the market, of both small and high power. Differing from other similar proposals, the caverns will have, as much as possible, sub-horizontal access (in hill slopes) in order to avoid any loss of charge in the circuits and, if realised under the terrain surface in plain areas, will have systems for connecting the underground facilities to the surface through inclines and/or service shaft. As far as the economic-financial model of the investment necessary for constructing a plant of the kind here proposed, the solution according to the invention offers the opportunity of taking into account from the beginning the times and costs for the decommissioning. Moreover, these costs can be assumed as substantially negligible (5-10% less than the costs of a conventional decommissioning) and the decommissioning will be performed certainly. For a conventional, superficially nuclear power plant, its decommissioning strongly depends on the availability of a radioactive waste repository, on the necessity of completely dismantling the nuclear power plant and returning the hosting area to green field condition, on the distance between the nuclear power plant and the radioactive waste repository, and, above all, on the uncertainty of the times required for obtaining the relevant construction authorisation. All these factors make a correct a priori evaluation of the costs highly improbable. Other relevant savings are offered by the possibility of using at marginal costs, during the construction of the nuclear power plant and the management of the nuclear waste and/or radioactive material repositories, the same handling systems to connect the cavern with outside, that are already arranged for the nuclear isle. Therefore, the solution here proposed according to the invention permits the realisation of a supersafe and easily-/certainly decommissionable nuclear power plant, so that the recourse of the nuclear energy is made again acceptable by populations, from the proposal of installing nuclear reactors beneath the surface of the ground, as already proposed in the years 60-70. The plant according to the invention provides the installation of the nuclear isle of one or more nuclear reactors in caverns, and further the installation, side by side with them, of a centre for characterising, treating and conditioning radioactive wastes and two repositories in suitable caverns (of which a final one is for low-intermediate level nuclear wastes and a temporary one is for the spent fuel and high-level long-life radioactive materials). In particular, this allows an extremely simplified decommissioning immediately at the end of the service life of the nuclear power plant and the realisation of inviolable accesses to the underground facilities. In this manner: (i) the costs of decommissioning of the nuclear power plants (usually between 30 and 60% of the cost of construction) are drastically reduced; (ii) both the decommissioning at the end of the service life of the nuclear power plant and the final disposal of the low-intermediate level wastes become certain; (iii) the number of fuel transports for refueling the reactor is reduced; (iv) the handling of the radioactive wastes on the soil surface towards the characterisation-treatment-conditioning-disposal centres is avoided and, above all, the incomparable capacity of natural protection offered by the rocks and the possibility of making inviolable accesses to the underground works can be exploited; (v) nuclear releases towards the external environment, due to both malfunctions of the plants or leakage caused by (aerial or terrestrial) terrorist attacks or catastrophic natural events, are completely avoided; (vi) additional costs are avoided, the costs for excavations and works for transferring underground the nuclear isle being compensated by the saving obtained since external protection works should not be constructed (shelters, sacrificial basements, redundancies necessary for the physical protection and for the plural safeguards buildings, minor soil occupation . . . ), (vii) there are other savings because the characterisation-treatment-conditioning centre and the two radioactive material repositories can be constructed at marginal costs, since they could use the handling systems with outside and the systems for physical protection already implemented for the nuclear power plant in cavern, (viii) the nuclear power plant can be constructed with extreme simplicity because it is possible to take advantage from the high level of technology reached in mining, so that the components of the main allowable nuclear power plants (from the French-German EPR to the Westinghouse and Russian WER power plant) can find (with marginal modifications) an easy arrangement in caverns of adequate dimensions; modifications that, consequently, can be put into effect in very short times especially if compared to those required from other plans of new generation nuclear reactors that are developed at the moment, (ix) it is possible, during the service life of a nuclear power plant according to the present patent application (that would demand a surveying similar to that necessary for localisation of a final low-intermediate radioactive waste repository in order to be authorised), to have all the time to carry out procedures able to verify any suitability of the site to host finally, and not only, also high-level long-life radioactive wastes. However, such radioactive wastes would be hosted for 50-60 years in the best manners and in safety conditions, that are obviously higher than those offered by a conventional nuclear power plant, which is constructed on the surface. The present invention has been described in an illustrative and not limiting way, according to its preferred embodiments. It should be understood that variations and/or modifications could be made by skilled in the art without departing from the scope of the invention as defined in the enclosed claims.
	summary
054897378	description	DETAILED DESCRIPTION OF THE EMBODIMENTS Embodiment 1 One of the embodiments of the present invention is explained referring to FIGS. 1 and 2. The present embodiment is a process for solidifying three kinds of waste such as used ion exchange resin (waste resin), incinerated ashes, and concentrated waste liquid, all of which are generated from a boiling water reactor nuclear power plant, with cement, and radioactivity of the obtained package is controlled so as not to exceed a statutory determined upper limit. The statutory determined upper limit of radioactive concentration used in the present embodiment is as follows; TABLE 1 ______________________________________ Nuclides Upper limit concentration ______________________________________ H-3 3.1 .times. 10.sup.11 (Bq/t) C-14 8.5 .times. 10.sup.9 (Bq/t) Co-60 2.8 .times. 10.sup.12 (Bq/t) Ni-59 8,9 .times. 10.sup.9 (Bq/t) Ni-63 1.1 .times. 10.sup.12 (Bq/t) Sr-90 1.7 .times. 10.sup.10 (Bq/t) Nb-94 8.5 .times. 10.sup.7 (Bq/t) Tc-99 1.9 .times. 10.sup.7 (Bq/t) I-129 2.8 .times. 10.sup.5 (Bq/t) Cs-137 1.0 .times. 10.sup.11 (Bq/t) .alpha. nuclide 5.6 .times. 10.sup.8 (Bq/t) ______________________________________ Waste resin powder, which was generated mainly from clean up systems of nuclear reactors and stored in the sludge tank 1 having storing capacity of 100 m.sup.3, was transported to the adjusting tank 4 of capacity 5 m.sup.3 by the suction pump 3 through the valve 2 in a condition of 5% slurry. Subsequently, a part of the waste resin (concentration is 5%), about 600 kg, in the adjusting tank 4 was transported to the dehydrater 5, water content was decreased to 70% by centrifugal dehydration, and the waste resin in a cake-like condition was sent to the radioactivity measuring barrel 7 by the screw feeder 6. Semiconductor detectors 8 were so installed in the radioactivity measuring barrel 7 as to measure concentration of Co-60 and Cs-137 in the waste resin. Result of the measurement revealed that concentration of respective Co-60 and Cs-137 was 1.times.10.sup.15 Bq/t and 2.times.10.sup.9 Bq/t, and concentration of the Co-60 was higher by three orders than the statutory upper limit concentration shown in Table 1. If the waste resin was solidified without any further treatment, the obtained package could not be disposed in a land disposal. Accordingly, solidifying processing was discontinued, and the waste resin in the adjusting tank 4 and the radioactivity measuring barrel 7 was returned to the sludge tank 1 through the by-path line 9. Next, processing of waste resin stored in another tank was intended. Waste resin particles, which were mainly generated from condensate clean up systems and stored in the resin tank 10 having storing capacity of 50 m.sup.3, were transferred to the adjusting tank 4 by the suction pump 12 through the valve 11 in a condition of 20% slurry. Subsequently, a part of the waste resin (concentration is 20%), about 250 kg, in the adjusting tank 4 was transferred to the dehydrater 5, water content was decreased to 50% by centrifugal dehydration, and the waste resin was introduced to the radioactivity measuring barrel 7 by the screw feeder 6. Results of the measurement on concentration of Co-60 and Cs-137 revealed that concentration of respective Co-60 and Cs-137 were 2.times.10.sup.9 Bq/t and 4.times.10.sup.6 Bq/t, and the concentration both of the Co-60 and Cs-137 were remarkably lower than the statutory upper limit concentration shown in Table 1. Subsequently, the maximum value of radioactivity of other nuclides was estimated by so-called scaling factor (SF) method. Concretely saying, values for Ni-59, Ni-63, and Nb-94, all of which were corrosion products, were obtained by multiplying the concentration of Co-60 with SF values, such as 1.times.10.sup.8 Bq/t for Ni-59, 2.times.10.sup.10 Bq/t for Ni-63, and 4.times.10.sup.6 Bq/t for Nb-94 at the maximum. And, values for Sr-90, 1-129, and .alpha. nuclides, all of which were fission products, were obtained by multiplying the concentration of Cs-137 with SF values, such as 3.times.10.sup.7 Bq/t for Sr-90, 3.times.10.sup.1 Bq/t for 1-129, and 3.times.10.sup.6 Bq/t for .alpha. nuclides at the maximum. And, as values for H-3 and C-14, respectively 1.times.10 Bq/t and 7.times.10 were known from actual data obtained by so-called mean value method. As the result of evaluation on concentration of nuclides other than Co-60 and Cs-137 based on the observed values for Co-60 and Cs-137 in the manner as above described, it was revealed that each concentration of all nuclides in the specimen was lower than the upper limit values shown in Table 1. However, in the evaluation of concentration by SF method, safety margin of 100 times was added to the observed data on SF values. In accordance with the above described result, it was decided to supply the above described data to the computer 13 and to execute the solidifying process because it was assumed that the obtained package would be acceptable for disposal in land when the waste resin was processed for solidification. The waste resin (weight 100 kg., moisture content 50% in the radioactivity measuring barrel 7 was transmitted to the solidifying vessel 15 by the screw feeder 14, and 120 kg. of solidifying agents (fiber reinforced cement disclosed in JP appl. No. 1-221502 (1989) was used in the present embodiment) which was mainly composed from cement and 60 kg. of kneading water including 1 kg. of water reducing agent were supplied from the cement silo 16 and the water tank 17 respectively. The solidifying vessel 15 was furnished with the agitator 18, which was used for preparation of cement paste by agitation, and the package was obtained. At that time, data on waste resin, solidifying agents, and kneading water, all of which were supplied into the solidifying vessel 15, were put in the computer 13. The package obtained by the above described method was sufficient in mechanical properties such as strength, and had lower radioactive concentration for each nuclide than the upper limit concentration for each nuclide shown in Table 1. Embodiment 2 Next, processing procedure on data supplied to the computer 13 is explained referring to FIG. 2. To the computer 13, initial conditions (such as ID number, processing date, sort of objective waste for solidification, and background of generation etc.) relating to the solidifying process are previously supplied. Subsequently, radioactive concentration of each nuclide in the objective waste is recorded based on measured result at the radioactivity measuring barrel 7. At that time, data on evaluating method of the concentration are also stored. That is, data recording that Co-60 and Cs-137 are directly measured, Ni-59, Ni-63, Nb-94, Sr-90, I-129, and a nuclides are evaluated by SF method, and H-3 and C-14 are evaluated by mean value method in the present embodiment are stored. Next, data on the solidifying method of the waste are stored. That means, sort and quantity of waste solidifying agents and kneading water used in the operation, operating condition of the solidifying apparatus, and specification of the solidifying vessel used in the operation are recorded. In accordance with the above described data, physical properties of the finally obtained package can be evaluated by the following procedure. First, radioactive concentration of each nuclide can be easily calculated from data on the concentration of the waste itself and on composition of the package (weight percentage of waste/solidifying agents/kneading water). Mechanical properties of the package (such as unconfined compression strength, specific gravity, porosity etc.) can be evaluated from the data on composition of the package and the operating condition of the solidifying apparatus, and are recorded depending on necessity. And, data on finally obtained package such as weight and surface dose are preferably recorded and stored. The above described data are arranged by a predetermined format, and the package itself is recorded and labelled with the first input ID number by stamping etc. By the above described procedure, it will become possible to grasp exact history including radioactive concentration of the package when, for example, the package will be transported for disposal in land in future. Embodiment 3 Next, a case of processing on concentrated waste liquid (CWL), of which main composition is Na.sub.2 SO.sub.4, is explained referring to FIG. 1. About 100 kg of CWL, of which concentration is about 25%, stored in the CWL tank 19 having storing capacity of 150 m.sup.3 was transmitted to the radioactivity measuring barrel 7 by the pump 21 through the valve 20. At the radioactivity measuring barrel 7, concentration of Co-60 and Cs-137, both of which are gamma nuclides and concentration is easily determined by non contact measurement, were measured by the semiconductor detector 8. The result were 1.times.10.sup.11 Bq/t for Co-60 and 4.times.10.sup.5 Bq/t for Cs-137, and it was revealed that both of the above described concentration were less than the upper limit shown in Table 1. Subsequently, concentration of nuclides other than Co-60 and Cs-137 were evaluated by SF method. Concentration of Ni-59, Ni-63, and Nb-94 were obtained by multiplying SF values to the concentration of Co-60 as 5.times.10.sup.9 Bq/t for Ni-59, 1.times.10.sup.12 Bq/t for Ni-63, and 2.times.10.sup.8 Bq/t for Nb-94 at the maximum, and it was revealed that concentration of Ni-59 and Ni-63 had a possibility to exceed the upper limit shown in Table 1. (Respective concentration of Sr-90, 1-129, and .alpha. nuclides was evaluated by SF method based on the concentration of Cs-137, and respective concentration of H-3 and C-14 was evaluated by mean value method, but all of the above described concentration were far lower than the upper limit shown in Table 1). Subsequently, 100 ml of CWL was taken as a sample through the sampling port furnished to the radioactivity measuring barrel 7, and concentration of Ni-59 and Ni-63, both of which are .beta. nuclides, were measured by a liquid scintillator. As a result, it was revealed that concentration of Ni-59 was 7.times.10.sup.7 Bq/t and of Ni-63 was 3.times.10.sup.10 Bq/t, both of which were lower than the upper limit shown in Table 1 by two orders. The reason that the observed value becomes smaller than the value estimated by SF method is based on safety margin of about 100 times in the value estimated by SF method. As radioactive concentration of the CWL was confirmed to be less than the upper limit shown in Table 1 as above described manner, the CWL (concentration 25%, 100 kg) was transmitted to the solidifying vessel 15 and was solidified by supplying 300 kg of solidifying agents from the cement silo 16 and kneading by the agitator 18. The obtained package was confirmed to be disposable in land, for example, in both aspects of radioactive concentration and mechanical properties. Subsequently, as same as the previously described package of waste resin, various data on the package was recorded and filed to the computer 13, and the package was so labelled with the ID number as to correspond to the record and stored in a waste storage facility. Embodiment 4 Next, a case of processing for incinerated ashes obtained by incineration of burnable miscellaneous solid bodies is explained referring to FIG. 1 again. About 100 kg of incinerated ashes stored in a drum was transferred to the radioactivity measuring barrel 7. As radioactive concentration of the incinerated ashes was previously estimated to be relatively high, small amount of specimen was taken as a measuring sample from the sampling port 22 furnished to the radioactivity measuring barrel 7 and measured concentration of 11 nuclides shown in Table 1 respectively. As a result, concentration of three nuclides, Ni-59, Sr-90, and .alpha. nuclides were found to exceed the upper limit shown in Table 1. Concretely saying, their concentration were 5.times.10.sup.10 Bq/t for Ni-59, 2.times.10.sup.11 Bq/t for Sr-90, and 6.times.10.sup.8 bq/t for a nuclides. On the other hand, calculation on solidification with cement revealed that final radioactive concentration would exceed the upper limit shown in Table 1 if 100 kg of the incinerated ashes at maximum per a 200 liters drum is solidified. In the present embodiment, the above calculation was performed by the computer 13 based on the above described result of measurement on radioactive concentration. Concurrently, calculation by the computer 13 revealed that final radioactive concentration of the package would be lower than the upper limit shown in Table 1 if filling amount of the incinerated ashes per a 200 liters drum was reduced to 20 kg. Consequently, it was decided that 100 kg of the incinerated ashes in the radioactivity measuring barrel 7 would be divided into five portions, 20 kg per a portion, and five packages would be prepared (although 100 kg of the incinerated ashes could be solidified physically in a body, concentration of the radioactivity would exceed the upper limit shown in Table 1 as previously described). After 20 kg of the incinerated ashes in the radioactivity measuring barrel 7 was transferred to the solidifying vessel 15 by the screw feeder 14, 250 kg of solidifying agents from the cement silo 16 and 130 kg of kneading water from the water tank 17 were supplied to the solidifying vessel 15, and a package was prepared by operation of the agitator 18. Radioactive concentration of the package obtained in the manner as above described could be calculated from the composition of the package (incinerated ashes/cement/water=5%/62.5%/32.5%) and radioactive concentration of the incinerated ashes(especially important nuclides Ni-59 was 5.times.10.sup.10 Bq/t, Sr-90 was 2.times.10.sup.11 Bq/t, and a nuclides was 6.times.10.sup.8 Bq/t) as 3.times.10.sup.9 Bq/t for Ni-59, 1.times.10.sup.10 Bq/t for Sr-90, and 3.times.10.sup.7 Bq/t for .alpha. nuclides at maximum, and it was revealed that the concentration of all of the above described nuclides were less than the upper limit shown in Table 1. Accordingly, packages managing data having the same content as the one shown in FIG. 2 were prepared, and the package was labelled with the ID number and stored in a storage facility. In accordance with the above described method, not only the radioactive concentration of each package and processing history of the waste can be grasped exactly, but also such problems as difficulty in disposal of the packages on account of high radioactivity when the packages will be intended to dispose in land, for example, in future can be previously resolved. Embodiment 5 Another concrete embodiment of the present invention is explained referring to FIG. 3 hereinafter. Although the present invention is applicable to general radioactive waste generated from nuclear facilities, radioactive waste resin generated from nuclear power plants is taken as an example in the present embodiment. At a nuclear power plant, ion exchange resin is used in a reactor clean up system and a condensate clean up system in order to keep properties of reactor water stable. As for waste resin generated from the condensate clean up system, solidifying process has been adopted in practical use. However, waste resin generated from the reactor clean up system, so-called an intermediate level radioactive resin, has been stored in tanks at site of the power station. The present embodiment relates to processing for the stored intermediate level radioactive resin and low level radioactive resin generated from the condensate clean up system, which has an experience to be solidified in practical use. Radioactive waste resin generated from nuclear power plant is stored in the waste resin storing tank 23 having average storing capacity of 300 m.sup.3. The radioactive waste resin (used ion exchange resin) stored in the waste resin storing tank 23 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, a small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and transuranium elements (TRU), is measured. Depending on the result of the measurement, all amount of waste resin containing Co-60 and Cs-137 of which total radioactivity exceeds a predetermined level, waste resin containing the eleven nuclides of which radioactivity exceeds a predetermined level, and waste resin containing any of the eleven nuclides of which radioactivity exceeds a predetermined level, are returned to the waste resin storing tank 23 from the adjusting barrel 24 by the returning pump 33 through the returning path 34. All amount of the waste resin of which radioactivity does not exceed the predetermined level in the adjusting barrel 24 is transferred to the dehydrator 26, and excess water is removed. Subsequently, the waste resin is introduced to the receiving tank 27. The receiving tank 27 is furnished with the sampling port 25 for radioactivity measurement, and a small amount of the waste resin is taken out from the receiving tank 27 through the sampling port 25. Radioactivities of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, 1-129, Nb-94, C-14, H-3,and TRU in the sample are measured. Depending on the result of the measurement, all amount of waste resin containing Co-60 and Cs-137 of which total radioactivity exceeds a predetermined level, waste resin containing the eleven nuclides of which radioactivity exceeds a predetermined level, and waste resin containing any of the eleven nuclides of which radioactivity exceeds a predetermined level, can be returned to the waste resin storing tank 23 from the receiving tank 27. The measurement of the radioactivity is performed at least at any one of the adjusting barrel 24 and the receiving tank 27. The waste resin of which radioactivity does not exceed the predetermined level is transferred to the kneading barrel 31, and is processed for solidification. For the solidification, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 31. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent are introduced into the kneading barrel 31 from the additives tank 30, and are sufficiently mixed and kneaded. When filling amount of the resin is increased, the solidification can be preferably performed by preliminary kneading with cement and subsequent main kneading after a few hours or a few days from the preliminary kneading. After the kneading, the mixture is rapidly poured into solidifying vessel 32 (a drum or a PIC vessel). After curing, the completely package is transferred to the package inspection apparatus 35, and weight and surface dose of the package are measured. The obtained data are compared with the data before the solidifying process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 60 kg/drum at the maximum. Embodiment 6 Next, another embodiment of the present invention is explained referring to FIG. 4. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive waste resin generated from nuclear power plants is taken as an example in the present embodiment. At a nuclear power plant, ion exchange resin is used in a reactor clean up system and a condensate clean up system in order to keep properties of the reactor water stable. As for waste resin generated from the condensate clean up system, solidification processing has been adopted in practical use. However, waste resin generated from the reactor clean up system, so-called an intermediate level radioactive resin, has been stored in tanks at site of the power station. The present embodiment relates to processing, especially to pellet solidification, for the storing intermediate level radioactive resin and low level radioactive resin generated from the condensate clean up system, which has experience to be solidified in practical use. Radioactive waste resin generated from the nuclear power plant is stored in the waste resin storing tank 23 having average storing capacity of 300 m.sup.3. The radioactive waste resin (used ion exchange resin) stored in the waste resin storing tank 23 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, a small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and transuranium elements (TRU), is measured. Depending on the result of the measurement, all amount of waste resin containing Co-60 and Cs-137 of which total radioactivity exceeds a predetermined level, waste resin containing the eleven nuclides of which radioactivity exceeds a predetermined level, and waste resin containing any of the eleven nuclides of which radioactivity exceeds a predetermined level, are returned to the waste resin storing tank 23 from the adjusting barrel 24 by the returning pump 33 through the returning path 34. All amount of the waste resin of which radioactivity does not exceed the predetermined level in the adjusting barrel 24 is introduced into the thin film dryer 36 and pulverized. Subsequently, the waste resin is transferred to the receiving tank 27, and after sufficiently mixed with a small amount of pelletizing binder, the waste resin is introduced to the pelletizer 38 and is pelletized. The pelletized waste resin is rapidly transferred to the solidifying vessel 32 (a drum or a PIC vessel). As for solidifying agent, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 39. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent are introduced into the kneading barrel 39 from the additives tank 30 , and are sufficiently mixed and kneaded. After the kneading, the mixture is poured into the solidifying vessel 32 (a drum or a PIC vessel) where the pellets are placed. After curing, the completely package is transferred to the package inspection apparatus 35, and weight and surface dose of the package are measured. The obtained data are compared with the data before solidification process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 120 kg/drum at the maximum. Embodiment 7 Farther, another embodiment of the present invention is explained referring to FIG. 5. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive incinerated ashes generated from nuclear power plants are taken as examples in the present embodiment. At a nuclear power plant, burnable waste is incinerated for volume reduction in order to decrease amount of generated waste from the nuclear power plant. The present embodiment deals with the above described incinerated ashes as for objectives, especially for pellet solidification of the ashes. Radioactive incinerated ashes stored in the incinerated ashes storing tank 41 are transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, small amounts of incinerated ashes are taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. The obtained data are stored. After the measurement, the ashes are transferred to the receiving tank 43. When the observed radioactivity is lower than a predetermined level, the ashes are introduced to solidifying processing system. On the other hand, when the observed radioactivity is higher than the predetermined level, adjustments to decrease filling amount of the ashes in a package or to transmit a predetermined amount of ashes to the receiving tank 43 from the low level incinerated ashes tank 42. The ashes passed the radioactivity measurement or adjusted in the manner as above described are introduced into the pelletizer 38 and pelletized. The pelletized incinerated ashes are rapidly transferred into the solidifying vessel 32 (a drum or a PIC vessel). As for solidifying agent, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 39. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent etc. are introduced into the kneading barrel 39 from the additives tank 30 , and are sufficiently mixed and kneaded. After the kneading, the mixture is poured into the solidifying vessel 32 (a drum or a PIC vessel) where the pellets are placed. After curing, the completely package is transferred to the package inspection apparatus 35, and weight, surface dose, and radioactivity of the package is measured. The obtained data are compared with the data before solidifying process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 200 kg/drum at the maximum. Embodiment 8 Further, another embodiment of the present invention is explained referring to FIG. 6. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive incinerated ashes generated from nuclear power plants are taken as examples in the present embodiment. At a nuclear power plant, burnable waste is incinerated for volume reduction in order to decrease amount of generated waste from the nuclear power plant. The present embodiment deals with the above described incinerated ashes as for objectives, especially for homogeneous solidification method (direct solidification method) of the ashes. Radioactive incinerated ashes stored in the incinerated ashes storing tank 41 are transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, small amounts of incinerated ashes are taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. The obtained data are stored. After the measurement, the ashes are transferred to the receiving tank 43. When the observed radioactivity is lower than a predetermined level, the ashes are introduced into solidifying processing system. On the other hand, when the observed radioactivity is higher than the predetermined level, adjustments to decrease filling amount of the ashes in a package or to transmit a predetermined amount of ashes to the receiving tank 43 from the low level incinerated ashes tank 42. After agitating and mixing the ashes homogeneously in the receiving tank 43, the mixture was poured into the kneading barrel 39. As for solidifying agent, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 39. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent etc. are introduced into the kneading barrel 39 from the additives tank 30, and are sufficiently mixed and kneaded with the incinerated ashes. After the kneading, the mixture is poured into the solidifying vessel 32 (a drum or a PIC vessel). After curing, the completely package is transferred to the package inspection apparatus 35, and weight, surface dose, and radioactivity of the package are measured. The obtained data are compared with the data before solidification process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 100 kg/drum at the maximum. Embodiment 9 Next, another embodiment of the present invention is explained referring to FIG. 7. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive waste resin generated from nuclear power plants is taken as an example in the present embodiment. At a nuclear power plant, ion exchange resin is used in a reactor clean up system and a condensate clean up system in order to keep properties of the reactor water stable. As for waste resin generated from the condensate clean up system, solidifying process has been adopted in practical use. However, waste resin generated from the reactor clean up system, so-called an intermediate level radioactive resin, has been stored in tanks at site of the power station. The present embodiment explains processing method for the stored intermediate level radioactive resin and low level radioactive resin generated from the condensate clean up system, which has experience to be solidified in practical use, especially for mixed solidification of the intermediate level resin and the low level resin. Among radioactive waste resin generated from the nuclear power plant, the waste resin having relatively high level radioactivity is stored in the waste resin storing tank 23 having average storing capacity of 300 m.sup.3. The radioactive waste resin (used ion exchange resin) stored in the waste resin storing tank 23 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, a small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Subsequently, total amount of the resin in the adjusting barrel 24 is transferred to the dehydrator 26 and excessive water is removed. After the dehydration, the waste resin is transferred to the receiving tank 27. The receiving tank 27 is furnished with the sampling port 25 for measurement of radioactivity, and through the sampling port 25, a small amount of waste resin is again taken out from the receiving tank 27 as a sample. And each radioactivity of the eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Depending on the result of the measurement, all amount of waste resin containing Co-60 and Cs-137 of which total radioactivity exceeds a predetermined level, waste resin containing the eleven nuclides of which radioactivity exceeds a predetermined level, and waste resin containing any of the eleven nuclides of which radioactivity exceeds a predetermined level, can be returned to the waste resin storing tank 23 from the receiving tank 27. Measurement of the radioactivity is performed at least either of the adjusting barrel 24 or the receiving tank 27. On the other hand, waste resin having relatively low radioactivity generated from the condensate clean up system is stored in the waste resin storing tank 48, and is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25 for measurement of radioactivity. Through the sampling port 25, a small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Subsequently, the waste resin is transferred to the receiving tank 27. The receiving tank 27 is furnished with the sampling port 25 for measurement of radioactivity, and through the sampling port 25, a small amount of waste resin is again taken out from the receiving tank 27 as a sample. And each radioactivity of the eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Depending on the result of the measurement, all amount of waste resin containing Co-60 and Cs-137 of which total radioactivity exceeds a predetermined level, waste resin containing the eleven nuclides of which radioactivity exceeds a predetermined level, and waste resin containing any of the eleven nuclides of which radioactivity exceeds a predetermined level, can be returned to the relatively high level waste resin storing tank 23 from the receiving tank 27. Measurement of the radioactivity is performed at least either of the adjusting barrel 24 or the receiving tank 27. The obtained data are stored. Subsequently, each of the waste resins are introduced into the kneading barrel 31 respectively, and processed for solidification. As solidifying agents, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 31. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent are introduced into the kneading barrel 31 from the additives tank 30 , and are sufficiently mixed and kneaded. When filling amount of the waste resin is increased, solidification is preferably performed by preliminary kneading with cement and main kneading at a few hours or a few days after the preliminary kneading. After the kneading, the mixture is rapidly introduced into the solidifying vessel 32 (a drum or a PIC vessel). After curing, the completely package is transferred to the package inspection apparatus 35, and weight and surface dose of the package are measured. The obtained data are compared with the data before solidification process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 60 kg/drum at the maximum. In the present embodiment, low level waste in the waste resin tank 48 can be other than the waste resin, for example, incinerated ashes and condensed waste liquid. Embodiment 10 Next, another embodiment of the present invention is explained referring to FIG. 8. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive waste resin generated from nuclear power plants is taken as an example in the present embodiment. Especially, the present embodiment explains mixed solidification of intermediate level radioactive waste resin and low level radioactive waste resin. Among radioactive waste resin generated from the nuclear power plant, the waste resin having relatively high level radioactivity is stored in the waste resin storing tank 23 having average storing capacity of 300 m.sup.3. The radioactive waste resin (used ion exchange resin) stored in the waste resin storing tank 23 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. A small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Subsequently, total amount of the resin in the adjusting barrel 24 is transferred to the receiving tank 27. On the other hand, the low level waste resin stored in the waste resin storing tank 48 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. A small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured. Subsequently, total amount of the resin in the adjusting barrel 24 is transferred to the receiving tank 43. The obtained measurement data are stored. Subsequently, the both waste resin are introduced into the mixing barrel 44, and mixed to be sufficiently homogeneous. After the mixing, the mixture is introduced into the thin film dryer 36, and dried and pulverized. The dried and pulverized mixture is introduced into the pelletizer 38 through the powder receiving tank 37, and pelletized. The pelletized waste resin is rapidly transferred to the solidifying vessel 32 (a drum or a PIC vessel). As for solidifying agents, cement from the solidifying agent tank 28 and a predetermined amount of kneading water from the water tank 29 are introduced into the kneading barrel 39. Depending on necessity, carbon, metal fiber, absorbing agent, fluidizing agent, and water reducing agent are introduced into the kneading barrel 39 from the additives tank 30, and are sufficiently mixed and kneaded. After the kneading, the mixture is rapidly poured into the solidifying vessel 32 (a drum or a PIC vessel) in which the pellets are placed. After curing, the completely package is transferred to the package inspection apparatus 35, and weight and surface dose of the package are measured. The obtained data are compared with the data before the solidifying process, and both data are stored. The package through the inspection apparatus 35 becomes an objective for transportation or intermediate storage. When the intermediate storage is elected, the procedure with the package inspection apparatus 35 is not necessarily performed. Filling rate of the radioactive waste resin package prepared in the manner as above described reached to 120 kg/drum at the maximum. In the present embodiment, low level waste in the waste resin tank 38 can be other than the waste resin such as incinerated ashes and condensed waste liquid. Embodiment 11 Next, another embodiment of the present invention is explained referring to FIG. 9. The present embodiment is also applicable to general radioactive waste generated from nuclear facilities, but radioactive waste resin generated from nuclear power plants is taken as an example in the present embodiment. Especially, the present embodiment explains preliminary processing system before solidifying processing system. Among radioactive waste resin generated from the nuclear power plant, the waste resin having relatively high level radioactivity is stored in the waste resin storing tank 23 having average storing capacity of 300 m.sup.3. The radioactive waste resin (used ion exchange resin) stored in the waste resin storing tank 23 is transferred to the adjusting barrel (adjusting tank) 24 having capacity of from a few cubic meters to tens cubic meters. The adjusting barrel 24 is furnished with the sampling port 25, and a small amount of waste resin is taken out from the adjusting barrel 24, and each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured, and the obtained measurement data are stored. Subsequently, total amount of the resin in the adjusting barrel 24 is introduced into the dehydrator 26, and excessive water is removed. Next, the waste resin is transferred to the dissolution tank 45, and radioactive nuclides and the resin are separated by dissolution. The solution including the radioactive nuclides is dried and pulverized by the thin film drier 36, and transferred to the powder receiving tank 37. The powder receiving tank 37 is furnished with the sampling port 25, and a small amount of the powder is taken out as a sample from the powder receiving tank 37. And each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured, and the obtained measurement data are stored as a record. On the other hand, the processed waste resin by the dissolution is introduced into the incinerator 46, and incinerated. The incinerated ashes are introduced into the incinerated ashes receiving tank 47, which is furnished with the sampling port 25, and a small amount of the incinerated ashes is taken from the incinerated ashes receiving tank 47 as a sample. And each radioactivity of eleven nuclides such as Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and TRU, is measured, and the obtained measurement data are stored as a record. Both of the powder in the powder receiving tank 37 and the incinerated ashes in the incinerated ashes receiving tank 47 are transmitted into the incinerated ashes storing tank 41 together, and are supplied to the solidifying processing system.
	description	The present invention resides in the surprising discovery that secondary hydriding in nuclear reactor fuel rods can be sigificantly mitigated, and in some instances substantially eliminated, by providing in the interior of the fuel rod an effective amount of one or more metal oxides. The oxides may be those of iron, nickel, tin, bismuth, copper, colbalt, chromium, manganese and/or combinations of such oxides. Specific examples of suitable metal oxides are iron oxides (Fe3O4; Fe2O3), nickel oxide (NiO), tin oxide (SnO2), copper oxide (CuO) and bismuth oxide (Bi2O3). The invention finds particular application to uranium oxide fuel contained within zirconium-alloy based cladding. Such fuel rods are commonly employed in LWRs. When steam enters the interior of the fuel cladding, an oxidation reaction occurs with the fuel/cladding which results in the generation of hydrogen. This may be described generally as follows: xH2O+Zr=ZrOx+xH2 xe2x80x83xH2O+UO2=UO2+x+xH2 The hydrogen:steam ratio within the rod in the region near the metal oxide will be dictated by the thermodynamic equilibrium for the Metal/Metal Oxide (M/MOx) couple, evaluated at the temperature within the fuel rod where the metal oxide is located. If the hydrogen fraction rises above the equilibrium condition for the M/MOx couple, back reaction between the hydrogen and MOx will generate steam and maintain the interior at the posited equilibrium. It is to be noted that in some instances the equilibrium could correspond to a couple such as MOz/MOx where MOz is a lower oxide, that is z less than x, and not pure metal. The back reaction is therefore described as follows: MOx+H2=M (or MOz, z less than x)+H2O Provided the ratio of steam to hydrogen in equilibrium with the metal oxide is such that the steam fraction is above the threshold level for secondary hydriding, secondary hydriding will be mitigated. Since the steam generated by the back reaction between the hydrogen and metal oxide can easily diffuse over a certain length, secondary hydriding can be mitigated even if the metal oxide is present only at discrete intervals. The presence of metal oxide may be in accordance with several possible embodiments. In a first embodiment, the metal oxide may be present as a coating on the cladding interior surface. The metal oxide may be selected from iron oxides (Fe3O4; Fe2O3), nickel oxide (NiO), tin oxide (SnO2), copper oxide (CuO) and bismuth oxide (Bi2O3). Bismuth oxide is generally employed as it has a lower cross-section for absorption of neutrons than the other oxides. Generally, the coating is applied to a thickness in the range of 1 mil (25 microns) or less, for example 0.25-0.5 mil. As a second embodiment, the metal oxide may be present as a coating on the fuel pellet surfaces. The metal oxide may be selected from iron oxides (Fe3O4; Fe2O3), nickel oxide (NiO), tin oxide (SnO2), copper oxide (CuO) and bismuth oxide (Bi2O3). Bismuth oxide is generally employed as it has a lower cross-section for absorption of neutrons than the other oxides. Generally, the coating is applied to a thickness in the range of 1 mil (25 microns) or less, for example 0.25-0.5 mil. As a third embodiment, the metal oxide may be present as individual pellets or as wafers between fuel pellets, or at the bottom of the fuel stack or at the top of the fuel stack or combinations thereof. Generally, the individual pellets or wafers will be of nearly the same geometry (diameter) as the pellet, possibly a little larger. In the instance where they are present between the fuel pellets, the pellet or wafer thickness will depend upon the number of pellets or wafers used. The pellets or wafers are generally fabricated by sintering the metal oxide powder selected from iron oxides (Fe3O4; Fe2O3), nickel oxide (NiO), tin oxide (SnO2), copper oxide (CuO) and bismuth oxide (Bi2O3). As a fourth embodiment, reference is made to the accompanying FIG. 1 showing a container 2 with perforations or slits 4 in the wall thereof which provide free access to the surrounding gases. The container is typically fabricated of a material that does not react with the metal oxide, such as stainless steel. The container wall has a thickness of 10 mils or less and an outside diameter which is essentially the same as fuel pellets, or slightly larger. The metal oxide may be present in the container 2 as a powder or pellet, as described above. In a fifth embodiment, the metal oxide may be discretely distributed (rather than in a continuous manner) along the fuel rod. The metal oxide may be in any of the configurations described in the first through fourth embodiments above. In a sixth embodiment, referring to FIG. 2, there is shown a fuel rod 6 comprising an outer cladding 8 and a fuel pellet stack 10. A container 2 as described above, is provided at the bottom of and retained in place by bottom end cap 12 and the fuel stack 10 containing metal oxide. FIG. 2 illustrates the situation where the container is at the bottom of the fuel stack. However, a similar container may also be placed at the top of the fuel stack. In the usual arrangement, a container is placed at the bottom of the fuel stack, and a further container may optionally be present at the top of the stack. When a container is at the top of the fuel stack, there is a plenum and a retainer spring (not shown) which presses down on the container to hold it in place. The specific metal oxide to be used for secondary hydriding mitigation may be selected from the oxides of Ni, Fe, Sn, Bi, Cu, Co, Cr, and Mn. The metal oxide is typically present in each fuel rod in an amount of up to about 15 grams, more usually up to about 12 grams, for example 2 to 10 grams. The specific metal oxide to be chosen is to be based on whether the metal oxide reacts with hydrogen rapidly enough. The rapidity of this reaction must be such that the rate is sufticiently fast so that it can counteract the rate at which hydrogen is produced in the forward reaction. A further factor in the choice of metal oxide is whether the equilibrium hydrogen:steam ratio is sufficiently rich in steam to avert secondary hydriding. Generally, if the pressure of steam is greater than about 5% of the hydrogen pressure, it is believed that hydriding can be avoided. Generally, the oxides of iron, nickel, tin, bismuth and copper are employed. Bismuth oxide (Bi2O3) is typically employed when the metal oxide is to be placed in the fuel pellet column space as it minimizes parasitic neutron absorption from the introduction of metal oxide into the core. Copper oxide (CuO) is typically employed when the metal oxide is to be located at the bottom or at the top of the fuel column where parasitic neutron absorption is not a prime consideration. Oxides of specific isotopes of these materials that minimize parasitic absorption may also be employed. The following example serves to illustrate the present invention. Tests have been conducted where a zirconium strip was placed in a confined space within a stainless stell container and hydrogen admitted to the confined space through a very small hole in the container. The strip was shown to be massively hydrided within one day at 400xc2x0 C. However, when specific metal oxides were present within the confined space, in addition to the zirconium strip, no hydriding was evident when tested under the same configuration and test conditions. The tests were conducted with Fe2O3, Fe3O4, CuO, Bi2O3, NiO and SnO2. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
054328292	summary	BACKGROUND OF THE INVENTION The present invention relates to a fuel assembly and a reactor core, and more particularly to a fuel assembly for use in a boiling water reactor and a core of such a reactor. A conventional fuel assembly loaded in a boiling water reactor comprises a channel box in the form of a rectangular tube and a fuel bundle housed in the channel box. The fuel bundle comprises upper and lower tie plates respectively fitted to upper and lower portions of the channel box, a plurality of spacers installed in the channel box with intervals therebetween in the axial direction, a number of fuel rods penetrating through the spacers and arrayed in a square lattice pattern with their opposite ends fixed to the tie plates, and at least one water rod. Recently, raising a degree of burn-up of a fuel assembly has been attempted from the standpoints of prolonging the operating time, effectively utilizing uranium resource, and reducing the amount of spent fuel generated. For achieving a higher degree of burn-up, it is required to increase enrichment of a fuel assembly. With enrichment increasing, however, larger mean energy of neutrons has raised the problem that reactivity change due to void variations may increase, or effective utilization of fissionable material (fuel economy) may be impeded. The increased reactivity change due to void variations not only enlarges an absolute value of the void coefficient and lowers core stability, but also reduces a shutdown margin because of an increase in the hot-cold swing. Such a tendency is dealt with by increasing a moderator proportion (i.e., a ratio of moderator to fuel) in the fuel assembly and reducing mean energy of neutrons (i.e., making the neutron spectrum softer). In a boiling water reactor, control rods and neutron detecting counters are disposed outside the channel box. Therefore, a gap is defined between fuel assemblies for allowing those units to be inserted therein. Since the gap is filled with saturation water, those fuel rods which are positioned in a peripheral portion of the fuel assembly (i.e., in a region nearer to the gap) and those fuel rods which are positioned in a central portion of the fuel assembly are affected by the saturation water in the gap in different ways. Specifically, in the peripheral portion of the fuel assembly nearer to the gap, the ratio of moderator to fuel is so large as to increase a moderating effect, thus making nuclear fissions in the fuel rods at such a position more active. On the contrary, the fuel rods positioned in the central portion of the fuel assembly are less affected by a moderating effect due to the saturation water in the gap. Thus, the ratio of moderator to fuel, as a factor of determining nuclear characteristics of a fuel assembly, is different depending on the position of the fuel assembly. There are two methods of raising the ratio of moderator to fuel; i.e., a method of reducing a fuel inventory and a method of increasing a moderator region or moderator density. Practically, these methods are carried out by (1) increasing a boiling water region (e.g., diminishing the number of fuel rods or thinning the diameter of fuel rods), and (2) increasing a non-boiling water region (water rod region or gap water region). In the fuel assembly prepared by adopting one of the above methods, however, the fuel inventory is reduced in any case, meaning that fuel economy is improved from the aspect of enlarging the ratio of moderator to fuel, but fuel economy is impeded in terms of the fuel inventory. Eventually, an improvement in fuel economy is not achieved. Furthermore, the above-mentioned methods give rise to new problems. With the method (1), the reduced total length of fuel rods increases a linear heat generation rate and decreases a thermal margin. With the method (2), the reduced area of flow paths for a coolant makes a pressure drop larger. In a conventional fuel assembly, fuel rods are arrayed into a lattice pattern of 8 rows and 8 columns (i.e., 8.times.8). If the number of unit lattices in the fuel rod array is increased to 9.times.9 or 10.times.10, it would be possible to reduce a linear heat generation rate, enlarge a heat conducting area, and increase a thermal margin. Also, as illustrated in FIGS. 3 and 4 of JP, A, 52-50498, it is known to construct a fuel assembly by using partial length fuel rods which have a shorter fuel effective length. With this type fuel assembly, since the flow path area of a two-phase flow (in an upper portion of the core) having a large friction loss is enlarged, the pressure drop can be suppressed without reducing the fuel inventory. Consequently, by adopting the aforesaid methods (1) and (2) in addition to those two approaches, a fuel assembly can be obtained which is suitable for raising a degree of burn-up. In view of the above, there has been proposed a fuel assembly that fuel rods are arrayed in a lattice pattern of 9.times.9 or 10.times.10 with each fuel rod having a larger outer diameter but the number of fuel rods being increased, the cross-sectional area of a water rod is made larger than that of a unit lattice, and further a plurality of partial length fuel rods are arranged, as disclosed in JP, A, 62-276493, JP, A, 64-31089 and U.S. Pat. No. 5,068,082, for instance. More specifically, JP, A, 62-276493 discloses a fuel assembly having the increased number of unit lattices in the fuel rod array, in which a number of water rods or large-diameter water rods are arranged, and a plurality of partial length fuel rods are arranged in a row along a diagonal including corners of the lattice array of fuel rods. The partial length fuel rods are denoted by 14 FIGS. 1 and 5. In FIGS. 1, 7 and 8, etc. of JP, A, 64-31089, there is disclosed a fuel assembly having the increased number of unit lattices in the fuel rod array, in which large-diameter water rods are arranged and one or a plurality of partial length fuel rods P are arranged at one or more corners of the lattice array of fuel rods. In FIGS. 41 to 56 of U.S. Pat. No. 5,068,082, there are disclosed a fuel assembly having the increased number of unit lattices in the fuel rod array, in which a plurality of partial length fuel rods P are arranged together adjacently to large-diameter water rods. U.S. Pat. No. 5,068,082 also describes a fuel assembly having the increased number of unit lattices in the fuel rod array, in which large-diameter water rods are arranged, and a plurality of partial length fuel rods P are arranged in a row along a line bisecting each side of the lattice array of fuel rods at its outermost layer (e.g., FIGS. 2B, 6 and 10, etc.) or along a diagonal including corners of the lattice array of fuel rods (FIG. 5 and 12, etc.). Furthermore, U.S. Pat. No. 5,068,082 discloses another layout example of the partial length fuel rods P in which the partial length fuel rods P are arranged at each corner and the middle of each side of the lattice array of fuel rods in its outermost layer. SUMMARY OF THE INVENTION A first object of the present invention is to provide a fuel assembly which enables a higher degree of burn-up and reduces a void coefficient without lowering reactivity, and a reactor core loading such a fuel assembly therein. A second object of the present invention is to provide a fuel assembly which enables a higher degree of burn-up, reduces a void coefficient without lowering reactivity, and further makes a local power peaking flat. A third object of the present invention is to provide a fuel assembly which enables a higher degree of burn-up, reduces a void coefficient without lowering reactivity, and further makes a moderator's cross-sectional area of a neutron moderating rod optimum. A feature of the present invention to achieve a fuel assembly meeting the above first object resides in that the fuel rods include a plurality of first fuel rods and one or more second fuel rods having a shorter fuel effective length than said first fuel rods; (b) said second fuel rods are arranged in an outermost layer of a fuel rod array in the square lattice pattern at positions other than corners of the outermost layer; and (c) among the fuel rods inside said outermost layer of said fuel rod array in the square lattice pattern and arranged in a layer adjacent to said outermost layer, those fuel rods adjacent to said second fuel rods in said outermost layer are said first fuel rods. A feature of the present invention to achieve a reactor core meeting the above first object resides in that (a) said core includes a plurality of first fuel assemblies and a plurality of second fuel assemblies; (b) said first fuel assemblies each comprise a number of fuel rods arrayed in a square lattice pattern and at least one neutron moderating rod having a cross-sectional area of a moderator larger than a cross-sectional area of a unit lattice of the fuel rod array, said fuel rods including a plurality of first fuel rods and one or more second fuel rods having a shorter fuel effective length than said first fuel rods, said second fuel rods being arranged in an outermost layer of said fuel rod array in the square lattice pattern at positions other than corners of the outermost layer, among the fuel rods inside said outermost layer of said fuel rod array in the square lattice pattern and arranged in a layer adjacent to said outermost layer, those fuel rods adjacent to said second fuel rods in said outermost layer being said first fuel rods; and (c) said first fuel assemblies and said second fuel assemblies are loaded in a core central portion and a core circumferential portion, said first fuel assemblies having a smaller loading ratio in the core central portion than in the core circumferential portion. The second object of the present invention is achieved by that when said neutron moderating rod is projected in two directions orthogonal to each other in said fuel rod array in a square lattice pattern, said second fuel rods arranged in said outermost layer are lacated inside a projected range of said neutron moderating rod including the outermost opposite regions of the projected range. The third object of the present invention is achieved by setting the cross-sectional area of the moderator in said neutron moderating rods to 7-14 cm.sup.2. The present invention has been made based on the following studies. A description will now be given of results of the studies. Taking into account application to existing cores, it is required in development of fuel assemblies having a higher degree of burn-up that a pressure drop, a thermal margin (linear heat generation rate, critical power) and other parameters remain the same as those in existing fuel assemblies. As stated before, thinning the outer diameter of fuel rods and increasing the number of lattice cells in the fuel rod array are advantageous in achieving a fuel assembly with a higher degree of burn-up. However, such approaches raise new problems. If the number of lattice cells in the fuel rod array is simply enlarged, the degree of freedom in fuel array becomes larger, but the length of the peripheral edges is increased and so is a pressure drop. Also, the reduced outer diameter of fuel rods increases a time constant of fuel rods and hence makes stability (channel stability, core stability) more marginal. In order to solve those problems, an absolute value of the void coefficient is required to be smaller than that of the existing fuel assembly. Stated otherwise, while reducing an absolute value of the void coefficient has been discussed so far from the standpoint of increasing an enrichment of fuel, an absolute value of the void coefficient must be made still smaller in the case of enlarging the number of unit lattices in the fuel rod array. As stated in connection with the above methods (1) and (2), a reduction in the reactivity coefficient such as the void coefficient requires increasing the ratio of moderator to fuel, i.e., increasing the water rod region and reducing the fuel inventory. However, reducing the fuel inventory impedes fuel economy and hence should be avoided. Accordingly, it is important in development of fuel assemblies having a higher degree of burn-up to realize a new reactivity control method (for reducing an absolute value of the void coefficient and a hot-cold swing) with no need of reducing the fuel inventory. One reactivity control method without reducing the fuel inventory is to select positions of partial length fuel rods, as disclosed in the above-cited JP, A, 62-276493, JP, A, 64-31089 and U.S. Pat. No. 5,068,082. In one part of U.S. Pat. No. 5,068,082 and JP, A, 64-31089, a plurality of partial length fuel rods are arranged together at positions adjacent to large-diameter water rods or at corners of the lattice array of fuel rods. In another part of U.S. Pat. No. 5,068,082 and JP, A, 62-276493, a plurality of partial length fuel rods are arranged in a row along a diagonal including corners of the lattice array of fuel rods or along a line bisecting each side of the lattice array of fuel rods at its outermost layer. Any of these prior arts intends to promote a neutron moderating effect and reduce both the void coefficient and the hot-cold swing, by making the non-boiling water region (water rod region or gap water region) and the partial length fuel rods adjacent to each other. With those prior art schemes, a reactivity control capability is improved with a reduction in the reactivity coefficient such as the void coefficient, but sufficient cares have not been paid to change in reactivity itself and a local power peaking depending on positions where partial length fuel rods are set. More specifically, with the prior arts, singe a plurality of partial length fuel rods are arranged together in such a manner that at least a part of the partial length fuel rods is adjacent to the water rod or the gap water region at the corner, there accompanies a problem that resonance neutrons are absorbed more and the reactivity loss is so increased as to impede fuel economy. Also, in a cross-section above the partial length fuel rods, there arises a problem that the local power peaking of the fuel rods adjacent to the partial length fuel rods is increased and hence the thermal margin is reduced. Further, the above-cited prior arts have paid no considerations on how to make, in fuel assemblies aiming a higher degree of burn-up, the cross-sectional area of the water rods optimum in relation to the arrangement of the partial length fuel rods. Meanwhile, the following has been found from studies conducted by the inventors of this application. In the case of restrictively arranging or localizing the non-boiling water region, as a moderator, in a fuel assembly that the number of unit lattices in the fuel rod array is increased to 9.times.9 or more, localizing the moderator in the outer region of the fuel assembly facing the channel box is more effective (to provide higher sensitivity) than localizing it in the inner region of the fuel assembly facing the water rods for the purpose of improving the reactivity control capability (i.e., making smaller reactivity change due to variations in the void coefficient and hot-to-cold transition, and reducing the void coefficient in its absolute value) (see FIG. 1). Also, in the case of arranging those fuel rods which have a shorter fuel effective length than ordinary fuel rods, namely, partial length fuel rods, this is effective in a cross-section above the partial length fuel rods for improving the reactivity control capability (i.e., reducing the void coefficient) similarly to the above case of localizing the non-boiling water region. The sensitivity, which represents a rate of reduction in the void coefficient depending on positions where the partial length fuel rods are set, changes in the following order from a higher to lower level (see FIG. 2): (1) Fuel at corners of an outermost layer of the fuel assembly facing the channel box, PA1 (2) Fuel in the outermost layer of the fuel assembly facing the channel box other than (1), PA1 (3) Fuel in the inner region of the fuel assembly adjacent to the water rods, and PA1 (4) Fuel adjacent to neither the channel box nor the water rods. Moreover, in the case of arranging the partial length fuel rods in the outermost layer of the fuel rod array, a control rod worth is affected depending on their set positions such that the larger control rod worth is obtained by arranging the partial length fuel rods in the outermost layer at any positions facing the channel box rather than arranging them adjacently to the water rods (see FIG. 3). Accordingly, by arranging one or more second fuel rods in the form of partial length fuel rods in the outermost layer of the fuel rod array in the square lattice array, an effect of reducing the void coefficient can be obtained to improve the reactivity control capability. An effect of enhancing the control rod worth can also be expected, which contributes to an improvement in safety. Furthermore, for the case of arranging partial length fuel rods in the outermost layer of the fuel rod array in a fuel assembly that the number of unit lattices in the fuel rod array is increased to 9.times.9 or more, the following has been found about an influence of set positions of the partial length fuel rods upon reactivity and a local power peaking from studies conducted by the inventors of this application (see FIG. 4). When a partial length fuel rod is arranged at a corner position of the outermost layer, the reactivity loss and the local power peaking of that fuel rod which is adjacent to the partial length fuel rod are both large. When a partial length fuel rod is arranged at a lattice position adjacent to the corner position of the outermost layer, the reactivity loss is remarkably improved, but the local power peaking of both the fuel rod adjacent to the partial length fuel rod and the corner fuel rod (i.e., the fuel rod positioned at the corner) remains substantially large. When a partial length fuel rod is arranged at a third lattice position counting from the corner position of the outermost layer, the reactivity loss is further improved, but the local power peaking of both the fuel rod adjacent to the partial length fuel rod and the corner fuel rod is still large. When a partial length fuel rod is arranged at a fourth lattice position counting from the corner position of the outermost layer, i.e., when a partial length fuel rod is arranged inside a projected range of the water rods including both lattice positions at the outermost opposite regions of the projected range, the reactivity loss is almost zero and the local power peaking of both the fuel rod adjacent to the partial length fuel rod and the corner fuel rod is reduced to a large extent. Consequently, the reactivity loss is reduced by arranging the second fuel rods in the form of partial length fuel rods in the outermost layer of the fuel rod array in a square lattice array other than its corner positions. Also, the reactivity loss is reduced by arranging the second fuel rods in the outermost layer of the fuel rod array in a square lattice array other than its corner positions and those positions adjacent to the corner positions. More preferably, when the water rods or neutron moderating rods are projected in two directions orthogonal to each other in the fuel rod array in the square lattice pattern, by arranging the second fuel rods in the outermost layer of the fuel rod array in a lattice pattern inside the projected range including the outermost opposite regions of the projected range, it is possible to reduce both the reactivity loss and the local power peaking, thereby improving fuel economy and a thermal margin. In addition, the following has been found from studies conducted by the inventors of this application. Specifically, to maximally utilize an effect resulted from localizing the moderator region, ordinary fuel rods are required to be localized. The localization of the ordinary fuel rods reduces probability that resonance neutrons are absorbed, and hence contributes to a further improvement in fuel economy. If, among the fuel rods inside the outermost layer of the fuel rod array in the square lattice pattern and arranged in a layer adjacent to the outermost layer, those fuel rods adjacent to the second fuel rods in the outermost layer are the first fuel rods, the region of the first fuel rods is surrounded by the moderator region and, as a result, thermal neutrons efficiently decelerated through the moderator region are caused to flow into the region of the first fuel rods with higher efficiency. Therefore, resonance absorption is reduced to improve not only the reactivity control capability but also fuel economy. This effect is further enhanced by making the region, where the first fuel rods are arranged, spread over one entire layer adjacent to the outermost layer, and is still further enhanced by making that region spread over two layers adjacent to the outermost layer. Meanwhile, preferably, by arranging, inside the outermost layer of the fuel rod array in the lattice pattern, one or more third fuel rods in the form of partial length fuel rods in a layer adjacent to the outermost layer at its corners, there is obtained an effect of rendering distribution of a coolant flow rate and distribution of a vapor volume rate more uniform within the channel box. In the region facing the channel box, particularly, in the region near its corners, friction resistance is generally so large that the coolant flow rate tends to decrease. This tendency can be overcome by arranging the partial length fuel rods at respective corners of the layer adjacent to the outermost layer. Incidentally, the expression "A is adjacent to B" used here in connection with the arrangement of fuel rods implies all such conditions that A is adjacent to B not only in the row and column directions, but also in oblique directions. Further, based on studies conducted by the inventors of this application, the following has been found about how fuel economy is affected by localizing the moderator region (i.e., the non-boiling water region), and how the cross-sectional area and shape of water rods are made optimum. For a fuel assembly that the number of unit lattices in the fuel rod array is increased to 9.times.9 or more, comparing the case of enlarging the water rod region inside the fuel rod assembly and the case of enlarging the gap water region outside the fuel rod assembly on a condition that the fuel inventory is kept constant, the water rod region is more effective (to provide higher sensitivity) than the gap water region in a point of increasing a neutron infinite multiplication factor (i.e., improving fuel economy). Accordingly, the cross-sectional area of the water rods requires to be enlarged for an improvement in fuel economy (see FIG. 5). In this case, an optimum range of the cross-sectional area of the water rods is from 7 to 14 cm.sup.2. Making the cross-sectional area of the water rods optimum is also related to stability. Stability is evaluated in terms of two modes; i.e., channel stability and core stability. An increase in both the uranium inventory and the cross-sectional area of the water rods reduces a margin of the channel stability, while an increase in the uranium inventory and a decrease in the cross-sectional area of the water rods degrades the core stability. Therefore, the allowable zone from the viewpoint of stability ranges from 9 to 11 cm.sup.2 in terms of the cross-sectional area of the water rods, the range being defined by a limit line of the channel stability and a limit line of the core stability (see FIG. 6). To enlarge the cross-sectional area of the water rods, adopting a large-size water rod is advantageous in reducing the number of fuel rods which must be sacrificed, and reducing the coolant flow passage area which is less effective to cool fuel rods (i.e., increasing a critical power). Assuming that the spacings between the water rods and the fuel rods adjacent to the water rods are constant, it is most preferable in the case of circular water rods to use the unit lattices of 2.times.2 as a water rod for effective utilization of the space. Therefore, preferably, by setting the moderator's cross-sectional area of one or more neutron moderating rods to the range of 7 to 14 cm.sup.2 the reactivity is enhanced and fuel economy is further improved. Also, by setting the moderator's cross-sectional area of the neutron moderating rods to the range of 9 to 11 cm.sup.2 the channel stability, the core stability, as well as fuel economy are improved. Additionally, the improved stability renders equipment installed for higher stability unnecessary. Moreover, by arranging the neutron moderating rods in such a region as able to accommodate 7 to 12 fuel rods, and locating the water rod region such that two or more of four lattice positions adjacent to each of the fuel lattice positions in the water rod region are those positions where the water rod region adjoins, a large-size circular water rod with the size corresponding to 2.times.2 cells can be arranged three or four in a fuel assembly having the fuel rod array of 10.times.10, and two in a fuel assembly having the fuel rod array of 9.times.9. Therefore, the coolant flow passage area which is less effective to cool fuel rods is diminished and the critical power is increased. In addition, the following has been found by considering the above results of the studies together. Arranging the partial length fuel rods adjacently to the water rods is substantially equivalent to enlarging the water rod region at the center of the fuel assembly and, therefore, has an effect of increasing a neutron infinite multiplication factor of the fuel assembly. Also, by making the number of the partial length fuel rods adjacent to the channel box larger than the number of partial length fuel rods adjacent to the water rods, the number of partial length fuel rods required in terms of the reactivity control capability can be cut down, which provides an effect of increasing the control rod worth. Therefore, preferably, by arranging one or more third fuel rods in the form of partial length fuel rods adjacently to the neutron moderating rods, the neutron infinite multiplication factor is further increased and both the reactivity and fuel economy are improved. Also, preferably, by making the number of second fuel rods arranged in the outer layer of the fuel rod array in a square lattice array larger than the number of third fuel rods given by the partial length fuel rods arranged adjacently to the neutron moderating rods, the number of partial length fuel rods used can be cut down while ensuring a necessary level of the reactivity and increasing the control rod worth. In addition, it has been found from studies conducted by the inventors of this application that, by arranging the partial length fuel rods adjacently to each other, there can be obtained a greater effect of improving both the reactivity control capability and the control rod worth than resulted from summing an effect obtainable with one partial length fuel rod alone. Therefore, preferably, by arranging two second fuel rods in the form of partial length fuel rods in at least one side of the outermost layer of the fuel rod array in a square lattice pattern, the effect of enhancing both the reactivity control capability and the control rod worth is doubled. Also, by arranging those two second fuel rods adjacently to each other, there can be obtained an effect twice or more as much as that in the case of arranging two second fuel rods not adjacently to each other, in point of enhancing both the reactivity control capability and the control rod worth. Additionally, it has been found from studies conducted by the inventors of this application that, to improve the channel stability and the core stability, upper ends of the partial length fuel rods are advantageously positioned at a level from 4th-stage spacer to 6th-stage spacer. This spacer level corresponds to 1/2-3/4 in terms of a ratio of the fuel effective length to the full fuel rod length. Therefore, preferably, by setting the second fuel rods in the form of partial length fuel rods to have a fuel effective length in a range of 1/2 to 3/4 of that of the ordinary fuel rods, there can be obtained an effect of improving the channel stability and core stability. Reactor cores are primarily grouped into C lattice cores of the type that the gap water region on the side, through which a crucial control rod is inserted, has the same gap width as that of the gap water region on the opposite side, and D lattice cores of the type that the gap water region on the side, through which a cross-shaped control rod is inserted, has a larger gap width than that of the gap water region on the opposite side. Preferably, by arranging second fuel rods at least one for each of two adjacent sides of the outermost layer of the fuel rod array in a square lattice pattern, there can obtained a fuel assembly suitable for being loaded into D lattice cores. A spectral shift rod can adjust a neutron moderating effect with a water level therein changed depending on the core flow rate and, as a result, it can be utilized to control reactivity or power. Meanwhile, in a BWR fuel assembly, burn-up reactivity is generally controlled by gadolinia. To effectively perform reactivity control or power control with a water level in the spectral shift rod, the amount of gadolinia requires to be reduced. Since the reactivity control effect is enhanced and the shutdown margin is improved by arranging the partial length fuel rods, the amount of gadolinia can be reduced. As a result, it is possible to improve fuel economy and achieve the best use of an effect of the spectral shift rod. If the number of unit lattices in the fuel rod array is increased from 8.times.8 to 9.times.9 or more, the number of layers of fuel rods constituting a fuel assembly is enlarged and hence time degree of freedom in layout for distributing the fuel rods in the fuel assembly is increased. Accordingly, fuel or moderators can be localized in the fuel assembly, meaning that the above-explained arrangement can easily be realized. Incidentally, the term "localization (localized or localizing" used herein implies that, in the fuel or moderator region surrounded by boundary lines between fuel and moderators, the length of the boundary lines per unit volume is shortened. Finally, a method of loading fuel assemblies according to the present invention will now be described. The fuel assembly (first fuel assembly) of the present invention has a feature that, since many partial length fuel rods are used, the fuel inventory largely varies in the axial direction. Accordingly, supposing a retrofitted core based on an existing core in which conventional fuel assemblies (second fuel assemblies) are loaded, an effect of axial neutron flux distribution due to an axial difference in fuel inventory must be taken into consideration. If the first fuel assembly is loaded among the second fuel assemblies having no partial length fuel rods or the smaller number of partial length fuel rods than the first fuel assembly, there is found a tendency for the second fuel assemblies to increase the power in a core lower portion and, on the contrary, for the first fuel assembly to increase the power in a core upper portion. With the method of loading fuel assemblies according to the present invention, the first fuel assemblies and the second fuel assemblies are loaded in the core central portion and the core circumferential portion such that the first fuel assemblies have a smaller loading ratio in the core central portion than in the core circumferential portion, whereby the linear power generation rate can be held not larger than a set value.
043483526	summary	This invention relates to a rack for intermediate storage of nuclear reactor fuel element bundles. Racks for intermediate storage of new fuel element bundles or of bundles removed from nuclear reactors according to the head concept of the above claim 1 are known, where the receiving tubes stand loosely on a bottom plate of a tank and are braced laterally by a structure which is firmly connected with the bottom plate and is braced laterally against the tank walls. Since for safety reasons--in view of earthquakes--considerable horizontal accelerations must be expected, the supporting structure, to be designed for buckling, becomes very massive if the intermediate storage areas are large. Thus, there result relatively large distances between the individual receiving tubes. It is an object of the invention to provide an intermediate storage rack which at equal earthquake safety has smaller receiving tube distances, so that it is possible to lodge a given number of tube bundles in a smaller water tank or, in a water tank of given size, a larger number of fuel element bundles. Briefly, the invention provides a storage rack for fuel element bundles. The rack is comprised of a bottom plate, a plurality of vertically disposed square receiving tubes each of which has an inwardly directed flange resting on the bottom plate and which defines a bore for centering a fuel element bundle and screws passing through each flange and into the bottom plate to secure the tubes to the plate. This construction not only requires less space than previously known racks, but also eliminates, either entirely or partly, any need for lateral bracing of a support sturcture against a tank wall. Thus, continuous channels disposed laterally of the rack for transfer of the fuel element bundles from the reactor safety tank and/or for manipulating the fuel element bundles can remain free. The bottom plate may also be provided with passage openings aligned with each bore so that a first circulation path for the water cooling the fuel element in natural circulation is provided. A plurality of ribs may also be secured to the underside of the plate to stiffen the plate. The ribs which extend at least in the direction of the longer dimension of the rack, provide a certain carrying capacity against buckling with a smaller thickness of the bottom plate and hence lower cost of material. Vertically adjustable screws may also be secured to the above plate for vertically adjusting the plate. In this way, by an exact leveling of the bottom plate, not only are difficulties in applying the manipulator avoided, but moreover the danger of the bottom plate buckling in case of earthquakes is reduced. Each tube may also be provided with lateral inlet openings for the passage of circulating water. This opens up additional paths for natural circulation of the water cooling the fuel elements. The cooling is thereby made more uniform and intensified. A cover plate may also be mounted over the tubes with square openings coaxially of the tubes with a depending frame secured to the plate and vertical angle sections securing the frame to the bottom plate. This cover plate reduces by its damping the ability of the system to swing in case of earthquakes. Besides, it permits a simple lateral bracing of the rack elements in the region of the upper ends of the receiving tubes, whether against each other or against the tank wall. Diagonally disposed flat rods may also be secured at each end to a respective end of an adjacent pair of angle sections. In this way, the cover plate is able to absorb higher transverse forces. The rack may also be sub-divided into a plurality of units with connecting elements screwed into the cover plates to secure adjacent units together. In addition, the cover plates are rigidly connected with one another also in the direction of traction. The connecting elements may also function as guides for installation and removal of the fuel element bundles. By transferring the guide function to the connecting elements, the cover plate is saved from additional chip-removing machining.
046648739	abstract	The invention is directed to remotely-controlled manipulator carrier systems for use in maintaining and servicing process equipment housed in large-area cells affected by radioactivity wherein industrial processes are conducted. The cells are part of a nuclear facility for reprocessing irradiated nuclear fuel materials. The process equipment is arranged along mutually adjacent walls of the cell and so defines a canyon-like passageway extending in the direction of the longitudinal axis of the cell. The system includes a first overhead bridge crane having a trolley movable thereon in a direction transverse to said longitudinal axis. The trolley includes a hoist for lowering and raising a hook for engaging and moving a component of the process equipment in a first vertical plane transverse to said axis. A second overhead bridge crane is disposed beneath the first bridge crane. Tracks are provided for guiding the bridge cranes in the enclosure in respective horizontal planes and in the direction of said longitudinal axis. The second overhead bridge crane includes an elongated supporting member arranged transversely to said axis above the passageway and engaging the track for movement therealong. A manipulator assembly includes a mast connected to the elongated supporting member and extends downwardly into the canyon-like passageway from the supporting member. The supporting member and the mast conjointly define a second vertical plane transverse to said axis. The manipulator assembly further includes a manipulator or performing manual-like operations on the process equipment, the manipulator being mounted asymmetrically on the mast so as to be on one side thereof and in a third vertical plane transverse to said axis. The first bridge crane is movable along the track to bring the first vertical plane into coincidence with the third vertical plane so as to permit movement of the hook in the third vertical plane clear of the transverse elongated supporting member whereby both the manipulator and the hook can be brought simultaneously to a predetermined work location at the process equipment.
	abstract	In a method for operating a computed tomography apparatus, having an x-ray source rotatable around a system axis and a radiation detector with a detector-proximate beam-gating diaphragm, and a patient support, a spiral scan of a patient on the patient support is conducted by rotating the x-ray source around the system axis while moving the subject on the patient support substantially parallel to the system axis. The diaphragm have movable absorber elements that are curved, and are moved independently of each other toward and away from each other in a direction substantially parallel to the system axis during the spiral scan. The absorber elements are dynamically adjusted in an asymmetrical manner during the spiral scan to reduce overexposure of the examination subject to x-rays.
062663933	summary	FIELD OF THE INVENTION The present invention relates generally to a multiple layer multileaf collimator for use during radiation treatment to shape and control the spatial distribution of a radiation beam. BACKGROUND OF THE INVENTION During conventional radiation therapy treatment, radiation beams of varying angles and intensities are directed at a target in a patient. Normal tissue and organs located in the path of the radiation beams must be taken into account for safety reasons, thereby limiting the dose that can be delivered to the target. Many techniques are known for shaping the radiation beams so that the radiation is concentrated at the target and is minimized or eliminated at the normal tissues. One of the techniques is conformal radiation therapy wherein the beam aperture varies from angle to angle via a multileaf collimator which employs a multiplicity of radiation blockers, called leaves. Each individual leaf in a multileaf collimator can be positioned independently, allowing the user to create an infinite amount of irregularly shaped fields. The radiation beams are directed between the ends of opposing arrays of the radiation blocking collimator leaves, thereby shaping the beam to closely match the shape of the desired treatment area, while shielding the normal tissue and organs. An example of such a system is U.S. Pat. No. 5,166.531 to Huntzinger which describes a multileaf collimator arrangement positioned about the central axis of a radiation emitting head for shaping an emitted radiation beam. The collimator includes two opposing arrays of side-by-side elongated radiation blocking collimator leaves. Each leaf in each opposing array can be moved longitudinally towards or away from the central axis of the beam, thus defining a desired shape through which the radiation beam will pass. However, because the adjoining leaves must be tightly positioned side-by-side in order to minimize radiation leakage between the leaves, friction is an inherent problem, creating complications in maintaining a set position of one leaf while re-positioning an adjacent leaf, such repositioning being frequently required in conformal therapy. If friction between the adjacent leaves is reduced by providing a looser fit between adjacent leaves, unacceptable radiation leakage through spaces between the adjacent leaves will result. On the other hand, maintaining a tight leaf fit between the adjacent leaves and providing a lubricating layer in the contact area of the adjacent leaves is also not an acceptable solution because the lower density of the lubricating layer, as compared to the high density of the collimator leaves, will allow an unacceptable amount of radiation leakage to occur. U.S. Pat. No. 5,591,983 to Yao, the disclosure of which is incorporated herein by reference, attempts to solve the friction problem by providing a collimator that comprises first and second layers of a plurality of elongated radiation blocking leaves. The leaves of each layer are arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and are movable in a longitudinal direction (Y) which is generally traverse to the direction of the beam so as to define a radiation beam shaping field between the opposed ends of the leaves. The layers are arranged one above another in the beam direction and offset in a lateral direction (X) generally transverse to the beam direction and orthogonal to the longitudinal direction (Y) so that spaces between adjacent leaves of the first and second layers are positioned over and under, respectively, leaves of the second and first layers, respectively. One disadvantage of the Yao system is that an irregularly shaped target is poorly covered by the two layers of leaves. In addition, the overall thickness of the leaves in the beam direction is quite large. Other patent documents related to multileaf collimators include U.S. Pat. Nos. 4,233,519 to Coad, 4,534,052 to Milcamps, 4,672,212 to Brahme, 4,739,173 to Blosser et al., 4,754,147 to Maughan et al., 4,794,629 to Pastyr et al., 4,868,843 and 4,868,844 to Nunan, 5,012,506 to Span et al., 5,165,106 to Barthelmes et al., 5,207,223 to Adler, 5,343,048 to Paster, 5,351,280 to Swerdloff et al., 5,427,097 to Depp, 5,438,991 to Yu et al., 5,442,675 to Swerdloff et al., and 5,555,283 to Shiu et al. SUMMARY OF THE INVENTION The present invention seeks to provide a multiple layer multileaf collimator with non-parallel layers that cross over each other. For example, a first layer employs radiation blocking leaves that are movable in a longitudinal direction (Y) which is generally traverse to the direction of the beam so as to define a radiation beam shaping field between the opposed ends of the leaves. A second layer of leaves positioned above the first layer employs radiation blocking leaves that are movable in a cross-over direction (X) which is generally traverse to the direction of the beam so as to define a radiation beam shaping field between the opposed ends of the leaves, the cross-over direction (X) being generally orthogonal to longitudinal direction (Y). Alternatively, the first and second layers can be angled with respect to each other at an angle other than 90.degree.. Although gaps in the X-Y plane are formed between the crossed-over layers these gaps are sufficiently small so as to pass a negligibly safe amount of radiation. Unlike the prior art, any irregularly shaped target can be accurately covered by the two layers of leaves. In addition, the overall thickness of the leaves in the beam direction is significantly less than the prior art. There is thus provided in accordance with a preferred embodiment of the present invention a multileaf collimator for use in a radiation system providing a radiation beam in a given beam direction, including a first layer of a plurality of radiation blocking leaves, the leaves being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a longitudinal direction (Y) which is generally traverse to the beam direction so as to define a radiation beam shaping field between the opposed ends of the leaves, a second layer of a plurality of radiation blocking leaves, the leaves of the second layer being arranged adjacent one another so as to form two opposed rows of adjacently positioned leaves and being movable in a cross-over direction (X) which is generally traverse to the beam direction and angled with respect to the longitudinal direction (Y) so as to define a radiation beam shaping field between the opposed ends of the leaves of the second layer, and actuator apparatus for moving the leaves of the first layer in the longitudinal direction (Y) and the leaves of the second layer in the cross-over direction (X), wherein the first and second layers are arranged one above another in an overlapping manner in the beam direction. In accordance with a preferred embodiment of the present invention gaps are formed generally traverse to the beam direction and generally in a plane of the X and Y directions, each of the gaps only allowing an amount of radiation to pass therethrough below a predetermined threshold. Further in accordance with a preferred embodiment of the present invention the cross-over direction is generally orthogonal to the longitudinal direction. Still further in accordance with a preferred embodiment of the present invention the leaves of the first layer and the leaves of the second layer are housed in a frame. Additionally in accordance with a preferred embodiment of the present invention there is provided a source of radiation for providing a radiation beam in the given beam direction. Imaging apparatus may also be provided for imaging a target irradiated by the radiation beam. In accordance with a preferred embodiment of the present invention an optical control device is provided that monitors travel of any of the leaves and signals the actuator apparatus to stop moving the leaves. The multileaf collimator may include a plurality of the first layers of radiation blocking leaves and/or a plurality of the second layers of radiation blocking leaves.
	claims	1. A charged particle beam irradiation apparatus comprising:a transport line configured to transport a charged particle beam; anda rotating gantry rotatable around a rotation axis,wherein the transport line comprisesan inclined section configured to make the charged particle beam advancing in a direction of the rotation axis advance to be inclined so as to become more distant from the rotation axis,a rotation section configured to turn the charged particle beam advanced in the inclined section to a rotational direction of the rotation axis; anda bending section configured to bend the charged particle beam turned to the rotational direction to the rotation axis,the rotating gantry is formed of a tubular body which accommodates an irradiated body and supports the transport line, andthe inclined section of the transport line is disposed to pass through the inside of the tubular body of the rotating gantry,wherein the rotating gantry has an irradiation chamber in which one end side in a direction in which the rotation axis extends is closed by a back panel, andthe inclined section of the transport line is disposed to pass through the back panel and the inside of the tubular body of the rotating gantry. 2. The charged particle beam irradiation apparatus according to claim 1, wherein the transport line includesa first bent section configured to change a traveling direction of the charged particle beam advancing in the direction of the rotation axis and introduce the charged particle beam into the inclined section,a second bent section that is provided downstream of the inclined section and configured to change the traveling direction of the charged particle beam to a direction orthogonal to the rotation axis,a third bent section that is provided downstream of the second bent section and configured to change the traveling direction of the charged particle beam to the rotational direction of the rotation axis, anda fourth bent section configured to bend the charged particle beam passed through the inside of the third bent section to the rotation axis side. 3. A charged particle beam irradiation apparatus comprising:a transport line configured to transport a charged particle beam; anda rotating gantry rotatable around a rotation axis,wherein the transport line has at least an inclined section configured to make the charged particle beam advancing in a direction of the rotation axis advance to be inclined so as to become more distant from the rotation axis,the rotating gantry has an irradiation chamber of a tubular body in which one end side in a direction in which the rotation axis extends is closed by a back panel, andthe inclined section of the transport line is disposed to pass through the back panel and the inside of the tubular body of the rotating gantry.
	claims	1. A method for use with a radiation therapy treatment platform having a multi-layer multi-leaf collimation system wherein a first layer of collimation leaves are laterally offset with respect to a second layer of collimation leaves, the method comprising:by a control circuit:generating a preliminary radiation treatment plan using a model of a radiation therapy treatment platform comprising a single-layer multi-leaf collimation system having a single layer of collimation leaves that may be the same as or different than the first layer or the second layer of collimation leaves of the multi-layer multi-leaf collimation system;using the preliminary radiation treatment plan to generate a final radiation treatment plan for the radiation therapy treatment platform that takes into account at least two layers of collimation leaves. 2. The method of claim 1 wherein the collimation leaves of the first layer have a same width as the collimation leaves of the second layer, and wherein the single-layer multi-leaf collimation system of the model of the radiation therapy treatment platform presumes that the single-layer multi-leaf collimation system has collimation leaves having a width that is less than the width of the collimation leaves of the first layer and the second layer. 3. The method of claim 2 wherein the width of the collimation leaves for the single-layer multi-leaf collimation system is approximately one half the width of the collimation leaves of the first layer and the second layer. 4. The method of claim 1 wherein the single-layer multi-leaf collimation system of the model of the radiation therapy treatment platform presumes that the single-layer multi-leaf collimation system has a greater number of collimation leaves than either of the first layer and the second layer. 5. The method of claim 4 wherein the single-layer multi-leaf collimation system of the model of the radiation therapy treatment platform presumes that the single-layer multi-leaf collimation system has at least twice as many collimation leaves as the first layer or the second layer. 6. The method of claim 1 wherein the single-layer multi-leaf collimation system of the model of the radiation therapy treatment platform presumes that the single-layer multi-leaf collimation system has collimation leaves that are less in width and greater in number than the collimation leaves of the first layer and the second layer. 7. The method of claim 6 wherein generating the final radiation treatment plan that takes into account a second layer of collimation leaves comprises, at least in part, modifying the preliminary radiation treatment plan to avoid apertures that the multi-layer multi-leaf collimation system cannot produce. 8. The method of claim 6 wherein generating the final radiation treatment plan that takes into account a second layer of collimation leaves comprises, at least in part, determining how to produce planned apertures formed by the single-layer multi-leaf collimation system using the multi-layer multi-leaf collimation system. 9. The method of claim 8 further comprising:using the final radiation treatment plan to administer radiation to a patient using the multi-layer multi-leaf collimation system. 10. The method of claim 1 wherein the collimation leaves of the single-layer multi-leaf collimation system are presumed to have better transmission parameters than the collimation leaves of either the first layer or the second layer. 11. The method of claim 10 wherein generating the final radiation treatment plan that takes into account a second layer of collimation leaves comprises, at least in part, refining apertures defined in the preliminary radiation treatment plan to provide higher resolution by optimizing collimation leaf positions in the second layer of collimation leaves. 12. An apparatus comprising:a control circuit configured to generate a radiation treatment plan by:generating a preliminary version of the radiation treatment plan using a model of a radiation therapy treatment platform comprising a single-layer multi-leaf collimation system;using the preliminary version of the radiation treatment plan to generate a final version of the radiation treatment plan that takes into account a second layer of collimation leaves. 13. The apparatus of claim 12 further comprising:a multi-layer multi-leaf collimation system wherein a first layer of collimation leaves are laterally offset with respect to a second layer of collimation leaves, wherein the collimation leaves of the first layer have a same width as the collimation leaves of the second layer; andwherein the single-layer multi-leaf collimation system of the model of the radiation therapy treatment platform presumes that the single-layer multi-leaf collimation system has collimation leaves having a width that is less than the width of the collimation leaves of the first layer and the second layer and that the single-layer multi-leaf collimation system has a greater number of collimation leaves than either of the first layer and the second layer. 14. The apparatus of claim 13 wherein the width of the collimation leaves for the single-layer multi-leaf collimation system is approximately one half the width of the collimation leaves of the first layer and the second layer. 15. The apparatus of claim 13 wherein the control circuit is configured to generate the final radiation treatment plan that takes into account a second layer of collimation leaves by, at least in part, modifying the preliminary radiation treatment plan to avoid apertures that the multi-layer multi-leaf collimation system cannot produce. 16. The method of claim 15 wherein the control circuit is configured to generate the final radiation treatment plan that takes into account a second layer of collimation leaves by, at least in part, determining how to produce planned apertures formed by the single-layer multi-leaf collimation system using the multi-layer multi-leaf collimation system. 17. The apparatus of claim 16 further comprising:a radiation therapy treatment platform configured to use the final radiation treatment plan to administer radiation to a patient using the multi-layer multi-leaf collimation system. 18. The apparatus of claim 12 wherein the collimation leaves of the single-layer multi-leaf collimation system are presumed to have better transmission parameters than the collimation leaves of a multi-layer multi-leaf collimation system. 19. The apparatus of claim 18 wherein the control circuit is configured to generate the final radiation treatment plan that takes into account a second layer of collimation leaves by, at least in part, refining apertures defined in the preliminary radiation treatment plan to provide higher resolution by optimizing collimation leaf positions in the second layer of collimation leaves.
	claims	1. A semiconductor manufacturing method, comprising:registering an exposure device as a device for performing a pattern writing processing and an electron beam direct writing device as an alternative to the exposure device, when creating process flow information by sequentially registering processing conditions of processings in a semiconductor manufacturing process with a semiconductor production management system;searching for information on the pattern writing processing based on the process flow information before the pattern writing processing;determining whether or not a mask used by the exposure device for performing the pattern writing processing searched for is installed in the exposure device; andsetting the exposure device to perform the pattern writing processing in the case where it has been determined that the mask is installed in the exposure device, or setting the electron beam direct writing device to perform the pattern writing processing in the case where it has been determined that the mask is not installed in the exposure device. 2. The semiconductor manufacturing method according to claim 1, wherein, when creating the process flow information, as the timing of setting of the device for performing the pattern writing processing, a timing of before the pattern writing processing or a timing of a desired number of processings before the pattern writing processing is registered, andwhen setting the device for performing the pattern writing device, it is determined at the registered timing whether or not the mask is installed in the exposure device, and according to the result of the determination, the exposure device or the electron beam direct writing device is set to perform the pattern writing processing. 3. The semiconductor manufacturing method according to claim 2, wherein, when creating the process flow information, as the timing of setting of the device for performing the pattern writing processing, any of a timing of before the pattern writing processing, a timing of a desired number of processings before the pattern writing processing and a timing of a desired number of days before the pattern writing processing is selected and registered, andwhen setting the device for performing the pattern writing device, in the case where the timing of a desired number of days before the pattern writing processing is registered as the timing of setting of the device for performing the pattern writing processing, the timing of setting of the device for performing the pattern writing processing is determined based on the processing date and hour of each of the processings. 4. The semiconductor manufacturing method according to claim 1, wherein, when setting the device for performing the pattern writing processing, in the case where the mask is installed in the exposure device after the electron beam direct writing device is set to perform the pattern writing processing the exposure device is set again to perform the pattern writing processing. 5. The semiconductor manufacturing method according to claim 1, wherein, when setting the device for performing the pattern writing processing, at a timing when the device for performing the pattern writing processing is switched from the exposure device to the electron beam direct writing device or switched from the electron beam direct writing device to the exposure device, information indicating the switching of the device for performing the pattern writing device is transmitted to a predetermined destination. 6. The semiconductor manufacturing method according to claim 1, wherein, when setting the device for performing the pattern writing processing, in the case where only the exposure device is registered as the device for performing the pattern writing processing, the exposure device is set to perform the pattern writing processing regardless of whether or not the mask is installed in the exposure device. 7. A semiconductor manufacturing method, comprising:registering an exposure device as a device for performing a pattern writing processing and an electron beam direct writing device as an alternative to the exposure device, when creating process flow information by sequentially registering processing conditions of processings in a semiconductor manufacturing process with a semiconductor production management system;searching for information on the pattern writing processing based on the process flow information before the pattern writing processing;determining whether or not a mask used by the exposure device for performing the pattern writing processing searched for is installed in the exposure device;setting the exposure device to perform the pattern writing processing in the case where it has been determined that the mask is installed in the exposure device, or setting the electron beam direct writing device to perform the pattern writing processing in the case where it has been determined that the mask is not installed in the exposure device; andwriting a desired circuit pattern on a semiconductor substrate by the exposure device performing the pattern writing processing in the case where the mask is installed in the exposure device, or writing the circuit pattern on the semiconductor substrate by the electron beam direct writing device performing the pattern writing processing in the case where the mask is not installed in the exposure device. 8. The semiconductor manufacturing method according to claim 7, wherein, when creating the process flow information, as the timing of setting of the device for performing the pattern writing processing, a timing of before the pattern writing processing or a timing of a desired number of processings before the pattern writing processing is registered, andwhen setting the device for performing the pattern writing device, it is determined at the registered timing whether or not the mask is installed in the exposure device, and according to the result of the determination, the exposure device or the electron beam direct writing device is set to perform the pattern writing processing. 9. The semiconductor manufacturing method according to claim 8, wherein, when creating the process flow information, as the timing of setting of the device for performing the pattern writing processing, any of a timing of before the pattern writing processing, a timing of a desired number of processings before the pattern writing processing and a timing of a desired number of days before the pattern writing processing is selected and registered, andwhen setting the device for performing the pattern writing device, in the case where the timing of a desired number of days before the pattern writing processing is registered as the timing of setting of the device for performing the pattern writing processing, the timing of setting of the device for performing the pattern writing processing is determined based on the processing date and hour of each of the processings. 10. The semiconductor manufacturing method according to claim 7, wherein, when setting of the device for performing the pattern writing processing, in the case where the mask is installed in the exposure device after the electron beam direct writing device is set to perform the pattern writing processing the exposure device is set again to perform the pattern writing processing. 11. The semiconductor manufacturing method according to claim 7, wherein, when setting the device for performing the pattern writing processing, at a timing when the device for performing the pattern writing processing is switched from the exposure device to the electron beam direct writing device or switched from the electron beam direct writing device to the exposure device, information indicating the switching of the device for performing the pattern writing device is transmitted to a predetermined destination. 12. The semiconductor manufacturing method according to claim 7, wherein, when setting the device for performing the pattern writing processing, in the case where only the exposure device is registered as the device for performing the pattern writing processing, the exposure device is set to perform the pattern writing processing regardless of whether or not the mask is installed in the exposure device. 13. A semiconductor manufacturing apparatus, comprising:a process flow information creating section which registers an exposure device as a device for performing the pattern writing processing and an electron beam direct writing device as an alternative to the exposure device, when creating process flow information by sequentially registering processing conditions of processings in a semiconductor manufacturing process; anda control section which searches for information on the pattern writing processing based on the process flow information before the pattern writing processing, determines whether or not a mask used by the exposure device for performing the pattern writing processing searched for is installed in the exposure device, and sets the exposure device to perform the pattern writing processing in the case where it has been determined that the mask is installed in the exposure device, or sets the electron beam direct writing device to perform the pattern writing processing in the case where it has been determined that the mask is not installed in the exposure device. 14. The semiconductor manufacturing apparatus according to claim 13, wherein, as the timing of setting of the device for performing the pattern writing processing, the process flow information creating section registers a timing of before the pattern writing processing or a timing of a desired number of processings before the pattern writing processing, andthe control section determines at the registered timing whether or not the mask is installed in the exposure device and, according to the result of the determination, sets the exposure device or the electron beam direct writing device to perform the pattern writing processing. 15. The semiconductor manufacturing apparatus according to claim 14, wherein, as the timing of setting of the device for performing the pattern writing processing, the process flow information creating section selects and registers any of a timing of before the pattern writing processing, a timing of a desired number of processings before the pattern writing processing and a timing of a desired number of days before the pattern writing processing, andin the case where the timing of a desired number of days before the pattern writing processing is registered as the timing of setting of the device for performing the pattern writing processing, the control section determines the timing of setting of the device for performing the pattern writing processing based on the processing date and hour of each of the processings. 16. The semiconductor manufacturing apparatus according to claim 13, wherein, in the case where the mask is installed in the exposure device after the electron beam direct writing device is set to perform the pattern writing processing, the control section sets the exposure device again to perform the pattern writing processing. 17. The semiconductor manufacturing apparatus according to claim 13, wherein, at a timing when the device for performing the pattern writing processing is switched from the exposure device to the electron beam direct writing device or switched from the electron beam direct writing device to the exposure device, the control section transmits information indicating the switching of the device for performing the pattern writing device to a predetermined destination. 18. The semiconductor manufacturing apparatus according to claim 13, wherein, in the case where only the exposure device is registered as the device for performing the pattern writing processing, the control section sets the exposure device to perform the pattern writing processing regardless of whether or not the mask is installed in the exposure device.
042591546	claims	1. A nuclear reactor containment structure comprising a closure casing, a pedestal within said closure casing for supporting a pressure vessel, a diaphragm floor dividing said closure casing into a drywell and a pressure suppression chamber and supported at its inner peripheral end by said pedestal and at portions near the outer peripheral end of the floor by vertical columns, said outer peripheral end of the diaphragm floor being free from contact with the inner surface of said closure casing, said diaphragm floor including structural steel beams radially extending between said pedestal and said closure casing, and a plurality of shear keys secured to the inner surface of said closure casing and associated with the respective outer ends of said steel beams to permit the vertical and radial movements between said steel beams and said closure casing but restrain the circumferential movement therebetween, said diaphragm floor further including concrete layers disposed between the respective adjacent steel beams. 2. A nuclear reactor containment structure as set forth in claim 1, wherein said concrete layers have the substantially same thickness as the height of said steel beams to form substantially flat upper and lower surfaces of said diaphragm floor. 3. A nuclear reactor containment structure as set forth in claim 1 wherein said diaphragm floor includes a steel plate layer positioned below the concrete layers. 4. A nuclear reactor containment structure as set forth in claim 1 wherein there is provided means for connecting an inner peripheral end of said diaghragm floor to said pedestal. 5. A nuclear reactor containment structure as set forth in claim 3 wherein there is provided means for connecting an inner peripheral end of said diaphragm floor to said pedestal. 6. A nuclear reactor containment structure as set forth in claim 4 wherein said connecting means includes anchor bolts. 7. A nuclear reactor containment structure as set forth in claim 4 wherein said connecting means includes anchor plates. 8. A nuclear reactor containment structure as set forth in claim 5 wherein said connecting means includes anchor bolts. 9. A nuclear reactor containment structure as set forth in claim 5 wherein said connecting means includes anchor plates. 10. A nuclear reactor containment structure as set forth in claim 1, wherein an annular elastic sealing element is provided between the outer peripheral end of said diaphragm floor and the closure casing. 11. A nuclear reactor containment structure as set forth in claim 1, wherein said structural steel beams are H-shaped structural steel beams. 12. A nuclear reactor containment structure as set forth in claim 1, wherein said concrete layers comprise two outer layers of unreinforced concrete, for insulating heat, and a layer of reinforced concrete sandwiched therebetween. 13. A nuclear reactor containment structure as set forth in claim 12, wherein said concrete layers have the substantially same thickness as the height of said steel beams to form substantially flat upper and lower surfaces of said diaphragm floor. 14. A nuclear reactor containment structure as set forth in claim 1, wherein the diaphragm floor includes a steel plate layer positioned on the upper side of the structural steel beams. 15. A nuclear reactor containment structure as set forth in claim 11, wherein steel plates are positioned to bridge across the upper side of the lower flanges of the H-shaped structural steel beams.
	abstract	An apparatus for deriving X-ray absorbing and phase information comprises; a splitting element for splitting spatially an X-ray, a detector for detecting intensities of the X-rays transmitted through an object, the intensity of the X-rays changing according to X-ray phase and also position changes, and an calculating unit for calculating an X-ray transmittance image, and an X-ray differential phase contrast or phase sift contrast image as the phase information. The X-ray is split into two or more X-rays having different widths, and emitted onto the detector unit. And, the calculating unit calculates the X-ray absorbing and phase information based on a difference, between the two or more X-rays, in correlation between the changing of the phase of the X-ray and the changing the intensity of the X-ray in the detector unit.
	claims	1. An apparatus for directing positively charged particles at least into a patient, comprising:an accelerator configured to accelerate the positively charged particles;a beam transport system configured to transport the positively charged particles from said accelerator to an output nozzle, said output nozzle supported by a gantry in a treatment room; andmultiple patient interface controllers, comprising:a first pendant, said first pendant tethered and moveable within the treatment room; anda second pendant, said second pendant tethered and moveable within a control room by a window, said treatment room separated by said window from said control room,wherein each of said first pendant and said second pendant provide an operator interface for control of: (1) the positively charged particles and (2) a position of a patient positioner configured to position the patient during use. 2. The apparatus of claim 1, wherein said second pendant comprises all controls, for control of the positively charged particles, of said first pendant. 3. The apparatus of claim 2, further comprising:a treatment delivery control system communicatively linked to both said first pendant and said second pendant. 4. The apparatus of claim 3, said first pendant further comprising:a flow process control unit, said process control unit configured to control all of:a beam state modifier positioned in said output nozzle;a patient positioner position;an imaging system at least partially supported by said gantry; andtransport of the positively charged particles from said accelerator. 5. The apparatus of claim 1, said multiple patient interface controllers further comprising:a first fixed position motion control system workstation. 6. The apparatus of claim 5, further comprising:identical controls of said first pendant and said second pendant, wherein said identical controls comprise a subset of controls of said fixed position motion control system workstation. 7. The apparatus of claim 6, said multiple patient interface controllers further comprisinga second fixed position motion control system workstation, said first fixed position motion control system positioned in said control room, said second fixed position motion control system positioned in said treatment room, said first fixed position motion control system comprising redundant control, with said first pendant, of the positively charged particles. 8. The apparatus of claim 1, said second pendant further comprising:a workflow process control selector comprising control of:positioning the patient relative to the positively charged particles in a process of using the positively charged particles to generate a tomographic image of the tumor. 9. The apparatus of claim 8, said second pendant in said control room further comprising control of:imaging the patient with the positively charged particles; andusing the positively charged particles to treat the patient. 10. The apparatus of claim 9, said workflow process control selector further comprising control of:a first imaging element mounted to said gantry at a first location and configured to move with rotation of said gantry;a second imaging element configured to move with rotation of said gantry, a first imaging beamline passing through said first location and the patient, a second imaging beamline passing through said second imaging element and the patient, the first beamline and the second beamline forming an angle of greater than seventy degrees and less than one hundred ten degrees. 11. The apparatus of claim 8, said workflow process control selector further comprising:position control of a tray retractable into said output nozzle, said tray comprising: a tray insert and an electromechanical communicator configured to connect to a receptor, said communicator configured with information about said tray insert. 12. A method for directing positively charged particles at least into a patient, comprising the steps of:accelerating the positively charged particles using an accelerator;transporting the positively charged particles from said accelerator to an output nozzle using a beam transport system;supporting said output nozzle using a gantry in a treatment room; andusing multiple patient interface controllers, comprising:a first pendant, said first pendant tethered and moveable within the treatment room; anda second pendant, said second pendant tethered and moveable within a control room by a window, said treatment room separated by said window from said control room,said step of using multiple patient interface controllers further comprising the step of: using each of said first pendant and said second pendant as an operator interface for control of: (1) the positively charged particles and (2) a position of a patient positioner configured to position the patient during use. 13. The method of claim 12, further comprising a step of:a first operator using said first pendant in said treatment room in a first control step of loading a patient specific beam state alteration tray into said output nozzle. 14. The method of claim 13, further comprising the step of:the first operator using said first pendant in said treatment room in a second control step of moving a patient positioner, constraining movement of the patient, into a treatment position. 15. The method of claim 12, further comprising the step of:a first operator using said second pendant in said control room in a first control step of directing the positively charged particles into the patient. 16. The method of claim 15, further comprising the steps of:the first operator using said second pendant in said control room in a second control step of imaging the patient; andusing output from said step of imaging the patient in repeating said step of directing the positively charged particles into the patient. 17. The method of claim 12, further comprising the step of:using said first pendant in steps of:imaging the patient with a first imaging beam passing through said gantry; andimaging the patient with a second imaging beam passing through said gantry, the first imaging beam, the patient, and the second imaging beam forming an angle of greater than forty degrees. 18. The method of claim 17, further comprising the step of:imaging the patient with the positively charged particles. 19. The method of claim 18, further comprising the step of:using at least two imaging material emitting photons upon passage of the positively charged particles in identifying a vector of the positively charged particles. 20. The method of claim 17, the positively charged particles substantially following a part of the first imaging beam in said treatment room, said first imaging beam comprising X-rays.
	claims	1. A decontamination formulation comprising: (a) from about 1% to about 15% by weight of chloroisocyanuric acid; (b) from about 1% to about 10% by volume of a co-solvent selected from the group consisting of polypropylene glycol, polyethylene glycol , and derivative thereof and mixtures thereof; (c) from about 1% to about 15% by volume of a surfactant; (d) a buffer to maintain said formulation at a pH from about 11 and which fails over time allowing the pH to fall to a pH of about 8.5; and (e) the balance being water. 2. A decontamination formulation of claim 1 , wherein the chloroisocyanuric acid is in an amount from about 3% to about 9% by weight. claim 1 3. The decontamination formulation of claim 1 , wherein said chloroisocyanuric acid is selected from the group consisting of an alkali metal salt of monochloroisocyanuric acid, dichloroisocyanuric acid, and a combination thereof with cyanuric acid. claim 1 4. The decontamination formulation of claim 3 , wherein said alkali metal salt of dichloroisocyanuric acid is sodium dichloroisocyanurate. claim 3 5. The decontamination formulation of claim 1 , wherein the buffer maintains the pH of the formulation above 8.5 for at least 30 minutes. claim 1 6. The decontamination formulation of claim 1 , wherein the co-solvent is in an amount of from about 6% to about 10% by volume. claim 1 7. The decontamination formulation of claim 1 , wherein polypropylene glycol has the chemical formula R 1 xe2x80x94(OCH(CH 3 )CH 2 ) n xe2x80x94OR 2 , where R 1 and R 2 are independently H, an alkyl, or an ester group and n greater than 1. claim 1 8. The decontamination formulation of claim 7 , wherein said alkyl group representing R 1 or R 2 are independently selected from the group consisting of a methyl, ethyl, propyl, and butyl group. claim 7 9. The decontamination formulation of claim 7 , wherein at least one of said R 1 or R 2 is hydrogen. claim 7 10. The decontamination formulation of claim 7 , wherein R 1 and R 2 are hydrogen. claim 7 11. The decontamination formulation of claim 1 , wherein said polypropylene glycol derivative is a partially etherified polypropylene glycol. claim 1 12. The decontamination formulation of claim 11 , wherein said partially etherified polypropylene glycol has the formulae R 1 xe2x80x94(OCH(CH 3 )CH 2 ) n xe2x80x94OR 2 , where one of R 1 or R 2 is independently H, or an alkyl group and nxe2x89xa71. claim 11 13. The decontamination formulation of claim 12 , wherein said R 1 or R 2 are independently selected from the group consisting of a methyl, ethyl, propyl, and butyl group. claim 12 14. The decontamination formulation of claim 13 , wherein at least one of R 1 or R 2 is hydrogen. claim 13 15. The decontamination formulation of claim 1 , wherein said buffer is capable of initially maintaining said formulation at a pH of from about 10 to about 11. claim 1 16. The decontamination formulation of claim 1 , wherein said buffer comprises a mixture of sodium tetraborate decahydrate, anhydrous sodium carbonate and sodium hydroxide. claim 1 17. The decontamination formulation of claim 1 , wherein said buffer comprises a mixture of sodium tetraborate decahydrate, anhydrous sodium carbonate and sodium metasilicate pentahydrate. claim 1 18. The decontamination formulation of claim 1 , wherein said surfactant comprises a composition of the formulae [R(OCH 2 CH 2 ) n X] a M b , where R is an alkyl group having from eight to eighteen carbon atoms; n is an integer from 0 to 10; X is selected from the group consisting of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 ; M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [R(OCH 2 CH 2 ) n X]. claim 1 19. The decontamination formulation of claim 1 wherein said surfactant comprises a composition of the formulae [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X] a M b where R is an alkyl group having from eight to eighteen carbon atoms; m is an integer from 0 to 3; X is selected from the group consisting of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 , M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X]. claim 1 20. The decontamination formulation of claim 1 wherein said surfactant further comprises a compound of the formula Rxe2x80x94OH, where R is an alkyl groups having from eight to sixteen carbon atoms. claim 1 21. The decontamination formulation of claim 1 , further comprising lithium hypochlorite in an amount of from about 5% to about 10% by weight of said chloroisocyanuric acid salt. claim 1 22. A method of preparing a decontamination formulation comprising the steps of adding to a stream of water: (a) a first aqueous solution comprising of up to about 30% by weight of chloroisocyanuric acid; (b) a second aqueous solution comprising a mixture of inorganic buffer salts adjusted to an initial pH of about 10 to 11 and capable of maintaining the pH of said decontamination formulation from about 11 and which fails over time, allowing the pH to fall to a pH of about 8.5; (c) a co-solvent selected from the group consisting of polypropylene glycol, polyethylene glycol and a derivative thereof; and (d) a surfactant. 23. The method of claim 22 , wherein the surfactant is a foaming agent. claim 22 24. The method of claim 22 , wherein the surfactant is a foaming agent comprising a composition of the formula [R(OCH 2 CH 2 ) n X] a M b , where R is an alkyl group having from eight to eighteen carbon atoms; n is an integer from 0 to 10; X is selected from the group of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 ; M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [R(OCH 2 CH 2 ) n X]. claim 22 25. The method of claim 22 , wherein the surfactant is a foaming agent comprising a composition of the formula [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X] a M b , where R is an alkyl group having from eight to eighteen carbon atoms; m is an integer from 0 to 3; X is selected from the group of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 , M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X]. claim 22 26. The method of claim 22 wherein said surfactant comprises a compound of the formula Rxe2x80x94OH, where R is an alkyl group having from eight to sixteen carbon atoms. claim 22 27. The method of claim 22 , wherein said first aqueous solution additionally comprises a lithium hypochlorite in amounts of up to 10% of the chloroisocyanuric acid salt. claim 22 28. A kit for providing a decontamination composition comprising the following components in packaged form: (a) a decontaminant comprising chloroisocyanuric acid; or its alkali metal or alkaline earth metal salt or a substance thereof; (b) a co-solvent selected from the group consisting of polypropylene glycols, polyethylene glycols, and derivatives and mixtures thereof; (c) a surfactant; and (d) a mixture of sodium tetraborate decahydrate, anhydrous sodium carbonate and sodium hydroxide. 29. A kit as claimed in claim 28 , wherein said decontaminant further includes lithium hypochlorite. claim 28 30. A kit as claimed in claim 28 , wherein said chloroisocyanuric acid is sodium dichloroisocyanurate. claim 28 31. A kit as claimed in claim 28 , wherein said surfactant comprises a composition of the formulae [R(OCH 2 CH 2 ) n X] a M b , where R is an alkyl group having from eight to eighteen carbon atoms; n is an integer from 0 to 10; X is selected from the group of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 ; M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [R(OCH 2 CH 2 ) n X]. claim 28 32. A kit as claimed in claim 28 , wherein said surfactant comprises a composition of the formulae [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X] a M b where R is an alkyl group having from eight to eighteen carbon atoms; m is an integer from 0 to 3; X is selected from the group of SO 3 2xe2x88x92 , SO 4 2xe2x88x92 , CO 3 2xe2x88x92 and PO 4 3xe2x88x92 , M is an alkali metal, alkaline earth metal, ammonium derivative or amine derivative; a is the valence of M and b is the valence of [Rxe2x80x94CHxe2x95x90CH(CH 2 ) m xe2x80x94X]. claim 28 33. A kit as claimed in claim 28 , wherein said surfactant comprises a compound of the formula Rxe2x80x94OH, where R is an alkyl group having from eight to sixteen carbon atoms. claim 28 34. A kit as claimed in claim 28 , wherein said composition components (a) and (b) are individually packaged and components (c) and (d) are packaged as a mixture or components (a) and (d) are packaged as a mixture and components (b) and (c) are packaged as a mixture. claim 28 35. A kit as claimed in claim 28 , wherein said composition components are individually packaged. claim 28 36. A method for decontaminating surfaces comprising the steps of: (a) preparing a decontamination formulation of from about 1% to about 15% by weight of a chloroisocyanuric acid salt, from about 1% to about 10% by volume of a co-solvent selected from the group consisting of polypropylene glycol, polyethylene glycol, and derivatives thereof and mixtures thereof, from about 1% to about 15% by volume of a surfactant, and a buffer to initially maintain said formulation at a pH from about 11 and which fails over time allowing the pH to fall to a pH of about 8.5 and water to form an aqueous solution; and (b) applying the aqueous solution to contaminated surfaces. 37. The decontamination method of claim 36 wherein the buffer maintains the pH of the aqueous solution above 8.5 for at least 30 minutes. claim 36 38. The decontamination method of claim 36 further comprising the steps of: claim 36 (a) foaming the aqueous solution; then (b) applying the foamed aqueous solution to the contaminated surface. 39. The decontamination method of claim 38 wherein the foaming step comprises dispensing the aqueous solution through an aeration nozzle. claim 38 40. The decontamination method of claim 39 wherein the buffer maintains the pH of the aqueous solution above 8.5 for at least 30 minutes. claim 39 41. The decontamination method as recited in claim 36 wherein all of the chloroisocyanuric acid salt, co-solvent, surfactant, and the buffer are combined with water before applying to the contaminated surface. claim 36 42. The decontamination method as recited in claim 36 wherein claim 36 (a) the co-solvent and surfactant are combined with water to form a non-degrading solution; and (b) the buffer and the chloroisocyanuric acid salt are added separately to the non-degrading solution before applying to the contaminated surface.
059404647	claims	1. A tube of zirconium-base alloy for constituting all or the outside portion of cladding for a nuclear fuel rod or of a guide tube for a nuclear fuel assembly, made of a zirconium-base alloy containing, by weight, 0.8% to 1.8% niobium, 0.2% to 0.6% tin, and 0.02% to 0.4% iron, plus inevitable impurities, and having a carbon content controlled to lie in the range 30 ppm to 180 ppm, a silicon content in the range 10 ppm to 120 ppm, and an oxygen content in the range 600 ppm to 1800 ppm. 2. A tube according to claim 1, wherein the alloy is in recrystallized state. 3. A tube according to claim 1, wherein the alloy is in relaxed state. 4. A tube according to claim 1, wherein the alloy has set contents: 0.9% to 1.1% niobium, 0.25% to 0.35% tin, and 0.2% to 0.3% iron. 5. A method of manufacturing a tube according to claim 1, including the following steps of: making a bar of an alloy containing 0.8% to 1.8% niobium, 0.2% to 0.6% tin, and 0.02% to 0.4% iron; after heating in the bar to a temperature in the range 1000.degree. C. to 1200.degree. C., quenching the bar in water; drawing the bar into a blank after heating to a temperature in the range 600.degree. C. to 800.degree. C.; annealing the drawn blank at a temperature in the range 590.degree. C. to 650.degree. C.; and cold rolling the annealed blank in at least four passes into a tube, with intermediate heat treatments at temperatures in the range 560.degree. C. to 620.degree. C. 6. A method according to claim 5, wherein the rolling passes are performed on tubes having increasing recrystallization ratios. 7. A method according to claim 5, further including a recrystallizing final heat treatment step at a temperature in the range 560.degree. C. to 620.degree. C. 8. A method according to claim 5, further including a strain relieving final heat treatment step at a temperature in the range from about 470.degree. C. to 500.degree. C.
039740293	claims	1. A gas-cooled nuclear reactor having A. a main circulatory system which includes a portion passing through the reactor core, B. a gaseous coolant flowing through said circulatory system, C. said circulatory system including at least one main energy converting unit, and D. an auxiliary circulatory system for the gaseous coolant, 2. A nuclear reactor according to claim 1 having at least one closed cycle gas turbine as a main energy converting unit, and in which the gaseous coolant is used directly as the working fluid for the turbine. 3. A nuclear reactor according to claim 1 in which said auxiliary steam turbine is arranged to run continuously in operation of the reactor. 4. A nuclear reactor according to claim 3 in which the steam turbine is used to drive an auxiliary generator. 5. A nuclear reactor according to claim 2 in which a plurality of gas turbines are disposed around the reactor core, and a plurality of auxiliary boilers disposed around the reactor core between the gas turbines. 6. A nuclear reactor according to claim 5 wherein the gas turbines and the auxiliary boilers are accommodated in chambers formed within the thickness of a pressure vessel wall surrounding the reactor core. 7. A nuclear reactor according to claim 5 wherein the gas turbines and the auxiliary boilers are disposed within the space between a primary pressure vessel accommodating the reactor core and a secondary pressure vessel enclosing at least part of the wall of the primary pressure vessel. 8. A high temperature gas-cooled nuclear reactor according to claim 1 having means for cooling the gas between the reactor core and the auxiliary boiler or boilers. 9. A nuclear reactor according to claim 8 wherein the cooling means includes means for mixing the gas emerging from the reactor core with gas at a lower temperature. 10. A nuclear reactor according to claim 9 in which the or each said auxiliary circulatory system incorporates a by-pass circuit by which some of the gas leaving an associated auxiliary boiler is caused to by-pass the reactor core and mix with the heated gas leaving the core before the latter is fed back to the boiler. 11. A nuclear reactor according to claim 8, in which the thermal capacity of the auxiliary circulatory/steam turbine system is about 16% to 20% of the thermal capacity of the core.
054003733	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 illustrates an elevational view of a preferred embodiment of the grid assembly fixture of this invention. As shown in FIG. 1, the fixture is comprised of a rectangular grid assembly plate 1 mounted on a pedestal 2 attached to a work surface 3, such as a work bench. The pedestal 2 is adjustable using the handle 4, which allows the grid assembly plate to be raised or lowered or tilted relative to the work surface 3, and then locked in place to provide a position most comfortable to the person assembling the grid. A suitable pedestal 2 work positioner is the Model 24001 manufactured by Marshall Industries, Dayton, Ohio. As best shown in FIGS. 3 and 4, the grid assembly plate 1 is a rectangular metal plate having four sides, each side having a flange 5 extending along a major portion of the length of each side. The top surface of the grid assembly plate is provided with a plurality of spaced slots 6 extending across the plate 1 in both directions. The slots 6 are adapted to receive the thin metal grid straps 7 that are interconnected at right angles to each other to form a network of interconnected grid cells similar to an "egg crate divider". The periphery or border of the grid is comprised of four grid straps 7 attached to each other and to the ends of the internal grid straps 7 to form a rectangular metal grid of interconnected open grid cells. The four border grid straps that make up the periphery of the grid are positioned along the four sides of the grid assembly plate 1 and are supported on the top surfaces of the four flanges 5, as illustrated in FIG. 1. As shown in FIGS. 1 and 2, a clamp support plate 8 is attached to the underside of each flange 5 on the sides of the grid assembly plate 1 and projects outwardly away from the flange 5. A toggle action clamp assembly 9 is attached to the end of each clamp support plate 8. The toggle action clamp assembly 9 is a commercially available assembly by which movement of the handle 10 moves a spindle 11 inwardly or outwardly relative to the sides of the grid assembly plate 1. The movement of the toggle action clamp assembly handle 10 may be manual or by a pneumatic piston and cylinder. For this embodiment, a Model 601 toggle action clamp assembly manufactured by Destaco Division, Drover Corporation, Detroit, Mich. was used. The end of the spindle 11 of each toggle action clamp assembly 9 is provided with an elongated clamping pad 12 adapted to move tightly against each of the sides of the grid assembly plate 1 just above the side flange 5. The clamping pad 12 is preferably about the length of the side flange 5 and preferably is provided with a face 17 for the clamping pad 12, as best illustrated in FIGS. 7 and 8. The toggle action clamp assemblies 9 and the clamping pads 12 are designed to temporarily press against and hold the four peripheral or border grid straps 7 tightly in place against the four sides of the grid assembly plate 1 and against the ends of the interconnected internal grid straps 7. After the four peripheral or border grid straps 7 are held tightly in place using the toggle action clamp assemblies 9 and the clamping pads 12, a rectangular strap retention assembly 13 is placed around the outside of the four peripheral or border grid straps 7 to hold the peripheral or border grid straps 7 in the proper position for welding of the grid strap 7 connections. When the rectangular strap retention assembly 13 is tightly in place holding the four peripheral or border grip straps 7 tightly against the sides of the grid assembly plate 1, the toggle action clamp assemblies 9 are released allowing the clamping pads 12 to be withdrawn from contact with the peripheral or border grid straps 7. As best illustrated in FIGS. 5 and 6, the rectangular strap retention assembly 13 is comprised of four rails 14 and four corner brackets 15 that allow the rails to be quickly connected to each other around the sides of the grid assembly plate 1 using openings in the corner brackets and knurled thumb screws 16. The clamping pads 12 are preferably provided with a face 17 for the clamping pad 12 that allows the rails 14 to be easily inserted through the face 17 of the clamping pads 12 while they are clamped against the border straps 7 and the sides of the grid assembly plate 1. This invention allows the efficient assembly of nuclear fuel assembly grids and is readily adapted to the assembly of grids of any geometry or size. It also provides a more accurate and precise alignment of the grid straps and components for further fabrication and welding. It is believed that the present invention and its advantages will be understood from the above description and the accompanying drawings and it will be apparent that changes may be made in the form, construction and arrangement of the apparatus of this invention without departing from the scope of this invention. PARTS IDENTIFICATION LIST 1. grid assembly plate PA0 2. pedestal PA0 3. work surface PA0 4. handle PA0 5. flanges on grid assembly plate PA0 6. slots on grid assembly plate PA0 7. grid straps PA0 8. clamp support plate PA0 9. toggle action clamp assemblies PA0 10. handle on clamp assemblies PA0 11. spindle PA0 12. clamping pad PA0 13. strap retention assembly PA0 14. rails PA0 15. corner brackets PA0 16. thumb screws PA0 17. face of clamping pad (12)
	claims	1. A method of fabricating a doped III-nitride semiconductor body, comprising:placing a substrate in a reactor chamber;pumping reactant gas for growing said III-nitride semiconductor body into said reactor chamber; andimplanting ions from an implanter, disposed in an implanter chamber in communication with said reactor chamber, into said III-nitride semiconductor body to dope said III-nitride semiconductor body as said III-nitride semiconductor body is continually grown, said implanter chamber separate from said reactor chamber, and said implanter chamber comprising a plurality of subchambers;wherein said ions pass through a space under a near vacuum prior to reaching said substrate. 2. The method of claim 1, wherein said near vacuum is obtained through differential pumping of said space. 3. The method of claim 1, wherein said space is said implanter chamber that further includes an ion path extending from said implanter to said reactor chamber, said ion path being in communication with said reactor chamber. 4. The method of claim 1, further comprising measuring a dosage of said ions while implanting takes place. 5. The method of claim 4, wherein said dosage is measured using a Faraday cup. 6. The method of claim 1, further comprising changing a direction of travel of said ions. 7. The method of claim 6, wherein deflection plates are used to change direction of travel of said ions. 8. The method of claim 1, further comprising implanting said ions through a metal mask. 9. The method of claim 1, wherein said ions are implanted through rastering. 10. An apparatus for fabrication of a doped III-nitride semiconductor body, comprising:a reactor chamber to receive a substrate and reactant gas to grow a III-nitride semiconductor body on said substrate;an implanter disposed in an implanter chamber that is in communication with said reactor chamber to implant ions into said 111-nitride semiconductor body and to dope said III-nitride semiconductor body during the growth of said III-nitride semiconductor body, said implanter chamber separate from said reactor chamber, and said implanter chamber comprising a plurality of subchambers; andat least one pump coupled to said implanter chamber configured to create a near vacuum condition in said implanter chamber. 11. The apparatus of claim 10, wherein said at least one pump performs differential pumping. 12. The apparatus of claim 10, further comprising a Faraday cup disposed within said reactor chamber. 13. The apparatus of claim 10, further comprising deflection plates to change the direction of travel of said ions.
062020382	claims	1. A method of monitoring a source of data for determining an operating condition of a selected system, comprising the steps of: providing reference data characteristic of an operating condition of a reference system; collecting selected data from said source of data and which is characteristic of an operating condition of a selected system; performing a bounded angle ratio test procedure on said reference data and said selected data to determine whether there is a deviation of said selected data for said selected system relative to said reference data of said reference system; and generating an indication upon determining the deviation. at least one first computer module operative to provide at least one of a second computer module operative to perform a similarity angle analysis on said reference data and said selected data for determining similarity angle data characteristic of a similarity value; and a third computer program module operative to receive and operate on the similarity value to determine whether a deviation exists for the monitored system relative to the model system. providing reference data characteristic of an operating condition of a reference system; collecting selected data from a source of data with said selected data characteristic of an operating condition of a selected system; performing a bounded angle ratio test procedure on said reference data and said selected data to determine a measure of similarity of said selected data for said selected system relative to said reference data of said reference system; and analyzing said measure of similarity to determine the operating condition of said selected system relative to said reference system. a first data source for providing reference data characteristic of an operating condition of a reference system; a second data source for providing selected data characteristic of an operating condition of a selected system; and a computer module operative to perform a bounded angle ratio test procedure on said reference data and said selected data to determine a measure of similarity of said selected data for said selected system relative to said reference data of said reference system and further operative to analyze a deviation of the selected system from the reference system, thereby enabling a user to act responsive to any said deviation. a monitored operational system selected from the group consisting of a biological system, a financial system, an industrial system, a chemical system and a physical system; at least one first computer module operative to accumulate reference data characteristic of learned states of a reference operational condition of a model system of said group and to accumulate selected data characteristic of an operational condition of a selected system of said group; a second computer module operative to perform a similarity angle analysis on said reference data and on said selected data for determining similarity angle data characteristic of a similarity value; and a third computer program module operative to receive and operate on the similarity angle data to determine whether a deviation exists for the monitored operational system relative to the model system. a data source for providing current data of a selected system; a data source for providing reference data of a reference system; and a computer module operative to render a measure of statistical similarity between the current data and the reference data, the computer module determining a statistical combination of a set of similarity values for corresponding data values of the current data and the reference data, wherein the similarity values are determined by a computer program which compares the data values from the current data to the corresponding data values from the reference data by performing a bounded angle ratio test. 2. The method as defined in claim 1 wherein the deviation is an item selected from the group consisting of said selected system and the source of the data. 3. The method as defined in claim 2 wherein the source of the data comprises at least one of a sensor and a data base. 4. The method as defined in claim 1 wherein said indication comprises an alarm signal. 5. The method as defined in claim 1 wherein the step of performing a bounded angle ratio test procedure comprises comparing; a first angle in a first triangle having a base opposite said first angle with a length along said base proportional to the difference between corresponding values comprised of a value of said selected data and a value in said reference data, to a second angle in a second triangle having a base opposite said second angle with a length proportional to the range over all values in said reference data. 6. The method as defined in claim 5 wherein the first and second triangles share a common altitude line segment. 7. The method as defined in claim 1 wherein the step of determining a deviation of said selected data relative to said reference data includes calculating a similarity angle. 8. The method as defined in claim 1 wherein the selected data from the source of data is being processed in substantially real time. 9. The method as defined in claim 1 wherein the selected data from the source of data are derived at least in part from previously accumulated data. 10. The method as defined in claim 1 wherein the method includes another step of performing a sequential probability ratio test on said selected data characteristic of the operating condition of the selected system. 11. An apparatus for monitoring a data source for determining a selected operating condition of a monitored system, comprising: 12. The apparatus as defined in claim 11 wherein the reference operational condition of the model system comprises a normal operating condition. 13. The apparatus as defined in claim 11 wherein said third computer program module includes a computer program which performs a sequential probability ratio test on the similarity angle data. 14. The apparatus as defined in claim 11 wherein said similarity angle analysis of said second module computer program comprises a bounded angle ratio test to determine a similarity angle characteristic of the operating condition of the monitored system relative to the operational condition of the model system. 15. The apparatus as defined in claim 14 wherein the bounded angle ratio test includes a computer program to establish a reference point R positioned adjacent a similarity domain line characteristic of a similarity domain with the point R at a distance h of closest approach to said similarity domain line. 16. The apparatus as defined in claim 15 wherein the second computer module establishes a minimum value X.sub.min and a maximum value X.sub.max over a statistical distribution over the similarity domain. 17. The apparatus as defined in claim 16 wherein a line from the reference point R to said similarity domain intersects at point X.sub.med, a median over all corresponding values in said reference data. 18. The apparatus as defined in claim 17 wherein an angle defined by lines from the reference point R to the values X.sub.min and X.sub.max is 90.degree.. 19. The apparatus as defined in claim 18 wherein the similarity angle is defined by: .theta.=arc tan(X.sub.1 /h)-arc tan(X.sub.0 /h), where X.sub.1 is a larger value and X.sub.0 is a smaller value in a set formed by a first value from said selected data and a second corresponding value selected from said reference data. 20. The apparatus as defined in claim 11 wherein the data source comprises at least two sources of data and said first computer module includes a computer program which operates to monitor at least two sources of data separately when the at least two sources of data are uncorrelated. 21. The apparatus as defined in claim 11 wherein the data source comprises at least two sources of data and said at least one first computer module includes a computer program which operates to monitor said at least two sources of data and perform a regression analysis on the similarity angle data. 22. The apparatus as defined in claim 21 wherein the regression analysis comprises a sequential probability ratio test. 23. The apparatus as defined in claim 11 wherein the similarity angle data is used in a computer program to compute estimated data characteristic of the operating condition of said selected system. 24. The apparatus as defined in claim 11 wherein a source of estimated data for the selected system is provided from a preaccumulated data source based upon determining a sensor of the data source has malfunctioned. 25. The apparatus as defined in claim 24 wherein the data source derives from a plurality of individual sources of data which provide signals selected from the group of non-linearly and linearly related signals. 26. A method of monitoring a source of data for determining an operating condition of a selected system relative to a reference system, comprising the steps of: 27. A method according to claim 26 further comprising the step of generating an estimate of said selected data based on said measure of similarity. 28. A method according to claim 26 further comprising the step of performing a statistical hypothesis test on said selected data and said estimate thereof, to determine if there is a statistically significant deviation between them. 29. A method according to claim 28 wherein said statistical hypothesis test comprises a sequential probability ratio test. 30. The method as defined in claim 26 further including the step of carrying out an action responsive to analyzing said measure of similarity. 31. An apparatus for monitoring an operating condition of a selected system relative to a reference system, comprising: 32. An apparatus according to claim 31 further comprising a computer module operative to generate an estimate of said selected data based on said measure of similarity. 33. An apparatus according to claim 32 further comprising a computer module operative to perform a statistical hypothesis test on said selected data and said estimate thereof, to determine if there is a statistically significant deviation between them. 34. An apparatus according to claim 33 wherein said statistical hypothesis test is a sequential probability ratio test. 35. An interconnected system for monitoring a data source for determining all operating condition of a monitored system relative to a model system, comprising: 36. The interconnected system as defined in claim 35 wherein said reference operational condition of said model system is selected from the group consisting of a normal operational condition and a deviating operational condition. 37. The interconnected system as defined in claim 35 wherein said second computer module includes a computer program which performs a sequential probability ratio test on said similarity angle data for at least one of said reference data and said selected data. 38. The interconnected system as defined in claim 35 wherein said similarity angle analysis of said second module computer program comprises a bounded angle ratio test to determine a similarity angle characteristic of the operating condition of the monitored system relative to the operational condition of the model system. 39. The interconnected system as defined in claim 38 wherein the bounded angle ratio test includes a computer program to establish a reference point R positioned adjacent a similarity domain line characteristic of a similarity domain with the point R at a distance h of closest approach to said similarity domain line. 40. The interconnected system as defined in claim 39 wherein the second computer module establishes a minimum value X.sub.min and a maximum value X.sub.max over a statistical distribution over the similarity domain. 41. The interconnected system as defined in claim 40 wherein a line from the reference point R to said similarity domain intersects at point X.sub.med, a median over all corresponding values in said reference data. 42. The interconnected system as defined in claim 41 wherein an angle defined by lines from the reference point R to the values X.sub.min and X.sub.max is 90.degree.. 43. The interconnected system as defined in claim 42 wherein the similarity angle is defined by: .theta.=arc tan(X.sub.1 /h)-arc tan(X.sub.0 /h), where X.sub.1 is a larger value and X.sub.0 is a smaller value in a set formed by a first value from said selected data and a second corresponding value selected from said reference data. 44. The interconnected system as defined in claim 35 wherein the data source comprises at least two sources of data and said first computer module includes a computer program which operates to monitor at least two sources of data separately when the at least two sources of data are uncorrelated. 45. The interconnected system as defined in claim 35 wherein the data source comprises at least two sources of data and said first computer module includes a computer program which operates to monitor all the sources of data and perform a regression analysis on the similarity angle data. 46. The interconnected system as defined in claim 45 wherein the regression analysis comprises a sequential probability ratio test. 47. The interconnected system as defined in claim 35 wherein estimated data for the monitored system is provided from preaccumulated data source based upon determining a sensor of the data source has malfunctioned. 48. The interconnected system as defined in claim 47 wherein the data source derives from a plurality of individual sources of data which provide signals selected from the group of non-linearly and linearly related signals. 49. An apparatus for determining statistical similarity between a reference system and a selected system, comprising: 50. The apparatus as defined in claim 49 wherein said computer module is further operative to conclude whether or not a deviated state exists for the selected system relative to the reference system. 51. The apparatus as defined in claim 49 wherein said first computer module is operative to compare: a first angle in a first triangle having a base opposite said first angle with a length along the base proportional to the difference between said corresponding data values comprised of a value of said current data and a value of said reference data, to a second angle in a second triangle having a base opposite said second angle with a length proportional to a range over all said corresponding data values in said reference data. 52. The apparatus as defined in claim 49 comprising a computer module operative to carry out at least one of (a) generate an estimate of said selected data based on the measure of similarity and (b) generate an estimate of said selected data based on the measure of similarity and perform a statistical hypothesis test on said selected data and said estimate to determine any statistical deviation.
	summary
	claims	1. An extreme ultraviolet light generation apparatus comprising:a chamber in which extreme ultraviolet light is generated from plasma, the plasma being generated by irradiating a target supplied into the chamber with a laser beam;a target generator configured to supply the target into the chamber as a droplet;a droplet measurement unit configured to measure the droplet supplied from the target generator into the chamber; anda shielding member configured to shield the droplet measurement unit from electromagnetic waves emitted from the plasma,the droplet measurement unit including:a light source configured to emit continuous light to the droplet;a window provided in the chamber and configured to allow the continuous light to transmit therethrough; andan optical sensor configured to receive the continuous light via the window, andthe shielding member including a shielding body provided on the chamber side with respect to the window configured to cover an optical path of the continuous light and having a light passing opening for passing the continuous light therethrough, and the shielding member is provided within the interior of the chamber, configured separately from a wall that constitutes the chamber. 2. The extreme ultraviolet light generation apparatus according to claim 1, wherein the shielding body prevents light of the electromagnetic waves emitted from the plasma from entering the droplet measurement unit. 3. The extreme ultraviolet light generation apparatus according to claim 2, wherein the shielding member further includes a shield configured to prevent noise of the electromagnetic waves emitted from the plasma from entering the droplet measurement unit. 4. The extreme ultraviolet light generation apparatus according to claim 2, wherein the droplet measurement unit further includes a transfer optical system configured to transfer an image of the droplet irradiated with the continuous light to the optical sensor. 5. The extreme ultraviolet light generation apparatus according to claim 2, wherein:the droplet measurement unit measures the droplet passing through a predetermined position in the chamber; andthe shielding body includes:a cylinder part provided between the predetermined position and the window and configured to cover the optical path of the continuous light; andan opening formed at an end of the cylinder part on the predetermined position side. 6. The extreme ultraviolet light generation apparatus according to claim 5, further comprising a gas lock mechanism configured to supply gas into the cylinder part and discharge the gas in the cylinder part from the opening, the gas lock mechanism being provided within the interior of the chamber. 7. The extreme ultraviolet light generation apparatus according to claim 6, wherein:the shielding member further includes a shield configured to prevent noise of the electromagnetic waves emitted from the plasma from entering the droplet measurement unit; andthe shield is disposed between the window and the optical sensor. 8. The extreme ultraviolet light generation apparatus according to claim 6, wherein:the shielding member further includes a shield configured to prevent noise of the electromagnetic waves emitted from the plasma from entering the droplet measurement unit; andthe shield is disposed to close a slit provided on an end of the cylinder part on the window side. 9. The extreme ultraviolet light generation apparatus according to claim 8, further comprising a heater configured to heat the shield. 10. The extreme ultraviolet light generation apparatus according to claim 5, wherein the droplet measurement unit further includes an optical filter disposed to close the opening and configured to allow the continuous light to transmit therethrough, the optical filter and the opening are provided within the interior of the chamber. 11. The extreme ultraviolet light generation apparatus according to claim 5, wherein the droplet measurement unit further includes a spatial filter disposed between the window and the optical sensor and configured to prevent passage of light not traveling on the optical path of the continuous light. 12. The extreme ultraviolet light generation apparatus according to claim 5, wherein:the droplet measurement unit outputs a passage timing signal indicating a timing at which the droplet is passing through the predetermined position; anda line filter is provided on a signal wire through which the passage timing signal outputted from the droplet measurement unit is transmitted to attenuate noise of the electromagnetic waves. 13. An extreme ultraviolet light generation apparatus comprising:a chamber in which extreme ultraviolet light is generated from plasma, the plasma being generated by irradiating a target supplied into the chamber with a laser beam;a target generator configured to supply the target into the chamber as a droplet; anda droplet measurement unit configured to measure the droplet supplied from the target generator into the chamber,the droplet measurement unit including:a light source configured to emit continuous light to the droplet;a window provided in the chamber and configured to allow the continuous light to transmit therethrough;an optical sensor configured to receive the continuous light via the window; anda transfer optical system disposed between the window and the optical sensor and configured to transfer an image of the droplet irradiated with the continuous light to the optical sensor and prevent light of electromagnetic waves emitted from the plasma from entering the optical sensor, wherein:the droplet measurement unit measures the droplet passing through a predetermined position in the chamber and outputs a passage timing signal indicating a timing at which the droplet is passing through the predetermined position; anda line filter is provided on a signal wire through which the passage timing signal outputted from the droplet measurement unit is transmitted to attenuate noise of the electromagnetic waves emitted from the plasma, and the line filter includes an analog circuit. 14. The extreme ultraviolet light generation apparatus according to claim 3, wherein:the droplet measurement unit measures the droplet passing through a predetermined position in the chamber; and the shielding body includes:a cylinder part provided between the predetermined position and the window and configured to cover the optical path of the continuous light; andan opening formed at an end of the cylinder part on the predetermined position side. 15. The extreme ultraviolet light generation apparatus according to claim 4, wherein:the droplet measurement unit measures the droplet passing through a predetermined position in the chamber; and the shielding body includes:a cylinder part provided between the predetermined position and the window and configured to cover the optical path of the continuous light; andan opening formed at an end of the cylinder part on the predetermined position side. 16. The extreme ultraviolet light generation apparatus according to claim 6, wherein:the gas lock mechanism is configured to supply gas to the interior of the cylinder part while causing the gas to contact the surface of the window toward the side of the chamber when supplying the gas to the interior of the cylinder part. 17. The extreme ultraviolet light generation apparatus according to claim 13, wherein:the line filter is a filter circuit, which is one of a low pass filter, a band pass filter, and a band eliminating filter. 18. The extreme ultraviolet light generation apparatus according to claim 13, wherein:the line filter attenuates signal components having frequencies within a range from 12 MHz to 18 MHz. 19. The extreme ultraviolet light generation apparatus according to claim 1, wherein:a shielding surface of a portion of the shielding body at which the light passing opening is formed is inclined toward the side of the optical sensor as a distance from a region where the plasma is generated becomes greater. 20. The extreme ultraviolet light generation apparatus according to claim 1, wherein:a shielding surface of a portion of the shielding body at which the light passing opening is formed is substantially parallel to a target traveling path.
040615332	summary	This invention relates to a control system for a nuclear power producing unit having a reactor in which a coolant, such as water under high pressure, is heated and circulated in parallel through a plurality of steam generators supplying steam to a prime mover such as a turbine generator. As an order of magnitude, the reactor in such a unit may have a heat output of upwards of 3,400 Mw and a net electric output of 1,200 Mw. In accordance with the invention a primary feed forward control signal corresponding to the desired or demand power output adjusts, in parallel, through separate discrete control loops, the reactor heat output required to satisfy the power demand, and the total rate of feedwater flow to and steam flow from the steam generators required to maintain critical system parameters at set point. Further in accordance with the invention the feed forward control signal to each discrete control loop is modified by the time integral of the difference between demand and actual power outputs to thereby continuously calibrate, under steady state conditions, changes in reactor heat output required to satisfy the power demand because of changes in cycle efficiency and the corresponding changes in total rate of feedwater flow to and steam flow from the steam generators required to maintain critical system parameters at set point. Further in accordance with the invention the feed forward control signal to each discrete control loop is further modified in proportion to transient changes in the difference between demand and actual power outputs and critical system parameters. Further in accordance with the invention the relative rates of feedwater flow to the steam generators are additionally adjusted in proportion to changes in the relative rates of coolant flows through the steam generators. Further in accordance with the invention the relative rates of feedwater flows to the steam generators are further adjusted in accordance with the difference in temperatures of the feedwater entering the steam generators. Further in accordance with the invention the relative rates of feedwater flows to the steam generators are additionally adjusted in accordance with the time integral of the difference between the average coolant temperatures in the steam generators. These and further objectives of the invention will be apparent as the description proceeds in connection with the drawings, in which:
	summary
058752200	abstract	A process for the production of radiostrontium consists in that a target of metallic rubidium is bombarded by a flow of accelerating charged particles. The target of irradiated rubidium is melted, whereas the extraction of radiostrontium is carried out by sorption on the surface of a sorbing material immersed into the irradiated molten metallic rubidium. As the sorbing material, use is made of materials selected from the group consisting of heat-resistant metals or metallic oxides or silicon which are inert with respect to rubidium. The resultant radiostrontium is extracted from the irradiated rubidium. The temperature of the sorbing material is selected to be close to the optimum one for the sorption of radiostrontium which is within the range of from the melting point of metallic rubidium to 220.degree. C. And the temperature of molten rubidium is selected to be close to the optimum one for the desorption of radiostrontium within the range of from 220.degree. C. to 270.degree. C.
	abstract	Separation and the like of an excised specimen from a specimen are automatically performed. Marks for improving image recognition accuracy are provided in a region that becomes an excised specimen in a specimen and a region other than said region, or in a transfer means for transferring the excised specimen and a specimen holder capable of holding the excised specimen, and the relative movement of the excised specimen and the specimen, and the like are recognized with high accuracy by image recognition. In the sampling of a minute specimen using a focused ion beam, the detection of an end point of processing for separation of the excised specimen from the specimen, and the like are automatically performed. Thus, for example, unmanned specimen excision becomes possible, and preparation of a lot of specimens becomes possible.
054266777	claims	1. A pressurizer vessel containing a liquid and steam both of which function to pressurize a reactor coolant system of a nuclear plant, the pressurizer vessel comprising: a) a heater support assembly disposed in an interior portion of the pressurizer vessel; b) a plurality of heaters mating with said heater support assembly for heating the liquid; c) a temperature detector operatively connected to said heater support assembly in a structural arrangement which measures the temperature of the liquid in the pressurizer vessel at preselected elevations; and wherein said temperature detector includes temperature measuring means for measuring a plurality of temperature readings of the liquid at preselected elevations of the liquid. a) a plurality of electrical heaters positioned in an interior portion of the pressurizer vessel for heating the liquid; b) a heater support assembly disposed in the interior portion of the pressurizer vessel and operable to receive said plurality of electrical heaters; and c) a temperature detector operatively mating to said heater support assembly for interchangeably replacing at least one of said electrical heaters which measure the temperature of the liquid; wherein said temperature detector includes temperature measuring means for measuring a plurality of temperature readings of the liquid at preselected elevations of the liquid. a) installing a plurality of electrical heaters in a plurality of receiving receptacles of a heater support assembly disposed in an interior portion of the pressurizer vessel for heating the liquid; and b) placing at least one temperature detector, having means for measuring the temperature of the liquid at preselected elevations, in a substantially vertical position in a receiving receptacle of the heater support assembly for measuring the temperature of the liquid. 2. The pressurizer vessel as in claim 1, wherein said temperature measuring means is a plurality of spaced apart thermocouples extending along said temperature detector for measuring the plurality of temperature readings. 3. The pressurizer vessel as in claim 2, wherein said temperature detector is positioned with its longitudinal length substantially perpendicular with a heater support plate for measuring the temperature of the liquid. 4. The pressurizer vessel as in claim 3, wherein said temperature detector is positioned in a heater hole for measuring the temperature of the liquid. 5. A pressurizer vessel containing a liquid and steam both of which function to pressurize a reactor coolant system of a nuclear power plant comprising: 6. The pressurizer vessel as in claim 5, wherein said temperature measuring means is a plurality of spaced apart thermocouples each extending along a longitudinal length of said temperature detector for measuring the plurality of temperature readings. 7. A method for measuring temperature of a liquid in a pressurizer vessel of a nuclear power plant comprising, which contains a liquid and steam both of which function to pressurize a reactor coolant system, the method comprising the steps of: 8. The method as in claim 7, wherein said placing of step (b) includes initially installing a temperature detector in a receiving receptacle. 9. The method as in claim 7, wherein said placing of step (b) includes replacing an installed electrical heater with a temperature detector. 10. The method as in claim 9 further comprising the step of measuring the temperature of the liquid with a plurality of spaced apart thermocouples each positioned in an interior portion of the temperature detector.
	description	The present invention relates to a core monitoring system for a nuclear plant. In a conventional core monitoring system for a nuclear plant, a TIP (Traversing Incore Probe) device is used about once every month to measure an amount of neutron while moving a detector inside a reactor in an axial direction thereof. And data of a reactor power distribution at each height position in a core axial direction are transmitted to a process computer. The process computer uses the data to maintain accuracy of the power distribution calculation performed every stipulated time period. A conventional core monitoring system is illustrated in FIG. 7 (Patent Documents 1 and 2). In FIG. 7, a TIP device drives a TIP detector 1-4 in TIP guide tubes 1-3 installed inside a core 1-2 of a reactor 1-1 to generate a TIP position signal corresponding to a travel distance when pulling the TIP detector 1-4 from the core top toward the core bottom, and reads a TIP level signal in synchronization with the TIP position signal to thereby measure neutron distribution along axial direction of the core. The TIP detector 1-4 is driven based on a drive signal (detector insertion/pull-out signal) that a process computer A and a TIP panel C output, while cooperating with each other, to a TIP drive device D. Along with loading of TIP position signal 5 and TIP level signal 6, the process computer A reads an LPRM (Local Power Range Monitoring) level signal and an APRM (Average Power Range Monitor) level signal from a neutron monitoring panel B at a predetermined timing to increase accuracy of a power distribution calculation performed in a core performance calculation or LPRM/APRM gain calibration. Further, the process computer A inputs thereto from the TIP device the TIP position signal 5 indicating a height position of the TIP detector and the TIP level signal 6 corresponding to the height position, and a time [ms] when the above signals are obtained and inputs thereto from the neutron monitoring panel B the LPRM level signal and the APRM level signal 7, plant data (feed water flow rate, pressure, temperature, etc.) and control rod position data required for heat balance calculation or power distribution calculation which are to be calculated in the core performance calculation. Patent Document 1: Japanese Patent Application Laid-Open Publication No. 2005-24383 Patent Document 2: Japanese Patent Application Laid-Open Publication No. 10-111379 Problems of the above conventional core monitoring system will be described below. First, in the conventional monitoring system, the data other than the TIP level signal are data obtained at a timing of data collection periodically performed by the process computer A, so that an error due to time lag exists between each of the collected datum. That is, in the conventional core monitoring system, the TIP level signal 6 and time data indicating a time at which the level signal is read are transmitted from the TIP panel C, whereas the LPRM/APRM level signals 7 from the neutron monitoring panel B are not attached with the time data. Thus, the process computer A periodically monitors the input of the TIP position signal 5. And the process computer A loads, based on the input TIP position signal 5, the APRM level signal measured at the core top, the core center, and the core bottom and LPRM level signal 7 measured at the LPRM height. Therefore, due to signal transmission delay between the TIP panel C and process computer A or input detection delay of the TIP position signal 5 in the process computer A, a time lag occurs between the TIP level signal 6 and the LPRM/APRM level signals 7 to be acquired at the same timing as the TIP position signal 5, resulting in occurrence of an error in the power distribution calculation. Secondly, in the power distribution calculation to be performed in the core performance calculation, the latest data are used to calculate the heat balance of, e.g., thermal output or the plant data, so that in the power distribution calculation of TIP learning using TIP data at a time of execution of TIP or LPRM learning after LPRM calibration, a time lag occurs between the TIP data and the plant data, resulting in occurrence of an error in the power distribution calculation. Thirdly, in a case where an instantaneous value is used without change as the LPRM level signal and the APRM level signal 7 to be input from the neutron monitoring panel B, an error is superposed on the LPRM/APRM level signals 7 during execution of the TIP due to fluctuation of the plant, leading to accuracy degradation of the power distribution calculation. The neutron monitoring panel B uses the LPRM/APRM level signals 7 in a plant interlock such as scram, so that the averaging processing cannot be applied. Fourthly, the control rod position data used in the power distribution calculation is not attached with the time data, so that the TIP data, the LPRM/APRM level signals, the plant data (heat balance to be calculated from the plant data), and the control rod position data are not time-consistent with each other. It follows that when a control rod is operated at an exertion time of the power distribution calculation, control rod data before the operation may be used, depending on the timing, in the power distribution calculation, which may degrade accuracy of the calculation. Therefore, the control rod cannot be operated from several minutes before activation of the punctual power distribution calculation until completion thereof, and an operator needs to confirm not only an operation state of the control rod but also the time. Fifthly, in the conventional core monitoring system, all the data (TIP data, the LPRM/APRM level signals, the plant data (heat balance to be calculated from the plant data), the control rod position data) required in the core performance calculation are not time-consistent with each other, so that the power distribution calculation using past data cannot be performed, and an analysis has been made based on predictive calculation using certain reasonable values. The present invention has been made to solve the above problems, and an object thereof is to provide a core monitoring system capable of performing accurate power distribution calculation by preventing an error due to a time lag from occurring between the data to be used in the calculation. According to an embodiment, there is provided a core monitoring system comprising: a TIP (Traversing Incore Probe) that measures an amount of neutrons in a reactor core; a TIP drive device; a TIP panel; a neutron monitoring panel; and a process computer, the TIP panel including a TIP level processing section that processes a TIP level signal input from the TIP drive device, a TIP position processing section that processes a TIP position signal input from the TIP drive device, a TIP panel time setting section that sets a TIP panel time for synchronizing the TIP level signal and TIP position signal, and a TIP level data storage section that stores synchronized TIP level data, the neutron monitoring panel including a neutron monitoring panel time setting section that sets a neutron monitoring panel time corresponding to a collection time of a LPRM (local power range monitoring) level signal and an APRM (average power range monitor) level signal, the process computer including a TIP level data database that compares a TIP time attached to the TIP level data transmitted from the TIP panel and a nuclear implementation console time attached to the LPRM level signal and the APRM level signal transmitted from the nuclear implementation console and stores the TIP level data and the LPRM level signal and the APRM level signal corresponding in time to each other and a core performance calculating section that calculates core performance based on the data stored in the TIP level data database. According to the present invention, the core power distribution calculation can be performed with high accuracy by preventing an error due to a time lag from occurring between the data to be used in the calculation. Embodiments of a core monitoring system according to the present invention will be described below with reference to the accompanying drawings. A core monitoring system according to a first embodiment will be described with reference to FIGS. 1 and 2. (Configuration) The core monitoring system according to the first embodiment includes a process computer A, a neutron monitoring panel B, a TIP panel C, a TIP drive device D, and a data transmitting device E. The process computer A includes: an input unit 2-1 thereof that an operator operates in an interactive manner; a TIP scanning section 2-2 that performs TIP scan according to a TIP scan request from the operator; a TIP level data receiving section 2-3 that receives a signal from the neutron monitoring panel B, and TIP level data stored in the TIP panel C and a current TIP position signal through the data transmitting device E, and synchronizes the received signals; a TIP level data DB 2-4 that stores the TIP level data; and a core performance calculating section 2-13. The neutron monitoring panel B includes a time setting section 2-7 that sets a time in the LPRM level signal and the APRM level signal 7 so as to synchronize the signals in the process computer A and transmits the LPRM level signal and the APRM level signal 7 and the time data set in the time setting section 2-7 to the process computer A. The TIP panel C includes a TIP level processing section 2-5 that inputs the TIP level signal 6 from the TIP drive device D; a TIP position processing section 2-6 that inputs the TIP position signal 5 from the TIP drive device D; a TIP level data storing section 2-8 that inputs the TIP level signal 6 in synchronization with the TIP position signal 5 and stores TIP level data 8; a TIP level data transmitting section 2-9 that transmits the TIP level data 8 to the process computer A through the data transmitting device E, and a time setting section 2-14 that sets the time data in the TIP level signal 6 and the TIP position signal 5. The data transmitting device E passes the TIP level data 8 and the TIP position signal 5 stored in the TIP panel C from the TIP panel C to the process computer A. (Operation) Operation of thus configured core monitoring system according to the present embodiment will be described below. The input unit 2-1 is provided in the process computer A, and the operator uses this device to request execution of the TIP. The request from the operator is processed in the TIP scanning section 2-2 of the process computer A to activate the TIP drive device D through the TIP panel C, thereby driving a TIP detector. An amount of neutrons measured by the TIP detector is input to the TIP level data storing section 2-8 as the TIP level signal 6 through the TIP level processing section 2-5 of the TIP panel C. A height position of the TIP detector moved from the core top to the core bottom is input, as the TIP position signal 5 generated when the TIP detector is inserted from the core bottom into the core top and when the TIP detector is pulled from the core top to the core bottom, to the TIP level data storing section 2-8 through the TIP position processing section 2-6 of the TIP panel 6. Further, in order to synchronize the TIP level signal 6 and the TIP position signal 5, times at which the respective signals are collected are set in the time setting section 2-14 and input to the TIP level data storing section 2-8. The TIP level data storing section 2-8 of the TIP panel C synchronizes the TIP position signal 5 and the TIP level signal 6 based on the times set in the time setting section 2-14 and stores therein the synchronized data as the TIP level data 8. The TIP level data 8 stored in the TIP level data storing section 2-8 is passed from the TIP level data transmitting section 2-9 to the TIP level data receiving section 2-3 of the process computer A through the data transmitting device E at a scan completion time of each string. The TIP position signal 5 stored in the TIP level data storing section 2-8 is constantly passed from the TIP level data transmitting section 2-9 to the TIP level data receiving section 2-3 of the process computer A through the data transmitting device E. The time setting section 2-7 of the neutron monitoring panel B sets times at which the LPRM level signal and the APRM level signal are collected for synchronization to be achieved in the process computer A. Further, the LPRM level signal, the APRM level signal, and the time data of these signals set in the time setting section 2-7 are constantly transmitted to the TIP level data receiving section 2-3 of the process computer A. As described above, the TIP level data receiving section 2-3 of the process computer A reads the LPRM level signal, the APRM level signal 7 and the time data of the respective signals from the neutron monitoring panel in addition to the TIP level data 8 input from the transmitting device E. Further, the TIP level data receiving section 2-3 determines whether the TIP detector is situated at the core bottom, the core center, or the core top based on the current TIP position signal and retrieves the time data corresponding to each point. The TIP level data receiving section 2-3 compares the retrieved the time data with the time data of the LPRM/APRM level signals 7 transmitted from the neutron monitoring panel B and stores the LPRM level signal and the APRM level signal 7a corresponding to the retrieved time data in the TIP level data DB 2-4. The above determination processing is repeatedly performed at a high speed, and the LPRM level and the APRM level stored in the TIP level data DB 2-4 are stored for all channels for each TIP detector (machine). An example of a time synchronization process flow (S1 to S8) between the TIP level data 8 from the TIP panel C, and the LPRM level signal and the APRM level signal 7 from the neutron monitoring panel B, which is performed in the TIP level data receiving section 2-3, is illustrated in FIG. 2. FIG. 2 illustrates a process flow of a TIP level data receiving section according to the first embodiment. As illustrated in FIG. 2, a process (S4 and S7) of storing the LPRM/APRM signal corresponding in time to the TIP data measured by the TIP detectors at the various height positions in the TIP level data DB 2-4 is repeatedly performed at a high speed. The measurement data stored in the TIP level data DB 2-4 of the process computer Are displayed on a display 2-11 by the TIP scanning section 2-2, printed as a log by a printer 2-12, and passed to the core performance calculating section 2-13 for the power distribution calculation. (Effects) As described above, the TIP level signal 6, the TIP position signal 5, and the time data are collected while they are synchronized with each other in the TIP level data storing section 2-8 of the TIP panel C, the TIP position signal 5 and the TIP level data 8 are constantly transmitted from the TIP panel C to the process computer A, and the LPRM level signal, the APRM level signal, and the time data are constantly transmitted from the neutron monitoring panel B to the process computer A. This allows achievement of the time synchronization based on the time data in the process computer A, which in turn allows accurate measurement free from signal transmission delay or signal detection delay. As a result, the LPRM level signal and the APRM level signal essential for the power distribution calculation can be obtained with an accuracy equivalent to or better than ever before. As described above, according to this first embodiment, the LPRM level signal and the APRM level signal are transmitted to the process computer with a function of adding the time data to the LPRM level signal and the APRM level signal provided, and the process computer retrieves, based on the TIP level signal and the time data, the LPRM/APRM data corresponding in time to the TIP level signal. This allows the power distribution calculation to be free from signal transmission delay or signal detection delay, thereby increasing accuracy of the power distribution calculation. A core monitoring system according to a second embodiment will be described with reference to FIGS. 3 and 4. The same reference numerals are given to the same components as those in the first embodiment, and the repeated description will be omitted. (Configuration) A core monitoring system according to a second embodiment includes, in addition to the components of the core monitoring system of the first embodiment, a plant data collector/calculator F that transmits the plant data 9 such as flow rate, reactor pressure, and temperature to the process computer A together with time at which the plant data 9 is collected which is set using a time setting section 2-20, a heat balance calculating section 2-17 that performs heat balance calculation, and a heat balance data section 2-18 that stores a result of the heat balance calculation and passes the result to the core performance calculating section 2-13. The heat balance calculating section 2-17 and the heat balance data section 2-18 are provided in the process computer A. (Operation) Operation of thus configured core monitoring system will be described below. The input unit 2-1 is provided in the process computer A, and the operator uses this device to request execution of the TIP. The request from the operator is processed in the TIP scanning section 2-2 of the process computer A to activate the TIP drive device D through the TIP panel C, thereby driving the TIP detector. The plant data collector/calculator F collects the plant data 9 such as flow rate, reactor pressure, and temperature and constantly transmits the collected the plant data 9 to the heat balance calculating section 2-17 together with the time at which the plant data 9 is collected. The heat balance calculating section 2-17 periodically calculates heat balance based on the plant data 9 transmitted from the plant data collector/calculator F. The plant data 9 and a result of the heat balance calculation are transmitted from the heat balance data section 2-18 to the core performance calculating section 2-13. A set of the plant data 9, the result of the heat balance calculated based on the plant data 9, and the time at which the plant data 9 are collected are referred to as “heat balance data”. For the TIP data, TIP data synchronized based on the time data are stored in the TIP level data DB 2-4. The stored TIP data is transmitted to the core performance calculating section 2-13 through the TIP scanning section 2-2. The core performance calculating section 2-13 searches for the time-corresponding data based on the time data of the transmitted the heat balance data and the time data retained in the TIP level data DB 2-4, makes TIP detectors scan in the same guide tube (string) to make outputs of the TIP detectors coincide with each other, and calculates machine normalization factors, thereby performing the power distribution calculation. In the calculation of the machine normalization factors, an average value of the APRM signals 6 at the core top, the core center, and the core bottom is used based on an APRM calibration factor calculated from the time-corresponding the heat balance data and the time-corresponding TIP level data DB 2-4. This advantageously eliminates the use of data collected at different times, so that an error due to a time lag is not included in the calculation. A flow of time synchronization process (Steps S1 to S5) in the core performance calculating section 2-13 is illustrated in FIG. 4. As illustrated in FIG. 4, after the TIP scan has been completed, the core performance calculating section 2-13 determines whether the operator has issued a scan termination request. When there has been issued the scan termination request from the operator (Step S1), the core performance calculating section 2-13 retrieves the time data at which the last scan of the string is terminated from the TIP level data DB 2-4 (Step S2), retrieves the heat balance data corresponding to the retrieved the time data from the heat balance data section 2-18 (Step S3), and performs the power distribution calculation (Step S5). This allows data immediately after completion of the TIP to be used in the power distribution calculation, thereby preventing occurrence of an error due to fluctuations of the plant data 9 from the termination of the TIP to start of the power distribution calculation. In the absence of the scan termination request from the operator, the core performance calculating section 2-13 performs punctual or on-demand power distribution calculation, and thus, retrieves the heat balance data at a request time (Step S4) so as to perform the power distribution calculation (Step S5). (Effects) As described above, the core performance calculating section 2-13 in the process computer A performs the time synchronization processing between the TIP level data and the heat balance data, thereby performing accurate power distribution calculation in a manner correctly reflecting the power distribution and the plant data 9 at the time of termination of the TIP. As described above, according to the second embodiment, there are added a function that adds the time data to the plant data and a function that searches the TIP data at the time of execution of the TIP, the plant data and the heat balance data used in the TIP learning and LPRM learning by using time as a search key to retrieve the time-corresponding data in the core performance calculation. This allows achievement of the time synchronization among the data to be used in the power distribution calculation, thereby increasing accuracy of the power distribution calculation. A core monitoring system according to a third embodiment will be described with reference to FIG. 5. The same reference numerals are given to the same components as those in the first and second embodiments, and the repeated description will be omitted. (Configuration) A core monitoring system according to a third embodiment additionally includes, in the process computer A of the second embodiment, an LPRM/APRM level averaging section 2-19 that suppresses fluctuations of the LPRM/APRM levels due to fluctuation of the plant. (Operation) Operation of thus configured core monitoring system will be described below. The LPRM/APRM level averaging section 2-19 can set a filtering constant for suppressing the fluctuation and a sampling period for the LPRM level signal and the APRM level signal 7 from the neutron monitoring panel B, performs averaging according to the filtering constant and the sampling period that have been set, and transmits the resultant signals to the TIP level data receiving section 2-3. In addition to the above-mentioned averaging, averaging at an arbitrary period and with an arbitrary number of collected data, such as one-minute averaging (using 12 points with intervals of 5 seconds therebetween) or 15 second averaging (using 3 points with intervals of 5 seconds therebetween) can be adopted. The TIP level data receiving section 2-3 determines, based on the current TIP position signal, whether the TIP detectors have been moved to be situated at the core bottom, the core center, and the core top, respectively, and retrieves the time data at respective points. Then, the TIP level data receiving section 2-3 compares the retrieved the time data and the time data of the averaged LPRM level signal and APRM level signal 7a transmitted from the LPRM/APRM level averaging section 2-19 and stores the averaged LPRM level signal and APRM level signal 7 corresponding to the retrieved the time data in the TIP level data DB 2-4. The core performance calculating section 2-13 reads the averaged LPRM level signal and APRM level signal 7a and a heat balance calculation result from the TIP level data DB 2-4 and the heat balance data 2-18, respectively and performs the power distribution calculation, thereby preventing an error due to fluctuation of the plant. (Effects) As described above, the process computer A has a function of averaging the LPRM/APRM levels and thus it is possible to suppress fluctuations of the value of the LPRM/APRM levels due to fluctuation of plant without influencing a function that uses the LPRM/APRM level signals in a plant interlock such as scram, thereby performing the power distribution calculation with high accuracy. As described above, according to the third embodiment, adding the averaging function to the process computer so as to suppress fluctuation can eliminate an error due to the fluctuation without exerting influence on the plant interlock such as a scram that uses the LPRM/APRM level signals, thereby performing the power distribution calculation with high accuracy. A core monitoring system according to a fourth embodiment will be described with reference to FIG. 6. The same reference numerals are given to the same components as those in the first to third embodiments, and the repeated description will be omitted. (Configuration) A core monitoring system according to the fourth embodiment additionally includes, in the configuration of the third embodiment, a control rod position information device G provided with the control rod position data 10 and a time setting section 2-21 so as to set the time data also for the control rod position data 10. (Operation) Operation of the thus configured core monitoring system will be described below. The plant data collector/calculator F collects the plant data (flow rate, reactor pressure, and temperature) 9, sets time at which the plant data have been collected, and constantly transmits the resultant collected the plant data to the heat balance calculating section 2-17. The control rod position information device G collects the control rod position data 10, sets time at which the control rod position data 10 has been collected, and constantly transmits the resultant data to the heat balance calculating section 2-17. Based on the received the plant data and the control rod data, the heat balance calculating section 2-17 performs the heat balance calculation using the data corresponding to periodical calculation timing and stores a result of the calculation in the heat balance data section 2-18. The core performance calculating section 2-13 searches the heat balance data section 2-18 for data at the execution time of the periodical power distribution calculation and performs calculation of the retrieved data. In the power distribution calculation after execution of the TIP, the core performance calculating section 2-13 searches for the time-corresponding data based on the time data of the heat balance data and the time data retained in the TIP level data DB 2-4 and performs the power distribution calculation using the time-corresponding data. This allows achievement of the time synchronization among all the data to be used in the power distribution calculation, thereby increasing accuracy of the core performance calculation. (Effects) As described above, in addition to synchronization of the TIP level data and the heat balance data, synchronization of the control rod data can be achieved by the time setting section of the control rod position information device G, thereby performing accurate power distribution calculation in a manner correctly reflecting the power distribution and the plant data at the time of termination of the TIP. As described above, according to the fourth embodiment, there is provided a function of adding the time data to the control rod data, and the control rod data added with the time data is transferred to the process computer and whereby time consistency among all the data (TIP data, the LPRM/APRM level signals, the plant data, the heat balance data calculated from the plant data, the control rod position data) required for the core performance calculation can be achieved in the process computer, thereby increasing accuracy of the power distribution calculation. While certain embodiments have been described, these embodiments have been presented by way of example only, and are not intended to limit the scope of inventions. Indeed, the novel methods and systems described herein may be embodied in a variety of other forms; furthermore, various omissions, substitutions and changes in the form of the methods and systems described herein may be made without departing from the spirit of the inventions. The accompanying claims and their equivalents are intended to cover such forms or modifications as would fall within the scope and spirit of the inventions. A process computer B neutron monitoring panel C TIP panel D TIP drive device E data transmitting device F plant data collector/calculator G control rod position information device 1-1 reactor 1-2 core 1-3 TIP guide tubes 1-4 TIP detector 2-1 input unit 2-2 TIP scanning section 2-3 TIP level data receiving section 2-4 TIP level data DB 2-5 TIP level processing section 2-6 TIP position processing section 2-7 time setting section 2-8 TIP level data storing section 2-9 TIP level data transmitting section 2-11 display 2-12 printer 2-13 core performance calculating section 2-14 time setting section 2-17 heat balance calculating section 2-18 heat balance data section 2-19 LPRM/APRM level averaging section 2-20 time setting section 2-21 time setting section 5 TIP position signal 6 TIP level signal 7 LPRM/APRM level signal 8 TIP level data 9 plant data 10 control rod position data
044341300	abstract	A fusion reaction system wherein a compressed spiral beam of electrons forms a cylindrical electron sheath and wherein oppositely directed cylindrical beams of fusible ions are projected through said electron sheath and are forced into a common thin cylindrical path located where the potential gradient in electron sheath is minimum.
	description	This application is a National Stage of International patent application PCT/EP2011/057286, filed on May 6, 2011, which claims priority to foreign French patent application No. FR 1053653, filed on May 11, 2010, the disclosures of which are incorporated by reference in their entirety. The field of the invention is that of the determination (or characterization) of so-called void rate in a biphase gas/liquid medium, corresponding to the determination (or characterization) of volume of gas bubbles within a volume of liquid. There exist numerous fields in which the determination of void rate is of interest and notably in fields as varied as the nuclear industry, the food-processing industry, the oil industry, the chemical industry, cryogenic applications, medicine (imaging and problems of decompression sickness) or else the field of underwater acoustics. More precisely in the nuclear field, and notably for the fourth generation of fast neutron nuclear reactor (or “FBR” standing for “Fast Breeder Reactor”), the SFR (“Sodium Fast Reactor”) reactor appears very promising. This family of reactors presents several challenges, in particular from the point of view of improving monitoring. Among the checks to be performed in the vessel of SFRs, there is one which was not taken into account within the framework of the development and exploitation of the Phenix and Superphenix SFRs: measurement of the continuous engassment of the primary sodium. More precisely, “engassment” is defined as the presence of gas in a liquid (or a solid) in the form of bubbles (free gas). The gases dissolved in the liquid phase are not considered as belonging to the gas volume but as belonging to the liquid volume, when evaluating a so-called void rate. Generally, an “FBR” is a reactor whose core is not moderated. Fast spectrum operation presents a certain number of advantages such as the possibility of implementing supergeneration or transmutation of minor actinides but it requires the use of a heat-exchanging fluid with low neutron capture cross-section such as liquid sodium. FIG. 1 illustrates the diagram of such a type of reactor according to the known art. Indeed, liquid sodium possesses the properties expected of a heat-exchanging fluid, namely good thermal properties, low noxiousness, low cost etc. Its main drawbacks are its reactivity to air and especially to water and its opacity which renders the inspectability of the reactors more difficult than in water. The gas present in free form in the liquid sodium of SFRs can have diverse origins and be of diverse kinds. There exist two possible sites of existence of gas bubbles in the sodium: the primary circuit (the main vessel in which the core is immersed) and the secondary circuit (circuit of the exchangers). SFR type reactors use liquid sodium as heat-exchanging fluid. This fluid phase, present in the primary vessel of the reactor, circulates through the core, the pumps and the exchangers so as to extract the heat emanating from nuclear fission. This sodium pool is surmounted by a cover gas, also called a pile headspace (generally argon). Ideally, this liquid sodium is perfectly pure and monophase. In reality, this is not the case: in addition to comprising a few impurities and dissolved gases, the sodium continuously comprises bubbles of free gas. This continuous engassment nevertheless presents several negative consequences and notably the presence of bubbles in a liquid which very greatly modifies its acoustic properties (speed, attenuation, diffusion, nonlinear properties, etc.). The deployment of acoustic measurement procedures for continuous monitoring, which is performed at the nominal power (measurement of the displacement of assembly heads, ultrasound thermometry based on flight time measurement), or the periodic checks operating in the shutdown regime (ultrasound telemetry, surface metrology, volume checking, etc.) requires a knowledge of an order of magnitude of the attenuation coefficients, so as to prove a priori that the amplitude of the signal is sufficient, as well as an order of magnitude of the lack of homogeneity of the spatial distribution, to prove that the speed calibrations carried out some distance from the effective measurement point, remain usable. The aforementioned measurements thus necessarily demand a knowledge of the void rate value, backed up if appropriate with certain data relating to the histogram of the radii of the bubbles (at least the bounds). If the evaluated void rate is not in itself directly deleterious in relation to the operation of the core, it is indirectly so if it participates in the generation of gas pockets at high points of the submerged structures. The characterization of the continuous primary engassment in a reactor can thus serve as input data for trials or calculations for the formation and relaxation of these gas pockets. It must be pointed out that the abrupt relaxation of accumulated pockets of gas formed part of the scenarios envisaged for explaining the series of emergency shutdowns on reactors that has operated in the past. Moreover, the continuous tracking of the engassment rate seems necessary for controlling the non-exceeding of several thresholds and notably: the neutronic perturbation threshold (a priori too high to be attainable under the normal operating conditions of the reactor: of the order of several percent); the blinding threshold of the systems for measuring activity in the pile headspace. Currently, in order to determine void rates, optical techniques are usable in translucent liquids, but these are no longer transposable into opaque media such as liquid sodium. Linear acoustic techniques based on the attenuation or the spreading of an acoustic wave are usable but exhibit an ambiguity—between resonant bubble and big bubble—which is impossible to resolve without a priori knowledge about the bubble cloud. For certain ranges of void rate and of bubble sizes, speed measurements are sometimes implemented to determine the void rate. In this context and to address the problematic issue of determining void rate applicable in a gas/liquid biphase non-translucent medium and notably of possibly very low void rate, typically of the order of 10−6 to 10−8, the present invention proposes to exploit the so-called “NRUS” procedure, frequently encountered in the literature under the acronym standing for “Nonlinear Resonant Ultrasound Spectroscopy”. It is a nonlinear acoustics technique used mainly in the field of the acoustics of solids. Generally, a mechanical system possesses resonant modes, all associated with a natural resonant frequency. As a general rule, these resonant frequencies are dependent on the geometric characteristics and the speed of the waves in the medium constituting the system. Now, the speed in a medium is dependent on its density and its compressibility. In the case of nonlinear acoustics, the modulus of elasticity is not constant and is dependent on the applied stress. It follows from this that the resonant frequency of a nonlinear mechanical system varies as a function of the applied stress and therefore of the acoustic excitation amplitude. Nonlinear resonant ultrasound spectroscopy consists in observing this type of phenomenon by exciting the mechanical system considered while performing a frequency scan at various amplitudes. A shift between the resonance peaks then appears. The authors K. Van Den Abeele et al. have notably proposed in the article, “On the quasi-analytic treatment of hysteretic nonlinear response in elastic wave propagation”—J. Acoust. Soc. Am. 101 (4), April 1997 1885-1898, the following model of the nonlinear modulus of elasticity: K ⁡ ( ɛ , ∂ ɛ ∂ t ) = K 0 ⁡ [ 1 + βɛ + α ⁡ ( Δɛ + sign ⁡ ( ∂ ɛ ∂ t ) ⁢ ɛ ) ] With β the conventional nonlinear parameter and α the nonconventional nonlinear parameter, ε being the instantaneous strain and Δε the amplitude of the strain. If f0 is the linear resonant frequency of a mechanical system (measured at low amplitudes) and f the resonant frequency measured for waves of larger amplitude and by considering the parameter α to be strongly predominant over the parameter β, this seeming to be confirmed by the experiments described in the articles by K. Van Den Abeele et al.—Nonlinear ElasticWave Spectroscopy (NEWS) Techniques to Discern Material Damage, Part I: Nonlinear Wave Modulation Spectroscopy (NWMS), Part II: Single Mode Nonlinear Resonance Acoustic Spectroscopy—Res Nondestr Eval (2000) 12: 17-42 or Micro-damage diagnostics using nonlinear elastic wave spectroscopy (NEWS)—NDT&E International 34 (2001) 239-248, the following relation is obtained: f 0 - f f 0 ≈ αΔɛ The NRUS procedure consists in measuring a frequency shift which turns out to be proportional to the nonconventional nonlinear parameter by frequency scanning. The frequency shift observed is a fast dynamics phenomenon. The field of nonlinear elasticity of materials and notably of materials such as rocks has already been explored for a long time, but the technique today called NRUS began to be studied in depth and exploited for the characterization of media really toward the middle of the 1990s. An NRUS procedure for characterizing damage to materials has notably been described in U.S. Pat. No. 6,330,827. This patent entails applying the NRUS procedure and deducing, from the frequency shift, damage to the material tested. The article by M. Muller et al.—Nonlinear resonant ultrasound spectroscopy (NRUS) applied to damage assesment in bone—J. Acous. Soc. Am., Vol. 118(6), p. 3946-3952, December. 2005, presents another interesting application of the NRUS technique: the detection of fractures in bones. Spectroscopy of a healthy bone exhibits a constancy of the resonant frequency whereas a fractured bone exhibits a frequency shift. The use of the RNUS procedure for detecting defects in materials has also been described in the article by Payan et al.: “Applying nonlinear resonant ultrasound spectroscopy to improving thermal damage assessment in concrete”, 13 Mar. 2007. Thus, according to the known art, the procedures of NRUS type are employed to detect defects constituting discontinuities which are the source of nonlinearities in solid media. Concerning biphase media, the inventors have mentioned in a publication: Cavaro M. ET AL: “Towards in-service acoustic characterization of gaseous microbubbles applied to liquid sodium” 2009 1ST INTERNATIONAL CONFERENCE ON ADVANCEMENTS IN NUCLEAR INSTRUMENTATION MEASUREMENTS METHODS AND THEIR APPLICATIONS” XP031704404, the possibility of using acoustic nonlinearities notably to detect bubble sizes by virtue of the presence of two acoustic wave transducers emitting in a biphase medium, a first transducer emitting an acoustic wave at a fixed first frequency f1, a second transducer emitting an acoustic wave at a variable second frequency f2. The bubbles present in the bubbly medium generate an acoustic wave with a frequency difference Δ(f1−f2) detected by a hydrophone, frequency scanning thus making it possible to detect various frequency differences and thus various sizes of bubble. In this article, the authors mention the possibility of using the procedure of RNUS type but without proposing any solution making it possible to implement such a procedure and to do so in order to determine a void rate in a biphase medium. This is why, in this context the subject of the present invention is a method of determining void rate exploiting the use of a bulk elastic wave resonator. More precisely the subject of the present invention is a method of determining the void rate in a biphase gas/liquid medium, corresponding to the volume fraction of gas corresponding to the presence of bubbles in the liquid medium in a total volume of gas and liquid, characterized in that it comprises the following steps: the deployment of a bulk elastic wave resonator in contact and coupled acoustically with the biphase medium; the measurement by nonlinear resonant ultrasound spectroscopy of the biphase medium comprising the scanning in terms of frequencies and amplitudes of acoustic excitation in a given range of frequencies and in a given range of amplitudes, of bulk elastic waves emitted and detected at said resonator and leading to the obtaining of a set of resonance curves exhibiting maxima; the determination of a straight line defined by the set of maxima of said resonance curves for different excitation amplitudes and of the slope (α) of said straight line; the determination of the void rate on the basis of said slope. According to a variant of the invention, the method comprises a prior step of determining the resonant frequency of said gas bubbles. According to a variant of the invention, the prior step of determining the size of the biggest bubbles present—so as to choose a resonator whose resonant frequency is appropriate—is performed by optical measurement. According to a variant of the invention, the bubbles having a radius of the order of a hundred microns, the frequency scan by the implementation of the NRUS procedure is performed below 33 KHz, the resonant frequency of 100-micron air bubbles in water. According to a variant of the invention, the bulk wave resonator comprises a first metallic plate connected to an emitter, a second metallic plate connected to a receiver. According to a variant of the invention, the first plate is connected to a transducer. According to a variant of the invention, the second plate is connected to a hydrophone. According to a variant of the invention, the bulk wave resonator is of Helmholtz type. According to a variant of the invention, the liquid is a metal in the liquid state. According to a variant of the invention, the liquid is sodium. The subject of the invention is also the application of the method of determining the void rate in a biphase gas/liquid medium according to the invention, in a nuclear reactor so as to determine a void rate in a heat-exchanging liquid. According to a variant of the invention, the nuclear reactor is a fast neutron reactor. According to a variant of the invention, the resonator is of plate type and is placed in the primary circuit of the reactor. According to a variant of the invention, the resonator is of Helmoltz resonator type and is placed branched off from the primary circuit or from the secondary circuit of the reactor. The method of determining void rate of the invention is applied in a liquid/gas biphase medium, the liquid possibly being notably a heat-exchanging fluid laden with gas bubbles or indeed with air bubbles. The bubbles present in said liquid generate acoustics-related nonlinearities and the nonlinearity coefficient is determined via resonant spectroscopy, that is to say a scan in terms of frequency and amplitude. The nonlinearities related to the presence of bubbles in the liquid medium induce a frequency shift. It is this shift which makes it possible to deduce a nonlinearity coefficient. According to the known art, the problematic issue of the propagation of an acoustic wave in a bubbly liquid is often treated in a linear manner. This approach is valid for low-amplitude oscillations. When it is applicable, it makes it possible moreover to establish a good approximation of the speeds, attenuations, diffusions etc. in biphase media. Like any mechanical system, a bubble exhibits a resonant frequency. Let us consider a spherical air bubble in a volume of water. This system possesses, on account of the respective compressibilities of the air and of the water, an infinity of degrees of freedom and therefore an infinity of natural modes of oscillation. Nonetheless an air bubble in water possesses a particular feature: the presence of a fundamental radial resonance lying at a frequency corresponding to wavelengths in air and in water which are very large compared with the size of the bubble. This fundamental radial mode is in general the only one considered when one speaks of resonance of bubbles since it is very strongly predominant from an energy point of view. It is possible to envisage just the spherically symmetric volume pulsations of the bubbles (mode 0) only when k.a<<1. Only mode 0 gives rise to volume variations as described in the article by A. Bouakaz, P. Palanchon, N. de Jong—Dynamique de la microbulle [Dynamics of the microbubble]—Chapter on Contrast Echography, Springer, 2007. The frequency corresponding to the fundamental mode of resonance of a bubble is named the Minnaert frequency as described in the article—On musical air bubbles and the sound of running water—Phil. Mag., 16(7), p. 235 to 248, 1933. Minnaert resonance involves large acoustic wavelengths compared with the radius of the bubbles. The vibratory phenomenon of the bubbles in a fluid can then be considered to be a harmonic oscillator model (mass-spring system, the mass being constituted by the fluid surrounding the bubble and the spring by the compressible gas of the bubble). The Minnaert model, the oldest describing the resonant frequency of a bubble of radius R is as follows (let us point out that in many cases, this simple linear model suffices): f res = 1 2 ⁢ π ⁢ ⁢ R ⁢ 3 ⁢ γ ⁢ ⁢ p 0 ρ l with: y=isentropic exponent of the gas (no unit) p0=static pressure (Pa) p1=density of the liquid (kg·m−3) For the air-water system, this gives a coefficient of:frequency*radius=3.29 SI This therefore constitutes the criterion for choosing the frequency of the resonator. The presence of bubbles will then cause the resonant frequency of the resonator to decrease, in a manner dependent on the void rate. Having no a priori knowledge of the void rate (since one is seeking to measure it), it must therefore be performed from 0 to the frequency calculated previously+about 20%. Example of Determining Rate of Air Bubbles in an Aqueous Medium: For a cloud of bubbles therefore the bubbles exhibit a radius of less than 100 microns, the resonant frequency of the biggest bubbles is 333 kHZ. A much smaller resonant frequency of the resonator will be chosen: (divided by about 5) i.e. about 6 kHz. The frequency scan by implementing the NRUS procedure can thus be performed from 0 to 7.2 kHz (6 kHz+20%). By way of illustration of the method of the invention, FIG. 2 shows diagrammatically how to determine via a set of curves giving the amplitude of the bulk waves analyzed according to a frequency scan, and via the maxima of said curves, the coefficient α making it possible to determine a void rate of the medium, (measurements performed by virtue of a plate resonator whose resonant frequency lies at round about 15 kHz). It should be noted that the method of the invention makes it possible to access a wide range of measurable void rates, makes it possible to avoid measurement ambiguity in contradistinction to the linear procedures and, the nonlinearities of the bubbles very strongly dominating all the other nonlinearities present (electronic, water, etc.), the latter then become negligible. First Exemplary Application of a Device Allowing the Implementation of the Method of the Invention, Positioned in a First Location of a Nuclear Reactor Primary Circuit The method of the present invention makes it possible to determine the rate of bubbles present in a liquid sodium heat-exchanging fluid. A biphase fluid sheet is set resonating by virtue of the use of a resonator consisting of two plates, as illustrated in FIG. 3, that may typically be spaced a few tens of millimeters apart. A first plate is connected to an emitter, a second plate is connected to a hydrophone. The emitter is a transducer emitting for example at 100 kHz. A signals generator 1 feeds a power amplifier 2, connected to a transducer capable of generating bulk elastic waves 10, the bulk wave resonator is formed by two plates 20. On output the bulk waves are sensed by a low-frequency hydrophone 30, linked for analysis to a charge amplifier 3, coupled to an oscilloscope 4. Second Exemplary Application of a Device Allowing the Implementation of the Method of the Invention, Positioned in a Second Location Branching off from the Primary or Secondary Circuit of a Nuclear Reactor. A resonator of Helmoltz type is defined in the following manner: 1—a closed cavity of volume V which communicates with the exterior by way of a small tube of length L and of cross-section A; 2—the aforementioned dimensions are small compared with the length of the acoustic waves considered. A Helmholtz resonator makes it possible to obtain an acoustic resonance at the resonant frequency of the Helmholtz resonator defined by: f = c 2 ⁢ π ⁢ A VL With A, V and L the aforementioned dimensions and c the acoustic speed in the medium considered. As illustrated in FIG. 4, an emitter 11 is placed at an end of the resonator of Helmoltz type, so as to generate bulk waves, received at acoustic receivers 31 after resonance in the cavity 21.
	summary
	claims	1. A method for determining bus performance characteristics of a system bus, comprising, during execution of a program which generates data exchanges:(a) measuring bus performance of the system bus, comprising:(i) capturing events indicative of the data exchanges by at least one interface monitor between at least two devices of the system via a system bus of the system;(ii) interpreting the captured events and calculating performance metrics for those captured events;(iii) storing the calculated performance metrics in a database; and(b) verifying bus performance of the system bus, comprising:(i) determining whether the calculated performance metrics fall within a determined performance range, wherein the determined performance range comprises at least a self-learned performance range; wherein the self-learned performance range is generated by:(1) querying the database to receive a sample of previously stored performance metrics, wherein the sample of previously stored performance metrics comprises at least some performance metrics stored prior to the execution of the program and during the execution of the program; and(2) calculating the self-learned performance range based on values of the sample;(c) in response to the determining, varying a rate at which one or more events occur on the bus during the execution of the program, resulting in the generation of potential events which fall outside the predetermined performance range and the generation of those events more frequently. 2. The method of claim 1, wherein whether to vary the rate is determined by querying the database.
048204781	description	DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a nuclear reactor fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 10. Basically, the fuel assembly 10 includes a lower end structure or bottom nozzle 12 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 14 which project upwardly from the bottom nozzle 12. The assembly 10 further includes a plurality of transverse grids 16 axially spaced along the guide thimbles 14 and an organized array of elongated fuel rods 18 transversely spaced and supported by the grids 16. Also, the assembly 10 has an instrumentation tube 20 located in the center thereof and an upper end structure or top nozzle 22 removably attached to the upper ends of the guide thimbles 14 to form an integral assembly capable of being conventionally handled without damaging the assembly parts. As mentioned above, the fuel rods 18 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 16 spaced along the fuel assembly length. Each fuel rod 18 includes nuclear fuel pellets 24 and the opposite ends of the rod are closed by upper and lower end plugs 26,28 to hermetically seal the rod. Commonly, a plenum spring 30 is disposed between the upper end plug 26 and the pellets 24 to maintain the pellets in a tight, stacked relationship within the rod 18. The fuel pellets 24 composed of fissile material are responsible for creating the reactive power of the nuclear reactor. A liquid moderator/coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract heat generated therein for the production of useful work. Control Rod With Uniformly Changeable Axial Worth Turning now to FIGS. 2 to 12, there is shown the preferred embodiment of the control rod of the present invention, being generally designated 32, which can be used in some of the fuel assemblies 10 of the reactor core to compensate for xenon transients, such as occur at load follow when the reactivity of the core is reduced. As will become clear, the control rod 32 constitutes means which compensates for the reduction in reactivity in a manner which matches it generally uniformly in a transverse direction and symmetric in the axial direction. Basically, the control rod 32 includes an elongated inner cylindrical member 34 and an elongated outer cylindrical member 36 surrounding the inner member, with each of the members 34,36 being composed respectively of alternating poison and nonpoison regions 38,40 and 42,44 (in FIGS. 4 and 6, the poison regions have an "X" on them) and with one of the inner and outer members 34,36 being axially movable relative to the other to adjust the degree to which the poison regions 38,42 of the members 34,36 overlap with the nonpoison regions 40,44 thereof and thereby change the overall worth of the rod 32. More particularly, as seen in FIG. 5, the inner cylindrical member 34 from end to end has a solid cross-sectional configuration, whereas, as seen in FIG. 7, the outer cylindrical member 36 from end to end has an annular cross-sectional configuration and concentrically surrounds and is generally coextensive with the inner member 34, as seen in FIGS. 8 and 9. The poison regions 38,42 of the respective members 34,36 are composed exclusively of black poison material, such as unclad hafnium, and extend axially in an alternating arrangement with the nonpoison regions 40,44 of the members which are composed exclusively of a nonpoison material, such as Zirc-4. The regions of the members can be in the form of poison and nonpoison annular and solid pellets and thus the members themselves are formed by connecting the respective pellets together in any suitable manner, such as by welding them. It will be observed that there is no cladding material needed, and no plenum is needed since no oxide material is used and thus no reaction gas is given off. As will be observed in FIG. 8, each of the black poison regions 38,42 of the inner and outer members 34,36 has substantially the same axial height. Also, each of the nonpoison regions 40,44 of the members 34,36 has substantially the same axial height. Additionally, preferably the axial heights of the poison regions 38,42 are about the same as the nonpoison regions 40,44. A height of one foot for the regions was arbitrarily chosen. Such height was considered to be short enough to not cause significant axial offset changes with rod movement. The inner member 34 has a conical shaped lower head 46 which defines an outwardly projecting annular ledge 48 at the lower end of the member 34 upon which rests on an inner portion of a lower edge 50 of the outer member 36, as seen in FIGS. 8 and 10. In uch manner, the inner cylindrical member 34 supports the outer cylindrical member 36 prior to insertion of the control rod 32 into one of the guide thimbles 14 of the fuel assembly 10. The control rod 32 is placed in one guide thimble 14 of the fuel assembly 10 at beginning of core life (BOL). The inner or central member 34 is connected at an upper threaded end 51 to the control rod drive line, in the same manner as a conventional control rod, through a radially extending fluke or arm 52 of a conventional spider assembly 54, as seen in FIG. 1. As the inner cylindrical member 34 is lowered into the guide thimble 14, the outer cylindrical member 36 rides along with the inner member 34 by sitting on the lower ledge 48 of the inner member. When the two members 34,36 are almost completely inserted into the core, i.e., the guide thimble 14 of the fuel assembly 10, an outer portion of the lower edge 50 of the outer cylindrical member 36 encounters an annular stop 56 in the form of a sleeve in a lower portion of the guide thimble 14, for instance, fixed to the thimble 14 just above a dashpot 58 defined in its lower portion. The sleeve 56 is sized to support the outer portion of the lower edge 50 of the outer member 36 and has a central hole 60 sized to allow passage of the inner member 34 therethrough. While the outer member 36 encounters the stop 56, the inner member 34 can still continue down relative to the outer member 36 an additional foot to full insertion, as seen in FIG. 12. In such offset position of the poison and nonpoison regions 38,40 and 42,44 of the members 34,36, the control rod 32 has its maximum worth. At full reactor power, the members 34,36 are in the offset position so that they have maximum worth. To adjust the degree to which the poison regions 38,42 of the cylindrical members 34,36 overlap with the nonpoison regions 40,44 thereof and thereby change the overall worth of the rod 32, the inner member 34 is withdrawn upwardly relative to the outer member 36. At the aligned position of the poison and nonpoison regions 38,40 and 42,44 of the members 34,36, as seen in FIG. 11, the control rod 32 has its minimum worth, since the poison regions 42 in the outer member 36 shields the poison regions 38 in the inner member 34 from neutrons. In summary, therefore, the inner member 34 is axially movable relative to the outer member 36 between an upper axially displaced position (FIG. 12), in which the black poison regions 38,42 of the members 34,36 are disposed in side-by-side alignment and the nonpoison regions 40,44 thereof are also disposed in side-by-side alignment, and a lower axially displaced position (FIG. 11), in which the black poison regions 38,42 of the members 34,36 are disposed in side-by-side alignment with the nonpoison regions 40,44 of the members. In such manner, the overall worth of the control rod 32 can be changed in a substantially axially uniform manner. While the annular sleeve 56 fixed in the guide thimble 14 is the means illustrated herein for supporting and retaining the outer member 36 in a stationary position in the guide thimble 14 while the inner member 34 is movable axially thereto between the aligned and offset positions of the poison and nonpoison regions 38,42 and 40,44, respectively shown in FIGS. 11 and 12, it will be readily understood that other means could be used to accomplish this same purpose. For example, the outer cylindrical member 36 could engage a stop associated with the top nozzle to hold it stationary in the guide thimble 14. The xenon compensating control rods 32 are not used for power reduction because they do not control the change in axial offset which results as power is decreased. However, the rods might be used to also compensate for the Doppler reactivity which is changed during load follow since these changes are also symmetric in the axial direction. If Doppler control is to be performed, then the rods would not be in their most absorbing condition at full power so that their worth could be increased at reduced power to cancel the Doppler reactivity increase. For use just in xenon control, the rods are at their highest worth position at full power, as mentioned above. At reduced power, their worth would be decreased to compensate for the xenon concentration buildup. Since the regular control rods inserted to reduce power have not had to be withdrawn to compensate for xenon, there is enough worth in the core so that it can be returned to full power as fast as permissible. The xenon compensating rods have, therefore, preserved return to power capability and have removed the requirement of changing dissolved boron concentration in the core during load follow. Also, the rods are removed before EOL so that there is no fuel cycle penalty associated with their being in the core. They may not all be withdrawn at once, but in stages. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof.
	description	The present invention relates to a device for treating a packaging material by means of UV radiation. UV radiation has long been used in a wide range of applications. In the food industry, for example, it is commonly used for disinfecting or sterilizing packaging material, or for surface treating the food products themselves. UV radiation is also used for disinfecting work environments. The device according to the present invention may conveniently be used, though not exclusively, in a packaging material sterilizing unit of a pourable food product packaging machine, to which application reference is made in the following description purely by way of a non-limiting example. Various types of machines are known for packaging various types of pourable food products, such as fruit juice, wine, tomato sauce, pasteurized or long-storage (UHT) milk, etc. Such machines have different characteristics, depending on the type of package used, e.g. packages made of strip or sheet material, cups, bottles, tubs, etc. One of the best-known packaging machines is the one marketed under the registered trademark Tetra Brik®—referred to purely by way of a non-limiting example—in which the packages or packs are formed from a continuous tube of packaging material defined by a longitudinally sealed web. The packaging material has a multilayer structure comprising a layer of paper material covered on both sides with layers of heat-seal material, e.g. polyethylene. In the case of aseptic packages for long-storage products, such as UHT milk, the packaging material comprises a layer of barrier material, e.g. aluminium foil, which is superimposed on a layer of heat-seal plastic material, and is in turn covered with another layer of heat-seal plastic material eventually defining the inner face of the package and therefore contacting the food product. To produce aseptic packages, the web of packaging material is unwound off a reel and fed through a sterilizing unit, in which it is sterilized, for example, by immersion in a bath of liquid sterilizing agent, such as a concentrated solution of hydrogen peroxide and water. Alternatively, or in addition to being treated with a liquid sterilizing agent, the web of packaging material may be treated by exposure to one or more sources of UV electromagnetic radiation, as described, for example, in European Patent Application EP-A-919246. Downstream from the sterilizing unit, the web of packaging material is maintained in an aseptic chamber, in which it is dried, folded into a cylinder, and sealed longitudinally to form a continuous vertical tube. In other words, the tube of packaging material forms an extension of the aseptic chamber, is filled continuously with the pourable food product, and is then fed to a forming and (transverse) sealing unit for producing the individual packages, and in which the tube is gripped between pairs of jaws and sealed transversely to form aseptic pillow packs. The pillow packs are separated by cutting the sealed portions in between, and are then fed to a final folding station where they are folded mechanically into the finished form. Packaging machines of the type described above are used widely and satisfactorily in a wide range of food industries for producing aseptic packages from strip packaging material. Performance of the sterilizing unit, in particular, amply ensures conformance with regulations governing the sterility of the packages. Within the industry, however, a demand exists for further improvement, particularly as regards the safety of UV devices, which may be used for both disinfecting and sterilizing various types of packaging material, such as strip and sheet material, cups, bottles, tubs, etc. UV devices substantially comprise a UV radiation source housed in a casing and protected at the front by a screen made of material resistant and permeable to UV radiation. In commonly marketed devices, the screen is defined by a quartz plate. Though perfectly suitable in terms of physical-chemical properties, quartz has various drawbacks. In particular, it is extremely expensive. Moreover, a quartz screen is fragile and, if broken, tends to form extremely hard, sharp fragments. In known machines, the UV device, and therefore the quartz screen, is so located as to be protected against impact, so that the risk of it breaking and leaving trace fragments of quartz on the packaging material is highly unlikely. Nevertheless, at present, the possibility cannot be entirely excluded. Though unlikely, breakage may be caused by anomalous vibration or thermal stress, by flaws in the structure of the material, by accidental forcing or impact during assembly, or by a combination of any of these. Another drawback of quartz in this type of application is its tendency to dirty easily in normal operating conditions, and where it is extremely difficult, expensive, or even impossible to clean, whereas replacement is far from cheap. Quartz also has the further drawback of any flaws, cracks, dirt, etc. locally affecting its optical properties, thus possibly resulting in uneven irradiation of the material being treated. It is an object of the present invention to provide a device for UV irradiation treatment of a packaging material, designed to eliminate the aforementioned hazards typically associated with known devices. According to the present invention, there is provided a device for treating a packaging material by means of UV radiation, the device comprising a source of said radiation, and a screen for protecting said source and which is interposed between the source and the material for treatment; characterized in that said screen comprises a film of a polymer resistant and permeable to said UV radiation. Using a film of a polymer resistant and permeable to UV radiation provides for meeting the requirements of chemical/physical compatibility with the workplace, and at the same time for eliminating the drawbacks typically associated with quartz plates. In particular, the flexibility of the polymer material substantially eliminates any risk of tearing caused by vibration or thermal stress. And, even in the highly unlikely event of the film tearing, no fragments are formed, thus making the material perfectly suitable for use in the food industry. Moreover, a film of polymer material is much cheaper than a quartz plate, and can therefore be replaced quickly and cheaply in the event of soiling. Finally, in the event of flaws in the structure, i.e. on the surface, of the polymer film, uniform irradiation of the material being treated is unaffected, the polymer film being by nature slightly translucent and therefore still capable of ensuring adequate UV radiation diffusion. In a preferred embodiment of the present invention, the polymer is fluorinated or, better still, completely fluorinated and in the perfluoroalkoxy (PFA) class. More preferably, the fluorinated polymer is an MFA, the characteristics of which are particularly favourable in terms of UV radiation transmission, even, for example, in the case of UV radiation with a 222 nm wavelength, as described in the aforementioned European Patent Application EP-A-919246. In one possible embodiment of the invention, the polymer film is supported between substantially rigid grilles. Number 1 in the accompanying drawings indicates as a whole a device for treating a packaging material 2 for producing packages of a pourable food product. By way of a non-limiting example, the packaging material is a sheet material. Material 2 is fed continuously in known manner along a path P in a plane a of the material. Material 2 may be defined by a different type of material, such as a tub, cup, bottle, etc. Device 1 extends crosswise to material 2, i.e. perpendicularly to path P and parallel to plane α, and comprises an elongated outer casing 3 open on the side facing material 2 so as to form a window 4 defined by a peripheral flange 5. Device 1 also comprises a UV radiation source 6 housed longitudinally in casing 3, which is provided in conventional manner with a lining 7 of reflecting material. Source 6 emits UV radiation of a wavelength, for example, of 222 nm. The device also comprises a protective screen 8 fixed at the front to flange 5 so as to close window 4. According to the present invention, screen 8 comprises a film 9 of a polymer material resistant and permeable to the UV radiation emitted by source 6. The polymer material may be of any type, providing it is transparent to UV radiation (T≧80%) and resistant to said UV radiation in the operating conditions and for the time period between two successive replacements. The polymer material may, for example, be in the polyolefin group, such as PE or PP. Depending on the different resistance to UV radiation of different possible types of polymer film, provision may be made for appropriate, programmed, periodic replacement of film 9 of polymer material, so as to ensure constant mechanical and optical properties of the film. Replacement may be made, for example, either at given times, or continuously, using a strip of polymer material moving continuously between two appropriately powered reels. In a preferred embodiment, the material is a fluorinated polymer, or a completely fluorinated polymer in the perfluoroalkoxy (PFA) class, and even more preferably is an MFA, e.g. the MFA produced by AUSIMONT® under the trade name HYFLON® MFA. Film 9 is conveniently 20 to 200 μm thick, is preferably 40 μm, 50 μm, or 100 μm thick, and is in the form of a flexible, slightly translucent film capable of diffuse transmission of a fraction of the incident radiation. Though film 9, by virtue of its physical characteristics, may be used alone and simply fixed along the edges to flange 5, it is preferably interposed between two substantially rigid metal supporting grilles 10, which are fixed to flange 5 by means of respective peripheral frames 11, and provide for protecting and keeping the film flat. Film 9 has excellent characteristics in terms of transmittance and resistance to ageing under UV radiation, as shown in the following examples. A 50 μm thick film of Hyflon® MFA was exposed to 300 KJ/cm2 UV radiation of 222 nm wavelength for 800 hours. After exposure, the film showed no visible alteration, and 90% transmittance referred to the above 222 nm wavelength. A 100 μm thick film of Hyflon® MFA was exposed to 300 KJ/cm2 UV radiation of 222 nm wavelength for 800 hours. After exposure, the film showed no visible alteration, and 82% transmittance referred to the above 222 nm wavelength. Film 9 does not tear in the event of anomalous thermal stress or vibration, and, even if torn by accidental impact, which is substantially impossible in working conditions, does not produce fragments. Film 9 is also much cheaper than a conventional quartz plate. Clearly, changes may be made to the present invention without, however, departing from scope of the invention itself. In particular, the screen may be made of a different polymer material, e.g. PE, PP, PFA or many others. Also, grilles 10 may be different or even dispensed with.
042082064	abstract	Castings of superior surface quality and internal quality can be produced by:. (1) transferring the melt from the furnace into a separate refining vessel provided with submerged tuyeres, and PA1 (2) refining the melt by (a) injecting into the melt through the tuyeres an oxygen-containing gas which may contain up to 90% of a dilution gas, and (b) thereafter injecting a sparging gas into the melt through the tuyeres.. Preferably, the oxygen-containing gas is surrounded by an annular stream of a protective fluid. Argon is preferred for dilution, protection as well as sparging.
	description	1. Field of Invention The present invention relates to a method for selecting and configuring spent nuclear fuel bundles for casks so that the heat load for each of the casks is about the average heat load for all of the casks. 2. Related Prior Art Spent nuclear fuel bundles are disposed in casks. There has not been any method for selecting and configuring the spent nuclear fuel bundles for the casks. The present invention is therefore intended to obviate or at least alleviate the problems encountered in prior art. It is the primary objective of the present invention to provide a method for selecting and configuring spent nuclear fuel bundles for casks so that the heat load for each of the casks is about the average heat load for all of the casks. In the method according to the present invention, the spent nuclear fuel bundles are arranged in order based on their values of decay heat so that the spent nuclear fuel bundles with lower values of decay heat are given smaller numbers. Any spent nuclear fuel bundles with values of decay heat higher than a limit of decay heat for the casks are removed. The mean value of decay heat of the remaining all spent nuclear fuel bundles is calculated. It is determined if the number of cells of each cask is odd or even. It is determined if the number of the remaining spent nuclear fuel bundles is odd or even. If the number of the cells of each cask is odd or even, and if the number of the remaining spent nuclear fuel bundles is even, the remaining spent nuclear fuel bundles are matched. The spent nuclear fuel bundle with the highest value of decay heat is matched with the spent nuclear fuel bundle with the lowest value decay heat. The spent nuclear fuel bundle with the second highest value of decay heat is matched with the spent nuclear fuel bundle with the second lowest value of decay heat. The spent nuclear fuel bundle with the third highest value of decay heat is matched with the spent nuclear fuel bundle with the third lowest value of decay heat, and so on. The mean value of decay heat of each spent nuclear fuel bundle pair is calculated. The spent nuclear fuel bundle pairs are arranged in order based on their mean values of decay heat. The difference between the mean value of decay heat of each spent nuclear fuel bundle pair and the mean value of decay heat of all spent nuclear fuel bundles is calculated. It is determined if the number of the cells of each cask is odd or even. A first or second way is selected to dispose the remaining spent nuclear fuel bundles in the casks. It is determined if the heat load on each cask is smaller than a limit of heat load. If the heat load on each cask is not smaller than the limit of heat load, the limit of decay heat is reduced and the process is returned to the step of removing any spent nuclear fuel bundles with values of decay heat higher than a limit of decay heat for the casks. Other objectives, advantages and features of the present invention will become apparent from the following description referring to the attached drawings. There is provided a method for selecting and configuring spent nuclear fuel bundles for casks according to the preferred embodiment of the present invention. The method includes a subroutine for selecting the spent nuclear fuel bundles for the casks referring to FIG. 1 and another subroutine for configuring the spent nuclear fuel bundles for the casks referring to FIG. 2. The number of the casks is N1. Each cask includes a number of cells. The number of the cells of each cask is N2. The number of the spent nuclear fuel bundles is N3. At 11, the spent nuclear fuel bundles are numbered in order based on their values of decay heat. The spent nuclear fuel bundles with lower values of decay heat are given smaller numbers. At 12, any spent nuclear fuel bundles with values of decay heat higher than a limit of decay heat for the casks are removed. The number of the remaining spent nuclear fuel bundles is N4. N4 is equal to N3 if no spent nuclear fuel bundle is removed. The limit of decay heat is determined during the design of the casks. At 13, the mean value of decay heat of the remaining all spent nuclear fuel bundles is calculated. At 14, it is determined if N2 is odd. The process goes to 141 if N2 is odd, and goes to 15 if otherwise. No spent nuclear fuel bundle is reserved if N2 is even so that the number of the non-reserved spent nuclear fuel bundles, N5, is equal to N4. At 141, some of the spent nuclear fuel bundles with values of decay heat closest to the mean value of decay heat of the spent nuclear fuel bundles are reserved for the casks. The number of the reserved spent nuclear fuel bundles is equal to N2. The number of the non-reserved spent nuclear fuel bundles, N5, is equal to N4 minus N2. Then, the process goes to 15. At 15, it is determined if N5 is odd. The process goes to 151 if N5 is odd. Otherwise, the process goes to 16. An even number, N6, is calculated before the process goes to 16. N6 is equal to N5 if N5 is even. At 151, the spent nuclear fuel bundle with the highest value of decay heat is removed from the non-reserved spent nuclear fuel bundles so that N6 is equal to N5 minus 1. At 16, the N6 spent nuclear fuel bundles are matched, thus providing a number of spent nuclear fuel bundle pairs, N7. N7 is equal to N6 divided by 2. The N6 spent nuclear fuel bundles are arranged in order based on their values of decay heat. The spent nuclear fuel bundle with the highest value of decay heat is matched with the spent nuclear fuel bundle with the lowest value decay heat. The spent nuclear fuel bundle with the second highest value of decay heat is matched with the spent nuclear fuel bundle with the second lowest value of decay heat. The spent nuclear fuel bundle with the third highest value of decay heat is matched with the spent nuclear fuel bundle with the third lowest value of decay heat. Similarly, the N6 spent nuclear fuel bundles is matched, thus providing the N7 spent nuclear fuel bundle pairs. The sum of decay heat of each spent nuclear fuel bundle pair is calculated. At 17, the sum of decay heat of each spent nuclear fuel bundle pair is divided by 2, thus providing the mean value of decay heat of each spent nuclear fuel bundle pair. The N7 spent nuclear fuel bundle pairs are arranged in order based on their mean values of decay heat. At 18, the mean value of all spent nuclear fuel bundles is subtracted from the mean value of decay heat of each spent nuclear fuel bundle pair, thus providing the difference between the mean value of decay heat of each spent nuclear fuel bundle pair and the mean value of decay heat of all spent nuclear fuel bundles. At 19, it is determined if N2 is odd. The process goes to 191 if N2 is odd, and goes to 192 if otherwise. Either way, a number, N8, of spent nuclear fuel bundle pairs are selected for each cask. At 191, N8 is equal to N2 minus 1 and then divided by 2. Each reserved spent fuel bundle is reserved for a related cask. The difference between the value of decay heat of each reserved spent nuclear bundle and the mean value of decay heat of all spent nuclear fuel bundles is calculated. N8 spent nuclear fuel bundle pairs are selected for each cask so that the total of the difference between the mean value of decay heat of each selected spent nuclear fuel bundle pair and the mean value of decay heat of all spent nuclear fuel bundles plus the difference between the value of decay heat of the only reserved spent nuclear fuel bundle and the mean value of decay heat of all spent nuclear fuel bundles is close to zero. At 192, N8 is equal to N2 divided by 2. N8 spent nuclear fuel bundle pairs are selected for each cask so that the total of the difference between the mean value of decay heat of each selected spent nuclear fuel bundle pair and the mean value of decay heat of all spent nuclear fuel bundles is close to zero. At 20, it is determined if the heat load on each cask is smaller than a limit of heat load. The process goes to 21 if so, and returns to 12 if otherwise. Should the process return to 12, the limit of decay heat would be reduced. At 21, for each cask, there are N9 spent nuclear fuel bundles wherein N9 is equal to N8 multiplied by 2. The N9 spent nuclear fuel bundles are arranged in order based on their values of decay heat. At 22, the N9 spent nuclear fuel bundles are disposed in the cask so that the spent nuclear fuel bundles with higher values of decay heat are located closer to the center of the cask. The value of decay heat of each spent nuclear fuel bundle is closest to the value of decay heat of another spent nuclear fuel bundle at a same distance to the center of the cask in a diagonal line. The sum of decay heat of the spent nuclear fuel bundles in each quadrant is close to the sum of decay heat of the spent nuclear fuel bundles in any other quadrant. Referring to FIGS. 3 through 11, there is shown a working environment for the execution of the method. There are 2 casks, i.e., N1 is 2. Each cask includes 56 cells, i.e., N2 is 56. There are 150 spent nuclear fuel bundles, i.e., N3 is 150. The limit of heat load on each cask is 13 kilo watts. The limit of the decay heat is 232.14 watts. Referring to FIG. 3, there is shown a table of 150 spent nuclear fuel bundles numbered in order based on the decay heat. This table is provided at 11. Referring to FIG. 4, there is shown a table of 75 spent nuclear fuel bundle pairs. This table is provided at 16. Referring to FIG. 5, there is shown another table of the 75 spent nuclear fuel bundle pairs arranged in order based on their mean values of decay heat. This table is provided at 17. Referring to FIG. 6, there is shown another table of the 75 spent nuclear fuel bundle pairs arranged in the order shown in FIG. 5 and their differences from the mean value of decay heat of the 150 spent nuclear fuel bundles. This table is provided at 18. Referring to FIG. 7, there is shown a table of 28 spent nuclear fuel bundle pairs for the first cask. This table is provided at 192. Referring to FIG. 8, there is shown a table of another 28 spent nuclear fuel bundle pairs for the second cask. This table is provided at 192. Referring to FIG. 9, there is shown a table of the 56 spent nuclear fuel bundles for the first cask. This table is provided at 21. Referring to FIG. 10, at 22, the 56 spent nuclear fuel bundles are disposed in the first cask. Referring to FIG. 11, the heat load on each quadrant is like the heat load on any other quadrant. The spent nuclear fuel bundles are selected and configured in the casks as low as reasonably achievable regarding the heat load. The present invention has been described via the detailed illustration of the preferred embodiment. Those skilled in the art can derive variations from the preferred embodiment without departing from the scope of the present invention. Therefore, the preferred embodiment shall not limit the scope of the present invention defined in the claims.
	claims	1. A component for conducting or receiving a fluid, the component comprising:a wall including:a carrying structure composed of a glass fiber reinforced plastic, said carrying structure having inner and outer surfaces;electrically insulating inner and outer protective layers each disposed on a respective one of said inner and outer surfaces of said carrying structure;an electrically conductive inner intermediate layer having an electrical terminal and being disposed between said inner protective layer and said carrying structure; andan electrically conductive outer intermediate layer having an electrical terminal, being insulated electrically from said inner intermediate layer and being disposed between said outer protective layer and said carrying structure;means for admixing an additive for a measurement duration to increase a conductivity of the fluid conducted or received by the component during a measurement process; andmeans for detecting an electrical resistance between the outer and inner intermediate layers. 2. The component according to claim 1, wherein said intermediate layers are formed by an electrically conductive fabric. 3. The component according to claim 1, wherein the wall is part of a component of a fluid-conducting line system of an industrial plant. 4. The component according to claim 1, wherein the wall is part of a component of a line system of a tertiary cooling circuit of a nuclear power station. 5. A method for testing a component, the method comprising the following steps:providing a component for conducting or receiving a fluid, the component having a wall including:a carrying structure composed of a glass fiber reinforced plastic, the carrying structure having inner and outer surfaces;electrically insulating inner and outer protective layers each disposed on a respective one of the inner and outer surfaces of the carrying structure;an electrically conductive inner intermediate layer having an electrical terminal and being disposed between the inner protective layer and the carrying structure; andan electrically conductive outer intermediate layer having an electrical terminal, being insulated electrically from the inner intermediate layer and being disposed between the outer protective layer and the carrying structure;admixing an additive for a measurement duration to increase a conductivity of the fluid conducted or received by the component during a measurement process; anddetecting an electrical resistance between the outer and inner intermediate layers.
048184764	summary	TECHNICAL FIELD The invention relates to a nuclear pressure vessel closure head, and more particularly to a nuclear reactor vessel stud thread protector to protect the exposed threaded section of a stud bolt and an associated nut projecting above the vessel head. BACKGROUND OF THE INVENTION In a typical nuclear reactor pressure vessel, a removable closure head is secured to the pressure vessel by a multitude of stud bolts, each having an associated nut and washer. One design of a typical four-loop plant has 54 such stud bolts. Stud bolts typically have threaded sections at both ends: the lower threaded section passes through an aperture in the vessel head and is received within a threaded bore in the reactor vessel flange; and the upper threaded section projects above the vessel head, upon which section the nut is torqued down against an associated washer and the vessel head to compress a seal between the head and the vessel. In this manner, the vessel head is securely held in sealing engagement with the reactor vessel. For uniform nut loading on the studs, the studs are tensioned by a process well known in the art. The studs are tensioned and the nuts securely threaded thereon and torqued at a predetermined level, which procedure prevents inadvertent loosening of the nuts during reactor operation. A measuring rod is received within a vertical bore which extends the entire length of the stud, and is used to measure stud elongation to ensure proper tensioning. After the head has been so installed the vertical bore is sealed by means of a screw or bolt which is threaded into the top portion of the stud in order to prevent accumulation of water or other material therein. Projecting through the top of the vessel head are a plurality of control rod drive mechanism housings. For a typical four loop plant, there are about 80 such ports. An example of one such design is shown in FIGS. 1A and 1B. Normally, there are on the order of 60 fuel assemblies out of the approximately 200 within the reactor vessel core which have control rods associated therewith. Depending upon such factors as power level of the reactor, enrichment and depletion of fuel in the core, control rod drive mechanisms are inserted through preselected housings. The remaining housings are utilized as instrumentation ports or spare penetrations. The instrumentation ports are for the introduction of instrumentation devices, such as thermocouples, into the reactor vessel. The control rod drive mechanism housing is a stainless steel pressure housing attached to a head adapter or port projecting upward from the reactor vessel head. The adapter is welded to the reactor vessel head and constitutes, in effect, an integral part of the vessel. A typical pressurized water reactor is operated at an internal pressure of about 15 MPa (2250 psia); the design pressure of the reactor vessel and associated components is about 17 MPa (2500 psia). Control rods secured to the lower end of the control rod drive mechanism are periodically inserted into and withdrawn from a fuel assembly, depending on power demands in the reactor core. Reactor coolant water may leak from the mechanical flange joint on the housing because of such large internal pressures. When the drive mechanism is withdrawn, a thin film of liquid coolant around the drive mechanism may be withdrawn with it. This coolant may then drip onto reactor vessel components, specifically the exposed portion of the stud bolt projecting above the vessel head. Therefore, although the vessel and associated components are welded and sealed as best as they can be, it may be possible for some liquid coolant, typically water, to drip from the control rod drive mechanism housings. The coolant within the reactor vessel is slightly acidic due to the presence of boric acid which is dissolved within the coolant. Boric acid is a neutron absorber used as a variable reactivity control over the long term operation of the plant, whereas the control rods provide rapid reactivity control for shutdown and other rapid reactivity changes. The number and placement of control rods is dependent upon numerous operating characteristics of a nuclear power plant. Even though there are regulatory limits on the allowable amount of coolant which be emitted from the reactor vessel, components on the exterior and in close proximity to the reactor vessel need to be protected from any possible corrosive leakage impingement due to the presence of borated water. One purpose of the top closure screw, referred to previously, disposed within the bore of the stud is to prevent accumulation of this potentially corrosive borated water within the stud. Although the internal vertical bore of the stud bolt is sealed off from this corrosive spray leakage, the exposed external threads of the stud projecting above the vessel head, as well as the associated nut and washer, are not. If the exposed threads are damaged, it may become difficult to quickly and easily remove the nuts therefrom. This is undesirable since removal and replacement of the vessel head is to be completed in as short a time period as possible. First, since the reactor vessel in a containment building of a nuclear power plant defines an irradiated environment, it is advantageous to provide for rapid maintenance procedures to reduce the time in which maintenance personnel are required to spend in and around such environment, thereby reducing individual man-rem exposure. Secondly, the quicker and easier such routine maintenance can be performed, the less down time experienced by the nuclear reactor. Since a nuclear reactor power plant operator cannot generate electricity when the plant is not operating, it must purchase replacement power elsewhere. Plus the more rapidly routine maintenance can be completed, that much more time can be devoted to actual power plant operation thereby improving the efficiency of the plant as well as increasing revenues generated by its operator. There exists in the prior art conventional protective caps for bolts and nuts. An example of which is U.S. Pat. No. 3,548,704 issued to Kutryk on Dec. 22, 1970. Although such devices are satisfactory for use in a normal, everyday environment, a nuclear power plant presents additional concerns and considerations. Such a plant is a highly sophisticated and complex machine, damage to which must be prevented with great diligence. Also, any coolant that may be emitted from the reactor vessel needs to be safely dealt with and controlled, and not allowed to merely run off the protector to the environment. Thus a stud protector for use in association with a nuclear power plant must not only be a reliable device which cannot be accidentally removed, but must also be able to control the leak path o the coolant. It must be able to withstand the relatively harsh environment, especially with respect to the borated coolant. Furthermore, it is highly desirable to have available a stud thread and nut protector which can add significantly to other safety aspects associated with a nuclear power plant. It is therefore an object of the present invention to provide a reactor vessel stud thread and nut protector which can be used in the highly specialized environment of a nuclear reactor pressure vessel. It is another object of the present invention to provide such a device which will prevent damage to exposed threads and nuts resulting from potentially corrosive spray leakage of borated and irradiated coolant. It is a further object of the present invention to provide a reactor vessel stud thread protector which can also collect any such coolant for return to the reactor's coolant system. It is a still further object of the present invention to provide a device which will provide protection from accidental deformation to the stud, nut and washer due to operator maintenance around adjacent studs. DISCLOSURE OF THE INVENTION In combination with a nuclear reactor pressure vessel having a removable closure head, the closure head being sealingly engaged with the pressure vessel by a plurality of stud bolts, said bolts each having a lower end threadingly engaged within a flange section of the vessel and an upper end passing through a corresponding aperture in the closure head and projecting thereabove, the upper end also having a threaded section for threadingly engaging a nut and thereby sealingly engaging the closure head with the pressure vessel. A vertical bore is disposed within the stud bolt, the bore being internally threaded at its upper end. The combination further includes a reactor vessel stud thread protector which encloses the exposed portion of the bolt and nut projecting above the vessel head, wherein the protector is comprised of a tubular wall portion being open at its lower end and substantially closed at its upper end. The upper end has a hole therethrough adapted to receive a closure screw, the screw having a head of a diameter larger than the hole and a threaded portion on its lower end adapted to be received in the internally threaded portion of the upper end of the stud bolt. A drip pan associated with the outer surface of the protector is disposed radially inwardly with respect to the outer periphery of the vessel head, whereby the drip pan collects any fluid being emitted from the reactor vessel.
	claims	1. A CT detector comprising:a scintillator module including at least one scintillator configured to be impinged with radiographic energy from a radiographic energy source;at least one indexing pin connected to the scintillator module; anda collimator assembly having at least one comb, wherein the collimator assembly is configured to position the at least one comb, and wherein the at least one comb has a plurality of teeth configured to engage the at least one indexing pin. 2. The CT detector of claim 1 wherein the at least one indexing pin further comprises at least one diamond-shaped pin. 3. The CT detector of claim 1 wherein the plurality of teeth are further configured to position a plurality of collimator plates. 4. The CT detector of claim 1 wherein the plurality of teeth have a first set of teeth extending a first length and a second set of teeth extending a second length, wherein the second set of teeth is longer than the first set of teeth. 5. The CT detector of claim 4 wherein the second set of teeth are constructed to flank the at least one indexing pin. 6. The CT detector of claim 5 wherein the at least one indexing pin is generally flanked by at least two teeth of the second set of teeth, and has a side surface constructed to abut a side surface of the second set of teeth. 7. The CT detector of claim 1 wherein the at least one scintillator includes a plurality of scintillators uniformly arranged in a scintillator array. 8. The CT detector of claim 1 incorporated into a rotatable gantry of a CT imaging system which is extensible in the Z-direction. 9. A scintillator-collimator combination comprising:a plurality of collimator plates configured to collimate x-rays projected thereat;a scintillator module having a locating pin and a scintillator pack formed of a material configured to illuminate upon reception of x-rays; anda comb having a first set of teeth and a second set of teeth, the first set of teeth and the second set of teeth constructed to align the plurality of collimator plates, and the second set of teeth constructed to engage the locating pin on the scintillator module and align the scintillator module, via the locating pin, relative to the plurality of collimator plates. 10. The scintillator-collimator combination of claim 9 wherein the locating pin is constructed to snuggly engage a recess of the comb, wherein the recess is defined between a pair of the second set of teeth. 11. The scintillator-collimator combination of claim 9 wherein the locating pin has a diamond shape. 12. The scintillator-collimator combination of claim 9 wherein the locating pin is configured to align the scintillator pack with respect to the plurality of collimator plates. 13. The scintillator-collimator combination of claim 9 further comprising a photodiode array, wherein the scintillator pack is configured to be optically coupled to the photodiode array and configured to detect illumination from the scintillator pack and output electrical signals responsive thereto. 14. The scintillator-collimator combination of claim 9 incorporated into a CT imaging system designed to acquire diagnostic data of a medical patient. 15. A scintillator-collimator combination comprising:a plurality of collimator plates configured to collimate x-rays projected thereat;a scintillator module having:a scintillator pack formed of a material configured to illuminate upon reception of x-rays; andat least one locating pin; anda comb having a first set of teeth and a second set of teeth, the first set of teeth and the second set of teeth constructed to align the plurality of collimator plates, and the second set of teeth constructed to engage the scintillator module and align the scintillator module relative to the plurality of collimator plates;wherein the comb comprises a recess defined between a pair of the second set of teeth, andwherein the at least one locating pin is constructed to snuggly engage the recess. 16. The scintillator-collimator combination of claim 15 wherein the locating pin has a diamond shape. 17. The scintillator-collimator combination of claim 15 wherein the locating pin is configured to align the scintillator pack with respect to the plurality of collimator plates. 18. A CT system comprising:a rotatable gantry having a bore centrally disposed therein;a table movable fore and aft through the bore and configured to position a subject for CT data acquisition;a high frequency electromagnetic energy projection source positioned within the rotatable gantry and configured to project high frequency electromagnetic energy toward the subject; anda detector array disposed within the rotatable gantry and configured to detect high frequency electromagnetic energy projected by the projection source, the detector array including:a plurality of scintillator modules, each having a scintillator array and at least one indexing pin;a collimator assembly having a plurality of collimator plates; anda detector support having at least one comb of alignment teeth, the alignment teeth constructed to align the plurality of collimator plates, and the alignment teeth constructed to engage the at least one indexing pin to align a scintillator array with the plurality of collimator plates. 19. The CT system of claim 18 wherein the at least one comb includes a first set of teeth extending a first length and a second set of teeth extending a second length, wherein the second set of teeth is longer than the first set of teeth. 20. The CT system of claim 19 wherein the comb defines a uniform spacing between collimator plates of the plurality of collimator plates. 21. The CT system of claim 19 wherein the second set of teeth extends beyond an edge of the collimator plates. 22. The CT system of claim 19 wherein the second set of teeth flank the at least one indexing pin. 23. The CT system of claim 18 wherein the at least one indexing pin is hexagonally-shaped with a tapered top section. 24. The CT system of claim 18 wherein the at least one indexing pin is laterally positioned beyond an end of a respective scintillator array. 25. A method of manufacturing a CT detector comprising the steps of:providing a scintillator array having at least one locator element extending beyond the scintillator array;providing a comb having a plurality of teeth constructed to define a spacing between collimating elements of a collimator; andpositioning the at least one locator element between at least two of the plurality of teeth.
048083704	summary	The invention relates to a gas-cooled, high-temperature nuclear reactor with a metallic core barrel having a lining formed of graphite or carbon blocks, and a hot gas line having an outer metallic pipe for confining pressure and a ceramic flow guidance pipe separated therefrom by insulation. Such a nuclear reactor is known from European Patent No. 0 039 016 and a further development is described in German Published, Non-Prosecuted Application DE-OS No. 33 45 457, corresponding to U.S. application Ser. No. 681,544, filed Dec. 14, 1984. Such hot gas lines have exhibited disadvantages stemming from the large temperature gradients, which will be described below. It is accordingly an object of the invention to provide a connection of the hot gas line to the core barrel of a gas-cooled, high-temperature nuclear reactor, which overcomes the hereinaforementioned disadvantages of the heretofore-known devices of this general type. This connection presents particular difficulties since it must withstand cooling gas which has been heated in the nuclear reactor to temperatures up to 950.degree. C.; hot gas leaks that might occur at the connection could therefore endanger metallic components located in the vicinity thereof (for instance at the corc barrel) In addition, leakage streams could lead to an erosion of the heat insulation material which is provided for the necessary temperature breakdown between the flow guidance pipe exposed to the hot gas temperature and a metallic pipe which surrounds the flow guidance pipe and provides a pressure seal. Due to the large temperature differences to which the components are subjected by the transition from the shut down to the operating phase, changes of position relative to each other which can be traced to different thermal expansion, must be expected; an angular offset between the core barrel and the hot gas line also occurs in this case, which would lead to excessive stresses if the two parts were rigidly connected to each other. In addition, it should be possible to replace components that have become defective, even under conditions of difficult access and if only remotely-controlled tools are used. With the foregoing and other objects in view there is provided, in accordance with the invention, a gas-cooled, high-temperature nuclear reactor, comprising a metallic core barrel, a graphite or carbon block lining disposed in the core barrel, a hot gas line including an outer pressure-confining metallic pipe and a ceramic flow guidance pipe, insulation separating the metallic pipe from the ceramic pipe, a stub concentric with the hot gas line, means for detachably connecting the stub to the core barrel, the metallic pipe being tightly disposed in the stub, means for detachably fastening the metallic pipe to the stub, a sleeve, means for detachably fastening the sleeve to the lining, a bellows compensator being disposed in the stub and having one end tightly fastened to the stub and another end, and means for connecting the other end to the sleeve It is known to use bellows-type compensators for equalizing position changes of components. The structure according to the invention utilizes these properties while at the same time protecting the parts which are necessarily metallic from the temperature of the hot gas. The joint produced in this manner is additionally completely gastight, and the gas-tightness thereof can be checked after installation without difficulty, since the hot gas line is only assembled later by a simple insertion into the stub Since the stub is detachably fastened to the core barrel it can be replaced, if necessary, tigether with the hot gas line, so that fit problems between these two parts after reassembly are avoided. In accordance with another feature of the invention, the ceramic pipe and the sleeve are formed of carbon fiber-reinforced carbon. Thus, not only the flow guidance pipe but also the sleeve which serves for breaking down the temperature between the hot lining of the core barrel and the bellows is made of carbon reinforced with carbon fibers (also known under the designation CFC). This material has sufficient strength even at the high temperatures prevailing at that location and is protected against oxidation, i.e., burn-up, by the provision that it is used only in an inert gas atmosphere (for instance helium). In accordance with a further feature of the invention, the stub and the bellows compensator define a space therebetween and the stub has openings formed therein and including a cold gas line coaxially surrounding the hot gas line. In nuclear reactors which are equipped like those described in the abovementioned publications, in other words with a cold gas line concentrically outside the hot gas line, the openings in the stub enclosing the bellows conduct a small shunt flow (in the order of 1/1000) of the cold gas to the outside of the bellows. This contributes to cooling the bellows, especially for compensating the temperatures which may vary over the periphery thereof. In accordance with an added feature of the invention, the means for detachably fastening the sleeve to the lining are in the form of carbon screws reinforced with carbon fibers In accordance with an additional feature of the invention, the metallic pipe has an outer surface and the outer surface of the metallic pipe and the stub have regions of mutual contact, and including a coating disposed on the regions for preventing friction welding of the metallic pipe to the stub in the cooling gas atmosphere. This is done to ensure the later disassembly of the pressure pipe from the stub. Welding of the two parts together cannot be precluded in a helium atmosphere. In accordance with again another feature of the invention, the stub has a flange to which the means for detachably fastening the stub to the core barrel are connected, the flange being offset or stepped back from the wall of the core barrel defining a gap therebetween for accomodating tools for cutting the fastening means. This is done in order to assure that the stub can be detached from the core barrel even if the basically detachable screws used for fastening can no longer be removed because they are welded fast or due to other reasons The cutoff stumps of the screws then remain in the core barrel and during reassembly, new tapped holes are cut in the core barrel which are offset by a given angle relative to the old ones. In accordance with a concomitant feature of the invention, the means for detachably fastening the metallic pipe to the stub include means for making the fastening means accessible to removal by milling tools, preferably including the screws This is done since hot gas pipes and stubs may in some cases be replaced together. This raises no special problem, even when working from the inside of the flow guidance pipe. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a connection of a hot gas line to the core barrel of a gas-cooled high-temperature nuclear rector, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.
	summary
	claims	1. A method of generating a two-level pattern for lithographic processing by multiple beamlets, the method comprising:providing a pattern in vector format;converting the vector format pattern into a pattern in pixmap format; andforming a two-level pattern by application of error diffusion on the pixmap format pattern. 2. The method of claim 1, wherein the pixmap comprises an array of pixel cells, and wherein a multi-level value is assigned to each pixel cell. 3. The method of claim 2, wherein providing multi-level values to pixel cells is based on relative coverage of the vector-format pattern by the respective pixel cell. 4. The method of claim 2, wherein providing multi-level values to pixel cells is based on dose level values of the vector format pattern. 5. The method of claim 1, wherein the vector-format pattern is formed by two-level values. 6. The method of claim 1, wherein application of error diffusion includes:dividing the array of pixels in portions, each portion being assigned to be patterned by a different beamlet;determining an error diffusion parameter value for each portion;assigning a two-level value to the pixel cells within each portion using said error diffusion parameter value as determined. 7. The method of claim 6, wherein determining the error diffusion parameter value is based on beamlet current measurements. 8. The method of claim 6, wherein said error diffusion parameter value is a threshold value, and wherein said assigning a two-level value to the pixel cells within a portion is based on comparison with the threshold value determined for said portion. 9. The method of claim 6, wherein said error diffusion parameter is a value representing the higher level of the two-level value. 10. The method of claim 1, wherein said error diffusion is a type of one-dimensional, 1D, error diffusion. 11. The method of claim 1, wherein said error diffusion is a type of two-dimensional, 2D, error diffusion. 12. The method of claim 11, wherein the 2D-error diffusion uses a Floyd-Steinberg kernel. 13. The method of claim 2, wherein the application of error diffusion is further restricted by disallowing diffusion towards one or more pixel cells that fulfill a no-shift condition. 14. The method of claim 13, wherein the no-shift condition is that a multi-level value assigned to said one or more pixels is equal to or below a further threshold value. 15. The method of claim 14, wherein said further threshold value equals zero. 16. The method of claim 13, wherein the no-shift condition is that said one or more pixels are located outside a feature. 17. A computer readable medium for performing, when executed by a processor, the method of generating a rasterized two-level pattern as defined by claim 1. 18. A pattern generator comprising:an input for receiving a pattern in vector format;a processing unit for performing the method of generating a two-level pattern for lithographic processing according to claim 1; andan output for supplying the two-level pattern. 19. The pattern generator of claim 18, further comprising a memory for storing a pattern in pixmap format, the memory being communicatively coupled to the processing unit. 20. A charged particle multi-beamlet system for exposing a target using a plurality of beamlets, the system comprising:a beamlet modulation system for modulating the plurality of beamlets so as to form an exposure pattern;a projection system for projecting the modulated beamlets on to the surface of the target;a deflector array for deflecting the plurality of beamlets in a first direction;a substrate support member for supporting the target to be exposed;a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a second direction and movement of the group of beamlets in the first direction such that the target can be exposed in accordance with an array of pixel cells;wherein the charged-particle multi-beamlet system further comprises a beamlet pattern generator of claim 18. 21. The system of claim 20, wherein the projection system comprises an array of projection lens systems. 22. The system of claim 21, wherein the plurality of beamlets is arranged in groups of beamlets, and each projection lens system corresponds with a group of beamlets. 23. The system of claim 21, wherein the deflector array comprises a plurality of deflectors, each deflector being arranged to deflect a corresponding group of beamlets. 24. Lithographic system comprising:a preprocessing unit;a charged particle multi-beamlet system for exposing a target using a plurality of beamlets in accordance with a two-level pattern;wherein the preprocessing unit comprises a beamlet pattern generator of claim 18. 25. The lithographic system of claim 24, wherein the charged particle multi-beamlet system comprises:a beamlet modulation system for modulating the plurality of beamlets so as to form an exposure pattern;a projection system for projecting the modulated beamlets on to the surface of the target;a deflector array for deflecting the plurality of beamlets in a first direction;a substrate support member for supporting the target to be exposed;a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a second direction and movement of the group of beamlets in the first direction such that the target can be exposed in accordance with an array of pixel cells. 26. The system of claim 25, wherein the projection system comprises an array of projection lens systems. 27. The system of claim 26, wherein the plurality of beamlets is arranged in groups of beamlets, and each projection lens system corresponds with a group of beamlets. 28. The system of claim 26, wherein the deflector array comprises a plurality of deflectors, each deflector being arranged to deflect a corresponding group of beamlets.
054266769	summary	BACKGROUND This invention generally relates to seals and more particularly relates to an extrusion-resistant seal assembly for sealing a gap defined between a first structure spaced-apart from a second structure, which first structure and second structure may be a first flange and a second flange belonging to an instrumentation column of the kind typically found penetrating nuclear reactor pressure vessels. Before discussing the current state of the art, it is instructive first to briefly describe the structure and operation of a typical nuclear reactor pressure vessel and its associated instrumentation columns. In this regard, a nuclear reactor pressure vessel is a device for producing heat by controlled fission of fissile material contained in a plurality fuel rods grouped to form a plurality of fuel assemblies disposed in the pressure vessel. The plurality of fuel assemblies define a nuclear reactor core in the pressure vessel. The pressure vessel itself includes a shell having an open top end and a closure head sealingly capping the open top end of the shell, so that the pressure vessel may be suitably pressurized thereby. A plurality of absorber rods slidably extend into each fuel assembly for controlling the fission process therein. Liquid moderator coolant (i.e., demineralized borated water), which may be pressurized to a pressure of approximately 2,500 psia during normal operation or approximately 3,000 psia during off-normal operation (e.g., during "overpressure" transients), is caused to flow over the fuel rods disposed in the pressure vessel for assisting in the fission process and for removing the heat produced by the fission process. During operation of the nuclear reactor, heat due to fission of the fissile fuel material is carried from the fuel rods by the liquid moderator coolant flowing over the fuel rods, which liquid moderator coolant becomes radioactive as it flows over the fuel rods. The heat carried away by the liquid moderator coolant is ultimately transferred to a turbine-generator set for generating electricity in a manner well known in the art. Penetrating the closure head are a plurality of elongated instrumentation columns, each instrumentation column having a longitudinal bore therethrough for receiving instrumentation wiring connected to an instrumentation probe located in the reactor core. The bore of the instrumentation column is typically in fluid communication with the pressurized reactor coolant circulating through the pressure vessel. Each probe is designed to measure predetermined core physics quantities (e.g., temperature, neutron flux, etc.) in the reactor core. Moreover, the instrumentation column is segmented for ease of assembly and servicing. The ends of at least some of the segments include opposing generally circular flanges, which are capable of being moved into close proximity and connected together for completing the assembly of the segmented instrumentation column. At least one of these pair of opposing flanges is located externally to the pressure vessel. As stated hereinabove, the pressurized radioactive coolant is in fluid communication with the bore of the instrumentation column. Therefore, for safety reasons, it is prudent to provide suitable seals at the interface of the flanges to prevent leakage of the pressurized radioactive coolant from between the flanges. In this regard, it is known that graphite is useable as a seal material because when clamped between opposing structures, it has a relatively higher compressibility than, for example, an all-steel seal. More specifically, it is known that an all-graphite seal can compress approximately 30% more than an all-steel seal. However, it is also known that graphite seals are most suitable to seal against relatively low fluid pressures of about 600 psia. When exposed to higher fluid pressures, such all-graphite seals tend to experience what is termed in the art as "blow-out". That is, when an all-graphite seal "blows-out", it tends to extrude, thereby compromising the ability of the all-graphite seal to maintain its sealing function. Therefore, although all-graphite seals can withstand relatively high compression, they tend to extrude at the relatively high fluid pressures (e.g. 2,500-3,000 psia) achievable in nuclear reactor pressure vessels. Hence, a problem in the art is to provide a suitable graphite seal for use in nuclear reactor pressure vessel instrumentation columns, the seal being capable of withstanding relatively high pressure without extrusion or "blow-out". A graphite seal is disclosed in U.S. Pat. No. 3,564,400 titled "Nuclear Magnetic Resonance Flowmeter Employing Ceramic Tube" issued Feb. 16, 1971 in the name of Ronald L. Pike et al. The Pike et al. patent relates to a nuclear magnetic resonance (or NMR) flowmeter, and particularly to such a flowmeter in which a ceramic tube is employed as the central conduit for conveying paramagnetic fluid. This patent discloses a flange secured to an end plate by means of bolts. A graphite sealing ring is disposed between the end plate and the flange in an annular recess provided in the exterior surface of the end plate. The interior side of the flange is provided with a boss, which has dimensions comparable to those of the annular recess. When the bolts are tightened, the graphite ring is compressed in the recess between the end plate and the boss, and is distorted under pressure to flow outwardly from the recess and fill up any space between the tube and the surfaces of the end plate and the flange. Therefore, this patent does not appear to disclose a graphite seal assembly configured to resist extrusion. Although the above recited prior art discloses a graphite seal, the above recited prior art does not appear to disclose a high pressure extrusion-resistant seal assembly for sealing a gap defined between a first structure spaced-apart from a second structure, which first structure and second structure may be a first flange and a second flange, respectively, belonging to an instrumentation column of the kind typically found penetrating nuclear reactor pressure vessels. Therefore, what is needed is an extrusion-resistant seal assembly for sealing a gap defined between a first structure spaced-apart from a second structure, which first structure and second structure may be a first flange and a second flange, respectively, belonging to an instrumentation column of the kind typically found penetrating nuclear reactor pressure vessels.
	description	This application claims priority to European Patent Application EP 09305921.0, filed on Sep. 30, 2009, the entire disclosure of which is incorporated by reference herein. The present invention relates to nuclear fuel assemblies. An object of the invention is to reduce costs related to development, manufacture and use of nuclear fuel assemblies. To this end, the invention provides a module for forming a nuclear fuel assembly, of the type comprising a casing extending in a longitudinal direction, a bundle of fuel rods encased in and supported by the casing and connection means provided on the casing for connecting the casing side-by-side to the casing of at least one other module to obtain a nuclear fuel assembly having a channel box defined by the casings of the assembled modules and of larger cross-section than that of the casing of each of the assembled modules and a bundle of fuel rods of larger cross-section than that of each of the assembled modules. In other embodiments, the module comprises one or several of the following features, taken in isolation or in any technically feasible combination: the casing has a cross-section of polygonal shape with one bevelled corner for delimiting a space for a water channel between the casings of assembled modules; the casing has a cross-section of regular polygonal shape with one bevelled corner, namely a cross-section of quadrilateral shape; the bevelled corner is opened or is closed by a bevel wall of the casing; the connection means are provided on longitudinal edges of the casing edging the bevelled corner; the connection means comprise at least one sleeve aligned in the longitudinal direction with the missing edge of the polygonal cross-section of the casing; the casing comprises at least one first side wall adapted to separate two sub-channels in a channel box defined by the casings of assembled modules; each first side wall comprises at least one groove on the outer face of the first side wall; each first side wall is adapted to define with first side walls of other modules assembled to the module a cross-shaped partition in a channel box defined by the casings of the assembled modules. The invention also relates to a nuclear fuel assembly formed of a plurality of modules as defined above assembled together side-by-side. In other embodiments, the nuclear fuel assembly comprises one or several of the following features, taken in isolation or in any technically feasible combination: it comprises a water channel delimited by bevel walls of the casings of the modules each closing a bevelled corner of a respective casing exhibiting a polygonal cross-section with a bevelled corner; it comprises a water channel delimited by a tube inserted in a spaced formed by bevelled corners of the casings of the modules exhibiting a polygonal cross-section with a bevelled corner; it comprises a channel box defined by the casings of the assembled modules and a partition of cross-shaped cross-section dividing the channel box in sub-channels receiving a sub-bundle of fuel rods; and it comprises an outer tubular housing surrounding the casings of the modules. The module 2 for forming a nuclear fuel assembly illustrated on FIGS. 1 and 2 comprises a tubular casing 4 extending in a longitudinal direction and a bundle of fuel rods 6 accommodated inside the casing 4. The module 2 is elongated in the longitudinal direction, only a longitudinal section of the module 2 being illustrated on FIG. 2 for the sake of clarity. The casing 4 is adapted to allow a coolant to flow longitudinally from a longitudinal lower end towards a longitudinal upper end of the casing 4 around the fuel rods 6. The casing 4 is opened at its longitudinal ends. The casing 4 exhibits a closed transverse cross-section of square shape with one bevelled corner 8. The casing 4 comprises four side walls 10, 12 arranged in a square. The side walls 10, 12 comprise two first side walls 10 adjacent the bevelled corner 8 and two second side walls 12 opposite the bevelled corner 8. Each first side wall 10 extends from the adjacent second side wall 12 towards the bevelled corner 8 and is of smaller width than the opposed second side wall 12. The bevelled corner 8 is edged by one longitudinal edge 16 of each first side wall 10. The casing 4 comprises a bevel wall 14 closing the bevelled corner 8. The bevel wall 14 connects the longitudinal edges 16 of the first side walls 10. The bevel wall 14 extends at an angle of 45° relative to each first side wall 10 edging the bevelled corner 8. In alternative, the angles between the bevel wall 14 and each one of the first side wall 10 are different. The module 2 comprises connection means 18 for connecting the casing 4 to the casing of another module identical or similar to module 2 and provided with corresponding connection means. The connection means 18 comprise connection members 20 provided on the longitudinal edges 16 of the first side walls 10 edging the bevelled corner 8. Each connection member 20 comprises a tab 22 projecting in cantilever from one of the longitudinal edges 16 and an anchoring rib 24 disposed at the free end of the tab 22. The tab 22 extends from the longitudinal edge 16 in the plane of the corresponding first side wall 10, towards the fictive missing edge of the square section of the casing 4. The tab 22 ends at a distance from the fictive missing edge. The rib 24 of each connection member 20 is elongated longitudinally and protrudes from the tab 22 towards the interior of the casing 4. The connection means 18 comprise a plurality of connection members 20 distributed along each one of the longitudinal edges 16 edging the bevelled corner 8. Each one of the connection members 20 provided on one longitudinal edge 16 is arranged at the same longitudinal position than one connection member 20 provided on the other longitudinal edge 16. In alternative, the connection means 18 comprise one single connection member 20 on each longitudinal edge 16. The fuel rods 6 of the module 2 are encased in the casing 4. The length of the casing 4 is substantially equal or superior to the length of the fuel rods 6. Each fuel rod 6 comprises in a know manner an elongated tubular cladding, nuclear fuel pellets stacked in the cladding and a pair of plugs closing the ends of the cladding. The fuel rods 6 extend parallel to each other in the longitudinal direction inside the casing 4 and are supported by the casing 4. The fuel rods 6 are arranged in a lattice and maintained transversely in spaced relationship inside the casing 4. The lattice is a 5×5 lattice of regular pitch with one fuel rod omitted in one corner due to the bevelled corner 8 of the casing 4. In alternative, the lattice may have a different amount of fuel rods 6, exhibit a varying pitch and/or have more than one fuel rod omitted depending on the size of the bevelled corner 8. In a known manner, some fuel rods 6 may be replaced in the lattice by part length fuel rods or water rods or by guide tubes. The module 2 comprises spacer grids 26 (FIG. 1) for supporting the fuel rods 6 longitudinally and transversely inside the casing 4. In the illustrated embodiment, each spacer grid 26 comprises intersecting strips 28 defining a plurality of cells 30 arranged in a lattice, each cell 30 being intended to receive one respective fuel rod 6. In a know manner, each cell 30 of the spacer grid 26 is provided with means for supporting one fuel rod 6 extending through the cell 30, such as springs and/or dimples formed in the strips 28 and/or assembled to the strips 28. Each spacer grid 26 is connected for instance to the side walls 10, 12 and optionally the bevel wall 14 of the casing 4 in a known manner, e.g. by complementary fittings and/or welding. Several spacer grids 26 are distributed along the length of the fuel rods 6. The module 2 is able to be handled as a single individual unit. The nuclear fuel assembly 32 illustrated on FIG. 3 is formed of four identical modules 2 as illustrated on FIGS. 1 and 2 disposed parallel and side-by-side in a 2×2 pattern and mutually connected. Each module 2 defines a portion of the cross-section of the fuel assembly 32. The modules 2 are oriented such that their bevelled corners 8 are adjacent thus defining a water channel 34 extending along a longitudinal axis A of the fuel assembly 32. The bevel walls 14 of the modules 2 define side walls of the water channel 34 which is thus closed laterally. The casing 4 of each module 2 is in contact by one of his first side walls 10 with the one of the first side walls 10 of the casing 4 of another module 2. The casings 4 of the modules 2 define together a channel box 36 of the fuel assembly 32 for conducting coolant flow along the fuel rods 6 in the longitudinal direction. The channel box 36 exhibits an outer square-shaped cross-section defined by the second side walls 12 of the casings 4. The channel box 36 is of larger cross-section that the casing 4 of each of the assembled modules 2. The channel box 36 is divided in four sub-channels 38 separated by a cross-shaped partition 40 defined by the first side walls 10 of the casing 4 and the bevel walls 14 of the casings 4. Each pair of first side walls 10 in contact defines one branch 42 of the partition 40. Each branch 42 extends from the water channel 34 to the channel box 36. The fuel rods 6 of each module 2 define a sub-bundle 44 of fuel rods 6 of the fuel assembly 32. Each sub-bundle 44 extends in a respective sub-channel 38. The fuel assembly 32 thus has a bundle of fuel rods 6 of larger cross-section than that of each of the assembled modules 2. The casing 4 of each pair of adjacent modules 2 are connected using their connection members 20. For each pair of adjacent first side walls 10, the connection members 20 of the two first side walls 10 are adjacent and located at the same longitudinal position. The connection members 20 of each pair are connected by positioning a connection piece 46 around the ribs 24 of the connection members 20 to prevent spacing between the connection members 20. The connection piece 46 is fixed to the connection members 20, e.g. by welding and/or crimping the connection piece 46 around the ribs 24. The fuel assembly 32 as illustrated on FIG. 4 differs from that of FIG. 3 in that it further comprises a tubular housing 48 surrounding the modules 2 for stiffening the fuel assembly 32. The housing 48 is tubular and has a closed square-shaped cross-section corresponding to the outer cross-square of the channel box 36 section defined by the second side walls 12 of the assembled modules 2. The modules 2 are connected to the housing 48 in a known manner, e.g. by form fittings and/or welding. Upon assembly, the modules 2 are assembled together, and then introduced into the housing 48 by sliding in the longitudinal direction The module 2 illustrated on FIGS. 5 and 6 differs from that of FIGS. 1 and 2 in that each first side wall 10 is formed with a groove 50 on the outer surface of the first side wall 10, said groove 50 extending in the longitudinal direction from the lower end to the upper end of the first side wall 10. The groove 50 is of general U-shape cross section with a large width and a relative small depth. The fuel assembly 32 illustrated on FIG. 7 results from the connection of four modules 2 as illustrated on FIGS. 5 and 6. The grooves 50 provided on each pair of first side walls 10 in contact define between said pair of first side walls 10 an internal water duct 52 extending longitudinally. Consequently, each branch 42 of the cross-shaped partition 40 is provided with one internal water duct 52, thus increasing the amount of coolant flowing separately from the sub-channels 38 and the fuel rods 6. The module 2 of FIGS. 8 and 9 differs from that of FIGS. 1 and 2 in that the casing 4 is deprived of bevel wall for closing the bevelled corner 8. The casing 4 thus has a longitudinal opening 54 delimited between the longitudinal edges 16 edging the bevelled corner 8. The fuel assembly 32 illustrated on FIG. 10 results from the connection of four open modules 2 as illustrated on FIGS. 8 and 9. The fuel assembly 32 comprises a tube 56 inserted in the free space defined by the bevelled corners 8 of the assembled modules 2 to define the water channel 34. As illustrated, the tube 56 extends longitudinally and has a square-shaped cross-section. The tube 56 is oriented to have its walls parallel to that of the channel box 36. The tube 56 can be provided with a larger area in cross-section than a water channel 34 formed by inclined bevel walls closing the bevelled corner 8. The tube 56 is optionally provided with a profiled reinforcing member 58 inserted inside the tube 56. The reinforcing member 58 is elongated in the longitudinal direction and has cross shaped cross-section with four branches extending from the centre of the tube 56 to the walls thereof. The tube 56 is connected to the first side walls 10 of the modules 2 in a known manner, e.g. by form fittings and/or welding. The module 2 of FIGS. 11 and 12 differs from that of FIGS. 1 and 2 by the connection means 18. The connection means 18 comprise first connection members 60 distributed along one of the longitudinal edges 16 edging the bevelled corner 8, and second connection members 62 distributed along the other one of longitudinal edges 16 edging the bevelled corner 8. Each first connection member 60 comprises a tab 64 projecting from the corresponding longitudinal edge 16 towards the fictive missing edge of the square section of the casing 4, and a sleeve 66 provided at the free edge of the tab 64 and adapted to be fitted onto a support rod as it will be detailed below. The sleeve 66 is tubular and extends in a longitudinal axis corresponding to the fictive missing edge of the square section of the casing 4. The tab 64 comprises an opening 68. The opening 68 is of rectangular outline elongated in the longitudinal direction. Each second connection member 62 is hook-shaped and adapted to be hooked in the opening 68 of a first connection member 60 of connection means 18 of another module 2 upon assembling the modules 2. Each second connection member 62 comprises a leg 70 extending from the corresponding longitudinal edge 16 towards the fictive missing edge of the square section of the casing 4 substantially in the plane of the adjacent first side wall 10 and a prong 72 extending from the leg 70 at an obtuse angle relative to the leg 70, towards the outside of the casing 4. Each second connection member 62 ends at a distance from the fictive missing edge of the square section of the casing 4. As illustrated on FIGS. 13 and 14, four modules 2 similar to that of FIGS. 11 and 12 are assembled to form a nuclear fuel assembly 32 having a bundle of fuel rods 6 of larger cross-section than each of the modules 2 forming the fuel assembly 32. Upon assembling the modules 2, each second connection member 62 of each module 2 engages into the opening 68 of a first connection member 60 of an adjacent module 2. Then, as illustrated on FIG. 14, the four modules 2 are arranged side-by-side in the longitudinal direction such that the sleeves 66 of their first connection members 60 are aligned in a longitudinal axis A and a support rod 74 is inserted through the sleeves 66 of the different modules 2. The sleeves 66 are fixed to the support rod 74, e.g. by form fittings and/or by welding. Consequently, each module 2 is connected to a common support rod 74 and to the adjacent modules 2. The connection between the modules 2 is stiff and the fuel assembly 32 is stiff. The modules 2 differs by the position of their first and second connection members 60, 62 to allow alignment of the sleeves 66 of the different modules 2 in a longitudinal axis and engagement of the second connection members 62 into the corresponding first connection members 60. As illustrated on FIG. 12, the first connection members 60 of each module 2 are spaced longitudinally with a pitch at least four times the longitudinal height of the sleeves 66. The second connection members 62 are spaced with the same pitch with being offset longitudinally relative to the first connection members 60 by a quarter of said pitch. Further, the connection members 60, 62 of each of the plurality of modules 2 to be assembled to form one fuel assembly 32 are offset longitudinally relative to the connection members 60, 62 of the other modules 2 such that the connection members 60, 62 of the different modules 2 insert between each other. Hence, a module 2 to be arranged beside the module 2 of FIG. 12 on the side of the first side wall 10 provided with the first connection members 60 is provided with second connection members 62 at the same longitudinal position than the first connection members 60 of module 2 of FIG. 12, and another module 2 to be arranged beside the module 2 of FIG. 12 on the side of the first side wall 10 provided with the second connection members 62 is provided with first connection members 60 at the same longitudinal position than the second connection members 62 of module 2 of FIG. 12. FIG. 15 illustrates the insertion and cooperation of the first and second connection members 60, 62 of the four modules 2 assembled to form the fuel assembly 32 of FIG. 14. The modules 2 and their components are differentiated by suffix letters A, B, C and D added to the numeral references. In alternative, the distribution of the first and second connection members 60, 62 may be different. For example, each longitudinal edge 16 edging the bevelled corner 8 of the casing 4 of each module 2 may be provided with both first and second connection members 60, 62 and/or the distribution of the connection members 60, 62 in the longitudinal direction may be regular or irregular. The sleeves 66 of the module(s) 2 of FIGS. 11-15 have a closed cross-section. In alternative, connection means 18 have sleeves 66 of open cross-section with a lateral aperture, namely hemi-cylindrical sleeves 66, to allow laterally inserting the support rod 74 into the sleeves 66 to ease assembly of several modules 2. Fuel assemblies have to undergo tests before they can be fabricated, delivered and used industrially in nuclear power plant. Providing a fuel assembly of modular conception formed of several modules having a general structure which is substantially identical, allows reducing tests to only one module or testing several different solutions, for instance one per module, thus reducing development and test costs for the fuel assembly. The features involved in the performance of the fuel assembly are namely the cross-section of the casing of each module and the lattice arrangement of the fuel rods of each module. Spacer grids may also influence performance, as well as optional mixing grids, which are similar to spacer grids but do not provide support for the fuel rods. In the different embodiments, each module can be handled as a unit thus making transport and/or handling easier. The modular conception also allows reducing maintenance or operation costs. For example, in case of failure of one fuel rod in a fuel assembly, the modular conception makes it easier to replace this fuel rod by replacing the module comprising the incriminated fuel rod. The modular conception further allows reducing costs of fuel cycle: one fuel assembly having burnt and partly burnt fuel rods can be reused after replacing one single module, namely the module including the burnt fuel rods. In the illustrated embodiments, the modules have casings of generally square shaped cross-section with one bevelled corner for delimiting a coolant channel in the centre of the fuel assembly. It is also possible to provide modules of generally rectangular shape with one bevelled corner, for example 4×5 or 5×6 modules. In a general manner, the modules have casings of generally quadrilateral shape. More generally, the modules have a casing exhibiting a cross-section having the shape of an angular sector of a polygon, preferably a regular polygon. The casings of such assembled modules define a channel box having the polygonal cross-section. For example, the casings exhibit cross-section of isosceles triangles with one bevelled corner, to obtain a hexagonal channel box. Modules can be provided in different sizes in cross-section whilst having complementary connection means. Modules having identical or different sizes can be assembled to form fuel assemblies of various sizes. For example a 8×8 fuel assembly may be obtained by assembling four 4×4 modules, a 10×10 fuel assembly may be obtained by assembling four 5×5 modules or by assembling one 6×6 module with two 6×4 modules and one 4×4 module or by assembling two 6×5 modules with two 4×5 modules. Similarly a 9×9 fuel assembly may be obtained by assembling one 5×5 module with two 4×5 modules and one 4×4 module or by assembling nine 3×3 modules allowing delimitation of one, two, three and even four water channels depending on the presence or not of one bevelled corner on each module. The examples above are given for the sole purpose of illustration and any fuel assembly lattices can be obtained be combination of adequate modules. Forming a fuel assembly by assembling modules of different sizes allow providing a water channel offset with respect to the central axis of the fuel assembly. Hence, a limited number of modules allows obtaining fuel assemblies of different sizes in cross-section, with various positions for a water channel. In a general manner, fuel assemblies are obtained by assembling modules of the same type having corresponding connection means allowing assembling the modules side-by-side to form a fuel assembly of larger cross-section than each module. The modules of the same type are identical or similar and differing e.g. by their general shape and/or size in cross-section and/or their connections means. The invention is particularly suitable for fuel assemblies for Boiling Water Reactors (BWR) since the casings of the assembled modules define a channel box for conducting coolant flow, and is also suitable for Pressurized Water Reactors (PWR).
06163588&	abstract	A modular differential pressure measuring system for a boiling water nuclear reactor pressure vessel is described. The modular pressure differential system includes a plurality of pressure lines having a plurality of pressure line sections. The system also includes a shroud having at least one replaceable shroud section. Each shroud section includes at least one pressure line section which is configured to connect to and disconnect from corresponding pressure line sections in adjacent shroud sections without welding. Additionally, the system includes a reactor bottom head petal section having a shroud support flange and a plurality of bores defining pressure line sections wherein at least one pressure line section of said bottom head petal is configured to couple with a corresponding pressure line section of an adjacent shroud section. The modular pressure system does not require cutting of the pressure lines or pressure line supports for replacement of the replaceable shroud sections. Additionally, the modular differential pressure system does not require welding of pressure lines and/or pressure line supports during installation of a replaceable shroud section.
	description	Referring now to FIG. 1, the essential principle of operation for the devices of the present invention is illustrated. FIG. 1 is a conceptual cross section view of a single neutron detector comprising a means for detecting neutrons 10 stacked on an absorbing layer 11. The absorbing layer 11, being composed of a first material that absorbs protons, such as titanium, is stacked on a hydrogenous substrate 12. Hydrogenous substrate 12 is composed of a second material having hydrogen atoms interacting with an unknown source of neutrons, indicated by box 13. When a single neutron detector is placed in a field of a neutron spectrum, the incident neutrons, indicated by arrow 14, from suspected neutron source 13 interact with hydrogen atoms within hydrogenous substrate 12. This interaction produces proton recoils that travel in fairly straight lines, one of which is indicated by arrow 15, through the absorber layer 11 and the detector means 10. Scattered neutrons, indicated by arrow 16, are deflected away from the hydrogenous substrate 12. Detector means 10 is connected to a data processing means, indicated by box 17, and a ground 18. The data processing means 17 includes a means for proton distribution. Using several detector means 10 with each absorbing layer 11 having a different thickness allows protons with energies and corresponding ranges greater than the thickness of a particular absorbing layer 11 to reach detector means 10 and produce proton counts. The amount of absorber layers 11 and their thickness can be selected to correspond to ranges of protons from a low value for 1 MeV and larger thicknesses of 250 MeV. Hydrogenous substrate 12 converts part of the kinetic neutron energy to energy of the recoil protons 15 and the detector means 10 detects protons passing through the absorbing layer 12. This approach is demonstrated by considering the energy transfer behavior of neutrons and protons. The maximum energy a neutron of energy En can transfer to a proton Ep (max) equals En (1,2). For this example, assume an absorbing layer 11 thickness of d. For monoenergetic neutrons (En), the number of recoil protons reaching detecting means 10 and producing proton counts decreases as energy En decreases. The number of protons will eventually equal zero when the range of maximum energy recoil protons becomes smaller than d. Recoil particles due to elastic scattering do occur in the higher atomic number non-hydrogenous absorber but, except for very high En, they do not contribute to the counts due to their small range and the unfavorable quantum energy transfer in elastic scattering. Having a system with K units, each with a different d and exposing them to a neutron spectrum, one obtains data which consist of K counts or count rate values Ci(di) i=1, 2, . . . K where for dixe2x88x921 less than di less than di+1, Cixe2x88x921 (dixe2x88x921) greater than Ci greater than Ci+1. From these numbers one can unfold the incident spectrum of neutrons. The detector means 10 can be of any shape or configuration and can be any type of solid state device. The inventors herein have employed a depleted n/p diode used to measure alpha particles, which was relatively insensitive to beta particles because of their low LET (Linear Energy Transfer) values as a detector means 10. Spectroscopic grade detectors are not required for this device since only event counting is required and data describing the energy spectrum are not needed. In considering the thicknesses of absorbing layers 11 and the ranges of protons to be measured, an energy range of 1 to 250 MeV was selected to match the expected neutron spectrum distribution. One solution to achieve this objective is to fabricate an instrument that converts a distribution of neutrons to one of recoil protons, which are charged particles that can be easily counted. By employing 12 detector means 10 within a given chamber, the recoil protons are essentially sorted into 12 bins where they can be readily counted. Said absorber layers 11 can be constructed of aluminum for detecting the lower energy levels or tantalum for the higher values. The hydrogenous substrate 12 for each detector means 10 could be constructed of polyethylene. The data processing means 17 and its means for proton distribution provides a hitherto unavailable capability to determine a proton distribution pattern to construct a neutron spectrum indicating the spectrum of neutrons from an unknown source of neutrons 13. In operation, results of a spectral measurement are a set of pairs from the detector means 10 and the absorbing layer 11 that allows protons with energies and corresponding ranges greater than the absorbing layer 11""s thickness to reach the detector means 10 and produce proton recoil counts. One data processing means 17 successfully employed by the present inventors is a 3-dimensional Monte Carlo Adjoint Transport code, NOVICE, which is described in Jordan, T., xe2x80x9cNovice, A Radiation Transport and Shielding Codexe2x80x9d, Experimental and Mathematical Physics Consultant, Report EMP. L 82.001, Jan. 1982. FIG. 2 is a chart showing plots of counts in the detector versus proton energy with different thicknesses indicated as a parameter on the curves, and these results were obtained using the NOVICE program and a flat spectrometer 20 depicted in FIG. 6, which will be described below. The FIG. 2 plots are counts in the detector versus proton energy with the aluminum and tantalum thicknesses indicated as a parameter on the curves. In this preliminary assessment of the feasibility of neutron monitor with multiple neutron detectors, an incident neutron spectrum and the subsequent unfolding software were not included in the code""s run. The proton recoil spectrum was assumed to exist in the converter material of hydrogenous substrate 12. The separation or resolution of proton energy shown in FIG. 2 provides useful information about detecting 12 ranges of neutron energy. The flat configuration of monitor 20, depicted in FIG. 6, along with the use of tantalum for the absorber layers 11 and for the chamber 21 make it too heavy for spacecraft or other airborne applications. Using a data processing device with the NOVICE computer software to analyze the monitor revealed other more useful potential configurations for neutron spectrometers, which were modeled and analyzed by the computer. One configuration suggested by the FIG. 2 NOVICE results is a pentagon dodecahedron, which allows for a full measurement range because of its 12 surfaces, each supporting a detector-absorber pair with different absorber layer thicknesses. FIGS. 3A and 3B, are perspective drawings depicting a detector means 41 stacked on a pentagonal absorbing layer 42 and a dodecahedron neutron spectrometer monitor 40, respectively. Referring now to FIG. 3A, which depicts a perspective view of a neutron detector comprising a detector means 41 stacked on an absorbing layer 42. Absorbing layer 42 is composed of a first material that absorbs protons, such as titanium. By placing this assembly on an appropriate hydrogenous substrate, a neutron detector is provided. Referring now to FIG. 3B, dodecahedron neutron spectrometer monitor 40 is depicted with 11 of 12 of the absorbing layers 42 with varying thicknesses stacked on a surface facet of a solid dodecahedron substrate 43, which provides the hydrogenous substrate. Dodecahedron substrate 43 is shown partially exposed without one absorbing layer for illustrative purposes. FIG. 4 is a front view drawing of the dodecahedron neutron spectrometer monitor 40 with all absorbing layers 51-62, respectively, covering each of the 12 facets of substrate 43 and representative dimensions. For the sake of clarity, only one detector means 41 is shown stacked on absorbing layer 54, with 11 other detector means 41 for the other 11 absorbing layers 51-53 and 55-62, respectively, not shown. Each of the 12 absorbing layers 51-62 are constructed with a varying thickness and are stacked on a surface facet of the solid dodecahedron substrate 43. Substrate 43 is composed of a hydrogenous material, such as polyethylene, having hydrogen atoms and functions as a neutron converter when interacting with said absorbing layers 51-62 in the presence of an unknown energy distribution, indicated by box 44, which emits incident neutrons, indicated by arrow 63. In operation, said hydrogenous substrate 43 converts said neutrons to recoil protons and each of said detector means 41 detects recoil protons passing through each absorbing layer 51-62, respectively. Each absorbing layer 51-62, respectively has a different thickness, as depicted in FIG. 5, to absorb neutron energies from 1 to 250 MeV. Returning now to FIG. 4, the hydrogenous substrate 43 is housed in a concentrically hollow spherical chamber, indicated by broken line 45. Each detector means 41 is coupled to a means for data processing, indicated by box 46, outside the spherical chamber 45, which provides a count of recoil protons to a means for proton distribution, not shown, residing within said data processing means 46. The means for proton distribution determines a proton distribution pattern to construct a neutron spectrum pattern indicating the spectrum of neutrons from said suspected source of neutron radiation 44. FIG. 4 also includes representative dimensions. Each absorbing layer 51-62 is pentagonally shaped in this embodiment, with each side 2.03 cm in length. Each of said detector means 41 are circular and 0.5xe2x80x3 wide and 0.015xe2x80x3 thick. Covered hydrogenous substrate 43 is 4.47 cm in height and housed concentrically within hollow spherical chamber 45. Hydrogenous substrate 43 was fabricated from a solid block of Lucite(trademark). The hollow spherical chamber 45 is composed of titanium in this embodiment with an inner diameter of 10.8 cm and a wall thickness of 2.5 cm. Each of said 12 absorbing layers 51-62 is composed of titanium in this embodiment with a varying thickness ranging from 0.00105 cm to 2.4217 cm, as described in Table I below. Detector means 41 can be constructed from a depleted n/p diode. It should be understand to those skilled in the art that these dimensions are merely representative and numerous other choices of dimensions are possible. FIG. 5 is a perspective drawing of hydrogenous substrate 43, using like numerals for similar structural elements, illustrating a number of absorbing layers with a varying thickness. In this drawing, covered hydrogenous substrate 43 is shown removed from the hollow spherical shell 45 to better illustrate each absorbing layer having a different thickness. Referring back to FIG. 2, which is the chart showing plots of counts in the detector versus proton energy with different thicknesses indicated as a parameter on the curves from the NOVICE program. Those plots from the FIG. 6 flat spectrometer. 20, which will be described shortly, are based on using aluminum and tantalum as absorber material. These results suggested using titanium as the preferred absorber material for the FIG. 4 absorbing layers 51-62 for all energy levels, because titanium is lighter than tantalum and its neutrons do not generate nuclear interactions. Only elastic scattering takes place. The proton energy resolution from this embodiment is also relatively good. The FIG. 2 results also indicate that aluminum absorbers produced a slightly better energy resolution for the lower range of energies, 1 to 10 MeV. The size of this dodecahedron configuration is small and light in weight and very practical for a spacecraft application. In order to insure that an unknown neutron spectrum has an isotropic distribution, the spectrometer 40 can also be located at the center of a titanium sphere with a diameter of 3 inches. FIG. 6 is a perspective conceptual drawing of the flat embodiment of the present invention""s neutron spectrometer monitor 70. Monitor 70 comprises a group of the FIG. 1 neutron detector means 10 arranged in a chamber 71. As described above, having several detector means 10 stacked onto absorbing layers, not shown, each having a different thickness, allows protons with energies and corresponding ranges greater than the thickness of each absorbing layer to reach the detector means 10 and produce proton counts. FIG. 6 depicts 12 detector means 10 which correspond to 12 energy bins and thus detect protons with ranges corresponding to energies from 1 MeV up to 250 MeV. The floor of chamber 71 serves as the hydrogenous substrate. Monitor 70 is placed in proximity to an unknown source of neutrons, shown as box 76. Detecting means 10 is coupled to a means for data processing, indicated by box 77, and provides a separate count of recoil protons for each different thickness employed in the absorbing layers. The data processing means 77 transmits the count of recoil protons to a means for proton distribution, not shown, residing within the data processing means 77. The means for proton distribution determines a proton distribution pattern to construct a neutron spectrum pattern indicating the spectrum of neutrons from the suspected concentration of neutrons 76. Bulkhead output connector 72 on the chamber 71 allows correction of voltage to the detector as well as correction of output counts to counting instruments. In the flat configuration, said chamber 71 is shown in a rectangular shape, and its walls 78, lid, not shown, and unit compartments 79 can be composed of tantalum. Each detector means 10 in the egg-crate-like structure is numbered 1xe2x80x2-12xe2x80x2, respectively, to correspond with readings shown in the FIG. 2 chart. Detector means 7xe2x80x2 is depicted with representative dimensions of 2 cm in width and 2 cm in length. A gap 80 between detector means 11xe2x80x2 and 12xe2x80x2 is 0.471 cm. The thickness of each wall 78 is 1 cm and its height is about 3 cm. The chamber 71 is depicted as 15 cm in length and 5.41 cm in width. These dimensions are merely representative and numerous other choices of dimensions are possible, however, it is critical that each absorber layer is constructed with a different thickness according to the minimum and maximum energies of neutrons in the spectrum. Similarly, the materials used for constructing the absorber layers, detector means 10 and chamber 71 can also be varied according to the minimum and maximum energies of neutrons in the spectrum. It is to be understood that details concerning materials, shapes and dimensions are merely illustrative, and that other combinations of materials, shapes and dimensions can also be advantageously employed and are considered to be within the contemplation of the present invention. We also wish it to be understood that we do not desire to be limited to the exact details of construction shown and described. It will be apparent that various structural modifications may be made without departing from the spirit of the invention and the scope of the appended claims.
	abstract	A method and apparatus are disclosed for accelerating ions in an ion implantation system. An ion accelerator is provided which comprises a plurality of energizable electrodes energized by a variable frequency power source, in order to accelerate ions from an ion source. The variable frequency power source allows the ion accelerator to be adapted to accelerate a wide range of ion species to desired energy levels for implantation onto a workpiece, while reducing the cost and size of an ion implantation accelerator.
	summary
052746828	description	DETAILED DESCRIPTION In FIGS. 1A and 1B there is shown both a PWR and a BWR fuel rod assembly 11, comprising a plurality of fuel rods 12 mounted in a skeleton 13 which comprises a top end member 14, a plurality of grids 16, and a plurality of guide tubes 17 which extend along the approximately fourteen foot length of the fuel assembly 11. The BWR fuel rod assembly 11 is surrounded by a removable can or shroud in a rod assembly or cell 18 of generally rectangular cross-section. This can or shroud is removed prior to any consolidation operation. Member 14 is removed, as by cutting or reaming prior to removing the rods from the assembly 11 on PWR's. A BWR fuel rod assembly is similar except fuel rods (typically 8) are used in place of the guide tubes to tie the top and bottom end members 9 and 14 together. The top nuts are typically removed to allow the top member 14 to be placed aside prior to removing the rule rods. The present invention is directed toward the method of removing the rods from skeleton 13 and precisely packing them in a fuel rod storage canister, and to the fuel rod storage canister which is not shown in FIG. 1. In FIGS. 1C and 1D there are shown typical PWR and BWR fuel rods each comprising an elongated hollow tube 19 of suitable material, such as Zincalory, filled with fuel pellets, not shown. At the top end of the tube 19 is a first cap member 21. The BWR cap member 21 is provided with an elongated tip portion 22 which supports a coil spring 23, crimped in place on the elongated portion 22. When the rod 19 is in the fuel cell within the reactor, it tends to grow in length. Spring 23 insures that the rod 19 remains firmly in place within the cell despite the growth. Cap member 21 is welded to and seals the top portion of tube 19, resulting in a weld bead 24 which is slightly larger in diameter than the tube 19. The lower end of tube 19 is sealed by a second cap 26 having an elongated tip portion 27 ending in a tapered tip 28. Cap 26 is welded to tube 19, resulting in a weld bead 29. Weld beads 24 and 29 can prevent fuel rods from being positioned closely together within a fuel storage canister, except where, as with the present invention, the rods are vertically displaced relative to adjacent rods. FIG. 2 depicts the various elements of the rod consolidation system 30, a majority of the components of which remain below the water line 31 of the storage pool 32. At the top of pool is a deck 33 adjacent thereto which has thereon the major control components of the system 30. These components comprise a computer 34 which controls a five or six axis commercially available robot 36 which, in turn, handles the long reach tools for the system 30 and for the method of rod consolidation. All of the functions of the system and steps of the method, with a few exceptions, are performed by the robot 36 and its associated long reach tools under control of the computer 34. Adjacent computer 34 is a monitoring station 37 which includes a closed circuit television monitor 38, the signals for which are received from a plurality of underwater television cameras 39, as will be explained more fully hereinafter. A protective wire cage 42 protects both the operator and the equipment from any accidental contact with the robot 36. Attached to the free or distal end 43 of the arm of robot 36 is a shaft 44 having, at its lower or distal end 46 a quick change coupler 47. Quick change couplers are commercially available items, and any of a number of types of such couplers may be used. A bracket 48 mounted on a curb at the top of the pool 32 has mounted thereto first and second tool racks 49 and 51 for holding a plurality of long reach tools 52, 52 and 53, 53, each having, at its top end, a quick change coupler 54 and 56 that matches quick change coupler 47. Each of the tools 52 and 53 is designed to perform a specific task, and when that task is to be performed, the robot removes that tool from the rack by means of the coupling, causes it to perform the task, and returns it to the rack. This arrangement has the important advantage of enabling almost all of the steps of the consolidation process to be performed within the pool, without the necessity of active human intervention. Also mounted to bracket 48 is a depending frame member 57 to which is mounted a work table 58, shown exploded in FIG. 2. Alternatively, the work table 58 can be supported by the header 64 and joined thereto by an appropriate connecting structure. Work table 58 has mounted in apertures therein four fuel rod assembly holders 18, 18, 18 two scrap canister holders 61, 61, and two fuel rod canister holders 62 and 63. Holders 18, 62, and 63 rest in apertures in a support base or header 64, which also functions as a manifold for a pair of vacuum filter assemblies 66 and 67, each comprising a pump 68, 69 and filtering element 71, 72. Also mounted on work table 58 adjacent one of the scrap canister holders 61 is a grid compacting apparatus 73 and mounted on table 58 adjacent another of the scrap canister holders 61 is a guide tube chopper and compactor 74. Both compactors 73 and 74 have foldable chutes (not shown) for emptying the compacted trash into its adjacent scrap canister. The remaining trash canister in its holder 61 is for other scrap that is not compacted. Television cameras 39, 39 and 39 are mounted to frame 57 above the table 58. The cameras are commercially available items having zoom lenses and integral lighting contained in waterproof housings. The cameras monitor the operation of the system, and more particularly, the location of the long reach tools 52 and 53 during operation. It is possible, using an appropriate tool calibrator fitted with proximity switches and located at the work table elevation in conjunction with the computer 34 and the robot 36, to position the distal or operative end of each tool to within twenty one-thousandths of an inch, thereby exceeding any accuracy obtainable when the tools are manipulated by other means. Mounted on table 58 at the corners thereof are locator pins 83. One of the long reach tools carries an electro-magnetic locator member thereon. Before operations are begun, this locator is placed over each of the locator pins 83 in turn and it generates an electrical signal which is transmitted to the computer. The combined inputs of the locating pins 83 enables the computer to determine the precise location of all of the various elements of the work table. FIG. 3 is an perspective partial view of the apparatus of the present invention, as used for locating a pulling tool over a fuel rod assembly, and for locating the tool over a fuel rod canister in which the rods are to be compacted. As shown in FIG. 3 a funnel plate guide 101 having a top portion 103 and a plurality of spaced indexing pins 100 on the under side thereof is placed over a fuel rod assembly holder 18. At least two of holders 18 have flanges 50 formed at the top thereof. Flanges 50 have a plurality of indexing holes 55 drilled therein into which indexing pins 100 fit. The spacing of holes 55 and pins 100 is such that guide plate 101 can be indexed to any of four positions in the embodiment of FIG. 3. As shown in FIG. 3, top portion 103 has an array of funneled holes 113, 113 therein and extending through plate 101. These funnels 113 function to guide the rod removing tool so that it is centered over a fuel rod in holder 18. Because of the funnel shape of the holes 113, guide plate 101 can only guide the tool over every fourth rod in the fuel rod assembly, but with the four position indexing provided by holes 55 and pins 100, the tool eventually accesses all of the rods. The guide plate 101 has an upwardly extending pin 76 as seen in FIG. 3 and is moved by one of the long reach tools adapted to pick it up by pin 76 and move it to a new position where pins 100 engage different holes 55. The rod pulling tool has a rod grasping means, such as a collet, not shown, which grasps the rod over which it is centered and pulls it up inside of the tool for substantially its entire length. As the rod is pulled out of the rod assembly, radioactive "crud" is scraped from the rod, which is pulled down through the holder by the water current created by pumps 68 and 69 and forced through filter elements 71 and 72. This action assures that the water will stay sufficiently clear for the monitoring cameras 39 to create a clear picture and, more importantly, does not allow radioactive crud to contaminate the pool water. After a rod has been pulled from a fuel rod assembly and drawn up into the rod pulling tool, the tool is swung to a position over the canister holder 63 containing a storage canister 116. Mounted on the top edges of a canister 116 within holer 63 is a guide plate 77. Plate 77 has a plurality of funnel shaped holes 78, 78 and a grasping pin 79 the same as with plate 101 in FIG. 3. On the underside of plate 77 are downwardly extending spaced ridges 81 which form locating grooves (not shown) extending parallel to each side of plate 77. The top edges of the canister 116 fit within the grooves, thereby giving plate 77 a plurality of possible index positions. The ridges or flanges 81 do not extend to the full length of each side, thereby providing clearance for the corners of the canister. Other means of locating and indexing guide plates 101 and 77 may be used as shown, for example, in the aforementioned U.S. patent application Ser. No. 07/831,404. After the rods have been removed from the rod assemblies and placed in the canister 116 in holder 63, the skeleton remaining in holder 18 is converted to scrap, by means of long reach cutting tools an compactors 73 and 74. As portions of the fuel assembly are cut away, means, not shown, are provided for raising the skeleton in holder 18 so that the cutting tools have access thereto. The operation of the apparatus of FIGS. 2 and 3 is as follows. An empty fuel rod canister 116 is placed in holder 63 and three or four fuel rod assemblies 11 are transferred, under water, to holders 18, and empty scrap canisters are transferred to holders 61 less their lids. A stack of lids for the fuel rod canisters and a second stack of lids for the scrap canisters have been previously stored in wells (not shown) located in the work table. The robot next selects a long reach tool having a cutter or cutters on the distal end thereof, and the top 14 of the PWR fuel rod assembly 11 is cut away and placed into a scrap canister. Alternately, a long reach tool having a nut runner on the distal end thereof is used to remove the retaining nuts and the top 14 of a BWR fuel rod assembly is removed and placed into a scrap canister. After returning the cutting tool or nut running tool to the rack 49 or 51, the computer directs the robot to couple with a suitable tool to place funnel guide plates 101 onto holder 18 by means of grasping pins 76 and guide plate 77 onto holder 63 and then return the tool to rack 49 and 51. The robot then couples to a fuel rod transfer tool, the distal end of which is guided to the fuel rods by guide plate 101. The tool grasps the top of a rod and pulls it up out of the rod assembly into the tool. After the rod is within the tool, robot 36 swings the tool to a position over guide plate 77 so that the tool may release the rod into holder 63. The canister 116 in holder 63 is preferably, although not necessarily, filled from the center outwardly. After enough rods have been removed and deposited in the canister to fill one-fourth of the rod positions, guide plate 77 is indexed to a second position. The depositing of the rods and the indexing continues until the canister contains all of the rods of the rod assembly. After all of the rods exposed by the holes 113 in plate 101 have been removed from the fuel rod assembly, plate 101 is indexed to a different position, thereby exposing a new set of rods. The process continues until all of the rods have been removed from the fuel assembly, and then operations are started on a second rod assembly and continued until the canister is filled. In practice, the rods of two or, in some cases, the rods of slightly more than two fuel assemblies can be consolidated into a single canister, thus making better use of the available storage space. In FIGS. 4A, 4B and 5 there is shown the bottom portion of the storage canister 116 of the invention as mounted in canister holder 63, which, in turn, is mounted on header or support base 64. The bottom end of holder 63 includes ducting 117 for introducing a current of water into canister 116 and for flushing it, as will be discussed hereinafter. Canister 116 comprises side walls 118 and an end wall 119 having a plurality of centrally located apertures 121, 121, as best seen in FIG. 4B. Mounted in place over holes 121 is a screen filter 122 for filtering "crud" and other matter. Apertures 121, 121 can be replaced by a single large centrally located aperture, but a plurality of smaller apertures affords better support for filter 122, which is of fine mesh and is somewhat fragile. Also located in end wall 119 are a plurality of spaced apertures 123, 123 as best seen in FIG. 4B. Within canister 116 and vertically slidable with respect thereto is an apertured base plate 124, the apertures 126, 126 of which are spaced forty-five degrees from the apertures 123 of end wall 119, as best seen in FIG. 4B. Base plate 124 also has a centrally located aperture 127 and a plurality of mounting holes 128, 128 for mounting pedestals 129, 129 in the form of shoulder screws and riser caps 131, 131 thereon. The orientation of the various apertures in base plate 124 is best seen in FIG. 4C. Mounted on base plate 124 and spaced therefrom by spacers 132, 132 is an array of spaced apertured parallel plates comprising a bottom plate 133, an intermediate plate 134, and a top plate 136, maintained in spaced relationship by spacers 137, 137 mounted on pedestals 129, 129, as best seen in FIG. 4A. Alternatively, the combination of pedestals, spacers, plates and caps may be in the form of welded or riveted construction. As best seen in FIGS. 4D, 4E and 4F, bottom plate 133 has a plurality of spaced apertures 138, 138 therein of sufficient diameter to permit passage therethrough of a first group of rod caps 26 to where the tips 28 rest upon the top surface of base plate 124. Intermediate plate 134, as seen in FIG. 4E has a greater number of spaced apertures 139, 139, a first group of which are aligned with apertures 138, 138 in plate 133, and a second group of which are offset therefrom. Thus, some rod caps or tips 28 pass through both plates 133 and 134 to rest upon plate 124, while others pass through plate 134 to rest upon the top surface of plate 133. However, adjacent rods do not have their tips 28 resting on the same surface. In a like manner, top plate 136, as seen in FIG. 4F, has a plurality of apertures 141, 141, a first group of which are aligned with both the apertures 138 and 139, a second group of which are aligned with the apertures 139 of the second group that are not aligned with apertures 138, and a third group of which are offset from the apertures 139 so that some rod tips 28 pass through apertures 141 to rest upon the top surface of plate 134. Finally, some rod tips 28 rest upon the top surface of top plate 136. As was pointed out heretofore, adjacent rods are vertically staggered with respect to each other so that the weld beads 24 and 29 do not interfere, and so that the grasping tool may grasp a rod without interference. Thus, the rods 19 can be placed close together and are readily insertable within the canister 116. Thus, for every pair of adjacent rods on a line in any direction, one rod is elevated with respect to the other. Riser caps 131 are provided for support of the typically eight special end caps 26 which are furnished with threads on elongated tip portion 26 to screw into bottom end member 9. The threaded portion is cut free from end cap 26 to release end member 9. The risers 131 are of sufficient length to locate the weld beads 24 and 29 at the desired elevation relative to other weld beads. Incorporated into holder 63 are spaced riser pins 142, 142 which pass through apertures 143, 143 in end wall 119 into the interior of canister 116 when the canister is placed in holder 63, thereby elevating base plate 124 and the array of plates 133, 134, and 136, during the loading operation as best seen in FIG. 4A. With the base plate 124 thus elevated, the primary downward flow as required to keep crud from escaping into the pool may flow within the canister walls 118, around edges of plates 133, 134 and 136 through openings 126 and 127 in base plate 124, through holes 123 in the bottom canister wall 119, through large conduit 145, which leads to the system filters and pumps. In addition, the top ends of the rods are elevated above the top end of the canister 118 to allow free access for the rod grasping tool. Also, with base plate 124 thus elevated, a smaller secondary water current may flow through conduit 117 and apertures 121 and out apertures 123, as shown by the arrows, thereby keeping the "crud" from clogging up screen 122 during the loading operation. As shown in FIG. 5, when the canister has been fully loaded, and is then lifted off of holder 63, base plate 124 is no longer supported by pins 142, and therefore it rests against the top surface of end wall 119, sealing off apertures 123, 123, so that water flow thereafter, while the loaded canister 116 rests in the pool, is through the aperture or apertures 121 equipped with screen filter 122 and upward through the canister. To facilitate this flow, plates 133, 134 and 136 preferably have scalloped edges, as shown in FIGS. 4D, 4E and 4F. Riser pins 142 are shown as being part of holder 63, being affixed thereto. Other means for elevating base plate 124 and its array may readily be used, however. For example, pins 142 may be made vertically movable by suitable mechanical or other means, or something other than riser pins might be used. The principal purpose is to permit sufficient water flow to remove "crud" during the canister loading operation and to allow plate 124 to rest against the top surface of end wall 119 after removal of canister 116 from holder 63. In FIG. 6A, there is shown a plan view of the interior of canister 116 showing the orientation and placement of the rod location and support means. A plurality of pairs of corrugated sheets preferably of stainless steel and extending along a major portion of the length of the canister 116 are affixed to the walls of the canister by, for example, rivets 146, 146, or other suitable means. Each pair of corrugated sheets comprises a first sheet 147 having indentations 148, 148 facing in one direction and a substantially identical corrugated sheet 149 having indentations 151, 151 facing in the opposite direction so that together the two sheets of each pair define a row of rod locations 152, 152. Also, the reverse side of each sheet 149 forms, with the reverse side of the first sheet 147 of the next adjacent pair, a row of rod locations 153, 153. In addition, the reverse side of the first sheet 147 of the first pair defines, with the wall 118 of the canister 116 a row of rod locations 154, 154, and the last sheet 147 also forms, in conjunction with wall 118, a row of rod locations 156, 156. As shown in FIG. 6A, six sheets 147 and five sheets 149 define twelve rows of rod locations. In order that insertion of the spent fuel rods into canister 116 in the rod locations defined by the corrugated sheets may be facilitated, sheets 149 are vertically offset relative to sheets 147, as shown in FIG. 6B so that they extend upward past the ends of sheets 147. The last corrugated sheet 147 on the right, as viewed in FIGS. 6A and 6B has the adjacent canister side wall 118 extending past its upper end. In operation, when the robot 36 inserts a rod into canister 116, the rod is pressed slightly against the higher extending sheet 149 at the desired rod location, thereby insuring that the tip 28 of the rod will be guided as it is lowered between the sheets 147 and 149. After canister 116 has been loaded with rods, as discussed hereinbefore, canister 116 is closed and capped by a lockable closure member. In FIG. 7A and FIG. 8, there is shown a closure member 161 with associated tamper indicators 162, with member 161 being lockable within slots 163 in the walls of canister 116 adjacent the top edge 164 thereof. Closure member 161 comprises a top plate 166 and a bottom plate 167 affixed to, and maintained in spaced relationship by a spacer member 168. At the four corners of member 168 are posts 169, 169 each having a pair of bores 171, 171 extending therethrough. Plate 167 is dimensioned to fit snugly within canister 116, and plate 161 is made slightly larger to rest upon top edge 164 of canister 116. Top plate 161 has affixed to the top surface a bail member 172 for facilitating lifting of closure member 161, an when member 161 is locked to canister 116, lifting of the entire assembly. The lifting bail is similar to the bail wheel and is an integral part of a BWR fuel assembly. Other lifting members having a different configuration may be used when consolidating PWR fuel assemblies. Also located on the top surface of plate 161 are a pair of stand-offs 173, 173 of the same height as member 172, so that a second canister lid may be supported on top of canister 116, where stacking of unused canister lids is feasible. Journaled within the bores 171, 171 are a plurality of cam lock member 174, 174, the configuration of each of which is best seen in FIGS. 7A and 7B. Each member 174 comprises a tubular member 176 which receives a shaft 177 which is journaled within bores 171, 171 in member 168. Four screws, not shown, used to attach member 167 to 161 also serve as stops to retain shaft 177 within 161. Near either end of the tubular member 176 and extending along a portion of the length of tubular member 176 are two locking members 178, 178 and a centrally located actuating member 179, oriented as shown. Members 178, 178 are dimensioned to fit within their corresponding slots 163, 163. Threaded through top plate 166 and located so that their ends bear against the actuating members 179 are actuating bolts or screws 181, 181. Each bolt or screw 181 has affixed, near the top thereof, a tamper indicator 162, and at the top end a shaped member 182 for providing purchase for a long reach tool used to rotate the screws or bolts 181. As each bolt 181 is screwed into and through top plate 166, its end forces actuating members 179 down, causing shaft 176 to rotate and the locking members 178, 178 to rotate upward into slots 163, 163 until they bear against the top edge of slots 163, 163. When the four members 174 are rotated to where each locking member 178 bears against the top edge of its corresponding slot 163, closure member 161 is firmly locked in place, and the canister 116 is sealed by plate 167. This plate is equipped with a tubular screen filter 170 to prevent coolant flow from carrying crud particles into the surrounding pool. The coolant flow therefor enters at the base screen filter 122, flows upward through the canister via natural convection, through the top screen 170 and radially outward through spaces provided along sides of locking members 178 and their wider slots 163. In assembly of the locking mechanism, each screw is driven through plate 161 and a keeper 185 is affixed to the end thereof. Once the canister 116 containing the spent rods 21 is sealed, the U.S. government requires that it not be reopened when in its storage mode or phase. It is also necessary that any attempt to re-open it produce an irreversible indication that such an attempt was made. To this end, each tamper indicator 162, as shown in FIGS. 9A and 9B, is shaped like a spoked wheel having an outer ring 183, hub 184, and spokes 186. Hub 184 has a central bore 187 to allow member 162 to be fitted on and affixed to bolt 181. Member 162 is preferably made of a reasonably soft metal such as fully annealed stainless steel. Located within plate member 166 adjacent each of the bolts 181 is a spring loaded pin 188, and on the underside of each indicator member 162, in the outer ring 183, is a detent recess 189. Each spring loaded pin 188 is spaced from the screw a distance such that it is directly under outer ring 183. Each indicator member 162 is located on screw 181 such that just when locking member 178 engages the upper edge of slot 163, pin 188 pops into the detent recess 189. Any subsequent torque applied to bolt 181, in either direction, will result in a permanent distortion of indicator member 162, inasmuch as it is prevented from turning with bolt 181 by pin 188 riding in recess 189. It may be the case that at some stage prior to loading rods into canister 116 it is necessary to check, for example, the proper positioning of member 162 on bolt 181, without distorting member 162. To this end, outer ring 183 is provided with a pin hole 191 extending therethrough to the recess 189. When spring loaded, pin 188 is engaged in recess 189, it can be depressed out of engagement by a suitable tool inserted into pin hole 191 so that bolt 181 and member 162 can be backed off. However, the hole 191 is sealed shut as by soldering or braising prior to the actual loading operation. This seal also affords a tamper indication inasmuch as it is removed or drilled, the evidence of tampering is apparent. The present invention has been disclosed in a preferred illustrative embodiment thereof. Various changes, modifications, or alterations may occur to workers in the art without departure from the spirit and scope of the invention.
	summary
	description	This application claims priority to U.S. Provisional Application Ser. No. 61/479,082, filed Apr. 26, 2011, which is incorporated by reference herein in its entirety. Spent nuclear fuel has historically been stored in deep reservoirs of water, called spent fuel pools, within nuclear power plants. This spent fuel storage technology is often termed “wet storage.” Spent fuel pools at reactors are reaching their spent fuel capacity limits, causing concerns about the need to shut down reactors because there is no more room for the spent fuel. Dry nuclear spent fuel storage technology (termed “dry storage”) is deployed throughout the world to expand the capabilities of nuclear power plants to discharge and store nuclear spent fuel external to a reactor's spent fuel pool, thereby extending the operating lives of the power plants. Two classes of technology are used in dry spent fuel storage: metal casks with final closure lids that are bolted closed at the power plants after loading with spent fuel, and concrete storage casks containing metal canisters having canister final closure lids that are welded closed or sealed with mechanical methods at the power plants following spent fuel loading. This latter dry storage technology is referred to as canister-based concrete spent fuel storage. The concrete cask serves as an enclosure or overpack structure that provides mechanical protection, heat removal features, and radiation shielding for the inner metal canister that encloses the radioactive material. The use of this technology tends to have significant capital cost and other economic advantage over the use of metal cask technology for storage. However, for transport of spent nuclear fuel, metal casks are still the preferred technology. For dry, spent nuclear fuel transport, two fundamental classes of technology are used: (i) metal casks with final closure lids (or lid) that are bolted closed at power plants or other facilities after loading of the spent fuel into open compartments of a separate structure nested within the cask body (termed the “basket”); this technology when used for spent fuel shipment is termed “bare fuel” transport; and (ii) similar metal casks with bolted final closure lids (or lid) having the metal canister used in dry storage within the cask body, the canister containing the basket structure and the final closure lid (or lids) installed at the power plants or other facility following spent fuel loading; this technology when used for spent fuel shipment is called “canistered fuel” transport. Throughout the technology considerations involving dry storage and transport of spent nuclear fuel discussed above, there is an issue that is common for each and that is of prime importance in assuring the long term integrity of the casks and canisters, as well as the spent nuclear fuel in dry storage or during transport. That consideration is the drying process for the fuel and containers that must be performed prior to the final closure of the canister or the metal cask at the power plant. Arriving at a condition within the spent fuel container (cask or canister) where all moisture below a specified minimum value has been removed is a most important step for the long term integrity of the container and the spent fuel. The process to accomplish this involves a careful balance of moisture removal and heat removal from the spent fuel container so that the cladding of the spent fuel is not subject to elevated temperatures or temperature cycling that may result in cladding degradation and a limitation on that cladding's structural stability, a critical characteristic that is determinative of its long term integrity. Over the last four decades, the process that has proven itself for moisture removal from spent fuel containers has been vacuum drying. That process is still the most efficient and safest process for insuring sufficient moisture removal from spent fuel containers. However, because the nuclear fuel cycle has now advanced to a condition where spent fuel has been burned in reactors at higher uranium (235 U) enrichments and for much longer periods (as measured in terms of the megawatt-days per ton of uranium produced by the fuel) and where it is necessary to place spent fuel into dry storage or transport containers after much shorter periods of cooling, the heat generation rate (heat-rate) of the spent fuel within containers is much higher than has been the case for the last 4 decades. With such a developing situation, one of the key strengths of vacuum drying, the production of fuel heat within the container, which greatly assists in moisture evaporation, could become a liability because fuel temperatures may reach unacceptable levels, or an exclusive vacuum drying process may allow fuel temperatures to vary outside of acceptable ranges. It is, therefore, the purpose of this method to provide an integrated system of vacuum drying with ancillary features and functions that make full use of the proven and safe drying features of current systems, while integrating such features/functions with new and unique methods, systems, processes and procedures that allow vacuum drying to work at its productive best while protecting the spent fuel and enhancing the vacuum drying moisture removal. Dry spent fuel storage and transport system designs must generally comply with governmental regulatory requirements. As part of the implementation of these requirements, regulatory bodies have issued design limitations on the allowable temperature of the spent nuclear fuel during the loading, drying, and closure of the canisters or transport casks, during the storage of the systems at the power plants, and during the transport of the spent fuel away from the plant. Drying of the spent fuel and the inside of the containers that store or transport it is a significant operational and regulatory consideration in order to assure there is little, if any, retained moisture that could produce long term degradation of the fuel or of the system that contains the fuel. When dry storage canisters and/or metal casks are being prepared for closure and moisture is being removed to dry the spent fuel in preparation for storage or transport, the regulatory requirements stipulate both a maximum allowable fuel temperature and a maximum range of temperature variation that the spent fuel is allowed to experience. Storage and transport systems, in combination with the ancillary drying systems used to remove the moisture, are designed to limit fuel temperatures and temperature variations while still fulfilling the drying function. These temperature limits help assure that the material properties of the spent fuel cladding are maintained in a safe and predictable range. However, controlling the maximum temperature of the spent fuel and the range of its variation during fuel drying in preparation for storage or transport can become a difficult technical design task when the spent fuel has in-reactor burn up and/or post-reactor cooling period characteristics that cause it to have high heat generation rates. Such is now more often the case, as utilities are generating large amounts of spent nuclear fuel that have been exposed to longer periods in-reactor to extract more energy from the contained uranium (known as high burn up fuel), and that cannot be kept in wet storage for as long as desired because of spent fuel pool capacity or other regulatory limits. Of the methods for drying closed spent fuel containers, vacuum drying is the foremost process having a demonstrated effectiveness over many years to achieve spent fuel and container dryness. Vacuum drying uses simple and proven systems and equipment, does not require any special or unique re-orientation of the spent fuel container, and introduces no special chemicals that might raise concerns of material interactions and integrity. For these reasons, vacuum drying is also the most widely used process for drying. Vacuum drying uses pumping systems to reduce the pressure within the closed spent fuel container system, so that, even with very low heat rates from the spent fuel, liquid water will flash to water vapor and, together with any other gases, be removed by the very system that is establishing the vacuum within the spent fuel container. In one example, the pressure can be reduced from approximately atmospheric pressure to a pressure range of a few millibars (e.g., tenths of an inch Hg, ten millibars, etc.). Dryness is also easily measured once a vacuum has been established. If the system pressure, without pumping, remains stable over a period of time, no further conversion of water to vapor is occurring and the closed system is dry. A further and unique advantage of vacuum drying is displayed when some spent fuel rods having small cracks or holes that permit water from the spent fuel pool to leak into the rods during wet storage (such rods are termed “water-logged rods”) must be placed into dry storage or transported. The vacuum drying process results in a pressure differential across the spent fuel cladding of water-logged rods, that pressure differential providing the motive force to expel water from inside the rod into the container, so that it can be removed as vapor to enhance dryness within the container. In such a case, vacuum drying is important because such water-logged rods are, very often, older and colder rods that do not generate sufficient internal heat to expel the water as vapor. If they remain water-logged during dry storage or transport, the opportunity for spent fuel or container degradation through corrosion or other mechanisms, in clear violation of regulations, increases. Finally, the vacuum drying process presents no threat to the spent fuel or to the spent fuel container, since the maximum of one atmosphere of differential pressure that these fuel rods and canisters experience is far less than the normal, off-normal, and accident condition loadings for which they are designed for service in a reactor. Other proposed drying processes that simply pump a heated gas through the container may very well fail to remove such moisture trapped in spent fuel rods because there is neither a sufficient heat source within the fuel nor a sufficient pressure differential across the fuel cladding to remove the moisture. Indeed, with a system that pumps a heated and pressurized gas through the container for drying, the increase in the pressure differential caused by the gas tends to suppress moisture vaporization in the fuel and container and to force the moisture to remain within the water-logged fuel. Further, there are other “moisture hide-out” conditions that may pertain to spent fuel and its containers that a pressurized gas drying system will have great difficulty in eliminating. These conditions include the use of damaged fuel cans within spent fuel containers and water that is trapped within spent fuel assembly guide tubes, among others. With all the advantages of vacuum drying, its major drawback occurs when very hot spent fuel must be placed into dry storage or transported. With very hot spent fuel under vacuum, the cooling atmosphere no longer surrounds the fuel and, without a conducting medium to remove fuel heat, fuel temperatures can rise rapidly over a period of hours. Such a temperature rise threatens the operators at the power plant with exceeding both the fuel high temperature and the fuel temperature variation regulatory limits. For this reason, it is vitally important to the industry to have a modified system for drying very hot spent fuel within a container, so that proper dryness can be achieved while maintaining fuel temperatures within regulatory requirements. The method described herein achieves this outcome without compromising the very useful application and effectiveness of vacuum drying, which has been demonstrated over several decades to assure spent fuel dryness for long term material integrity and stability. Embodiments of the disclosure facilitate proper spent fuel container dryness while controlling and limiting the fuel temperature rise that may result from using proven vacuum drying methods alone. A method according to the disclosure comprises the use of both a vacuum drying system and a non-reactive gas cooling system that works in combination with the vacuum drying system. The vacuum drying system is comprised of outlet paths and connections from the spent fuel container, vacuum drying pump or pumps, pump discharge paths and connections to radioactive liquid and gas disposal systems, other pump discharge paths and connections as well as associated piping, valving, and monitoring instrumentation. In one embodiment, a gas cooling system according to the disclosure may be comprised of inlet paths and connections to the spent fuel container, outlet paths and connections from the spent fuel container, which may be integrated with the vacuum drying system, non-reactive gas circulating pump or pumps, which may be integrated with the drying system vacuum pump or pumps, pump discharge paths and connections to the spent fuel container and to the heat exchanger, other pump discharge paths and connections, a heat exchanger with appropriate cooling medium inlet, outlet, and moisture drain features, such heat exchanger being located in the main non-reactive gas circulation flow path and/or in a parallel path, a non-reactive gas supply and supply connection, and associated piping, valving, and monitoring instrumentation. The non-reactive gases that may be used include, but are not limited to, all inert gases (helium, argon, etc.), carbon dioxide, nitrogen, or other gases having high thermal stability with good heat transfer properties (high moisture absorption characteristics may also be considered as a valuable parameter). Upon initiation of the container and spent fuel drying process, the heat generation rate of the spent fuel in the container is known, and fuel heat-up rates and increase in the resulting temperatures of the spent fuel have been calculated based upon the design conditions of the container. Typically, this disclosed drying method will be used with spent fuel containers having contained fuel heat rates in the range of 25 kW to 45 kW, although with lower or higher container heat rates, the method can still be applied effectively. With the spent fuel and container heat rates and temperatures known from calculation or measurement, an initial vacuum is drawn on the system, which begins the removal of moisture. Because of a fuel clad temperature variation range limitation of approximately 50° C. to 100° C., or because of a peak spent fuel clad temperature limit that can range from approximately 250° C. to 400° C., depending on fuel or regulatory restrictions, the initial drying period will be predetermined, based upon the temperature of the spent fuel in the container at vacuum drying initiation and the heat generation rate of the spent fuel. This period will typically be less than 10 hours, but vacuum drying continues until both the calculated fuel temperature and the range of fuel temperature increase after the initial period of vacuum drying reach pre-determined values based upon the initial conditions. When vacuum drying is stopped, gas cooling is initiated by back-filling the system with a non-reactive gas at approximately atmospheric or somewhat higher pressure and using the cooling system pump or pumps to circulate the gas through the spent fuel container. For purposes of meeting regulatory non-condensable/corrosive gas limitations in the canister, the non-reactive gas should meet chemical gas purity standards or have impurities typically totaling less than 0.5%, depending on the impurity(ies). A spent fuel container, with the internal spent fuel basket and the rod-and-grid array of the spent fuel assemblies, makes a desirable arrangement for a typical tube-and-shell heat exchanger to remove spent fuel heat, thereby providing excellent control of both fuel temperature peaks and ranges of variation. If the fuel container is generating heat above a rate that requires other than ambient cooling of the non-reactive gas in the cooling circuit (typically less than 30 kW), the non-reactive gas heat exchange system may also be used, and cooling water flow to the heat exchanger will be started. As with heat-up, the fuel temperature range must also be controlled during cool-down with the non-reactive gas, and this will be done by calculation using the parameters of the fuel, container, and drying system previously discussed. As the gas circulates through the spent fuel container, it removes heat from the fuel and, as the gas is heated by the spent fuel, it can also absorb moisture from the spent fuel container. The circulating gas primarily serves the purpose to cool the spent fuel in the container, and the gas is cooled by the heat exchanger, thereby allowing the fuel temperature to be controlled by the gas flow rate appropriate for the wide range of possible spent fuel loading conditions and the amount of sub-cooling established by the heat exchanger, based upon pre-determined levels established by calculation. An ancillary feature of the gas chilling function performed by the heat exchanger (as with even a simple air-conditioning system) is that the moisture from the spent fuel container retained by the heated, non-reactive gas will condense and be removed from the gas in the heat exchanger while the fuel temperature is being stabilized at a pre-determined level. Once the spent fuel and gas have achieved a condition of homeostasis at the desired temperature within the range of approximately 250° C. to approximately 400° C. and within the accepted range of temperature variation of approximately 50° C. to 100° C., as determined by gas flow rate and spent fuel container inlet gas temperature, gas cooling period, spent fuel heat generation rate, and spent fuel container outlet gas temperature, then gas cooling may be stopped and vacuum drying may be resumed. This drying and heat removal process cycle may be repeated as desired, and a vacuum dryness testing process (e.g., the pressure rise in the spent fuel container over a given period when isolated from vacuum drying as a result of additional moisture evaporating into the vapor state) may be performed at any time. With proper management of multiple cycles of drying and heat removal, sufficient dryness while assuring the regulatory limits of spent fuel temperature or temperature variation can be achieved because the non-reactive gas cooling system may always be relied upon for a specified period of time to actually control the spent fuel temperatures and range of variation so that additional drying and heat removal cycles can be performed until proper dryness is achieved. The method for use of the combined vacuum drying and heat removal systems can utilize a known heat generation rate of the spent fuel within the container, the approximate initial temperature of the spent fuel before the initiation of vacuum drying, the inlet temperature and the flow rate of non-reactive gas through the spent fuel container, and the spent fuel container non-reactive gas outlet temperature. These parameters may all be measured or conservatively calculated as part of the procedures to implement the method. All resulting system conditions and parameters can be conservatively calculated on the basis of known initial conditions and procedures that vary in accordance with these known initial conditions may be established based upon such calculations so that proper spent fuel container dryness with spent fuel temperatures remaining in compliance with regulatory directions is assured. Accordingly, FIG. 1 shows one embodiment of a system 100 according to the present disclosure. In the embodiment shown in FIG. 1, spent fuel (e.g., spent nuclear fuel) is disposed within a spent fuel container 103. In an embodiment, the spent fuel container further comprises a spent fuel canister 156 disposed within another cask used to handle the canister. The spent fuel within the spent fuel container 103 can initially be allowed to reach a pre-determined temperature in a non-reactive gas, the temperature being approximately that of normal operation for the heat generation rate of the spent fuel once it is in dry storage or in transport. Such a temperature would typically be in the range of approximately 150° C. to approximately 300° C. In many embodiments, the pre-determined temperature can be one that permits a container temperature that causes vaporization of residual moisture within the spent fuel container. The spent fuel container 103 is also coupled to at least one inlet path 105 as well as at least one outlet path 107, which communicatively couple the spent fuel container 103 to a circulation path 109 typically comprising a high quality metal that is corrosion, diffusion and leakage resistant (e.g., stainless steel), ceramic, or other high integrity material that is configured to circulate a non-reactive gas among the various components of the system. The circulation path 109 can be communicatively coupled to the various components of the depicted system 100 and permit the flow of gas between the various components as depicted in the non-limiting example of FIG. 1. The depicted system 100 can also include a spent fuel container bypass valve 110 that can substantially isolate the spent fuel container from the circulation path 109. Additionally, the circulation path 109 can include one or more shutoff valves 112 that can cause circulation of gas through the circulation path 109 to cease. At least one circulation pump 113, when the system is in a vacuum drying mode, can be used to cause a reduction of pressure within the spent fuel container 103 as well as the rest of the circulation path 109 relative to the atmosphere. In some embodiments, the circulation pump 113 can cause near-vacuum conditions within the spent fuel container 103 and/or the circulation path 109. The reduction of pressure within the spent fuel container in the vacuum drying mode 103 can cause moisture extraction from the spent fuel within the spent fuel container and from the container itself. Additionally, when the circulation pump is used for non-reactive gas cooling of the spent fuel, the non-reactive gas absorbs heat from the spent fuel. Accordingly, a heat exchanger 117 coupled to the circulation path 109 is configured to remove heat from the gas exiting the spent fuel container 103. Therefore, the heat exchanger 117 can be coupled to a cooling inlet 118 as well as a cooling outlet 119 for circulation of a coolant through the heat exchanger. The heat exchanger 117 can also be coupled with a liquid waste removal system that can remove potentially radioactive waste liquids from the gas exiting the spent fuel container 103 via a liquid waste outlet 120. The circulation path 109 can also be configured with a heat exchanger bypass valve system 122 that can be configured to isolate the heat exchanger from the circulation path 109 if necessary. The liquid waste outlet 120 can likewise be configured with a liquid waste shutoff valve 125 to isolate the liquid waste outlet 120 from the heat exchanger 117. During a subsequent cooling mode, a non-reactive gas source 131 coupled to the circulation path 109 to provide the non-reactive gas circulating throughout the system 100. The non-reactive gas source would typically be pressurized to a value in the range of 100 psig or greater in order to supply the volume of gas desired for system and container backfill and system flow rates. In one embodiment, such a source would be typically provided as bottled-gas that may be replaced when empty. Further, as noted above for the vacuum drying mode, the circulation pump 113 can cause a reduction of pressure within the circulation path 109 relative to the atmosphere. Accordingly, such a pressure differential can draw the non-reactive gas from the container and circulation system at the re-initiation of vacuum drying and discharge that gas to a radioactive waste gas system or through a vent to another storage (tanks or similar) system for use during the next gas cooling mode. A non-reactive gas shutoff valve 133 can also be provided, which can isolate the non-reactive gas source 131 from the circulation path 109. A four-way flow splitter 141 (e.g., a four-way splitter ball valve) can also be employed that is coupled to the circulation path 109 to facilitate enhancement of non-reactive gas flow rates or to supplement moisture extraction rates during vacuum drying. A shutoff valve system 155 can isolate the four-way splitter 141 from the circulation path 109 as well as halt circulation of gases through the circulation path 109. A radioactive waste gas outlet 145 can also be coupled to the circulation path 109 to remove any radioactive waste gases that may persist in the gas exiting the spent fuel container 103 from the circulation path during vacuum drying, prior to starting recirculation of the non-reactive gas to the at least one inlet path 105 coupled to the spent fuel container 103. In an embodiment, a radioactive waste gas disposal system 157 is coupled to the circulation path, the radioactive waste gas disposal system configured to remove radioactive waste gas from the circulation path. Such disposition of potentially radioactive gases from within the spent fuel container into a power plant's radioactive waste gas system is a typical operation for the management of plant radioactive waste. A radioactive waste gas outlet shutoff valve 147 can isolate the radioactive waste gas outlet from the circulation path 109. The circulation pump(s) may also be coupled to a venting system 149 for storing (recycling) of non-reactive gas from one cooling cycle to the next, and a venting shutoff valve 151 can isolate the venting system from the circulating pump(s). In some embodiments, an additional circulation pump 161 can be employed. In various embodiments, a first pump can be employed for vacuum drying and a second pump for gas recirculation flow throughout the system. A method using vacuum drying combined with non-reactive gas cooling of hot spent nuclear fuel in dry storage and transport containers can allow operators of such systems to meet regulatory fuel temperature requirements while achieving proper spent fuel container dryness using the most efficient and proven vacuum drying systems and methods. FIG. 2 illustrates a flowchart that depicts one example of a method according to an embodiment of the disclosure. The method illustrated in FIG. 2 can be implemented in a system according to the present disclosure. First, in box 201, spent nuclear fuel can be disposed or stored in a spent fuel container that is coupled to a circulation path. As noted above, a spent fuel container can comprise a canister with a closure lid that can, in turn, be disposed in a storage and/or transport cask. In box 203, a negative pressure can be imparted on the spent fuel container. In some embodiments, a negative pressure can be generated by a vacuum pump and/or recirculation pump that is coupled to the circulation path. In box 204, gas can be discharged from the spent fuel container to a radioactive waste gas system and/or vent. Next, in box 205, a non-reactive gas can be released and/or injected into the spent fuel container and/or the circulation path. In box 209, a gas exiting the spent fuel container can be cooled by a heat exchanger or other apparatus to remove heat from the gas. In box 210, the non-reactive gas can be circulated through the recirculation path. The circulation of the non-reactive gas can facilitate the controlling and/or modulation of temperature levels in the spent fuel container. In box 211, vacuum drying of the spent fuel container can be resumed. In box 212, condensate waste can be removed from the system. In some embodiments, condensate waste may be radioactive in nature and thus properly disposed of. Finally, it should be noted that commercial spent nuclear fuel has been generated and accumulated for more than 50 years at power plants in the U.S. That means there is an extensive variation in fuel design, heat generation rates, cooling periods, materials of fabrication, uranium enrichments, burn-ups, and other parameters that all play significant roles in establishing the design characteristics of the subject method. Where typical values are provided, it must be understood that the large variation in commercial spent nuclear fuel designs, materials, and radioactive decay periods means there will also be a number of method and system approaches that will be outside the typical ranges provided herein. Although the flowchart of FIG. 2 shows a specific order of execution, it is understood that the order of execution may differ from that which is depicted. For example, the order of execution of two or more blocks may be scrambled relative to the order shown. Also, two or more blocks shown in succession in FIG. 2 may be executed concurrently or with partial concurrence. Further, in some embodiments, one or more of the blocks shown in FIG. 2 may be skipped or omitted. It should be noted that ratios, concentrations, amounts, and other numerical data may be expressed herein in a range format. It is to be understood that such a range format is used for convenience and brevity, and thus, should be interpreted in a flexible manner to include not only the numerical values explicitly recited as the limits of the range, but also to include all the individual numerical values or sub-ranges encompassed within that range as if each numerical value and sub-range is explicitly recited. To illustrate, a concentration range of “about 0.1% to about 5%” should be interpreted to include not only the explicitly recited concentration of about 0.1 wt % to about 5 wt %, but also include individual concentrations (e.g., 1%, 2%, 3%, and 4%) and the sub-ranges (e.g., 0.5%, 1.1%, 2.2%, 3.3%, and 4.4%) within the indicated range. In an embodiment, the term “about” can include traditional rounding according to significant figures of the numerical value. In addition, the phrase “about ‘x’ to ‘y’” includes “about ‘x’ to about ‘y’”. Additionally, where components of embodiments of the disclosure are shown and/or discussed as being coupled, communicatively coupled and/or connected to one another, it should be appreciated that these components may not be in direct coupling to one another, and that intermediary components or elements can be employed between the coupled components. It should be emphasized that the above-described embodiments of the present disclosure are merely possible examples of implementations set forth for a clear understanding of the principles of the disclosure. Many variations and modifications may be made to the above-described embodiment(s) without departing substantially from the spirit and principles of the disclosure. All such modifications and variations are intended to be included herein within the scope of this disclosure and protected by the following claims.
	claims	1. A source storing apparatus, comprising: a source tank and a shielding plug, the source tank being provided with an opening and an accommodating cavity, the accommodating cavity being configured to accommodate a cobalt source box, the shielding plug being configured to seal an opening of the accommodating cavity; whereina first connecting structure is provided on the cobalt source box;a second connecting structure is provided on an outer side of the shielding plug anda pickup structure is provided on an inner side of the shielding plug, and the first connecting structure is detachably connected to the pickup structure,wherein the first connecting structure is a connecting slot, the pickup structure is snappable into the connecting slot, the connecting slot comprises a first opening and a second opening that are in communication, wherein the first opening is in a first direction, and the second opening is in a second direction; and the pickup structure is configured to snap fit into the connecting slot through the first opening and get out of the connecting slot through the second opening. 2. The source storing apparatus according to claim 1, wherein the pickup structure is directly connected to the shielding plug; or the pickup structure is connected to the shielding plug through a connecting rod. 3. The source storing apparatus according to claim 1, wherein the connecting slot comprises a clamping slot and a compressing slot, wherein a maximum size of the clamping slot is larger than the first opening. 4. The source storing apparatus according to claim 3, wherein when the pickup structure comprises the elastic member, the first opening is larger than a minimum compression size of the elastic member and smaller than a maximum extension size of the elastic member. 5. The source storing apparatus according to claim 3, wherein a maximum size of the clamping slot is larger than the maximum extension size of the elastic member. 6. The source storing apparatus according to claim 3, wherein a maximum size of the compressing slot is smaller than the maximum extension size of the elastic member. 7. The source storing apparatus according to claim 1, wherein a bottom size of the second opening is larger than a top size of the second opening, and a maximum opening size of the second opening is larger than or equal to a minimum compression size of the pickup structure. 8. The source storing apparatus according to claim 4, wherein a size of the second opening at a position corresponding to the compressing slot is larger than or equal to the minimum compression size of the pickup structure. 9. A source guiding system, being configured to guide a cobalt source box from a first source storing apparatus into a second source storing apparatus, the cobalt source box further comprising a third connecting structure; andthe source guiding system comprising: a source guiding tank, the source guiding tank comprising a tank body, a first pull rod and a first opening, a second pull rod and a second opening, wherein the tank body comprises a source-carrying cavity, the first pull rod and the first opening are located on opposite sides of the source-carrying cavity along a first direction, the second pull rod and the second opening are located on opposite sides of the source-carrying cavity along a second direction, the first pull rod moves along the first direction and is connectable to a second connecting structure of a shielding plug of the first source storing apparatus, a pickup structure of the shielding plug is detachably connected to a first connecting structure of the cobalt source box; and the second pull rod moves along the second direction and is connectable to the third connecting structure of the cobalt source box. 10. The source guiding system according to claim 9, further comprising a first shielding door configured to open and close the first opening; and/orthe source guiding system further comprising a second shielding door configured to open and close the second opening. 11. The source guiding system according to claim 9, wherein the first direction is perpendicular to the second direction. 12. The source guiding system according to claim 9, wherein the first pull rod and/or the second pull rod are/is provided with a position restricting slot, and the source guiding system further comprises a fastener fixable with the position restricting slot to prevent the first pull rod and the second pull rod from rotating. 13. The source guiding system according to claim 9, wherein:a glass window is provided on the source guiding tank; orat least one of a camera or a detection lamp is further provided within the source-carrying cavity of the source guiding tank. 14. The source guiding system according to claim 9, wherein the source guiding tank comprises a plurality of components, and the plurality of components is fixedly connected. 15. A source guiding method, being used for guiding a cobalt source box from a source storing apparatus into a radiotherapy device, the method comprising:connecting a first connecting structure of the cobalt source box and a pickup structure of a shielding plug of the source storing apparatus;driving a first pull rod to connect the first pull rod to a second connecting structure of the shielding plug of the source storing apparatus, and lifting the shielding plug to a source-carrying cavity of a source guiding tank;driving a second pull rod to connect the second pull rod to a third connecting structure of the cobalt source box;driving the first pull rod to coordinate with the second pull rod, such that the first connecting structure of the cobalt source box separates from the pickup structure; anddriving the second pull rod to enable the cobalt source box to enter the radiotherapy device and fix the cobalt source box with the radiotherapy device. 16. The method according to claim 15, wherein the driving the first pull rod to coordinate with the second pull rod, such that the first connecting structure of the cobalt source box separates from the pickup structure specifically comprises:driving the first pull rod to move along a first direction such that an elastic member is in a compressing slot; anddriving the second pull rod to move along a second direction, such that the pickup structure is disengaged from a connecting slot through a second opening. 17. The method according to claim 16, further comprising: driving the radiotherapy device to switch off and shield a radioactive source. 18. A source guiding method, being applied to the source guiding system according to claim 9, wherein the first source storing apparatus is a radiotherapy device, the method comprising:driving the first pull rod to connect the first pull rod to the second connecting structure of the shielding plug of the second source storing apparatus, and lifting the shielding plug to the source-carrying cavity of the source guiding tank;driving the second pull rod to connect the second pull rod to the third connecting structure of the cobalt source box, and pulling the cobalt source box to the source-carrying cavity of the source guiding tank;driving the first pull rod to coordinate with the second pull rod, such that the first connecting structure of the cobalt source box is connected to the pickup structure;driving the second pull rod to separate the second pull rod from the cobalt source box; anddriving the first pull rod to send the cobalt source box into the second source storing apparatus.
	abstract	Systems and methods are provided for scanning an item utilizing an X-ray scanner in order to facilitate a determination of whether the X-ray radiation penetrated through the entirety of the scanned item. Various embodiments comprise a conveying mechanism, an X-ray emitter, a detector, and an X-ray penetration grid (XPG). The XPG may comprise a radiopaque grid that may serve as a reference for determining whether radiation passes through the scanned item, the grid oriented such that the grid members are neither parallel nor perpendicular to the direction of travel. Such orientation may minimize or eliminate “ghosted” radiation signals included in a visual display of the radiation received by the detector. A scanned item may be oriented with the XPG such that radiation emitted by the X-ray emitter that passes through a portion of the scanned item must also pass through the XPG before being received by the detector.
	claims	1. A nuclear fuel rod for a boiling water nuclear reactor, comprising:a cladding tube, defining a closed inner space and which is manufactured from at least one of the materials in the group zirconium and a zirconium-based alloy, the material of the cladding tube comprising a plurality of sites in which hydrogen is capable of being adsorbed;a plurality of nuclear fuel pellets, arranged in the inner space in the cladding tube so that the nuclear fuel pellets fill pan of the inner space;an initial fill gas arranged in the closed inner space in order to fill the rest of the inner space;wherein the initial fill gas contains a proportion of inert gas and a proportion of carbon monoxide, the carbon monoxide being located in the sites in which hydrogen is capable of being adsorbed, thereby blocking the sites; and whereinthe internal pressure (Pfill) of the initial gas in the nuclear fuel rod amounts to at least about 7 bar (abs) at room temperature (TR) and the proportion of carbon monoxide is at least 4.7 volume percent of the initial fill gas; andwherein the cladding tube has an inner surface that faces the inner space and the material in the cladding tube nearest the inner surface is pre-oxidized to provide a surface layer that comprises zirconium oxide. 2. A nuclear fuel rod according to claim 1, wherein the proportion of carbon monoxide constitutes at least 5 volume percent of the initial fill gas. 3. A nuclear fuel rod according to claim 2, wherein the proportion of carbon monoxide constitutes at least 6 volume percent of the initial fill gas. 4. A nuclear fuel rod according to claim 1, wherein the inert gas consists substantially of helium. 5. A nuclear fuel assembly for a boiling water nuclear reactor, said nuclear fuel assembly comprising a plurality of nuclear fuel rods, each fuel rod including:a cladding tube, defining a closed inner space and which is manufactured from at least one of the materials in the group zirconium and a zirconium-based alloy, the material of the cladding tube comprising a plurality of sites in which hydrogen is capable of being adsorbed;a plurality of nuclear fuel pellets, arranged in the inner space in the cladding tube so that the nuclear fuel pellets fill pan of the inner space;an initial fill gas arranged in the closed inner space in order to fill the rest of the inner space;wherein the initial fill gas contains a proportion of inert gas and a proportion of carbon monoxide, the carbon monoxide being located in the sites in which hydrogen is capable of being adsorbed, thereby blocking the sites; and whereinthe internal pressure (Pfill) of the fill gas in the nuclear fuel rod amounts to at least about 7 bar (abs) at room temperature (TR) and the proportion of carbon monoxide is at least 4.7 volume percent of the initial fill gas; andwherein the cladding tube has an inner surface that faces the inner space and the material in the cladding tube nearest the inner surface is pre-oxidized to provide a surface layer that comprises zirconium oxide. 6. A nuclear fuel rod for a boiling water nuclear reactor, comprising:a cladding tube, defining a closed inner space and which is manufactured from at least one of the materials in the group zirconium and a zirconium-based alloy, the material of the cladding tube comprising a plurality of sites in which hydrogen is capable of being adsorbed;a plurality of nuclear fuel pellets, arranged in the inner space in the cladding tube so that the nuclear fuel pellets fill pan of the inner space;an initial fill gas arranged in the closed inner space in order to fill the rest of the inner space;wherein the initial fill gas contains a proportion of inert gas and a proportion of carbon monoxide, the carbon monoxide being located in the sites in which hydrogen is capable of being adsorbed, thereby blocking the sites; and whereinthe internal pressure (Pfill) of the initial gas in the nuclear fuel rod amounts to at least about 6 bar (abs) at room temperature (TR) and the proportion of carbon monoxide is at least 5.7 volume percent of the initial fill gas; andwherein the cladding tube has an inner surface that faces the inner space and the material in the cladding tube nearest the inner surface is pre-oxidized to provide a surface layer that comprises zirconium oxide. 7. A nuclear fuel assembly for a boiling water nuclear reactor, said nuclear fuel assembly comprising a plurality of nuclear fuel rods, each fuel rod including:a cladding tube, defining a closed inner space and which is manufactured from at least one of the materials in the group zirconium and a zirconium-based alloy, the material of the cladding tube comprising a plurality of sites in which hydrogen is capable of being adsorbed;a plurality of nuclear fuel pellets, arranged in the inner space in the cladding tube so that the nuclear fuel pellets fill pan of the inner space;an initial fill gas arranged in the closed inner space in order to fill the rest of the inner space;wherein the initial fill gas contains a proportion of inert gas and a proportion of carbon monoxide, the carbon monoxide being located in the sites in which hydrogen is capable of being adsorbed, thereby blocking the sites; and whereinthe internal pressure (Pfill) of the initial gas in the nuclear fuel rod amounts to at least about 6 bar (abs) at room temperature (TR) and the proportion of carbon monoxide is at least 5.7 volume percent of the initial fill gas; andwherein the cladding tube has an inner surface that faces the inner space and the material in the cladding tube nearest the inner surface is pre-oxidized to provide a surface layer that comprises zirconium oxide.
	claims	1. A semiconductor sensor, comprising:a semiconductor layer having a top surface;a plurality of pixel surface coatings positioned above the top surface of the semiconductor layer to form a corresponding plurality of pixels, wherein each pixel surface coating is separated from each adjacent pixel surface coating, wherein the plurality of pixel surface coatings are conductive, wherein electrons incident one of the plurality of pixel surface coatings are absorbed by the one of the plurality of pixel surface coatings resulting in a charge associated with the pixel corresponding to the one of the plurality of pixel surface coatings such that the charge produces a readable voltage associated with the pixel corresponding to the one of the plurality of pixel surface coatings; anda conductive layer positioned above the top surface of the semiconductor layer and positioned so that electrons passing between adjacent pixel surface coatings are incident on the conductive layer and prevented from penetrating into the semiconductor layer, wherein the conductive layer is insulated from the plurality of pixel surface coatings. 2. The semiconductor sensor according to claims 1, wherein each of the plurality of pixel surface coatings comprises light impervious material wherein each of the plurality of pixel surface coatings are light impervious. 3. The semiconductor sensor according to claim 2, wherein each of the plurality of pixel surface coatings comprises aluminum. 4. The semiconductor sensor according to claims 1, wherein the conductive layer acts as a capacitor electrode. 5. The semiconductor sensor according to claims 1, wherein a potential is applied to the conductive layer. 6. The semiconductor sensor according to claim 1, further comprising a plurality of detection surfaces positioned above the plurality of pixel surface coatings, each detection surface corresponding to one of the plurality of pixel surface coatings wherein each detection surface comprises an electron-intensifying coating, wherein electron multiplication occurs when an electron is incident on the electron-intensifying coating. 7. The semiconductor sensor according to claim 6, further comprising a plurality of second conductive layers wherein each of the plurality of second conductive layers is positioned above a corresponding pixel surface coating and below a corresponding electron-intensifying coating wherein a first additional electric potential is applied to each of the plurality of second conductive layers wherein the conductive layer comprises at least one extension positioned above the plurality of detection surfaces, wherein a second additional electric potential is applied to each of the at least one extension. 8. The semiconductor according to claim 1, wherein neighboring pixel surfaces coatings have different electric potentials. 9. The semiconductor sensor according to claim 1, wherein the conductive layer comprises a light impervious material, wherein the conductive layer is light impervious. 10. The semiconductor sensor according to claim 9, wherein the second conductive layer comprises aluminum. 11. The semiconductor sensor according to claim 6, wherein each detection surface comprises a transmit channel which allows electrons to pass through the transmit channel to the corresponding pixel surface coating.
	claims	1. A shutter assembly for mitigating debris in a laser-produced plasma device, the shutter assembly comprising:a rotatable shutter having at least one aperture that provides a line-of sight between a radiation source and an exit of the device during a first period of rotation of the shutter, and wherein the shutter obstructs the line-of-sight between the radiation source and the exit during a second period of rotation; anda motor configured to rotate the shutter to permit passage of radiation through the at least one aperture during the first period of rotation, and to obstruct the passage of debris generated at the radiation source through the at least one aperture during the second period of rotation, wherein the shutter further comprises a plurality of vanes arranged in parallel defining a plurality of parallel apertures. 2. A shutter assembly according to claim 1, wherein the shutter further comprises a plurality of apertures radially extending from a center of the shutter. 3. A shutter assembly according to claim 1, further comprising a synchronization device configured to synchronize the first period of rotation with a radiation generation event at the radiation source. 4. A shutter assembly according to claim 3, wherein the synchronization device comprises an encoded motor configured to generate a synchronization signal. 5. A shutter assembly according to claim 3, wherein the synchronization device comprises a blade coupled to the shutter, wherein the blade is configured to synchronize the first period of rotation with the radiation generating event by blocking a laser beam during the first or second period of rotation. 6. A shutter assembly according to claim 1, wherein the motor is further configured to rotate the shutter at about 300 RPM when the radiation source generates X-rays and the debris generated at the radiation source has a velocity of about 105 cm/sec or slower. 7. A shutter assembly according to claim 1, wherein the motor is further configured to rotate the shutter at about 3000 RPM when the radiation source generates X-rays and the debris generated at the radiation source has a velocity of about 106 cm/sec or slower. 8. A method for mitigating debris in a laser-produced plasma device, the method comprising:providing a rotatable shutter having at least one aperture;rotating the shutter to provide a line-of-sight between a radiation source and an exit of the device through the at least one aperture during a first period of rotation;rotating the shutter to obstruct the line-of-sight between the radiation source and the exit during the second period of rotation; andwhereby rotating the shutter during the first period of rotation permits the passage of radiation generated at the radiation source through the at least one aperture to the exit, and obstructs the passage of debris generated at the radiation source through the at least one aperture during the second period of rotation,wherein the shutter further comprises a plurality of vanes arranged in parallel defining a plurality of parallel apertures. 9. A method according to claim 8, wherein the shutter further comprises a plurality of apertures radially extending from a center of the shutter. 10. A method according to claim 8, further comprising synchronizing the first period of rotation with a radiation generating event at the radiation source. 11. A method according to claim 10, wherein synchronizing the first period of rotation further comprises generating a synchronization signal with an encoded motor configured to rotate the shutter. 12. A method according to claim 10, wherein synchronizing the first period of rotation further comprises obstructing a laser beam during the first or second portion of the rotation with a blade coupled to the shutter, the obstruction configured to synchronize the first period of rotation with the radiation generating event. 13. A method according to claim 8, wherein the shutter is rotated at about 300 RPM wherein the radiation source generates X-rays and the debris generated at the radiation source has a velocity of about 105 cm/sec or slower. 14. A method according to claim 8, wherein the shutter is rotated at about 3000 RPM and wherein the radiation source generates X-rays and the debris generated at the radiation source has a velocity of about 106 cm/sec or slower. 15. A laser-produced plasma device, comprising:a laser source for laser irradiation of a target;a radiation source comprising a target that forms a plasma that generates short-wavelength radiation and generates debris when irradiated by the laser; anda shutter assembly for obstructing the passage of debris through an exit, the shutter assembly comprising:a rotatable shutter having at least on aperture that provides a line-of-sight between the radiation source and the exit of the device during a first period of rotation of the shutter, and obstructs a line-of-sight between the radiation source and the exit during a second period of rotation, anda motor configured to rotate the shutter to permit passage of the radiation through the at least one aperture during the first period of rotation, and to obstruct the passage of the debris through the at least one aperture during the second period of rotation, wherein the shutter further comprises a plurality of vanes arranged in parallel defining a plurality of parallel apertures. 16. A laser-produced plasma device according to claim 15, wherein the shutter further comprises a plurality of apertures radially extending from a center of the shutter. 17. A shutter assembly according to claim 15, further comprising a synchronization device configured to synchronize the first period of rotation with the radiation generating event.
047605898	claims	1. An apparatus for holding an X-ray film cassette in position comprising in combination: a. a cabinet including a compartment for slidably receiving a planar tray and an opening covered by a planar grid having a plurality of X-ray opaque grid lines disposed in a parallel relationship, wherein said planar grid is movable over a short distance in a first direction generally perpendicular to the direction of said grid lines: b. means for reciprocating said planar grid in directions parallel to said first direction; c. a planar tray included in said compartment of said cabinet, said planar tray including means for receiving the X-ray film cassette, said planar tray further including: a. a cabinet including a compartment for receiving a planar tray holding the X-ray film cassette, said cabinet including an opening for receiving X-ray therethrough; b. a planar grid with a plurality of X-ray opaque grid lines disposed in a parallel relationship covering said opening, said planar grid disposed in a movable arrangement in directions parallel to a first direction generally perpendicular to the direction of said grid lines on a plane including said planar grid; c. a spring bias means disposed adjacent to one extremity of said planar grid for exerting pressure on said one extremity, said one extremity being one of two extremities in said first direction; and d. a drive means disposed adjacent to the other extremity of said planar grid opposite to said one extremity for imposing reciprocating motions on said planar grid in directions generally parallel to said first direction, said other extremity being the other of two extremities in said first direction; wherein said drive means comprises variable speed motor, a noncircular cam nonrotatably mounted on a shaft of said variable speed motor; and a pivoting arm driven by said cam with a first extremity of said pivoting arm pivotably secured to the cabinet and a record of said extremity pivoting arm under a pressurized contact with said other extremity of said planar grid; wherein said drive means includes a bushing rotatably secured to said pivoting arm and under a pressurized contact with said cam, wherein said bushing has an offset outside diameter and a ratio of circumference of said bushing to circumference of said cam is an odd number. a. a cabinet including a compartment with an opening; b. a planar grid with a plurality of X-ray opaque grid lines disposed in parallel relationship in a first direction covering said an opening for receiving X-ray therethrough said planar grid resting on a pair of slide bearings respectively disposed along two opposite extremities of said planar grid parallel to a second direction which is perpendicular to said first direction wherein each slide bearing of said pair comprises a hollow bar with a slitted opening disposed parallel to said second direction and affixed to said cabinet and a slide bar supporting said grid, said slide bar having a cross section including two opposite right angle bends and slidably engaging said hollow bar, wherein combination of said hollow bar and said slide bar includes two parallel longitudinal cavities respectively confining two sets of plurality of bearing balls providing a frictionless sliding movement between said hollow bar and said slide bar; c. means for reciprocating said planar grid in directions parallel to said second direction. 1. a first pair of positioning means for positioning and holding the X-ray cassette therebetween, said first pair of positioning means connected to said planar tray in a sliding relationship in said first direction; 2. The combination as set forth in claim 1 wherein said combination includes a first actuator linkage with one extremity connected to the combination of said first pair of positioning means and having an actuating means included in the other extremity thereof, wherein said actuating means actuates a first position transducer means affixed to said cabinet, said first position transducer means indicating positions of said first pair of positioning means; and a second actuator linkage with one extremity connected to the combination of said second pair of positioning means and having an actuating means included in the other extremity thereof, wherein said actuating means actuates a second position transducer means affixed to said cabinet, said second position transducer means indicating positions of said second pair of positioning means. 3. The combination as set forth in claim 2 wherein said combination includes an X-ray collimator shutter control means controlled by signals from said first and second position transducer means, wherein said control means automatically positions the X-ray collimator shutters in a relationship consistent with the X-ray film cassette held by said planar tray. 4. The combination as set forth in claim 2 wherein said combination includes a handle means for pulling open said first pair of positioning means and means for disconnecting said limiting connector. 5. The combination as set forth in claim 1 wherein said combination includes mechanical linkage means coupling said first and second pair of positioning means to an X-ray collimator shutter control means, wherein said mechanical linkage means automatically positions the X-ray collimator shutters in a relationship consistent with the X-ray film cassette held by said planar tray. 6. The combination as set forth in claim 1 wherein each of said pair of positioning means connected to said planar tray in a sliding relationship includes an elongated sliding member supported by a pair of slide bearings respectively disposed along two edges thereof, wherein each of said pair of slide bearings comprises a first V-groove disposed following one of said edges of said elongated sliding member and a second V-groove included in said planar tray wherein said first and second V-grooves disposed in a closely spaced parallel relationship confines a plurality of bearing balls providing a frictionless sliding movement between said elongated sliding member and said planar tray. 7. An apparatus for holding X-ray film cassette during radiographic exposure comprising in combination: 8. The combination as set forth in claim 7 wherein said drive means comprising a variable speed motor, a noncircular cam nonrotatably mounted on a shaft of said variable speed motor; and a pivoting arm driven by said cam with one extremity pivotably secured to the cabinet and the other extremity under a pressurized contact with said other extremity of said planar grid. 9. The combination as set forth in claim 8 wherein said combination includes a bushing rotatably secured to said pivoting arm and under a pressurized contact with said cam, wherein said bushing has an offset outside diameter and a ratio of circumference of said bushing to circumference of said cam is an odd number. 10. The combination as set forth in claim 19 wherein said noncircular cam in combination with said bushing having the offset outside diameter provides a means for controlling amplitudes of reciprocating motion of said planar grid. 11. An apparatus for holding X-ray film cassette in position during X-ray examination, said apparatus comprising in combination: 12. The combination as set forth in claim 11 wherein said combination includes a planar tray included in said compartment of said cabinet in a sliding relationship in said second direction, wherein said planar tray is slidably supported by a pair of slide bearings respectively included in two opposite extremities of said planar tray parallel to said second direction, wherein each slide bearing of said pair comprises a hollow bar with a slitted opening disposed parallel to said second direction and affixed to said cabinet and a slide bar with a cross section including two opposite right angle bends affixed to said planar tray and slidably engaging said hollow bar wherein combination of said hollow bar and said slide bar includes two parallel longitudinal cavities respectively confining two sets of plurality of bearing balls providing a frictionless sliding movement between said hollow bar and said slide bar.
	abstract	Using a beam current of an ion beam, and a dose amount to a substrate, and an initial value of a scan number of the substrate set to 1, a scan speed of the substrate is calculated. If the scan speed is within the range, the current scan number and the current scan speed are set as a practical scan number and a practical scan speed, respectively. If the scan speed is higher than the upper limit of the range, the calculation process is aborted. If the scan speed is lower than the lower limit of the range, the scan number is incremented by one to calculate a corrected scan number. A corrected scan speed is calculated by using the corrected scan number, etc. The above steps are repeated until the corrected scan speed is within the allowable scan speed range.
	description	This application claims priority, under Section 371 and/or as a continuation under Section 120, to PCT Application No. PCT/IB2011/002190, filed on Jun. 2, 2011. The present invention generally relates to systems, methods and containers for storing hazardous waste material and, more particularly, filling devices, systems and methods for transferring hazardous waste material into a sealable container. Despite a proliferation of systems for handling and storing hazardous waste materials, prior art systems are still unable to effectively confine and control the unnecessary spread of hazardous waste contamination to areas remotely located from the hazardous waste material filling stations. Therefore, an urgent need exists for hazardous waste processing/storing systems that effectively minimize and/or eliminate unnecessary hazardous material contamination. There is disclosed herein a system for transferring hazardous waste material into a sealable container, the system comprising: a filling nozzle having: a valve body having a distal end and an outer surface, the valve body including a valve seat proximate the distal end, the outer surface proximate the distal end being configured to sealingly and removeably couple the valve body to an inner surface of a filling port of the container, a valve head having a valve face configured to form a seal with the valve seat in a closed configuration, the valve head configured to allow the valve body and the container to be fluidly coupled with one another in an open configuration, and a valve stem extending axially from the valve head through at least a portion of the valve body. Preferably, the system further comprises: a container configured to sealingly contain the hazardous waste material, the container including the filling port. Preferably, the system further comprises: a hopper; a first scale coupled to the hopper and configured to determine an initial hopper weight; a second scale coupled to the container and configured to determine a container fill weight; and a processor coupled to the first scale and the second scale and configured to compare the initial hopper weight to the container fill weight. Preferably, the hopper includes a volume substantially equal to a volume of the container. Preferably the system further comprises: at least one vibrator coupled to the hopper. Preferably, the system further comprises: at least one vibrator coupled to a bottom of the container. Preferably, the system further comprises: at least one vibrator coupled to a sidewall of the container. Preferably, the system further comprises: a lift mechanism configured to lift the container toward the fill nozzle Preferably, the lift mechanism including at least one damper. Preferably, the system further comprises: a sensor disposed in the valve head. Preferably, the sensor is configured to determine a level of hazardous material in the container. Preferably, the sensor extends distally from the valve body. Preferably, the sensor is coupled to a wire that extends through the valve stem. Preferably, the valve body includes: a first branch section configured to couple to a hopper, and a second branch section including the distal end and having a proximal end, the proximal end coupled to a drive mechanism configured to move the valve stem. Preferably, the drive mechanism includes a pneumatic cylinder. Preferably, the valve stem extends through the proximal end of the second branch section, the proximal end including a seal coupled to a portion of the valve stem. Preferably, the system further comprises: a vacuum nozzle configured to be in fluid communication with the container. Preferably, the vacuum nozzle extends through the distal end of the valve body. Preferably, the vacuum nozzle includes a filter proximate the distal end of the valve body. Preferably, the container includes an exhaust port. Preferably, the exhaust port includes a filter. Preferably, the system further comprises a vacuum nozzle sealingly and removeably couplable with the exhaust port, the vacuum nozzle being in sealed fluid communication with the valve body in a filling configuration. Preferably, the outer surface proximate the distal end includes at least one seal. Preferably, the at least one seal includes at least one o-ring. Preferably, the valve head extends distally from the valve body and into the container in the open configuration. Preferably, the container is at least initially under negative pressure. Preferably the filing port of the container is configured to be sealed closed after decoupling the valve body from the filling port. A method of transferring hazardous waste material into a sealable container, the method comprising: coupling an outer surface of a filling nozzle with an inner surface of a filling port of a container to form a first seal; opening a valve of a filling nozzle to add hazardous waste material into the container, the valve being proximate the first seal; closing the valve of the filling nozzle; decoupling the filling port from the filling nozzle and inserting a fill plug into the filling port, the fill plug forming a second seal with the inner surface of the filling port, the second seal being distally spaced from at least a portion of the first seal with respect to the container. Preferably the valve includes: a valve body having a distal end and an outer surface, the valve body including a valve seat proximate the distal end, the outer surface proximate the distal end being configured to sealingly and removeably couple the valve body to the filling port of the container, a valve head having a valve face configured to form a seal with the valve seat in a closed configuration, the valve head configured to allow the valve body and the container to be fluidly coupled with one another in an open configuration, and a valve stem extending axially from the valve head through at least a portion of the valve body. Preferably, the container includes an evacuation port. Preferably, the evacuation port includes an evacuation plug threadably coupled to the evacuation port, the method further comprising: allowing air and/or gas to pass through the filter and between the evacuation plug and the evacuation port in a filling configuration and a heating configuration; and closing the evacuation port with the evacuation plug in a closed configuration. Preferably, the evacuation port includes a filter. Preferably, the method further comprises: drawing air within the container displaced by the hazardous material through an evacuation nozzle coupled to the container, the evacuation nozzle being in sealed fluid communication with the valve body via the container. Preferably, the method further comprises: lifting the container toward the filling nozzle via a lifting mechanism to couple the filling port and the filling nozzle. Preferably, the method further comprises: weighing a hopper containing the hazardous material to determine an initial hopper weight; weighing the container while adding the hazardous material to determine a container fill weight; and comparing, via a processor, the difference between the initial hopper weight to the container fill weight. Preferably, the method further comprises: closing the valve once the container fill weight equals the initial hopper weight. Preferably, the method further comprises: vibrating the hopper via at least one vibrator while adding the hazardous material to the container. Preferably, the method further comprises: vibrating the container via at least one vibrator coupled to the container while adding the hazardous material to the container. Preferably, the method further comprises: measuring the level of hazardous material in the container via a sensor disposed in the valve head. Preferably, first seal includes at least one o-ring. Preferably, the second seal includes a gasket, the gasket being comprised of one or more of metal, ceramic or graphite. Preferably, the method further comprises: applying a vacuum to the container before or during adding of the hazardous material. Preferably, the method further comprises: permanently sealing the fill plug to the filling port; and heating and reducing the volume of the container after permanently sealing the fill plug to the filling port. Preferably, the hazardous waste material includes calcined material. In some embodiments, there are systems, methods and devices for storing and/or disposing of hazardous waste material. In some embodiments, the hazardous waste material includes nuclear waste such as calcined material. In one embodiment, there is system for transferring hazardous waste material into a sealable container, the system includes a filling nozzle having (a) a valve body having a distal end and an outer surface, the valve body including a valve seat proximate the distal end, the outer surface proximate the distal end being configured to sealingly and removeably couple the valve body to an inner surface of a filling port of the container, (b) a valve head having a valve face configured to form a seal with the valve seat in a closed configuration, the valve head configured to allow the valve body and the container to be fluidly coupled with one another in an open configuration, and (c) a valve stem extending axially from the valve head through at least a portion of the valve body. In a further embodiment, the system includes a container configured to sealingly contain the hazardous waste material, the container including the filling port. In a further embodiment, the system includes a hopper, a first scale coupled to the hopper and configured to determine an initial hopper weight, a second scale coupled to the container and configured to determine a container fill weight, and a processor coupled to the first scale and the second scale and configured to compare the initial hopper weight to the container fill weight. In one embodiment, the hopper includes a volume substantially equal to a volume of the container. In a further embodiment, the system includes at least one vibrator coupled to the hopper. In a further embodiment, the system includes at least one vibrator coupled to a bottom of the container. In a further embodiment, the system includes at least one vibrator coupled to a sidewall of the container. In a further embodiment, the system includes a lift mechanism configured to lift the container toward the fill nozzle. In one embodiment, the lift mechanism including at least one damper. In a further embodiment, the system includes a sensor disposed in the valve head. In one embodiment, the sensor is configured to determine a level of hazardous material in the container. In one embodiment, the sensor extends distally from the valve body. In one embodiment, the sensor is coupled to a wire that extends through the valve stem. In one embodiment, the valve body includes a first branch section configured to couple to a hopper, and a second branch section including the distal end and having a proximal end, the proximal end coupled to a drive mechanism configured to move the valve stem. In one embodiment, the drive mechanism includes a pneumatic cylinder. In one embodiment, the valve stem extends through the proximal end of the second branch section, the proximal end including a seal coupled to a portion of the valve stem. In a further embodiment, the system includes a vacuum nozzle configured to be in fluid communication with the container. In one embodiment, the vacuum nozzle extends through the distal end of the valve body. In one embodiment, the vacuum nozzle includes a filter proximate the distal end of the valve body. In one embodiment, the container includes an exhaust port. In one embodiment, the exhaust port includes a filter. In a further embodiment, the system includes a vacuum nozzle sealingly and removeably couplable with the exhaust port, the vacuum nozzle being in sealed fluid communication with the valve body in a filling configuration. In one embodiment, the outer surface proximate the distal end includes at least one seal. In one embodiment, the at least one seal includes at least one o-ring. In one embodiment, the valve head extends distally from the valve body and into the container in the open configuration. In one embodiment, the container is at least initially under negative pressure. In one embodiment, the filling port of the container is configured to be sealed closed after decoupling the valve body from the filling port. In another embodiment, there is a method of transferring hazardous waste material into a sealable container, the method comprising (a) coupling an outer surface of a filling nozzle with an inner surface of a filling port of a container to form a first seal, (b) opening a valve of a filling nozzle to add hazardous waste material into the container, the valve being proximate the first seal, (c) closing the valve of the filling nozzle, (d) decoupling the filling port from the filling nozzle and (e) inserting a fill plug into the filling port, the fill plug forming a second seal with the inner surface of the filling port, the second seal being distally spaced from at least a portion of the first seal with respect to the container. In one embodiment, the valve includes, a valve body having a distal end and an outer surface, the valve body including a valve seat proximate the distal end, the outer surface proximate the distal end being configured to sealingly and removeably couple the valve body to the filling port of the container, a valve head having a valve face configured to form a seal with the valve seat in a closed configuration, the valve head configured to allow the valve body and the container to be fluidly coupled with one another in an open configuration, and a valve stem extending axially from the valve head through at least a portion of the valve body. In one embodiment, the container includes an evacuation port. In one embodiment, the evacuation port includes an evacuation plug threadably coupled to the evacuation port and the method further comprises (f) allowing air and/or gas to pass through the filter and between the evacuation plug and the evacuation port in a filling configuration and a heating configuration, and (g) closing the evacuation port with the evacuation plug in a closed configuration. In one embodiment, the evacuation port includes a filter. In a further embodiment, the method includes drawing air within the container displaced by the hazardous material through an evacuation nozzle coupled to the container, the evacuation nozzle being in sealed fluid communication with the valve body via the container. In a further embodiment, the method includes lifting the container toward the filling nozzle via a lifting mechanism to couple the filling port and the filling nozzle. In a further embodiment, the method includes (t) weighing a hopper containing the hazardous material to determine an initial hopper weight, (g) weighing the container while adding the hazardous material to determine a container fill weight, and (h) comparing, via a processor, the difference between the initial hopper weight to the container fill weight. In a further embodiment, the method includes closing the valve once the container fill weight equals the initial hopper weight. In a further embodiment, the method includes vibrating the hopper via at least one vibrator while adding the hazardous material to the container. In a further embodiment, the method includes, vibrating the container via at least one vibrator coupled to the container while adding the hazardous material to the container. In a further embodiment, the method includes measuring the level of hazardous material in the container via a sensor disposed in the valve head. In one embodiment, wherein the first seal includes at least one o-ring. In one embodiment, the second seal includes a gasket, the gasket being comprised of one or more of metal, ceramic or graphite. In a further embodiment, the method includes applying a vacuum to the container before or during adding of the hazardous material. In a further embodiment, the method includes (f) permanently sealing the fill plug to the filling port, and (g) heating and reducing the volume of the container after permanently sealing the fill plug to the filling port. Reference will now be made in detail to the various embodiments of the present disclosure, examples of which are illustrated in the accompanying drawings FIGS. 2-17. Wherever possible, the same reference numbers will be used throughout the drawings to refer to the same or like parts. Nuclear waste, such as radioactive calcined material, can be immobilized in a container that allows the waste to be safely transported in a process known as hot isostatic pressing (HIP). In general, this process involves combining the waste material in particulate or powdered form with certain minerals and subjecting the mixture to high temperature and high pressure to cause compaction of the material. In some instances, the HIP process produces a glass-ceramic waste form that contains several natural minerals that together incorporate into their crystal structures nearly all of the elements present in HLW calcined material. The main minerals in the glass-ceramic can include, for example, hollandite (BaAl2Ti6O16), zirconolite (CaZrTi2O7), and perovskite (CaTiO3). Zirconolite and perovskite are the major hosts for long-lived actinides, such as plutonium, though perovskite principally immobilizes strontium and barium. Hollandite principally immobilizes cesium, along with potassiume, rubidium, and barium. Treating radioactive calcined material with the HIP process involves, for example, filling a container with the calcined material and minerals. The filled container is evacuated and sealed, then placed into a HIP furnace, such as an insulated resistance-heated furnace, which is surrounded by a pressure vessel. The vessel is then closed, heated, and pressurized. The pressure is applied isostatically, for example, via argon gas, which, at pressure, also is an efficient conductor of heat. The combined effect of heat and pressure consolidates and immobilizes the waste into a dense monolithic glass-ceramic sealed within the container. FIGS. 1A and 1B respectively show an example container, generally designated 100, before and after HIP processing. Container 100 has a body 110 defining an interior volume for containing waste material. Body 110 includes sections 112 each having a first diameter and a section 114 having a second diameter that may be less than the first diameter. Container 100 further has a lid 120 positioned at a top end of body 110 and a tube 140 extending from lid 120 which communicates with the interior volume of body 110. The interior volume of body 110 is filled with waste material via tube 140. Following hot isostatic pressing, as shown in FIG. 1B, the volume of body 110 is substantially reduced and container 100 is then sealed. Typically, tube 140 is crimped, cut, and welded by linear seam welding. One drawback in such a process is that cutting of tube 140 can create secondary waste as the removed portion of tube 140 may contain amounts of residual waste material which must then be disposed of in a proper manner. Moreover, the tools used for cutting tube 140 may be exposed to the residual waste material and/or require regular maintenance or replacement due to wear. Also, this system requires complex mechanical or hydraulic systems to be in the hot cell (radioactive environment) near the can to be sealed reducing the life of seals on hydraulic rams and the equipment is bulky taking up additional space in the hot cell. It is therefore desirable to have systems, methods, filling equipment and containers for storing hazardous waste material that can avoid one or more of these drawbacks. FIG. 2 schematically represents an exemplary process flow 200 used to dispose of nuclear waste, such as calcined material, in accordance with the present invention. Process 200 may be performed using a modular system 400, exemplary embodiments of which are illustrated in subsequent figures, wherein the hazardous waste is processed or moved in a series of isolated cells. Modular system 400 may be referred to as including the “hot cell” or “hot cells”. In some embodiments, each cell is isolated from the outside environment and other cells such that any spillage of hazardous waste may be contained within the cell in which the spill occurred. Modular system 400 in accordance with the present invention may be used to process liquid or solid hazardous waste material. The hazardous waste material may be a radioactive waste material. A radioactive liquid waste may include aqueous wastes resulting from the operation of a first cycle solvent extraction system, and/or the concentrated wastes from subsequent extraction cycles in a facility for reprocessing irradiated nuclear reactor fuels. These waste materials may contain virtually all of the nonvolatile fission products, and/or detectable concentrations of uranium and plutonium originating from spent fuels, and/or all actinides formed by transmutation of the uranium and plutonium as normally produced in a nuclear reactor. In one embodiment, the hazardous waste material includes calcined material. Modular system 400 may be divided into two or more cells. In one embodiment, modular system 400 includes at least four separate cells. In one embodiment, modular system 400 includes four separate cells. In one such embodiment, the series of cells include a first cell 217, which may be a filling cell, a second cell 218, which may be a bake-out and vacuum scaling cell, a third cell 232 which may be a process cell, and a fourth cell 230 which may be a cooling and packaging cell, each of which will be discussed in more detail below. In one embodiment, first cell 217 includes a feed blender 212 configured to mix a hazardous waste material with one or more additives. In one embodiment, a container feed hopper 214 is coupled to feed blender 212. In one embodiment, container feed hopper 214 is coupled with a fill system to transfer the hazardous waste material and additive mixture into container 216. In some embodiments, calcined material is transferred from a surge tank 205 to a calcined material receipt hopper 207 configured to supply feed blender 212. In some embodiments, additives are supplied to feed blender 212 from hopper 210. In some embodiments, the additives are transferred to hopper 210 from tank 201. After being filled, container 216 is removed from first cell 217 and transferred to second cell 218 where bake-out and vacuum sealing steps take place. In some embodiments, the bake-out process includes heating container 216 in a furnace 290 to remove excess water, for example, at a temperature of about 400° C. to about 500° C. In some embodiments, off-gas is removed from container 216 during the bake-out process and routed through line 206, which may include one or more filters 204 or traps 219 to remove particulates or other materials. In further embodiments, a vacuum is established in container 216 during the bake-out process and container 216 is sealed to maintain the vacuum. After the bake-out and sealing steps, according to some embodiments, container 216 is transferred to third cell 232 where the container 216 is subjected to hot isostatic pressing or HIP, for example, at elevated temperature of 1000° C.-1250° C. and elevated argon pressure supplied from a compressor 234 and argon source 236. In some embodiments, hot isostatic pressing results in compaction of container 216 and the waste material contained therein. After the hot isostatic pressing, according to some embodiments, container 216 is transferred to fourth cell 230 for cooling and/or packaging for subsequent loading 203 for transport and storage. Modular system 400 may be configured in numerous ways depending on the spatial arrangement of the plurality of cells. In an embodiment, the plurality of cells may have any suitable spatial arrangement, including a lateral arrangement of cells, a vertical arrangement of cells or a combination of laterally arranged cells and vertical arranged cells. In one embodiment, modular system 400 comprises a plurality of cells spatially arranged in a single row of contiguous cells, wherein each cell is isolated from an adjacent cell. In another embodiment, the plurality of cells may be spatially arranged in a single row of contiguous cells, wherein each cell may be isolated from an adjacent cell by at least one common side wall. In another embodiment, the plurality of cells may be arranged vertically in space in single column of contiguous cells, wherein each cell is isolated from an adjacent cell by at least one common wall. In yet another embodiment, the plurality of cells may be spatially arranged in a plurality of rows of contiguous cells. In one embodiment, modular system 400 includes a first cell 217, a second cell 218, and a third cell 232, first cell 217 being adjacent second cell 218 and contiguous therewith, and third cell 232 being adjacent to second cell 218 and being contiguous therewith, wherein first cell 217, second cell 218 and third cell 232 are spatially arranged in a single row of cells. Modular system 400 may contain one or more assembly lines that move containers 216 sequentially through modular system 400. As illustrated in FIGS. 2-4, an exemplary modular system 400 for processing and/or storing and/or disposing of a hazardous waste material includes parallel assembly lines of a plurality of cells for manipulating container 216. In some embodiments, as described above, the plurality of cells for manipulating container 216 includes at least first cell 217, second cell 218, third cell 232 and fourth cell 230. In other embodiments, any number of cells may be provided. In some embodiments, the cells may be held at different pressures relative to adjacent cells to control contamination from spreading between cells. For example, each subsequent cell may have a higher pressure than the previous cell such that any air flow between cells flows toward the beginning of the process. In some embodiments, first cell 217 is held at a first pressure P1 and second cell 218 is held at a second pressure P2. In one embodiment, first pressure P1 is less than second pressure P2. In such embodiments, first cell 217 does not exchange air with second cell 218 at least during the time when container 216 is being manipulated in first cell 217. In another such embodiment, an air interlock 241 (see FIG. 12), as described in further detail below, couples first cell 217 to second cell 218 and is configured to allow transfer of container 216 from first cell 217 to second cell 218 while maintaining at least one seal between first cell 217 and second cell 218. In another embodiment, first cell 217 is held at first pressure P1, second cell is held at second pressure P2 and third cell 232 is held at a third pressure P3, where third pressure P3 is greater than second pressure P2 which is greater than first pressure P1. In such embodiments, third cell 232 is isolated from first cell 217 and second cell 218, wherein second cell 218 and third cell 232 are configured to allow transfer of container 216 from second cell 218 to third cell 232. In yet another embodiment, first cell 217 is held at first pressure P1, second cell 218 is held at second pressure P2, third cell 232 is held at third pressure P3 and fourth cell 230 is held at a fourth pressure P4, wherein fourth pressure P4 is greater than third pressure P3, third pressure P3 is greater than second pressure P2 which is greater than first pressure P1. In such embodiments, fourth cell 230 is isolated from first cell 217, second cell 218 and third cell 232, wherein third cell 232 and fourth cell 230 are configured to allow transfer of container 216 from third cell 232 to the fourth cell 230. In one embodiment, each pressure P1, P2, P3 and/or P4 is negative relative to normal atmospheric pressure. In some embodiments, the pressure difference between first cell 217 and second cell 218 is about 10 KPa to about 20 KPa. In some embodiments, the pressure difference between second cell 218 and third cell 232 is about 10 KPa to about 20 KPa. In some embodiments, the pressure difference between third cell 232 and fourth cell 230 is about 10 KPa to about 20 KPa. I. First Cell Exemplary embodiments of first cell 217 are illustrated in FIGS. 3, 4 and 7. In one embodiment, first cell 217 is a filling cell which allows for filling a container 216 with hazardous waste with minimal contamination of the exterior of container 216. In one embodiment, empty containers 216 are first introduced into the modular system 400. In one embodiment, empty containers 216 are placed in first cell 217 and first cell 217 is sealed before transferring any hazardous waste material within first cell 217. In one embodiment, once first cell 217 is scaled and contains one or more empty containers 216, first cell 217 is brought to pressure P1. Container and Method of Filling a Container Containers of various designs may be used in accordance with the various embodiments of the present disclosure. A schematic container 216, which may be a HIP can, is shown throughout in FIGS. 2, 3, 4, 7, 13, 15, 16 and 17. Container 216 may have any suitable configuration known in the art for HIP processing. In some embodiments, container 216 is provided with a single port. In other embodiments, container 216 is provided with a plurality of ports. Some particular configurations for containers 216 that may be used in accordance with the various embodiments of the present invention are shown in FIGS. 5A, 5B, 6A and 6B, which illustrate exemplary containers configured to sealingly contain hazardous waste material in accordance with the present disclosure. FIGS. 5A and 6A show one embodiment of a container, generally designated 500, for containment and storage of nuclear waste materials or other desired contents in accordance with an exemplary embodiment of the present invention. Container 500, in some embodiments, is particularly useful in HIP processing of waste materials. It should however be understood that container 500 can be used to contain and store other materials including nonnuclear and other waste materials. According to some embodiments, container 500 generally includes body 510, lid 520, filling port 540, and evacuation port 560. In some embodiments, container 500 also includes filling plug 550 configured to engage with filling port 540. In further embodiments, container 500 also includes evacuation plug 570 configured to engage with evacuation port 560. In yet further embodiments, container 500 includes lifting member 530. Body 510 has a central longitudinal axis 511 and defines interior volume 516 for containing nuclear waste materials or other materials according to certain embodiments of the invention. In some embodiments, a vacuum can be applied to interior volume 516. In some embodiments, body 510 has a cylindrical or a generally cylindrical configuration having closed bottom end 515. In some embodiments, body 510 is substantially radially symmetric about central longitudinal axis 511. In some embodiments, body 510 may be configured to have the shape of any of the containers described in U.S. Pat. No. 5,248,453, which is incorporated herein by reference in its entirety. In some embodiments, body 510 is configured similarly to body 110 of container 100 shown in FIG. 1. Referring to FIG. 5A, in some embodiments body 510 has one or more sections 512 having a first diameter alternating along central longitudinal axis 511 with one or more sections 514 having a smaller second diameter. Body 510 may have any suitable size. In some embodiments, body 510 has a diameter in a range of about 60 mm to about 600 mm. In some embodiments, body 510 has a height in a range of about 120 mm to about 1200 mm. In some embodiments, body 510 has a wall thickness of about 1 mm to about 5 mm. Body 510 may be constructed from any suitable material known in the art useful in hot isostatic pressing of nuclear waste materials. In some embodiments, body 510 is constructed of material capable of maintaining a vacuum within body 500. In some embodiments, body 510 is constructed from a material that is resistant to corrosion. In some embodiments, body 510 is made from a metal or metal alloy, for example, stainless steel, copper, aluminum, nickel, titanium, and alloys thereof. In some embodiments, container 500 includes a lid 520 opposite closed bottom end 515. Lid 520, in some embodiments, is integrally formed with body 510. In other embodiments, lid 520 is formed separately from body 510 and secured thereto, for example, via welding, soldering, brazing, fusing or other known techniques in the art to form a hermetic seal circumferentially around lid 520. In some embodiments, lid 520 is permanently secured to body 510. Referring to FIG. 6A, lid 520 includes interior surface 524 facing interior volume 516 and exterior surface 526 opposite interior surface 524. In some embodiments, central longitudinal axis 511 is substantially perpendicular to interior surface 524 and exterior surface 526. In some embodiments, central longitudinal axis 511 extends through a center point of interior surface 524 and exterior surface 526. In some embodiments, container 500 further includes a flange 522 surrounding exterior surface 526. In some embodiments, container 500 further includes a filling port 540 having an outer surface 547, an inner surface 548 defining a passageway in communication with interior volume 516, and configured to couple with a filling nozzle. In some embodiments, the nuclear waste material to be contained by container 500 is transferred into interior volume 516 through filling port 540 via the filling nozzle. In some embodiments, filling port 540 is configured to at least partially receive the filling nozzle therein. In some embodiments, inner surface 548 of filling port 540 is configured to form a tight seal with a filling nozzle so as to prevent nuclear waste material from exiting interior volume 516 between inner surface 548 of filling port 540 and the filling nozzle during filling of container 500. Filling port 540 may extend from lid 520 as shown in the exemplary embodiment of FIGS. 5A and 6A. In some embodiments, filling port 540 may be integrally formed with lid 520. In other embodiments, filling port 540 is formed separately from lid 520 and secured thereto, for example, by welding. In some embodiments, filling port 540 is constructed from metal or metal alloy, and may be made from the same material as body 510 and/or lid 520. Referring particularly to FIG. 6A, filling port 540 has a generally tubular configuration with inner surface 548 extending from first end 542 towards second end 543. According to some embodiments, filling port 540 extends from lid 520 along an axis 541 substantially parallel to central longitudinal axis 511. In some embodiments, inner surface 548 is radially disposed about axis 541. In some embodiments, first end 542 of filling port 540 defines an opening in lid 520 and has an internal diameter Df1. In some embodiments, second end 543 of filling port 540 has an internal diameter Df2 which may be different than diameter Df1. In some embodiments, Df2 is larger than Df1. In one embodiment, for example, Df1 is about 33 mm and Df2 is about 38 mm. In some embodiments, a stepped portion 549 is provided on the exterior of filling port 540. In some embodiments, stepped portion can be used for positioning an orbital welder (e.g., orbital welder 242 described herein below). Container 500, in some embodiments, further includes a filling plug 550 configured to couple with filling port 540. In some embodiments, filling plug 550 is configured and dimensioned to be at least partially received in filling port 540 as generally shown in FIG. 6A. In some embodiments, filling plug 550 is radially disposed about axis 541 when coupled with filling port 540. In some embodiments, filling plug 550 is configured to close and seal filling port 540 to prevent material from exiting interior volume 516 via filling port 540. Filling plug 550, in some embodiments, is configured to abut inner surface 548 when coupled to filling port 540. In some embodiments, filling plug 550 includes a portion having a diameter substantially equal to an internal diameter of filling port 540. In some embodiments, filling plug 550 includes a first portion 552 having a diameter substantially equal to Df1. In some embodiments, filling plug 550 alternatively or additionally includes a second portion 553 having a diameter substantially equal to Df2. In some embodiments, second portion 553 is configured to abut surface 544 when filling plug 550 is coupled with filling port 540. In some embodiments, filling plug 550 further abuts end surface 545 when filling plug 550 is coupled with filling port 540. In some embodiments, filling plug 550 when coupled with filling port 540 creates a seam 546. In some embodiments, seam 546 is formed at an interface between filling plug 550 and end surface 545 of second end 543 of filling port 540. In some embodiments, seam 546 is located between external surface 551 of filling plug 550 and external surface 547 of filling port 540. In some embodiments, external surface 551 of filling plug 550 is substantially flush with external surface 547 of filling port 540 proximate seam 546. Seam 546 extends circumferentially around a portion of filling plug 550 according to some embodiments. Filling port 540 and filling plug 550 may be secured together according to some embodiments by any suitable method known in the art. In some embodiments, filling plug 550 is threadably coupled with filling port 540. According to some of these embodiments, at least a portion of inner surface 548 is provided with internal threads that are configured to engage with external threads provided on at least a portion of filling plug 550 such that, for example, filling plug 550 may be screwed into filling port 540. In some embodiments, one or more of portions 552 and 553 may be provided with external threads that engage with internal threads provided on inner surface 548 of filling port 540. In other embodiments, filling port 540 and filling plug may be coupled via an interference or friction fit. In some embodiments, container 500 includes a gasket (not shown) positioned within filling port 540 to aid in sealing filling port 540 with filling plug 550. In some embodiments, a gasket is positioned between filling plug 550 and surface 544. In some embodiments, filling port 540 and filling plug 550 may be permanently secured together after filling of container 500 with the nuclear waste material or other desired contents. In some embodiments, filling port 540 and filling plug 550 may be mechanically secured together. In some embodiments, filling port 540 may be fused with filling plug 550. In some embodiments, filling port 540 and filling plug 550 may be soldered or brazed together. In some embodiments, filling port 540 and filling plug 550 may be welded together along seam 546, for example, by orbital welding. In other embodiments, an adhesive or cement may be introduced into seam 546 to seal filling port 540 and filling plug 550 together. In some embodiments, container 500 includes an evacuation port 560 having an outer surface 567 and an inner surface 568 defining a passageway in communication with interior volume 516. In some embodiments, evacuation port 560 is configured to allow venting of air or other gas from interior volume 516. In some embodiments, evacuation port 560 is configured to couple with an evacuation nozzle, as described further below, for evacuating air or other gas from interior volume 516. In some embodiments, the evacuation nozzle is connected with a ventilation or vacuum system capable of drawing air or other gas from interior volume 516 through evacuation port 560. Evacuation port 560 may extend from lid 520 as shown in the exemplary embodiment of FIGS. 5A and 6A. In some embodiments, evacuation port 560 may be integrally formed with lid 520. In other embodiments, evacuation port 560 is formed separately from lid 520 and secured thereto, for example, by welding, soldering, brazing, or the like. In some embodiments, evacuation port 560 is constructed from metal or metal alloy, and may be made from the same material as body 510 and/or lid 520. Referring particularly to FIG. 6A, evacuation port 560 has a generally tubular configuration with inner surface 568 extending from first end 562 towards second end 563. According to some embodiments, evacuation port 560 extends from lid 520 along an axis 561 substantially parallel to central longitudinal axis 511. In some embodiments, axis 561 is coplanar with central longitudinal axis 511 and axis 541 of filling port 540. In some embodiments, inner surface 568 is radially disposed about axis 561. In some embodiments, first end 562 of evacuation port 560 defines an opening in lid 520 and has an internal diameter De1. In some embodiments, second end 563 of evacuation port 560 has an internal diameter De2 which may be different than diameter De1. In some embodiments, De2 is larger than De1. In some embodiments, evacuation port 560 may further include one or more intermediate sections positioned between first end 562 and second end 563 defining internal diameters different than De1 and De2. In the exemplary embodiment shown in FIG. 6A, evacuation port 560 includes intermediate sections 564 and 565 respectively have internal diameters De3 and De4 and configured such that De1<De3<De4<De2. In some embodiments, evacuation port 560 has the same external diameter as filling port 540. In some embodiments, a stepped portion 569 is provided on the exterior of evacuation port 560. In some embodiments, stepped portion 569 can be used for positioning an orbital welder (e.g. orbital welder 242 described therein below). In some embodiments, stepped portion 569 can be used for positioning the evacuation nozzle. According to some embodiments of the invention, evacuation port 560 is provided with a filter 590. In some embodiments, filter 590 is sized to span across the passageway defined by evacuation port 560. In some embodiments, filter 590 is positioned within evacuation port 560 at or proximate to first end 562 and has a diameter substantially equal to De1. In some embodiments, the filter 590 is sealingly engaged to inner surface 568 of evacuation port 560. In some embodiments, the filter 590 is secured to inner surface 568 of evacuation port 560, for example, via welding, soldering, brazing, or the like. In one embodiment, filter 590 is a high efficiency particulate air (HEPA) filter. In some embodiments, filter 590 is a single layer of material. In some embodiments, filter 590 is multi-layer material. In some embodiments, filter 590 is made from sintered material. In some embodiments, filter 590 is made from metal or metal alloy, for example, stainless steel, copper, aluminum, iron, titanium, tantalum, nickel, and alloys thereof. In some embodiments, filter 590 is made from a ceramic, for example, aluminum oxide (Al2O3) and zirconium oxide (ZrO2). In some embodiments, filter 590 includes carbon or a carbon compound, for example, graphite. In some embodiments, the material of filter 590 is chosen so that upon heating the filter densifies into a solid and non-porous material. In some embodiments, the material of filter 590 is chosen wherein at a first temperature filter 590 is porous to air and/or gas but prevents passage of particles and at a second temperature filter 590 densifies into a non-porous material, wherein the second temperature is greater than the first temperature. In some embodiments, filter 590 is configured to prevent passage of particles having a predetermined dimension through evacuation port 560 while allowing passage of air or other gas. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 100 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 75 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 50 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 25 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 20 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 15 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 12 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 10 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 8 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 5 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 1 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 0.5 μm through evacuation port 560. In some embodiments, filter 590 is configured to prevent passage of particles having a dimension greater than 0.3 μm through evacuation port 560. Container 500, in some embodiments, further includes an evacuation plug 570 configured to couple with evacuation port 560. In some embodiments, evacuation plug 570 is configured and dimensioned to be at least partially received in evacuation port 560 as generally shown in FIG. 6A. In some embodiments, evacuation plug 570 is radially disposed about axis 561 when coupled with filling port 560. In some embodiments, evacuation plug 570 is configured to allow air and/or other gas to pass through evacuation port 560 in a filling configuration and to close filling evacuation port 560 in a closed configuration to prevent air and/or other gas from passing through evacuation port 560. In some embodiments, evacuation plug 570 includes a portion having a diameter substantially equal to or slightly less than an internal diameter of evacuation port 560. In some embodiments, evacuation plug 570 includes a first portion 572 having a diameter substantially equal to or slightly less than De1. In some embodiments, evacuation plug 570 alternatively or additionally includes a second portion 573 having a diameter substantially equal to De2. In some embodiments, evacuation plug 570 alternatively or additionally includes intermediate portions 574 and 575 having respective diameters substantially equal to or slightly less than De3 and De4. In some embodiments, evacuation plug 570 when coupled with evacuation port 550 creates a seam 566. In some embodiments, seam 566 is formed at an interface between evacuation plug 570 and second end 563 of evacuation port 560. In some embodiments, seam 566 is located between external surface 571 of evacuation plug 570 and external surface 567 of evacuation port 560. In some embodiments, external surface 571 of evacuation plug 570 is substantially flush with external surface 567 of evacuation port 560 proximate seam 566. Seam 566 extends circumferentially around a portion of evacuation plug 570 according to some embodiments. According to some embodiments of the invention, evacuation plug 570 is configured to be at least partially received within evacuation port 560 in a filling configuration such that air and/or other gas is allowed to exit from interior volume 516 of container 500 through filter 590 and through evacuation port 560 between inner surface 568 of evacuation port 560 and evacuation plug 570. In some embodiments, evacuation plug 570 and evacuation port 560 are coupled in the filling configuration such that a gap 582 of sufficient dimension to allow for air and/or other gas to pass there through is maintained between evacuation plug 570 and evacuation port 560 to provide a pathway for air and/or other gas to evacuated from interior volume 516. In some embodiments, gap 582 extends circumferentially around at least a portion of evacuation plug 570. In some embodiments, air and/or other gas is allowed to pass through gap 582 and through seam 566 in the filling configuration. In some embodiments, evacuation plug 570 and evacuation port 560 are coupled in the filling configuration such that a space 581 is maintained between evacuation plug 570 and filter 590. When present, space 581 should be of sufficient distance along the axial direction (e.g., along axis 561) to allow for air and/or other gas to pass through filter 590. In some embodiments, container 500 is further configured to transition from the filling configuration to a closed configuration wherein the evacuation plug 570 is coupled with evacuation port 560 such that air and/or other gas is not allowed to pass through evacuation port 560. In some embodiments, evacuation port 560 is hermetically sealed by the evacuation plug 570 in the closed configuration. In some embodiments, the closed configuration allows a vacuum to be maintained in interior volume 516. In some embodiments, in the closed configuration, evacuation plug 570 is at least partially received within evacuation port 560 to close and seal the passageway defined by evacuation port 560 to prevent material from passing therethrough. In some embodiments, a gasket 580 is provided between evacuation port 560 and evacuation plug 570. In some embodiments, gasket 580 aids in sealing the evacuation port 560 with the evacuation plug 570 in the closed configuration. Gasket 580, in some embodiments, surrounds at least a portion of evacuation plug 570. In the embodiment of FIG. 6A, gasket 580 is shown surrounding portion 575 of evacuation plug 570 and is positioned between and configured to abut second portion 573 of evacuation plug 570 and intermediate section 565 of evacuation port 560. In some embodiments, gasket 580 can be made from a metal or metal alloy, for example stainless steel, copper, aluminum, iron, titanium, tantalum, nickel, and alloys thereof. In some embodiments, gasket 580 is made from a ceramic, for example, aluminum oxide (Al2O3) and zirconium oxide (ZrO2). In some embodiments, gasket 580 includes carbon or a carbon compound, for example, graphite. In some embodiments, evacuation plug 570 is threadably coupled with evacuation port 560. According to some of these embodiments, at least a portion of inner surface 568 is provided with internal threads that are configured to engage with external threads provided on at least a portion of evacuation plug 570 such that, for example, evacuation plug 570 may be screwed into evacuation port 560. In some embodiments, one or more of portions 572, 573, 574, and 575 may be provided with external threads that engage with internal threads provided on inner surface 568 of evacuation port 560. In some embodiments, the filling configuration includes partially engaging the external threads of evacuation plug 570 with the internal threads of evacuation port 560 (e.g., partially screwing evacuation plug 570 into evacuation port 560) and the closed configuration includes fully engaging the external threads of evacuation plug 570 with the internal threads of evacuation port 560 (e.g., fully screwing evacuation plug 570 into evacuation port 560). In some embodiments, evacuation port 560 and evacuation plug 570 may be permanently secured together. In some embodiments, evacuation port 560 and evacuation plug 570 may be mechanically secured together. In some embodiments, evacuation port 560 may be fused with evacuation plug 570. In some embodiments, evacuation port 560 and evacuation plug 570 may be soldered or brazed together. In some embodiments, evacuation port 560 and evacuation plug 570 may be welded together along seam 566, for example, by orbital welding. In such embodiments, the weld is placed between the evacuation port 560 and evacuation plug 570 away from the gasket 580 so not to disrupt the hermetic seal maintaining the atmosphere in the container 500. In other embodiments, an adhesive or cement may be introduced into seam 566 to seal evacuation port 560 and evacuation plug 550 together. Referring to FIGS. 5A and 6A, container 500, in some embodiments, includes lifting member 530 which is configured to engage with a carrier for lifting and/or transporting container 500. Lifting member 530, according to some embodiments, is securely attached to and extends from exterior surface 526 of lid 520. In some embodiments, lifting member 530 is positioned centrally on exterior surface 526 of lid 520. In some embodiments, lifting member 530 is integrally formed with lid 520. In other embodiments, lifting member is formed separately from lid 520 and secured thereto, for example, by welding, soldering, brazing, or the like. In some embodiments, lifting member 530 is constructed from metal or metal alloy, and may be made from the same material as body 510 and/or lid 520. In the exemplary embodiment shown, lifting member 530 includes a generally cylindrical projection 532 extending from lid 520 substantially co-axial with central longitudinal axis 511. In some embodiments, lifting member 530 is radially symmetric about central longitudinal axis 511. In some embodiments, lifting member 530 is positioned on lid 520 between filling port 540 and evacuation port 560. In some embodiments, lifting member 530 includes a groove 533 that extends at least partially around the circumference of projection 532. In further embodiments, lifting member 530 includes a flange 534 that partially defines groove 533. FIGS. 5B and 6B show another embodiment of a container, generally designated 600, for containment and storage of nuclear waste materials or other desired contents in accordance with an exemplary embodiment of the present invention. Container 600, in some embodiments, is particularly useful in hot isostatic pressing of waste materials. In some embodiments, body 610 is constructed of material capable of maintaining a vacuum within body 600. According to some embodiments, container 600 generally includes body 610, lid 620, and filling port 640. In some embodiments, container 600 also includes filling plug 650 configured to engage with filling port 640. Body 610 has a central longitudinal axis 611 and defines interior volume 616 for containing nuclear waste materials or other materials according to certain embodiments of the invention. In some embodiments, a vacuum can be applied to interior volume 616. In some embodiments, body 610 has a cylindrical or a generally cylindrical configuration having closed bottom end 615. In some embodiments, body 610 is substantially radially symmetric about central longitudinal axis 611. In some embodiments, body 610 may be configured to have the shape of any of the containers described in U.S. Pat. No. 5,248,453, which is incorporated herein by reference in its entirety. In some embodiments, body 610 is configured similarly to body 110 of container 100 shown in FIG. 1. Referring to FIG. 5B, in some embodiments body 610 has one or more sections 612 having a first diameter alternating along central longitudinal axis 611 with one or more sections 614 having a smaller second diameter. Body 610 may have the same configuration and dimensions described herein for body 510. Body 610 may be constructed from any suitable material known in the art useful in hot isostatic pressing of nuclear waste materials. In some embodiments, body 610 is constructed from a material that is resistant to corrosion. In some embodiments, body 610 is made from a metal or metal alloy, for example, stainless steel, copper, aluminum, nickel, titanium, and alloys thereof. In some embodiments, container 600 includes a lid 620 opposite closed bottom end 615, Lid 620, in some embodiments, is integrally formed with body 610. In other embodiments, lid 620 is formed separately from body 610 and secured thereto, for example, via welding, soldering, brazing, fusing or other known techniques in the art to form a hermetic seal circumferentially around lid 620. In some embodiments, lid 620 is permanently secured to body 610. Referring to FIG. 6B, lid 620 includes interior surface 624 facing interior volume 616 and exterior surface 626 opposite interior surface 624. In some embodiments, central longitudinal axis 611 is substantially perpendicular to interior surface 624 and exterior surface 626. In some embodiments, central longitudinal axis 611 extends through a center point of interior surface 624 and exterior surface 626. In some embodiments, container 600 further includes a flange 622 surrounding exterior surface 626. In some embodiments, container 600 further includes a filling port 640 having an outer surface, a stepwise inner surface 647 and a lower inner surface 648 defining a passageway in communication with interior volume 616, and configured to couple with a filling nozzle. In some embodiments, the nuclear waste material to be contained by container 600 is transferred into interior volume 616 through filling port 640 via the filling nozzle. In some embodiments, filling port 640 is configured to at least partially receive the filling nozzle therein. In some embodiments, stepwise inner surface 647 and/or lower inner surface 648 of filling port 640 is configured to form a tight seal with a filling nozzle so as to prevent nuclear waste material from exiting interior volume 616 between stepwise inner surface 647 and lower inner surface 648 of filling port 640 and the filling nozzle during filling of container 600. Filling port 640 may extend from lid 620 as shown in the exemplary embodiment of FIGS. 5B and 6B. In some embodiments, filling port 640 may be integrally formed with lid 620. In other embodiments, filling port 640 is formed separately from lid 620 and secured thereto, for example, by welding. In some embodiments, filling port 640 is constructed from metal or metal alloy, and may be made from the same material as body 610 and/or lid 620. Referring particularly to FIG. 6B, filling port 640 has a generally step wise tubular configuration with stepwise inner surface 647 and lower inner surface 648 extending from first end 642 towards second end 643. According to some embodiments, filling port 640 extends from lid 620 along an axis 641 substantially coaxial to central longitudinal axis 611. In some embodiments, stepwise inner surface 647 is radially disposed about axis 641. In some embodiments, lower inner surface 648 is radially disposed about axis 641. In some embodiments, first end 642 of filling port 640 defines an opening in lid 620 and has an internal diameter Dg1. In some embodiments, second end 643 of filling port 640 has an internal diameter Dg2 which may be different than diameter Dg1. In some embodiments, Dg2 is larger than Dg1. In some embodiments, filling port 640 is provided with a flange 634 at least partially defining a groove 633. In some embodiments, flange 634 and groove 633 extend circumferentially around filling port 640. In some embodiments, flange 634 and groove 633 are radially symmetric about axis 641. In some embodiments, flange 634 and/or groove 633 are configured to engage with a carrier for lifting or transporting container 600. Container 600, in some embodiments, further includes a filling plug 650 configured to couple with filling port 640. In some embodiments, filling plug 650 is configured and dimensioned to be at least partially received in filling port 640 as generally shown in FIG. 6B. In some embodiments, filling plug 650 is radially disposed about axis 641 when coupled with filling port 640. In some embodiments, filling plug 650 is configured to close and seal filling port 640 to prevent material from exiting interior volume 616 via filling port 640. In some embodiments, filling plug 650 is configured for hermetically sealing filling port 640. Filling plug 650, in some embodiments, is configured to abut stepwise inner surface 647 when coupled to filling port 640. In some embodiments, filling plug 650 includes a first portion 673 having a diameter substantially equal to Dg2. In some embodiments, filling plug 650 alternatively or additionally includes a second portion 675 having a diameter substantially equal to Dg3. In some embodiments, filling plug 650 alternatively or additionally includes a third portion 674 having a diameter substantially equal to Dg4. In some embodiments, first portion 673 is configured to abut surface 649 when filling plug 650 is coupled with filling port 640. In some embodiments, filling plug 650 when coupled with filling port 640 creates a seam 646. In some embodiments, seam 646 is formed at an interface between filling plug 650 and end surface 645 of second end 643 of filling port 640. In some embodiments, seam 646 is located between an external surface of filling plug 650 and an external surface of filling port 640. In some embodiments, the external surface of filling plug 650 is substantially flush with the external surface of filling port 640 proximate seam 646. Seam 646 extends circumferentially around a portion of filling plug 650 according to some embodiments. Filling port 640 and filling plug 650 may be secured together according to some embodiments by any suitable method known in the art. In some embodiments, filling plug 650 is threadably coupled with filling port 640. According to some of these embodiments, at least a portion of inner surface 648 is provided with internal threads that are configured to engage with external threads provided on at least a portion of filling plug 650 such that, for example, filling plug 650 may be screwed into filling port 640. In some embodiments, one or more of portions 652 and 653 may be provided with external threads that engage with internal threads provided on inner surface 648 of filling port 640. In other embodiments, filling port 640 and filling plug may be coupled via an interference or friction fit. In some embodiments, a gasket 680 is provided between filling port 640 and filling plug 650. In some embodiments, gasket 680 aids in sealing the filling port 640 with the filling plug 650 in a closed configuration. Gasket 680, in some embodiments, surrounds at least a portion of filling plug 650. In the embodiment of FIG. 6B, gasket 680 is shown surrounding portion 675 of filling plug 650 and is positioned between and configured to abut portion 673 of filling plug 650 and filling port 640. In some embodiments, gasket 680 can be made from a metal or metal alloy, for example stainless steel, copper, aluminum, iron, titanium, tantalum, nickel, and alloys thereof. In some embodiments, gasket 680 is made from a ceramic, for example, aluminum oxide (Al2O3) and zirconium oxide (ZrO2). In some embodiments, gasket 680 includes carbon or a carbon compound, for example, graphite. In some embodiments, filling port 640 and filling plug 650 may be permanently secured together after filling of container 600 with the nuclear waste material or other desired contents. In some embodiments, filling port 640 and filling plug 650 may be mechanically secured together. In some embodiments, filling port 640 may be fused with filling plug 650. In some embodiments, filling port 640 and filling plug 650 may be soldered or brazed together. In some embodiments, filling port 640 and filling plug 650 are configured to provide a hermetic seal. In some embodiments, filling port 640 and filling plug 650 may be welded together along seam 646, for example, by orbital welding. In such embodiments, the weld is placed between the filling plug 650 and filling port 640 away from the gasket 680 so as not to disrupt the hermetic seal maintaining the atmosphere in the container 600. In other embodiments, an adhesive or cement may be introduced into seam 646 to seal filling port 640 and filling plug 650 together. According to some embodiments of the invention, filling plug 650 is provided with a filter 690. In some embodiments, filter 690 is sized to span the circular end section 670 of filling port 650. In some embodiments, the filter 690 is sealingly engaged to circular end section 670 of filling plug 650. In some embodiments, the filter 690 is secured to circular end section 670 of filling plug 650, for example, via welding, soldering, brazing, or the like. In some embodiments, filter 690 is secured to filling plug 650 with a mechanical fastener 695, such as a screw, nail, bolt, staple, or the like. In one embodiment, filter 690 is a high efficiency particulate air (HEPA) filter. In some embodiments, filter 690 is a single layer of material. In some embodiments, filter 690 is multi-layer material. In some embodiments, filter 690 is made from sintered material. In some embodiments, filter 690 is made from metal or metal alloy, for example, stainless steel, copper, aluminum, iron, titanium, tantalum, nickel, and alloys thereof. In some embodiments, filter 690 is made from a ceramic, for example, aluminum oxide (Al2O3), aluminosilicates (eg. Al2SiO5) and zirconium oxide (ZrO2). In some embodiments, filter 690 includes carbon or a carbon compound, for example, graphite. In some embodiments, the material of filter 690 is chosen so that upon heating the filter densifies into a solid and non-porous material. In some embodiments, the material of filter 690 is chosen wherein at a first temperature filter 690 is porous to air and/or gas but prevents passage of particles and at a second temperature filter 690 densifies into a non-porous material, wherein the second temperature is greater than the first temperature. In some embodiments, filter 690 is configured to prevent passage of particles having a predetermined dimension through filling port 640 while allowing passage of air or other gas when filling plug 650 is coupled with filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 100 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 75 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 50 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 25 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 20 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 15 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 12 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 10 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 8 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 5 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 1 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 0.5 μm through filling port 640. In some embodiments, filter 690 is configured to prevent passage of particles having a dimension greater than 0.3 μm through filling port 640. According to some embodiments of the invention, filling plug 650 is configured to be at least partially received within filling port 640 in a filling configuration such that air and/or other gas is allowed to exit from interior volume 616 of container 600 through filter 690 and between stepwise inner surface 647 of filling port 640 and filling plug 650. In some embodiments, filling plug 650 and filling port 640 are coupled in the filling configuration such that a gap (not shown) of sufficient dimension to provide a pathway for air and/or other gas to evacuated from interior volume 616. In some embodiments, the gap extends circumferentially around at least a portion of filling plug 650. In some embodiments, air and/or other gas is allowed to pass through the gap and through seam 646 in the filling configuration. In operation, the interior volume of a container 216 is filled with material by coupling a filling port 540 to a filling nozzle 260 wherein container 216 is place under a negative pressure prior to filling or container 216 is simultaneously evacuated during the filling process according to some embodiments. In some embodiments, the filling port 540 is configured to tightly fit around the filling nozzle 260 to prevent material from exiting container 216 between the filling port 540 and the filling nozzle 260. In some embodiments, the filling of container 216 continues until the desired amount of material has been added to container 216. In some embodiments, a predetermined volume of material is added to container 216. In some embodiments, a predetermined weight of material is added to container 216. With reference to FIG. 6A, material to be stored (e.g., nuclear waste or calcined material) is added to interior volume 516 of container 500 via a filling nozzle 260 coupled to filling port 540 according to some embodiments. In some embodiments, the filling port 540 is configured to tightly fit around filling nozzle 260 to prevent material from exiting container 500 between the filling port 540 and filling nozzle 260. In some embodiments, as container 516 is being filled, air and/or other gas contained in interior volume 516 is evacuated from container 500 via evacuation port 560 provided with filter 590. In some embodiments, filter 590 prevents all or at least most non-gaseous materials from exiting container 500 through evacuation port 560 while the air and/or other gas is being evacuated from interior volume 516. In some embodiments, filter 590 is configured to prevent particles having a diameter of at least 10 μm from exiting interior volume 516 through evacuation port 560 during filling of waste material and air/gas evacuation. Evacuation of the air and/or other gas, in some embodiments, can be facilitated by coupling evacuation port 560 with an evacuation nozzle 300. Evacuation nozzle 300 may be coupled with an evacuation line or system (e.g., a vacuum source). In some embodiments, the evacuation line is operated at vacuum levels of about 25 to about 500 millitorr. After filling container 500 with the desired amount of material, filling nozzle 260 is replaced with filling plug 550 to close and seal filling port 540. In some embodiments, filling port 540 is hermitically sealed with filling plug 550. In some embodiments, filling plug 550 is welded to filling port 540. In some embodiments, an orbital welder 242 is used to weld filling plug 550 to filling port 540. In some embodiments, evacuation port 560 may be provided with evacuation plug 570. As previously described, evacuation plug 570 may be threadably coupled with evacuation port 560 in a first open configuration to allow air and/or other gas to pass through filter 590 and between evacuation plug 570 and evacuation port 560 and in a second closed configuration to hermitically seal and close evacuation port 560. In some embodiments, after filling is complete, evacuation port 560 is closed by evacuation plug 570. In some embodiments, evacuation port 560 is closed while evacuation nozzle 300 is coupled to evacuation port 560. With reference to FIG. 6B, container 600 is evacuated by coupling filling port 640 with an evacuation line or system (e.g., a vacuum source). Material is then added to interior volume 616 of container 600 via a filling nozzle 260 coupled to filling port 640. In some embodiments, the filling port 640 is configured to tightly fit around filling nozzle 260 to prevent material from exiting container 600 between the filling port 640 and filling nozzle 260. In some embodiments, container 600 is evacuated to a pressure of about 750 millitorr to about 1000 millitorr prior to filling. After filling container 600 with the desired amount of material, filling nozzle 260 is replaced with filling plug 650 to close and seal filling port 640 according to some embodiments. In some embodiments, container 600 is returned to the atmospheric pressure (e.g. the pressure of first cell 217) after filling. FIGS. 8-11 illustrate an exemplary filling system 299 for transferring hazardous waste material into a container 216 in accordance with various embodiments of the present invention. Filling system 299, in accordance with some embodiments of the present invention, is designed to prevent contamination of equipment and container exterior and elimination of secondary waste. The design features include, but are not limited to: container structure to allow container filling under vacuum; weight verification system and/or volume verification system; and filling nozzle structure. As illustrated, in FIGS. 8-10, in some embodiments, system 299 for transferring hazardous waste material into a sealable container 216 includes a filling nozzle 260, at least one hopper 214, a pneumatic cylinder 285, a seal 284, a vibrator 281, a lift mechanism 282, a damper 283, a first scale 277, a second scale 278 and a processor 280. The system of FIGS. 8-11 may be used with a container having a single port, such as container 600, or a container having two ports, such as container 500, as described above herein. FIG. 8 illustrates a filling nozzle 260 relative to an exemplary container 216 having a single port 291. FIG. 9 illustrates a filling nozzle 260 relative to an exemplary container 216 having two ports, a filling port 292 and an evacuation port 293. In some embodiments, filling port 292 and evacuation port 293 may have the configuration of filling port 540 and evacuation port 560 of container 500 illustrated in FIGS. 5A and 6A. In one embodiment, the evacuation port 293 includes a filter 350. In some embodiments, filter 350 prevents the escape of hazardous waste particles from the container. Exemplary filter materials are discussed above herein. In some embodiments, filter 350 has the configuration of filter 590 as described above herein. In some embodiments, the transfer of hazardous waste is performed to prevent overpressure of container 216. In some embodiments, container 216 is at least initially under negative pressure before transfer of hazardous waste begins. In other embodiments, container 216 is under negative pressure simultaneously with the transfer of hazardous waste. In yet other embodiments, container 216 is initially under negative pressure before the filling process begins and is intermittently placed under negative pressure with the transfer of hazardous waste. In another embodiment, filling port 292 of container 216 is configured to be sealed closed after decoupling valve body 261 from filling port 292. In some embodiments, container 216 is filled at about 25° C. to about 35° C. In other embodiments, container 216 is filled at a temperature up to 100° C. Referring to FIGS. 2 and 11, in one embodiment, additive from the additive feed hopper 210 is added to the feed blender 212. In one such embodiment, the amount of additive is metered using an additive feed screw (not shown). Feed blender 212 is actuated to mix the calcined material with the additive. In one embodiment, feed blender 212 is a mechanical paddle-type mixer with the motor drives external to the cell. Referring to FIG. 8, in one embodiment a rotary airlock or ball valve 298, located between the feed blender 212 and hopper 214, transfers the mixed calcined material to feed hopper 214. In another embodiment, a rotary air lock or ball valve 298 is positioned between feed hopper 214 and container 216 to control transfer of material therebetween. Referring to FIG. 7, in some embodiments, a fixed volume of the mixed calcined material is transferred from feed hopper 214 to container 216 which is located in first cell 217. In one embodiment, container 216 has two ports, a fill and an evacuation port, as described herein. In another embodiment, container 216 has a single port as described herein. Fill port 540, 640, attached to the top of container 216, is mated to a fill nozzle, discussed below herein, that is designed to eliminate spilling any of the hazardous material on the exterior of container 216. In one embodiment, fill nozzle 260 and fill port 540, 640 are configured to prevent contamination with waste material of the seal between a filling plug 550 and the interior of fill port 540, 640. In one embodiment, the amount of hazardous material transferred to a container is carefully controlled to ensure that container 216 is substantially full without overfilling container 216. In some embodiments, a weight verification system connected to hopper 214 and container 216 ensures that the proper amount of material is transferred. In some embodiments, equal volumes between hopper and container in combination with weight verification system connected to hopper 214 and container 216 ensure that the proper amount of material is transferred. In some embodiments, the weight verification system includes a processor 280 and a plurality of weigh scales 277. In some embodiments, a first scale 277 is coupled to the hopper 214 and configured to determine an initial hopper weight; a second scale 278 is coupled to the container 216 and configured to determine a container fill weight; and a processor 280 is coupled to the first scale 277 and the second scale 278 and configured to compare the initial hopper weight to the container fill weight. In some embodiments, initial hopper weight is the weight between flange 294 and flange 295 including hopper 214. In some embodiments, initial hopper weight means the weight of hazardous material within the hopper prior to filling container 216. In some embodiments, container fill weight means the weight of hazardous material in container 216 during the filling process and/or at the end of the filling process. In one embodiment, hopper 214 includes a volume substantially equal to a volume of container 216. In some embodiments, one or more vibrators 281 are provided to one or more components of filling system 299 to help ensure that all of the material is transferred from hopper 214 to container 216. In some embodiments, one or more vibrators 281 are configured to apply a vibrating force to one or more components of system 299 in order to assist in transferring the material to container 216. In some embodiments, vibrators 281 are configured to provide at least a force in a vertical direction. In some embodiments, vibrators 281 are configured to provide at least a force in a lateral direction. In one embodiment, at least one vibrator 281 is coupled to hopper 214, for example, to shake material from hopper 214 to container 216. In one embodiment, at least one vibrator 281 is coupled to a bottom of container 216. In one such embodiment, vibrator 281 coupled to bottom of container 216 is configured to provide vibration to container 216 in at least a vertical direction. In one embodiment, at least one vibrator 281 is coupled to a sidewall of the container 216. In one such embodiment, vibrator 281 coupled to the sidewall of container 216 is configured to provide vibration to container 216 in at least a lateral direction. The one or more vibrators 281, in some embodiments, are coupled a processor configured to control activation and/or operation (e.g., frequency) of vibrators 281. In some embodiments, processor 280 is coupled to the one or more vibrators 281. In some embodiments, one or more vibrators 281 are activated if container 216 is determined to be under-filled, for example, where the material to be transferred has been held up inside the system. In one embodiment, one or more vibrators 281 are activated if the container fill weight is less than the initial hopper weight. Referring to FIGS. 8 and 10, in one embodiment, filling nozzle 260 includes a valve body 261, a valve head 265 and a valve stem 267. Valve body 261 includes a distal end 262 and an outer surface 263, valve body 261 including a valve seat 264 proximate distal end 262, outer surface 263 proximate distal end 262 configured to sealingly and removeably couple valve body 261 to a filling port 272 of a container 216. In certain embodiments, valve body 261 includes a first branch section 270 configured to couple to hopper 214. In one embodiment, a second branch section 269 includes the distal end 262 of the filling nozzle 260 and has a proximal end 288. In one embodiment, the proximal end 288 is coupled to a drive mechanism 289 configured to move the valve stem 267. In one embodiment, valve head 265 includes a valve face 266 configured to form a seal with the valve seat 264 in a closed configuration. In one embodiment, valve head 265 is configured to allow valve body 261 and container 216 to be fluidly coupled with one another in an open configuration. In certain embodiments, valve head 265 extends distally from valve body 261 and into container 216 in the open configuration. Valve stem 267 extends co-axially with axis 276 from valve head 265 through at least a portion of valve body 261. In a further embodiment, valve stem 267 extends through proximal end 288 of second branch section 269, proximal end 288 including a seal 284 coupled to a portion of valve stem 267. In some embodiments, filling nozzle 260 is sealed with filling port 272 of container 216 to prevent spilling of the hazardous waste material from container 216. In one embodiment, filling nozzle 260 extends into filling port 272 to prevent waste material from interfering with the seal between a filling plug (e.g. filling plug 650) and filling port 272 after removing filling nozzle 260. In some embodiments, outer surface 263 of distal end 262 includes at least one seal 273 to form a seal with filling port 272. In another embodiment, at least one seal 273 includes at least one o-ring. In one embodiment, at least one seal 273 includes two o-ring seals. In some embodiments, outer surface 263 includes a second seal 275 to form a seal with filling port 272. In some embodiments, filling port 272 has the configuration of filling port 640 of container 600, and at least one of seals 273 and 275 engages with lower inner surface 648 to form a seal therewith. In some embodiments, at least one of seals 273 and 275 engages with lower inner surface 648 at a position between first end 642 and where filter 690 engages filling port 640 as shown in FIG. 6B. In some embodiments, at least one of seals 273 and 275 engages with stepwise inner surface 647 at a position between first end 642 and gasket 680. In one embodiment, filling nozzle 260 further includes a sensor 274 disposed in valve head 265. In one embodiment, sensor 274 is configured to determine a level of hazardous material in container 216. In one embodiment, sensor 274 extends distally from valve body 261. In another embodiment, sensor 274 is coupled to a wire 268 that extends through valve stem 267. In one embodiment, sensor 274 is coupled to a wire 268 that extends through valve stem 267. Suitable sensors may include contact type sensors including displacement transducer or force transducer. In such embodiments, a displacement transducer senses filling powder height. In such embodiments, a force transducer includes a stain gauge on thin membrane that is deflected by the filling powder front. Suitable sensors may also include non contact type sensors including sonar, ultrasonic, and microwave. In one embodiment, a drive mechanism operates valve stem 267. In one embodiment, drive mechanism 289 includes a pneumatic cylinder 285. In some embodiments, a lift mechanism 282 is configured to lift container 216 toward filling nozzle 262. In one embodiment, lift mechanism 282 includes at least one damper 283. In one embodiment, the system for transferring hazardous waste material into the sealable container further comprises a vacuum nozzle 271 configured to be in fluid communication with container 216. In one embodiment, vacuum nozzle 271 extends through distal end 262 of valve body 261. In another embodiment, vacuum nozzle 271 includes a filter 279 proximate the distal end 262 of valve body 261. In certain embodiments, the system in accordance with the present invention further comprises a vacuum nozzle 271 sealingly and removeably couplable with the exhaust port 286, vacuum nozzle 271 being in sealed fluid communication with the valve body 261 in a filling configuration. In one embodiment, first cell 217 does not exchange air with subsequent cells while at least container 216 is being filled by the filling system 299. Referring to FIG. 7, in one embodiment, first cell 217 includes an off-gas sub-system 206 coupled to filling system 299 wherein off-gas sub-system 206 has a vacuum nozzle configured to couple to container 216. Referring to FIG. 12, in a further embodiment, first cell 217 is coupled to the second, subsequent cell 218 with one or more sealable doors 240. In one embodiment, the second, subsequent cell 218 is a bake-out and vacuum sealing cell. In one embodiment, first cell 217 is coupled to second cell 218 via an air interlock 241. In one embodiment, air interlock 241 is configured to allow container 216 to be transferred from first cell 217 to second cell 218. II. Second Cell Exemplary embodiments of second cell 218 and certain components thereof are illustrated in FIGS. 2, 3, 4, 12, 13, 14 and 16. In one embodiment, second cell 218 is a bake-out and vacuum sealing cell which allows for heating and evacuating container 216 followed by sealing of container 216. In one embodiment, first cell 217 is held at a first pressure P1 and second cell 218 is held at a second pressure P1, where the first pressure P1 is less than the second pressure P2. First cell 217 and second cell 218 are interconnected via the sealable door 240 according to some embodiments. In one embodiment, second cell 218 includes a baking and sealing station 243. In certain embodiments, second cell 218 further includes a welding station. Referring to FIG. 2, in one embodiment, second cell 218 includes a bake-out furnace 290, an off-gas system 206 having a vacuum nozzle configured to couple to the container 216. In some embodiments, as shown in FIG. 16, second cell 218 further includes an orbital welder 242 configured to apply a weld to container 216. In one embodiment, referring to FIGS. 3 and 12, second cell 218 includes an interlock 241, interlock 241 coupling first cell 217 to second cell 218 and configured to allow container 216 to be transferred from first cell 217 to second cell 218 while maintaining at least one seal between the first cell 217 and second cell 218. In one embodiment, interlock 241 includes decontamination equipment. In another embodiment, first cell 217 and interlock 241 may be communicatively interconnected via sealable door 240, allowing container 216 to be transferred from first cell 217 to interlock 241. In a further embodiment, first cell 217 and second cell 218 include a roller conveyer 246 configured to allow containers 216 to be loaded thereon and transported within and/or between each cell. Referring again to FIG. 2, in some embodiments, second cell 218 includes a furnace 290 configured for heating container 216 in a bake-out process. In some embodiments, the bake-out process includes heating container 216 in furnace 290 to remove excess water and/or other materials, for example, at a temperature of about 400° C. to about 500° C. for several hours. In some embodiments, a vacuum is established on container 216 and any off-gas is removed from container 216 during the bake-out process. The off-gas may include air from container 216 and/or other gas released from the waste material during the bake-out process. In some embodiments, the off-gas removed from container 216 is routed through line 206, which may lead out of second cell 218 and may be connected to a further ventilation system. Line 206, in some embodiments, includes one or more filters 204 to capture particulates entrained in the off-gas. Filters 204 may include HEPA filters according to some embodiments. In further embodiments, line 206 includes one or more traps 219 for removing materials such as mercury that may not be desirable to vent. For example, trap 219 in one embodiment may include a sulfur impregnated carbon bed trap configured to trap mercury contained in the off-gas from container 216. In further embodiments, a vacuum is established in container 216 during the bake-out process and container 216 may then be sealed to maintain the vacuum. Evacuation of the air and/or other gas from container 216, in some embodiments, is achieved by coupling container 216 with an evacuation system. FIG. 13 illustrates an exemplary evacuation system that can be used in accordance with embodiments of the invention shown coupled to filling plug 640 of container 600 as described above herein. It should be understood that the evacuation system depicted in FIG. 13, in other embodiments, may be coupled to containers having other configurations. For example, the evacuation system may be coupled to evacuation port 560 of container 500 shown in FIGS. 5A and 6A. Referring again to FIG. 13, the evacuation system shown includes an evacuation nozzle 300, which may be coupled with an evacuation line or other a vacuum source. In some embodiments, evacuation nozzle 300 is further coupled to a vacuum transducer 301 configured to measure the vacuum level in container 600. In some embodiments, evacuation nozzle 300 is coupled to a valve 302. In some embodiments, valve 302 is configured to isolate container 600 from the vacuum source, which in turn allows for the detection of leaks in container 600 or detection of gas being evolved from interior volume 616. The detection can be accomplished, for example, by measuring pressure change (e.g. using vacuum transducer 301) as a function of time. An increase in pressure (or loss of vacuum) in container 600 over time may indicate, for example, a possible leak or gas generation from interior volume 616. In some embodiments, evacuation nozzle 300 further includes a filter configured to prevent passage of particulate matter there through. As illustrated, evacuation nozzle 300 in some embodiments is coupled to filling plug 650 and/or filling port 640 of container 600. In some embodiments, evacuation nozzle 300 fits around filling plug 650 and filling port 640. In some embodiments, evacuation nozzle 300 is configured to at least partially surround filling plug 650 and filling port 640 when filling plug 650 is coupled with filling port 640. In some embodiments, evacuation nozzle 300 forms a circumferential seal with filling port 640 when coupled thereto. In some embodiments, evacuation nozzle 300 seats against flange 634. In some embodiments, evacuation nozzle 300 includes a gasket that engages with an external surface of filling port 640 to form a hermitic seal therewith when evacuation nozzle is coupled with filling port 640. In some embodiments, filling plug 650 may be threadably coupled with filling port 640 in a first open configuration to allow air and/or other gas to pass through filter 690 and between filling plug 650 and filling port 640 and in a second closed configuration to hermitically seal and close filling port 640. In some embodiments, air and/or other gas is allowed to pass between filling plug 650 and filling port 640 and through seam 646. In some embodiments, evacuation nozzle 300 is configured to withdraw air and/or other gas from interior volume 616 of container 600 when filling plug 650 and filling port 640 are in the first open configuration. In some embodiments, after air and/or other gas is withdrawn from interior volume 616, a vacuum is created within interior volume 616 and filling plug 650 is used to hermetically seal filling port 640 in the closed configuration so as to maintain the vacuum. In some embodiments evacuation nozzle 300 is fitted with a torque 304 having a stem 303. In some embodiments, stem 303 has a proximal end and a distal end, said distal end being configure to mate with a recess in filling plug 650, and the proximal end being coupled to a handle. In some embodiments, the handle of torque 304 is manipulated to threadably tighten filling plug 650 to filling port 640, thereby forming a tight seal between the filing plug 650 and filling port 640. In some embodiments, torque 304 is manipulated with a drive shaft. In some embodiments, when the bake-out process is completed, the vacuum is maintained on container 600 through the evacuation system. In some embodiments, when the vacuum reaches a set point, the vacuum is verified, for example using vacuum transducer 301 as described above herein, and filling port 640 is closed (e.g., hermetically scaled) by filling plug 650 and the evacuation system is removed. In some embodiments, filling plug 650 is then welded to filling port 640. In some embodiments, filling plug 650 is welded to filling port 640 by an orbital welder 242, which may be positioned in a welding station in second cell 218. An embodiment of an orbital welding station is illustrated in FIG. 14, which shows orbital welder 242 configured to weld filling plug 650 onto filling port 640 of container 600 at seam 646. In some embodiments, orbital welder 242 is remotely operated. In some embodiments, welds applied by orbital welder 242 are visually inspected. While the foregoing description of the evacuation system and orbital welder 242 makes reference to container 600, it should be understood that these elements may be similarly used on other configurations for container 216. For example, in other embodiments, these elements may be similarly used to evacuate, seal, and weld container 500 at evacuation port 560. In these embodiments, where container 500 also includes a separate filling port 540, filling port 540 may be similarly closed (e.g., by filling plug 550) and welded sealed by orbital welder 242 prior to the bake-out process. With reference again to FIG. 2, following the bake-out process, container 216, in some embodiments, is placed in containment 231 after being removed from furnace 290. In some embodiments, containment 231 provides for further contamination control in case of leakage or rupture of container 216. In some embodiments, containment 231 may be pre-staged on roller conveyor 246 for subsequent transport to third cell 232. III. Third Cell Exemplary embodiments of third cell 232 are illustrated in FIGS. 3, 4 and 15. In one embodiment, third cell 232 is a HIP process cell which allows for hot isostatic pressing of container 216. In one embodiment, third cell 232 includes a hot isostatic pressing station. In one embodiment, first cell 217 is held at a first pressure P1, second cell 218 is held at a second pressure P2 and third cell 232 is held at a third pressure P3. In one embodiment, first pressure P1 is less than second pressure P2 which is less than third pressure P3. Referring to FIGS. 3, 4 and 16, in one embodiment, modular system 400 in accordance with the present invention includes third cell 232, wherein third cell 232 is isolated from first cell 217 and second cell 218, and wherein second cell 218 and third cell 232 are configured to allow container 216 to be transferred from second cell 218 to third cell 232. In some embodiments, container 216 is transferred from second cell 218 to third cell 232 in containment 231. In some embodiments, container 216 is subjected to hot isostatic pressing in third cell 232. In some embodiments, container 216 is subjected to hot isostatic pressing while in containment 231. In some embodiments, third cell 232 includes a hot isostatic pressing station 249. In one embodiment, hot isostatic pressing station 249 includes a HIP support frame 245, a hot isostatic pressing vessel 251 secured to support frame 245, and a pedestal mounted pick and place machine (robotic arm) 252 secured to the HIP support frame 245, robotic arm 252 configured to manipulate within hot isostatic pressing station 249. In one embodiment, robotic arm 252 is configured to lift and transfer container 216 from roller conveyer 246 into isostatic process vessel 251. In a further embodiment, third cell 232 includes a sealable door 240. In one embodiment, sealable door 240 couples third 232 and second cell 218 and is configured to allow container 216 to be transferred from second cell 218 to third cell 232. In a further embodiment, second cell 218 and third cell 232 each include a roller conveyer 246 configured to allow container 216 to be loaded thereon and transported within and/or between second 218 and third cell 232. Hot isostatic pressing, according to some embodiments, includes positioning containment 231 holding container 216 in a hot isostatic pressing vessel 251. In some embodiments, container 231 is positioned by robotic arms 252. In some embodiments, the hot isostatic pressing vessel 251 is provided with an argon atmosphere (e.g., from argon source 236 via argon line 202) which can be heated and pressurized. In some embodiments, for example, the hot isostatic pressing process is performed by heating containment 231 holding container 216 to about 1000° C. to about 1250° C. in the hot isostatic pressing vessel 251 for about 2 hours to about 6 hours. In some embodiments, the pressure inside the hot isostatic pressing vessel 251 is controlled to be about 4300 psi to about 15000 psi during the hot isostatic pressing process. In some embodiments, compressors (e.g., 234) protected by in-line filtration are used to control the argon atmosphere of the hot isostatic pressing vessel 251. In some embodiments, the argon used during the hot isostatic pressing process is filtered and stored in a manner that conserves both argon and pressure. Referring to FIG. 2, in some embodiments, the argon is recycled to argon source 236 via pump 238. The recycled argon, in some embodiments, passes through filter 233. With reference to container embodiments illustrated in FIGS. 5A, 5B, 6A and 6B, the material of filter 590 and/or filter 690 is chosen so that upon heating during hot isostatic pressing the filter densifies into a solid and non-porous material forming a weld with container, container evacuation port and/or container filling port. In some embodiments, the material of filter 590 and/or 690 is chosen wherein at a filling temperature filter 590 and/or 690 is porous to air and/or gas but densifies into a non-porous material during hot isostatic pressing. In some embodiments, after hot isostatic pressing is complete, containment 231 and container 216 is allowed to cool within the hot isostatic pressing vessel 251 to a temperature sufficient for removal (e.g., about 600 EC). In some embodiments, hot isostatic isostatic pressing vessel 251 includes a cooling jacket having cooling fluid (e.g., water) flowing therethrough. In some embodiments, the cooling jacket is supplied with cooling water at a rate of about 80 gpm to about 100 gpm. In some embodiments, containment 231 holding container 216 is removed from hot isostatic pressing vessel 251 and transferred to a cooling cabinet for cooling. In some embodiments, the cooling cabinet is supplied with a cooling fluid (e.g., water). In some embodiments, the cooling cabinet is supplied with cooling water at a rate of about 10 gpm. In some embodiments, containment 231 and container 216 are allowed to cool in the cooling cabinet for about 12 hours. Following cooling in the cooling cabinet, containment 231 holding container 216 is placed on a roller conveyor 246 for transport to fourth cell 230. IV. Fourth Cell Exemplary embodiments of fourth cell 230 are illustrated in FIGS. 3, 4 and 17. In one embodiment, fourth cell 230 is a cooling cell which allows for further cooling of container 216 after the hot isostatic pressing (HIP) process. In some embodiments, container 216 is packaged in fourth cell 230 for subsequent storage. In a further embodiment, referring to FIGS. 3, 4 and 17, modular system 400 in accordance with the present invention includes fourth cell 230, which may be a cooling cell. In one embodiment, fourth cell 230 is isolated from first 217, second cell 218 and third cell 220. In one embodiment, third 232 and fourth cell 230 are configured to allow container 216 to be transferred from third cell 232 to fourth cell 230. In one embodiment, first cell 217 is held at a first pressure P1, bake-out and second cell 218 is held at a second pressure P2, third cell 232 is held at a third pressure P3 and fourth cell 230 is held at a fourth pressure P4. In one embodiment, first pressure P1 is less than second pressure P2 which is less than third pressure P3 which is less than fourth pressure P4. In a further embodiment, fourth cell 230 includes a moveable shielded isolation door 240. In one embodiment, sealable door 240 is coupled to fourth cell 230 and third cell 232 and is configured to allow container 216 to be transferred from third cell 232 to fourth cell 230. In a further embodiment, each of third cell 232 and fourth cell 230 includes a roller conveyer 246 configured to allow container 216 to be loaded thereon and transported within and/or between third cell 232 and fourth cell 230. In yet another embodiment, fourth cell 230 includes an orbital welder 255. In some embodiments, after transport to fourth cell 230, containment 231 is opened and container 216 checked for evidence of container failure (e.g., deformation, expansion, breakage, etc.). In the event of failure of container 216, according to some embodiments, container 216 and containment 231 are moved to a decontamination chamber within fourth cell 230, decontaminated and returned to second cell 218 for possible recovery. If there is no evidence of failure of container 216, container 216 is removed from containment 231 and transferred to a cooling and packing station 250 in fourth cell 230 according to some embodiments. In a further embodiment, cooling and packing station 250 includes a set of at least one or more cooling stations. In one embodiment, at least one or more cooling stations 253 configured to receive and hold processed container 216 for final cooling. In some embodiments, container 216 is passively cooled in cooling station 253. In some embodiments, container 216 is actively cooled in cooling station 253. In some embodiments, after final cooling, container 216 is packaged in fourth cell 230 for transport and storage. In some embodiments, one or more cooled containers 216 are placed in a canister. In some embodiments, the canister containing one or more containers 216 is then welded shut, for example, using an orbital welder 255. In some embodiments, the canister can then be transported for storage. Referring to FIG. 2, any one of the cells of the modular system 400 may include any suitable number of vacuum lines, including no vacuum line at all. As illustrated in FIG. 2, first cell 217, second cell 218, third cell 232 and fourth cell 230 may each include a set of one or more vacuum lines. Moreover, as illustrated in FIGS. 2, 3, 4, 5 and 10, first cell 217, second cell 218, third cell 232 and fourth cell 230 may each be equipped with a set of at least one or more remotely operated overhead bridge cranes 239. In one embodiment, in addition to their material handling roles, each of these remotely operated overhead bridge cranes 239 are designed to be available for use in accomplishing either remote or manned maintenance of the equipment within the various cells. In another embodiment, each of the in-cell cranes may be configured to be capable of being remotely removed from the cell via a larger crane provided for maintenance purposes. In some embodiments, secondary waste produced by modular system 400 of the present invention may be collected and transferred to containers 216 for processing in accordance with steps of process flow 200. In some embodiments, for example, secondary waste is added to feed blender 212, mixed with calcined materials and/or additives, and transferred to a container 216 via a filling nozzle for subsequent hot isostatic pressing. Secondary waste, as used herein according to certain embodiments, refers to hazardous waste materials which are removed from container 216 and/or materials which are contaminated with hazardous waste materials during steps of the present invention. In some embodiments, the secondary waste is converted to a form suitable for transferring via the filling nozzle before introducing the secondary waste into a container 216. In some embodiments, secondary waste includes materials filtered or trapped from the off gases evacuated from container 216. In one such embodiment, secondary waste includes mercury captured from off gas evacuated from a container 216 during processing, for example, by one or more traps 219 as described above herein. The mercury may be transformed into an amalgam by mixing the mercury with one or more other metals and transferred to another container 216 for further processing according to one example of this embodiment. In some embodiments, secondary waste further includes system components which may have been contaminated by or in direct contact with hazardous waste material. The contaminated components may be combusted, crushed, pulverized, and/or treated in another manner prior to feeding to a container 216. In one such example, secondary waste includes a used cell or exhaust line filter (e.g., filter 204), which may contain hazardous waste materials. In some embodiments, the used filter may be combusted and the resulting ashes are fed to a container 216 for further processing. In some embodiments, at least 50% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 60% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 70% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 80% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 90% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 95% by weight of the secondary waste produced by modular system 400 is collected for processing. In some embodiments, at least 99% by weight of the secondary waste produced by modular system 400 is collected for processing. Method of Processing Hazardous Waste Using a Modular System In some embodiments, the systems, method and components described herein provide for a method of storing hazardous waste material comprising a plurality of steps and performed in a modular system. In some embodiments, one or more of the steps described herein can be performed in an automated manner. In a first cell, hazardous waste material is added to a container via a filling nozzle coupled to a filling port of the container. Various embodiments of such filling nozzle are described herein. The container is configured to sealingly contain the hazardous waste material. In one embodiment, the container further includes an evacuation port. In one embodiment, the container is evacuated prior to adding the hazardous waste material by connecting a filling nozzle having a connector coupled to a vacuum system to thereby place the container under a negative pressure. In another embodiment, the container is evacuated during adding of the hazardous waste material via an evacuation nozzle coupled to an evacuation port of the container to thereby maintain the container under a negative pressure during the adding step. In some embodiments, the amount of hazardous waste material added to the container is verified by measuring the weight of the container after filling. Various embodiments of weight verification systems are described herein. In some embodiments, the amount of hazardous waste material added to the container is verified by comparing the weight (or change in weight) of the container after filling to the weight of hazardous waste material prior to filling. In one embodiment, a filling plug is inserted into the filling port to form a plugged container after the hazardous waste material is added to the container to close the filling port. In another embodiment, a filling plug is inserted into the filling port and an evacuation plug is inserted into the evacuation port prior to sealing the filling port to form a plugged container. The plugged container is then transferred from the first cell to the second cell via the moveable shielded isolation door. In one embodiment, the plugged cell is transferred from the first cell to the second cell via the moveable shielded isolation door and then into an interlock area containing contamination equipment. In the second cell, the plugged container is connected to an evacuation nozzle coupled to an evacuation system and the container is heated. In some embodiments, the container is heated in a bake-out furnace to remove excess water and/or other materials. In some embodiments, off-gas including air and/or other gas is removed from container during heating, for example, through the use of the evacuation nozzle. In one embodiment, the evacuation nozzle is coupled to the evacuation port of the container. In such an embodiment, the evacuation plug is closed while the evacuation nozzle is couple to the evacuation nozzle. In one such embodiment, the evacuation port includes an evacuation plug which is threadably coupled to the evacuation port. The evacuation plug allows air and/or gas to pass through a filter, located in the evacuation port, and between the evacuation plug and the evacuation port in a heating configuration. Prior to heating the container, the evacuation port is at least partially opened. The container is then heated. Following the heating step, the evacuation port is placed in a closed configuration and is sealed in one embodiment. In one such embodiment, the vacuum on the container is maintained for a period of time following the heating step prior to sealing. Optionally, the maintenance of the vacuum in the container is verified. In one such embodiment, the sealing step is performed by welding an evacuation plug to the evacuation port to seal the evacuation port. In such an embodiment, the welding is performed using an orbital welder. In another embodiment, the evacuation nozzle is coupled to the filling port of the container. In such an embodiment, the filling plug is closed while the evacuation nozzle is couple to the evacuation nozzle. In one such embodiment, the filling port includes a filling plug which is threadably coupled to the filling port. The filling plug allows air and/or gas to pass through a filter, located in the filling plug, and between the filling plug and the filling port in a heating configuration. Prior to heating the container, the filling port is at least partially opened. The evacuated container is then heated. Following the heating step, the filling port is closed in a closed configuration and is sealed. In one such embodiment, the vacuum on the container is maintained for a period of time following the heating step prior to sealing. Optionally, the maintenance of the vacuum in the container is verified. In one such embodiment, the sealing step is performed by welding the filling plug to the filling port to seal the filling port. In such an embodiment, the welding is performed using an orbital welder. Following the sealing step, the sealed container is transferred from the second cell to the third cell via a second moveable shielded isolation door. In some embodiments, the sealed container is transferred from the second cell to the third cell inside a containment. The sealed container is then subjected to hot isostatic pressing. In some embodiments, the sealed container is subjected to hot isostatic pressing while inside the containment. In some embodiments, hot isostatic pressing includes subjecting the sealed container to a high temperature, high pressure argon atmosphere. In some embodiments, the sealed container is initially cooled in a cooling cabinet after hot isostatic pressing. Following the hot isostatic pressing, the container is transferred from the third cell to the fourth cell via a third moveable shielded isolation door. In the fourth cell, according to some embodiments, the container undergoes final cooling. In further embodiments, the container is packaged in a canister for transport and storage. It will be appreciated by those skilled in the art that changes could be made to the exemplary embodiments shown and described above without departing from the broad inventive concept thereof. It is understood, therefore, that this invention is not limited to the exemplary embodiments shown and described, but it is intended to cover modifications within the spirit and scope of the present invention as defined by the claims. For example, specific features of the exemplary embodiments may or may not be part of the claimed invention and features of the disclosed embodiments may be combined. Unless specifically set forth herein, the terms “a”, “an” and “the” are not limited to one element but instead should be read as meaning “at least one”. It is to be understood that at least some of the figures and descriptions of the invention have been simplified to focus on elements that are relevant for a clear understanding of the invention, while eliminating, for purposes of clarity, other elements that those of ordinary skill in the art will appreciate may also comprise a portion of the invention. However, because such elements are well known in the art, and because they do not necessarily facilitate a better understanding of the invention, a description of such elements is not provided herein. Further, to the extent that the method does not rely on the particular order of steps set forth herein, the particular order of the steps should not be construed as limitation on the claims. The claims directed to the method of the present invention should not be limited to the performance of their steps in the order written, and one skilled in the art can readily appreciate that the steps may be varied and still remain within the spirit and scope of the present invention.
052176827	summary	FIELD OF THE INVENTION The invention relates to a nuclear reactor having a primary cooling circuit with a primary heat sink for removing heat generated in the reactor core during normal operation and an additional means for dissipating decay heat which is produced in the core of the reactor after the reactor has been shutdown. The means for removal of decay heat from the reactor core is one that is automatically activated as soon as the primary heat sink becomes unavailable as a heat sink. BACKGROUND OF THE INVENTION An emergency or normal shutdown of any nuclear power reactor system requires a system to remove decay heat which is produced in the reactor core after shutdown and thereby prevent damage to the reactor and associated systems. The system for removal of decay heat from the reactor core must be one which can remain operational for a lengthy period of time. In power generating nuclear reactor systems, such as a CANDU reactor, a steam generator is present in the primary cooling circuit which acts as a heat sink during normal operation. However, if the steam generator becomes unavailable as a heat sink and the reactor is shutdown, another means must be present to dissipate decay heat which continues to be produced in the reactor core. U.S. Pat. No. 4,689,194 shows one type of decay heat removal system for a gas cooled reactor. Circulating blowers cause a cooling gas (helium) to flow up through the reactor core and a central hot gas line downward through steam generators and decay heat exchangers back to the blowers. If the circulating blowers are not operational, decay heat from the core is removed by a natural convection flow of the cooling gas in the same direction as the flow during normal operation of the reactor. The decay heat exchangers are each connected with an external re-cooling heat exchanger at a geodetically higher location by means of two legs which form a water circulation loop. If the steam generators are no longer available for the removal of heat from the primary cooling path, they are traversed by hot gas which subsequently passes through the decay heat exchanges. This causes a rise in temperature at the inlet of the decay heat exchangers which leads to evaporation taking place in the water circulation loops whereby natural convection flow in these loops is enhanced and a sufficient amount of heat is removed from the primary cooling path through the decay heat exchangers. U.S. Pat. No. 4,830,815 shows other types of shutdown cooling systems for pressurized boiling water reactors which include a separate shutdown cooling heat exchanger in a cooling pond. Valves are including in the piping to those heat exchangers and additional pumps are used to pump coolant from the reactor to the shutdown cooling heat exchanger and back to the reactor. These types of systems require valves to be opened and pumps activated before they are operational. SUMMARY OF THE INVENTION It is an object of the invention to provide an improved decay heat removal system for a nuclear reactor which is automatically activated when the main heat removal component for the reactor system becomes inoperational and which causes very little heat to be lost from the reactor system during normal operation of the reactor. A nuclear reactor system, according to a preferred embodiment of the invention, consists of a reactor core and a main heat transport path containing a main heat removal component, at least one main coolant pump and a first coolant wherein, during normal operation, the first coolant is pumped by the main coolant pump through the reactor core to the main heat removal component and back to the reactor core to transport heat generated in the reactor core to the main heat removal components; the main heat transport path including a heat exchanger located in the path after the main heat removal component's outlet with the secondary side of the heat exchanger being included in a decay heat removal loop having a vapor separator connected to an outlet of the heat exchanger, the vapour separator's outlet being connected to an inlet of a further heat exchanger located in a reservoir of coolant which forms a heat sink, the further heat exchanger's outlet being connected to the heat exchanger's inlet; and wherein the further heat exchanger is located at a higher elevation than the heat exchanger whereby a natural convection flow can occur in the decay heat removal path, the vapor/liquid interface in the vapour separator being at a higher elevation than the heat sink which prevents any significant natural convection flow until boiling of the second liquid coolant occurs in the heat exchanger.
047587261	abstract	A collimator exchanging system for exchanging collimators between an active holding device (14) and a storage device (8) includes a cart (5) which is provided with a catching/locking device (26) for the collimator. The device (26) is constructed so that the collimator can be exchanged only in the correct orientation of the cart relative to the active device or the storage device. An unlocking pin (34) of the cart enables the exchange with the exchanging system otherwise being blocked.
056688474	description	DETAILED DESCRIPTION The invention is described below with primary reference to a system for delivering X-ray radiation to a field of a patient, and for delimiting the field using at least one movable plate in the beam path from a radiation source. This is by way of example only. The invention may be used to adjust the delivery of any type of energy, for example, electrons (instead of X-rays), to any type of object (not just a human patient), provided the amount of energy delivered to the field can be sensed or estimated. FIG. 1 shows a radiation treatment device 2 of common design, in which plates 4 and a control unit in a housing 9 and a treatment unit 100 constructed in accordance with the principles of the invention are used. The radiation treatment device 2 comprises a gantry 6 which can be swiveled around a horizontal axis of rotation 8 in the course of a therapeutic treatment. Plates 4 are fastened to a projection of gantry 6. To generate the high-powered radiation required for the therapy, a linear accelerator is located in gantry 6. The axis of the radiation bundle emitted from the linear accelerator and gantry 6 is designated by 10. Electron, photon, or any other detectable radiation can be used for the therapy. During the treatment the radiation beam is trained on a zone 12 of an object 13, for example, a patient who is to be treated, and who lies at the isocenter of the gantry rotation. The rotational axis 8 of gantry 6, the rotational axis 14 of a treatment table 16, and the beam axis 10 all preferably intersect in the isocenter. The construction of such a radiation treatment device is described in general in a brochure "Digital Systems for Radiation Oncology", Siemens Medical Laboratories, Inc. A91004-M2630-B358-01-4A00, September 1991. The area of the patient that is irradiated is known as the field. As is well known, the plates 4 are substantially impervious to the emitted radiation. They are mounted between the radiation source and the patient in order to delimit the field. Areas of the body, for example, healthy tissue, are therefore subjected to as little radiation as possible, and preferably to none at all. In the preferred embodiment of the invention, at least one of the plates is movable so that the distribution of radiation over the field need not be uniform (one region can be given a higher dose than another); furthermore the gantry can preferably be rotated so as to allow different beam angles and radiation distributions without having to move the patient around. Neither or these features is necessary according to the invention: the invention may also be used with fixed-field devices (no movable plates), with constant radiation delivery rates, and with fixed-angle beams (no rotatable gantry). Radiation treatment device 2 also includes a central treatment processing or control unit 100, which is usually located apart from radiation treatment device 2. The radiation treatment device 2 is normally located in a different room to protect the therapist from radiation. Treatment unit 100 includes output devices, such as at least one visual display unit or monitor 70, and an input device such as a keyboard 19, although data can be input also through data carriers, such as data storage devices, or an verification and recording or automatic set-up system 102, which is described below. The treatment processing unit 100 is typically operated by the therapist who administers actual delivery of a radiation treatment as prescribed by an oncologist. By utilizing the keyboard 19, or other input device, the therapist enters into a control unit 76 of the treatment unit 100 the data that defines the radiation to be delivered to the patient, for example, according to the prescription of the oncologist. The program can also be input via another input device like a data storage device, through data transmission, or using the automatic set-up system 102. On the screen of a monitor 70 various data can be displayed before and during the treatment. FIG. 2 shows portions of an illustrative radiation treatment device 2 and portions of treatment unit 100 in more detail. An electron beam 1 is generated in an electron accelerator 20. Accelerator 20 comprises an electron gun 21, a wave guide 22 and an evacuated envelope or guide magnet 23. A trigger system 3 generates injector trigger signals and supplies them to injector 5. Based on these injector trigger signals, injector 5 generates injector pulses which are fed to electron gun 21 in accelerator 20 for generating electron beam 1. Electron beam 1 is accelerated and guided by wave guide 22. For this purpose, a high frequency (HF) source (not shown) is provided which supplies radio frequency (RF) signals for the generation of an electromagnetic field supplied to wave guide 22. The electrons injected by injector 5 and emitted by electron gun 21 are accelerated by this electromagnetic field in wave guide 22 and exit at the end opposite to electron gun 21 as electron beam 1. Electron beam 1 then enters a guide magnet 23, and from there is guided through a window 7 along axis 10. After passing through a first scattering foil 15, the beam goes through a passageway 51 of a shield block 50 and encounters a second scattering foil 17. Next, it is sent through a measuring chamber 60, in which the dose is ascertained. If the scattering foils are replaced by a target, the radiation beam is an X-ray beam. Finally, the aperture plate arrangement 4 is provided in the path of radiation beam 1, by which the irradiated field of the subject of investigation is determined. Aperture plate arrangement 4 includes a pair of plates 41 and 42. As is described above, this is just one example of a beam-shielding arrangement that can be used in the invention. The invention will work with others also as long as there is an aperture plate arrangement that defines an irradiated field. Plate arrangement 4 comprises a pair of aperture plates 41 and 42 and an additional pair of aperture plates (not shown) arranged perpendicular to plates 41 and 42. In order to change the size of the irradiated field the aperture plate can be moved with respect to axis 10 by a drive unit 43 which is indicated in FIG. 2 only with respect to plate 41. Drive unit 43 comprises an electric motor which is coupled to plates 41 and 42 and which is controlled by a motor controller 40. Position sensors 44 and 45 are also coupled to plates 41 and 42, respectively, for sensing their positions. This is just one example of such a system. The invention will work with other systems also, as long as them is a beam-shielding arrangement that defines an irradiated field and as long as sensors are provided to indicate the field size. Motor controller 40 is coupled to a dose control unit 61 which includes a dosimetry controller and which is coupled to a central processing unit 18 for providing set values for the radiation beam for achieving given isodose curves. The output of the radiation beam is measured by a measuring chamber 60. In response to the deviation between the set values and the actual values, dose control unit 61 supplies signals to trigger system 3, which changes in a known manner the pulse repetition frequency so that the deviation between the set values and the actual values of the radiation beam output is minimized. In such a radiation treatment device the dose absorbed by object 13 is dependent on the type of filter used for shaping the radiation beam. If a wedge filter built from absorbing material is inserted in the trajectory of the radiation beam, then the preset dose has to be increased according to the wedge factor in order to supply the desired dose to object 13. FIG. 3 shows isodose curves for a conventional wedge filter 46 in the path of the radiation beam emitted from radiation source 17 to object 13. The radiation beam is shaped on the one hand by the wedge filter and on the other hand by aperture plates 41 and 42. Due to the absorbing material of wedge filter 46, the isodose curve in the center 10 of the beam on object 13 has a maximum value of Dmax, which is the maximum value at a spot in center 10 of the beam on the surface of object 13 without wedge filter 46. In the illustrated example, Dmax is roughly 72%. The wedge factor defined as the ratio of doses with and without wedge filter 46 is thus, in this case 0.72. FIG. 4 shows isodose curves in a radiation treatment device according to the invention. Instead of including a wedge-shaped absorber in the path of the radiation beam, the filter function is performed by changing the radiation output of the radiation beam and by simultaneously moving at least one plate 41 and keeping the other plates of plate arrangement 4 stationary. A radiation treatment device having such a filter arrangement is disclosed in U.S. Pat. No. 5,148,032. Although this U.S. Patent describes the possibility of moving any plate, in the following, the invention is described in connection with only one plate being moved and the other plates being kept stationary. This is for the sake of simplicity only. The invention may be used for multiple moving plates as well. When in FIG. 4 plate 41 moves in the direction of arrow A toward plate 42 and at the same time the radiation output is changed according to a desired wedge angle, by adjusting the speed of plate 41 and/or correspondingly, the value of the isodose curve through the center of the beam on the surface of object 13 equals Dmax=100%. Thus, by using a wedge function instead of a wedge-shaped absorber an efficiency factor of "1" or 100% can be established; in other words, the dose delivered at that point is 100% of the prescribed dose, although the same relative isodose profiles are maintained. That means that the therapist does not have to take into account a wedge factor when defining the treatment, although wedge shaped isodose curves are established. FIG. 2 shows those portions of treatment unit 100 which are necessary to carry out the invention. Treatment unit 100 comprises central processing unit 18 and which is programmed by the therapist according to the instructions of the oncologist so that the radiation treatment device carries out the prescribed radiation treatment. Through keyboard 19 the prescribed delivery of the radiation treatment is input. Central processing unit 18 is connected, on the one hand, with the input device, such as the keyboard 19, for inputting the prescribed delivery of the radiation treatment and, on the other hand, with a dose control unit 61 that generates the desired values of radiation for the controlling trigger system 3. Trigger system 3 then adapts the pulse repetition frequency or other parameters in a corresponding, conventional manner. The ability to change the radiation output is generally known and it is particularly advantageous to use a digital dosimetry system became then it can easily be controlled by the digital output of central processing unit 18. Central processing unit 18 includes control unit 76 which controls the execution of the program and which supplies position signals P for controlling the opening of plate arrangement 4 and nominal dose signals D (corresponding to the plate position that would be demanded using prior art methods, that is, without regard to output factor compensation) for adjusting the radiation output at the output of radiation source 17. A memory 77 is also provided in or is connected to the central processing unit 18 for supplying correction signals C, which the processing unit uses to adjust the radiation output dependent on the position signals P supplied from position sensors 44 and 45 in order to achieve the predetermined constant output factor. The preferred arrangement of the memory unit is that, for each plate position (field size), it has stored a corresponding wedge correction signal C. The memory thus stores a table of wedge correction factors. If more than one set of movable plates is included in the system, then the table will be correspondingly multi-dimensional, and arranged using any known data structure, so that a wedge correction factor is available for any combination of plate positions. Control unit 76 and memory 77 apply the nominal dose and wedge correction signals D and C, respectively, to a combination circuit 78, which combines the values to generate set signals S. The set signals S are in turn applied to the dose control unit 61, which sets the radiation output. The combination circuit 78 will depend on the form in which the wedge correction signals are generated and stored. Assume that the wedge correction signals C are stored as additive offsets to the set radiation output. In this case, the combination circuit will be an adder which adds the wedge correction signals C to nominal dose signals D. This is the preferred embodiment, since it is simplest. If, however, the wedge correction factors are multipliers, for example, an increase in radiation output from a sensed value of 72% would require a multiplicative correction signal of about 139%. Instead of storing actual values of the wedge correction signals C, it is also possible to store the parameters of a wedge correction function for the various field sizes. The processing unit would then evaluate the wedge correction function for each current field size using the parameters stored in the memory, and would then generate the wedge correction signals (additive or multiplicative) itself. The wedge correction signals are determined before actual treatment of a patient in one or more calibration runs. To determine relative wedge correction values, a reference surface is irradiated with a known reference plate position, and the radiation output over the surface is sensed by a conventional sensing device (not shown), which generates radiation output signals, which are applied to the processing unit 18. In particular, the radiation output at a reference point (for example, at the center of the beam) is sensed. The reference surface need not lie in the patient plane, although if it does the calibration will typically be easier and more accurate. The plates are then moved to a new opening position, the radiation output is sensed and the needed amount of adjustment is determined to create the proper isodose profile for that position. This process is continued until correction values are stored for the reference surface over the entire range of motion of the plates. If more than one set of movable plates is included, then correction values will be sensed and stored for each combination of plate positions; the number of combinations will depend on the desired or required resolution. The correction values indicate by how much the radiation output (for example, dose rate) is to be changed (via the wedge correction signals) such that the delivered dose distribution is equal to the desired dose distribution, that is, the isodose profiles are generated corresponding to what they would be if the radiation output were held constant and a physical wedge were included in the beam path. During actual treatment, for each plate position, the processing unit adjusts the radiation output to correspond to what is needed to generate the correct isodose profile. Since no actual physical wedge is included, however, and the system is calibrated for 100% output at the reference point, the therapist need not perform any calculations to adjust for a wedge factor. If additive offsets are chosen for the wedge correction factors, then the difference between the sensed output values and the desired output value is stored. If multiplicative correction factors are chosen, then ratios are stored. Alternatively, any known function approximation method may be used to generate the parameters of an approximating function of the additive or multiplicative wedge correction factors required. A "course" of radiation treatment may, and often does, have more than one field, and may run over several different sessions. In some cases, hundreds of different (and, in some cases, fixed) sequential fields with different wedges are used during a course, for example, to provide proper irradiation of a field that has a complicated geometry or prescribed dose profile, to lessen discomfort to the patient, or to adjust the field as a tumor shrinks during treatment. The invention therefore also comprises an optional verification and recording or "auto set-up" system 102 (see FIG. 2), which stores and downloads to the radiation system (via the CPU 18 or directly into the memory) the parameters, for example, of the geometry, of the various fields of the course of treatment, and/or the tables of wedge correction factors that were derived during earlier calibration runs for the various fields.
	claims	1. A system for irradiating a product with a source of radiation comprising, in combination, a) a source providing radiation to penetrate and irradiate the product; some of the radiation exiting the product; and b) a reflector of a high density, low Z material positioned to receive radiation exiting the product and to reflect back some portion of the radiation exiting the product to re-irradiate said product. 2. A system as in claim 1 wherein the product has a top surface, an opposite bottom surface, and side surfaces wherein claim 1 a) the source of radiation is positioned to irradiate the top surface of the product and penetrate the product; some of said radiation exiting on said opposite bottom surface and the side surfaces of the product; and wherein b) said reflector is positioned to receive and reflect back radiation exiting said product to re-irradiate the product from said bottom and side surfaces. 3. A method of irradiating a selected product comprising, in combination, a) directing radiation of sufficient energy to cause some of said radiation to penetrate and exit the product; b) positioning a reflector of a selected high density, low Z material at least three quarters inch thick to receive radiation exiting the product and to reflect said radiation; and c) directing the reflected radiation back to irradiate said product. 4. A system as in claim 1 wherein said reflector comprises boron carbide of at least three quarters inch in thickness. claim 1 5. A system as in claim 1 wherein said reflector comprises boron of at least three quarters inch in thickness. claim 1 6. A system as in claim 1 wherein said reflector comprises carbon of at least three quarters inch in thickness. claim 1 7. A system as in claim 1 wherein claim 1 a) the product has a top surface, a bottom surface and side surfaces and the radiation enter the top surface; b) said reflector comprises a low Z, high density material configured to reflect ray to the bottom surface of the product as well as to the sides of the product. 8. A system as in claim 1 wherein the reflector may be of boron, boron carbide or carbon of at least 10 cm in thickness to reflect X-rays or gamma rays. claim 1 9. A system as in claim 8 wherein the reflector is configured to reflect the radiation to selected areas of the product being irradiated. claim 8 10. A system for irradiating with X-rays a product which product has top, bottom and sides surfaces comprising, in combination, a) a source for providing X-rays directed to irradiate the top surface of the product; b) said source of X-rays providing X-rays suitable for penetrating at least 4 cms of water equivalent product; c) a reflector of a high density, low Z material positioned to receive X-rays exiting the product and to reflect back a major portion of the X-rays exiting the product to re-irradiate said product; d) said reflector being of boron carbide and being of a thickness of at least 10 cms in thickness, e) said reflector being configured to reflect X-rays back to the sides of the product as well as to the bottom of the product; and f) said reflector being positioned adjacent the bottom surface and side surfaces. 11. A system as in claim 10 wherein said reflected X-rays are selectively directed to specific areas of the product. claim 10 12. A system as in claim 1 wherein said reflector is formed as recess for containing the product. claim 1
	claims	1. An apparatus for cooling a spent fuel pool, the apparatus comprising:the spent fuel pool containing cooling water and cooling a spent fuel stored therein;a cooling water pool positioned above the spent fuel pool and also containing cooling water therein;a heat exchanger disposed in the cooling water pool and lowering temperature of the cooling water from the spent fuel pool;a cooling water recovery pipe connecting the spent fuel pool and a first end of the heat exchanger and allowing the cooling water from the spent fuel pool to flow into the heat exchanger;a cooling water circulation pump disposed in the cooling water recovery pipe;a cooling water supply pipe connecting a second end of the heat exchanger and the spent fuel pool and allowing the cooling water from the heat exchanger to flow to the spent fuel pool;an emergency cooling water supply pipe connecting a bottom side of the cooling water pool and the spent fuel pool and supplying the cooling water in the cooling water pool to the spent fuel pool by gravity, the emergency cooling water supply pipe being separated from the cooling water supply pipe;a floating valve disposed in the emergency cooling water supply pipe;a floating device disposed in the spent fuel pool and opening and closing the floating valve according to a level of the cooling water in the spent fuel pool; anda transportable cooling water circulation pipe having a first end connected to the cooling water supply pipe upstream of the cooling water circulation pump and a second end connected to the spent fuel pool, the transportable cooling water circulation pipe including a transportable pump, a first valve and a transportable heat exchanger connected in series along the transportable cooling water circulation pipe between the supply pipe and the spent fuel pool. 2. The apparatus according to claim 1, further comprising:an emergency cooling water circulation pipe connecting the cooling water pool and the spent fuel pool separately from the cooling water supply pipe and the emergency cooling water supply pipe and circulating the cooling water of the spent fuel pool and the cooling water pool. 3. The apparatus according to claim 1, wherein the cooling water pool comprises an air cooling pipe cooling the cooling water in the cooling water pool by air. 4. The apparatus according to claim 1, wherein the cooling water pool comprises a cooling water supplementing pipe supplying the cooling water from an outside of the cooling water pool into the cooling water pool. 5. The apparatus according to claim 1, further comprising:a second valve connecting the cooling water supply pipe downstream of the cooling water circulation pump and the transportable pump.
	summary
	abstract	A nuclear fuel pellet for a nuclear reactor is disclosed. The pellet comprises a metallic matrix and ceramic fuel particles of a fissile material dispersed in the metallic matrix. The metallic matrix is an alloy consisting of the principle elements U, Zr, Nb and Ti, and of possible rest elements. The concentration of each of the principle elements in the metallic matrix is at the most 50 molar-%.
055747582	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a method for measuring gamma rays of trace amounts of radionuclides (radioisotopes), such as iodine-131, cobalt-60, etc., particularly in primary water of a nuclear reactor, which further contains other radionuclides such as nitrogen-13, fluorine-18, cobalt-58, etc. each emitting a pair of annihilation gamma-rays. More particularly, the invention is concerned with an improved gamma-ray spectrometric method for measuring selectively gamma-rays of the aforementioned radionuclides by diminishing the annihilation gamma-rays emitted by radionuclides coexisting in the primary water, thus significantly elevating the detection limits of the gamma-rays. 2. Statement of Related Art For instance, at nuclear power plants, with a view toward safe operation thereof, leakage of nuclear fuel assemblies is always kept under surveillance, for example, by measuring gamma-rays of .sup.131 I .sup.60 Co ,etc. contained in trace amounts in water of a primary coolant of an individual nuclear reactor. However, other radionuclides such as .sup.13 N, .sup.18 F, .sup.58 Co, etc., which are unstable radionuclides emitting .beta..sup.+(e.sup.+) or positron, are also contained in the primary water, and soon .beta..sup.+ decay at a low energy level by absorption in a substance and bonding with electrons therein at the end of their ranges. At that time, one positron and one electron are annihilated, emitting annihilation gamma-rays of 0.511 MeV in diametrically opposite directions. The coexistence of the radionuclides emitting the annihilation gamma-rays is a major disturbing factor for the measurement of the intended gamma-rays, particularly gamma-ray of .sup.131 I, which has a close energy level (0.364 MeV) to the annihilation gamma-rays. Additionally, Compton scattering caused inevitably in a gamma-ray spectrometry also interferes with the intended measurement. As a consequence, it is essential for gamma-ray spectrometric measurement of gamma-rays of the intended radionuclides (.sup.131 I, .sup.60 Co, etc. ) in the primary water that the annihilation gamma-rays be minimized while Compton backgrounds or continua of the resulting gamma-spectra due to the gamma-rays and annihilation gamma-rays are suppressed. Hitherto, iodine-131 and other radionuclides emitting gamma-rays in water of a primary coolant has been measured by means of a germanium (Ge) detector or a scintillation detector of NaI(T1) (sodium iodide activated by thallium) or Bi.sub.4 Ge.sub.3 O.sub.12 (bismuth germanate known as BGO), or a gamma-ray sepctrometric measurement system wherein a scintillation detector is disposed around a germanium detector. The method using the Ge detector was poor in detection limit of .sup.131 I owing to the effect of Compton backgrounds produced from .sup.131 I gamma-ray, .sup.60 Co gamma-ray the annihilation gamma-rays, etc., so that trace amounts of .sup.131 I and other radionuclides emitting gamma-rays in the primary water couldn't be measured. Only in the event that .sup.131 I, .sup.60 Co, etc. were leaked from a fuel assembly into the primary water, increased concentrations of them enabled the measurement. Again, the method using the NaI(T1) detector was too inferior to the Ge detector method in resolution power. The method using both Ge detector and scintillation detector has been improved more or less over the preceding methods, but it has still not been possible to measure extremely slight concentrations of .sup.131 I, .sup.60 Co and others. Thus, any of the known gamma-ray spectrometric methods has not been satisfactory and feasible because the annihilation gamma-rays from coexsiting radionuclides in the primary water have interfered with the measurement of the intended radionuclides, e.g., .sup.131 I, etc. Another method for measuring .sup.131 I and other radionuclides by chemical analysis has been known, but has yielded disadvantageously awkward radioactive wastes, which should be handled or disposed of with great care. Hence, this is not suitable for frequent or continuous measurement. In view of the drawbacks or problems as encountered in the prior art measurement methods of gamma-rays in primary water of a nuclear reactor as stated above or gamma-rays in another radioactive substances, the present invention is designed to provide a gamma-ray spectrometric measurement method which enables to significantly enhance detection limits of gamma-ray-emetting radionuclides, particularly in the primary water. That is to say, it is a primary object of the invention to provide an improved method for measuring selectively gamma-rays of radionuclides (iodine-131, cobalt-60, etc.), particularly in the primary water contained in micro-quantities by excluding disturbing factors to the measurement, namely, the aforesaid annihilation gamma-rays emitted by other radionuclides coexisting in the primary water, and Compton effects due to the gamma-rays and annihilation gamma rays as far as possible. Another object of this invention is to provide a high-sensitive measurement method capable of detecting such extremely slight amounts of the radionuclides emitting gamma rays in the primary water that it has been not possible to detect hitherto. Further object is to provide a reliable measurement method which enables continuous surveillance of leakage of a nuclear fuel assembly, thereby assisting in early prevention of the risk. SUMMARY OF THE INVENTION The invention for achieving the foregoing objects resides generally in a method for measuring selectively gamma-rays of radionuclides of microquantities, particularly in primary water of a nuclear reactor, coexisting with radionuclides each emitting a pair of annihilation gamma-rays in diametrically opposite directions, using a gamma-ray spectrometric system which includes a primary detector for detecting photons of the gamma-rays and photons of the one annihilation gamma-rays in the one direction, a secondary detector for detecting at least photons of the other annihilation gamma-rays in the opposite direction, a 10 shield detector for detecting photons of Compton-scattered gamma-rays escaped from the primary detector to the shield detector, and an anticoincidence circuit connecting with the primary, secondary, and shield detectors, the primary detector and the secondary detector being located in opposed manner relative to the axis of a coolant pipe, through which the primary water flows, the shield detector being disposed to surround the primary detector except for its portion facing to the pipe on which the gamma-rays and the annihilation gamma-rays are incident. The method comprises: detecting the photons of the gamma-rays and the photons of the one annihilation gamma-rays on the primary detector as pulses while detecting the photons of the other annihilation gamma-rays on the secondary detector as pulses; and counting the pulses from the secondary detector in anticoincidence with the pulses from the primary detector thereby to reject the recording of the pulses of the annihilation gamma-rays, thus minimizing the annihilation gamma-rays; and subsequently measuring count numbers of the gamma-rays on the basis of the analysis of the pulses. More preferably, the method further comprises, simultaneously with the foregoing detecting step, detecting the photons of the Compton-scattered and escaped gamma-rays on the shield detector as pulses and counting the pulses from the shield detector in anticoincidence with pulses from the primary detector thereby to reject the recording of the pulses of the Compton-scattered gamma-rays, thus additionally diminishing the Compton gamma-rays. According to the method of this invention, when the photons of the annihilation gamma-rays emitted in diametrically opposite directions are coincidently detected on the primary and secondary detectors located in an opposed relation to each other, the resulting pulses are rejected by anticoincidence counting operation of the anticoincidence circuit, whereby the annihilation gamma-rays are vastly reduced from the primary detector. Consequently, it is possible to elevate significantly the detection limits of the intended gamma-rays of iodine-131, cobalt-60, etc. Further according to a preferred embodiment, when the photons of Compton-scatterered gamma-rays are coincidently detected on the primary and shield detectors, the resulting pulses are rejected by anticoincidence counting operation of the anticoincidence circuit, whereby the Compton gamma-rays are also significantly diminished from the primary detector, which enables to further enhance the detection limits of the intended gamma-rays of radionuclides.
	description	The present application is a continuation of U.S. patent application Ser. No. 15/838,414 filed Dec. 12, 2017, which is a continuation of U.S. patent application Ser. No. 14/760,215 filed Jul. 10, 2015 (now U.S. Pat. No. 9,852,822), which is a U.S. national stage application under 35 U.S.C. § 371 of PCT/US2014/010967 filed Jan. 10, 2014, which claims the benefit of U.S. Provisional Patent Application Ser. No. 61/750,986 filed Jan. 10, 2013; the entireties of which are incorporated herein by reference. The present invention relates to spent nuclear fuel and radioactive waste storage systems, and more particularly to such a system suitable for consolidated interim waste storage. Used or spent nuclear fuel and radioactive waste materials are presently stored on an interim basis “on site” at commissioned and some decommissioned nuclear generating plants until the federal government provides a central permanent repository. For example, spent nuclear fuel is stored in the reactor fuel pool after removal from the core where it continues to generate decay heat. The fuel can be transferred after a period of cooling in the pool to canisters which are placed in dry storage casks (i.e. overpacks) typically constructed of concrete, steel, and iron, etc. to provide containment and radiation shielding. The casks are stored on site at the generating plant. The concept of using consolidated interim storage (CIS) is intended to provide geographically distributed off-site storage facilities for spent nuclear fuel and radioactive wastes (collectively “waste”) gathered from a number of individual generating plant sites, thereby providing greater control over the widely dispersed waste stockpiles. The waste materials would initially be transported to the CIS facility from the generating plants for a period of time, with the eventual goal of a final move to a permanent nuclear waste repository when available. Such so called independent spent fuel storage installations (ISFSI) are facilities designed for the interim storage of spent nuclear fuel comprising solid, reactor-related, greater than Class C waste, in addition to other related radioactive materials. Each ISFSI facility would typically maintain an inventor of a multitude of waste canisters containing spent nuclear fuel and/or radioactive waste materials. The present disclosure provides a below-ground used or spent nuclear fuel storage system designed for the compact dry storage of a large number of used fuel canisters in a small land area. In a non-limiting exemplary embodiment, two or more elongated canisters may be stored in vertically oriented and stacked relationship in each of a plurality of underground vertical ventilated storage modules which provide an overpack. The storage modules may be diametrically sized to fit a single canister therein at a given elevation, as further described herein. The collective array of storage modules defines an independent spent fuel storage installation (ISFSI) facility suitable for CIS that may include any number and arrangement of modules. The canisters may contain both radioactive used nuclear fuel and/or non-fuel waste materials in some embodiments. In one embodiment, the canisters may be Multi-Purpose Canisters (MPCs) available from Holtec International of Marlton, N.J. The underground storage system is intended to provide vanishingly low site boundary radiation dose levels and safety during catastrophic events. As an underground system, the system takes advantage of the surrounding soil or subgrade to provide shielding, physical protection, and a low center of gravity for a stable storage installation. Each vertical storage module provides storage of canisters in a vertical configuration inside a cylindrical cavity located entirely below the top-of-grade in the storage facility. The vertical modules may each be generally comprised of a cavity enclosure container formed by an outer shell, an inner divider shell, and a top closure lid in addition to various interfacing structures and features, as further described herein. The canister storage system is further configured to provide passive heat removal from the canisters via natural convection during storage in the modules, thereby rejecting the used fuel's decay heat emitted to the ambient air above the module. Radiation emitted from the used nuclear fuel is substantially contained within the soil fill in which the modules are disposed and canisters stored. Advantageously, stacking two canisters in each vertical ventilated storage module according to the present disclosure ultimately cuts the required storage area in half. For example, a 14 acre ISFSI for CIS can store 4,000 canisters containing more than 50,000 tons of uranium. This significantly reduces siting requirements. The radiation released to the environment from such a CIS facility storing used fuel may be negligible. According to one exemplary embodiment, a system for vertically-stacked storage of nuclear waste canisters includes an elongated outer shell defining a vertical axis and an internal cavity; a first canister positioned in the cavity in a lower position; a second canister vertically stacked above the first canister in an upper position, the first and second canisters being concentrically aligned with the vertical axis; a centering and spacing ring assembly interspersed between the first and second canisters; and a removable top lid mounted on top of the outer shell covering the cavity. The centering and spacing ring assembly is arranged and operable to transfer weight of the second canister to the first canister. According to another embodiment, a storage module for vertically-stacked storage of nuclear waste canisters includes an elongated outer shell defining a vertical axis and an internal cavity; an elongated inner shell disposed in the internal cavity; a first annular space formed between the inner and outer shells, the first annular spacing defining a vertical downcomer ventilation shaft operable to convey ambient cooling air downwards to the cavity; a first canister positioned in the cavity in a lower position; a second canister vertically stacked above the first canister in an upper position, the first and second canisters being concentrically aligned with the vertical axis; a middle centering and spacing ring assembly interspersed between the first and second canisters, the middle centering and spacing ring assembly operable to transfer weight of the second canister to the first canister; a second annular space formed between the first and second canisters and the inner shell, the second annular space defining a vertical riser ventilation shaft operable to convey cooling air upwards across outer surfaces of the canisters; and a removable top lid mounted on top of the outer shell covering the cavity, the top lid being in fluid communication with the riser ventilation shaft and configured to form an airflow pathway to atmosphere through the lid. According to another embodiment, an underground storage module for vertically-stacked storage of nuclear waste canisters includes an elongated vertical outer shell defining vertical axis and an internal cavity, the outer shell having a top and a hermetically sealed bottom, the outer shell being disposed below grade for a majority of its height; a common inlet air plenum disposed at the top of the outer shell, the air plenum arranged to draw ambient cooling air through a plurality of air inlets in fluid communication with the air plenum; an annular-shaped vertical downcomer ventilation shaft arranged to convey the cooling air from the inlet air plenum downwards along the outer shell to a bottom of the cavity; a first canister positioned in the cavity in a lower position; a second canister vertically stacked above the first canister in an upper position, the first and second canisters being concentrically aligned with the vertical axis; an elongated inner shell disposed inside the outer shell and cavity; an annular-shaped vertical riser ventilation shaft formed between the inner shell and the canisters, the riser ventilation shaft being in fluid communication with the downcomer ventilation shaft near the bottom of the outer shell and arranged to convey cooling air upwards across outer sidewall surfaces of the canisters for removing decay heat; and a removable top lid mounted on top of the outer shell covering the cavity, the top lid in fluid communication with the riser ventilation shaft and configured to form an airflow pathway to atmosphere through the lid from the riser ventilation shaft. All drawings are schematic and not necessarily to scale. Parts given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. References herein to a figure number (e.g. FIG. 1) shall be construed to be a reference to all subpart figures in the group (e.g. FIGS. 1A, 1B, etc.) unless otherwise indicated. The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. FIG. 1 depicts an ISFSI facility forming a high-density subterranean Consolidated Interim Storage (CIS) system 100 comprising an array of underground vertical ventilated storage modules 110. In one embodiment, each storage module 110 houses at least two sealed canisters containing spent nuclear fuel and/or radioactive waste materials. The storage modules 110 are arranged in a tightly packed configuration to minimize spatial site requirements. The storage modules 110 are spaced apart by a grid of orthogonally intersecting aisles 102 formed of concrete slabs 104 to provide access for commercially-available motorized wheeled equipment operable to move and lift (i.e. raise/lower) the canisters for insertion into and removal from the modules. Such equipment is well known to those skilled in the art without further elaboration. The low exposed vertical profile of the storage modules 110 (as further described herein) allows the equipment to move over modules in a single row to the desired module for inserting or removing canisters. Each storage module 110 may include an associated concrete top pad 112 which is positioned and disposed between the aisles 102 of slabs 104. The top pads 112 may form a contiguous structure with slabs 104 to provide radiation shielding. The top pads 112 may be square-shaped in top plan view in one non-limiting example; however, other suitable polygonal and non-polygonal configurations (e.g. circular) may be used. FIG. 2 is a cross-sectional view of a single storage module 110 from FIG. 1. With additional reference to FIGS. 6 and 6A, storage module 110 is vertically elongated and includes a vertically-extending outer shell 120 defining an internal cavity 122 and an inner shell 130. In one embodiment, outer and inner shells 120, 130 are cylindrically and complementary shaped, albeit dimensionally different. The outer shell 120 provides an impermeable barrier against leakage of ground water through the earthen soil S fill into the storage module 110. Outer shell has an open top 121 and a closed bottom formed by bottom plate 123 at the bottom end of the shell. Bottom plate 123 may be substantially flat and is preferably seal welded to outer shell 120 in one embodiment to hermetically seal the bottom of the shell thereby forming an impermeable barrier to ingress of external ground water. Accordingly, all portions of storage module 110 below grade are sealed against the ingress of moisture or water from the environment transmitted through the soil. The bottom plate 123 of outer shell 120 may be positioned on and supported by a concrete base pad 106. The area adjacent the outer shell 120 between the top pad 112 and base pad 106 is filled with fill or soil “S”, thereby forming a cross-sectional composite structure of upper and lower concrete caps with soil disposed therebetween. A majority of the height of the outer and inner shells 120, 130 is disposed below grade as shown in FIGS. 2 and 6. The top portion of outer shell 120 is surrounded by the top pad 112 and embedded therein so that virtually none or only a small projection of the top end 121 of the outer shell protrudes above the concrete pad. Substantially the entire height of the outer shell 120 is therefore embedded in soil and/or concrete in the embodiment being described. It will be appreciated that in some embodiments, a monolithic concrete base pad 106 may extend beneath a plurality or cluster of individual storage modules 110 in lieu of individually poured pads. Similarly, a monolithic top pad 112 may be used to surround and extend between a plurality or cluster of individual storage modules 110 in lieu of individually poured pads. It will be appreciated that although the cross-sectional shape of the outer and inner shells 120, 130 may be cylindrical in the illustrated embodiment, the shells can have other suitable polygonal and non-polygonal shapes, including without limitation rectangular, conical, hexagonal, or irregularly shaped. In some embodiments, the outer and inner shells 120, 130 need not be concentrically aligned with each other. Outer and inner shells 120, 130 and bottom plate 123 are made of metal, such as steel in exemplary non-limiting embodiments. Outer shell 120, which provides a barrier between the soil S in which the outer shell is embedded, is preferably made of a corrosion resistant metal such as without limitation coated steel, stainless steel, etc. In one embodiment, inner shell 130 has an open top 131 and open bottom 132 (reference FIGS. 6 and 6A). The open top 131 allow insertion of canisters 140 into the storage module 110, as further described herein. Inner shell 130 may be coaxially and concentrically aligned with outer shell 120 about a vertical axis VA defined by storage module 110. Outer and inner shell 120, 130 have vertical heights that are substantially coextensive. The bottom 132 of inner shell 130 may rest on top of the bottom plate 123 of the outer shell 120 in one arrangement. The outer and inner shells 120, 130 have a sufficient height or depth suitable to allow at least two canisters 140 to be stored in a vertically stacked configuration or relationship. In other embodiments, the outer and inner shells 120, 130 may have heights or depths suitable for holding three or more vertically stacked canisters 140. Inner shell 130 has a smaller diameter than outer shell 120. Inner shell 130 is radially spaced apart inwards from the outer shell 120 and acts to divide the cavity 120 into an outer annular space 124 and an inner portion configured and dimensioned to hold canisters 140. The outer annular space 124 extends from the top 121 to bottom plate 123 of outer shell 120. The outer annular space 124 defines an annular-shaped vertical downcomer ventilation shaft 125 for introducing outside ambient cooling air into cavity 122 of storage module 110 to remove decay heat emitted from the spent nuclear fuel or radioactive waste material contained in canisters 140. To complete a natural convection heat removal airflow circuit, a second inner annular space 133 is defined between the outer cylindrical shell sidewall 141 of canister 140 and inner shell 130 which is radially spaced apart outwards from the canister. The second inner annular space 133 extends from the bottom 132 to top 131 of inner shell 130 and defines an annular-shaped vertical riser ventilation shaft 134 for removing heated cooling air from storage module 110. The inner shell 130 serves to separate the downcomer ventilation air from the up-flowing air heated by contact with the canister (see, e.g. airflow diagram of FIG. 5). The downcomer ventilation shaft 125 is fluidly (i.e. airflow) separated and isolated from riser ventilation shaft 134 by inner shell 130 for substantially the entire vertical height of storage module 110 along shells 120, 130 except near the bottom of the storage module. The downcomer and riser ventilation shafts 125, 134 respectively are in fluid communication through a plurality of circumferentially arranged and spaced apart airflow openings 135 formed at the bottom end of the inner shell 130 near the bottom 132. The bottom end of shell 130 may have a castellated configuration in one embodiment with the openings 135 having a generally square or rectangular shaped configuration. Other suitable shaped airflow openings may be used however. In one embodiment, the inner “divider” shell 130 is insulated being provided with an insulation layer 150 to minimize heat exchange between the incoming cooling (downcomer ventilation shaft 125) and outgoing heated (riser ventilation shaft 134) ventilation air in contact with the inner shells inner and outer surfaces, respectively (see, e.g. FIGS. 4 and 5, and best shown in detail in FIG. 4B). This keeps the ambient cooling air drawn into the storage module 110 by natural convention as cold as possible until it encounters the hotter outer shell sidewall 141 of the canisters 140 to maximize cooling efficiency. In one embodiment, the insulation layer 150 is disposed on the outer surface of inner shell 130 between the inner shell and outer shell 120 in the downcomer ventilation shaft 125 to prevent damage which could potentially occur from inserting and removing canisters from the storage module 110. Any suitable type of insulating material may be used including without limitation separately formed and applied pliable, semi-rigid, and rigid insulating materials and sprayed on types (e.g. hardening foams). The insulation 26 is preferably chosen so that it is water and radiation resistant without substantial degradation. Some examples of suitable forms of insulation include, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alumina and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). The desired thickness of the of insulation layer 150 will be dictated by such considerations such as the heat load (e.g. temperature differential between ambient air and canister external temperatures), the thickness of the shells, and the type of insulation used (e.g. K factor). In some embodiments, the insulation may have a representative non-limiting thickness in a range of about ½ to 6 inches for example. Canisters 140 stored in storage module 110 may be any type of canister, including without limitation Multi-Purpose Canisters (MPCs) available from Holtec International of Marlton, N.J. As shown in FIG. 5, canisters 140 have a generally hollow cylindrical shell sidewall 141 including a top 142 with removable and sealable lids 144 for inserting spent nuclear fuel and/or radioactive waste materials and an opposing bottom 143. The interior of canisters 140 may include racks or grids to contain and support spent fuel rods and waste materials. FIGS. 4 and 4A show the construction of the upper portion of storage module 110 and top pad 112 in greater detail. The tops 121 and 131 of outer and inner shells 120, 130 respectively penetrate top pad 112, thereby providing external above grade access to downcomer ventilation shaft 125, riser ventilation shaft 134, and cavity 122 of storage module 110 for inserting and removing canisters 140. The top pad 112, which may be formed of concrete in a one preferred embodiment, extends at least partially beyond the diameter of outer shell 120 as shown. In this non-limiting example, the top pad 112 may have a square shape (in top plan view) as previously described. The perimeter of the top pad 112 would be adjoined by the access aisles 102 formed between adjacent storage modules 110. With continuing reference to FIGS. 4 and 4A, top pad 112 includes one or more air inlets 160 which are in fluid communication with annular-shaped downcomer ventilation shaft 125 to introduce cooling ambient ventilation air into the storage module 110. A common recessed air plenum 161 covered by a removable cover plate 162 may be formed in top pad 112 around outer shell 120 to fluidly connect the air inlets to the ventilation shaft 125. Air plenum 161 further fluidly interconnects the air inlets 160 to each other. The air plenum 161 is formed by a recessed portion of top pad 112 and has a bottom surface 166 spaced vertically below the higher top surface 113 at peripheral portions of the pad 112, as shown. Air plenum 161 may have a complementary shape to top pad 112 (e.g. square in this embodiment), or a different shape. The air inlets 160 may be formed in one embodiment from short sections of pipe attached directly to and removable as a unit with the cover plate 162 (both of which preferably are formed of metal) by any suitable means (e.g. fasteners, welding, etc.). Cover plate 162 includes apertures 167 which fluidly communication with short pipe sections. The air inlet 160 pipe sections include lateral airflow openings 164 cut into the sides of the pipe and the open free end is covered by a weather cap 163 to prevent the direct ingress of rain and/or debris. The top end of shell 120 may include a plurality of circumferentially spaced apart airflow openings 165 which are in fluid communication with air plenum 161 to allow ambient ventilation air to flow through the air inlets 160 into the plenum and in turn down into the downcomer ventilation shaft 125 through the openings 165. The ambient ventilation cooling air is admitted through a plurality of air inlets 160 in the top pad 112 that are arranged to be non-preferential with respect to the horizontal direction of the wind to maximum cooling of the canisters in storage module 110. In one non-limiting embodiment, four air inlets 160 may be provided with one inlet being positioned at each of the four corners of the top pad 112 to ensure each quadrant of the storage module 110 via the downcomer ventilation shaft 125 receives an equal amount of ambient cooling ventilation air. The air plenum 161 advantageously serves to further distribute the ventilation air uniformly to all portions of the downcomer ventilation shaft 125. It will be appreciated that other suitable configurations and numbers of air inlets 160 and configurations of air plenum 161 may be provided depending on the configuration of top pad 112 used and other factors. The heated ventilation air exits riser ventilation shaft 134 from storage module 110 through a central airflow passageway 201 in the top lid 200 shown in FIGS. 4, 4A, and 5. Top lid 200 is a removable cover that closes storage module 110 and is positioned over the tops 121 and 131 of outer and inner shells 120, 130, respectively. The top lid 200 is a massive stepped-shaped circular shielded structure in one embodiment equipped with a diametrically enlarged upper portion 202 and smaller cylindrical bottom protrusion 203 extending downwards therefrom. The upper portion 202 has a larger diameter than the diameter of the outer shell 120 forming an annular shaped peripheral portion (in top plan view) overhanging the outer shell. In some embodiments, upper portion 202 is configured and dimensioned to close off both the open tops 121, 131 of the outer and inner shells 120, 130. This effectively seals off the top of vertical downcomer ventilation shaft 125 to prevent inlet cooling ventilation air entering from air plenum 161 via airflow openings 165 at the top of outer shell 120 from bypassing the shaft 125 and entering the riser ventilation shaft 134 at the top of the storage module 110. In some embodiments, a top seal plate may be used to seal the top of vertical downcomer ventilation shaft 125 in addition to top lid 200. An annular gasket 250 formed of a suitable material may be provided between the underside of the upper portion 202 of top lid 200 and the inner top 131 of the inner shell 130 providing a sealed lid-to-inner shell interface (see FIG. 4A). The bottom protrusion 203 has a diameter smaller than the inner shell 130 and forms a plug that is inserted at least partially into the inner shell into cavity 122 to keep the lid 200 from sliding in a lateral direction excessively during a seismic event (e.g. earthquake). The annular gap G formed between the bottom protrusion 203 and inner surface of inner shell 130 forms a continuation of vertical riser ventilation shaft 134 as best shown in FIGS. 4A and 5. In some embodiments, the top lid 150 may be a substantially hollow metal structure filled with a radiation absorbing material shielding such as concrete. The metal exoskeleton of top lid 150 can be constructed of a wide variety of materials, including without limitation various steel, stainless steel, aluminum, aluminum-alloys, and other metals. In some embodiments, the lid may be constructed of a single piece of material, such as concrete or steel for example. With continuing reference to FIGS. 4, 4A, and 5, upper portion 202 of top lid 200 is annular shaped which defines the central airflow passageway 201 to eject heated ventilation air from vertical riser ventilation shaft 134 to the ambient environment. To complete this airflow circuit, the bottom protrusion 203 of lid 200 includes a plurality of radially extending air passages 205 which are in fluid communication with the annular-shaped riser ventilation shaft 134 and central airflow passageway 201 in the upper portion 202 of the lid. An air outlet 210 extending upwards from top lid 200 may be mechanically thereto and fluidly coupled to central airflow passageway 201 to help vent heated ventilation air away from storage module 110. The air outlet 210 may be formed in one embodiment from a short section of pipe attached to the upper portion 202 of top lid 200 by any suitable means (e.g. fasteners, welding, etc.). The air outlet pipe sections include lateral airflow openings 211 cut into the sides of the pipe and the open free end is covered by a weather cap 212 to prevent the direct ingress of rain and/or debris. In one embodiment, top lid 200 may include four intersecting rigging plates 204 useable to raiser and lower the lid into position on storage module 110 (see, e.g. FIGS. 4 and 4A). The plates 204 may be welded to the metal exoskeleton plates of the lid 200. The plates 204 may extend from the bottom of bottom protrusion 203 to the top of upper portion 202, and in some embodiments have center extension sections 206 which extend radially inwards into central airflow passageway 201 and protrude upwards therefrom. Extension section of rigging plates 204 may therefore extend vertically upwards into and be covered by air outlet 210 when storage module 110 is in use. Extension sections 206 are configured for grappling by rigging and hoisting equipment (e.g. holes, clips, etc.) to facilitate manipulating and maneuvering the top lid 200. Other suitable configurations and arrangements for rigging lid 200 are possible. FIG. 5 is an airflow diagram showing the cooling ventilation air path through storage module 110 created by the features described above. As shown by the directional airflow arrows in this figure, the cooling ventilation air will travel in a generally U-shaped airflow path through the storage module 110 to remove and dissipate decay heat emitted from the canisters 140 by the spent nuclear fuel and/or radioactive waste stored therein. Airflow circulation is created by natural convection induced by the decay heat liberated. In operation, ambient cooling air is first drawn into air plenum 161 through each of the individual air inlets 160 and is mixed together. The inlet air circulates around and through the plenum. It should be noted that the air plenum 161 prevents ambient air flowing from the air inlets 160 directly into the annular vertical downcomer ventilation shaft 125. Advantageously, this mitigates the effects of preferential wind direction which otherwise might adversely affect the amounts of cooling ventilating airflow reaching certain portions of the storage module 110 and canisters 140 therein. Without the plenum and its airflow balancing effects, certain areas of the canisters 140 may be starved of cooling air while other portions receive cooling resulting in differential cooling of the canisters shell sidewalls 141. This would reduce the natural convection cooling efficiency. With continuing reference to FIG. 5, the cooling ventilation air leaves the air plenum 161 through the airflow openings 165 and enters the vertical downcomer ventilation shaft 125. The air flows downwards in the downcomer ventilation shaft 125 towards the bottom of the storage module 110. The cooling ventilation air travels through the plurality of airflow openings 135 formed at the bottom end of the inner shell 130 near its bottom 132 (see also FIGS. 4 and 4C) and enters the bottom of cavity 122 and the annular vertical riser ventilation shaft 134. The cooling air reverses direction and flows upward through the riser ventilation shaft 134 contacting the exposed outer circumferential surfaces of the first the bottom and then the top canister 140 in storage module 110 to draw away decay heat via convection. As the ventilation air flows vertically upward along the canisters 140 in the riser ventilation shaft 134, the air becomes progressively heated. The now heated ventilation air flows to and eventually reaches the top of the storage module 110 at the top of cavity 122 in the vertical riser ventilation shaft 134. The air flow changes direction and flows radially inwards through the radial air passages 205 in top lid 200 and is recombined in the central airflow passageway 201 in the upper portion 202 of the lid. The ventilation air then changes direction again and flows vertically upward entering air outlet 210 from which it is exhausted to atmosphere completing the ventilation airflow cycle. The support and placement of the multiple canisters 140 in storage module 110 will now further described. Referring to FIGS. 4 and 4C, the bottom or lower canister 140 (shown in, e.g. FIGS. 2 and 3) is horizontally supported and laterally restrained by a circumferentially spaced apart series of suitably shaped centering lugs 300. Lugs 300 are oriented and extend in a radial direction from the vertical axis VA of the storage module 110. When placed in the storage module 110, the lower canister 140 rests on bottom plate 123 welded to the outer shell 120 and transfers dead load (weight) of both the lower and top or upper canisters 140 housed in the storage module 110 to the concrete base pad 106. The canister 140 is positioned laterally adjacent to and inside the ring of radial lugs 300 as seen in FIG. 2. The lugs 300 are located proximate to the shell sidewall 141 of the canister 140 and positioned to fully engage the canister in the event of a lateral shift in position of the canister caused by a seismic event. This would stabilize the canister 140 and prevent excessive horizontal movement to protect the canister and its contents. Any suitable number of lugs 300 may be provided. Lugs 300 may be formed from generally flat steel plate in one embodiment and extend both upwards and inwards from the outer shell 120 towards the vertical axis VA (see, e.g. FIG. 6A). As shown, the lugs 300 may have a substantially greater radial width and axial height than thickness T (thickness being measured perpendicular to the width and height in a circumferential direction along the outer shell 120). At least a portion of the innermost axial edge 301 of lugs 300 is preferably straight or flat and arranged parallel to the shell sidewall 141 of canister 140 and vertical axis VA to prevent puncturing the canister in case of a seismic event. The lug 300 includes an angled edge 302 which adjoins the axial edge 301 that is angled downwards and inwards as shown in FIG. 6A. When the lower canister 140 is initially lowered into storage module 110, this angled edge 302 helps blindly guide and center the bottom 143 of the canister towards the centerline or vertical axis VA so that the canister becomes properly seated on the bottom plate 123 or ring 310 if provided. Centering and spacing lugs 300 may be attached to the outer shell 120 and/or inner shell 130 and are essentially not vertical load bearing structural members. In one exemplary arrangement, lugs 300 may be directly attached to the shell 120 (e.g. welded) through slots 136 formed through inner shell 130 at the location of each lug 300. The slots may be closed at the top and open at the bottom adjacent bottom 132 of the inner shell 130 to allow the inner shell to slide over the lugs when initially inserted into the outer shell 120 during fabrication. The bottom ends of the inner shell 130 may then rest on the flat bottom plate 123 affixed to the outer shell 120. In one embodiment shown in FIGS. 6 and 6A, an interfacing centering and spacing bottom ring 310 may be provided for some specific canister designs to engage the bottom 143 of the canister. Ring 310 is disposed inside the lugs 300 and may be fixedly attached thereto (e.g. welded) or a separate element in various embodiments. In the latter separate or loose construction, the ring 310 may have peripheral notches configured to engage the lugs 300 for preventing the ring from rotating in relation to the lugs and outer shell 120. On other embodiments, the ring 300 fits loosely inside lugs 300 without notches. The ring 300 rests on bottom plate 123 of outer shell 120 in abutting contact for transferring dead load (weight) of the canister 140 to the concrete base pad 106. Ring 310 is preferably made of metal, such as a suitable steel. In some embodiments, the top surface 311 of ring 310 may be castellated including a plurality of alternating arcuate raised segments 312 and arcuate recessed segments 314 having a complementary configuration to match and engage similarly configured features on the bottom 143 of a lower canister 140. The segments 312, 314 may extend radially from the inside to the outside of the ring 300 as best shown in FIG. 6A. Segments 312, 314 have a circumferentially measured arc width greater than the corresponding width of the lugs 300, and preferably may have a width at least coextensive with the radial depth of the segments (measured from the center of the ring outwards). The bottom surface 313 of ring 310 may be substantially flat. Referring to FIGS. 2, 2A, and 5, the lower and upper canisters 140 are horizontally supported and laterally restrained against the inner shell 130 by a centering and spacing middle ring 320. Ring 320 may be configured and constructed similarly to bottom support ring 310 described above and shown in FIGS. 6 and 6A. In one embodiment, middle ring 320 may be a composite structure formed of an upper ring 320A and lower ring 320B each shaped similarly to bottom ring 310. The two rings 320A, 320B may be attached together (e.g. welded) in back-to-back relationship with flat bottoms 313 in contact and exposed surfaces 311 with the raised and recessed segments 312, 314 facing axially outwards as best shown in FIG. 2A. It will be appreciated that in some embodiments, the upward and downward facing exposed top surfaces 311 of the middle ring 310 may be substantially flat without raised/recessed segments 312, 314 depending on the canister design used. If different configuration lower and upper canisters 140 are to be accommodated in the storage module 110, one of the rings 310A or 310B may be castellated (i.e. raised/recessed segments 312, 314) and the other may be flat on both surfaces. Accordingly, any combination may advantageously be used depending on the canister types to be stored in the storage module 110. Referring to FIGS. 2, 2A, and 5, centering and spacing middle ring 320 assembly further includes radially extending centering lugs 322 arranged in a circumferentially spaced pattern around the ring similarly to the bottom lugs 300 and ring 310 assembly already described. In one construction, the centering lugs 322 are welded around the perimeter of middle ring 320 and made integral therewith; both of which preferably are both made of suitable metal such as steel. The lugs 322 may not be fixedly attached to the inner shell 130 of storage module 110 such that the middle ring-lug assembly 320/322 is removable as a unit from the storage module with the canisters 140. This allows the first lower canister 140 to be first positioned in the bottom half of the storage module 110, the middle ring-lug assembly 320/322 then lowered and placed on top of the lower canister, and then the second upper canister 140 lowered and positioned into the top half of the storage module 110 to engage the middle ring-lug assembly 320/322. The middle ring-lug assembly 320/322 may alternatively be lowered into the inner shell 130 with the lower canister simultaneously allowing the ring-lug assembly to be placed on top of the lower canister before being lowered into the inner shell together in one step. FIG. 2 shows the lower canister 140 in position within the storage module 110 and ready for receiving the upper canister 140. The middle ring-lug assembly 320/322 is in position. The centering lugs 322 may have a similar side profile as lugs 310 already described including angled edges 302 to help guide and center the bottom 143 of the upper canister 140 when lowered into storage module 110 on top of the lower canister (see also FIG. 2A). Both the upper and lower portions of lugs 322 may include angled edges 302 as shown which helps center and properly position the middle ring-lug assembly 320/322 on the top 142 of the lower canister 140 when the assembly is lowered into place in the storage module 110. Accordingly, in one embodiment the upper and lower portions of lugs 322 are mirror images. It will be appreciated that middle ring-lug assembly 320/322 in conjunction with the lower canister 140 supports the upper canister 140 as shown in FIGS. 3 and 3A. Advantageously, from a structural standpoint, the middle ring 320 transfers and distributes the weight of the upper canister to the inherently stronger and stiffer cylindrical sidewalls of the shell sidewall 141 of the lower canister instead of onto the central portion of the structurally weaker canister lid. This enhances the load bearing capability of the lower canister for supporting the weight of the upper canister. The middle ring-lug assembly 320/322 also laterally restrains the bottom end 143 of the upper canister 140. Accordingly, the centering lugs 322 are configured, dimensioned, and positioned to engage both the top 142 of the lower canister 140 and the bottom 143 of the upper canister 140. Significantly, the middle ring-lug assembly 320/322 further serves to maintain the inner annular space 133 and vertical riser ventilation shaft 134 formed between the canisters 140 and inner shell 130 by providing proper horizontal aligned of the canisters along the vertical axis VA of the storage module 110. The middle ring-lug assembly 320/322 also provides some vertical spacing between the top 142 of the lower canister 140 and bottom of the upper canister 140 to permit cooling ventilation air to flow in the small space between the two canisters to enhance cooling the canisters. Referring to FIGS. 3 and 3A, a top ring-lug assembly 330/332 is also provided to laterally support and restrain the top 142 of the upper canister 140 against the inner shell 130. This assembly is comprised of a single support ring 330 having the arcuate raised/recessed segments 312, 314 facing downwards towards the upper canister. The plurality of centering lugs 332 are welded to the perimeter of ring 330 in a similar manner to the middle ring 320 as already described. Preferably, the lugs 332 are not fixedly attached to the inner shell 130 like the middle centering lugs 322 to allow the top ring-lug assembly 330/332 to be removable in the same manner from the storage module 110. In some embodiments, both the top and bottom surfaces 311, 313 of the top ring 330 may be substantially flat instead of castellated. The centering lugs 332 may be configured similarly to lugs 310 or 322 already described. It should be noted that the centering lugs 300, 322, and 332 laterally restrain and horizontally support the lower and upper canisters 140 inside storage module 110 during a seismic event (e.g. earthquake) against excessive movement. In addition, these lugs also maintain the inner annular space 133 along the entire height of the module to preserve the inner annular space 133 between the sidewalls of the canister shells 140 and inner shell 130 of the storage module 110 thereby protecting the integrity of the vertical riser ventilation shaft 134 for proper ventilated cooling of the canisters. It should be noted that the support rings 310, 320, and 330 with undulating top surfaces 311 having raised and recessed segments 312, 314 may be used with both canisters 140 having plain (i.e. flat) top and bottom ends, or with specially configured ends as described herein with complementary configured ends as the rings to provide an anti-rotation feature. In other possible embodiments, the rings 310, 320, 330 may be substantially flat on both the top surface 311 and opposing bottom surface 313. In some alternative constructions, the middle and top lugs 322, 332 may be attached (welded) to the inner shell 130 of the storage module 110 and rings 320, 330 may be separate and removable elements. FIGS. 3 and 3A show storage module 110 with both lower and upper canisters 140 in position and top lid 200 in place after insertion of the canisters. The lower and upper canisters 140 are concentrically aligned with the vertical axis VA of the storage module 110. In one embodiment, the diameter of the inner shell 130 and diameter of the internal cavity 122 are only wide enough to accommodate a single canister 140 at each elevation of the storage module 110 so that two canisters will not fit in side-by-side relationship in the storage module. The canisters 140 undergo cooling by natural convection via the ventilated cooling air system described above and shown by the directional airflow arrows in FIG. 5. It stands noting that the upper canister 140 does not directly contact the lower canister, but instead bears on middle ring-lug assembly 320/322 which vertically separates and spaces the two canisters. The dead load or weight of the upper canister is transferred through the middle ring-lug assembly 320/322 to the lower canister which bears the weight of the upper canister. It will be appreciated that the number of vertically stacked canisters in each storage module 110 may be limited by the load carrying capacity of the canisters themselves since each canister in the stack transmits and bears the weight of the canisters above; the lowermost canister 140 in the stack bearing the entire dead weight of the whole canister stack. Accordingly, a vertically deeper (higher) storage module 110 and internal cavity 122 with additional canisters can be deployed if the structural strength of the lowermost canister 140 and the support foundation were accordingly strengthened to support greater than two canisters. According to the present invention, it bears noting that the top and bottom canisters 140 can be of different diameters and heights within a range of limits which fit within the storage module 110. The centering and spacing rings 310, 320, 330 with lugs 300, 322, 332 as described herein can be customized to provide the necessary adaptation for varying canister diameters and different end type features. Accordingly, the storage modules 110 disclosed herein are highly customizable to accept numerous types and sizes of canisters from a number of different canister suppliers or sources. While the foregoing description and drawings represent exemplary embodiments of the present disclosure, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made within the scope of the present disclosure. One skilled in the art will further appreciate that the embodiments may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the disclosure, which are particularly adapted to specific environments and operative requirements without departing from the principles described herein. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive. The appended claims should be construed broadly, to include other variants and embodiments of the disclosure, which may be made by those skilled in the art without departing from the scope and range of equivalents.
	abstract	A 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 her 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.
	description	1. Field of the Invention The invention generally relates to an irradiating device in the art of radiation processing, and a method for controlling the same. 2. Description of Related Art Radiation processing is now used to prepare macromolecular materials, keep foods fresh, sterilize medical products and drugs, protect products from contaminating, color crystals and pearls, and treat environmental contaminants with high energy electron beams, X-ray generated by a target hit by electron beams or Gamma ray radiated by a radionuclide. The radiation processing, as an economic, energy saving, manpower saving and harmless new processing method, is widely applied to various fields such as agriculture, industry and medicine, and becomes increasingly important. Generally, high energy electron beams are generated by accelerators such as traveling or standing wave linear accelerators, and DC high voltage accelerators, which further include static accelerators, transformer type accelerators with insulating core, electron curtain accelerators, high frequency and high voltage accelerators, etc. As shown in FIG. 1, an irradiating device outputting electron beams comprises an electron linear accelerator 1, a scanning magnet 3 mounted on the electron linear accelerator 1 via a flange 2, and a scanning box 4 of triangle shape. The scanning box 4 is provided with an electron beam exit window 10 right in a direction of electron beams output from the electron linear accelerator, and a cooling fluid loop 9 at the bottom of the scanning box 4 for cooling the electron beam exit window 10. The cooling fluid loop 9 is externally connected to a cooling fluid system via an inlet 8 and an outlet 12. In operation, the scanning magnet 3 scans in bidirectional mode, that is, a positive current and a negative scanning current are respectively supplied to the scanning magnet 3 for half of a scanning period. The irradiating device configured in this way typically directs electron beams 5 through scanning box 4, and then carries out radiation processing. Although radiation processing with electron beams has advantages of great power, high efficiency, excellent safety and so on, it could only be used for small or thin articles due to the low processing depth of electron beams. It is not suitable for processing big articles that cannot be separated into smaller ones, such as logs to be cleared of pests. Furthermore, the scanning boxes of the conventional electron beam exit window type irradiating devices are required to have scanning magnets with good stability. If the scanning magnet 3 fails in operation, even for a very short time, the electron beams 5 would damage the electron beam exit window 10 greatly, and even damage the entire system including the electron linear accelerator 1 and the scanning box 4. Therefore, an irradiating device outputting X-rays is typically used to irradiate big articles that cannot be separated into smaller ones. As shown in FIG. 2, such irradiating device outputting X-rays comprises an electron linear accelerator 1, a shift section 13 mounted on the electron linear accelerator 1 via a flange 2, a target 7 mounted at a center of the outputting beams, and a cooling fluid loop 9 at bottom of the shift section 13 for cooling the target 7. The cooling fluid loop 9 is externally connected to a cooling fluid system via an inlet 8 and an outlet 12. The radiation processing is carried out utilizing X-rays generated by the target hit by the electron beams originated from the electron linear accelerator 1. X-rays can penetrate deeply, and thereby carry out radiation processing on bigger articles. However, its efficiency is lower than electron beams since X-rays are converted by electron beams impinging on the target. In conventional accelerator irradiating devices, the radiation is implemented by either electron beams or X-rays generated by a target hit by electron beams. The articles suitable for being irradiated by these devices are limited. An object of the present invention is to overcome the above disadvantages of the prior arts by providing an irradiating device capable of outputting two radiation sources, both the electron beams and X-rays. To achieve the above object, an aspect of the present invention provides an irradiating device comprising an electron accelerator, a scanning box connected to the electron accelerator, and a scanning magnet for controlling electron beams generated by the electron accelerator, wherein the scanning box is provided with both a target and an electron beam exit window, so that when the scanning magnet is not in operation, the electron beams impinge on the target and X-rays are generated to be output, and when the scanning magnet is in operation, the scanned electron beams are output via the electron beam exit window. The target can be positioned right in a direction of the electron beams generated by the electron accelerator. The electron beam exit window can be positioned at a left or right side of the target. The target can be positioned at an inner side of the electron beam exit window, forming an inner target structure. The scanning box can further be provided with a cooling fluid loop for cooling the target and the electron beam exit window. When the scanning magnet is in operation, the scanning center of the scanned electron beams can be deflected with respect to the direction of the electron beams generated by the electron accelerator by controlling a scanning current supplied to the scanning magnet. Another aspect of the present invention includes providing a method for controlling an irradiating device, which comprises an electron accelerator, a scanning box connected to the electron accelerator, and a scanning magnet for controlling electron beams generated by the electron accelerator, wherein the scanning box is provided with both a target and an electron beam exit window, the method comprising steps of: a) when the scanning magnet is not in operation, the electron beams impinge on the target to generate X-rays, so that the irradiating device outputs the X-rays; and b) when the scanning magnet is in operation, the scanned electron beams are deflected and pass through the electron beam exit window by supplying deflecting scanning current to the scanning magnet, so that irradiating device outputs the electron beams. The irradiating device capable of outputting both electron beams and X-rays of the present invention has many advantages. The configuration comprising an electron accelerator, a scanning box, a target and an electron beam exit window according to the invention integrates the functions of both the existing irradiating device outputting electron beams and those outputting X-rays. When the scanning magnet is in operation, the irradiating device outputs electron beams; and when the scanning magnet is not in operation, the irradiating device outputs X-rays. Therefore, an aspect of the device of the invention is to provide a system that is capable of outputting two radiation sources of electron beams and X-rays, so as to meet requirements for radiation-processing articles with different sizes. With such a device, much more applications are supported at substantially no additional cost. On the other hand, even when the scanning magnet fails in directing the electron beams, i.e., does not work, the electron beams will travel to impinge the target to generate X-rays, without any damage to the electron beam exit window. The safety of the system is improved, which assists in increasing the life and efficiency of the irradiating device of the present invention. These and other features, advantages and objects of the present invention will be further understood and appreciated by those skilled in the art by reference to the following specification, claims and appended drawings. Referring to FIG. 3, an irradiating device capable of outputting both electron beams and X-rays according to a first embodiment of the invention includes an electron linear accelerator 1, a scanning box 4 mounted on the electron linear accelerator 1 via a flange 2, and a scanning magnet 3 mounted at a position where the scanning box 4 is connected to the electron linear accelerator 1. The scanning box 4 is provided with a target 7, made of heavy metal materials such as tungsten or tungsten alloy, right in the outputting direction of the electron linear accelerator 1, and further provided with an electron beam exit window 10, made of metallic foil such as titanium, at a left or right side of the target 7. Although the scanning box is shown in the figure in a shape of triangle, its shape is not limited and may be any suitable shape. In the irradiating device capable of outputting both electron beams and X-rays of the invention, the scanning magnet 3 is scanning in a unidirectional manner, that is, the scanning current supplied thereto is always positive/negative. This scanning current may be obtained by superposing an original bidirectional scanning current with a positive or negative current. The target 7 is positioned right in front of the electron beams 5, while the electron beam exit window 10 is positioned right in front of the electron beams 6. The illustrated scanning box 4 is one with two different outputs from two outlets respectively. When the scanning magnet 3 is not in operation, the electron beams 5 travel in the original direction out of the accelerator, and impinge on the target 7 so as to generate X-rays, which are then output by the irradiating device. When the scanning magnet 3 is in operation, the electron beams are deflected from the direction of the electron beams 5 to be spread on one side of the beams 5, thereby forming pencil shaped electron beams 6. The electron beams 6 pass through the electron beam exit window 10 and are finally output by the irradiating device. A cooling fluid loop 9 is provided at bottom of the scanning box 4 for cooling the target 7 and the electron beam exit window 10. The cooling fluid loop 9 is externally connected to a cooling fluid system via an inlet 8 and an outlet 12. Referring to FIG. 4, in a second embodiment, the target 7 is provided inside the electron beam exit window 10, thus forming an inner target structure. When the scanning magnet 3 is not in operation, the electron beams 5 generated by the electron linear accelerator 1 impinge on the target 7 so as to generate X-rays, which are then output through the electron beam exit window. Here, since the electron beam exit window is very thin, for example tens of micrometers, the energy deposition of the X-rays on the electron beam exit window is so tiny that its effects can be neglected. In other words, the X-rays would not damage the electron beam exit window, which would not affect the X-rays either. When the scanning magnet 3 is in operation, the electron beams generated by the electron linear accelerator 1 are deflected from the original direction of the beams to be spread on one side of the beams 5, thereby forming a bundle of deflected electron beams 6. The electron beams 6 pass through the electron beam exit window 10 and are finally output by the irradiating device. While some embodiments of the invention have been described above, for the illustrative purpose only, it is to be understood that the invention is not limited to the details of the illustrated embodiments, but may be embodied with various changes, modifications or improvements, which may occur to those skilled in the art without departing from the spirit and scope of the invention. The above description is considered that of the preferred embodiments only. Modification of the invention will occur to those skilled in the art and to those who make or use the invention. Therefore, it is understood that the embodiments shown in the drawings and described above are merely for illustrative purposes and not intended to limit the scope of the invention, which is defined by the following claims as interpreted according to the principles of patent law, including the doctrine of equivalents.
050911430	abstract	The invention is directed to a natural circulation reactor having a reactor pressure vessel with a core housed therein, the core being disposed in such a location that a top portion of the core is submerged under coolant even in the event that any pipe connected to the reactor pressure vessel is broken and then a coolant level in the reactor pressure vessel is lowered due to flushing. This permits the reactor core to be submerged under coolant even in the event of breakage of any pipe connected to the reactor pressure vessel, ensuring to eliminate a possibility that the top portion of the reactor core is exposed temporarily during an intermediate period before actuation of an accumulated coolant injection system to start injecting of the coolant into the reactor pressure vessel after the end of flushing.
	abstract	Provided is a body, a method for manufacturing the body and a method of using of the body for nuclear shielding in a nuclear reactor. The body may include boron, iron, chromium, carbon and tungsten.
	summary
	claims	1. A multi-beam synchronous raster scanning lithography system comprising:a) a processor that generates electrical signals representing a desired exposure pattern at an output;b) a multi-beam source of exposing radiation that generates a plurality of exposure beam;c) a beam modulator having control inputs that are electrically connected to the output of the processor, the beam modulator receiving the electrical signals generated by the processor and modulating the plurality of exposing beams according to the desired exposure pattern;d) a beam deflector that deflects the plurality of exposure beams by a predetermined distance along a first axis, thereby exposing a plurality of pixels along the first axis with the desired exposure pattern; ande) a translation stage that supports a substrate to be exposed, the translation stage moving the substrate a predetermined distance along a second axis to position the substrate for a subsequent exposure of pixels along the first axis that results in a desired overlapping exposure dose profile, wherein a cycle time TC is inversely proportional to the square of a spot resolution Rmin and is independent of a positional grid of the multi-beam synchronous raster scanning lithography system. 2. The system of claim 1 wherein the multi-beam source of exposing radiation comprises a two-dimensional array multi-beam source. 3. The system of claim 1 wherein the multi-beam source of exposing radiation comprises a multi-electron beam source that generates a plurality of electron beams. 4. The system of claim 1 wherein the multi-beam source of exposing radiation comprises a field emission array comprising a plurality individually addressable electron emitters that emit a plurality of electron beams. 5. The system of claim 4 further comprising a secondary electron multiplication array that is positioned in a path of the plurality of electron beams emitted from the field emission array, the secondary electron multiplication array generating a plurality of focused electron beams. 6. The system of claim 4 wherein the beam deflector further comprises an electrostatic lens array that deflects the plurality of electron beams. 7. The system of claim 1 wherein the beam deflector comprises a control input that is electrically connected to the output of the processor, the processor generating a signal that instructs the beam deflector to deflect the plurality of exposure beams by a predetermined distance along a first axis that results in the desired overlapping exposure dose profile. 8. The system of claim 1 wherein the translation stage comprises a control input that is electrically connected to the output of the processor, the processor generating a signal that instructs the translation stage to move the substrate a distance along the second axis that results in the desired overlapping exposure dose profile. 9. The system of claim 1 wherein the predetermined distance along the second axis is less than a beam resolution of the system. 10. A method of synchronous raster scanning lithography, the method comprising:a) generating a plurality of exposure beams;b) modulating the plurality of exposure beams according to a first lithography pattern;c) deflecting the modulated plurality of exposure beams a predetermined distance along a first axis, thereby exposing a plurality of pixels at a first substrate position along the first axis with the first lithography pattern;d) translating the substrate a predetermined distance along a second axis to a second substrate position;e) modulating the plurality of exposure beams according to a second lithography pattern; andf) deflecting the modulated plurality of exposure beams a predetermined distance along the first axis, thereby exposing a plurality of pixels at the second substrate position along the first axis with the second lithography pattern, wherein at least one of the predetermined distance that the modulated plurality of exposure beams is deflected along the first axis and the predetermine distance that the substrate is translated along the second axis is chosen to achieve a desired total exposure dose profile for each of the plurality of pixels exposed along the first axis on the substrate and wherein a cycle time TC is inversely proportional to the square of a spot resolution Rmin and is independent of a positional grid. 11. The method of claim 10 wherein the generating a plurality of exposure beams comprises generating the plurality of exposure beams in two dimensions. 12. The method of claim 10 wherein the generating the plurality of exposing beams comprises generating a plurality of electron beams. 13. The method of claim 10 wherein the desired total exposure dose for each of the plurality of pixels exposed along the first axis on the substrate comprises an overlapping exposure dose. 14. The method of claim 10 wherein the modulated plurality of exposure beams is deflected by the predetermined distance along the first axis at a constant velocity. 15. The method of claim 10 wherein the substrate to be exposed is translated the predetermined distance along the second axis at a constant velocity. 16. The method of claim 10 wherein the second axis is perpendicular to the first axis. 17. The method of claim 10 wherein the desired total exposure dose for each of the plurality of pixels exposed along the first axis on the substrate comprises a blended dose profile. 18. The method of claim 10 further comprising adjusting at least one of the predetermined distance that the modulated plurality of beams is deflected along the first axis and the predetermine distance that the substrate is translated along the second axis to reduce effects of aberrations in the modulated plurality of beams. 19. The method of claim 10 further comprising generating instructions for modulating the plurality of beams according to the first and the second desired lithography pattern. 20. The method of claim 10 further comprising repeating the steps of deflecting the modulated plurality of exposure beams and translating the substrate until the desired lithographic pattern is exposed in a desired area of the substrate. 21. The method of claim 10 further comprising:a) translating the substrate a predetermined distance along the second axis to a third substrate position;b) modulating the plurality of exposure beams according to a third desired lithography pattern; andc) deflecting the modulated plurality of exposure beams a predetermined distance along the first axis, thereby exposing a plurality of pixels at the third substrate position along the first axis with the third desired lithography pattern,wherein at least one of the predetermined distance that the modulated plurality of exposure beams is deflected along the first axis and the predetermine distance that the substrate is translated along the second axis is chosen to achieve a desired total exposure dose for each of the plurality of pixels exposed along the first axis on the substrate. 22. A method of synchronous raster scanning lithography, the method comprising:a) generating a plurality of exposure beams that are modulated according to a first lithography pattern;b) deflecting the plurality of modulated exposure beams a predetermined distance along a first axis, thereby exposing a plurality of pixels at a first substrate position along the first axis;c) generating a second plurality of exposure beams that are modulated according to a second lithography pattern; andd) deflecting the plurality of modulated exposure beams a predetermined distance along the first axis, thereby exposing a plurality of pixels at a second substrate position along the first axis, wherein the exposures of the plurality of pixels at the first and the second substrate positions overlap so as to increase imaging throughput so that a cycle time TC is inversely proportional to the square of a spot resolution Rmin and is independent of a positional grid.
	claims	1. A filter system, comprising:a dedicated piping system configured to connect between an interior of a containment housing a nuclear reactor or a ventilation outlet of the containment and a spent fuel storage water pool outside the containment, the dedicated piping system configured to fluidly communicate any atmospheric effluent to be released from inside of the containment through the spent fuel storage water pool, the dedicated piping system comprising an outlet in a lower portion of the spent fuel storage water pool;one or more valves connected to the dedicated piping system, the one or more valves configured to control a release of the atmospheric effluent to be released;a chemical injection system configured to release a chemical into the spent fuel storage water pool to facilitate a reaction with the atmospheric effluent to be released to substantially neuter any deleterious environmental impact of the atmospheric effluent to be released, the chemical injection system comprising a chemical injection header;a control system connected to one or more of the chemical injection system and/or the one or more of the valves and configured to control the release of the chemical and/or the release of the atmospheric effluent; anda sparger connected to the outlet of the dedicated piping system, the chemical injection header supported above the sparger within the spent fuel storage water pool;wherein the outlet in the lower portion of the spent fuel storage water pool is configured to release the atmospheric effluent to the spent fuel storage water pool through the sparger; andwherein the chemical injection system is configured to release the chemical into the spent fuel storage water pool through the chemical injection header. 2. The filter system of claim 1 wherein the dedicated piping system includes a check valve configured to prevent spent fuel storage pool water from being drawn into the containment. 3. The filter system of claim 1 wherein the control system includes a manually actuated, remotely operated valve in the dedicated piping system, the manually actuated, remotely operated valve being configured to isolate the atmospheric effluent from the spent fuel storage water pool under normal operating conditions, unless actuated. 4. The filter system of claim 3 wherein, the manually actuated, remote operated valve is configured in the dedicated piping system to be in parallel with a passively operated valve structured to release the atmospheric effluent to the spent fuel storage water pool if a pressure is sensed within the containment in excess of a given pressure. 5. The filter system of claim 1 wherein the chemical injection header and the sparger are supported within the spent fuel storage water pool at an elevation below a level that is used to transfer fuel into and out of the spent fuel storage water pool. 6. A filter system, comprising:a piping system configured to couple a ventilation outlet of a containment housing a nuclear reactor and a spent fuel storage water pool outside the containment, the piping system configured to fluidly communicate atmospheric effluent from the containment to the spent fuel storage water pool, the piping system comprising an outlet in a lower portion of the spent fuel storage water pool;a valve system configured to control a release of the atmospheric effluent from the containment to the spent fuel storage water pool;a chemical injection system configured to release a chemical into the spent fuel storage water pool to facilitate a reaction with the atmospheric effluent, the chemical injection system comprising a chemical injection header;a control system configured to control:the release of the chemical from the chemical injection system; andthe release of the atmospheric effluent from containment to the spent fuel storage water pool; anda sparger connected to the outlet of the piping system, the chemical injection header supported above the sparger within the spent fuel storage water poolwherein the outlet in the lower portion of the spent fuel storage water pool is configured to release the atmospheric effluent to the spent fuel storage water pool through the sparger; andwherein the chemical injection system is configured to release the chemical into the spent fuel storage water pool through the chemical injection header. 7. The filter system of claim 6 wherein the chemical injection header and the sparger are supported within the spent fuel storage water pool at an elevation below a level that is used to transfer fuel into and out of the spent fuel storage water pool. 8. A filter system, comprising:a piping system configured to couple a ventilation outlet of a containment housing a nuclear reactor and a spent fuel storage water pool outside the containment, the piping system configured to fluidly communicate atmospheric effluent from the containment to the spent fuel storage water pool, the piping system comprising an outlet in a lower portion of the spent fuel storage water pool;a valve system configured to control a release of the atmospheric effluent from the containment to the spent fuel storage water pool;a chemical injection system configured to release a chemical into the spent fuel storage water pool, the chemical injection system comprising a chemical injection header;a control system configured to control at least one of:the release of the chemical from the chemical injection system; andthe release of the atmospheric effluent from containment to the spent fuel storage water pool via the valve system; anda sparger connected to the outlet of the piping system, the chemical injection header supported above the sparger within the spent fuel storage water poolwherein the outlet in the lower portion of the spent fuel storage water pool is configured to release the atmospheric effluent to the spent fuel storage water pool through the sparger; andwherein the chemical injection system is configured to release the chemical into the spent fuel storage water pool through the chemical injection header. 9. The filter system of claim 8 wherein the chemical injection header and the sparger are supported within the spent fuel storage water pool at an elevation below a level that is used to transfer fuel into and out of the spent fuel storage water pool.
	description	The following description of embodiments of the present invention refers to the accompanying drawings. Where appropriate, the same reference numbers in different drawings refer to the same or similar elements. In accordance with methods and systems consistent with the present invention, an improved processor performance instrumentation system is provided that allows a software tester to measure more performance indicators than there are hardware counters during a single execution of a tested program. The improved processor performance instrumentation system accomplishes this by xe2x80x9cmultiplexingxe2x80x9d performance indicators while executing the tested program. In using such an improved processor performance system, the user chooses the performance indicators to measure. While the tested software program is running, systems or methods consistent with the present invention select one or more of the performance indicators, and measure and record the selected performance indicator during a pre-determined time period. Such an improved processor performance system then selects another one or more performance indicators, which are and measured and recorded during the predetermined time period. Such an improved processor performance system selects new performance indicators until all the user""s chosen performance indicators are measured. After measuring all the performance indicators, such methods and systems start over and measures all the performance indicators. Such methods and systems repeat until the tested program stops executing. Thus, the performance indicators are multiplexed during the execution of the tested program. This allows the software tester to measure, in one pass of the tested program, a greater number of performance indicators than there are hardware counters. FIG. 1 depicts a data processing system suitable for use with methods and systems consistent with the present invention. Computer 118 includes a memory 102, a secondary storage device 104, a processor 105 such as a central processing unit (CPU), an input device 106, and an output device 108. Input device 106 may comprise a keyboard, a mouse, or both. Output device 108 may be a cathode ray tube (CRT) that can display a graphical user interface (GUI). Memory 102 stores an application 109 used to multiplex multiple performance indicators. Memory 102 also contains a data storage space 120 for storing data while multiplexing application 109 is running. Memory 102 also stores a program 122 that multiplexing application 109 tests. Processor 105 comprises a first performance counter 112 and a second a performance counter 114. First performance counter 112 and second performance counter 114 each are capable of measuring one performance indicator at a single time. In another embodiment consistent with the present invention, there are more than two performance counters. In yet another embodiment consistent with the present invention, there is only one performance counter. The hardware counters are xe2x80x9cPerformance Instrumentation Countersxe2x80x9d and they count xe2x80x9cperformance indicators.xe2x80x9d For example, in the ULTRASPARC(trademark) processor available from SUN MICROSYSTEMS(trademark), there are two PICs, each of which can measure twelve performance indicators, but one at a time. Table I provides the performance indicator variable name and description for the first PIC in an ULTRASPARC(trademark). Table II provides the performance indicator variable name and description for the second PIC in an ULTRASPARC(trademark). As evident from the tables, some performance indicators are common to both the first PIC and the second PIC. Some performance indicators are found only on one of the two PICs. This table and a description of Performance Instrumentation is found in Appendix B of the ULTRASPARC(trademark) User""s Manual from SUN MICROSYSTEMS(trademark), July, 1997, which is incorporated herein by reference. FIG. 2 depicts a flowchart of the steps performed by the multiplexing application 109 in accordance with methods and systems consistent with the present invention. The first step performed by the multiplexing application 109 is to allow the user to choose a plurality of performance indicators (step 202). The user may do this through a GUI with a keyboard or mouse, or both. A list of all possible performance indicators be may stored in a data storage space 120. FIG. 3 depicts a more detailed diagram of data storage space 120, including a table 302 that includes a list of all the possible performance indicators in a first column 304. Multiplexing program 109 flags user chosen performance indicators by inserting a xe2x80x9cTRUExe2x80x9d in a second column 306. Performance indicators not chosen by the user are flagged xe2x80x9cFALSExe2x80x9d in the second column 306. The user may choose one or more performance indicators, including a number that is larger than the number of hardware counters present. The user chosen performance indicators will be measured by multiplexing application 109 while tested program 122 runs. Performance indicators not chosen will be ignored by multiplexing application 109. After the user chooses performance indicators, multiplexing application 109 then performs the step of determining a measured time period for measuring the user chosen performance indicators (step 204). To do this, program 109 may use a number of different algorithms. In one embodiment, the algorithm is based on the bit length of performance counters 112, 114 and the speed of processor 105. For instance, the bit-length of the performance counter in the SUN ULTRASPARC(trademark) processor is thirty-two bits. This means that the performance counter can count from a minimum of zero to a maximum of 232xe2x88x921 occurrences of a performance indicator. Once the performance counter reaches 232xe2x88x921, it xe2x80x9cwraps aroundxe2x80x9d quietly and starts counting at zero again. Therefore, multiplexing application 109 reads and records the performance counter data before it resets to zero. In general, if xe2x80x9cnxe2x80x9d represents the bit length of the performance counter, the wrap around time in seconds is 2n divided by the processor speed in cycles per second. Because of possible xe2x80x9cinterrupts,xe2x80x9d the measured time period, in this embodiment, is set to 80% of the wrap around time. During step 204, multiplexing application 109 also initializes a recorded data table 402, which is shown in FIG. 4 that depicts a more detailed diagram of data storage space 120. Recorded data table 402 stores data accumulated while multiplexing application 109 is running so that application 109 can display results at a later time. Data table 402 may be stored in data storage space 120. Recorded data table 402 includes a list of all the possible performance indicators in a first column 404. A second column 406 stores the total number of occurrences of each performance indicator while each is measured. A third column 408 stores the total time each performance indicator is measured. In one embodiment, a fourth column 410 stores the total number of clock cycles that have occurred while each performance indicator is measured. During step 204, multiplexing application 109 sets all of these values in second column 406, third column 408, and fourth column 410 to zero. Multiplexing application 109 then executes the tested program 122 (step 206). Tested program 122 is the program whose performance is measured. Multiplexing application then flags all the chosen performance indicators as unmeasured (step 208). FIG. 3 depicts a more detailed diagram of data storage space 120, including table 302 that includes a list of whether a performance indicator has been measured. Multiplexing application 109 flags measured performance indicators by inserting a xe2x80x9cTRUExe2x80x9d in third column 308. Unmeasured performance indicators are flagged xe2x80x9cFALSE.xe2x80x9d Performance indicators that will not be measured at all because the user did not choose them are flagged xe2x80x9cNULL.xe2x80x9d Multiplexing application 109 then selects two unmeasured performance indicators from the user chosen performance indicators (step 210). In another embodiment, if there is only one performance counter then the application 109 would only select one unmeasured performance indicator. Likewise, in the embodiment where there are more than two performance counters, then multiplexing application 109 selects a number of performance indicators equal to the number of performance counters. Multiplexing application 109 then instructs processor 105 to initialize performance counters 112, 114 to zero (step 212), and instructs processor 105 to measure the selected performance indicators (step 214) that multiplexing application 109 selected in step 210. Tested program 122 continues to execute (step 216). When the measured time period expires or when the tested program ends, multiplexing application 109 reads the performance counter data (step 218). This data is summed to the value in second column 406 corresponding to the appropriate selected performance indicators. Likewise, multiplexing application 109 increments the total time in third column 408 by the time the performance indicators were measured. Multiplexing application 109 then flags the two selected performance indicators as measured (step 220) by inserting a xe2x80x9cTRUExe2x80x9d in third column 308 of table 302. For instance, in table 302, the first performance indicator and the second performance indicator are flagged as measured. The third, sixth, seventh, eighth, and ninth performance indicators are flagged as unmeasured. The fourth and fifth performance indicators are flagged xe2x80x9cNULLxe2x80x9d because the user did not choose them for measurement. If tested program 122 has not ended (step 222), then the multiplexing application 109 determines if all the performance indicators are flagged as measured (step 224). If all the performance indicators are flagged as measured, then multiplexing application 109 flags all the performance indicators as unmeasured (step 208) and continues to step 210 to gather more data. If there are unmeasured performance indicators (step 224), then multiplexing application 109 selects two unmeasured performance indicators from the user chosen performance indicators (step 210) and continues to step 212. In another embodiment, multiplexing application 109 selects one new unmeasured performance indicator and retains the other previously selected performance indicator. For instance, the retained performance indicator may be the clock cycle count. Therefore, multiplexing application 109 may record the number of clock cycles during each measured time period. Multiplexing application places this data in fourth column 410 of recorded data table 402. If tested program 122 ended (step 222), multiplexing application 109 displays the results (step 226), for example on display device 103. To display results, multiplexing program 109 displays recorded data table 402. After displaying the results, multiplexing application 109 ends. One skilled in the art will appreciate that numerous variations to this system exist. For example, the measured time period can be selected in any fashion that is prevents the counters from resetting and losing data. As another example, the data may be tabulated and displayed in any fashion. Although methods and systems consistent with the present invention have been described with reference to a preferred embodiment thereof, those skilled in the art knows various changes in form and detail which may be made without departing from the spirit and scope of the present invention as defined in the appended claims and their full scope of equivalents.
	summary
047028823	claims	1. In a fuel assembly having a top nozzle, a longitudinally extendign control rod guide thimble, and a spacer grid extending transversely about an upper regin of said fuel assembly an attaching structure for removably mounting the top nozzle on the fuel assembly comprising: first means for transferring compressive loads on said fuel assembly through said top nozzle directly to said guide thimble; second means for transferring tensive loads on said fuel assembly through said top nozzle and said spacer grid to said guide thimble. a bottom nozzle; a longitudinally extending control rod guide thimble having an upper end, and a lower end attached to said bottom nozzle; a plurality of transverse spacer grids and a top spacer grid axially space along said guide thimble for supporting an array of upstanding fuel rods; a top nozzle removably mounted on the upper end of the control rod guide thimble for permitting access to said fuel rods upon removal thereof; means attached to the upper end of the control rod guide thimble for transferring compressive loads on the fuel assembly through the top nozzle to the cnotrol rod guide thimble; means for transferring tensive loads on the fuel assembly through the top nozzle to the top spacer grid. providing means on the guide thimble for supporting compressive loads on the fuel assembly; providing means for resiliently latching the top spacer grid to the top nozzle: lowering the top nozzle onto the means for supporting compressive loads: and latching the top nozzle to the top spacer grid with said resilient latching means. holding said locking pin in a deflected position until said grid skirt extension aperture passes into alignment with said locking pin; and releasing said locking pin into said grid skirt extension aperture. withdrawing said locking pin from said grid skirt extension aperture; lifting said top nozzle from said fuel assembly to thereby access said plurality of upstanding fuel rods. deflecting said locking tang while lowering said top nozzle until said lock1ng tang engages said slot thereby latching said top nozzle to said grid skirt extension. inserting said tool into the aperture in said slot; deflecting said locking tang out of engagement with said slot by means of said tool; holding said locking tang out of engagement with said tang while lifting sald top nozzle by means of said tool. 2. The fuel assembly of claim 1 wherein said top nozzle includes a transversely extending adapter plate having an aperture therein, and wherein said guide thimble is sized ot clearance fit in siad aperture and said first means comprises a collar fixed to said guide thimble for transferring compessive loads through said adapter plate to said guide thimble. 3. The fuel assembly of claim 1 wherein said spacer grid has an upstanding skirt extension and said second means comprises a latch for securing said skirt extension to said top nozzle. 4. The fuel assembly of claim 2 wherein spacer grid has an upstanding skirt extension and said second means comprises a latch for securing said skirt extension to said top nozzle. 5. The fuel assembly of claim 3 wherein said top nozzle has a generally vertical sidewall and a tang extending generally parallel with respect to said sidewall and wherein said latch comprises a pin carried by said tang and a latching aperture formed in said grid skirt extension, said pin aligning with said latching aperture to latch said top nozzle to said grid skirt extension. 6. The fuel assembly of claim 5 wherein said sidewall includes an aperture adapted to be aligned with said latching aperture where said pin extends through said latching aperture and into said aperture. 7. The fuel assembly of claim 5 wherein said tang is formed of spring steel. 8. The fuel assembly of claim 3 wherein said top nozzle has a generally vertical sidewall and said latch comprises a slot formed in said sidewall and a complementary tang formed in said skirt extension. 9. The fuel assembly of claim 8 wherein said sidewall further includes an aperture in the region of said slot for passage of a tool and said tang further comprises an upstanding flange portion adapted to be engaged by said tool. 10. The fuel assembly of claim 8 wherein said spacer grid skirt extension is formed of spring steel and said tang is operable to assume a first stable position engaging said slot, and a second position resiliently deflected from said slot. 11. A fuel assembly for a nculear reaction comprising: 12. The fuel assembly of claim 11 wherein top nozzle further comprises a transversely extending adapter plate having an aperture sized for clearance fitting at least a portion of the upper end of said control rod guide thimble and wherein said means for transferring compressive loads comprises a load collar fixed to the control rod guide thimble adjacent the upper end thereof and radially dimentioned to bear against the adapter plate when said top nozzle is lowered onto said control rod guide thimble whereby the upper end of the control rod guide thimble is clearance fitted into said aperture. 13. The fuel assembly of claim 11 wherein said top nozzle has a sidewall and said top spacer grid further comprises a longitudinally extending grid skirt extension extending towards said sidewall, and means for resiliently latching said grid skirt extension to said sidewall. 14. The fuel assembly of claim 13 wherein said latching means comprises an aperture formed in said grid skirt extention and a resiliently biased locking pin attached to said top nozzle for capturing said grid skirt extension at said grid skirt extension aperture wherein tensive loads on said fuel assembly are transferred from said top nozzle to the top spacer grid through the grid skirt extension. 15. The fuel assembly of claim 13 wherein said grid skirt extension is formed of resilient material and said latching means comprises a tang formed at an end of said grid skirt extension, and a complementary slot formed in said sidewall for accepting said tang, said grid skirt extension being movable between a first stable position wherein said tang engages said slot and tensile loads on the fuel assembly are transferred from top nozzle to the top spacer grid through the grid skirt extension and a second resiliently deformed position wherein a clearance is formed between said tang and said slot and said top nozzle can be freely removed from said fuel assembly. 16. A method of assembling and disassembling a top nozzle of a fuel assembly having a guide thimble with an upper end, a top spacer grid, and a plurality of upstanding fuel rods, comprising the steps of: 17. The method of claim 16 wherein said means for supporting compressive loads comprises a load collar attached to the guide thimble adjacent the upper end thereof and wherein said top nozzle comprises an adapter plate having an aperture sized for clearance fitting the upper end of said guide thimble, and said step of lowering further comprises lowering said adapter plate aperture onto the upper end of the guide thimble until the adapter plate bears on said load collar. 18. The method of claim 16 wherein said top spacer grid includes a grid skirt extension having an aperture therein and said top nozzle includes a resiliently biased locking pin and said step of latching further comprises: 19. The method of claim 18 further comprising the steps of: 20. The method of claim 16 wherein said top spacer grid includes a grid skirt extension having a locking tang and said top nozzle includes a sidewall having a slot for receiving said locking tang and said step of latching comprises: 21. The method of claim 20 wherein said top nozzle sidewall further comprises an aperture through said slot for accepting a release and lift tool and further comprising the steps of:
	claims	1. A passively cooled storage module for spent nuclear fuel comprising:an elongated module body defining a top end, a bottom end, a sidewall, and an internal cavity extending between the ends along a longitudinal axis, the internal cavity being configured for holding a fuel storage canister;a plurality of cooling air inlet ducts spaced circumferentially apart around the body, the inlet ducts each forming a radially oriented air inlet passageway fluidly connecting ambient atmosphere with the internal cavity;each air inlet duct having an inlet end opening at an exterior surface of the sidewall and an outlet end opening at an interior surface of the sidewall adjoining the cavity;wherein the air inlet ducts each have a recurving configuration to draw cooling air radially inwards and initially upwards from ambient atmosphere, and then redirect the cooling air downwards through the air inlet duct into a lower part of the internal cavity of the module. 2. The module according to claim 1, wherein the inlet end openings of the air inlet ducts are located at a higher elevation than the outlet end openings. 3. The module according to claim 2, wherein the inlet end openings are located on a lower half of the module and spaced vertically apart from a baseplate affixed to the bottom end of the module by a distance of at least two feet. 4. The module according to claim 3, wherein the inlet end openings of each inlet air duct are located on the lower portion of the module at 25 percent or less than the height of the module. 5. The module according to claim 4, wherein the inlet end openings have a vertically staggered arrangement in which the elevation of the inlet end opening of each air inlet duct is at a different elevation than the inlet end opening of each adjacent inlet air duct. 6. The module according to claim 5, wherein the inlet end openings of the air inlet ducts are radially open directly to ambient atmosphere. 7. The module according to claim 6, wherein the air inlet ducts are formed of piping having a circular cross-sectional shape. 8. The module according to claim 2, wherein the outlet end openings of the air inlet ducts are arranged to discharge cooling air into the lowest-most part of the internal cavity through the sidewall and adjacent to a top surface of a baseplate attached to the bottom end of the module. 9. The module according to claim 8, wherein the outlet end openings of the air inlet ducts are disposed adjacent to the top surface of the base plate. 10. The module according to claim 9, further comprising a plurality of L-shaped support brackets mounted on the base plate, the support brackets being configured to both elevate the canister above the baseplate and center the canister radially in the internal cavity of the module to form a ventilation annulus between the canister and body of the module, and wherein the cooling air is radially discharged from each air inlet duct into a bottom of the annulus via the outlet end openings of the air inlet ducts and flows upwards alongside the canister in the annulus for dissipating heat emitted by the canister. 11. The module according to claim 2, wherein the module body further comprises a sidewall having a composite construction comprised of an inner shell, an outer shell, and a concrete liner disposed between the shells for radiation shielding. 12. The module according to claim 11, wherein each air inlet duct is embedded in the concrete liner. 13. The module according to claim 12, wherein the inner and outer shells of the module body are structurally tied together via a plurality of vertically oriented upper and lower radial shell interconnector plates welded to the shells and embedded in the concrete liner. 14. The module according to claim 2, wherein the air inlet ducts each have a mitered multi-angled configuration comprising a first linear section adjoining the inlet end opening and a second linear section adjoining the first linear section at a miter joint, the first and second linear sections being obliquely angled to each other and the longitudinal axis of the module. 15. The module according to claim 14, further comprising a perforated radiation shield disposed at the miter joint and oriented transversely the first and second linear sections. 16. The module according to claim 1, wherein each air inlet duct comprises an inlet section adjoining to the exterior inlet end opening which is obliquely angled in an upwards direction to the longitudinal axis of the module, an outlet section adjoining the outlet end opening, and an intermediate section extending between the inlet and outlet sections. 17. The module according to claim 1, wherein each inlet duct includes a radiation shielding member abuttingly attached to a top of the inlet duct, the radiation shielding member being configured to conform to the inlet duct. 18. The module according to claim 1, further comprising a lid detachably coupled to the top end of the module for enclosing the internal cavity, the lid being configured to form air outlet ducts at an interface between the lid and top end of the module oriented to radially discharge heated cooling air from the internal cavity of the module outwards to atmosphere. 19. A ventilated dry storage system for passive cooling of spent nuclear fuel, the system comprising:an elongated module defining a top end, a bottom end, and a sidewall extending between the ends defining an internal cavity extending along a longitudinal axis;the sidewall including an inner shell, an outer shell, a radiation shielding fill material disposed between the shells;a plurality of radially oriented interconnector plates embedded in the fill material and welded to the inner and outer shells to rigidly couple shells together;a base plate sealingly affixed to the bottom end of the module;a removable lid detachably coupled to the top end of the module;a fuel storage canister disposed in the internal cavity and containing heat-emitting spent nuclear fuel;a plurality of cooling air inlet ducts each forming a radially oriented air inlet passageway through the fill material of the module and configured to fluidly connect ambient atmosphere with the internal cavity;wherein each of the interconnector plates are disposed between adjacent ones of the air inlet ducts.

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