patent_number
stringlengths 0
9
| section
stringclasses 4
values | raw_text
stringlengths 0
954k
|
|---|---|---|
abstract | The present invention consists of a multilayer structure having at least one triad of layers where each of the three layers is a predetermined material. One of the materials is from a group including lanthanum, lanthanum oxide, or lanthanum-based alloys. A second material is disposed between the first material and a third material. The second material is from a group including carbon, silicon, boron, boron carbide or silicon carbide. The third material is from a group including boron or boron carbide. Alternatively, a fourth material is added to further strengthen and increase the water resistance of the multilayer structure. The fourth material is selected from a group including silicon, boron, boron carbide or silicon carbide. The fourth material is disposed between the third layer of multilayer period n and the first layer of multilayer period nxe2x88x921. |
|
summary | ||
051768771 | claims | 1. A nuclear fuel assembly for a nuclear reactor, having a plurality of vertically extending fuel rods arranged side by side in a square array and containing fissile material, said array having two adjacent first sides which, when the fuel assembly is in said reactor core, are next to a control rod region of said core and two adjacent second sides which, when the fuel assembly is in said reactor core, are next to a non-control rod region of the core, wherein, when said array as seen in horizontal cross section is divided into four regions a-d and excluding from any said regions the rods lying on the diagonal line joining opposite corners of the array at which a said first side meets a second side, said four regions being a: the rods in the row and column of said array adjoining said first sides; b: the rods lying between said region a and said diagonal line; c: the rods lying between said region d and said diagonal line, d: the rods in the row and column of said array adjoining said second sides and; a: the rods in the row and column of said array adjoining said first sides; b: the rods lying between said region a and said diagonal line; c: the rods lying between said region d and said diagonal line, d: the rods in the row and column of said array adjoining said second sides and; a: the rods in the row and column of said array adjoining said first sides; b: the rods lying between said region a and said diagonal line; d: the rods in the row and column of said array adjoining said second side and; c: the rods lying between said region d and said diagonal line, 2. A fuel assembly according to claim 1 wherein said average concentration of fissile material per fuel rod is higher in said region b than in said region c by an amount in the range of at least 5-10%. 3. A fuel assembly according to claim 1 wherein a plurality of said fuel rods contain burnable poison, there being more of said rods containing said burnable poison in said region c than in said region b. 4. A fuel assembly according to claim 1 having at least one water rod occupying a central position in said array symmetrical with respect to said diagonal line and with respect to the other diagonal line joining opposite corners of the array at which respectively said first sides meet each other and said second sides meet each other. 5. A fuel assembly according to claim 1 having at least one water rod located wholly within said region c, there being no water rod in said region b. 6. A fuel assembly according to claim 1 wherein the number of said fuel rods in said region b is larger than the number thereof in said region c. 7. A fuel assembly according to claim 1 wherein said fissile material is present in fuel pellets in said fuel rods, the average outer diameter of the fuel pellets being larger in said region b than in said region c. 8. A fuel assembly according to claim 1 wherein said fissile material is present in pellets in said fuel rods, and the average pellet density per fuel rod is higher in said region b than in said region c. 9. A fuel assembly according to claim 1 having at least one water rod located in said array, wherein at least one of (i) the number of water rods and (ii) the total flow passage area of said at least one water rod is larger in said region c than in said region b. 10. A fuel assembly according to claim 1 wherein said fissile material comprises at least one of U-235, Pu-239 and Pu-241. 11. A nuclear fuel assembly for a nuclear reactor, having a plurality of vertically extending fuel rods arranged side by side in a square array and containing fissile material, said array having two adjacent first sides which, when the fuel assembly is in said reactor core, are next to a control rod region of said core and two adjacent second sides which, when the fuel assembly is in said reactor core, are next to a non-control rod region of the core, wherein, when said array as seen in horizontal cross section is divided into four regions a-d and excluding from any said regions the rods lying on the diagonal line joining opposite corners of the array at which a said first side meets a said second side, said four regions being 12. A nuclear fuel assembly for a nuclear reactor, having a plurality of vertically extending fuel rods arranged side by side in a square array and containing fissile material, said array having two adjacent first sides which, when the fuel assembly is in said reactor core, are next to a control rod region of said core and two adjacent second sides which, when the fuel assembly is in said reactor core, are next to a non-control rod region of the core, wherein, when said array as seen in horizontal cross section is divided into four regions a-d and excluding from any said regions the rods lying on the diagonal line joining opposite corners of the array at which a said first side meets a said second side, said four regions being 13. A nuclear reactor core containing at least one fuel assembly according to claim 1. 14. A nuclear reactor core containing at least one nuclear fuel assembly according to claim 1, said core having a first water gap constituting a control rod region adjacent said two first sides of said fuel rod assembly and a second water gap comprising a non-control rod region adjacent said two second sides of said fuel assembly, said first water gap being wider than said second water gap. 15. A nuclear reactor core containing at least one nuclear fuel assembly according to claim 11. 16. A nuclear reactor core containing at least one nuclear fuel assembly according to claim 11 said core having a first water gap constituting a control rod region adjacent said two first sides of said fuel rod assembly and a second water gap comprising a non-control rod region adjacent said two second sides of said fuel assembly, said first water gap being wider than said second water gap. 17. A nuclear reactor core containing at least one nuclear fuel assembly according to claim 12. 18. A nuclear reactor core containing at least one nuclear fuel assembly according to claim 12 said core having a first water gap constituting a control rod region adjacent said two first sides of said fuel rod assembly and a second water gap comprising a non-control rod region adjacent said two second sides of said fuel assembly, said first water gap being wider than said second water gap. |
051006111 | description | In FIGS. 1-3, 1 designates a fuel channel with substantially square cross section. The fuel channel surrounds, with no significant play, an upper square portion of a bottom part 2 with a circular, downwardly-facing inlet opening 3 for cooling water and moderator water. In addition to supporting the fuel channel 1, the bottom part 2 also supports a supporting plate 4. At the lower part the fuel channel 1 has a relatively thick wall portion which is fixed to the bottom part 2 and the supporting plate 4 by means of a plurality of horizontal bolts, indicated by means of dash-dotted lines 5. By means of a hollow support member 7 of cruciform cross section, the fuel channel 1 is divided into four vertical tubular parts 6 having at least substantially square cross section. The support member 7 is welded to the four walls 1a, 1b, 1c and 1d of the fuel channel 1 and has four hollow wings 8. The central channel formed by the support member is designated 32 and is connected at its lower part to an inlet tube 9 for moderator water. Each tubular part 6 contains a bundle 25 of twenty-five fuel rods 10. The rods are arranged in a symmetrical lattice in five rows each containing five rods. Each rod is included in two rows perpendicular to each other. Each bundle is arranged with a bottom tie plate 11, a top tie plate 12 and a plurality of spacers 13. A fuel rod bundle 25 with a bottom tie plate 11, a top tie plate 12, spacers 13 and a casing part 6 forms a unit which in this application is referred to as a fuel assembly, whereas the device comprising four such fuel assemblies shown in FIGS. 1-3 is referred to as a composed fuel assembly. In the composed fuel assembly the four bottom tie plates 11 are supported by the supporting plate 4 and are partially each inserted into a corresponding square hole 14 therein. In each fuel assembly at least one of the fuel rods is provided with relatively long, threaded end plugs 33 and 34 of solid cladding material, the lower end plug 33 being passed through the bottom tie plate 11 and provided with a nut 15 and the upper end plug 34 being passed through the top tie plate 12 and provided with a nut 16. In the embodiment shown the centre rod 26 is formed in this way. This rod also serves as spacer holder rod. An upper end portion of the fuel channel 1 surrounds a cruciform lifting plate 17 with four horizontal arms 18, 19, 20 and 21, which extend from a common central portion. At the outer end each arm has an arrowhead-like portion 22, which in each respective corner of the fuel channel 1 makes contact with the inner wall surface of the fuel channel 1. A lifting handle 23 is fixed to the arms 20 and 21. The lifting plate 17 and the handle 23 together form a steel lifting member cast in one piece. The lifting plate 17 is fixed to the support member 7 by inserting four vertical bars 28 into respective wings 8 of the support member 7 and welding them thereto. At the upper end each bar 28 has a vertical, bolt-like portion 29 which is passed with a play through a corresponding hole in the mid-portion of the lifting plate 17 and provided with a nut 30. As will be clear from the figures, the fuel channel 1 is provided with indentations 31, intermittently arranged in the longitudinal direction, against which the support member 7 is welded. In FIG. 4, which shows three adjacently positioned fuel assemblies for a pressurized water reactor, the vertical fuel rods are designated 42, the rectangular top tie plates 43, the rectangular bottom tie plates 44 and the spacer members by means of which the fuel rods are positioned are designated 45. Guide tube members 46 for control rod pins are fixed at their upper ends to the top tie plates 43 and at the lower ends to the bottom tie plates 44. In addition, they are fixed to the spacer members 45. The bottom tie plate according to FIGS. 6 and 7 have through-holes 50 for conducting water through the bottom tie plate. The holes have parts 50a and 50b, the centre lines of which are displaced in relation to each other. One part 50a of a through-hole on the inlet side 51 of the plate for the water is common to several parts 50b of through-holes on the outlet side 52 of the plate. One part 50a of a through-hole on the inlet side 51 of the bottom tie plate has a larger cross section than one part 50b of a through-hole on the outlet side 52 of the bottom tie plate. Parts 50a of through-holes are in the exemplified case--but need not be so--arranged in open communication with one or more edge sides 53 by way of transverse channels 54 which at the edge have orifices 55 which are arranged in open communication with the spaces between the fuel rods in the same fuel assembly for a boiling water reactor or in the same and adjacent fuel assemblies in a pressurized water reactor. The open communication is achieved by means of recesses 56 in the edge side 53 of the bottom tie plate. When the bottom tie plate according to the FIGS. 6 and 7 is used in a fuel assembly for a boiling water reactor, for example in the fuel assembly shown in FIG. 1, the surface 57 makes contact with the supporting plate 4 and the edge side 53 makes contact with the casing part 1, which means that the water which has passed through the orifices 55 flows up to the space between the fuel rods 10. The water which passes through the holes 50a and 50b also flows up to the space between the fuel rods 10. In FIG. 1--for reasons of space--the holes 50a and 50b are drawn schematically as holes of a conventional kind. The holes for the end plugs of the fuel rods in the bottom tie plate are designated 59. When the bottom tie plate according to FIGS. 6 and 7 is used in a fuel assembly for a pressurized water reactor, for example the fuel assembly shown in FIG. 4, the bottom tie plate has no holes (59) for fuel rods and the edge side 53 makes contact with an edge side in another fuel assembly of the same kind. Nor do the recesses 56 extend all the way down to the lower edge of the edge side 53 but two adjacent bottom tie plates make contact with each other along their entire horizontal extension along an edge at the bottom of each edge side. The water which has passed through the orifices 55 thereby flows via the recesses 56 up to the spaces between the fuel rods 42 above the bottom tie plate and in spaces between adjacent fuel assemblies since there are no partitions between the fuel assemblies. Parts 50a of through-holes are also arranged in the illustrated case in open communication with each other via transverse channels 58. The embodiment of the bottom tie plate according to FIGS. 8 and 9 differs from the embodiment according to FIGS. 6 and 7 in that it has several transverse channels. In addition to transverse channels 54 which connect parts 50a of through-holes with edge sides 53 and transverse channels 58 which connect parts 50a with each other, there are transverse channels 60 which connect parts 50b of through-holes with each other and with edge sides 53. The bottom tie plate according to FIGS. 8 and 9 may be used in the same way as the bottom tie plate according to FIGS. 6 and 7, i.e. as such in a fuel assembly according to FIG. 1 and in modified form without holes (59) for fuel rods and with recesses (56) which are shut off at the bottom in a pressurized water reactor. In the bottom tie plates according to FIGS. 10-12, the inlet side for the water is designated 61 and the outlet side, facing the fuel rods, for the water is designated 62. The through-holes in the bottom tie plate according to FIG. 10 have parts 70a and 70b, the centre lines of which make an angle with each other. In the embodiment according to FIG. 11 the through-holes are arranged with a pocket 71 for capturing debris. In the bottom tie plate according to FIG. 12, a part which extends from one side, for example a part 70a extending from the side 61, is arranged in open communication with two or three parts which extend from the other side of the bottom tie plate, i.e. two or three parts 70b extending from the side 62. The bottom tie plate is also provided with pockets 71. A bottom tie plate of the kind illustrated in FIG. 12 has an exceedingly large open inner volume. The bottom tie plates according to FIGS. 10-12 are used in the illustrated embodiment in a fuel assembly according to FIG. 4. When using it in a fuel assembly according to FIG. 1, it is provided with holes corresponding to the holes 59 in the bottom tie plates according to FIGS. 6-9, for end plugs for fuel rods 10. Such holes are not visible in the shown cross section. It may be suitable to combine the use of the described bottom tie plates with the use of a separate strainer means with a low flow resistance to water to ensure that objects of debris with different shapes are captured and prevented from entering sensitive parts of the fuel assembly. The use of such a separate strainer means in the form of a strainer plate 80 is shown in dashed lines in FIG. 1. The strainer plate may, for example, be fixed to the end plug 33 which is extended with an extra nut 81. |
summary | ||
055526128 | abstract | A transport container for transporting a radiation shield member which is light and capable of preventing a radiation shield member from popping out from the transport container by accident. The container has a favorable operability, and is capable of easily separating a reusable portion and a portion other than the reusable portion. The transport container includes a cup-shaped sheath container, a radiation shield container accommodated in the sheath container and having concaves recesses, formed thereon, engaging wing-shaped holding members of the radiation shield member, a holding frame, mounted on the sheath container, for preventing the radiation shield container and the radiation shield member from being removed from the transport container, and a cover for closing the opening end of the sheath container. |
062884018 | summary | FIELD OF THE INVENTION The present invention relates to a field emission source used, for example, in an electron beam microcolumn, and in particular to the electrostatic alignment of a charged particle beam. BACKGROUND Miniature electron beam microcolumns ("microcolumns") are based on microfabricated electron "optical" components and field emission sources operating under principles similar to scanning tunneling microscope ("STM") aided alignment principles. Field emission sources are bright electron sources that are very small, making them ideal for use in microcolumns. One type of field emission source is a Schottky emitter, such as the type discussed in "Miniature Schottky Electron Source," H. S. Kim et al., Journal of Vacuum Science Technology Bulletin 13(6), pp. 2468-72, November/December 1995 incorporated herein by reference. For additional field emission sources and for information relating to microcolumns in general, see the following publications and patents: "Experimental Evaluation of a 20.times.20 mm Footprint Microcolumn," by E. Kratschmer et al., Journal of Vacuum Science Technology Bulletin 14(6), pp. 3792-96, November/December 1996; "Electron Beam Technology-SEM to Microcolumn," by T. H. P. Chang et al., Microelectronic Engineering 32, pp. 113-130, 1996; "Electron-Beam Microcolumns for Lithography and Related Applications," by T. H. P. Chang et al., Journal of Vacuum Science Technology Bulletin 14(6), pp. 3774-81, November/December 1996; "Electron Beam Microcolumn Technology And Applications," by T. H. P. Chang et al., Electron-Beam Sources and Charged-Particle Optics, SPIE Vol. 2522, pp. 4-12, 1995; "Lens and Deflector Design for Microcolumns," by M. G. R. Thomson and T. H. P. Chang, Journal of Vacuum Science Technology Bulletin 13(6), pp. 2445-49, November/December 1995; U.S. Pat. No. 5,122,663 to Chang et al.; and U.S. Pat. No. 5,155,412 to Chang et al., all of which are incorporated herein by reference. FIG. 1 shows a schematic cross sectional view of a conventional field emission source 10, which includes an electron emitter 12 and an extraction electrode 14. The electron emitter 12 is a Schottky emitter with a tungsten tip 16 serving as a cathode, which is spot welded on a filament 18. The filament 18 is mounted on two rods 20, which are held by a base 22, and is surrounded by a suppressor cap 24. The extraction electrode 14 defines a center aperture 15. The aperture 15 and following (downstream) lenses (not shown) in the microcolumn define the optical axis 26 for the field emission source 10. By applying a voltage Vc to the tip 16 and a voltage Ve to the extraction electrode 14, a resulting electric field causes the emission of electrons from tip 16. A voltage Vs applied to the suppressor cap 24 suppresses undesired thermionic electrons. An important consideration in the field emission source 10 is that the electron emitter 12 is aligned with the optical axis 26. Because the diameter of aperture 15 is typically 1-2 .mu.m (micrometers), a small misalignment, e.g., 1 .mu.m, will result in a large off-axis aberration and an undesirable increase in the total spot size. Thus, a small misalignment can severely deteriorate the performance of a microcolumn. Conventionally, to ensure proper alignment, the electron emitter 12 is mechanically aligned with the optical axis 26. Thus, electron emitter 12 is physically moved, as indicated by arrows 28, by the use of, e.g., alignment screws, a micrometer x-y stage, a piezoelectric stage, or a scanning tunneling microscope (STM) like positioner to align position electron emitter 12 with optical axis 26. Unfortunately, mechanical alignment is difficult to achieve and is difficult to maintain over extended periods of time due to drift problems. Thus, there is a need for a field emission source that can be easily aligned with the optical axis. SUMMARY A field emission source in accordance with the present invention produces a charged particle beam that is electrostatically aligned with the optical axis. The field emission source includes a charged particle emitter, such as a Schottky or cold-field emitter. Centering electrodes define an aperture through which a beam of charged particles from the emitter passes and which is approximately centered on the optical axis. The centering electrodes provide an electrostatic deflection field near the optical axis that aligns the beam of charged particles with the optical axis, i.e., the axis of the electron beam passes through the center of the next lens down stream. Thus the emitter need not be precisely aligned mechanically with the optical axis. The center electrodes may be, for example, a quadrupole (or higher multipole) arrangement of electrodes placed between the emitter and an extraction electrode. By applying centering potentials of equal amplitude and opposite polarity on opposing elements of the centering electrodes, an electrostatic deflection field is created near the optical axis. The electrostatic deflection field aligns the charged particle beam with the optical axis thereby obviating the need to mechanically align the emitter with the optical axis. A second set of centering electrodes may be used to further deflect the charged particle beam and to ensure that the charged particle beam is approximately parallel with the optical axis. The centering electrodes may be integrally formed on the extraction electrode with an insulating layer between the extraction electrode and the centering electrodes and between the first set of centering electrodes and the second set of centering electrodes if a second set is used. In another embodiment, the extraction electrode is split into a quadrupole (or higher multipole) arrangement. The extraction potential and the centering potentials are then superimposed. |
abstract | An X-ray illumination optical system includes a reflection type integrator, having cylindrical surfaces, for reflecting an X-ray beam, and a concave mirror, including a rotational parabolic surface mirror, for reflecting the X-ray beam reflected by the integrator and for illuminating an object with the X-ray beam. |
|
052681289 | claims | 1. A method for removing contaminants comprising heavy metals and radioactive compounds, singly and in combination, from a process stream used in removing said contaminants from contaminated particulate material, said method including the steps of: washing said contaminated particulate material with said process stream, thereby removing said contaminants from said particulate material, said process stream containing an oxidizing agent selected from the group Cl.sub.2, ClO.sub.2 and O.sub.3 and/or a reducing agent, said oxidizing or reducing agent rendering at least one of said contaminants more soluble in said process stream; separating clean particulate material from said process stream; treating said process stream with a precipitant to remove said contaminant rendered more soluble in said process stream by said oxidizing or reducing agent; and recycling said process stream for washing additional contaminated particulate material. washing said contaminated particulate material with said process stream, thereby removing said contaminates from said particulate material; separating clean particulate material from said process stream, while retaining said heavy metals in solution in said process stream; adjusting the pH of said process stream to 8-10; adding a precipitant to precipitate said heavy metals from said process stream; removing said heavy metals from said process stream; and recycling said leachate and/or said surfactant for washing additional contaminated particulate matter. means washing said contaminated particulate material with said process stream, thereby removing said contaminates from said particulate material; means introducing an oxidizing and/or reducing agent to said process stream, said oxidizing agent increasing the solubility of said heavy metals in said process stream; means separating clean particulate material from said process stream following removal of said contaminants from said particulate material by said process stream; means treating said process stream with a heavy metal precipitant following removal of said clean particulate material therefrom; means removing precipitated heavy metal from said process stream following precipitation of said heavy metal; means treating said process stream after removal of said heavy metal therefrom with makeup oxidizing and/or reducing agent; means recycling said process stream to said means washing said contaminated particulate material. 2. The method of claim 1 wherein said heavy metals are selected from the group Hg, U, Pb, Ag, As, Cd, Cr, Cu, Ra, Th. 3. The method of claim 1 wherein said process stream is water-based. 4. The method of claim 3 wherein said process stream further includes a leachate for the removal of said contaminants from said particulate matter. 5. The method of claim 3 wherein said process stream further includes a surfactant for the removal of said contaminants from said particulate matter. 6. The method of claim 5 wherein said surfactants are removed from said process stream following washing of said particulate matter. 7. The method of claim 6 wherein said surfactants are precipitated from said process stream. 8. The method of claim 7 wherein CaCl.sub.2 is used for precipitating said surfactants. 9. The method of claim 8 wherein said contaminant rendered more soluble comprises said heavy metals. 10. The method of claim 8 wherein following removal of said clean particulate matter, said heavy metals are precipitated from said process stream with a precipitant. 11. The method of claim 10 wherein said precipitant comprises Na.sub.2 SiO.sub.3. 12. The method of claim 11 wherein fines are recovered from said process stream prior to precipitation of said heavy metals. 13. The method of claim 12 wherein following removal of said fines said process stream is pH adjusted to about 8-10 to assist in precipitation and removal of said heavy metals from said process stream. 14. The method of claim 13 wherein said precipitated heavy metals are used as a feedstock for mining or smelting operations to recover said heavy metals. 15. The method of claim 13 wherein said precipitated heavy metals are mixed with a fixative material for disposal. 16. The method of claim 11 wherein fines are recovered with said precipitated heavy metals. 17. The method of claim 16 wherein said process stream is pH adjusted to 8-10 to assist in precipitation of said heavy metals. 18. The method of claim 17 wherein said precipitated heavy metals are used as a feedstock for mining or smelting operations to recover said heavy metals. 19. The method of claim 17 wherein said precipitated heavy metals and fines are mixed with a fixative material for disposal. 20. The method of claim 8 wherein said heavy metals are precipitated and removed from said process stream and said process stream is recycled for washing said particulate matter. 21. The method of claim 8 wherein additional oxidizing or reducing agent is added to said process stream following removal of said heavy metals. 22. The method of claim 21 wherein said contaminated particulate material further includes contaminants that are organic compounds and said additional oxidizing or reducing agent destroys said organic compounds. 23. The method of claim 22 wherein said process stream includes an aliphatic surfactant which is recycled for washing said contaminated particulate matter. 24. The method of claim 22 wherein said surfactants are removed from said process stream prior to treating said stream with a heavy metal precipitant and said surfactants are aliphatic when a reducing agent is used and are non-aliphatic when an oxidizing agent is used. 25. The method of claim 1 wherein said reducing agent is H.sub.2 and said oxidizing agent is Cl.sub.2. 26. The method of claim 1 wherein prior to treating said process stream with said precipitant fines are removed from said process stream. 27. A method for removing contaminants comprising heavy metals, radioactive compounds and organic compounds, singly and in combination, from a water-based process stream used in removing said contaminants from contaminated particulate material, said process stream containing a leachate and/or a surfactant, said process stream further including an oxidizing and/or a reducing agent, said oxidizing agent increasing the solubility of said heavy metal in said process stream, said method including the steps of: 28. The method of claim 27 wherein said reducing agent is H.sub.2 and said oxidizing agent is Cl.sub.2. 29. The method of claim 27 wherein said heavy metals are selected from the group Hg, U, Pb, Ag, As, Cd, Cr, Cu, Ra, Th. 30. The method of claim 27 wherein said precipitant comprises Na.sub.2 SiO.sub.3. 31. The method of claim 27 wherein said oxidizing agent comprises Cl.sub.2, fines are removed from said process stream prior to precipitation of said heavy metals and prior to adjusting said pH, said precipitated heavy metals are removed from said process stream yielding a recycle process stream, and makeup Cl.sub.2 is added to said recycle process stream, which is recycled for washing said contaminated particulate material. 32. The method of claim 27 wherein said oxidizing agent comprises Cl.sub.2, said precipitant flocculates any metal hydroxides and fines, precipitated heavy metals, flocculated metal hydroxides and fines are removed from the process stream, yielding a recycle stream, and makeup Cl.sub.2 is added to said recycle process stream, which is recycled for washing said contaminated particulate material. 33. The method of claim 27 wherein said oxidizing agent is used and comprises Cl.sub.2, and organic contaminants are precipitated with CaCl.sub.2 prior to precipitation of said heavy metals, and said precipitated heavy metals, organics and fines are removed from the process stream, yielding a recycle process stream, which is recycled for washing said contaminated particulate material. 34. The method of claim 27 wherein said oxidizing agent comprises Cl.sub.2 and, wherein prior to adding said heavy metal precipitant, organic contaminants are precipitated with CaCl.sub.2 and said precipitated organic contaminants and fines are removed, after which said heavy metal precipitant is added, said heavy metals are precipitated and removed from said process stream, yielding a recycle process stream, which is recycled for washing said contaminated particulate matter. 35. An apparatus for removing contaminants comprising heavy metals, organics and radioactive compounds, singly and in combination, from a process stream used in removing said contaminants from contaminated particulate material, said apparatus comprising: |
claims | 1. A radioactive substance container comprisinga forged bottomed container made of a forgeable material, the bottomed container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, the bottomed container being made by hot-dilating a metal billet in a container for forming, wherein sections of inner and outer circumferences of the thick bottomed container perpendicular to an axial direction of the thick bottomed container are octagonal. 2. The radioactive substance container according to claim 1, wherein the inner circumference of the thick bottomed container vertical to the axial direction of the bottomed container has an irregular octagonal shape which is modified from rectangular shape by chamfering four corners of the rectangular shape. 3. A radioactive substance container comprisinga forged bottomed container made of a forgeable material, the bottomed container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, the bottomed container being made by hot-dilating a metal billet in a container for forming, wherein inner circumference of the bottomed container perpendicular to an axial direction of the bottomed container is polygonal, wherein the polygonal is defined as a modified shape of rectangular by shaping each of the four corners of the rectangular shape having at least a step. 4. The radioactive substance container according to claim 1, wherein an outer diameter of the bottomed container is not less than 1000 mm to not more than 3000 mm and its thickness is not less than 150 mm to not more than 300 mm. 5. The radioactive substance container according to claim 1, wherein a spot facing section is further formed integrally with the bottom section at the time of forming the bottomed container. 6. The radioactive substance container according to claim 1, wherein a flange is further provided integrally with the body section of the bottomed container. 7. A radioactive substance container comprising:a forged bottomed container made of a forgeable material, the bottomed container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, the bottomed container being made by hot dilating a metal billet, wherein the metal billet is hot-dilated in a container for forming one end of the metal billet is left not hot-dilated so as to be a bottom section, wherein sections of inner and outer circumferences of the bottomed container perpendicular to an axial direction of the bottomed container are octagonal. 8. The radioactive substance container according to claim 7, wherein an outer diameter of the bottomed container is not less than 1000 mm to not more than 3000 mm and its thickness is not less than 150 mm to not more than 300 mm. 9. The radioactive substance container according to claim 7, wherein a spot facing section is further formed integrally with the bottom section at the time of forming the bottomed container. 10. The radioactive substance container according to claim 7, wherein a flange is further provided integrally with the body section of the bottomed container. 11. The radioactive substance container according to claim 7, wherein at least any one of an external section and an internal section of the bottomed container vertical to the axial direction is polygonal. 12. A radioactive substance container comprising:a bottomed container for storing a basket for used nuclear fuel aggregate, whereinthe bottomed container is forged and made of a forgeable material having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, wherein the bottomed container is made by hot dilation forming in a container for forming, wherein sections of inner and outer circumferences of the bottomed container perpendicular to an axial direction of the bottomed container are octagonal. 13. The radioactive substance container according to claim 12, wherein a spot facing section is further formed integrally with the bottom section at the time of forming the bottomed container. 14. The radioactive substance container according to claim 12, wherein a flange is further provided integrally with the body section of the bottomed container. 15. The radioactive substance container according to claim 12, wherein at least any one of sections of inner and outer circumferences of the bottomed container vertical to the axial direction is octagonal. 16. A radioactive substance container comprisinga forged bottomed container made of a forgeable material, the bottomed container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, wherein a dosage equivalent factor of γ rays on an outer wall surface of a substantially center portion of a side surface of the body is not more than 200 μSv/h in the case where radioactive substance is contained in the bottomed container, the bottomed container being made by hot dilation forming in a container for forming, wherein sections of inner and outer circumferences of the bottomed container perpendicular to an axial direction of the bottomed container are octagonal. 17. The radioactive substance container according to claim 16, wherein a spot facing section is further formed integrally with the bottom section at the time of forming the bottomed container. 18. The radioactive substance container according to claim 16, wherein a flange is further provided integrally with the body section of the bottomed container. 19. A container made of a forgeable material, the container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, wherein a metal billet is hot-dilated in a container for forming and a thick bottomed container is obtained, wherein sections of inner and outer circumferences of the thick bottomed container perpendicular to an axial direction of the thick bottomed container are octagonal. 20. The container according to claim 19, wherein an outer diameter of the bottomed container is not less than 200 mm to not more than 4000 mm, and a thickness is not less than 20 mm to not more than 400 mm. 21. The container according to claim 19, wherein the bottomed container is constituted so that at least any one of an external section and an internal section of the bottomed container vertical to the axial direction is octagonal. 22. A container made of a forgeable material, the container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, wherein whena metal billet is hot-dilated in a container for forming and the body section is worked,one end of the metal billet is left not hot-dilated so as to be the bottom section, wherein sections of inner and outer circumferences of the container perpendicular to an axial direction of the container are octagonal. 23. The container according to claim 22, wherein an outer diameter of the bottomed container is not less than 200 mm to not more than 4000 mm, and a thickness is not less than 20 mm to not more than 400 mm. 24. A container made of a forgeable material, the container having a bottom section and a body section configured integrally such that the bottom section and the body section have continuous metal flow, wherein a metal billet, where at least a section vertical to an axial direction on a pressing forward side is octagonal, is set into a container for forming, and a boring punch is pushed into the metal billet and the metal billet is hot-dilated to be formed and into a bottomed container where the bottom section and the body section are integral, wherein the bottomed container is constituted so that an external section of the bottomed container perpendicular to the axial direction is octagonal. |
|
summary | ||
063273225 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a device for transporting poison rod assemblies between fuel assemblies in a nuclear fuel storage facility and, more specifically, to a device for transferring a poison rod assembly which uses an overhead crane as its only lifting mechanism. 2. Description of the Prior Art Fuel for a nuclear reactor used to create steam and, ultimately, electricity, generally is in the form of fuel rods containing a fissile material. When fuel rods are being stored, the fuel rods are typically supported in nuclear fuel assemblies arranged as spaced parallel arrays. Fuel assemblies are stored in racks in a protective medium, such as water containing boric acid. In addition to the fuel rods, poison rods are disbursed throughout the fuel assemblies to control the fission process. Poison rods generally include a plurality of elongated rods, each containing a neutron absorbent material, which fit in longitudinal openings, or thimbles, defined in the fuel assemblies. The top end of each poison rod is attached to a web thereby forming a poison rod assembly. A T-shaped bar is affixed to the top of the web, creating an easily accessible handle for lifting the poison rod assembly so that it may be transferred from one fuel assembly to another. As shown in Hornak et al., U.S. Pat. No. 5,325,408, prior art transfer devices use a winch located at the top of the transfer device to lift the poison rod assembly. Movement of the transfer device itself between fuel cells, is accomplished by a gantry crane in the fuel cell storage facility. Thus, prior art lifting devices consist of at least two lifting means, one to lift the transfer device itself, and one to lift the poison rod assembly within the transfer device. Lifting of a poison rod assembly could be performed more efficiently by the crane used to move the transfer device. Additionally, elimination of the winch would reduce the cost of the transfer device and eliminate unnecessary parts that are subject to mechanical failure. Therefore, there is a need for a poison rod transfer device that does not require the use of a winch. There is a further need for a poison rod transfer device that uses a single lifting mechanism to remove the poison rod assembly from a fuel cell and transfer it to another fuel cell. SUMMARY OF THE INVENTION The present invention satisfies the above referenced needs and others by providing a poison rod assembly transfer device having a poison rod assembly lifting device which may be actuated by the overhead crane used to move the transfer device. Lifting of the poison rod assembly is accomplished by a lifting assembly coupled to the overhead crane. The lifting assembly consists of an elongated outer member, an inner member, a gripper, assembly and an interlock device. The crane is attached to the top of the inner member. The gripper assembly is connected to the bottom of the inner member. The inner member is slidably disposed within the elongated outer member. The elongated outer member is seated on the fuel cell from which the poison rod assembly will be lifted. When the interlock device is not engaged, raising or lowering the crane will slide the inner member within the elongated outer member between an upper locked position and a lower locked position. When the interlock is engaged, the inner and elongated outer members are coupled so that the inner member cannot slide within the elongated outer member and raising the crane will lift the entire transfer device. Thus, when the interlock is not engaged the crane may lift a poison rod assembly out of the fuel cell, when the interlock is engaged the crane will lift the entire transfer device. Accordingly, only a single lifting mechanism is required. The interlock device utilizes a pair of releasable latch members which pass through openings in the outer and inner members. The interlock device is designed to automatically lock the inner member in place each time the inner member is brought to the upper position. The interlock device is also designed to automatically lock the inner member in place when brought into the lower position if the gripper is not engaged with a poison rod assembly. When the gripper has engaged a poison rod assembly, a shield device prevents the interlock device from operating. Thus, when the inner member is in its lower position and the gripper has engaged a poison rod assembly, the inner member will not be locked in the lower position and, therefore, can be lifted without the operator having to release the interlock, thus simplifying the lifting operation. Typically, removal of a poison rod assembly will begin with the inner member locked in the lower position. When the inner member is locked in the lower position, the gripper is located near the bottom of the outer member. After the transfer device is seated on a fuel cell, the operator will release the interlock allowing the gripper to be lowered further to engage the T-bar on the poison rod assembly. When the gripper is rotated into the latched position, the shield device is engaged and will prevent the interlock device from reengaging when the inner member reaches the lower position. When the operator raises the overhead crane without the interlock engaged, the inner member slides vertically within the outer member, lifting the poison rod assembly out of the fuel cell. When the inner member reaches the upper position, the interlock device engages, preventing the inner member and poison rod assembly from moving relative to the outer member. At this point, as the crane continues to lift, the entire transfer device will be lifted off the fuel cell. When the transfer device is seated on another fuel cell, the operator can only lower the poison rod assembly by releasing the interlock device. When the interlock device is released, the inner member and poison rod assembly may move vertically relative to the outer member. When the crane is lowered, the inner member slides downwardly in the outer member as the poison rod assembly is lowered into the new fuel cell. When the poison rod assembly is seated within the fuel cell, the operator may unlatch the gripper, thereby also disengaging the shield device. When the inner member is raised to the lower position, the interlock assembly once again locks the inner member in the lower position. When the crane is raised, the entire device will lift with the inner member locked into the lower position. |
claims | 1. An X-ray topography apparatus that uses X-rays for form two-dimensional images in correspondence with an internal structure of a sample, comprising:an X-ray source that produces X-rays with which the sample is irradiated;a multilayer film mirror provided in a position between the sample and the X-ray source;a slit member provided in a position between the sample and the X-ray source and including a slit that limits a width of the X-rays;two-dimensional X-ray detection means for two-dimensionally detecting X-rays having exited out of the sample; andsample moving means for achieving stepwise movement of the sample relative to the X-rays with which the sample is irradiated to sequentially move the sample to a plurality of step positions, wherein:the X-ray source produces the X-rays from a minute focal spot,the multilayer film mirror converts the X-rays emitted from the X-ray source into monochromatic, collimated, high-intensity X-rays,the direction in which the multilayer film mirror collimates the X-rays coincides with a width direction of the slit of the slit member,the step size by which the sample moving means moves the sample is smaller than a width of the slit, andthe combination of the size of the minute focal spot, the width of the slit, and the intensity of the X-rays that exit out of the multilayer film mirror allows the contrast of an X-ray image produced when the two-dimensional X-ray detection means receives the X-rays for a predetermined period of 1 minute or shorter to be high enough for observation of the X-ray image. 2. The X-ray topography apparatus according to claim 1, further comprising a processor, wherein the processor is configured to:acquire a two-dimensional cross-sectional image associated with each of the plurality of step positions, wherein the two-dimensional cross-sectional image is produced by irradiating the sample with the X-rays in each of the plurality of step positions for the predetermined period and detecting X-rays having exited out of the sample irradiated with the X-rays with the two-dimensional X-ray detection means, thereby acquiring a plurality of two-dimensional cross-sectional images,form a three-dimensional image by arranging the plurality of two-dimensional cross-sectional images, andacquire a second two-dimensional image by extracting data along a flat plane different from measurement planes associated with the three-dimensional image. 3. The X-ray topography apparatus according to claim 2, wherein the processor is further configured to calculate dislocation density based on the second two-dimensional image. 4. The X-ray topography apparatus according to claim 3, wherein the minute focal spot comprises a focal spot so sized as to fall within a circle having a diameter of 100 μm, and the width of the slit ranges from 10 to 50 μm. 5. The X-ray topography apparatus according to claim 4, wherein the multilayer film mirror comprises a parabolic form, so as to allow X-rays incident on the sample to be diffracted in parallel to each other. 6. The X-ray topography apparatus according to claim 5, wherein interplanar spacing of lattice planes in the multilayer film mirror is so differentiated from each other location-to-location that the X-rays incident at different angles of incidence are reflected off the entire surface of the multilayer film mirror. |
|
claims | 1. A modular submersible repairing system comprising: a working unit; and a base unit; wherein the working unit includes: at least one type of tool module repairing structures in a reactor, a scanning/pitching module being selectively connected to or disconnected from the tool module, and provided with a scanning/pitching shaft for scanning or pitching the tool module, a submersible fan module being selectively connected to or disconnected from the scanning/pitching module, and a first buoyant module for keeping an orientation of the tool module; the base unit includes: a manipulator module internally provided with an actuator driving mechanism, an adsorbing module being detachably mounted on the manipulator module and a wall, and a second buoyant module for keeping an orientation of the manipulator module; each of at least the scanning/pitching module and the manipulator module is provided with a submersible connecting device being operated in water for engagement and disengagement; configuration and functions of the modular submersible repairing system can be changed or modified according to various purposes of work in the reactor by properly combining those modules; and the modules can be connected together in the reactor by remotely operating the submersible connecting devices. 2. The modular submersible repairing system according to claim 1 , wherein the submersible connecting device includes a male connecting unit provided with a taper member, and a female connecting unit provided with a taper hole complementary to the taper member. claim 1 3. The modular submersible repairing system according to claim 2 , wherein the submersible connecting device is provided with an ultrasonic distance measuring device measuring distance between the modules to be connected in a noncontact measuring mode, and a locking mechanism for preventing disengagement of the male connecting unit and the female connecting unit. claim 2 4. The modular submersible repairing system according to claim 1 , further including a hoisting device for suspending and submerging a desired module in water, the hoisting device comprising a gripper for gripping the module, and an engaging/disengaging device adapted to draw a first module to be connected to a second module toward the second module and to push the first module away from the second module. claim 1 5. The modular submersible repairing system according to claim 4 , wherein the engaging/disengaging device includes an axially movable arm, and a claw which engages the first module to draw the first module toward the second module or to push the first module away from the second module. claim 4 6. The modular submersible repairing system according to claim 1 , wherein the manipulator module is provided with an extension mechanism which is expandable, and the scanning/pitching module is connected to an extremity of the extension mechanism. claim 1 |
|
050849093 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The entire disclosure of U.S. Pat. No. 4,749,869 entitled "PROCESS FOR IRRADIATING TOPAZ AND THE PRODUCT RESULTING THEREFROM" granted June 7, 1988 to Richard Fournier is specifically incorporated herein by reference. The reason for the blue color in both electron and neutron irradiated topaz has not been widely known.It has been attributed merely to displacement of electrons (as in the discussion in U.S. Pat. No. 4,749,869) or to the displacements of atoms within the crystal lattice to form F center and/or other color centers. Furthermore, the steely-grey tones in London Blue have been attributed to neutrons as well as gamma rays. It appears that the actual cause of blue color in topaz may be the presence of phosphorous as an impurity which allows the formulation of color centers by irradiation strong enough to displace the electrons. London Blue may be caused first by the transformation of silicon into phosphorous by neutron bombardment. Only after the phosphorous is present in the crystal structure can the associated gamma rays caused by the reaction of the neutrons with the shielding materials in the reactor produce the blue and grey color centers. Since the blue and grey are persistent upon heating to a much higher temperature in reactor irradiated topaz than in that irradiated by gamma or electron, it may be that the neutrons also cause atomic displacements which assist in the formation and stabilization on the color centers. A difference is also indicated by the different (A+B) axes on which the color of London Blue forms. Natural topaz containing phosphorous from 10-150 ppm will turn light-deep blue upon irradiation by gamma rays or electrons, with color forming on the C-axis of the crystal. Thus, if sufficient phosphorus were to be added to topaz either by formation in situ by nuclear reaction; by ion bombardment, or by diffusion, blue color could be subsequently generated by gamma or electron bombardment. Such color centers could be stabilized to heat by bombardment with high energy neutrons to cause atomic displacements. Furthermore, the crystal can be "healed", restored, to be more like its former structure by electron or high energy gamma irradiation subsequent to bombardments causing atomic displacements. The in situ formation of phosphorous can be achieved by transformation of silicon into phosphorous through bombardment with either protons or neutrons or other sub-atomic particles. This California Blue can be produced without resort to the London Blue process or nuclear reactors, or even electron bombardment. My invention, makes use of the energy heretofore wasted to perform various gem treatments; especially to "pre-screen" or"pre-treat" the topaz with a low dose of radiation in order to select, by subsequent visual inspection, which individual pieces are profitable to treat sky blue with electron irradiation. My invention provides a general substitute for Co60 gamma irradiation. My new method is superior to Co60 for prescreening as some stones are not affected by Co60 gamma radiation. In accordance with my method, gemstones are placed in an intense gamma ray field of 0.5 to 30MEV photons, and more preferably 3 to 25 MEV, either in air, or in water, or other cooling medium. Quartz is then irradiated until a dose of 17 to 400 megarads is accumulated; Tourmaline for 20 to 400 megarads; Beryl for 400 to 8,000 megarads; Diamonds for 400 to 10,000 megarads; and Topaz for 200-10,000 megarads. The stones may then be heat-treated and/or exposed to sunlight, or other source of UV radiation, to eliminate undesirable or extraneous shades of color and/or to lighten the color back to the most marketable tone. In particular, quartz is heated to 160-400 degrees centigrade to eliminate or lighten the dark smoky color generated by irradiation and to leave a pleasing smokiness, amethyst or citrine yellow color. Tourmaline is heat treated to 200-900 degrees centigrade to reduce orange/brown color which may develop in addition to the desirable pink/red added by irradiation. Beryl is heated to 175-400 degrees centigrade to eliminate green/brown shades from the good yellow/golden colors produced by irradiation. Diamonds are heated to 500-1100 degrees centigrade to lighten green, or eliminate green to leave yellow, or to change green to red/brown color in irradiated diamonds. In the case of blue topaz produced from colorless or light blue colored stones by irradiation with a 200 to 10,000 megarad dose of 3-25MEV gamma rays, heating is performed in the standard way, after irradiation at 175-250 degrees centigrade to eliminate the brown color and leave the refined "sky blue" color. The most exceptionally intense sky blue is known as "California Blue." Since this color is not obtainable by Co60 or similar irradiation,and since electron irradiation is generally limited to stones of no more than three-quarters of an inch thickness, the present invention is able to produce a much higher yield of topaz which could be classified as the very desirable and marketable "California Blue." For the purposes of improving the color of London Blue topaz, the optimum heating temperature is in the range of 500-600 degrees centigrade for a very short period of time, i.e. just a few minutes. Higher temperatures, up to 700 degrees centigrade, begin to degrade the color, rather than improve it, at least, on the specimens I tested. Using lower temperatures requires considerably longer heating times, e.g. 420 degrees centigrade over 4-6 hours. Lower temperatures take even longer, but the results are not as good as the higher temperatures no matter how long the heating. These comments pertain to heating both before or after gamma irradiation. In some cases, no heating at all is required for the proper enhancement and evening out of the color in poorly colored London Blue topaz, the gamma irradiation by itself being sufficient. Unlike the "Super Blue" process, based on electron radiation, the relative low to high energies (1/2 to 30 MEV) of gamma radiation causes the color to intensify by eliminating the green and grey colors throughout the stone. Because my method uses gamma rays, and not electrons, the problem of electron discharge, which can damage the stone, is eliminated. Therefore, my method allows a larger quantity of gemstones to be processed at one time. The higher the MEVs of the gamma radiation, the better the effect it has on the stones to eliminate the greens and greys of the neutron treated stones. I have found that the best way for achieving higher energy gammas is to use a linear accelerator, cyletron, or any high energy particle accelerator, and by converting electrons to gamma rays by using any conventional or non-conventional target, like platinum or tungsten. For low energy gamma rays it is possible to use isotopes, like Co60, Cs137, etc. or low energy particle accelerators. One suitable commercially available linear accelerator is the BC-20 manufactured by Brobeck Corporation of 1235 - Tenth St., Berkeley, California 94710. It is designed to operate at an energy of 10 MEV and provide electron beam power to 20KW. In the case of blue topaz which has been irradiated with neutrons in a nuclear reactor to make "London Blue" topaz, my invention allows for heating the stones at 300 up to 600 degrees centigrade, prior to irradiation with 3-25MEV gamma rays to an approximate dose of 1,000-10,000 megarads, in order to produce California Blue topaz of virtually any size. My invention is not limited to gemstones of approximately three-quarters of an inch in thickness or less as is the similar electron process. The production of California Blue topaz by electron irradiation often results in partial or total breakage of London Blue topaz by electrical discharge. This frequently occurs even in stones much smaller than the three-quarter of an inch limitation on thickness limit. The cause of the breakage is not apparent from visual examination. The breakage problem can be somewhat ameliorated by proper placement of the stones during electron irradiation. However, this severely limits the amount that can be processed at once and does not completely eliminate the breakage. The steely-grey and or green color of London Blue topaz can be eliminated, and breakage during irradiation can be avoided, by annealing the London Blue topaz at 300-600 centigrade prior to the electron irradiation, rather than afterwards. Heating after electron irradiation does not help the survival of the stones during such irradiation. Yet the prior heating works as well or better to reduce the steely-grey and/or green color which impairs the marketability of some reactor neutron irradiated London Blue. With my invention, approximately 400 megarads may be obtained in 50 hours, and up to 10kg or more of material may be exposed at once, which is five times the weight processed by the electron bombardment at the rate of approximately 800 megarads/hr. So, the useful waste radiation amounts to approximately 5% of the total. Not only does my method more correctly imitate the linear accelerator action for prescreening purposes on topaz, but it may also be used to treat topaz and other gem minerals in ways not possible by any of the methods formerly employed. A single topaz crystal in excess of 1 kg which had been subjected to approximately 1500 megarads of Cs137 gamma rays did not indicate very good color potential. But after just one 50 hr, approximately 400 megarad dose of the 8-12MEV gamma rays, the color became very deep sky blue in the thick cross section, much superior to the color from electron bombardment. Just as important, it did not break from electrical discharge as any rough stone even a fraction that size would be sure to do under electron bombardment. Various color changes have been observed in amethyst, colorless or pale quartz and certain tourmalines which did not respond to Co60 gamma rays, and which were too fragile or too costly to irradiate in the electron linac. As normally practiced short doses of electron irradiation became expensive due to set up time and cooling off time before the personnel can enter the chamber. Nigerian topaz exposed to 8000 megarads of Co60 gamma rays failed to achieve worthwhile color. The 1 and 1.3 MEV gamma rays from Co60 is simply unable to activate the color change. Whereas, only 400 megarads of the more energetic 8-12MEV gamma rays will give a stronger, better quality blue than 20 times the dose of lower energy rays in the Nigerian and lower qualities of Brazilian topaz abundantly available. In a preferred embodiment of my invention, London Blue Topaz which has been neutron irradiated and is very greyish or otherwise poorly colored, is subjected to gamma rays. By subsequent, prolonged careful heating of the topaz, a bright, deep blue color is obtained. Sometimes it is beneficial to heat the topaz before, as well as after, the gamma irradiation. Gamma ray sources of between about 0.5 MEV and 30MEV will suffice. The heating is conducted between about 250 and 700 degrees centigrade for periods between about 15 minutes and several days, e.g. seventy-two hours. The lower temperatures requiring longer heating times, as well as the more "stubborn" topaz stones. Heretofore, gamma rays generated by linear accelerators, betatrons, and other devices were considered too expensive for gemstone treatments. This is because the primary linear accelerator generates electrons and the efficiency of the conversion to gamma is only about 10%. My invention utilizes a strong, useable (concentrated) gamma flux that is a byproduct of the ordinary electron topaz irradiation process. The high energy gamma radiaton which was formerly ten times as costly as electrons is thus essentially free and available. This valuable radiation source could be marketed for uses other than gem enhancement. It could also be used to generate a neutron flux when appropriate elements are present in the material being irradiated. While I have described preferred embodiments of my invention, it may be varied in detail. Therefore the protection afforded my invention should only be limited in accordance with the scope of the following claims. |
060834544 | summary | FIELD OF INVENTION The present invention is concerned with producing evenly sized and evenly shaped spheres or balls. In one aspect, the invention relates to making uniformly sized metallic spheres for use in solder compositions. BACKGROUND OF THE INVENTION Various methods have been proposed to form droplets of material having a known size. The present invention is an improvement over the method described in the Chun et al. U.S. Pat. No. 5,266,098 which is expressly incorporated herein by reference. The Chun et al. '098 describes placing a charge on a droplet in an attempt to produce uniformly sized droplets. The Chun et al. method produced a high percent of unacceptable and unusable pieces or bits of material having flat, oblong or irregular shapes. The Chun et al. method is limited to producing irregularly shaped droplets having undesirably rough and bumpy surfaces. Other previously considered methods for forming droplets by imparting a charge on the droplets include: Smith U.S. Pat. No. 5,560,543 describes a method for forming droplets and passing droplets through charged and grounded plates to selectively deflect droplets. Soviet Patent No. 541682039-A1 describes forming droplets which acquire a charge by passing through an electric field. Orme et al. U.S. Pat. Nos. 5,171,360; 5,226,948; 5,259,593 and 5,340,090 describe methods and apparatuses for forming a net product by directing a stream of uniformly sized droplets onto a collector of the shape of the desired product. C. H. Passow thesis, The Massachusetts Institute of Technology (MIT) May 1992, describes forming uniform droplets by using parallel plates positioned below a charging plate to selectively deflect some droplets off to one side where they are collected. Various other methods proposed to form droplets include: Hayes U.S. Pat. No. 5,411,602 describes ejecting solder drops from an ejection device into a flow of inert gas and catching the solidified solder balls; Hommel et al. U.S. Pat. No. 4,956,128 describes an aqueous calcium chloride hardening solution through which droplets are passed; Yabuki et al. U.S. Pat. No. 4,744,821 describes forming drops and passing the drops through layers of oil and water; Fulwyler et al. U.S. Pat. No. 4,302,166 describes a droplet forming apparatus where the droplets fall into an aqueous solution of a nonionic surfactant; Green et al. U.S. Pat. No. 4,628,040 describes forming droplets using a venturi process where the droplets pass through an oil to harden the droplets; Eylon et al. U.S. Pat. No. 4,787,935 describes a method for making powders using swirling, cooling fluids to harden the droplets; Anderson U.S. Pat. Nos. 4,216,718 and 4,419,303 describe forming sodium amalgam particles for high pressure discharge lamps where a vibrating discharge nozzle forms droplets which fall into a fluid; Rhim et al. U.S. Pat. No. 4,981,625 describes forming polymeric microspheres by ejecting droplets of monomers, charging the droplets, freezing the droplets in a cryogenic liquid, and thawing the droplets by irradiation to activate free radicals which polymerize the monomer. The prior art droplet formation methods have not been entirely satisfactory for a number of reasons. A major concern is the wide distribution of the particle size of the droplets. Inconsistent sizes in the droplets makes the use of the droplets more difficult in soldering applications. Another problem is that after the formation of the droplets, the droplets have to be cleaned to remove contaminants or oxides on the surface or to remove oils and solutions through which the droplets have been passed. The cleaning of the droplets adds to the manufacturing time and costs. Another major concern is that the droplets have irregular shapes and/or bumpy or rough surfaces. This lack of sphericity makes handling and use of the individual droplets more difficult. Therefore, it is an object of the present invention to develop an apparatus and method for manufacturing high quality uniformly sized and shaped droplets. The present invention further provides a process which does not involve the use of multiple formation steps and/or cleaning steps. BRIEF DESCRIPTION OF THE INVENTION The present invention provides a highly accurate method and apparatus for producing uniformly sized and shaped spheres of a desired material, or materials. The apparatus comprises a sphere generation means and a controlled temperature solidification environment. In preferred embodiments, the spheres are formed using a uniform sphere generation means where a low viscosity liquid material is supplied into a crucible or feed system. The crucible has a means for heating and melting the material. The low viscosity liquid material in the crucible is subjected to a certain periodic disturbance by a stimulation actuator means. The crucible has at least one orifice which permits passage of the material therethrough. The material is subjected to a pressure differential (preferably about 4-50 psi) which forces the material through the orifice as a stream. The periodic disturbance applied to the material causes a controlled breakup of the stream of material into uniformly sized spheres. As the spheres are formed, the spheres are subjected to a positive or negative charge by a charging means. When the charging means is held at a predetermined voltage with respect to the stream, the combination of the voltage and the capacitance between the charging means and the stream brings a charge to the leading point of the stream. Each sphere retains a charge that the sphere held before it broke free from the stream. The charge on the sphere causes each sphere to repel from adjacent spheres. The like charge on the individual spheres prevents the spheres from merging together in flight with neighboring or adjacent spheres. In an especially preferred embodiment, the stream is broken by introducing minute periodic disturbances (preferably from about 1 to 30 kHz) by the stimulation actuator means. In certain embodiments the stimulation actuator means comprises either an electromechanical or piezoelectrical transducer. In one especially preferred embodiment, the transducer comprises a stack of five piezoelectric crystals mounted on a top portion of the crucible. The bottom four crystals are mechanically connected in serial and are electrically connected in parallel to a high sinusoidal voltage source. The top piezoelectric crystal serves as a motion sensor. The output voltage is an indication of the amplitude of the stimulation actuator means. In an alternative embodiment, the periodic disturbance or stimulation of the material can be obtained from a monolithic multilayer piezoelectric stimulation actuator. One preferred type of multilayer actuator has the following dimensions: length 5 mm, height 5 mm and width 5 mm. The multilayer actuator contains about 29 piezoelectric ceramic layers cofired together. The multilayer actuator is capable of over 3 micron expansion. In certain embodiments, an extender means, such as a rod, is attached to the bottom of the stimulation actuator means and extends into the supply of the material. The stimulation actuator means transfers the minute periodic disturbances to the material through the rod. In a different embodiment, the stimulation actuator means can comprise a piezoelectric ceramic material having a nozzle which is connected to the voltage source. The nozzle has a fixed aspect ratio defining the orifice in the crucible. The sinusoidal voltage is directly applied on the nozzle causing a minute periodic disturbance which radiates into the stream of material through the nozzle wall. It is contemplated that in certain very high temperature applications (for example, where the temperature of the low viscosity liquid material is over about 300.degree. C.), a lithium niobate (LiNoO.sub.3) material can be used instead of a piezoelectric ceramic material. At least one pressure regulator means supplies a constant hydrostatic pressure in the crucible. In a preferred embodiment, the pressure regulator means keeps the crucible at a negative pressure before the operation of the sphere generation means. The negative pressure prevents the material from dripping out of the nozzle. In a preferred embodiment, the pressure regulator means keeps the crucible at a positive pressure during the operation of the sphere generation means by maintaining the low viscosity liquid material at a desired level in the crucible. In a preferred embodiment, the pressure regulator means supplies a dry and inert gas, such as nitrogen, onto the low viscosity liquid material in the crucible. The positive pressure forces the material out through the orifice in the crucible. The amount of pressure on the material controls the flow of the material through the orifice. During operation of the sphere generation means, the applied constant positive pressure forces the material out of the nozzle and forms the jet or stream. The sinusoidal frequency from the stimulation actuator causes a minute periodic disturbance to the stream. Due to the Rayleigh instability effect, the disturbance builds due to energy of momentum in the stream, breaking the stream into uniform sized and uniform spaced spheres. The space between two neighboring spheres .lambda. is a function of the jet speed v.sub.j and the stimulation frequency, .lambda.=v.sub.j /f. In certain embodiments, the present invention provides a deflection means which is in a spaced apart relationship to the charging means. The deflection means comprises at least one set of deflection surfaces which are spatially separated from each other. A high voltage is applied across the set of deflection surfaces to generate an electric field between the surfaces. The deflection means creates an electrical force field through which the spheres pass. Since the spheres are charged, the electric field deflects on the spheres, depending upon their polarity. The deflection means spatially separates the spheres in the plane perpendicular to a center or vertical axis, further preventing the spheres from merging, thus further maintaining size consistency. The deflection distance of the spheres is a function of the size and speed of the spheres, the charge of the spheres and the force of the deflection field. The bigger spheres remain close to the center axis through the deflection means while the smaller spheres are deflected further away from the center axis. A visual observation system is positioned in a spaced apart relationship to the orifice in the crucible to monitor the formation of the spheres and to measure the size of the spheres. The visual observation system, in preferred embodiments, is operatively connected to the stimulation actuator means to increase or decrease the periodic disturbance being supplied to the low viscosity liquid material. The visual observation system is also operatively attached to the deflection means which increases, maintains or decreases the charge on the spheres in response to the information being collected by the visual observation system. The present invention is an improvement over the Chun et al. '098 technology where the droplets are not formed in a temperature controlled environment. According to another aspect of the present invention, the sphere generation means is operatively positioned in a controlled temperature solidification environment. The controlled temperature solidification environment allows the spheres to be formed in a short amount of time and distance. Also, the controlled temperature solidification environment provides less risk of contamination, such as oxidation, from occurring on the surface of the spheres. The controlled temperature solidification environment contains at least one heat transfer medium, such as a cold or liquified gas. The heat transfer medium provides a solidification environment where the spheres are cooled and solidified in a controlled manner. In certain embodiments the controlled temperature solidification environment provides a thermal gradient which allows the spheres to be quickly cooled in a controlled manner so that the spheres being formed have a consistently round or spherical shape in addition to a uniform size. In an embodiment where a tin-lead alloy is being used to form spheres for solder, the thermal gradient in the controlled temperature solidification environment ranges from about room temperature to about -90.degree. C. (and in certain embodiments about 0.degree. C.) in an upper chamber adjacent the crucible and ranges from about -110 to about -170.degree. C. in a lower chamber adjacent a bottom of the controlled temperature solidification environment. In other embodiments where different low viscosity liquid materials are being formed in spheres, the temperatures in the solidification environment are regulated so that the spheres form both quickly and uniformly. The temperatures in the controlled temperature solidification environment are affected by the type of heat transfer medium being used. It is to be understood that various heat transfer media, including liquified gases, halo-carbon fluids, ammonia, water and steam are within the contemplated scope of the present invention and that the temperatures of the heat transfer media can range from about room temperature to about -200.degree. C., depending upon the type of low viscosity liquid material being used to form the spheres. In embodiments where the low viscosity liquid material comprises a material which solidifies by radiating heat, for example, metals such as copper and steel, different types of heat transfer media may be used and the heat transfer media may be supplied at different temperatures. For example, in embodiments where titanium spheres are being formed the heat transfer medium can comprise a heated gas or vapor. In certain embodiments, it has been found that the contact of the spheres by rapidly moving or flowing heat transfer medium tends to cause the spheres to have uneven shapes. Therefore, in certain preferred embodiments, it is desired that there be as little movement or flow of the heat transfer medium. The heat transfer medium present in the controlled temperature solidification environment is substantially at rest or still such that there are generally no currents or flows of medium to contact or misshape the spheres being formed. In certain embodiments, the controlled temperature solidification environment comprises a first or gaseous environment containing a first heat transfer medium and a second or liquid environment containing a second heat transfer medium. In other embodiments the controlled temperature solidification environment comprises the first or gaseous environment without the use of the second or liquid environment. However, the description herein will describe in detail the gaseous/liquid controlled temperature solidification environment in order to provide a full understanding of all embodiments of the present invention. It should be further understood that all the embodiments are within the contemplated scope of the present invention. The use of the gaseous/liquid controlled temperature solidification environment is especially useful in the formation of spheres from relatively soft or malleable materials and in the formation of spheres which have large diameters or have a high latent heat of fusion (i.e., the quantity of heat evolved in the transformation of the material from the liquid phase to the solid phase). As the spheres exit the deflection means the spheres begin to solidify by first forming a skin or shell on the outer surface of the sphere. Before the spheres exit the gaseous environment zone, the spheres have substantially solidified, i.e., the heat of fusion has been transferred from the spheres to the gaseous heat transfer medium, or has been radiated. The spheres may still be at a high temperature, rending the spheres malleable. The spheres then pass into the liquid environment which preferably comprises a low temperature inert liquid material such as a supply of liquid nitrogen. The liquid environment further cools the spheres to remove the specific heat and to harden the spheres. The liquid environment also acts as a cushioning medium to prevent the spheres from mechanically deforming by colliding with each other or the walls and the bottom of the upper and lower chambers. In one preferred embodiment a low temperature liquid material is dispensed both into a top portion and into a bottom portion of the controlled temperature solidification environment. The low temperature liquid dispensed from the top portion is dispensed into the gaseous environment and at least partially vaporizes thus keeping the gaseous environment at a preferred low temperature. In certain embodiments it is contemplated that the top dispensed liquid can contact the descending spheres to quicken their cooling. In a preferred embodiment, the bottom portion of the controlled temperature solidification environment comprises a funnel which contains the second or liquid environment. The liquid environment cushions the spheres prior to the spheres hitting the bottom of the funnel. No further processing steps need to be carried out on the finished spheres once the spheres have passed with the liquid environment pool. There is no need to remove any oil or other materials from the surface of the spheres. It is within the contemplated scope of the present invention to monitor and respond to any changes in the operating parameters for forming the uniformly sized spheres. A first thermocouple measures the temperature at the top portion of the controlled temperature solidification environment and a second thermocouple measures the temperature at the bottom portion of the controlled temperature solidification environment. A differential pressure sensor monitors the pressure of the crucible. A data acquisition/control system is operatively connected to the pressure sensors, the thermocouples, the sphere generation means and the visual observation system. The data acquisition/control system collects pressure and temperature measurement data and controls the sphere generation means. The data acquisition/control system and the visual observation system provide the feasibility of actively controlling the size of the spheres being generated. The size of the sphere is measured by the visual observation system. The data acquisition/control system receives continuous and updated information on the crucible pressure and the frequency provided by the stimulation actuator means so that the sphere diameter is kept at a predetermined size. It is to be understood that the actual size of the spheres being generated depends upon the end use requirements. According to the present invention, the formation of spheres and the control over sphere shape and diameter is accurate to within microns. The spheres have a diameter that are precise to within about 1% of each other. The present invention is a further improvement over the Chun et al. '098 technology since the spheres produced according to the present invention have a substantially spherical and smooth surface. The spheres of the present invention have high degree of sphericity where the diameters through any section of a sphere vary less than about 1.0%. The method and apparatus of the present invention are useful for forming uniformly sized and shaped spheres having diameters that range from about 12 to about 1000 microns. The present invention is particularly useful for forming large spheres having a diameter greater than about 500 and in certain preferred embodiments about 760 microns (0.030 inches.+-.0.0003). The enclosed controlled temperature solidification environment controls the rate of transition from the liquid state to the solid state of the spheres. The volume and surface ratio of the spheres being formed affects the rate of cooling of the spheres. The larger diameter spheres are solidified at a controlled rate within the controlled temperature solidification environment so that the spheres maintain a substantially round and smooth surface and have a uniform shape. The operating parameters, including orifice diameter, frequency and amplitude of the periodic disturbances of the spheres, can be varied so that spheres having different diameters can be formed. It is to be understood that the optimum diameter of the spheres depends, in part, upon the type of sphere being formed. Other parameters, such as feed rate of the metal into the crucible, crucible pressure, temperature of the material being formed into spheres, and the amount of charge on the spheres also affect the size and rate of formation of the uniformly sized and shaped spheres. It is to be understood that the temperature of the low viscosity liquid material itself affects the thermal state of the spheres. In certain embodiments, the temperature of the material can vary from just above the melting point, and in other embodiments can be, for example, about 50.degree. C. above the melting point of the material. This difference in the temperature of the material affects the rate of solid sphere formation. In certain embodiments of the present invention, another parameter which can be varied is the "stand-off" distance between the sphere generation means and the point in time where the sphere solidifies. The present invention allows the spheres to solidify more quickly than in a conventional droplet formation apparatus. In preferred embodiments, the spheres descend through the enclosed controlled temperature solidification for about 0.5 to about 1.5 seconds prior to contacting the bottom of the enclosed environment. The enclosed controlled temperature solidification environment allows the spheres to be formed in a much shorter distance (about 1 to about 5 meters versus 10 to 20 meters found in conventional sphere forming apparatuses) and in a much shorter time (about 0.5 to about 1.5, preferably about 0.8 seconds versus 7 to 10 seconds). The formation of the spheres in the controlled temperature solidification environment allows the spheres to be formed substantially without any contamination or oxidation. In preferred embodiments, the spheres are formed from a low viscosity fluid, including, for example, glasses, ceramics and metals. In certain embodiments, it is within the contemplated scope of the present invention that the spheres can be formed of a wide variety of metals including tin/lead solder alloys, gold, aluminum, steel or copper alloys. In addition, the spheres can be plated with precious metals such as silver, gold or palladium or can be coated with organic coatings to prevent oxidation after their formation. The spheres are especially useful in such applications as solder for interconnection of integrated circuits to printed circuit boards, especially ball grid arrays, chip scale packages and flip chip packages. In certain embodiments, the spheres formed according to the present invention are especially useful in a solder composition and the spheres do not need any additional flux materials to prevent oxide formation on the surface of the spheres. While certain preferred embodiments have been shown and described herein, it is to be understood that the invention is not limited thereto, but may be embodied with the scope of the following claims. |
description | This application claims priority of U.S. Provisional Patent Application Ser. No. 61/048,273, filed Apr. 28, 2008, the disclosure of which is hereby incorporated in its entirety by reference. The present disclosure is related to an apparatus and a method for detecting and controlling delamination of a film. Ion implantation is used to perform a variety of functions. One such function is the doping of a semiconductor material, such as a wafer, to change its electrical properties. The goal of this process is to create a region within the substrate that has a polarity that may be different from the surrounding area. This process is used to develop integrated circuits, used in electronic components such as processors, memories, and other devices. In this embodiment, techniques such as, but not limited to, CVD, PECVD, plasma immersion and beamline implantation, may be used to introduce ions to the substrate. More recently, interest has grown in using ion implantation to cleave a thin film of material from a bulk substrate. There are several methods of performing a cleave process, such as one referred to as “SmartCut”. This process is used for many applications, including the preparation of silicon-on-insulator (SOI). Briefly, a semiconductor substrate, such as a wafer, receives a surface treatment to oxide the surface. This creates an insulating layer around the substrate. An ion implantation of hydrogen and/or helium is then applied to the substrate. In some embodiments, the substrate is then flipped and bonded to a handle substrate. This handle substrate may be silicon, quartz or some other suitable material. The implanted hydrogen or helium ions tend to cause bubbles while the substrate is being annealed. These bubbles may aggregate to form a layer within the substrate. The depth of this layer is dependent on the concentration and energy of the hydrogen ions, as well as the anneal time. This layer weakens the substrate at that position, allowing it to be cleaved. This cleaved interface is then smoothed, using techniques such as chemical-mechanical polishing (CMP). The resulting film and handle substrate is then suitable for use in a SOI process. The remainder of the original semiconductor wafer can be reused to create another thin film. In addition to the SOI process, cleaving processes are also gaining interest for other applications, such as a method of fabricating solar cells. As with SOI, these thin films are susceptible to strain, which can deform or destroy the film. Accordingly, it may be desirable to detect and monitor the delamination process. Furthermore, in addition to monitoring the delamination process, it would be beneficial to control the thin film delamination process. Additionally, it would be desirable if these techniques were used to determine delamination of other surfaces, such as chamber walls and equipment. The problems in the prior art are overcome by the method and apparatus described herein. An interferometer is used to detect the onset and progression of thin film delamination. By projecting one or more wavelengths at a surface, and measuring the reflectance of these projected wavelengths, it is possible to monitor the progression of the delamination process. Testing has shown that different stages of the delamination process produce different reflectance graphs. This information can be used to establish implantation parameters, or can be used as an in situ monitor. The same techniques used to detect delamination of a thin film from a semiconductor substrate can be used for other applications. For example, in certain systems, such as a CVD reactor, a film of material may be deposited on the walls of the chamber. This film is not deleterious until it begins separating from the wall. The techniques described herein can be used to monitor this separation, and determine when preventative maintenance may be performed on the chamber. In the present disclosure, several embodiments of an apparatus and a method for detecting film delamination are introduced. For illustrative purpose, the present disclosure may be made in context to systems for manufacturing and/or processing thin films. However, those in the art will recognize that the present disclosure need not be so limited. Indeed, the present disclosure is applicable for detecting delamination of a thin film where such a delamination is desired or undesired. Accordingly, the present disclosure may also be applicable for systems for detecting delamination of the film from a bulk material caused by wear and tear. Among systems for manufacturing and/or processing thin films, the present disclosure will focus on a beam-line ion implantation system for a purpose of clarity. However, those in the art will recognize that the present disclosure may also be applicable to other types of systems for manufacturing and/or processing thin films. For example, the present disclosure may be applicable to a plasma based system including a plasma immersion ion implantation system. In addition, the present disclosure may also be applicable to optical based thin film processing system. In the present disclosure, the thin film may be conducting, semiconducting, or insulating material. For example, the thin film may be Aluminum (Al) thin film, silicon (Si) thin film, Gallium Arsenide (GaAs) thin film, Germanium (Ge) thin film, diamond thin film, organic or polymeric thin film. The thin film may be transparent to at least a portion of the electromagnetic spectrum. Meanwhile, the substrate may also be conducting, semiconducting, or insulating material. The ions disclosed in present disclosure may be atomic or molecular ions. Further, although the present disclosure may focus of system based on ions, the present disclosure may also be applicable to other particle based systems such neutral particle based system and photon based system. As such, the thin film, substrate, nor particles in the present disclosure need be limited to a particular type of film, substrate, nor particles. FIG. 1 illustrates an apparatus for manufacturing a thin film. The cleaving process, unlike conventional ion implantation processes, does not form or deposit the film on the substrate. Instead, the cleaving process cuts or separates the film from a bulk substrate. Among other tools, a beam-line ion implanter 100 may be used to perform the cleaving process. A block diagram of a conventional ion implanter 200 is shown in FIG. 1. The conventional ion implanter may include an ion source 102 for generating ions. The ion implanter 100 may also comprise a series of beam-line components through which ion beam 106 passes. Examples of the beam-line components may include extraction electrodes 104, a magnetic mass analyzer 108, a plurality of lenses 110, and a beam parallelizer 112. The ion implanter 100 may also include a platen 116 supporting the wafer 114 to be implanted. The wafer 114, meanwhile, may be moved in one or more dimensions (e.g., translate, rotate, and tilt) by a component, sometimes referred to as a “roplat” (not shown). During implantation, the ions of desired species, such as hydrogen and helium ions, are generated and extracted from the ion source 102. Thereafter, extracted ions 106 travel in a beam-like state along the beam-line components and implanted to the wafer 114. Much like a series of optical lenses that manipulate a light beam, the beam-line components manipulate the ion beam 106. The ion beam 106 manipulated by the beam-line components is directed. As described above, one application that uses a cleaving process is the creation of SOI substrates. As illustrated in FIG. 2a, ions such as protons or hydrogen and/or helium ions 106 directed toward the wafer 114 are implanted at a predetermined depth. In some embodiments, the wafer 114 has been treated so as to create an insulation layer 121 on the top surface. This may be achieved by oxidizing the top surface. The implanted ions coalesce to form an intermediate hydrogen layer 114b, between upper layer or thin film 114a and lower layer 114c or bulk of the wafer 114, and the film 114a may be delaminated or separated from the bulk 114c. As shown in FIG. 2b, to form a thin silicon-on-insulator (SOI) film, a substrate 120, such as silicon, quartz or polyethylene terephthalate (PET), may be affixed to the film 114a and the film and the substrate 120 may be delaminated or separated from the bulk 114c. After separation, a thin SOI film may result. As seen in FIG. 2b, the cleaving process results in a uneven or jagged edge. Typically, the resulting SOI wafer contains a substrate 120, and a thin film layer 114a, often with an insulating layer 121 separating these two substrates. The SOI wafer then undergoes a polishing step, such as chemical-mechanical polishing (CMP) to smooth the top surface. The resulting wafer is shown in FIG. 2c. The remainder of the unused wafer 114c may be used in a subsequent cleaving process. During implantation of ions 106 and formation of the intermediate layer 114b, the upper layer 114a may be under a strain. The film 114a may also experience strain as the film 114a is delaminated from the bulk 114c. As the film 114a is relatively thin, excessive strain may cause deformation or even catastrophic failure of the film 114a. Interferometry is a technique whereby light is incident on a sample, and the reflected intensity is monitored. FIG. 3 illustrates a diagram of this measurement technique in which a light is incident on a sample (at an angle) and the reflected intensity is monitored. The light can also be incident at a normal angle. The percentage of the incident light that is reflected, the reflectivity, is a function of the thickness of the film and its optical properties. As suggested by the equation shown in FIG. 3, the reflected intensity is a function of the incident light intensity, the wavelength and incident of that incident light, the thickness of the film, and the optical properties of the film and the underlying substrate. Reflectivity is a measure of the amplitude of the reflected light as a function of the incident light. Thus, a reflectivity value of 0.5 indicates that the reflected wave has an intensity that is 50% of the incident wave. FIG. 4 shows the two ways to monitor the reflectivity of the sample. In FIG. 4a, the reflectivity of a substrate as a function of wavelength may be monitored by passing the broadband light source through a monochromator. The resulting waveform shows the reflectivity of the sample as a function of wavelength or frequency. As shown in FIG. 4b, a single wavelength may be chosen, and the single wavelength may be monitored for change in reflectivity as a function of time. In the present disclosure, the wavelength is not limited to a single wavelength. Instead, a multi-wavelength optical beam may be used to detect a thin film delamination. In some embodiments, a multiwavelength measurement may be used, as this technique may provide more information. In other embodiments, the single wavelength implementation may be used, as it may be simpler to implement, as the detector used in single wavelength measure may be a photodetector. The choice of which technique to use is dependent on the application, and the present disclosure is not limited to a specific embodiment. In the cleaving process, ionized particles, such as hydrogen or helium, may be implanted. During subsequent thermal processing, these particles may diffuse from lattice site of the substrate and into voids/pockets formed by during the implant process. As the particles gather in such pockets/voids, the particles may form bubbles. If the internal pressure exerted by the pockets/voids exceeds the coherent strength of substrate, then the film may delaminate. In the present disclosure, this delamination process can be detected by thin film reflectance, as shown in FIG. 5. In FIG. 5, the top row shows the thin film reflectance curves for increasing implanted dose. The terms reflectance and reflectivity are used interchangeable throughout this disclosure, and both are defined to be a measure of the reflected light intensity in comparison to the incident light intensity. The particles used for the implantation may be a mixture of particles containing a substantial portion of helium and/or hydrogen. The substrates have all been annealed at 600° C. for 5 minutes. As the dose of the particles increases from 4e14/cm2 to 4e16/cm2, the reflectance shows a strong fringe, indicating creation of a free surface within the substrate lattice from which the reflection is quite strong. Beyond this dose, the amplitude of the fringe decreases indicating loss of coherence between the film and substrate and presence of delamination. The term fringe is used to indicate a change in the reflectivity at one or more wavelengths. The creation of bubbles within the substrate, as seen in the graph labeled 4e16, shows an unexpected large decrease in reflectivity at about 570 nm, and an increase in reflectivity at about 730 nm. This may be explained by the interaction of the incident wave with the substrate. As the substrate begins to bubble, there is strong interference from the internal surface (i.e. the bottom surface of the bubble). Due to the characteristics of the bubble, the light reflected from the internal surface may be highly coherent. This high degree of coherence results in large increases and decreases in reflectivity, depending on the relationship between the thickness of the film and the phase of the reflected light. As the film delaminates, the level of coherence decreases, due to the uneven gap between the film and the underlying substrate. This decreased coherence reduces the amplitude of both the downward and upward spikes. However, the delaminated film (i.e. the two rightmost graphs) still demonstrates a different profile than the substrate prior to delamination (i.e. the four leftmost graphs). These differences allow an operator to monitor the progression of the delamination process. The three stages of delamination shown in FIG. 5 (no bubbles, bubbles forming, film delaminated) were confirmed using an atomic force microscope (AFM). FIG. 6 shows 4 representative reflectance graphs from FIG. 5. For each of these graphs, the corresponding AFM picture is shown directly below it. Note that, for the two leftmost graphs, the reflectance shows no unusual behavior, and specifically shows a general decrease in reflectance as the wavelength increases. The minimum reflectance appears to be about 0.35. The corresponding AFMs for these graphs show no bubbles have formed yet, thus the reflectance is almost exclusively from the top surface. The third graph shows the greatest amplitude variation, with a large decrease in amplitude at 570 nm and an increase in amplitude at 730 nm. The AFM graph below shows the presence of large surface bubbles, which are the cause of this effect. Finally, the rightmost graph shows a film that has been delaminated. In contrast to the first two graphs, this graph shows an increase in reflectivity between the wavelengths of roughly 450 nm to about 730 nm. Again, the AFM picture appears below the graph and shows that the film has delaminated. To monitor the delamination of a wafer, a light source using a wavelength (or range of wavelengths) that may be substantially transparent to the substrate may be chosen. Based on the graphs shown in FIG. 5, if a single frequency source is used, light at a wavelength of 530 nm may be used to determine when delamination has occurred. For a multiwavelength light source, much of the infrared band would be applicable if the substrate of interest is silicon. This transparency is desirable as there should be at least some penetration into the bulk of the substrate by the light. In the present disclosure it may be desirable for the thin film to be thin enough to allow some transmission of the light. For example, it may be desirable for approximately 10% of incident light to be transmitted. As illustrated in FIG. 7A, the light should be incident on the substrate and at least a portion of the incident light should be reflected from the substrate 130. In this Figure, Ii represents the incident intensity and IR, the reflected intensity. The left-hand portion of the diagram shows the setup for normal or near-normal incidence. The optical beam may reflect from several interfaces—including the backside of the wafer and the internally separating interface. If the other reflections obscure the signal of interest, then an interferometer that strikes the substrate at an angle may be implemented. In this way, a reflected optical beam that is not of interest may be separated out (by confining the detector area to sample only the reflection of interest). As an example, FIG. 7B shows a light wave incident on a substrate 130 at an angle. In this example, light may be reflected from the top surface (IR1), the bottom surface (IR3), and the internal surface of interest (IR2). If the magnitude of IR1 or IR3 is too great, the changes in IR2 may not be easily detected. By projecting the light at an angle at the substrate, three distinct reflection paths are created. In such an instance, the detector 135 may be located in the path of IR2. By locating this detector in the path of IR2, such that it cannot detect reflected light IR1 or IR3, the accuracy of the detector 135 may be improved. As stated above, the graphs shown in FIGS. 5 and 6 were created by using different concentrations of implanted material, with a constant anneal time. However, the method described herein can be used in other applications. For example, in one embodiment, a predetermined concentration of ions, such as hydrogen and helium ions, can be implanted in the substrate. The substrate is then subjected to an anneal process. As the substrate is heated, it will undergo the various steps shown in FIG. 5, namely, after some time, bubbles will begin to form. Later, the thin film will delaminate. A beam of light, as described above, can be directed toward the substrate while it is being annealed. As the substrate is being annealed, the hydrogen and helium ions diffuse in the substrate, forming bubbles. As described above, these bubbles can be detected. Thus, by placing a light source and detector within the anneal chamber, it is possible to determine when the film has delaminated, based on the changes in the reflectivity or reflectance graph. Upon detection of this condition, the annealing process can be terminated. Alternatively, the interferometer may be placed outside the anneal chamber, using a transparent window to project light into the chamber. In another embodiment, the substrate is not annealed. Rather, ions are implanted into the substrate continuously, until the substrate begins to bubble as described above. A light source and detector are located so as to be incident on the substrate during the implant process. Once the film has delaminated, based on changes to its reflectance graph, the ion implantation process is terminated. Thus, the method described herein can be used to determine and control either ion implantation time or anneal time. The present method can be used in situ to determine the appropriate time required to achieve the desired level of delamination. The location of the light source and the detector is dependent upon the portion of the process that is to be monitored and controlled. For example, the manufacturing process may include a predetermined dose of particles be implanted into a substrate. The substrate is then annealed, where an interferometer is able to observe changes in the reflectance graph. Upon determination by the interferometer that the top film has delaminated, a controller, in communication with the interferometer may terminate the anneal cycle. In another embodiment, the method used above is not used in situ, but rather is used to determine the appropriate standardized process parameters. For example, a test environment may be created where the light source and detector are located so as to be incident on the substrate. As described above, the interferometer may monitor the substrate at a number of steps in the process, including but not limited to the implantation step or the anneal step. One or more tests are performed using this interferometer and the time required to achieve the desired degree of delamination is determined. Using the data collected during the test process, a standard process can be established, such that the light source and detector need not be used during normal operation. In other words, it may be determined that for a given dose, test data suggests that an anneal time of 5 minutes results in the proper amount of film delamination. In this case, the standard process can be established wherein the anneal time is set to 5 minutes. The present method is not limited to detection of delamination of a semiconductor wafer. As an example of a second implementation, present disclosure may be applicable to a surface where detection of film delamination is desired. Film delamination is a common problem in reactors in which deposition is present, including etch, deposition and some implanters (e.g. PLAD). A light source 205 may output a light beam toward the reactor wall 208 and monitor the reflectance for evidence of delamination. FIG. 8 shows a reactor chamber 200, having a window 210 through which light can pass. The light is reflected off the opposite wall 208 and back through the window 210. The light source 205 and detector 207 may both be located outside the chamber 200, such that they operate through the window 210. If a chamber wall 208 delaminates, there may be thin-film interference between the top and bottom surfaces of the film wall, which may cause a change in the reflectance or reflectivity. This change could be an increase or decrease in intensity, depending on the optical characteristics of the film and the wavelength(s) being used, as described above. When the system detects delamination, this may indicate that maintenance or in situ clean may be performed. Currently, cleaning of chamber walls is done as part of the preventative maintenance (PM). Typically, a PM cycle is recommended based on the number of cycles, or the number of elapsed hours since the last cleaning. In order to insure that flaking films do not contaminate the chamber, it is customary that the PM cycle will be performed more frequently than necessary. By incorporating a method of detecting the delamination of film from a chamber wall, the frequency of PM can be decreased, as it is now possible to determine exactly when the chamber wall will begin to delaminate. This method can be used for any solid that tends to accumulate gaseous species, including apparatus within the reactor chamber. In addition to monitoring film on the chamber walls, the present method can be used to monitor film buildup and delamination on the associated equipment, such as the platen 116, and other components of the ion beam line, such as the magnet 112. Porous materials, such as graphite, may be less susceptible to bubbling as the gas may be able to diffuse to the surface and out into the chamber environment, but they have their own challenges with suppressing particle formation. FIG. 9A shows a front view of the wafer 114 in a semiconductor process. Note that the platen 116 extends beyond the edge of the wafer 114, and is therefore exposed to the ion beam. As a result, film may develop on the exposed portion of the platen 116. FIG. 9B shows a top view of the platen 116 and the attached wafer 114. To monitor the surface of the platen 116, a light source 140 may be aimed and focused at the fixture, as shown in FIG. 9B. As described above, the choice of wavelength would depend on the material used for the platen. Preferably, the light should be of a frequency that is transparent to the material of the platen. Thus, if the platen is silicon, a wavelength (or range of wavelengths) in the visible and/or IR bands may be applicable. If the platen is metal or a semi-metal, such as graphite, a suitable wavelength may best be determined empirically. Alternatively, if the platen is made of an insulator such as quartz or a ceramic, then there are typically some suitable bands of transparency, which can be determined by referring to the optical properties of these materials. The reflected beam IR would then be sensed by the detector, so that its reflectance graph can be generated and monitored. The monitoring of the reflected beam may have to be synchronized with the motion of the platen. As noted earlier, the present disclosure is not limited to semiconductor manufacturing/processing system. On the contrary, the present disclosure may be equally applicable to detecting film lamination on a wall where a strip or film of the wall delaminates due to wear and tear. |
|
abstract | A focused ion source based on a Hall thruster with closed loop electron drift and a narrow acceleration zone is disclosed. The ion source of the invention has an ion focusing system consisting of two parts. The first part is a ballistic focusing system in which the aperture through which the beam exits the discharge channel is tilted. The second is a magnetic focusing system which focuses the ion beam exiting the discharge channel by canceling a divergent magnetic field present at the aperture through which the beam exits the discharge channel. The ion source of the invention also has an in-line hollow cathode capable of forming a self-sustaining discharge. The invention further reduces substrate contamination, while increasing the processing rate. Further the configuration disclosed allows the ion source to operate at lower operational gas pressures. |
|
claims | 1. A nuclear thermal propulsion system comprising:a nuclear reactor core including:an array of fuel elements;an array of tie tubes adjacent the array of fuel elements, each tie tube including:a propellant supply passage to flow a propellant;an inner tie tube layer surrounding the propellant supply passage;a moderator sleeve surrounding the inner tie tube layer;a propellant return passage surrounding the moderator sleeve; andan outer tie tube layer surrounding the propellant return passage; anda burnable poison dispersed in the nuclear reactor core, wherein:the burnable poison includes Gadolinium (Gd),the Gd is dispersed in an alloy that forms the outer tie tube layer, andthe Gd dispersed in the outer tie tube layer alloy is in a quantity of 10 parts per million (ppm) to 70 ppm. 2. The nuclear thermal propulsion system of claim 1, wherein:the propellant is hydrogen. 3. The nuclear thermal propulsion system of claim 2, wherein:the array of fuel elements includes a graphite composite fuel formed of low-enriched uranium (LEU) having 20% or lower 235U; andthe moderator sleeve includes a solid hydride. 4. The nuclear thermal propulsion system of claim 2, wherein:the array of fuel elements includes a tungsten ceramic and metal matrix (CERMET) fuel formed of low-enriched uranium (LEU) having 20% or lower 235U, andthe moderator sleeve includes a solid hydride. 5. The nuclear thermal propulsion system of claim 2, wherein each tie tube further comprises:an inner gap between the inner tie tube layer and the moderator sleeve;a graphite insulation layer surrounding the outer tie tube layer;a medial gap between the outer tie tube layer and the graphite insulation layer;an outer gap surrounding the graphite insulation layer;a graphite sleeve surrounding the outer gap; anda coating formed of zirconium carbide or niobium carbide surrounding the graphite sleeve. 6. The nuclear thermal propulsion system of claim 5, wherein the graphite insulation layer includes zirconium carbide (ZrC), titanium carbide, silicon carbide, tantalum carbide, hafnium carbide, ZrC—ZrB2 composite, or ZrC—ZrB2—SiC composite. 7. The nuclear thermal propulsion system of claim 2, wherein:the propellant supply passage, the inner tie tube layer, the moderator sleeve, the propellant return passage, and the outer tie tube layer are radially arranged; anda respective radial wall thickness of the outer tie tube layer is less than the inner tie tube layer. 8. The nuclear thermal propulsion system of claim 7, wherein the array of fuel elements are interspersed with the array of tie tubes and each of the tie tubes is in direct or indirect contact with at least one fuel element. 9. The nuclear thermal propulsion system of claim 7, further comprising:a plurality of circumferential control drums surrounding the array of fuel elements and the array of tie tubes to change reactivity of the nuclear reactor core. 10. The nuclear thermal propulsion system of claim 1, wherein the Gd in the outer tie tube layer alloy is enriched in 157Gd isotope. 11. The nuclear thermal propulsion system of claim 10, wherein the outer tie tube layer alloy further includes nickel, chromium, and iron. 12. The nuclear thermal propulsion system of claim 11, wherein the outer tie tube layer alloy further includes molybdenum, niobium, cobalt, manganese, copper, aluminum, titanium, silicon, carbon, sulfur, phosphorus, and boron. 13. The nuclear thermal propulsion system of claim 10, wherein the outer tie tube layer alloy is also formed of zirconium or zirconium carbide. |
|
044629540 | claims | 1. A self pulsating nuclear reactor plant comprising a pair of mobile nuclear mass blocks chosen for radially pendant motion in an arcuate cylinder filled with a liquid moderator, the arch of each mobile nuclear mass block being limited in a first position by direct surface contact between both mass blocks in vertical position where both mobile mass blocks become super critical, the resultant nuclear reaction moving both blocks apart to a second position, the arch of the mobile mass blocks limited by stationary nuclear mass blocks which again makes both sets of nuclear mass blocks super critical whereby the nuclear reaction repulses mobile mass blocks back towards the first position, this motion serving to pump a working fluid through the reactor to drive a liquid turbine or motor to do physical work, collecting high energy charged particles emanating from each nuclear excursion inside a copper primary coil filled with a coolant and directing said changed particles into a secondary coil inside an energy cell whereby secondary coil winds around a cathode through which the working fluid circulates, thus creating a induction current, the reactor comprising: (A) A pair of mobile nuclear mass blocks situated between a vertical and horizontal position inside an arcuate cylinder, (B) A pair of stationary nuclear mass blocks limiting the arch of mobile nuclear mass blocks, (C) A reactor housing in form of an arcuate cylinder with inlet and outlet valves to direct a working fluid through the reactor by the motion introduced by the nuclear reaction of the mobile nuclear mass blocks, (D) A cooling coil made from copper tubing circulating a coolant which carries radiated energy, (E) A magnetic filter to remove nuclear particles from the working fluid, (F) A liquid turbine motor driven by the working fluid accelerated inside the reactor, the motor being connected to a drive means, (H) A storage tank to store the working fluid, (I) An energy cell containing a circulation tank of the working fluid whereby the circulation tank acts as a cathode by a positive coil winding around cathode inside an active solution which is stimulated by impulses of electro magnetic forces and chemical interactions, (J) zinc plates installed around the cathode which interacts galvanically with the copper winding of the coil depending on the frequency of the nuclear impulses inside reactor housing. 2. A self pulsating nuclear reactor of claim 1 wherein the pair of mobile mass blocks are mounted on an upper pin whereby gravitical forces tends to introduce surface contact in center of the arcuate cylinder whereby both mobile nuclear mass blocks become super critical to repulse each other in a pendant radially motion from vertical position towards horizontal position whereby the arch of the mobile mass blocks is limited by a pair of stationary nuclear mass blocks which when in contact with mobile mass blocks will also become super critical and thereby repulsing mobile mass blocks to first position. 3. A self pulsating nuclear reactor of claim 1 wherein an energy cell contains a cathode in a form of a moderator circulation tank which is surrounded by a positive coil submerged in a solution which is able to absorb and emanate electro magnetic energy. 4. A self pulsating nuclear reactor of claim 3 wherein an independent valve body with several openings which can be mounted or dismounted on top of the reactor to regulate pump and suction action of the working fluid inside reactor into a steady pressure flow in one direction. 5. A self pulsating nuclear reactor of claim 4 wherein a pair of nuclear mass blocks acting as pistons to circulate a fluid by radial pendant motion whereby fluid is extracted from a storage tank into a pump cylinder so when nuclear mass blocks move away from each other meaning from vertical position towards horizontal position, said motion creates suction action, when the motion is reversed the fluid inbetween the mass blocks is then channeled through a series of one way valves and channels behind the mobile mass blocks, whereby when the nuclear mass blocks move outwards towards horizontal position the fluid trapped between the mass blocks and cylinder housing becomes pressurized and thereby accelerated through a series of one way valves and channels into a main pressure pipe to do physical work. |
abstract | A traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. In a traveling wave nuclear fission reactor, a nuclear fission reactor fuel assembly comprises a plurality of nuclear fission fuel rods that are exposed to a deflagration wave burnfront that, in turn, travels through the fuel rods. The excess reactivity is controlled by a plurality of movable neutron absorber structures that are selectively inserted into and withdrawn from the fuel assembly in order to control the excess reactivity and thus the location, speed and shape of the burnfront. Controlling location, speed and shape of the burnfront manages neutron fluence seen by fuel assembly structural materials in order to reduce risk of temperature and irradiation damage to the structural materials. |
|
052157040 | claims | 1. A method of determining the extent of fouling in a heat exchanger including a plurality of tubes, the tubes of said heat exchanger being fouled to some degree, said method comprising the steps of: connecting a test reservoir to an inlet port and an outlet port of one of said tubes, said one of said tubes remaining in place in said heat exchanger; circulating a service fluid from said test reservoir through said one of said tubes; altering the heat content of said service fluid in said test reservoir; and detecting (a) the temperature of said service fluid at the respective ends of said tube, (b) the flow rate of said service fluid, and (c) the temperature of a process fluid on the outside of said tube; using the values of said items (a), (b) and (c) to compute a heat transfer efficiency for said one of said tubes; using the heat transfer efficiency of said one of said tubes to determine the extent of fouling of the tubes in said heat exchanger. a heat exchanger for removing heat from a heated fluid in the event of an accident in said nuclear power plant, said heat exchanger comprising a plurality of tubes, said heat exchanger transferring a substantially lower quantity of heat when said nuclear power plant is in normal operation than said heat exchanger is required to transfer in the event of an accident in said nuclear power plant; apparatus conducting a test to determine the extent of fouling of said heat exchanger, said apparatus comprising; 2. The method of claim 1, said method being used to determine the extent of fouling of a heat exchanger in a nuclear power plant, said heat exchanger being intended to remove heat from said process fluid in the event of an accident in said nuclear power plant, said heat exchanger transferring a substantially lower quantity of heat when said nuclear power plant is in normal operation than said heat exchanger is required to transfer in the event of an accident in said nuclear power plant. 3. The method of claim 1 comprising the step of determining the fouling resistance (r.sub.f) of said one of said tubes. 4. A method of determining the extent of fouling of a heat exchanger comprising performing the method of claim 1 on a plurality but substantially less than all of the tubes in a heat exchanger. 5. The method of claim 4 comprising performing the method of claim 1 on at least six tubes but less than 10% of the total number of tubes in said heat exchanger. 6. The method of claim 5 wherein said tubes are selected randomly. 7. A combination in a nuclear power plant, said combination comprising: 8. The arrangement of claim 7 comprising a computer, said detection means being connected to said computer. 9. The arrangement of claim 7 wherein said means for exchanging the heat content of said service fluid comprises a heater. 10. The arrangement of claim 7 wherein said means for exchanging the heat content of said service fluid comprises a chiller. 11. The arrangement of claim 7 wherein said service fluid and said process fluid are liquids. 12. The arrangement of claim 7 wherein said apparatus is configured so as to be disconnected from said test exchanger when said test has been completed. |
abstract | In a method of interactive manipulation of the dose distribution of a radiation treatment plan, after an initial candidate treatment plan has been obtained, a set of clinical goals are transferred into a set of constraints. Each constraint may be expressed in terms of a threshold value for a respective quality index of the dose distribution. The dose distribution can then be modified interactively by modifying the threshold values for the set of constraints. Re-optimization may be performed based on the modified threshold values. A user may assign relative priorities among the set of constraints. When a certain constraint is modified, a re-optimized treatment plan may not violate those constraints that have priorities that are higher than that of the modified constraint, but may violate those constraints that have priorities that are lower than that of the modified constraint. |
|
abstract | The present invention relates to a connecting apparatus for a steam generator disposed between a steam generator and a flow mixing header to fasten the steam generator to the flow mixing header in a sealed manner, and an integral reactor including the same. Fastening the steam generator to the flow mixing header in a sealing manner includes: a base plate mounted on the flow mixing header having a through hole formed at the center thereof; and a steam generator connecting portion protruding along the circumference of the through hole in the base plate allowing an outlet of the steam generator to be inserted and fastened thereto. Since the connection for the steam generator is tightly fastened to the flow mixing header, leakage of a coolant therebetween is prevented, and since the steam generator is horizontally disposed in the flow mixing header, structural stabilization may be achieved. |
|
description | The present application claims the benefit of U.S. Provisional Patent Application Ser. No. 62/029,931, filed Jul. 28, 2034, the disclosure of which is incorporated herein by reference in its entirety. The field of the present invention relates to racks for the storage of nuclear fuel, and particularly for storing spent nuclear fuel. In the nuclear power industry, the nuclear energy source is in the form of hollow zircaloy tubes filled with enriched uranium, known as fuel assemblies. Upon being depleted to a certain level, spent fuel assemblies are removed from a reactor. At this time, the fuel assemblies not only emit extremely dangerous levels of neutrons and gamma photons (i.e., neutron and gamma radiation) but also produce considerable amounts of heat that must be dissipated. It is necessary that the neutron and gamma radiation emitted from the spent fuel assemblies be adequately contained at all times upon being removed from the reactor. It is also necessary that the spent fuel assemblies be cooled. Because water is an excellent radiation absorber, spent fuel assemblies are typically submerged under water in a pool promptly after being removed from the reactor. The pool water also serves to cool the spent fuel assemblies by drawing the heat load away from the fuel assemblies. The water may also contain a dissolved neutron shielding substance. The submerged fuel assemblies are typically supported in the fuel pools in a generally upright orientation in rack structures, commonly referred to as fuel racks. It is well known that neutronic interaction between fuel assemblies increases when the distance between the fuel assemblies is reduced. Thus, in order to avoid criticality (or the danger thereof) that can result from the mutual inter-reaction of adjacent fuel assemblies in the racks, it is necessary that the fuel racks support the fuel assemblies in a spaced manner that allows sufficient neutron absorbing material to exist between adjacent fuel assemblies. The neutron absorbing material can be the pool water, a structure containing a neutron absorbing material, or combinations thereof. Fuel racks for high density storage of fuel assemblies are commonly of cellular construction with neutron absorbing plate structures (i.e., shields) placed between the storage cells in the form of solid sheets. For fuel assemblies that have a square horizontal cross-sectional profile, the storage cells are usually long vertical square tubes which are open at the top through which the fuel elements are inserted. In order to maximize the number of fuel assemblies that can be stored in a single rack, the fuel racks for these square tubes are formed by a rectilinear array of the square tubes. Similarly, for fuel assemblies that have a hexagonal horizontal cross-sectional profile, the storage cells are usually long vertical hexagonal tubes which are open at the top through which the fuel elements are inserted. For such storage cells, in order to maximize the number of fuel assemblies that can be stored in a single rack, the fuel racks for these hexagonal tubes are formed by a honeycomb array of the hexagonal tubes. Regardless of whether the storage cells are square tubes or hexagonal tubes, the storage cells of some fuel racks may include double walls that can serve two functions. The first function of a double cell wall may be to encapsulate neutron shield sheets to protect the neutron shield from corrosion or other deterioration resulting from contact with water. The second function of a double cell wall may be to provide flux traps to better prevent undesirable heat build-up within the array of storage cells. When both of these double-wall functions are incorporated into a fuel rack array, it necessarily decreases the storage density capability. Thus, improvements are desired in design a fuel racks that provide both these functions and improve the overall storage density capability. The present invention is directed to an apparatus for supporting spent nuclear fuel. Specifically, the apparatus enables the high density storage of spent nuclear fuel. In a first separate aspect of the invention, a fuel rack apparatus includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to the upper surface of the base plate in a side-by-side arrangement to form a rectilinear array of the storage tubes. Each of the storage tubes extends along a longitudinal axis and includes: a rectangular outer tube having an inner surface defining an inner cavity; a first chevron plate comprising a first wall plate and a second wall plate; and a second chevron plate comprising a first wall plate and a second wall plate. The first and second chevron plates are positioned in the inner cavity in opposing relation to divide the inner cavity into: (1) a first chamber formed between the first wall plate of the first chevron plate and a first corner section of the rectangular outer tube; (2) a second chamber formed between the second wall plate of the first chevron plate and a second corner section of the rectangular outer tube; (3) a third chamber formed between the first wall plate of the second chevron plate and a third corner section of the rectangular outer tube; (4) a fourth chamber formed between the second wall plate of the second chevron plate and a fourth corner section of the rectangular outer tube; and (5) a fuel storage cell having a hexagonal transverse cross-section and configured to receive a fuel assembly containing spent nuclear fuel. In a second separate aspect of the invention, a fuel rack apparatus for storing spent nuclear fuel includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to and extending upward from the upper surface of the base plate, the storage tubes arranged in a side-by-side arrangement to form an array of the storage tubes. Each of the storage tubes extend along a longitudinal axis and include: an outer tube having an inner surface defining an inner cavity; and an inner plate-assemblage positioned within the outer tube that divides the inner cavity into a plurality of interior flux trap chambers and a fuel storage cell. In a third separate aspect of the invention, a fuel rack apparatus includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to the upper surface of the base plate in a side-by-side arrangement to form a rectilinear array of the storage tubes. Each of the storage tubes extends along a longitudinal axis and includes: a rectangular outer tube having an inner surface defining an inner cavity; and a plurality of wall plates positioned in the inner cavity that divide the inner cavity into: (1) a first interior flux chamber formed between a first one of the wall plates and a first corner section of the rectangular outer tube; (2) a second interior flux chamber formed between a second one of the wall plates and a second corner section of the rectangular outer tube; (3) a third interior flux chamber formed between a third one of the wall plates and a third corner section of the rectangular outer tube; (4) a fourth interior flux chamber formed between a fourth one of wall plates and a fourth corner section of the rectangular outer tube; and (5) a fuel storage cell having a hexagonal transverse cross-section and configured to receive a fuel assembly containing spent nuclear fuel. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,” “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation unless explicitly indicated as such. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Moreover, the features and benefits of the invention are illustrated by reference to the exemplified embodiments. Accordingly, the invention expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features; the scope of the invention being defined by the claims appended hereto. Turning in detail to the drawings, FIG. 1 schematically shows a fuel rack 101, according to one embodiment of the invention, placed in a cooling pool 103 for the storage of spent nuclear fuel. As is known in the art, the cooling pool 103 may include treated water to aid in neutron absorption and heat dispersion, with examples including demineralized water and borated water. The fuel rack 101, as shown in FIG. 2, includes a rectilinear array of hexagonal fuel storage cells 105. The fuel rack 101 is a cellular, upright, prismatic module. The illustrated embodiment of the fuel rack 101 is specifically designed to accommodate hexagonal fuel assemblies, such as VVER 1000 fuel assemblies. To this extent, each fuel storage cell 105 of the fuel rack 101 also has a hexagonal cross-sectional profile so as to geometrically accommodate no more than a single hexagonal fuel assembly. In certain embodiments, the hexagonal cross-sectional profile of the storage cell 101 may have a shape that is other than a regular hexagon. It is to be understood that the concepts of the present invention can be modified to accommodate any shaped fuel assembly, including rectangular, octagonal, round, among others. The fuel rack 101 includes a base plate 111, support pedestals 131, and a plurality of storage tubes 151 placed together in a side-by-side arrangement to form a rectilinear array as shown in FIG. 3A. The support pedestals 131 are affixed to a bottom surface 113 of the base plate 111, and the array of storage tubes 151 are affixed to the top surface 115 of the base plate 111 in a substantially vertical orientation. Each storage tube 151 extends along its own longitudinal axis LA, and in addition to being substantially vertical, each longitudinal axis LA is also substantially perpendicular to the top surface 115 of the base plate 111. The connection between each of the storage tubes 151 and the base plate 111 is achieved by welding the bottom edge of each of the storage tubes 151 to the top surface 115 of the base plate 111. Similarly, the connection between each of the support pedestals 131 and the base plate 111 is achieved by welding each of the support pedestals 131 to the bottom surface 113 of the base plate 111. By welding the storage tubes 151 to the base plate 111, the flexural strength of the base plate 11 may be increased, thereby making it possible to support the combined weight of the fuel rack and fuel assemblies with the support pedestals 131 located only near the edges of the base plate 111. Of course, other connection techniques can be utilized for either or both of the storage tubes 151 and the support pedestals 131 with minor modification, including mechanical connections such as bolting, clamping, threading, and the like. As shown in FIGS. 3A-D, the storage tubes 151 are connected to the base plate 111 to form a plurality of rows 153 and a plurality of columns 155. The storage tubes 151 within each row 153 are placed in a spaced apart manner, with the spacing between adjacent storage tubes 151 in a row 153 being maintained by spacers 157. Spacers 157 are placed between all adjacent storage tubes 151 within a row 153, with several spacers 157 being used to separate two adjacent storage tubes 151. The spacers 157 are welded in place to each of the adjacent storage tubes 151. Several spacers 157 are placed between each of the aligned longitudinal edges of adjacent storage tubes 151, with spacers 157 being placed at the top and bottom of aligned longitudinal edges, and the other spacers being spaced along the aligned longitudinal edges. The number of spacers 157 included between adjacent storage tubes 151 may vary depending on factors such as the desired fluid flow between adjacent storage tubes 151 and/or between adjacent columns 155, space considerations, and weight of the entire fuel rack, among other considerations. By having the spacers 157 distributed in this manner, the space between adjacent columns 155 forms flux traps 159, not only between adjacent ones of the storage tubes 151 within each row 153, but also between entire columns 155. These flux traps 159 are exterior to each of the storage tubes 151, and because the flux trap 159 of one row 153 is not partitioned from the flux trap 159 of an adjacent row 153, adjacent ones of the flux traps 159 effectively separate one column 155 from another. The width of the spacers 157, and thus the width of the flux traps 159, may be selected to tailor the ability to control criticality of the nuclear fuel stored within the fuel rack 101. The storage tubes 151 within each column 155 are placed adjacent each other so that the outer walls of adjacent storage tubes 151 within the respective column 155 are in surface contact with one another. Each of the aligned longitudinal edges of adjacent storage tubes 151 within a column 155 may be contiguously welded together to provide additional stability to the overall structure of the fuel rack 101. With the rectilinear array of the fuel rack 101 formed with the plurality of rows 153 and columns 155 as described above, the longitudinal axes LA of each of the storage tubes 151 in each of the rows 153 and in each of the columns 155 align to form reference planes RP. Also, the longitudinal axes LA of adjacent storage tubes 151 in one of the rows 153 may be separated from one another by a distance D1, and the longitudinal axes LA of adjacent storage tubes 151 in one of the columns 155 may be separated from one another by a distance D2, which may different, and even greater, than the distance D1. The distance D1 separating adjacent storage tubes 151 within a row 153 may be controlled within a design by appropriate selection of either the width of the storage tubes 151 or the width of the spacers 157. The distance D2 separating adjacent storage tubes 151 within a column 155 may be controlled within a design by appropriate selection of the length of the storage tubes 151. An exemplary storage tube 151 is shown in FIG. 4A. The storage tube 151 includes an outer tube 161 having a rectangular cross-section, as can be seen in FIG. 4B. The top end of the storage tube 151 remains open so that a fuel assembly can be inserted into the hexagonal fuel storage cell 105 formed therein. The storage tube 151 includes a first pair of opposing wall plates 163, 165 and a second pair of opposing wall plates 167, 169. The outer walls of the first pair of wall plates 163, 165 are placed into surface contact with respective outer walls of wall plates 163, 165 of adjacent storage tubes 151 to form the columns 155 of the rectilinear array, as discussed above. The storage tube 151 defines a longitudinal axis LA, which is the center point of the rectangular cross-section, and the wall plates 163, 165, 167, 169 each have an overall height H1. The top of each of the second pair of opposing wall plates 167, 169 includes a guide plate 171. The guide plate 171 for each wall plate 167, 169 extends at an angle up from the respective wall plate 167, 169 and away from the longitudinal axis LA of the storage tube 151. The guide plates 171 provide a surface to aid in guiding a fuel assembly into the fuel storage cell 105 formed within the storage tube 151. The guide plates 171 also help reduce the amount of wear and/or damage caused to the top edge of the wall plates 167, 169 during the process of loading a fuel assembly into the fuel storage cell 105. The guide plates 171 may be integrally formed with the wall plates 167, 169, or they be mounted as part of a separate structure to the external walls of the wall plates 167, 169. The outer walls of the second pair of opposing wall plates 167, 169 each have a neutron-absorbing plate 173 coupled thereto, and the neutron-absorbing plate 173 is secured in place against the outer walls of the second pair of opposing wall plates 167, 169 by an outer sheath 175. The outer sheath 175 encloses the neutron-absorbing plate 173 in a pocket 177, which is also shown in FIG. 4C, to protect the pool water from possible deterioration of the neutron-absorbing plate 173. The neutron-absorbing plate 173 and the outer sheath 175 extend a height H2, which is less than the height H1. The height H2 may be the equivalent of the height of a fuel assembly positioned for storage within the fuel storage cell 105. Of course, the height H2 of the neutron-absorbing plate 173 and the outer sheath 175 may, in certain embodiments, be as great as the height H1 of the outer tube 161. An inner plate-assemblage 191 is positioned within the outer tube 161 to help form the fuel storage cell 105. The inner plate-assemblage 191 includes two chevron plates 193a, 193b, which may be of identical design. An exemplary chevron plate 193, representative of both chevron plates 193a, 193b, is shown in FIG. 3B. The chevron plate 193 includes two wall plates 195 adjoined at an apex edge 197, and each wall plate 195 may have a height H3, which is slightly less than the height H1 of the wall plates 163, 165, 167, 169 of the storage tube 151. The top of each wall plate 195 includes a guide plate 199. The guide plate 199 for each wall plate 195 extends at an angle up from the respective wall plate 195, such that when the chevron plate 193 is in place within the outer tube 161 of the storage tube 151, the guide plates 199 also extend away from the longitudinal axis LA of the storage tube 151. The guide plates 199 provide a surface to aid in guiding a fuel assembly into the fuel storage cell 105 formed within the storage tube 151. The guide plates 199 also help reduce the amount of wear and/or damage caused to the top edge of the wall plates 195 during the process of loading a fuel assembly into the fuel storage cell 105. The guide plates 199 may be integrally formed with the wall plates 195, or they be mounted as part of a separate structure to the external walls of the wall plates 195. The outer walls of the wall plates 195 each have a neutron-absorbing plate 201 coupled thereto, and the neutron-absorbing plate 201 is secured in place against the outer walls of the wall plates 195 by an outer sheath 203. Each outer sheath 203 encloses the respective neutron-absorbing plate 201 in a pocket 205, which is also shown in FIG. 4C, to protect the pool water from possible deterioration of the neutron-absorbing plate 201. The neutron-absorbing plate 201 and the outer sheaths 203 extend a height H2, which is less than the height H3 of the wall plates 195. The height H2 may be the equivalent of the height of a fuel assembly positioned for storage within the fuel storage cell 105. Of course, the height H2 of the neutron-absorbing plate 201 and the outer sheaths 203 may, in certain embodiments, be as great as the height H3 of the wall plates 195. The dimension and position of the neutron-absorbing plate 173 on the wall plates 167, 169 of the outer tube 161, and the neutron-absorbing plate 201 on the wall plates 195 of the chevron plates 193, may be determined by the position and dimension of a fuel assembly positioned for storage within the fuel storage cell 105, and more particularly by the position and dimension of fuel rods contained within any such fuel storage assembly. The neutron-absorbing plates 173, 201 are generally placed on the respective wall plates 167, 169, 195 and dimensioned so that the height H2 is at least as great as the height of stored fuel rods within the fuel storage cell 105. Such dimensioning of the neutron-absorbing plates 173, 201 helps ensure that neutron emissions, directed toward any of the wall plates 167, 169, 195 from the fuel assembly within the fuel storage cell 105, are incident on the neutron-absorbing plates 173, 201. The outer sheaths 175, 203 on the wall plates 167, 169, 195 are dimensioned to provide a sufficiently large enclosure to secure the neutron-absorbing plates 173, 201 to the respective wall plates 167, 169, 195. The neutron-absorbing plate 173, 201 may be formed of a material containing a neutron absorber isotope embedded in the microstructure, such as elemental boron or boron carbide. Metamic, produced by Metamic, LLC, which is made of an aluminum alloy matrix with embedded boron carbide, is an example of an acceptable material. In certain embodiments, the outer sheaths 175, 203 may be formed of materials such as stainless steel, borated stainless steel, or any other type of steel appropriate for use in the long term storage environment for spent nuclear fuel. In certain embodiments, particularly those in which the neutron-absorbing plates 173, 201 are not formed of a material which is brittle or becomes brittle over time, thereby presenting a risk of deterioration and contamination of the pool water, the neutron-absorbing plates 173, 201 may be secured directly to the respective wall plates 167, 169, 195. In such embodiments, the outer sheaths 175, 203 may be omitted, or alternatively, the outer sheaths 175, 203 may be configured to couple the neutron-absorbing plates 173, 201 to the respective wall plates 167, 169, 195 without enclosing the neutron-absorbing plates 173, 201 in an envelope. FIG. 4C shows a cross-section of an exemplary storage tube 151. The outer tube 161 has a width W in the row direction and a length L in the column direction, and the length L in the column direction is greater than the width w in the row direction. The inner surface 211 of the outer tube 161 of the storage tube 151 defines an inner cavity 213, and a hexagonal fuel storage cell 105 is formed within the inner cavity 213 of the storage tube 151. The profile of a hexagonal fuel assembly 109 is shown for reference within the fuel storage cell 105. In certain embodiments, the gap between the fuel assembly 109 and the walls forming the fuel storage cell 105 is less than about 4 mm around all sides of the fuel assembly 109. The inner plate-assemblage 191 is positioned within the outer tube 161 to divide the inner cavity 213 into a plurality of interior flux trap chambers 215a-d and the fuel storage cell 105. In the rectilinear array of the storage tubes 151, these flux trap chambers 215a-d serve as interior flux trap chambers between the fuel storage cells 105 of adjacent storage tubes 151 in the fuel rack 101. Thus, storage tubes 151 that are adjacent within a row have their respective fuel storage cells 105 separated by four flux trap chambers, two from each of the adjacent storage tubes 151. The inner plate-assemblage 191 includes two chevron plates 193a, 193b. Each chevron plate 193a, 193b includes two wall plates 195a-d, and each wall plate 195a-d is oblique to and extends between adjacent sides of the outer tube 161 to form the plurality of interior flux trap chambers 215a-d within the inner cavity 213. Specifically, the wall plate 195a of the chevron plate 193a extends between the wall plate 167 of the outer tube 161 and the wall plate 163 of the outer tube 161 to form the interior flux trap chamber 215a. With the wall plate 195a positioned in this manner, the interior flux trap chamber 215a is formed between the wall plate 195a of the chevron plate 193a and a corner section formed at the intersection of wall plates 163, 167 of the outer tube 161. The wall plate 195b of the chevron plate 193a extends between the wall plate 169 of the outer tube 161 and the wall plate 163 of the outer tube 161 to form the interior flux trap chamber 215b. With the wall plate 195b positioned in this manner, the interior flux trap chamber 215b is formed between the wall plate 195b of the chevron plate 193a and a corner section formed at the intersection of wall plates 163, 169 of the outer tube 161. The wall plate 195a and the wall plate 195b are joined at an apex edge 197a of the chevron plate 193a. The edges of the wall plates 195a, 195b that are positioned against the wall plates 167, 169, respectively, are contiguously welded to the inner surface 211 of the rectangular outer tube 161. Similarly, the wall plate 195c of the chevron plate 193b extends between the wall plate 169 of the outer tube 161 and the wall plate 165 of the outer tube 161 to form the interior flux trap chamber 215c. With the wall plate 195c positioned in this manner, the interior flux trap chamber 215c is formed between the wall plate 195c of the chevron plate 193b and a corner section formed at the intersection of wall plates 165, 169 of the outer tube 161. The wall plate 195d of the chevron plate 193b extends between the wall plate 167 of the outer tube 161 and the wall plate 165 of the outer tube 161 to form the interior flux trap chamber 215d. With the wall plate 195d positioned in this manner, the interior flux trap chamber 215d is formed between the wall plate 195d of the chevron plate 193b and a corner section formed at the intersection of wall plates 165, 167 of the outer tube 161. The wall plate 195c and the wall plate 195d are joined at an apex edge 197b of the chevron plate 193a. The edges of the wall plates 195c, 195d that are positioned against the wall plates 167, 169, respectively, are contiguously welded to the inner surface 211 of the rectangular outer tube 161. With this configuration of the chevron plates 193a, 193b within the outer tube 161, the hexagonal fuel storage cell 105 is defined by: the inner surface 217a of the first wall plate 195a of the first chevron plate 193a; the inner surface 217b of the second wall plate 195b of the first chevron plate 193a; the inner surface 217c of the first wall plate 195c of the second chevron plate 193b; the inner surface 217d of the second wall plate 195d of the second chevron plate 193b; a portion of the inner surface 211 of the wall plate 167 of the outer tube 161; and a portion of the inner surface 211 of the wall plate 169 of the outer tube 161. Each of the flux trap chambers 215a-d formed by this configuration of the chevron plates 193a, 193b have triangular transverse cross-sections. The size and hexagonal cross-sectional shape of the fuel storage cell 105 is designed and constructed so that the fuel storage cell 105 can accommodate no more than one fuel assembly 109. Due to the different cross-sectional shape of the flux trap chambers 215a-d, as compared to the cross-sectional shape of the typical fuel storage assembly, the flux trap chambers 215a-d are not able to accommodate a fuel assembly that has a square or hexagonal transverse cross-section. The apex edges 197a, 197b of each of the chevron plates 193a, 193b are located in a reference plane RP that is defined by including the longitudinal axis LA of the storage tube 151 and being perpendicular to the wall plates 163, 165 of the outer tube 161. The apex edges 197a, 197b may form an angle of 120°, so that the resulting hexagonal cross-sectional shape of the fuel storage cell 105 forms a regular hexagon. In alternative embodiments, the apex edges 197a, 197b may form an angle α of slightly less than 120°, within the range of about 120°-115°, so that the resulting hexagonal cross-sectional shape of the fuel storage cell 105 varies slightly away from the form of a regular hexagon. When the hexagonal fuel assembly is placed within the fuel storage cell 105, the fuel assembly may rattle undesirably during a seismic or other rattling event. By having the apex edges 197a, 197b forming an angle of slightly less than 120°, the acute edges of the fuel assembly that face the apex edges 197a, 197b are prevented from impacting the apex edges 197a, 197b during a seismic or other rattling event. A cross-section of the storage tube 151 is shown in FIG. 4D with a schematic representation of a fuel assembly 109 disposed within the fuel storage cell 105. Similar to hexagonal fuel assemblies commonly in use, the fuel assembly 109 includes a top handle 233, a body portion 235, in which a plurality of nuclear fuel rods (not shown) are housed, and a tapered bottom portion 237. The handle 233 and the tapered bottom portion 237 facilitate inserting the fuel assembly 109 into the fuel storage cell 105 of the storage tube 151. When the fuel assembly 109 is being inserted into the storage tube 151, the tapered bottom portion 237 may engage the guide plates 171, 199 to aid in centering the fuel assembly 109 within the fuel storage cell 105. As shown, with the fuel assembly 109 fully inserted into the fuel storage cell 105, the height H1 of the outer tube 161 is greater than the overall height H4 of the fuel assembly 109. The height H3 of the chevron plates 193a, 193b is also less than the height H1 of the outer tube 161. The lower edges of the chevron plates 193a, 193b do not extend to the lower edge of the outer tube 161, so that a gap is formed at the lower end of the storage tube 151 for cooling fluid to flow into the flux trap chambers 215a-d. In certain embodiments, the chevron plates 193a, 193b may include apertures at their bottom edges for cooling fluid to flow into the flux trap chambers 215a-d, and in such embodiments, the height H3 of the chevron plates 193a, 193b may be the same as the height H1 of the outer tube 161. The height H2 of the neutron-absorbing plates 201 coupled to the chevron plates 193a, 193b (and the neutron-absorbing plates 173 coupled to the outer tube 161 as shown in FIG. 4C) is substantially the same as the height of the body portion 235 of the fuel assembly 109. In certain embodiments, the height H2 of the of the neutron-absorbing plates 201 (and 173) may be less than the height of the body portion 235 of the fuel assembly 109. The height H2 of the neutron-absorbing plates 201 (and 173) may be designed to provide appropriate shielding of adjacent fuel assemblies from one another. This is because adjacent spent nuclear fuel rods may not extend the entire length of the body portion 235 of the fuel assembly 109, and the height of the neutron-absorbing plates 201 (and 173) need only be high as the nuclear fuel rods when the fuel assembly 109 is positioned within the storage tube 151. The base plate 111, which is shown in FIG. 5, includes a plurality of flow holes 117 extending through the base plate 111 from the bottom surface 113 to the top surface 115. The base plate 111 also includes four oblong holes 119 (second row in from the corners) for lifting and installing the fuel rack 101 within the fuel pool 103. Typically, a special lifting beam with four long reach rods is used to interact with the oblong holes 119 to grapple the fuel rack 101 for transfer into or out of, or movement within, the pool 103. The flow holes 117 (and oblong holes 119) create passageways from below the base plate 111 into the bottom ends of the fuel storage cells 105 formed by the storage assemblies 151. As shown, a single flow hole 117 is provided for each storage assembly 151. In certain embodiments, multiple flow holes 117 may be provided for each storage assembly 151 to provide cooling fluid to the fuel storage cell 105 and each of the flux trap chambers 215a-d. The flow holes 117 serve as fluid inlets to facilitate natural thermosiphon flow of pool water through the fuel storage cells 105 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the fuel storage cells 105 in a submerged environment, the water within the fuel storage cells 105, and within the flux trap chambers 215a-d, surrounding the fuel assemblies becomes heated, thereby rising due to increased buoyancy. As this heated water rises and exits the storage assemblies 151 via their open top ends, cool water is drawn into the bottom of the fuel storage cells 105 and the flux trap chambers 215a-d via the flow holes 117. This heat induced water flow along the fuel assemblies then continues naturally. A support pedestal 131 for the fuel rack 101 is shown in FIG. 6. The support pedestals 131 affixed to the bottom surface 113 of the base plate 111 ensure that a space exists between the floor of the pool 103 and the bottom surface 113 of the base plate 111, thereby creating an inlet plenum for water to flow through the flow holes 117. The support pedestal 131 includes a base portion 133 and a riser portion 135 formed about an interior flow space 139. The riser portion 135 includes flow apertures 141 through which water from the pool 103 may pass from a space external to the support pedestal 131 into the interior flow space 139. Water passing into the interior flow space 139 may then pass up through a flow hole 117 in the base plate 111 to enable the cooling process described above. Although the riser portion 135 is depicted as being annular, in certain embodiments the riser portion 135 may have any geometrical configuration which supports the base plate 111 above the floor of the pool 103 and permits water from the pool 103 to flow into any flow holes 117 in the base plate 111 near which the support pedestal 131 may be affixed. The fuel rack 101 described above with reference to FIGS. 1-6 is intended to be placed free standing in a pool 103, without being coupled to sides or the bottom of the pool. However, in certain embodiments, a coupler may be used to aid in securing the position of the fuel rack 101 within the pool 103 during a seismic or other rattling event. Other than the neutron absorbing material described above, the fuel rack may be formed entirely from austenitic stainless steel. Although other materials may be used, some materials, such as borated stainless steel, are not preferred for a free standing fuel rack 101 within a pool 103, as the greater weight of materials such as borated steel aggravate the seismic response of the fuel rack 101, thus forcing the fuel rack 101 to be anchored. An alternative embodiment of a fuel rack 301 is shown in FIG. 7. This fuel rack 301 includes a plurality of storage tubes 303 affixed to the top surface of a base plate 309, and support pedestals 311 affixed to the bottom surface of the base plate 309. The storage tubes 303 each include a fuel storage cell 305, and they are placed together in a side-by-side arrangement to form a plurality of rows 305 and a plurality of columns 307 as part of a rectilinear array, in the manner described above. A plurality of auxiliary flow apertures 313 are included in the storage tubes 303 at or near their bottom edges. In certain embodiments, at least one auxiliary flow aperture 313 is included in each face of the storage tubes 303, even those faces of storage tubes 303 that are placed in surface contact with the face of an adjacent storage tube 303. The auxiliary flow apertures 313 act as additional inlet openings (when combined with flow holes in the base plate 309) for incoming pool water to facilitate the thermosiphon flow during the cooling process. While an auxiliary flow aperture 313 is shown in each face of each and every storage tube 303 in the fuel rack 301, in certain embodiments the auxiliary flow aperture 313 may be omitted from a select subset of faces for select storage tubes 303. Various other modifications of the embodiments of the present invention will readily be apparent to those skilled in the art and are encompassed within the scope of the invention as defined in the appended claims. |
|
summary | ||
051788233 | description | GENERAL DESCRIPTION OF THE INVENTION Schematically shown in FIG. 1 are the operational units of this invention consisting of a first housing 10, a mobile platform 11, a carriage 12, and a cart 13. Each of these units are capable of being moved independently relative to each other, while connected together for synchronized operation as hereinafter made apparent. The housing 10 contains the fluid heating unit and pressurizing pump by which optimum heated pressurized fluid is produced and entrained into a supply line 14 leading to a suitable discharge tool not shown. Within the housing 10 are conventional heat and pressure safety cut-out switches all of which are mounted upon a convenient control panel 15 displayed to an operator of the apparatus. To better understand the operation of the apparatus the cooperative units will be described in the order of their relation to the vacuum recovered contaminated cleaning fluid. As such it is shown that the cart 13 supports the first stage of the separating and filtering recovery system. A liquid separator 16 is mounted on the cart 13 and consists of a conical liquid hopper 17 that has an internal construction by which is critically safe with respect to the radioactive mass it initially recovers from the cleaned decontaminated surfaces. This critical safe feature in the liquid hopper 17 is achieved by providing the hopper 17 with a volume reducing filler core 18 having the same conical shape as the hopper 17. A tangently placed vacuum induced intake 19, connected by a suitable hose to a remote cleaning tool or head not shown, has open communication with the area exposed between the inner wall of the hopper 17 and the volume reducing core 18. As such the vacuum induced flow of recovered contaminated material is caused to spiral within the hopper 17 in an agitated manner so as to separate the liquid from the recovered contaminates causing the same to be deposited at the apex 20 of the hopper 17 for forced removal by a pump 21 through a suitable discharge 22. As shown the core 18 is provided with a center tube 23, the free lower end of which is disposed in spaced relation to the apex 20 of the hopper 17. The opposite end of the tube 23 has open communication with the bottom of a chamber 24 formed in the center of the core 18. A vacuum induced discharge tube 25 is carried by the cover 26 of the hopper 17 and has one open end 27 thereof disposed within the chamber 24. By a suitable wicker basket type container 28 a liquid level cut-off ball 29 is movably positioned relative to the open end 27 of the discharge tube 25. By this arrangement the vacuum induced intake of the hopper will be disrupted at any time the liquid level in the hopper 18 reaches a volume that has been predetermined as a critical mass of radioactive material. Without the safety cut-off and the volume reducing core 18 a critical mass of radioactive material by volume could accumulate in this first recovery stage with hazardous consequences. The moisture laden contaminates which have been separated from the liquid by the first stage separator will exit through the discharge tube 25 and be vacuum induced into the second stage demister filter unit 30, mounted on the carriage 12, as shown in FIG. 6. Unit 30 consists of a container 31 compartmentalized as at 32 and 33. These compartments 32 and 33 provide open tops which are adapted to be closed by a cover member 34, while each of their bottom portions are open to provide unrestricted communication with a liquid collection tank 35. The cover member 34 provides a pair of spaced truncated risers 36 and 37 for closing the top portion of each compartment. The riser 36 provides an inlet port 38 while the riser 37 provides a discharge port 39. The tapered walls of the riser 36 functions as a deflector against the vacuum drawn moisture laden fluids exhausted from the liquid separator 16, which are deflected into a downward path into the compartment 32. Within the compartment 32 and supported upon a set of rails is a demister 40. This demister 40 will coagulate the larger particles of contaminates into liquid particles. These liquid particles will be carried by the vacuum in a downward direction where they will impinge upon one tapered wall or baffle 41 of the collection tank 35 and into the fluid reservoir provided therein. Any remaining air borne contaminates will be drawn over the top of the collected liquids in the tank 35 and be deflected by the opposite tapered wall 42 into an upward path through the compartment 33. The upward path of the air flow will be drawn through a high efficiency air particle filter 43. The now demisted and filtered air will continue in an upward path until it impinges upon the tapered walls of the riser 37 and discharged through the discharge port 39. Each of the compartments 32 and 33 are readily assessable though removable side walls 44 and 45 whereby the demister 40 and the filter 43 may be readily replaced as needed. The collector tank 35 is provided with a discharge pump 46 by which the collected contaminated liquid may be discharged therefrom for safe disposal. The collector tank 35 may include a float switch 47 by which the volume of the radioactive contaminated liquid collected therein may be controlled against a critical mass criteria. A safe amount of filtered liquid can be discharged through an exhaust nozzle 48. Referring to FIG. 1 there is illustrated a vacuum creating apparatus 49 mounted on the mobile platform 11. The vacuum creating apparatus includes a liquid ring pump 50 providing a manifold 51 which includes an intake port 52 which by a suitable hose 53 has open communication with the discharge port 39 of the demister filter unit 30. Not shown the manifold 51 provides communication with the final liquid stage recovery tank 54. Essentially the working parts of the liquid ring pump 50 consists of a multibladed impeller eccentrically mounted in a round casing 55 which provides a liquid well that is partially filled. As the impeller blades are caused to rotate through energization of an electric motor 56, the liquid in the well is drawn by centrifical force created by the rotating blades, to form a liquid ring which is concentric with the casing 55. The space between the impeller blades will fill with liquid during their rotation and air trapped therein is compressed and discharged thus creating a vacuum. The liquid pump 50 is electrically controlled and is in circuit with the float switch 47 of the demister filter unit 30 whereby when the recovered liquid as collected in tank 35 reaches a predetermined level, indicating a critical mass collection, it will de-energize the liquid ring pump 50 terminating the created vacuum recovery flow through the entire apparatus. Utilization of the decontamination apparatus of this invention removes decontamination at its source. The apparatus provides an unique three stage decontamination function with each stage providing an independent safety control system against critical mass build up. The functional units of this apparatus, wherein critical mass build up is susceptible, each contain an independent control for deactivating the recovery vacuum flow throughout all interfaced functional units of the apparatus. These safety features make this decontamination apparatus particularly useful in the decontaminating of radioactive contaminated surfaces. While I have illustrated and described the preferred form of construction for carrying my invention into effect, this is capable of variation and modification without departing from the spirit of the invention. I therefore, do not wish to be limited to the precise details of construction as set forth, but desire to avail myself of such variations and modifications as come within the scope of the appended claims. |
abstract | A method of drying high-level radioactive wastes and device thereof contains a suspension mechanism for hanging a manual elevating mechanism, and a shielding cover. The manual elevating mechanism couples with a basket for accommodating wastes by using a hanging rope, and the shielding cover is fixed below the suspension mechanism and in a moving path of the basket. The basket is moved to a storage tank containing water in which radioactive wastes are stored, and the radioactive wastes are pumped into the basket by means of a pump. The basket is then lifted above a water surface of the storage tank and received in the shielding cover, and then the shielding cover is moved into a water holder so as to drain waters in the basket. The basket is further moved onto a heating seat to be heated and a vacuum equipment is started to dry the radioactive wastes. |
|
051732513 | summary | The present invention relates to a mixing apparatus for a plurality of turbulently flowing fluid flows of varying temperature and/or composition, having a mixing chamber with a predetermined cross section, a straight or a curved center line and an outlet opening for the exit of a mixed fluid flow. Such a mixing apparatus is of particular importance for large throughput quantities at low pressure loss, for instance with a straight center line for heating or air conditioning systems, or with a circular center line for a gas-cooled high-temperature nuclear reactor through which helium flows from top to bottom. In such high-temperature nuclear reactors, in which the core container is made up of graphite blocks and the mean gas exit temperature is approximately 950.degree. C., the local maximum-possible gas temperatures must not be more than slightly above the mean value, because the temperature resistance of the metal components connected to it is limited to approximately 1000.degree. C. Moreover, changing temperatures put added stress on structural parts, yet for the sake of the service life of the parts the stress should be as slight as possible. An attempt is therefore made to operate such reactors with local temperature differences that are as slight as possible over time. British Patent No. 1,183,355 describes an apparatus for mass and heat transfer between solid or liquid particles and gases that has a rotationally symmetrical vortex chamber. The chamber has a plurality of inlet conduits that are staggered in the longitudinal direction of a swirl chamber, and because of their tangential inflow, they induce a rotary flow with freely floating rotating rings of particles in the swirl chamber. The gas, having been cleaned of particles, leaves the swirl chamber through an outlet. The particles themselves are carried away through a different outlet. It is not the object of such an apparatus to mix a plurality of fluid flows. German Patent DE 25 06 293 C2 describes a gas-cooled high-temperature nuclear reactor in which direct neutron irradiation from the hot-gas collecting chamber into the hot-gas lines and an attendant activation of the metal fixtures in such lines are prevented. That is attained by providing a neutron shield, which is in the form of an annular wall in the hot gas collecting chamber and which extends over its entire height. The shield is coaxial with the reactor cavern and spaced apart from it by a distance, in such a way that the resultant annular chamber is dimensioned adequately for the coolant gas flow, and the annular wall has a plurality of rows of slits for the passage of the coolant gas, with the slits being offset from the coolant gas outlet connection pieces. Another advantage of such a configuration is considered to be that forced guidance of the hot cooling gas, which improves the temperature distribution at the coolant gas outlet connection pieces, is attained by the way in which the annular wall and the predominantly radially extending slits are disposed. German Patent DE 27 42 847 C2, which also relates to a gas-cooled high-temperature nuclear reactor, achieves improved mixing of the hot coolant gas prior to its entry into the hot-gas conduits by means of high columns that have radial bores, which are disposed in a plurality of planes distributed over the entire length of the columns and through which the interior of the columns communicates with the hot-gas collecting chamber. Since the hot gas can flow out uniformly in all of the planes of the hot-gas collecting chamber, the development of laminations in the flow is avoided, and good mixing of the hot gas arriving from the various core regions is attained. An annular chamber with the structure described in the previous paragraph is not employed. German Published, Non-Prosecuted Patent Application DE 37 07 379 A1, which is again applicable for a gas-cooled high-temperature nuclear reactor, prevents the creation of hot-gas streams of varying temperature in a hot-gas collecting chamber by means of two vertical baffle walls protruding into the hot-gas collecting chamber. Once again, an annular chamber with numerous slits is not provided. In the sense of the present invention, "turbulent" describes fluid flows that definitely do not flow laminarly, or in other words have a Reynolds number of more than 5000, with respect to the hydraulic diameter of the mixing chamber. The Reynolds number Re is dimensionless and in fluids, it represents the ratio between the forces of inertia and the forces of viscosity. The hydraulic diameter of an arbitrary cross section equals the diameter of a circle of the same area. The "center line" in the sense of the present invention is defined as a line that joins the various centroids of the mixing chamber cross sections. It may be straight or curved. The "cross-sectional plane" of the mixing chamber is defined as a plane that is penetrated perpendicularly by the center line. A flow is defined as "tangential" if it enters the vortex produced in the mixing chamber at the circumference of the vortex, in the direction of rotation of the vortex. "Radial" is intended to pertain only to the curved center line. It is accordingly an object of the invention to provide a mixing apparatus for a plurality of turbulently flowing fluid flows of varying temperature and/or composition, having a mixing chamber of predefined cross section with a straight or curved center line and having an outlet opening for the exit of a mixed fluid flow, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type. A special object of this invention is to disclose a mixing apparatus for a gas-cooled high-temperature nuclear reactor with a radially and/or azimuthally different release of heat in the core. Hot and cold gas streams in the ensuing outflow line, which leads to one or more heat exchangers, should be avoided as a result. Gas streams that emerge from unavoidable gaps between graphite blocks in a high-temperature nuclear reactor or which are needed to cool reflector rods or a packed bed of the fuel assembly in a pebble discharge tube, should be intimately mixed with the primary gas flow so that the structural parts through which there is a subsequent flow are not threatened. With the foregoing and other objects in view, in order to attain the first object given above, there is provided, in accordance with the invention, a mixing apparatus for a plurality of turbulently flowing fluid flows varying in temperature and/or composition, comprising a mixing chamber having a non-circular or non-round cross section and a straight center line, a plurality of single-conduit and parallel deflector elements being disposed beside the mixing chamber and staggered in the direction of the center line, each of the deflector elements having means for receiving a fluid flow being oriented at an angle relative to the center line and staggered laterally and means for deflecting or diverting the fluid flow tangentially into the mixing chamber, and the mixing chamber having an outlet opening for exiting of a mixed fluid flow. With the objects of the invention in view, in order to attain the first object given above, there is also provided a mixing apparatus for a plurality of turbulently flowing fluid flows varying in temperature and/or composition, comprising a mixing chamber having a predetermined cross section and a curved center line, a plurality of single-conduit and radial deflector elements being disposed beside the mixing chamber and staggered in the direction of the center line, the deflector elements having means for receiving one or a plurality of fluid flows being oriented at an angle relative to the center line and staggered laterally and means for deflecting or diverting the at least one fluid flow tangentially into the mixing chamber, and the mixing chamber having an outlet opening for exiting of a mixed fluid flow. With these configurations, a macroscopic vortex of predetermined rotary direction is created in the mixing chamber. The vortex moves helically to the outlet opening, and in so doing assures good mixing of the fluid flows entering from the following deflection elements and is kept in rotation by the kinetic energy of the fluid flows arriving in staggered fashion in the direction of the center line. As a result of the deflection beside the mixing chamber, or in other words immediately before the entry into the mixing chamber, the fluid flows are preferentially oriented in the desired direction at a tangent with respect to the vortex. The deflector elements preferably have a rectangular conduit cross section. In accordance with another feature of the invention, the deflector elements have a curved wall in a cross-sectional plane of the mixing chamber. The wall of the deflector elements that is curved in the cross-sectional plane of the mixing chamber, preferably curved circularly, also serves to provide the desired tangential inflow and reduces the pressure loss in the deflection element, in comparison with a rectangular deflection. In accordance with a further feature of the invention, the mixing chamber has a given full height, and the mixing chamber is joined over the given full height to the deflector elements. If the mixing chamber is joined to the deflector elements at its full height, the pressure loss is likewise reduced, and if the wall is curved, the tangential inflow is promoted. In accordance with an added feature of the invention, the mixing chamber is horizontally oriented and has a top, and the deflector elements have vertical limitations penetrating part of the mixing chamber at the top. In accordance with an additional feature of the invention, the mixing chamber is horizontally oriented and has a cover with outer vertical openings partly penetrating the cover above the mixing chamber. Piercing of some deflector elements and part of the mixing chamber even prior to the deflection also promotes the creation and maintenance of the desired macroscopic vortex. However, this piercing should be present only in the peripheral region of the mixing chamber, so that the fluid flows there will also enter the vortex approximately at a tangent. In accordance with yet another feature of the invention, the mixing chamber has a rectangular cross section. The rectangular cross section of the mixing chamber according to the invention is important not only because it makes it easier to manufacture a large mixing apparatus, but also because there are considerable advantages for mixing of the fluid flows as well. In a conventional circular mixing chamber of the prior art, individual gas streams could propagate in a helical pattern without hindrance. However, in a mixing chamber with a rectangular cross section and a tangential inflow, the rotating vortex in its corners is constantly disturbed and braked, but is supplied with kinetic energy and kept in rotation by the fluid flows entering in staggered fashion. Other cross-sectional shapes, such as polygonal, rounded or oval ones, are also possible in principle. However, a circular cross-section undergoes less turbulence. In accordance with yet a further feature of the invention, the center line is curved and there is provided a distributor having the shape of a star or part of a star for distributing the fluid flows to a plurality of the radial deflector elements. A mixing apparatus with a circular or partially circular center line and with a distributor in the form of a star or part of a star, for distributing the fluid flows, which arrive in joined fashion at a ring or partial ring or circle or partial circle, to radial deflector elements, is advantageous as compared with a straight mixing chamber, because in this case the mixing apparatus can be made with a relatively small requisite amount of base area. In accordance with yet an added feature of the invention, the center line is curved and the deflector elements are widened in the radial direction toward the outside. Widening the deflector elements radially toward the outside takes into account the fact that the throughput of the fluid flows that arrive from a circular area, or the area of part of a circle, increases from the inside outward. The special object of the invention may also be attained for a gas-cooled nuclear reactor having a horizontal, annular mixing chamber beneath its core, which has radially disposed sectors and a radial outlet opening. The core of a high-temperature nuclear reactor is preferably constructed of spherical (pebble-type) or block-shaped fuel assemblies. In either case it is surrounded by a container that is formed of stacked graphite blocks in the vicinity of the core, which farther outward are insulated with coal stone, cooled, and held together by a steel container. If the core of the reactor is formed of pebble-type fuel assemblies, then the core container has a funnel-shaped bottom that is penetrated by numerous vertical gas conduits and ends in a circular pebble discharge conduit. The container and the bottom are formed of stacked graphite blocks, which necessarily have gaps between them, or else such gaps are even quite intentionally provided, for instance in order to cool the fuel assemblies located in the pebble discharge conduit or to cool absorber rods located in the blocks of the container. With the objects of the invention in view, in order to attain the special object of this invention for a gas-cooled high-temperature nuclear reactor, there is further more provided, in a gas-cooled, high-temperature nuclear reactor with a circular outline, a mixing apparatus for a plurality of turbulently flowing fluid flows varying in temperature and/or composition, comprising a horizontal, annular mixing chamber having a plurality of sectors, horizontal annular conduits for receiving at least one fluid flow, a plurality of vertical conduits being disposed above some of the sectors and having upper ends connected to the horizontal annular conduits and lower ends connected to the mixing chamber, an outlet opening communicating with the mixing chamber, and some of the sectors having no radial deflector elements but instead a plurality of bores formed therein in the vicinity of the outlet opening for receiving absorber material. In order to make it possible to provide approximately uniform shielding over its periphery against radioactive radiation outside the reactor, it is desirable to replace the shielding that is missing in the vicinity of the outlet opening with a very effective absorber material, such as boron carbide, in the adjacent graphite blocks. It is also logical for reasons having to do with mixing technology to leave out some of the radial deflector elements in the graphite blocks in the region of the outlet opening, so that hot or cold gas streams cannot reach the outflow line from there without having been adequately mixed. In order to enable the gas flows arriving in the region of the missing deflector elements to find some way to the outflow line, the vertical conduits present above the deflector elements are joined at their upper ends with the horizontal annular conduits, through which these gas flows can flow to other deflector elements. In accordance with another feature of the invention, one sector forms a plurality of deflector elements. In other words, for instance, two radial guide walls are each formed by a single sector-shaped graphite block between the radial deflector elements. This embodiment creates structural parts that can still be manufactured and transported because of their size, and that with a useful ratio of height to width can also be stacked one above the other without tipping over. With the blocks disposed below or above them, these blocks form mutually independent columns with gaps between them, for hindering differential expansions of adjacent columns in the vertical direction. In accordance with a further feature of the invention, in a gas-cooled high-temperature nuclear reactor on the pebble-bed principle with a central pebble discharge conduit, it is provided that the pebble discharge conduit is constructed by the sectors, i.e., the stacked graphite blocks. This avoids a separate self-contained pebble discharge tube. Due to the pressure differences prevailing during operation of such a reactor, the gaps present between the blocks assure a constant gas flow, which adequately cools the fuel assembly pebbles located in the pebble discharge conduit. Due to the ensuing good mixing according to the present invention, these gas flows are tolerable. In accordance with a concomitant feature of the invention, the graphite blocks have vertical bores for receiving absorber material in the vicinity of the pebble discharge conduit, so that there will be less power production in the highly radioactive fuel assemblies in the pebble discharge conduit. 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 mixing apparatus for fluids, 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. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
041347900 | summary | BACKGROUND OF THE INVENTION The present invention relates to nuclear fuel assemblies and more particularly to lock down devices for nuclear fuel assemblies. It is well known that nuclear power reactors typically contain a plurality of nuclear fuel assemblies arranged and supported between an upper core alignment plate and a lower core support plate. The upper and lower ends of the fuel assemblies are typically provided with end fittings which include alignment posts that extend outwardly from the end of the fuel assemblies and slidably engage post receiving openings in the plates. The weight of the fuel assemblies is typically borne by the core support plate. During operation of the reactor a fluid coolant such as water is forced vertically upward through the fuel assemblies to remove the heat generated therein. This upward flow produces a lifting force on the fuel assemblies which can exceed the weight of the fuel assembly itself. Consequently, various prior art devices have been used at the top or the bottom of the fuel assembly to prevent the assembly from lifting off the core support plate. If this upward motion of the fuel assembly is not prevented, damage to its fuel rods and to the upper alignment plate may result. In addition, lateral motion of the fuel assemblies can adversely affect the local power density in the core and can cause wear between the fuel rods and the grid structure of the fuel assembly. The devices which hold the fuel assemblies in place must also accommodate differential thermal expansion of each assembly and the various other core components. Thus the fuel assemblies are usually supported between the upper alignment plate and the core support plate in a manner which permits relative axial growth without overly stressing the components of the fuel assembly. Although, as described above, the fuel assembly must be held firmly in place during core operation, the periodic refueling of the reactor requires that the assembly be easily relocatable within the core when the reactor vessel head is unbolted and removed. One method of holding down fuel assemblies disclosed in the prior art uses a bayonet lock at the bottom of the center fuel element of a seven element hexogonal assembly. The lock is actuated by rotating the center fuel element. This device is impractical in modern reactors where closely packed square fuel assemblies do not provide clearance for rotation. Other bottom mounted locking devices permit simultaneous locking or unlocking of an entire row of fuel assemblies through the motion of an actuating rod that extends horizontally and in contact with a bottom extension of each fuel assembly. Each actuating rod is driven by means external to the reactor vessel thus requiring many penetrations in the lower portion of the reactor vessel. These penetrations are undesirable for reactors operating at the high pressures common in modern reactors, and external driving means require space around the reactor vessel which is not provided in current reactor cavity designs. In recent years the major nuclear power reactor suppliers in the United States have held assemblies in place with the use of various spring arrangements between the top of the fuel assembly and the upper core alignment plate. These designs typically require a substantial compressive pre-loading of the springs so that enough hold-down force is applied to the assemblies to resist the upward forces that exist during core operation. This pre-load force is transmitted from the top of the fuel assembly to the bottom of the assembly through the control element guide tubes. The guide tubes provide a path for insertion of the control rods, and the guide tubes and the fuel assembly spacer grids attached thereto provide the framework which maintains the proper spacing and alignment of the individual fuel rods in an assembly. In the past the control rod guide tubes were typically made from stainless steel, a material having the desirable characteristic of high compressive strength, but the undesirable characteristic of a high cross section for parisitic neutron absorption. In order to improve the neutron economy in the reactors, the guide tubes have more recently been fabricated from zircaloy. This material is less resistant to compressive stresses than is stainless steel. Consequently, there is a greater likelihood that the pre-loading of the fuel assemblies as described above can result in bowing of the guide tubes and as a result bowing of the entire fuel assembly. This problem has become more acute in recent years as the power densities and flow rates in the reactors have increased, thereby requiring larger hold-down springs and pre-loading forces. These top mounted springs have several other disadvantages. Their size not only contributes to the flow resistance offered by the fuel assembly, but also increases the required height of the reactor vessel by several inches and the cost of the vessel by several thousand dollars. In addition, since the pre-load force typically originates from the weight of the reactor vessel head and upper guide structure bearing down on the upper core alignment plate, if the total pre-load which the upper springs must provide is greater than the combined weight of the reactor vessel head and upper guide structure, special techniques are required for bolting down the reactor vessel head. A prior art improvement to the top mounted spring hold-down device moves the springs to the bottom of the fuel assembly where an upward compressive force is applied to the bottom of the control rod guide tubes. The compressive pre-load force on the guide tubes in this design is approximately equal to the weight of the fuel assembly in water, or about 1200 pounds. This is a significant reduction in compressive force relative to the upper spring design but the possibility of guide tube and fuel assembly bowing is still significant. The consequences of fuel assembly bowing can be quite severe. For example, the local power density surrounding individual fuel pins can be much higher than predicted under unbowed conditions. Also a bowed fuel assembly may not be relocatable because the distortions and non-rectangularity may prevent a proper fit in relation to adjacent assemblies. A further possibility is that the guide tube will be bowed enough to interfere with the dropping of a control rod therethrough. Another disadvantage of both the top mounted spring hold-down and the bottom mounted spring hold up devices disclosed in the prior art is the possibility of small but continual lateral motion of the end fittings and guide tubes during core operation. Since the fuel assembly spacer grids are attached to the guide tubes, these grids also will move laterally. The lateral motion of the grids can cause wear on the surface of the fuel rods which, after a year or two of operation, may significantly increase the susceptibility of the fuel rods to clad failure. Thus it can be seen that early fuel assembly hold-down devices included positive locking means at the bottom of the fuel assembly. However, as reactors became larger and more complex, space limitations and ease of refueling, in combination with the ability to accommodate different expansion rates of the various core components, led to the wide spread use of upper mounted spring hold-down devices and, more recently, lower mounted spring hold up devices. Although these latter devices adequately perform their intended hold-down function, they increase the possibility of problems resulting from compression of the control rod guide tubes and the small but continual lateral motion of the guide tubes. SUMMARY OF THE INVENTION It is an object of the present invention to provide an inexpensive apparatus for automatically securing a fuel assembly to the core support plate without compressively stressing the control rod guide tubes. Another object of the invention is to reduce the possibility of lateral motion in a fuel assembly secured to the core support plate. It is a further object of the invention to simplify the effort required to load and unload fuel assemblies into the reactor. The present invention employs spring-actuated latches between the fuel assembly lower alignment posts and the fuel assembly alignment pins on the core support plate to automatically lock the fuel assembly in place. Angles on the spring surfaces are selected so that the weight of the fuel assembly causes self-actuation of the device during insertion. The horizontal components of the spring load are used to restrain lateral motion of the fuel assembly. Removal of the fuel assembly is accomplished by applying an axial force with the refueling machine sufficient to overpower the spring load. Several advantages of the invention are immediately evident. The elements are inexpensive to manufacture and fabricate, yet provide simple, reliable operation. No vessel penetrations are needed and the locking and unlocking of the assembly can be accomplished with the typical refueling machine normally used for loading and unloading the assembly. The small size and vertical orientation of the elements require less space than most prior art devices and offer less flow restriction. In addition, material costs can be further reduced since the reactor vessel can be shortened by several inches relative to designs employing fuel assembly hold-down springs. Furthermore, the weight of the fuel assembly itself actuates the locking mechanism making unnecessary the use of the torque pressure of the vessel head applied to the upper core alignment plate for pre-loading the springs. Thus the special care and techniques used to tighten or remove the fuel alignment plate and reactor vessel head in designs requiring such pre-loading will never be needed with the present invention. Finally, no vertical compressive force is exerted on the guide tubes, and lateral vibration of the assembly is prevented. |
abstract | A fuel element handling system having a gripping device, the structure of which can be pushed in the horizontal position and uncoupled from the handling system if a failure occurs, uncoupling releasing the load formed by the fuel elements and their gripped support and putting them in a safe situation. |
|
048184756 | summary | The present invention relates to emergency core cooling system (ECCS) networks for nuclear power reactors. More particularly, the invention is preferably designed to compliment advanced boiling water reactor (BWR) designs known as simpIified boiling water reactors (SBWRs). Under the invention as appIied to SBWRs, reactor coolant inventory is replenished, in a backup mode to the safety-grade emergency core cooling system, at an early point following a loss-of-coolant accident. The spindown energy of the main turbine-generator is used to drive selected pumps such as the condensate pumps to achieve desired initial injections of emergency coolant into the reactor. BACKGROUND OF THE INVENTION The feedwater supply system for many conventional boiling water reactors and especially for simplified boiling water reactors (collectively referred to as BWRs) is a conventional yet simplified system, characterized by two stages of pumping. These pumping stages raise the feedwater from the below-atmospheric pressures at the source of the feedwater flow--namely, the condenser hotwell--to the pressures needed for injection into the BWR feedwater spargers positioned inside the reactor. The first or lowermost pumping stage, customarily termed the "condensate" stage, contains oondensate pumps having pump discharge pressures, at design flow, of approximately 500 psig, with shutoff heads of approximately 600 psig. The second or uppermost pumping stage, customarily termed the "feedwater" stage, contains feedwater pumps capable of increasing the feedwater supply pressure to approximately 1250 psig. The pumping burden, at both stages, is commonly shared by redundant pumps. For example, a configuration featuring three condensate pumps (and also three feedwater pumps) each having 50% rated flow capacity is one attractive configuration. One pump from each stage may be held in a standby mode, to be brought into service in case an operational pump requires shutdown for any reason. During normal operation, these pump units are motor-driven by power from the main station power supply, with the feedwater pumps having adjustable-speed drives to provide feedwater regulation to the reactor. However, during some loss-of-coolant accidents (LOCAs), the reactor must be supplied with additional coolant and cooled--that is, the reactor coolant level must be maintained high enough to cover all of the reactor nuclear fuel assemblies. Such additional coolant must be supplied by reliable emergency cooling systems which draw power from reliable alternative sources. Loss-of-coolant inventory conditions may occur because of a pipe break (i.e., a LOCA), loss of feedwater supply, or because a safety-relief valve has stuck-open and failed to reclose following a transient. Coolant must be maintained, or must be rapidly replenished following its loss during such accident conditions, to keep the reactor core supplied with coolant to counteract core decay heat generation. These systems which must function to prevent exceedence of core temperature limits comprise the "emergency core cooling network" (ECCS). Core decay heat generation results from the radioactive decay of fission products and continues even after the fission itself has stopped. In addition, coolant inventory is depleted within the reactor through processes of boiling and evaporation as the hot reactor coolant continues to receive decay heat from the core. As a result, an intermittent or even continuous replenishment of coolant is needed in the long term. The replenishment rates may be large immediately following an accident. Thereafter however, replenishment rates diminish as time goes on and the decay heat generation rate decreases. Exoept for very small LOCAs, replenishment of coolant must continue until the break can be isolated and normal coolant inventory level reestablished inside the reactor. For certain accidents, replenishment must continue until the region of the containment immediately outside the reactor pressure vessel can be flooded to an elevation above the top of t he core active fuel level or the break, whichever is higher. Several emergency core cooling systems with independent power supplies have evolved for responding to a LOCA for nuclear power reactors in general and for boiling water reactors in particular. Conventional BWR ECCS networks, for example the BWR/3 through the BWR/6 model BWR designs by GE Nuclear Energy, utilize a combination of pumping systems and power supplies to pump coolant into the reactor following any loss-of-coolant inventory condition. Water is typically used as the emergency coolant for BWRs. The source of water can be any available quantity of water within the power station or its premises. For example, the BWR/3 through BWR/6 reactors typically draw emergency coolant from a containment suppression pool. This suppression pool provides water which is assured, is available in large amounts, and is generally of a quality that is not particularly harmful to the reactor vessel or the nuclear steam supply system piping or equipment. Because the containment suppression pool is conventionally located low in the containment relative to the higher-elevation nuclear reactor, a break in certain pipes connecting to the reactor can allow injected coolant to be drained back out of the vessel. Such BWR designs result in extremely long pump-operating requirements for the pumping systems that provide the necessary emergency coolant inventory replenishment action. Thus, the conventional BWR designs have several drawbacks relating to emergency core cooling resulting from the extremely long pump duty cycles needed to meet coolant replenishment requirements. For example, both the pumps and the piping networks as well as the power supplies that power the ECCS have heretofore been costly dedicated systems having high reliability ratings. Such high reliability ECCS design often is achievable only by providing redundant components or even redundant pumping loops. Such redundancy in systems results in significant cost increases for the power station. It is possible to use the main system generator as a source of power for ECCS pumps during some LOCAs. However, in some important accident scenarios, electrical power from the main generator is hypothesized to be unavailable. For example, the main generator itself may be in a shorted condition (e.g., shorted windings), or the main generator may otherwise have been taken offline during the LOCA. For conventional BWRs, safety-grade diesel generators are installed that supply the necessary reliable ECCS network electrical power. These diesel generators are used where in-house electrical power has been interrupted from the generally two independent offsite grid power supplies into the station, as well as from the power station main turbine-generator. (The power station's in-house ("hotel") load can be furnished by the station main turbine-generator, but only if the reactor steam source has not become isolated.) A loss of power from these preferred sources would result in the automatic start-up of the diesel generators, and the subsequent progressive loading onto their emergency buses of the motor loads for ECCS pumps and other emergency equipment. For conventional systems, such diesel generators must be rated to operate continuously for as long as 90 days and typically must have an 8-day supply of fuel on hand. Advanced simplified types of BWRs--termed SBWRs--position the suppression pool previously discussed at a high elevation in the containment vessel relative to the core top-of-active-fuel (TAF) elevation. This elevation of the suppression pool overcomes the long-term need for continuous pumped coolant injection into the reactor. The suppression pool is connected via a plurality of pipes directly to the reactor, with valves--typically check valves--that prevent the discharge of high-pressure reactor coolant into the suppression pool during routine reactor power generation. This system of pipes and valves is termed a "gravity-driven cooling system (GDCS)", and along with associated venting systems, represents the entire ECCS network for certain SBWRs. If a loss-of-coolant inventory condition occurs, as detected by reactor water level measurements, the SBWR reactor is promptly depressurized to the suppression pool pressure level using a venting system. When the reactor pressure has fallen to a low pressure level (such as 30 psig), the hydrostatic head created by the elevated suppression pool initiates flow of suppression pool water into the reactor. The suppression pool includes sufficient water such that during a LOCA, both the reactor as well as the region of the containment external to the reactor (the "drywell") can be flooded to a level moderately higher than the TAF level. The maintenance of adequate reactor coolant inventory in these SBWRs thus no longer depends at any time on coolant inventory replenishment (pumping) by ECCS pumps. The flooding of the reactor and/or drywell by the GDCS using suppression pool water keeps the reactor core inundated. Any boiloff of evaporated coolant passes to the suppression pool through latched-open depressurization valves, and returns to the core by gravity refill via GDCS pipelines. A design goal for SBWR is not only to avoid exceeding core temperature limits during the course of any design basis accident, but also to provide ample margin against such occurrence. This assured margin is attained by specifying no core uncovery condition shall occur, even briefly, during such accidents. However, any added systems that provide this margin are not required to meet safety-grade design criteria, and these systems are taken as backups to, but not part of the ECCS network itself. The advantage to added or backup systems that are not required to be part of the ECCS network is that they can be designed to less-stringent criteria, which translates to less expense. At the same time these added or backup systems provide important enhanced investment protection to the power station because they further reduce the risk of core damage given an accident. To insure adequate coolant inventory (margin) in the short term while the reactor is undergoing depressurization--before the initiation of GDCS flow--the SBWR reactor vessel is designed to contain excess water, relative to conventional BWRs. This extra water is contained in a zone starting with the TAF and extending up to the water level at which reactor depressurization signals are initiated (termed "Level-1"). Thus, those SBWRs which use gravity driven cooling can undergo depressurization--which entails a reduction of steam/water inventory from inside the reactor--and still maintain a sufficient vessel residual coolant inventory. The coolant inventory maintains adequate coverage of the core as the reactor is depressurized to low pressure levels. The zone between TAF and Level-1 in such SBWR reactor designs contains an amount of water corresponding to approximately one minute of rated feedwater flow injection. This amount is substantially larger than in conventional BWR designs which rely on long term ECCS pumped water injection into the reactor during and following reactor depressurization. Unfortunately, this excess volume leads to a taller reactor vessel, which in turn leads to a larger drywell and larger suppression pool, and thus greater costs for both the reactor vessel and containment. SUMMARY OF THE lNVENTION According to the invention there is provided an improved, reliable, low-cost electrical power supply and coolant injection system useful in such applications as the ECCS network for SBWRs. The invention uses one or more dedicated auxiliary generators, of small size and generating capacity relative to the size/capacity of the power station's main generator, which are direct-coupled mechanically to the main turbine-generator. During normal plant operations, electrical power derived from these auxiliary generators is the preferred power supply to the station condensate pump-motors. During accident conditions, while the turbine-generator undergoes a spindown transient, these condensate pumps remain connected to their respective auxiliary generators to maintain pumping of condensate into the reactor. This continued pumping of condensate begins the emergency introduction of condensate from the moment when the depressurized condition inside the reactor exceeds the current shutoff head capacity corresponding to the condensate pump motor speed. As stated, the generators are used to supply the electrical power for the aforesaid loads--for example, condensate pumps--during normal station operation. During LOCAs, no start-up of alternate power sources is required to effect continuation of function, or switch-over of function to ECCS service. The invention thus eliminates one of the principal causes for unreliability for conventional ECCS networks--namely, the start-up of the diesel-generator. This configuration of condensate pumps powered by auxiliary generators (preferred source) also avoids the cost of providing separate dedicated emergency injection pumps and diesel generators. Due to provision of emergency coolant replenishment according to the invention, the cost of providing extra volume (described as TAF-to Level-1 volume) inside the reaotor vessel and the suppression pool is additionally minimized. In an alternative embodiment, the short term power supply permits slow-coastdown of upper-stage feedwater pumps during transients involving loss-of-offsite-power. In a preferred embodiment, the shaft-coupled auxiliary generator and the pump-motors to which they are connected are designed to non-safety-grade criteria. In an alternative embodiment, these components are all designed as safety-grade components. |
claims | 1. A method of repairing a connection between a first nozzle and a closed vessel, comprising:cutting through an entire thickness of the first nozzle at a location adjacent to a mid-wall of the vessel;completely removing from the mid-wall of the vessel a portion of the first nozzle;disposing a replacement nozzle in an opening that remains after removal of the portion of the first nozzle, wherein the replacement nozzle and the removed portion of the first nozzle are different pieces;welding at least a portion of the replacement nozzle to at least a portion of a surface of the mid-wall of the vessel; andpositioning the replacement nozzle vertically and horizontally in the opening, wherein positioning comprises: disposing a positioning tool in the opening prior to disposing the replacement nozzle in the opening; disposing the replacement nozzle on the positioning tool to axially and radially position the replacement nozzle; fixing the axial and radial positions of the replacement nozzle; and removing the positioning tool prior to welding at least the portion of the replacement nozzle, wherein fixing comprises fixing the axial and radial positions of the replacement nozzle with at least one clamping tool, wherein the clamping tool has an end portion clamped to the replacement nozzle at a location on the replacement nozzle that is disposed external to the mid wall and an opposite end portion clamped to another vessel nozzle at a location on the vessel nozzle that is also disposed external to the mid wall. 2. The method according to claim 1, wherein welding comprises forming a plurality of weld layers between the replacement nozzle and the surface of the mid-wall. 3. The method according to claim 2, wherein welding comprises forming at least three weld layers. 4. The method according to claim 3, wherein welding comprises forming the weld to have a total thickness about equal to the total thickness of the replacement nozzle. 5. The method according to claim 2, further comprising: evaluating the integrity of the weld. 6. The method according to claim 5, wherein evaluating comprises: liquid penetrant testing of the weld. 7. The method according to claim 5, wherein evaluating comprises: ultrasonically inspecting the weld; and comparing a result of the ultrasonic inspection to a plurality of results obtained from inspecting defective and defect-free welds. 8. The method according to claim 7, wherein ultrasonically inspecting comprises analyzing at least one of an echodynamic signature including response amplitude and time of flight of the ultrasonic signal. 9. The method according to claim 1, further comprising: reducing a total length of the first nozzle. 10. The method according to claim 1, further comprising: removing a component disposed within the first nozzle. 11. The method according to claim 10, wherein removing comprises removing a heater or other component disposed within the first nozzle. 12. The method according to claim 1, where fixing comprises fixing the positions of the replacement nozzle with three clamping tools. 13. The method according to claim 12, further comprising:preparing at least one of the portion of the first nozzle that remains with the vessel and the surface of the mid-wall of the vessel after removing the portion of the first nozzle and before disposing the replacement nozzle in the opening; anddye penetrant testing the surface of the mid-wall after preparing and before disposing the replacement nozzle in the opening. 14. A method of repairing a connection between a first nozzle and a closed vessel, comprising:cutting through an entire thickness of the first nozzle at a location adjacent to a mid-wall of the vessel;completely removing from the mid-wall of the vessel a portion of the first nozzle;disposing a replacement nozzle in an opening that remains after removal of the portion of the first nozzle, wherein the replacement nozzle and the removed portion of the first nozzle are different pieces;welding at least a portion of the replacement nozzle to at least a portion of a surface of the mid-wall of the vessel; andpositioning the replacement nozzle vertically and horizontally in the opening, wherein positioning comprises: disposing a positioning tool in the opening prior to disposing the replacement nozzle in the opening; disposing the replacement nozzle on the positioning tool to axially and radially position the replacement nozzle; fixing the axial and radial positions of the replacement nozzle; and removing the positioning tool prior to welding at least the portion of the replacement nozzle, wherein fixing comprises fixing the axial and radial positions of the replacement nozzle with at least one clamping tool, wherein the welding is performed without the use of a welding pad. |
|
abstract | A modular structure is disclosed that shields people from radiation typically encountered in radiotherapy rooms. The structure can be easily assembled without cracks. Moreover, the structure can be assembled and repaired quickly, without interrupting treatments for long periods of time. The structure is made of two general types of modules. A base module is made up of two cuboids fused along facing long edges and offset, one from the other, both vertically and horizontally. A complementary module is made of a single flatter shaped cuboid. It is designed to fill gaps that appear at the top and bottom of the assembled array of base modules. These modules, as configured, allows several structures to be assembled by quick and simple fitting and horizontal and vertical stacking of the modules. The modules can be manufactured from low cost materials such as metal casing filled with metal powder. |
|
abstract | A reflective optical element exhibits an increase in the maximum reflectivity at operating wavelengths in the extreme ultraviolet or soft x-ray wavelength range. A first additional intermediate layer (23a, 23b) and a second additional intermediate layer (24a, 24b) are provided between the absorber layer (22) and the spacer layer (21), wherein the first additional intermediate layer increases the reflectivity and the second additional intermediate layer (24a,b) prevents chemical interaction between the first additional intermediate layer (23a,b) and the adjoining spacer layer (21) and/or the absorber layer (22). |
|
description | The present invention relates to a scintillator panel and a radiation detector. As a conventional scintillator panel there is, for example, the one described in Patent Literature 1. In this conventional configuration, a 0.05-mm glass substrate is used as a support body for a scintillator layer. Furthermore, a buffer to relieve force from the outside of a housing and a high-stiffness member with stiffness higher than that of the scintillator layer are disposed between the housing and the scintillator layer. In the scintillator panel described in Patent Literature 2, a graphite substrate coated with a polyimide-based resin film or with a poly-para-xylylene film is used as a support body. Furthermore, in the scintillator panel described in Patent Literature 3, the entire surface of the substrate comprised of amorphous carbon or the like is covered by an intermediate film such as a poly-para-xylylene film. Patent Literature 1: Japanese Patent Application Laid-open Publication No. 2006-58124 Patent Literature 2: International Publication WO 2009/028275 Patent Literature 3: Japanese Patent Application Laid-open Publication No, 2007-279051 The scintillator panel applied, for example, to a solid-state detector such as a thin-film transistor (TFT) panel is required to have flexibility enough to satisfy shape-following capability to the solid-state detector. In addition, if there is a difference between the coefficient of thermal expansion of the TFT panel and the coefficient of thermal expansion of the substrate of the scintillator panel, fine flaws on the substrate of the scintillator panel or flaws made between the scintillator panel and the TFT panel by abnormally grown portions produced in formation of the scintillator layer by evaporation can transfer to the light receiving surface because of heat during operation, raising a problem that effort of calibration becomes troublesome. For solving the problem of flexibility and the problem of coefficient of thermal expansion as described above, it is conceivable to use extremely thin glass, e.g., in the thickness of not more than 150 μm as the substrate of the scintillator panel. However, when the extremely thin glass is used, there arises a problem that the end (edge part) of glass is brittle under an impact to chip or crack. The present invention has been accomplished in order to solve the above problems and it is an object of the present invention to provide a scintillator panel capable of ensuring satisfactory flexibility while preventing the glass substrate from chipping or cracking, and a radiation detector using it. In order to solve the above problems, a scintillator panel according to the present invention comprises: a glass substrate with a thickness of not more than 150 μm having radiotransparency; a first organic resin layer formed so as to cover the entire surface of the glass substrate; a scintillator layer formed on a one face side of the glass substrate on which the first organic resin layer is formed; and a moisture-resistant protection layer formed so as to cover the whole of the scintillator layer along with the glass substrate on which the first organic resin layer is formed. In this scintillator panel, the glass substrate with the thickness of not more than 150 μm serves as a support body, thereby to achieve excellent radiotransparency and flexibility and also relieve the problem of thermal expansion coefficient. In addition, in this scintillator panel the first organic resin layer is formed so as to cover the entire surface of the glass substrate. This reinforces the glass substrate, whereby the edge part thereof can be prevented from chipping or cracking. Furthermore, stray light can be prevented from entering the side face of the glass substrate, and warping of the glass substrate can be suppressed because the first organic resin layer is formed on the entire surface thereof. In the foregoing scintillator panel, the first organic resin layer may be selected from poly-para-xylylen and polyurea. Furthermore, preferably, a resin film layer is stuck between an other face side of the glass substrate on which the first organic resin layer is formed, and the protection layer. In this case, the glass substrate can be further reinforced by the resin film layer. The resin film layer is present on the other face side of the glass substrate, whereby internal stress of the scintillator layer can be cancelled, so as to more effectively suppress warping of the glass substrate. Furthermore, preferably, a resin film layer is stuck between the one face side of the glass substrate on which the first organic resin layer is formed, and the scintillator layer. In this case, the glass substrate can be further reinforced by the resin film layer. In addition, transparency can be ensured for light incident to the other face side of the glass substrate, so as to maintain resolution. In the foregoing scintillator panel, the resin film layer may be selected from PET, PEN, COP, and PI. Furthermore, preferably, a second organic resin layer is formed so as to cover an other face side and a side face side of the glass substrate on which the first organic resin layer is formed. This further reinforces the glass substrate, whereby the edge part thereof can be more effectively prevented from chipping or cracking. Furthermore, the second organic resin layer is formed on the other face side and on the side face side of the glass substrate, which can further enhance the effect of preventing stray light and the effect of suppressing warping of the glass substrate. Furthermore, preferably, a second organic resin layer is formed so as to cover the one face side and a side face side of the glass substrate on which the first organic resin layer is formed. This further reinforces the glass substrate, whereby the edge part thereof can be more effectively prevented from chipping or cracking. In addition, the second organic resin layer is formed on the one face side and on the side face side of the glass substrate, which can further enhance the effect of preventing stray light and which can ensure transparency for light incident to the other face side of the glass substrate, so as to maintain resolution. In the foregoing scintillator panel, the second organic resin layer may be selected from silicone resin, urethane resin, epoxy resin, and fluorine resin. Furthermore, a radiation detector according to the present invention comprises: the scintillator panel as described above; and a light receiving element arranged opposite to the scintillator layer on which the protection layer is formed. In this radiation detector, the glass substrate with the thickness of not more than 150 μm serves as a support body of the scintillator panel, thereby to achieve excellent radiotransparency and flexibility and also relieve the problem of thermal expansion coefficient. In addition, in this radiation detector the first organic resin layer is formed so as to cover the entire surface of the glass substrate. This reinforces the glass substrate, whereby the edge part thereof can be prevented from chipping or cracking. Furthermore, stray light can be prevented from entering the side face of the glass substrate, and warping of the glass substrate can be suppressed because the first organic resin layer is formed on the entire surface thereof. The present invention has made it feasible to ensure satisfactory flexibility while preventing the glass substrate from chipping or cracking. Preferred embodiments of the scintillator panel and the radiation detector according to the present invention will be described below in detail with reference to the drawings. FIG. 1 is a cross-sectional view showing a configuration of a radiation detector according to the first embodiment of the present invention. As shown in the same drawing, the radiation detector 1A is constructed by fixing a light receiving element 3 to a scintillator panel 2A. The light receiving element 3 is, for example, a TFT panel in which photodiodes (PD) and thin-film transistors (TFT) are arrayed on a glass substrate. The light receiving element 3 is stuck on a one face side of the scintillator panel 2A so that a light receiving surface 3a thereof is opposed to a below-described scintillator layer 13 in the scintillator panel 2A. The light receiving element 3 to be also used herein besides the TFT panel can be an element configured so that an image sensor such as CCD is connected through a fiber optic plate (FOP: an optical device composed of a bundle of several-micrometer optical fibers, e.g., J5734 available from Hamamatsu Photonics K.K.). The scintillator panel 2A is composed of a glass substrate 11 as a support body, an organic resin layer (first organic resin layer) 12 to protect the glass substrate 11, a scintillator layer 13 to convert incident radiation to visible light, and a moisture-resistant protection layer 14 to protect the scintillator layer 13 from moisture. The glass substrate 11 is, for example, an extremely thin substrate having the thickness of not more than 150 μm and preferably having the thickness of not more than 100 μm. Since the glass substrate 11 is extremely thin in thickness, it has sufficient radiotransparency and flexibility and ensures satisfactory shape-following capability of the scintillator panel 2A in sticking it on the light receiving surface 3a of the light receiving element 3. The organic resin layer 12 is formed, for example, of poly-para-xylylene or polyurea or the like by vapor-phase deposition (e.g., evaporation), so as to cover the entire surface of the glass substrate 11. The thickness of the organic resin layer 12 is, for example, approximately from ten to several ten μm. The scintillator layer 13 is formed on a one face 11a side of the glass substrate 11 on which the organic resin layer 12 is formed, for example, by growing and depositing columnar crystals of CsI doped with Tl by the evaporation method. The thickness of the scintillator layer 13 is, for example, 250 μm. The scintillator layer 13 is highly hygroscopic and could deliquesce with moisture in air if kept exposed to air. For this reason, the moisture-resistant protection layer 14 is needed for the scintillator layer 13. The protection layer 14 is formed, for example, by growing poly-para-xylylene or the like by the vapor phase deposition such as the CVD method, so as to cover the scintillator layer 13 along with the glass substrate 11 on which the organic resin layer 12 is formed. The thickness of the protection layer 14 is, for example, approximately 10 μm. In the radiation detector 1A having the configuration as described above, radiation incident from the glass substrate 11 side is converted to light in the scintillator layer 13 and the light is detected by the light receiving element 3. Since in the scintillator panel 2A the glass substrate 11 with the thickness of not more than 150 μm serves as a support body, it has excellent radiotransparency and flexibility. The glass substrate 11 has sufficient flexibility, thereby satisfying the shape-following capability in sticking the scintillator panel 2A to the light receiving surface 3a of the light receiving element 3. Furthermore, when the TFT panel is used as the light receiving element 3 and when the light receiving surface 3a is a glass panel, the coefficient of thermal expansion of the light receiving surface 3a can be made equal to that of the glass substrate 11 of the scintillator panel 2A. This can prevent fine flaws on the glass substrate 11 or flaws made between the scintillator panel and the TFT panel by abnormally grown portions produced during formation of the scintillator layer 13 by evaporation, from transferring to the light receiving surface 3a because of heat during operation, and can also avoid the need for troublesome effort of calibration. In addition, in this scintillator panel 2A the organic resin layer 12 is formed so as to cover the entire surface of the glass substrate 11. This reinforces the glass substrate 11, whereby the edge part thereof can be prevented from chipping or cracking. This also contributes to improvement in handling performance during manufacture and during use. Furthermore, stray light can be prevented from entering a side face 11c of the glass substrate 11 and, since the organic resin layer 12 is formed on the entire surface, it becomes possible to suppress warping of the glass substrate 11 due to internal stress after formation of the scintillator layer 13. The effect of suppressing warping of the glass substrate 11 becomes particularly prominent in a case where the glass substrate 11 is a small substrate of about 10 cm×10 cm. Moreover, since the organic resin layer 12 is formed so as to cover the entire surface of the glass substrate 11, it also becomes possible to adjust the surface condition of the glass substrate 11 so as to achieve appropriate surface energy and surface roughness in formation of the scintillator layer 13. FIG. 2 is cross-sectional views showing configurations of radiation detectors according to the second embodiment of the present invention. As shown in the same drawing, the radiation detectors 1B, 1C according to the second embodiment are different from the first embodiment in that in scintillator panels 2B, 2C, a resin film layer 16 is further arranged outside the glass substrate 11 on which the organic resin layer 12 is formed. More specifically, in the example shown in FIG. 2(a), the resin film layer 16 is stuck on the opposite face (other face 11b) side to the face where the scintillator layer 13 is formed, in the glass substrate 11 on which the organic resin layer 12 is formed, by means of a laminator or the like. Furthermore, in the example shown in FIG. 2(b), the resin film layer 16 is stuck on the face (one face 11a) side where the scintillator layer 13 is formed, in the glass substrate 11 on which the organic resin layer 12 is formed, by means of a laminator or the like. The resin film layer 16 is selected, for example, from PET (polyethylene terephthalate), PEN (polyethylene naphthalate), COP (cycloolefin polymer), and PI (polyimide). The thickness of the resin film layer 16 is, for example, approximately from ten to several ten μm as the organic resin layer 12 is. Furthermore, the edge of the resin film layer 16 is preferably coincident with the edge of the glass substrate 11 or slightly projects out therefrom. In this configuration, just as in the above embodiment, the glass substrate 11 is also reinforced by the organic resin layer 12, whereby the edge part thereof can be prevented from chipping or cracking. In addition, stray light can be prevented from entering the side face 11c of the glass substrate 11 and, since the organic resin layer 12 is formed on the entire surface, warping of the glass substrate 11 can be suppressed. Furthermore, in these radiation detectors 1B, 1C, the glass substrate 11 is further reinforced by addition of the resin film layer 16, whereby the edge part thereof can be more securely prevented from chipping or cracking. When the resin film layer 16 is arranged on the other face 11b side of the glass substrate 11 as shown in FIG. 2(a), internal stress of the scintillator layer 13 can be cancelled, so as to more effectively suppress warping of the glass substrate 11. When the resin film layer 16 is arranged on the one face 11a side of the glass substrate 11 as shown in FIG. 2(b), transparency can be ensured for light incident to the other face 11b side of the glass substrate 11, so as to decrease reflection toward the light receiving element 3, with the result that resolution can be maintained. FIG. 3 is cross-sectional views showing configurations of radiation detectors according to the third embodiment of the present invention. As shown in the same drawing, the radiation detectors 1D, 1E according to the third embodiment are different from the first embodiment in that in scintillator panels 2D, 2E, an organic resin layer (second organic resin layer) 17 is further arranged outside the glass substrate 11 on which the organic resin layer 12 is formed. More specifically, in the example shown in FIG. 3(a), the organic resin layer 17 is formed so as to cover the opposite face (other face 11b) to the face where the scintillator layer 13 is formed, and the side face 11c, in the glass substrate 11 on which the organic resin layer 12 is formed. Furthermore, in the example shown in FIG. 3(b), the organic resin layer 17 is formed so as to cover the face (one face 11a) where the scintillator layer 13 is formed, and the side face 11c, in the glass substrate 11 on which the organic resin layer 12 is formed. The organic resin layer 17 to be used herein can be, for example, silicone resin, urethane resin, epoxy resin, fluorine resin, or the like. A method for forming the organic resin layer 17 is, for example, coating by the spin coating method or the like. The thickness of the organic resin layer 17 is, for example, approximately from ten to several ten μm as the organic resin layer 12 is. In these configurations, just as in the above embodiments, the glass substrate 11 is also reinforced by the organic resin layer 12, whereby the edge part thereof can be prevented from chipping or cracking. In addition, stray light can be prevented from entering the side face 11c of the glass substrate 11 and, since the organic resin layer 12 is formed on the entire surface, warping of the glass substrate 11 can be suppressed. Furthermore, in these radiation detectors 1D, 1E, the glass substrate 11 is further reinforced by addition of the organic resin layer 17, whereby the edge part thereof can be more securely prevented from chipping or cracking. When the organic resin layer 17 is formed so as to cover the other face 11b and the side face 11c of the glass substrate 11 as shown in FIG. 3(a), it is feasible to further enhance the effect of preventing stray light from entering the side face 11c and the effect of suppressing warping of the glass substrate 11. When the organic resin layer 17 is formed so as to cover the one face 11a and the side face 11c of the glass substrate 11 as shown in FIG. 3(b), the effect of preventing stray light from entering the side face 11c can be enhanced, and transparency can be ensured for light incident to the other face 11b side of the glass substrate 11, so as to decrease reflection toward the light receiving element 3, with the result that resolution can be maintained. 1A-1E radiation detectors; 2A-2E scintillator panels; 3 light receiving element; 11 glass substrate; 11a one face; 11b other face; 11c side face; 12 organic resin layer (first organic resin layer); 13 scintillator layer; 14 protection layer; 16 resin film layer; 17 organic resin layer (second organic resin layer). |
|
052727335 | abstract | A control rod driving hydraulic system of a control rod driving system of a nuclear power plant, in utilization of condensate as a control rod driving water comprises a water pressure control unit arranged for each control rod and a water pressure supply unit for commonly supplying water pressure to all control rods. The water pressure supply unit comprising a pump mechanism operatively connected to a condensate supply source for driving the control rod driving water, a filter mechanism operatively connected to the condensate supply source and the pump mechanism for filtering the condensate as the control rod driving water, a valve unit for regulating quantity of flow of the control rod driving water, and a valve unit for regulating pressure thereof. The filter mechanism including at least one hollow fiber filter unit for purifying the control rod driving water. The hollow fiber filter unit includes a backwash regeneration equipment. The filter mechanism includes a suction filter unit and a driving water filter unit, the suction filter unit being connected to the condensate supply unit, the driving water filter unit being connected to the suction filter unit through the pump mechanism and connected to the flow quantity regulating mechanism and the suction filter unit or the driving water filter unit is substituted with the hollow fiber filter unit. |
description | This application is a filing under 35 U.S.C. 371 of international application number PCT/GB02/05624, filed Dec. 11, 2002, which claims priority to application number 0206550.6 filed Mar. 20, 2002, in Great Britain the entire disclosure of which is hereby incorporated by reference. The present invention relates to an interengaging component support which is particularly, but not exclusively, suited to implementation in a radioisotope generator of the type commonly used to generate radioisotopes such as technetium-99m (99mTc). The diagnosis and/or treatment of disease in nuclear medicine constitute one of the major applications of short-lived radioisotopes. It is estimated that in nuclear medicine over 90% of diagnostic procedures performed worldwide annually use 99mTc labelled radio-pharmaceuticals. Given the short half-life of diagnostic radio-pharmaceuticals, it is helpful to have the facility to generate suitable radioisotopes on site. Accordingly, the adoption of portable hospital/clinic size 99mTc generators has greatly increased over the years. Portable radioisotope generators are used to obtain a shorter-lived daughter radioisotope which is the product of radioactive decay of a longer-lived parent radioisotope, usually adsorbed on a bed in an ion exchange column. Conventionally, the radioisotope generator includes shielding around the ion exchange column containing the parent radioisotope along with means for eluting the daughter radioisotope from the column with an eluate, such as saline solution. In use, the eluate is passed through the ion exchange column and the daughter radioisotope is collected in solution with the eluate, to be used as required. In the case of 99mTc, this radioisotope is the principle product of the radioactive decay of 99Mo. Within the generator, conventionally the 99Mo is adsorbed on a bed of aluminium oxide and decays to generate 99mTc. As the 99mTc has a relatively short half-life it establishes a transient equilibrium within the ion exchange column after approximately twenty-four hours. Accordingly, the 99mTc can be eluted daily from the ion exchange column by flushing a solution of chloride ions, i.e. sterile saline solution through the ion exchange column. This prompts an ion exchange reaction, in which the chloride ions displace 99mTc but not 99Mo. In the case of radio-pharmaceuticals, it is highly desirable for the radioisotope generation process to be performed under aseptic conditions i.e. there should be no ingress of bacteria into the generator. Moreover, due to the fact that the isotopes used and generated with the generator are radioactive, and are thereby extremely hazardous if not handled in the correct manner, the radioisotope generation process also should be conducted under radiologically safe conditions. Naturally, it is desirable to ensure that when the elution process is performed, the radiological safety of the generator is not compromised. In particular, when the eluate is introduced into the generator, it is important for the radiological safety of the generator to be maintained. In trying to ensure adequate radiological protection, some known radioisotope generators have tended to be of a complicated construction incorporating a large number of components. However, the radiological protection afforded by such structures can be compromised where the interconnection of the various components is unreliable. Such complex structures also add to the cost of the generator. It is thus important that the actual construction of the generator is reliable and all component interconnections are secured to a high degree of certainty. U.S. Pat. No. 3,946,238 describes a shielded radioisotope generator comprising a cylindrical shielded housing for a central repository. The repository is bound by a removable top cover and side walls and a base which are made from lead and which act as the shielding. Within the repository a bottle is located which contains an ion exchange column in which 99Mo is absorbed. When it is desired to add saline solution to the system to prompt the elution of 99mTc, the top cover is removed, and the saline is introduced by way of a transfer pipette. The saline solution is introduced by means of the pipette to an annular region between the bottle and the inner surfaces of the shielding. From this annular region the saline solution flows in a controlled manner into the bottle containing the ion exchange bed via a series of radial openings in the wall of the bottle. The transfer pipette has a long handle designed such that a user's hands always remain outside the generator when saline is introduced into the annular region about the bottle. It is apparent, however, that the removal of the top cover for the purposes of introducing the saline solution constitutes an unacceptable radiological risk as the interior of the repository is radioactive. U.S. Pat. No. 3,564,256 describes a radioisotope generator having quick-coupling members for the elution process. The generator includes a cylindrical holder containing a radioactive substance bound to an ion exchange bed. The holder is closed by rubber plugs at both ends, and is surrounded by shielding having passages opposite each of the rubber plugs in which respective needles are located. At the outermost ends of the needles quick-coupling members are provided to enable a syringe vessel containing a saline solution to be quickly and easily connected to one of the needles and to enable a collection vessel to be connected to the other of the two needles. In use, each one of the rubber plugs of the cylindrical holder is pierced by one of the needles to prompt the elution of 99mTc from the ion exchange column. Suitable quick-coupling members proposed in the document are conventional detachable injection needle to injection syringe connections. U.S. Pat. No. 4,387,303 describes a radioisotope generator comprising a column having an elute inlet aperture and an elute outlet aperture and containing an ion exchange bed with the parent radioisotope. Both the elute inlet and outlet are in communication with channels in the surrounding shielding. One of the channels, that is in communication with the elute outlet, is connected to a tapping point on the generator via an eluate conduit. The tapping point is adapted to receive an evacuated elution vial for collection of the daughter radioisotope in solution and consists of a hollow needle that pierces the seal to the evacuated elution vial. The eluate conduit is also in communication with a source of sterile air and the generator includes a device for interrupting the elution process before the elution vial is filled by interrupting the flow of sterile air. No information is provided with regard to the construction of the generator and in particular no information is provided as to how the hollow needle at the tapping point is held in position. In view of the needs of the prior art, the present invention provides a component support for use in a radioisotope generator. The component support includes a latching member movable between an engaging position and an open position and further includes a bracing member mechanically associated with the latching member and adapted to prevent movement of the latching member to the open position. The present invention seeks to provide a component support that is simple in construction but provides greater reliability than existing simple component supports and so is particularly suited for use in radioisotope generators where there exists a need for a radioisotope generator that is simple in construction but which ensures the necessary degree of sterility and radiological protection. According to a first aspect of the present invention, there is provided a component support for use in a radioisotope generator, the component support comprising a latching member movable between an engaging position and an open position characterised by further including a bracing member mechanically associated with the latching member and adapted to prevent movement of the latching member to the open position. In a preferred embodiment of the present invention, the component support may include a first plate on which the latching member is mounted, with the first plate including an opening at or adjacent the latching member for receiving the bracing member. The opening in the first plate is preferably an aperture in the first plate adjacent the latching member on the side of the latching member facing the direction of movement of the latching member from the engaging position to the open position. Preferably, the opening is of non-circular cross-section and the bracing member has a corresponding non-circular cross-section. Also, the latching member may additionally include a camming surface engageable by the bracing member for urging the latching member away from the open position. More preferably the component support may also include a second plate on which the bracing member is mounted, the second plate being arranged to lie substantially parallel to the first plate when the bracing member is inserted through the opening in the first plate. The latching member is preferably a generally L-shaped structure consisting of a wall and a flange projecting therefrom, and in a preferred embodiment the latching member also includes a second flange arranged substantially parallel to the first flange for defining a slot therebetween. It is envisaged but by no means essential that the component support comprises at least two opposing latching members and respective bracing members. According to a second aspect of the present invention there is provided a radioisotope generator having one or more component supports as previously described. The latching member of the generator may be mounted on a closure plate of the generator which include an opening for receiving the bracing member, and wherein the bracing member is mounted on a cover plate of the generator such that insertion of the bracing member into the opening mounts the cover plate over the closure plate. Preferably, the radioisotope generator has two latching members mounted on the closure plate either side of a central component aperture and wherein the cover plate also includes a component aperture for alignment with the component aperture in the closure plate. The radioisotope generator may also include a fluid port comprising a hollow generally cylindrical body and a retaining plate, the hollow body being received in the component apertures in the closure plate and the cover plate and the retaining plate being engaged by the opposed latching members for securely holding the fluid port in position. In the preferred embodiment, the radioisotope generator includes a container consisting of a wall and a floor, with the opening to the container being closed by a closure plate. With this arrangement the latching member is located on the container wall and the closure plate includes a bracket for engagement with the latching member and an opening at or adjacent the bracket and the bracing member is provided on a cover plate such that insertion of the bracing member into the opening in the closure plate aligns the bracing member with the latching member thereby to prevent movement of the latching member to the open position. An embodiment of the present invention will now be described, by way of example only, with reference to the accompanying The component support is Illustrated generally by reference numeral 29, and the component illustrated in FIG. 1 is a spike 1 which projects through an aperture 2 in a plate 3 and has a planar mounting member 4 that is held in position by a pair of latching members 5. The latching members are movable between an engaging position in which they engage the planar mounting member and an open position in which the planar mounting member is not restrained by the latching members. Each of the pair of latching members 5 includes a wall 6 projecting outwardly from the surface of the plate 3 (downwardly as illustrated in FIGS. 1 and 2). The walls 6 are each spaced from the aperture 2 diametrically opposite one another across the aperture 2. A flange 7 is provided at the free end of each wall 6. The flanges 7 on each of the walls project away from the walls towards one another and extend substantially parallel to the plate 3. A second flange 8, substantially parallel to the first flange 7, is provided between the first flange 7 and the plate 3. The first 7 and second 8 flanges thus form a slot 9 suitable for receiving a planar member 4. The plate 3 is preferably made from a hard plastics material and the walls 6 and flanges 7, 8 are preferably moulded as a single unit with the plate 3. This results in the walls 6 and flanges 7, 8 having a small degree of resiliency sufficient to be suitable for “snap-fit” engagement of a planar member within the slot 9 defined by the first 7 and second 8 flanges. For this reason, as illustrated in FIG. 1, the first flange 7 has a camming surface 10 facing away from the plate 3 for guiding and centering a planar member 4 towards the slot 9 and for urging the small amount of flexure of the opposed walls 6 necessary to permit the planar member 4 to pass the periphery of the first flange 7 whereupon the walls 6 ‘snap’ back into position with the planar member 4 located and held in the slot 9 between the first and second flanges 7, 8. Such a snap-fit connection is generally well-known and provides a particularly quick method for securing two elements (in this case the planar member 4 and the plate 3) together. However, the fact that this manner of securement demands a small degree of flexure of the walls 6, generally renders such a means of securement undesirable in circumstances where the securement must be highly reliable. An external force applied to the plate 3 is capable of causing flexure of the walls 6 to the extent that the planar member 4 is accidentally freed from the slot 9. For this reason, snap-fit connections have not been considered suitable in the construction of radioisotope generators. The component supports 29 illustrated in FIGS. 1 and 2 however provide a greatly improved reliability of securement over convention snap-fit connectors, which renders the component supports 29 particularly suited for use in radioisotope generators. The component supports 29 include a cover 11 that is arranged to overlie the plate 3. The cover 11 has a component aperture 12 for alignment with the aperture 2 in the plate 3. The cover 11 also has a pair of bracing members 13 that project (downwardly in FIGS. 1 and 2) away from the cover 11. Also, adjacent each of the walls 6, on the opposite side of each of the walls 6 to the flanges 7, 8, respective brace apertures 14 are provided in the plate 3. The bracing members 13 on the cover 11 are positioned either side of the component aperture 12 so as to be aligned with the brace apertures 14 in the plate 3. The brace apertures 14 are sized to permit the passage of the bracing members 13 and preferably are non-circular in cross-section so that the bracing member 13 is keyed into the brace aperture 14. With the cover 11 positioned over the plate 3 and the bracing members 13 inserted into the brace apertures 14, the bracing members 13 are mechanically associated with the walls 6, and act as braces to the walls 6. This substantially prevents outward flexure of the walls 6. In this way, the reliability of the component support 29 is greatly enhanced. In a particularly preferred embodiment, each associated wall 6 and bracing member 13 have co-operable camming surfaces and followers. In FIG. 1 the camming surface 15 is on the wall 6 facing towards the bracing member 13. This enables the bracing member 13 to actively engage with and urge the wall 6 inwardly towards the planar member 4 when inserted in the slot 9 defined by the first and second flanges 7, 8. This further improves the reliability of the securement of the component provided by the component supports 29. FIG. 2 illustrates an implementation of the component supports in a radioisotope generator 16. The radioisotope generator 16 has an outer container 17, a closure plate, referred to herein as a top plate 3 which is sealingly secured to the outer container 17, and a separate top cover 11 which is secured to the outer container 17 over the top plate 3. Inside the outer container 17 a radioactive shield 18 is located which is preferably, but not exclusively, made from either lead or a depleted uranium core within a stainless steel shell. The radioactive shield 18 surrounds a tube 19 containing an ion exchange column 20. The ion exchange column 20 preferably consists of a mixture of aluminium and silica, onto which molybdenum in the form of its radioactive isotope, 99Mo is adsorbed. The tube 19 containing the ion exchange column 20 has frangible rubber seals 21 and 22 at opposing ends 23 and 24 which, as illustrated, when in use are pierced by respective hollow needles 25 and 26. Each of the hollow needles 25 and 26 are in fluid communication with respective fluid conduits 27, 28 which in turn are in respective fluid communication with an eluent inlet and an eluate outlet. The fluid conduits 27, 28 are preferably flexible plastics tubing and in the case of the tubing 27 that communicates with the hollow needle 25 at the top 23 of the ion exchange column 20, the length of the tubing 27 is much greater than the minimum required to connect the hollow needle 25 with the eluent inlet. The top plate 3 of the radioisotope generator 16 has a pair of apertures 2 through which the respective eluent inlet and eluate outlet components project. The eluent inlet and eluate outlet components are each hollow spikes 1 though in the case of the inlet component the hollow spike additionally includes a filtered air inlet 30. The hollow spike 1 consists of an elongate generally cylindrical spike body 31 and an annular retaining plate 32 which is attached to or is moulded as a single part with one end of the spike body 31. The opposing end of the spike body 31 is shaped to a point and has an aperture 33 communicating with the interior of the spike body 31 adjacent the point. This pointed end of the spike body 31 is shaped so that it is capable of piercing a sealing membrane of the type commonly found with sample vials. The annular retaining plate 32 forms a skirt projecting outwardly from the spike body 31 and may be continuous around the spike body 31 or discontinuous in the form of a plurality of discrete projections. The top cover 11 of the radioisotope generator 16 also includes a pair of apertures 12 arranged so as to align with the apertures 2 in the top plate 3 and shaped to allow through passage of the spike body 31. Thus, each of the hollow spikes 1 is arranged to be held and supported by its annular retaining plate 32 by latching members 5 located on the inside of the top plate 3 whilst the hollow spike body 31 projects through the apertures in both the top plate 3 and the top cover 11 to the exterior of the outer container 17. Each one of the apertures 12 in the top cover 11 is located at the bottom of a well 34 that is shaped to receive and support either an isotope collection vial 35 or a saline supply vial 36. Thus, both vials 35, 36 are housed outside of the outer container 17 and are not exposed to radiation from the ion exchange column 20. The hollow spikes 1 are held in place by the component supports 29 as described earlier with reference to FIG. 1. Thus, the spike body 31 projects through the aligned apertures in the top plate 3 and the top cover 11 and is securely held in position by engagement of the annular retaining plate 32 in the slot 9 defined by the first and second flanges 7, 8 of the latching members 5. Retention of the plate 32 in the slot 9 is maintained by the supporting action of the bracing members 13 outside of the walls 6 of the latching members 5 which substantially prevent outward flexure of the walls 6. When the radioisotope generator 16 is constructed, the spike body 31 is inserted through the aperture 2 in the top plate 3 and the annular retaining plate 32 contacts the camming surfaces 10 on an opposing pair of first flanges 7. Further pressure applied to the retaining plate 32 forces outward flexure of the walls 6 supporting the first flanges 7 until the retaining plate 32 is able to pass the free end of the first flanges 7. Once the retaining plate 32 has passed the first flanges 7 the external pressure on the walls 6 is eased and the walls 6 ‘snap’ back to their normal position locating the retaining plate 32 in the slots 9 defined by the first and second flanges 7 and 8. The top cover 11 is then positioned over the top plate 3 with the apertures 12 in the top cover 11 aligned with the spike body 31 and the bracing members 13 aligned with apertures 2 in the top plate 3 adjacent each of the walls 6. As the top cover 11 is brought into contact with the top plate 3 the bracing members 13 pass through the apertures 2 in the top plate 3 so as to be positioned next to, and preferably in contact with, the outer surfaces of the walls 6. The interaction of the bracing members 13 on the top cover 11 and the walls 6 of the top plate 3 thus provide reliable securement of the retaining plate 32 of the hollow spike 1 in the slot 9 defined by the first and second flanges 7, 8. The tubing 27 and 28 is then fluidly attached to the hollow spikes 1 and the outer container 17 is closed when the top plate 3 and the top cover 11 are secured to the container. When it is desired to attach a vial 35 or 36 to the hollow spike 1, a user positions the frangible seal of the vial over the pointed end of the spike and pushes the vial down onto the spike 1. This causes the seal on the vial 35 or 36 to be pierced establishing fluid communication between the spike 1 and the vial. Once the seal has been pieced by the spike 1, the vial is pushed down over the spike 1 until it rests and is supported by the well 34 in the top cover 11. In order to supply the ion exchange column 20 with the chloride ions required for elution of the radioisotope, saline solution 37 is drawn through the ion exchange column 20, by establishing a pressure differential across the ion exchange column 20. This is accomplished by connecting the saline supply vial 36 to the eluent inlet which is in fluid communication with the top end 23 of the ion exchange column 20 via the tubing 27 and hollow needle 25 and connecting an evacuated collection vial 35 to the eluate outlet which is in fluid communication with the bottom end 24 of the ion exchange column 20 via the tubing 28 and hollow needle 26. The pressure differential is established by virtue of the fluid pressure of the saline in the supply vial 36 and the extremely low pressure in the evacuated collection vial 35. This urges passage of the saline solution 37 through the ion exchange column 20 to the collection vial 35 carrying with it the daughter radioisotope. The component support is simple in design but by the interaction of the bracing member on one plate with the wall of the snap-fit component on the other plate and highly reliable component support is provided. Although reference has been made in the description to a component support suitable for a hollow spike, it will be apparent that the component support of the present invention may be employed with alternative components that are intended to be secured in a snap-fit holder. For example, the component support may be used as a means for attaching the top plate to the outer container of the radioisotope generator. With this arrangement, latching members are attached to the inner side walls of the outer container. Each latching member is spaced from the wall of the outer container by means of a bridge element so as to define a bracket receiving region between the latching member and the wall of the container. Thus, the wall of the latching member is arranged substantially parallel to the container wall and the slot defined by the paired flanges mounted on the wall of the latching member lies substantially perpendicular to the container wall. This arrangement also requires the top plate to have an equivalent number of brackets for location and engagement with respective latching members. Thus, as the top plate is lowered into position, the bracket attached to the periphery of the top plate and projecting downwardly therefrom, engages the first of the flanges on the latching member. The bracket urges the latching member to flex away from the container wall thereby enlarging the bracket receiving region until the bracket is capable of passing the periphery of the flange whereupon the latching member snaps back into position trapping part of the bracket in the slot defined by the two flanges. As described previously, the bracing member projects from the top cover and is locatable in an aperture in the top plate, such that, as before, it is mechanically associated with the latching member and acts to brace the latching member against flexure. It is not a requirement of the present invention that the bracing means is locatable through an aperture in the top plate such that it acts as an exterior abutment to the component support wall. For example, it is alternatively envisaged that the component support wall may include a blind bore, into which the bracing means is inserted, to provide the desired improved support for the latching member. Moreover, it is not a requirement of the present invention that the plates of the component support contain apertures through which the component passes. Instead, the component may extend away from the surface of the first plate bearing the walls of the component support (in the illustrated embodiment the top plate 3) in which case the second plate (in the illustrated embodiment the top cover 11) need only align the bracing members with the brace apertures in the first plate. Furthermore, although paired flanges defining a slot are illustrated above, it will be appreciated that the slot may be defined between a single flange and the surface of the first plate. Further and alternative features of the component support are envisaged without departing from the scope of the present invention as claimed. |
|
description | This application claims priority to U.S. Provisional Patent Application No. 62/270,741, filed on Dec. 22, 2015, and entitled “Integrated Mounting Feature for X-Ray Shielding Curtains,” and to U.S. Provisional Patent Application No. 62/312,066, filed on Mar. 23, 2016, and entitled “Molded in Features for X-Ray Shielding Curtains,” the disclosures of which are incorporated by reference. This disclosure relates in general to systems for efficiently and safely scanning luggage, packages, parcels, personal items, and the like, and, and, in particular, but not by way of limitation, to an electromagnetic radiation scanning system that includes shielding curtains with features to simplify manufacturing, assembly, and replacement of the shielding curtains. Electromagnetic radiation, for example X-ray radiation, is used to examine the contents of luggage and parcels prior to allowing such items to be taken on or loaded on transport vehicles or before allowing entry into buildings or other facilities. X-ray scanning machines continuously convey luggage, parcels, cargo, and personal items that are exposed to X-ray radiation that can penetrate the container and can be used to create an image of the contents of the container. Packages and luggage of all shapes and sizes are accommodated by the same scanning system. Radiation is contained within the scanning system by shielding curtains disposed at the entrance and exit of the scanning system. Conventional shielding curtains are fabricated in a laminated construction. Rolls of material scrim, lead vinyl, lead rubber, and Teflon/nylon are fed from rolls and each becomes a layer of a thin sheet of material. The lead vinyl is sandwiched between Teflon/nylon layers. The continuous strip is wound on a spool and then cut into individual strips. The individual strips are then secured by one or two metal bars or attachment devices and arranged adjacent to each other such that a series of parallel individual strips hang in front of an entrance or exit of the scanning machine and collectively contain or deflect the X-rays within the machine, such that workers are not exposed to potentially harmful X-rays. The lead content of the strips is selected to block the radiation generated in a particular application. The layered construction of the curtain strips forms uniform thickness strips that are free of surface texture. Sandwiching the individual strips of layered construction between two generally flat bars forms the X-ray shielding curtain. Each of the bars includes a plurality of through holes. A fastener is received through the front bar, and extends through a hole formed through the layered strip, and through the rear bar or attachment bar located on the X-ray scanning system. The holes in the layered construction strips are generally formed after the strips are constructed but before the strips are sandwiched between the clamping bars. Misplacement or misalignment of an individual curtain strip with respect to an adjacent curtain strip may lead to unwanted radiation leakage through a curtain bank. An example scanning system is disclosed in U.S. Pat. No. 4,020,346, issued on Apr. 26, 1977, entitled “X-Ray Inspection Device and Method,” which is hereby incorporated by reference. The '346 patent discloses a scanning system with two banks of shielding arranged parallel to each other to block the entrance to the scanning system, and two banks of shielding curtains arranged parallel to each other to block the exit to the scanning system. However, scanning systems for different applications, such as pre-shipping parcel or cargo inspection may have greater strength radiation, and therefore may have additional banks of radiation shielding curtains positioned at the entrance and exit. Parcels, luggage, or personal items that are conveyed through the scanning system displace the strips of curtains. In certain applications, a light parcel may be required to simultaneously displace two or more banks of curtains. If the parcel is too light to displace multiple curtain banks, a back-up may occur on the system that must be addressed by a worker. As should be obvious, curtains with a greater stiffness are not as easily displaced as curtains that are more flexible. Also, friction between the curtains and the parcel must be overcome so the parcel can move through the scanning system. Finally, the layered construction strip curtains wear, which can lead to unwanted material, including lead, being rubbed off onto the luggage or parcels. Of course, worn shielding curtains need to be replaced. Embodiments herein disclose a shielding curtain that is configured to block electromagnetic radiation from passing through it. The shielding curtain may be a flap portion of a larger shielding curtain or a single, unitary body that includes a single integrated mounting bead and a plurality of flaps. The shielding curtain is formed of a polymer material that has a uniformly dispersed particulate material. According to certain embodiments, the shielding curtain is molded from a composite polymer material that includes a thermosetting polymer material and the uniformly dispersed particulate material. Electromagnetic radiation emitted by an inspection system is blocked by the uniformly dispersed particulate material. A shielding curtain assembly includes a curtain suspending member with a slot that extends along a length of the curtain suspending member. A shielding curtain that blocks electromagnetic radiation is suspended by the curtain suspending member. The shielding curtain is formed of a polymer material, such as a thermosetting polymer, and a particulate filler material, such as Tungsten powder and/or Barium sulfate. According to certain embodiments, the shielding curtain includes a mounting bead that is received in the slot and a plurality of flaps that extend from the mounting bead. The mounting bead and the plurality of flaps may be a single, unitary body. Shielding curtains according to the present disclosure may be disposed at an entrance end or an exit end of an exposure station of an X-ray inspection system that emits electromagnetic radiation, for example X-ray radiation, to inspect the contents of luggage or shipping parcels. Each end of the inspection system may include multiple shielding curtains. Technical advantages of shielding curtains for electromagnetic radiation scanning systems according to the teachings of the present disclosure include mounting features that are directly molded into a unitary curtain with a plurality of flaps. The molded in mounting features facilitate easy installation, removal, and replacement of shielding curtains in existing inspection systems. In addition, the molded shielding curtains allow a surface texture of the flaps to be molded into the shielding curtain, which may reduce the coefficient of friction and/or the surface area of the shielding curtain that comes into contact with the package, parcel, or personal item to allow the item to more easily pass through the shielding curtain. Other technical advantages include the elimination of lead and replacement of lead containing curtains with a composite polymer material with a lead equivalency. The composite polymer material may be more flexible than conventional leaded layered construction curtains and may have a lower coefficient of friction. Lower frictional force and increased curtain flexibility results in increased throughput of packages, parcels, personal items, cargo, or luggage and also results in fewer jams or other stoppage of the inspection equipment. Other technical advantages will be readily apparent to one of ordinary skill in the art from the following figures, descriptions, and claims. Moreover, while specific advantages have been described above, various embodiments may include all, some, or none of the enumerated advantages. FIG. 1 is a perspective view of an electromagnetic radiation scanning system 10 according to the teachings of the present disclosure. The scanning system 10 may also be referred to as an inspection system. The electromagnetic radiation scanning system 10 includes a conveyor belt 12 that is supported by a support structure 14. The conveyor belt 12 conveys items 16 into an exposure station 18 where the item 16 is exposed to electromagnetic radiation that penetrates the item and provides an image of its contents. A worker views the image created by the penetrating electromagnetic radiation on a monitor 20 and can determine whether the item 16 should be further inspected. The item 16 may be luggage, a personal item, a package, or a parcel for shipping, or other container where an initial examination determines that the item is safe to transport or enter a facility and does not contain contraband. The item 16 may also be inspected to determine whether it contains items controlled by airport security regulations or other security protocol. For example the United States Transportation Security Administration may use an electromagnetic radiation scanning system 10 to inspect for explosive devices or other controlled items. The electromagnetic radiation may be in any suitable form for creating an image of the contents of a container. For example, the electromagnetic radiation may be x-rays, gamma rays, and the like. X-ray electromagnetic radiation is often used in scanning systems to inspect baggage and parcels. To protect individuals near the electromagnetic radiation scanning system 10, such as transportation, shipping, or security workers, the electromagnetic radiation should be contained within the exposure station 18. Therefore, the exposure station 18 is includes a material that is impenetrable by the particular emitted electromagnetic radiation. It is known to use lead to contain electromagnetic radiation, such as X-rays. The exposure station 18 includes an open entrance end 22 and exit end 24 that allow the conveyor belt 12 to continuously move the items 16 into and out of the exposure station 18. One or more shielding curtains 30 are disposed at the entrance end 22 and the exit end 24 of the exposure station 18 to block electromagnetic radiation from escaping into the ambient environment. In addition to blocking electromagnetic radiation, the shielding curtains are also configured to be displaced by the items 16 on the conveyor belt 12. Each shielding curtain 30 includes a plurality of flaps 32 that are displaced by the items 16. The shielding curtains 30 block the electromagnetic radiation from breaching the entrance end 22 and the exit end 22, but the flaps 32 of the shielding curtain 30 are flexible enough to be displaced by the items 16 moved by the conveyor belt 12. By displacing the flaps 32 of the shielding curtains 30 at the entrance end 22, the item 16 enters the exposure station 18 where it is safely exposed to electromagnetic radiation. After the exposure, the conveyor belt 12 moves the item 16 such that it displaces the flaps 32 of the shielding curtains 30 at the exit end 24 where the items 16 can be safely further handled. The shielding curtains 30 are coupled to the exposure station 18 such that they hang or are otherwise positioned to extend across and block the open entrance end 22 and the open exit end 24 of the exposure station 18. The shielding curtains 30 may be passive in that they hang and the item displaces the shielding curtain in order to pass through, or the shielding curtain 30 may be active in that mechanical actuation, usually automatic actuation, displaces the shielding curtain to allow items to pass. In certain embodiments, multiple shielding curtains 30 are disposed parallel to each other and each shielding curtain 30 must be traversed for an item 16 to be scanned by the system 10. This configuration further contains the electromagnetic radiation such that if the electromagnetic radiation escapes through an inner shielding curtain 30 that escaped electromagnetic radiation can be blocked by one or more outer shielding curtains 30. Any suitable number of shielding curtains may be positioned to block the entrance end 22 and the exit end 24. According to one embodiment, four to eight shielding curtains 30 are disposed parallel to each other at the entrance end 22 of the exposure station 18 and four to eight shielding curtains 30 are disposed at the exit end 24 of the exposure station 18. The slits 34 forming the individual flaps 32 of a shielding curtain 30 may be staggered with respect to adjacent shielding curtains 30 to further prevent the electromagnetic radiation from escaping the exposure station 18. According to alternate embodiments, the shielding curtain 30 may be mechanically actuated to open and close to allow the item 16 to pass through to a location where it can be exposed to electromagnetic radiation. Reference is now made to FIG. 2, which is a perspective view of a shielding curtain 31 according to the teachings of the present disclosure. The shielding curtain 31 is a single homogeneous, unitary body that is molded from a composite polymer material, as discussed in more detail below. According to one embodiment, the single, unitary body includes a plurality of flaps 32, as shown in FIG. 2. According to an alternate embodiment, the shielding curtain 31 may be formed from individually molded flaps that are molded from a composite polymer material. The shielding curtain 31 does not include molded-in mounting features. As such, the shielding curtain 31 may be mounted conventionally with fasteners received through the curtain and through a pair of clamping bars disposed on the front and the rear of the top edge of the shielding curtain 31. Certain advantages are obtained by molding the shielding curtain 31 including the plurality of flaps 32 or individual flaps 32 from a composite polymer material, as opposed to forming flaps using conventionally layered construction. For example, the composite polymer material may be more flexible than conventional leaded layered construction curtains and may have a lower coefficient of friction. Reference is made to FIGS. 3A and 3B, which are perspective views of a shielding curtain 30 with a molded-in mounting feature 36. The shielding curtain 30 may be a single, unitary body that includes a plurality of flaps, or it may be a single flap 32 with a portion of the molded-in mounting feature 36. The shielding curtain 30 includes an integrated mounting bead 36 as the molded-in mounting feature, and the shielding curtain 30 includes the plurality of flaps 32 extending from the mounting bead 36. The shielding curtain 30 and the shielding curtain 31 are each formed using a polymer fabrication process, such as injection molding, compression molding, casting, extrusion, and the like. The material that is molded or cast into the shielding curtain 30, 31 may be a composite polymer material, a lead vinyl material, or a lead rubber material. An exemplary composite polymer material includes a thermosetting polymer such as urethane and one or more heavy particulate filler, such as Tungsten powder, and/or one or more light particulate filler, such as Barium sulfate, and is sold under the trade name Brandonite. The filler material is in the form of particles or powder that is uniformly dispersed in the polymer material. Such composite polymer material is introduced into a mold as pellets or as liquid, and then formed into the desired flap or shielding curtain according to the teachings of the present disclosure. For example, a composite polymer material includes a filler material that includes either Tungsten powder or Barium sulfate or both materials in particle form that is uniformly dispersed in a urethane or other polymer. Other suitable polymers and particulate fillers are contemplated by the present disclosure. U.S. Pat. No. 8,487,029 to Wang and assigned to Globe Composite Solutions, Ltd., which is hereby incorporated by reference, describes materials and forming processes for composite polymer materials that result in a lead-free, non-toxic article that is particularly useful in radiation shielding applications. In addition, the composite polymer material is flexible to allow the item 16 to displace the flaps 32 of the shielding curtain 30, 31, while at the same time providing a barrier for the electromagnetic radiation. The shielding curtain 30, 31 formed of a composite polymer material may be compliant with the directive as to Restriction of Hazardous Substances (“RoHS”). The flaps 32 may be any thickness, for example, each flap 32 may be approximately 0.075 inches thick. Electromagnetic radiation shielding equivalency or lead equivalency corresponds to the thickness of the flaps 32 of the shielding curtain 30. For example, 1 millimeter in flap thickness corresponds to approximately 0.25 millimeters (0.010 inches) in lead equivalency. Certain embodiments of the shielding curtain 30, 31 have a uniform thickness of approximately 0.075 inches (1.9 millimeters), which corresponds to approximately 0.5 millimeters (0.020 inches) in lead equivalency. Accordingly, the shielding curtains 30, 31 can have any suitable thickness depending on the desired lead equivalency, provided that the flaps 32 remain flexible enough to be displaced by the items 16 as the items pass through the shielding curtain 30, 31. The mounting bead 36 is generally cylindrical or oblong and extends along the length of an upper edge of the shielding curtain 30. The flaps 32 are integral with the mounting bead 36 and hang from the mounting bead 36. According to an alternate embodiment, the mounting bead 36 may be molded around a reinforcing rod. Any suitable number of flaps 32 may extend from the mounting bead 36. For example, 10-16 flaps 32 or more may extend from the mounting bead 36. According to one embodiment, the mounting bead 36 and a pre-cut sheet extending from the mounting bead 36 is formed according to known polymer forming processes, such as molding, casting, or extrusion. The material formed may be a composite polymer material, a lead vinyl material, or a lead rubber material. Then, the sheet is cut to form a predetermined number of flaps 32 by cutting the slits 34 through the sheet such that the slits 34 extend from the bottom of the sheet to a location proximate the mounting bead 36, but the mounting bead 36 is not cut, such that the shielding curtain remains a single, unitary body. According to certain embodiments, the shielding curtain 30 is not cut into flaps 32. Rather, the shielding curtain 30 may be a single sheet extending from the mounting bead 36. The single sheet embodiment may be employed as an active shielding curtain, which may be useful shielding cargo that is exposed to electromagnetic radiation. In the active shielding curtain embodiment, the shielding curtain is automatically mechanically actuated to open and close to allow items to pass through. Returning to the multiple-flap embodiment, each slit 34 separates one flap 32 from an adjacent flap 32. The slits 34 may be made by an automated cutting system that is known in the machining art, such as a water jet, laser jet, cutting blade, and the like that automatically makes the flap forming slits 34 according to a software program. According to an alternate embodiment, a single flap 32 including a flap-sized mounting bead 36 may be formed, and then combined with other individually formed flaps 32 in an assembly according to the teachings of this disclosure to form a shielding curtain. With regard to the single, unitary body shielding curtain 30 with the plurality of flaps, either with or without (see FIG. 2) the mounting bead 36 or other molded-in mounting features (see FIGS. 6A-6C, a strain relief hole 38 may be formed at an upper end of the slit 34 proximate the upper edge of the shielding curtain 31 or the mounting bead 36. The strain relief holes 38 delimit each flap forming slit 34 and prevent the cut from propagating further toward the mounting bead 36 or the upper edge as the shielding curtain 30, 31 is flexed during use. The strain relief holes 38 may present a path for the electromagnetic radiation to breach a shielding curtain 30. Staggering the strain relief holes 38 in adjacent and/or successive shielding curtains 30 installed at the entrance end 22 or exit end 24 of the exposure station 18 helps prevent the electromagnetic radiation from escaping and entering the ambient environment. Additionally or in lieu of staggering the shielding curtains 30, the strain relief holes 38 may be aligned with a portion of the exposure station 18, which may prevent or reduce the electromagnetic radiation from passing through the strain relief holes 38. Reference is now made to FIGS. 4A and 4B, which are perspective views of an alternate flap configuration for the shielding curtain 31 without the mounting bead and for the shielding curtain 30, including the mounting bead 36 or other molded-in mounting feature. Each flap 32 of the shielding curtain may have a uniform thickness, as shown and described above with respect to FIGS. 2, 3A, and 3B, or a flap may have a varied or non-uniform thickness. A non-uniform thickness flap 40a is formed using the molding, casting, or extrusion processes of polymer forming and includes the mounting bead 36 or other molded-in mounting feature. And, a non-uniform thickness flap 40b does not include molded-in mounting features. Such non-uniform thickness flap 40a, 40b is an advantage over the layered strip flaps of conventional shielding curtains. The varied thickness in the flap may be implemented to provide varying lead equivalency for shielding against electromagnetic radiation. For example, the flap 40a, 40b may taper from a thicker, upper portion to a thinner, lower portion. A lower portion 42 of the varied thickness flap 40a, 40b may be thinner and have a lower lead equivalency and be less effective at blocking electromagnetic radiation than an upper portion 44. The upper portion 44 may have a greater thickness than the lower portion 42, and thus have a greater lead equivalency and be more effective in preventing electromagnetic radiation from penetrating the thicker portion of the flap 40a, 40b. Alternatively, the flap 40a, 40b may taper from a thicker, lower portion to a thinner. The upper portion 42 of the varied thickness flap 40a, 40b may be thinner and have a lower lead equivalency and be less effective at blocking electromagnetic radiation than a lower portion 44. The lower portion 44 may have a greater thickness than the upper portion 42, and thus have a greater lead equivalency and be more effective in preventing electromagnetic radiation from penetrating the thicker portion of the flap 40a, 40b. By employing a varied or non-uniform thickness flap 40a, 40b shielding curtain, different zones may be made thicker to shield more effectively against the electromagnetic radiation than other zones. The different zones may be selected to accommodate the particular shielding application depending on an emission pattern and strength of the electromagnetic radiation. In addition, the electromagnetic radiation scanning system 10 may be equipped with different varied thickness flaps 40 shielding curtains at different locations at the entrance end 22 and/or the exit end 24 of the exposure station 18. According to an alternate embodiment, individual varied thickness flaps 40a, 40b may be formed by molding, casting, or extrusion of a composite polymer material, a lead vinyl material, or a lead rubber material and then subsequently assembled to form a shielding curtain. Reference is now made to FIGS. 5A-5C, which show various surfaces of the flaps 32 of a shielding curtain 30, 31 according to embodiments of the present disclosure. The surfaces of the flaps are the surfaces that are contacted by the items 16 moved by the conveyor belt 12 through the electromagnetic radiation scanning system 10. For example, as shown in FIG. 5A, a flap 32 may have a surface feature in the form of raised contact projections 46 that extend either parallel or perpendicular to the slits 34. In another embodiment shown in FIG. 5B, a raised contact feature may be in the form of a plurality of raised bosses or dome-shaped projections 48. According to yet another embodiment shown in FIG. 5C, the raised contact features are raised parallelepipeds 50. Each of the raised contact features, the raised strips 46, the raised dome-shaped projections 48, and the raised parallelepipeds 50 provide a contact surface area that is reduced from the overall surface area of the flap 32. In this manner, friction and drag between the conveyed item 16 displacing the flaps 32 and the flaps 32 is reduced and wear of the flaps 32 may also be reduced over conventional layered shielding curtains. The raised surface features described herein could also be depressions molded into the flaps 32 of the shielding curtain 30. The surface features of FIGS. 5A-5C may be employed with any of the shielding curtain or individual flap embodiments disclosed herein. Such surface features are formed by creating a mold with the negative of the desired surface feature, then molding the curtain or individual flap from the composite polymer material including the filler material that blocks electromagnetic radiation but remains flexible to be displaced by the items. Surface area reducing surface features are not easily formed in the fabrication process of conventional layered construction shielding curtains. The raised features may also be used to indicate the level of wear of the shielding curtains in use. Reference is now made to FIGS. 6A-6C, which are perspective views of portions of a shielding curtain 33 with various molded-in mounting features that may be used in lieu of the molded-in mounting bead depending on the particular curtain mounting features associated with the scanning system where original or replacement shielding curtains or original or replacement individual flaps are installed. Molded-in mounting features as shown and described with respect to FIGS. 6A-6C are included in the mold and created when the composite polymer material is formed by the mold. In this manner, few or no additional fabrication operations may be necessary for the shielding curtain or an individual flap to be mounted to a shielding curtain assembly that is ultimately installed in an electromagnetic radiation scanning system. FIG. 6A illustrates through holes 51 that have been molded into an upper portion of the shielding curtain 33. The through holes 51 may also be molded into individual flaps 32 of the shielding curtain. The through holes 51 may be any shape or size such that they correspond to the mounting features for the shielding curtain assembly or to allow for horizontal or vertical adjustment of the shielding curtains with respect to the specific mounting configuration. Protrusions 53 or bosses as shown in FIG. 6B may also be molded into the top portion of the shielding curtain 33 or individual flaps 32. The protrusions 53 may be any suitable size and shape that corresponds with mounting features or to allow for horizontal or vertical adjustment of the shielding curtains with respect to the specific mounting configuration for the particular scanning system. FIG. 6C illustrates molded-in mounting hardware 55. The mounting hardware 55 may be a generally elongated flat bar that extends through the shielding curtain 33. According to certain embodiments, the mounting hardware 55 extends such that it is exposed on each side of the shielding curtain 33 where an exposed mounting feature 57, such as a through hole, may be used to secure the shielding curtain 33 or to allow for horizontal or vertical adjustment of the shielding curtain with respect to the specific mounting configuration of the scanning system. The composite polymer material is bonded to the mounting hardware 55 because the liquid composite polymer material in the mold forms around the mounting hardware such that when the piece is taken out of the mold, the shielding curtain 33 or an individual flap 32 is bonded to the mounting hardware 55. According to an alternate embodiment, the shielding curtain 33 may not envelop or encapsulate all sides of the mounting hardware, but rather may be molded to be bonded to one front or rear surface of the mounting hardware 55. Other mounting hardware that may be molded into the shielding curtain or individual flaps include, but are not limited to, threaded inserts, fasteners, washers, bushings, pins, and the like. Reference is now made to FIG. 7, which is an exploded, perspective view of a shielding curtain assembly 52 according to the teachings of the present disclosure. The shielding curtain assembly 52 includes the shielding curtain 30 and a multi-piece curtain suspending member 54 that supports the shielding curtain 30. When assembled, the curtain suspending member 54 receives the mounting bead 36 of the shielding curtain 30. The multi-piece curtain suspending member 54 is received by a mounting channel 56 that is secured to the electromagnetic radiation scanning system 10. According to certain embodiments, the mounting channel 56 is accessible through at least one access door disposed on one or both sides or on the top of the scanning system 10. The mounting channel 56 may be the same as in conventional electromagnetic scanning systems so as to allow the shielding curtain assembly 52 of the present disclosure to be easily retrofit to existing and in-use scanning systems. The multi-piece curtain suspending member 54 includes a front bar 58a and a rear bar 58b, where front and rear refer generally to the direction of travel of the items 16 on the conveyor belt 12 that encounter the shielding curtain 30. Each of the front and rear bars 58a, 58b defines a generally semi-circular recess 60a, 60b disposed at a lower portion of each bar 58a, 58b. Disposed above the semicircular recess 60a, 60b on each bar 58a, 58b is a plurality of fastening holes 62a, 62b. When the bars 58a, 58b are abutted together, fasteners are received through the fastening holes 62a, 62b to join the bars 58a, 58b to form the multi-piece curtain suspending member 54, which includes a bead receiving slot 64. The shape of the bead receiving slot 64 corresponds to the shape of the mounting bead 36 on the shielding curtain 30 such that the mounting bead 36 is received by and supported by the bead receiving slot 64. Unlike conventional shielding curtains that are clamped between generally flat bars and secured therebetween by fasteners that penetrate the shielding curtain, no fasteners penetrate the mounting bead 36 or any other part of the shielding curtain 30. Rather, an upward facing portion 66 of the bead receiving slot 64 contacts an underside of the mounting bead 36 and the weight of the shielding curtain 30 is opposed by the upward facing portion 66 of the bead receiving slot 64 and the mounting bead 36 is held in the bead receiving slot 64. In this manner, the shielding curtain 30 is more easily initially assembled and replaced than conventional shielding curtains. The mounting bead 36 and the corresponding bead receiving slot 64 need not be cylindrical, and any suitable shape for the mounting bead 36 and the corresponding bead receiving slot 64 is contemplated by this disclosure, including, but not limited to cross-sections of the mounting bead having a shape generally in the form of square, rectangle, oval, triangle, and the like. In addition, the shielding curtain formed with a composite polymer material allows the installed shielding curtain 30 to be curved. The mounting bead 36 may likewise be curved or wavy along the length of the shielding curtain 30. According to an alternate embodiment, the flexibility of the molded composite polymer material allows the mounting bead 36 and the shielding curtain 30 to be generally straight, but when installed into a curved or wavy mounting slot, the curtain then has a curved or wavy configuration as it extends across the entrance end 22 or the exit end 24 of the exposure station 18. The flaps 32 of the shielding curtain 30 are received through an incomplete portion 68 of the generally circular slot 64 disposed at the bottom of the slot 64. The slot 64 also functions as a pivot for the collective flaps 32. Thus, the slot 64 and mounting bead 36 junction provides rotational freedom for the movement of the collective flaps 32 of the shielding curtain 30, which may reduce stresses on the shielding curtain 30 imparted as the items 16 displace and flex the flaps 32 of the shielding curtain 30. Such stress relief may result in a longer useful life of the shielding curtain 30. The joining of the front and rear bars 58a, 58b also forms a generally elongated outer rectangular shape that corresponds to the shape of the mounting channel 56 of the electromagnetic radiation scanning system 10. According to an alternate embodiment, an exterior of the front and/or rear bars 58a, 58b or other curtain suspending member may include any suitable mounting feature that corresponds to the scanning system. For example, one or both of the bars 58a, 58b may include an angle bar that includes through holes that correspond to tapped or non-tapped through holes on the scanning system. The front bar 58a and the rear bar 58b may each be a metal part where the generally semi-circular recesses 60a, 60b and the fastener holes 62a, 62b are machined into a blank piece of metal, for example a blank of steel or aluminum, to form the final front and rear bars 58a, 58b. In one example, a fastener hole 62a, 62b in either the front or rear bar 58a, 58b may be tapped to receive a threaded fastener. According to other embodiments, the front bar 58a and the rear bar 58b may be formed of various plastics or fiberglass and may include a bearing-type material and/or a lubricant proximate the slot to facilitate rotation of the mounting bead 36 within the slot 64, as described above. According to an alternate embodiment, the multi-piece curtain suspending member receives individual flaps 32 that are each formed with a mounting bead 36 with a shape that corresponds to the bead receiving slot 68. The individual flaps 32 are positioned to be adjacent to each other to minimize a distance between adjacent flaps 32 through which electromagnetic radiation may pass, yet each individual flap 32 is free to flex and be displaced separately such that the item can pass through the shielding curtain 30. The receiving slot 68 may also allow the shielding curtain 30 to move laterally more freely to act as a swinging hinge to permit items to pass through the shielding curtain 30 and enter or exit the exposure station 18. Reference is now made to FIG. 8, which is an exploded, perspective view of an alternate embodiment of a shielding curtain assembly 70. The curtain receiving assembly 70 includes a curtain receiving bar 72, which functions as a curtain suspending member, and the shielding curtain 30. The curtain receiving bar 72 is a single, unitary elongated member that includes an incomplete circular slot 74, similar to that described above with respect to the multi-piece curtain support 54 of FIG. 5. The incomplete circular slot 74 is sized and shaped to receive the mounting bead 36 of the shielding curtain 30 to allow the collective flaps 32 to be suspended to block the entrance end 22 or the exit end 24 of the radiation exposure station 18. The mounting bead 36 and the slot 74 may be any suitable shape as describe above with respect to the embodiment shown in FIG. 7. Unlike conventional shielding curtains that are clamped between generally flat bars and secured therebetween by fasteners that penetrate the shielding curtain, no fasteners penetrate the mounting bead 36 or any other part of the shielding curtain 30. Rather, an upward facing portion 76 of the incomplete circular slot 74 contacts and underside of the mounting bead 36 and the weight of the shielding curtain 30 is opposed by the upward facing portion 76 of the incomplete circular slot 74 and the mounting bead 36 is held in the incomplete circular slot 74. In this manner, the shielding curtain 30 is more easily initially assembled and replaced than conventional shielding curtains. The mounting bead 36 and the corresponding incomplete circular slot 74 need not be cylindrical, and any suitable shape for the mounting bead 36 and the corresponding slot 74 is contemplated by this disclosure, including, but not limited to cross-sections of the mounting bead having a shape generally in the form of square, rectangle, oval, triangle, and the like. The flaps 32 of the shielding curtain 30 are received through an incomplete portion 78 of the incomplete circular slot 74 disposed at the bottom of the slot 74. The slot 74 also functions as a pivot for the collective flaps 32. Thus, the slot 74 and mounting bead 36 junction provides rotational freedom for the movement of the collective flaps 32 of the shielding curtain 30, which may reduce stresses on the shielding curtain 30 imparted as the items 16 displace and flex the flaps 32 of the shielding curtain 30. Such stress relief may result in a longer useful life of the shielding curtain 30. The outer shape of the curtain receiving bar 72 is generally shaped in an elongated rectangular shape to correspond to the mounting channel 56 secured above and across the entrance end 22 and the exit end 24 of the exposure station 18. As described above, the mounting channel 56 may be similar to those found in existing and in-use electromagnetic radiation scanning systems, which facilitates retrofitting existing systems with replacement shielding curtain assemblies 70 according to the teachings of the present disclosure. According to certain embodiments, the curtain receiving bar 72 is an elongated, thin walled member that may be formed by extrusion of a polymer or metallic material, such as aluminum, a composite polymer material, a thermosetting polymer, or a thermoplastic polymer. According to other embodiments, the curtain receiving bar 72 is a metallic or polymer material formed by a different molding process other than extrusion, such as injection molding. The curtain receiving bar 72 may be any suitable length, for example it may have a length of between 35 inches and 50 inches, for example approximately 40 inches. The curtain receiving bar 72 may be extruded and/or cut to any suitable length to span across the entrance end 22 or exit end 24 of the exposure station 18 of the electromagnetic radiation scanning system 10. According to an alternate embodiment, the curtain receiving bar 72 receives individual flaps 32 that are each formed with a mounting bead 36 with a shape that corresponds to the incomplete circular slot 74. The individual flaps 32 are positioned to be adjacent to each other to minimize a distance between adjacent flaps 32 through which electromagnetic radiation may pass, yet each individual flap 32 is free to flex and be displaced separately such that the item can pass through the shielding curtain 30. Reference is now made to FIG. 9, which illustrates an alternate embodiment of a single-piece curtain receiving bar 80. The single-piece curtain receiving bar 80 has a different profile geometry than the curtain receiving bar 72 shown in FIG. 8. The curtain receiving bar 80 includes a pair of flanges 82 extending proximate a top portion of the curtain receiving bar 80. The flanges 82 are configured to receive a fastener to secure the curtain receiving bar 80 to the exposure station 18 of the electromagnetic radiation scanning system 10. Similar to the embodiment shown in FIG. 6, the curtain receiving bar 80 includes a bead receiving slot 84, and it is a generally thin-walled part formed by injection molding, pultrusion, or extrusion of a polymer or a metallic material. This disclosure contemplates any suitable extrusion profile that can be mounted to the electromagnetic radiation scanning system 10 and includes a bead receiving slot 84 that receives the mounting bead 36 of the shielding curtain 30. A single unitary body shielding curtain 30 with a mounting bead 36 or individual flaps 32 of a shielding curtain may be received and held in place by the single-piece curtain receiving bar 80, similar to the embodiments described above with respect to FIGS. 7 and 8. According to an alternate embodiment, a top portion of the curtain receiving bar may be open to allow the shielding curtain to be dropped in from above the curtain receiving bar such that the bead receiving slot supports and suspends the mounting bead 36 or other integrated mounting feature. In the foregoing description of certain embodiments, specific terminology has been resorted to for the sake of clarity. However, the disclosure is not intended to be limited to the specific terms so selected, and it is to be understood that each specific term includes other technical equivalents which operate in a similar manner to accomplish a similar technical purpose. Terms such as “left” and right”, “front” and “rear”, “above” and “below,” “top” and “bottom” and the like are used as words of convenience to provide reference points and are not to be construed as limiting terms. In addition, the foregoing describes only some embodiments of the invention(s), and alterations, modifications, additions and/or changes can be made thereto without departing from the scope and spirit of the disclosed embodiments, the embodiments being illustrative and not restrictive. Furthermore, invention(s) have been described in connection with what are presently considered to be the most practical and preferred embodiments, it is to be understood that the invention is not to be limited to the disclosed embodiments, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the invention(s). Also, the various embodiments described above may be implemented in conjunction with other embodiments, e.g., aspects of one embodiment may be combined with aspects of another embodiment to realize yet other embodiments. Further, each independent feature or component of any given assembly may constitute an additional embodiment. |
|
050531905 | claims | 1. A water cooled reactor and pressuriser assembly comprising a reactor core, a primary coolant circuit, a pressuriser, a pressure vessel, at least one vent conduit and at least one surge conduit, the reactor core and at least a portion of the primary water coolant circuit being enclosed by the pressure vessel, the primary water coolant being arranged to cool the reactor core, the primary water coolant circuit having an upper portion and a lower portion, the lower portion being positioned below any normal effective range of water levels of the primary water coolant circuit, the upper portion being positioned above the lower portion, the pressuriser having a water spaced and a steam space, at least a portion of the water space of the pressuriser being positioned above the upper portion of the primary water coolant circuit. the at least one surge conduit communicating between the pressuriser and the primary water coolant circuit to connect the water space of the pressuriser with the lower portion of the primary water coolant circuit, the at least one surge conduit being arranged to have a relatively low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and a relatively high flow resistance for water from the primary water coolant circuit to the water space of the pressuriser, the at least one vent conduit communicating between the pressuriser and the primary water coolant circuit to connect the steam space of the pressuriser with the upper portion of the primary water coolant circuit whereby the at least one vent conduit which communicates between the steam space of the pressuriser and the upper portion of the primary water coolant circuit allows excess vapour formed in the primary water coolant circuit to flow to the steam space of the pressuriser to increase the stability of the assembly. the reactor core and at least a portion of the primary water coolant circuit being enclosed by the pressure vessel, the primary water coolant circuit being arranged to cool the reactor core, the primary water coolant circuit having an upper portion, and a lower portion, the lower portion being positioned below any normal effective range of water levels of the primary water coolant circuit, the upper portion being positioned above the lower portion, the pressuriser having a water space and a steam space, at least a portion of the water space of the pressuriser being positioned above the upper portion of the primary water coolant circuit, the at least one surge conduit communicating between the pressuriser and the primary water coolant circuit to connect the water space of the pressuriser with the lower portion of the primary water coolant circuit, the at least one surge conduit being arranged to have a relatively low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and a relatively high flow resistance for water from the primary water coolant circuit to the water space of the pressuriser, the plurality of vent conduits communicating between the pressuriser and the primary water coolant circuit to connect the steam space of the pressuriser with the upper portion of the primary water coolant circuit, each of the plurality of vent conduits extending into the primary water coolant circuit to a different elevation, the plurality of vent conduits allowing excess vapour formed in the primary water coolant circuit to flow to the steam space of the pressuriser to increase the stability of the assembly, the plurality of vent conduits controlling the effective water level in the primary water coolant circuit. 2. A water cooled nuclear reactor as claimed in claim 1 in which the pressure vessel has a lower region, and an upper region, the reactor core is arranged in the lower region of the pressure vessel, the primary coolant circuit comprising a riser passage and a downcomer passage, at least one heat exchanger, the riser passage conveys relatively hot water and steam to the at least one heat exchanger, the downcomer passage conveys relatively cool water from the at least one heat exchanger to the reactor core. 3. A water cooled nuclear reactor as claimed in claim 2 in which the at least one heat exchanger is positioned in the upper region of the downcomer passage. 4. A water cooled nuclear reactor as claimed in claim 3 in which the at least one heat exchanger is a steam generator. 5. A water cooled nuclear reactor as claimed in claim 1 in which the primary water coolant circuit comprises at least one pump to assist the circulation of primary water coolant. 6. A water cooled nuclear reactor as claimed in claim 3 in which the reactor core, the primary coolant circuit and the pressuriser are arranged as an integral unit enclosed by the pressure vessel, at least one casing being arranged in the pressure vessel to substantially divide the pressure vessel into a first chamber and a second chamber, the reactor core and the primary coolant circuit being arranged in the second chamber, the pressuriser being arranged in the first chamber, the casing preventing interaction between the water in the primary water coolant circuit and the water in the water space of the pressuriser. 7. A water cooled nuclear reactor as claimed in claim 6 in which the pressuriser forms a surge tank positioned in the first chamber, the surge tank being defined by the pressure vessel and the casing. 8. A water cooled nuclear reactor as claimed in claim 7 in which the casing comprise a peripheral region and an annular member, the annular member extends downwards from the peripheral region of the casing, an annular passage being formed between the annular member of the casing and the pressure vessel for the flow of water from the water space of the pressuriser to the primary coolant circuit. 9. A water cooled nuclear reactor as claimed in claim 7 in which the casing comprise a peripheral region and an annular member, the annular member extends downwards from the peripheral region of the casing, the annular member being secured to the pressure vessel to form an annular lower portion of the surge tank with the pressure vessel. 10. A water cooled nuclear reactor as claimed in claim 7 in which the casing comprises a peripheral region, a central region and an annular member, the annular member extends downwards from the central region of the casing, the peripheral region of the casing being secured to the pressure vessel, the annular member having a lower end, the annular member being sealed at its lower end to form a lower portion of the surge tank. 11. A water cooled nuclear reactor as claimed in claim 7 in which a peripheral region of the casing is secured to the pressure vessel. 12. A water cooled nuclear reactor as claimed in claim 6 in which the casing is arranged to divide the pressure vessel into a first vertically upper chamber and a second vertically lower chamber. 13. A water cooled nuclear reactor as claimed in claim 8 in which the casing comprises a bottom member positioned below the reactor core, the casing dividing the pressure vessel into a first outer chamber and a second inner chamber, the second inner chamber being substantially defined by the casing. 14. A water cooled nuclear reactor as claimed in claim 7, in which the water space of the surge tank has a lower portion, the at least one surge conduit connects the lower portion of the water space of the surge tank with a portion of the downcomer passage in the region of the heat exchanger. 15. A water cooled nuclear reactor as claimed in claim 7, in which the water space of the surge tank has a lower portion, the at least one surge conduit connects the lower portion of the water space of the surge tank with the primary water coolant circuit in the region of the reactor core. 16. A water cooled nuclear reactor as claimed in claim 15 in which the at least one surge conduit connects the lower portion of the water space of the surge tank with the primary water coolant circuit below the reactor core. 17. A water cooled nuclear reactor as claimed in claim 7, in which the water space of the surge tank has a lower portion, the at least one surge conduit connects the lower portion of the water space of the surge tank with a lower portion of the downcomer passage below the heat exchanger. 18. A water cooled nuclear reactor as claimed in claim 2 in which the riser passage is defined by a hollow cylindrical member, the downcomer passage being defined between the hollow cylindrical member and the at least a part of the pressure vessel. 19. A water cooled nuclear reactor as claimed in claim 13 in which the riser passage is defined by a hollow cylindrical member, the downcomer passage being defined between the hollow cylindrical member and the casing. 20. A water cooled nuclear reactor as claimed in claim 1 in which the at least one surge conduit comprises a re-entrant nozzle. 21. A water cooled nuclear reactor as claimed in claim 1 in which the at least one surge conduit comprises a hydraulic diode. 22. A water cooled nuclear reactor as claimed in claim 1 in which the at least one vent conduit which communicates between the pressuriser and the primary coolant circuit comprises at least one pipe which interconnects at least one port in the casing with the steam space in the pressuriser. 23. A water cooled nuclear reactor as claimed in claim 1 in which at least one of the vent conduits which communicate between the pressuriser and the primary water coolant circuit comprises a spray nozzle. 24. A water cooled nuclear reactor as claimed in claim 2 in which at least one of the means which communicates between the pressuriser and the primary water coolant circuit connects the steam space of the pressuriser with the primary water coolant circuit above the heat exchanger. 25. A water cooled nuclear reactor as claimed in claim 1 in which the water cooled nuclear reactor is an integral pressurised water reactor. 26. A water cooled nuclear reactor as claimed in claim 25 in which the pressuriser has heating means to heat the water in the water space. 27. A water cooled nuclear reactor as claimed in claim 1 in which the water cooled nuclear reactor is an integral indirect cycle boiling water reactor, the at least one vent conduit which communicates between the steam space of the pressuriser and the upper portion of the primary water coolant circuit controls the effective water level in the primary water coolant circuit. 28. A water cooled nuclear reactor as claimed in claim 1 in which the pressuriser is a separate pressuriser. 29. A water cooled nuclear reactor as claimed in claim 28 in which the water cooled nuclear reactor is an integral indirect cycle boiling water reactor, the at least one vent conduit which communicates between the steam space of the pressuriser and the upper portion of the primary water coolant circuit controls the effective water level in the primary water coolant circuit. 30. A water cooled nuclear reactor as claimed in claim 28 in which the water cooled nuclear reactor is an integral pressurised water reactor. 31. A water cooled nuclear reactor as claimed in claim 2 in which the at least one vent conduit extends into the pressuriser steam space by at least a distance equal to the head loss due to flow in the downcomer passage between the normal effective range of water levels of the primary water coolant circuit and the elevation at which the at least one surge conduit communicates with the lower portion of the primary water coolant circuit. 32. A water cooled nuclear reactor as claimed in claim 31 in which the at least one surge conduit communicates with the lower portion of the primary water coolant circuit at the highest practical evaluation below the normal effective range of water levels of the primary water coolant circuit to minimize the distance that the at least one vent conduit extends into the pressuriser steam space. 33. A water cooled nuclear reactor as claimed in claim 27 in which a plurality of vent conduits communicate between the steam space of the pressuriser and the upper portion of the primary water coolant circuit, the vent conduits extending into the primary water coolant circuit to different elevations. 34. A water cooled nuclear reactor as claimed in claim 29 in which a plurality of vent conduits communicate between the steam space of the pressuriser and the upper portion of the primary water coolant circuit, the vent conduits extending into the primary water coolant circuit to different elevations. 35. A water coolant integral indirect cycle boiling reactor and pressuriser assembly comprising a reactor core, a primary coolant circuit, a pressuriser, a pressure vessel, a plurality of vent conduits and at least one surge conduit, |
description | This application claims priority from Korean Patent Application No. 10-2006-0097972 filed on Oct. 9, 2006, the disclosure of which is incorporated herein by reference in its entirety. 1. Technical Field The present disclosure relates to a method of manufacturing a mask, and more particularly, to a method of manufacturing a mask for forming a pattern on a wafer using a self-aligning double patterning process. 2. Discussion of the Related Art A Self-Aligning Double Patterning (SADP) method is used to form a pitch smaller than a minimum pitch that can be formed using exposure equipment. In the SADP method, a mask data pattern is designed using a first pitch which can be formed using exposure equipment, and a first hard mask layer pattern is formed using a mask layer that corresponds to the mask data pattern. Thereafter, a sacrificial layer and a second hard mask layer are formed on the first hard mask layer pattern, and a hard mask layer pattern having a second pitch less than the first pitch is formed through a planarization process and an anisotropic etching process. Using the above SADP method, a pitch, which is less than the minimum pitch that can be formed using exposure equipment, can be formed so that fine patterns that facilitate high integration can be formed. According to an exemplary embodiment of the present invention, a method of manufacturing a mask includes designing a second mask data pattern for forming a first mask data pattern, creating a first emulation pattern, which is determined from the second mask data pattern, using a first emulation, creating a second emulation patter, which is determined from the first emulation pattern, using a second emulation, comparing a pattern, in which the first and second emulation patterns overlap, with the first mask data pattern, and manufacturing a mask layer, which corresponds to the second mask data pattern, according to the results of the comparison. According to an exemplary embodiment of the present invention, a method of manufacturing a mask includes designing a first mask data pattern, designing a second mask data pattern for forming the first mask data pattern, creating a first emulation pattern, which is determined from the second mask data pattern, using a layout-based SADP emulation, comparing the first emulation pattern with the first mask data pattern, modifying the second mask data pattern according to the results of the comparison, performing an Optical Proximity Correction (OPC) on the modified second mask data pattern, creating a second emulation pattern, which is determined from the second mask data pattern, on which the OPC has been performed, using an image-based emulation, creating a third emulation pattern, which is determined from the second emulation pattern, using the image-based SADP emulation, comparing a pattern, in which the second and third emulation patterns overlap, with the first mask data pattern, and manufacturing a first mask layer, which corresponds to the second mask data pattern, according to the results of the comparison. According to an exemplary embodiment of the present invention, a method of manufacturing a mask includes designing a first mask data pattern, designing a second mask data pattern for forming the first mask data pattern, creating a first emulation pattern, which is determined from the second mask data pattern, using a layout-based SADP emulation, comparing the first emulation pattern with the first mask data pattern, designing third mask data patterns, which are used to create partial patterns that are not formed using the second mask data pattern, according to the results of the comparison, performing OPC on the second and third mask data patterns, creating a second emulation pattern, which is determined from the second mask data pattern on which the OPC has been performed, using an image-based SADP emulation, creating a third emulation pattern, which is determined from the second emulation pattern, using the image-based SADP emulation, creating a fourth emulation pattern, which is determined from the third mask data patterns on which the OPC has been performed, using the image-based SADP emulation, comparing a pattern, in which the second to fourth emulation patterns overlap, with the first mask data pattern, and manufacturing first and second mask layers, which respectively correspond to the second and third mask data patterns, according to the results of the comparison. According to an exemplary embodiment of the present invention, a method of manufacturing a mask includes defining one or more parameters related to respective step processes of an SADP process, modeling each of the step processes using the parameters, and creating a second emulation pattern by applying step process models to a first emulation pattern. Exemplary embodiments of the present invention will be described below in more detail with reference to the accompanying drawings. The present invention may be embodied in many different forms and should not be construed as limited to the embodiments set forth herein. FIG. 1 is a flowchart for illustrating a method of manufacturing a mask according to an exemplary embodiment of the present invention. FIG. 2 is a flowchart illustrating an image-based SADP emulation of FIG. 1 according to an exemplary embodiment of the present invention. FIGS. 3A to 3K are diagrams for illustrating the method of manufacturing a mask according to an exemplary embodiment of the present invention. With reference to FIGS. 1 and 3A, a first mask data pattern 10 to be formed on a wafer is designed at step S110. The first mask data pattern 10 may include a plurality of patterns 11 to 16. The pitch between each pair of neighboring patterns among patterns 11 to 16 may be a first pitch P1 or a second pitch P2. The first pitch P1 is a value that can be achieved using typical exposure equipment, and the second pitch P2 is a value less than the minimum pitch that can be achieved using the typical exposure equipment The pitch between two patterns, for example, the patterns 11 and 12, is referred to as the shortest distance from the edge of a pattern to the edge of another pattern, for example, from the edge of the pattern 11 to the edge of another pattern 12. In FIGS. 1 and 3B, a second mask data pattern 20 for forming the first mask data pattern 10 using an SADP process is designed at step S120. The first mask data pattern 10 includes a main pattern 10a (represented by a dark dot pattern) and a sub-pattern 10b (represented by a light dot pattern) according to a decomposition guide. The main pattern 10a may correspond to the second mask data pattern 20. The decomposition guide is a rule for the classification of the first mask data pattern 10 designed at a prior stage to perform the SADP process. An example of the decomposition guide may include the definition “when the pitch between two neighboring patterns is less than a predetermined pitch (for example, P1), one pattern is the main pattern 10a and another pattern is the sub-pattern 10b.” As shown in FIG. 3B, when the pitch between each pair of neighboring patterns among the patterns 11 to 16 is the second pitch P2, patterns 11, 12a, 13, 14a and 15 may be the main pattern 10a, and patterns 12b, 14b and 16 may be the sub-pattern 10b. Using the decomposition guide as a method of designing the second mask data pattern 20 is an exemplary method. With reference to FIGS. 1 and 3C, a first emulation pattern 30, which is determined from the second mask data pattern 20, is created using a layout-based SADP emulation at step S130. The layout-based SADP emulation is an exemplary method of examining whether the first mask data pattern 10 can be normally formed from the second mask data pattern 20. In the layout-based SADP emulation, an emulation tool receives process conditions, such as pitches between patterns, the deposition thickness of a sacrificial layer, and the deposition thickness of a hard mask layer, and determines the first emulation pattern 30, which is formed from the second mask data pattern 20, using the above conditions. The sacrificial layer may comprise a silicon oxide layer, and the hard mask layer may comprise a polysilicon layer. The emulation tool receives pitches between patterns without receiving the deposition thickness of a sacrificial layer or the deposition thickness of a hard mask layer. When the pitch between two neighboring patterns in the second mask data pattern 20 is less than a predetermined pitch (for example, P1), a new pattern is created between the two patterns, and thus the first emulation pattern 30 may be determined. The layout-based SADP emulation may not use exposure conditions, such as the wavelength, dosage and energy of light. Process conditions used to perform the layout-based SADP emulation may not be used. The reason for this is because the layout-based SADP emulation is used to examine whether all of the patterns of the first mask data pattern 10 are normally formed using the second mask data pattern 20, but is not used to examine the detailed shapes of the patterns formed using the second mask data pattern 20. When it is found that all of the patterns of the first mask data pattern 10 are normally formed using the second mask data pattern 20 even though the exposure conditions are used, the exposure conditions may be used. Thereafter, the first emulation pattern 30 and the first mask data pattern 10 are compared with each other at step S140. Referring to FIG. 3C, in the first emulation pattern 30 formed using the layout-based SADP emulation, a pattern 31 is formed between the portions of the pattern 11, a pattern 33 is formed between the patterns 12a and 14a, a pattern 32a is formed between the patterns 11 and 13, and a pattern 34 is formed between the patterns 13 and 15. When the first emulation pattern 30 and the first mask data pattern 10 are compared with each other, it can be seen that patterns 32b and 36, which must be formed, are not formed, and the patterns 31 and 33, which must not be formed, are formed. The pattern 12a and the pattern 14a must be respectively connected with the pattern 32a and the pattern 34, but the patterns are not connected. Depending on the differences between the first emulation pattern 30 and the first mask data pattern 10, subsequent steps to be performed are changed. That is, the first mask data pattern 10 is redesigned at step S110 if it is determined that the first mask data pattern 10 has been erroneously designed at a prior stage. The second mask data pattern 20 is redesigned at step S120 if it is determined that the second mask data pattern 20 has been erroneously designed at a prior stage. The partial patterns of the second mask data pattern 20 are modified, or separate third mask data patterns 40 and 50 are designed if it is determined at step S150 that the differences between the first emulation pattern 30 and the first mask data pattern 10 are insignificant. With reference to FIGS. 3D and 3E, the second mask data pattern 20 is modified and the separate third mask data patterns 40 and 50 are designed at step S150. FIG. 3D shows a modified second mask data pattern 20a and the newly designed third mask data patterns 40 and 50 overlapping each other. FIG. 3E shows the modified second mask data pattern 20a and the newly designed third mask data patterns 40 and 50 separately. With reference to FIGS. 3D and 3E, the modification of adding the second mask data pattern 20 to dummy patterns 19 is performed to form the partial patterns 32b and 36 that were not formed in the first emulation pattern 30. The dummy patterns 19 are formed where the dummy patterns 19 do not cause a malfunction in the operation of the completed semiconductor device. Although the dummy patterns 19, may be formed to the second mask data pattern 20, the dummy patterns 19 may be designed in the form of separate third mask data patterns according to an exemplary embodiment of the present invention. To remove the patterns 31 and 33 that must not be formed in the first emulation pattern 30, a third mask data pattern 40, including trimming patterns 41, is designed. To respectively connect the partial patterns 12a and 14a of the first emulation pattern 30 with the partial patterns 32a and 34, the third mask data pattern 50, including connection patterns 51a, is designed. A process of forming a first emulation pattern, which is determined from the modified second mask data pattern 20a and the newly designed third mask data patterns 40 and 50, using the layout-based SADP emulation, and remodifying the modified second mask data pattern 20a or modifying the newly designed third mask data patterns 40 and 50 may be repeated. With reference to FIGS. 1 and 3F, Optical Proximity Correction (OPC) is performed on the modified second mask data pattern 20a and the third mask data patterns 40 and 50 at step S160. Second mask data pattern 20b, which is formed by performing the OPC on the modified second mask data pattern 20a, is shown in FIG. 3F. A rule table-based correction method is a method of listing the amounts of correction depending on the arrangements of patterns in a rule table, and correcting the patterns of the mask data pattern with reference to the rule table. Although the correction method has a simple correction sequence, it may be complicated to list the entire range of variation in an actual mask data pattern in the table. A model-based correction method is a method of correcting a mask data pattern to determine the shape that will be transferred onto a wafer based on pattern and wafer process conditions, and form a desired pattern on the wafer. That is, evaluation marks are formed on the edge of a mask data pattern to perform the OPC, and the edge is divided according to the evaluation marks. The intensity of light around the evaluation marks is calculated and the divided edge is moved, so that the mask data pattern is modified. Thereafter, the second mask data pattern 20b, on which the OPC has been performed, is verified, and the OPC is repeated until there is no pattern to be corrected at step S165. With reference to FIG. 1, an emulation pattern, which is determined from the second and third mask data patterns 20b, 40 and 50 on which the OPC has been performed, is created using an image-based SADP emulation. The image-based SADP emulation uses process conditions, such as pitches between patterns, the deposition thickness of a sacrificial layer, the deposition thickness of a hard mask layer, and exposure conditions, such as the wavelength, dosage and energy of light. The image-based SADP emulation uses the exposure conditions and the process conditions so that the actual pattern formed on the wafer can be formed using the second mask data pattern 20b and the third mask data patterns, on which the OPC has been performed. With reference to FIGS. 2, 3G and 3H, the image-based SADP emulation is described. FIG. 3G shows a second emulation pattern 60a, which is formed using the second mask data pattern 20b on which the OPC has been performed. FIG. 3H shows a pattern in which the second emulation pattern 60a, shown in FIG. 3G, and a third emulation pattern 60b, obtained from the second emulation pattern 60a, overlap. The second emulation pattern (e.g., the one indicated by reference numeral 60a in FIG. 3G), which is determined from the second mask data pattern 20b on which the OPC has been performed, is created using a general emulation at step S410. Emulation tools that can be used include, for example, the Mentor's Calibre and Synopsys's Proteus. The term ‘general emulation’ refers to determination of the emulation pattern 60a that can be created from the second mask data pattern 20b, on which the OPC has been performed, using the process conditions and the exposure conditions. Thereafter, the third emulation pattern 60b, which is determined from the second emulation pattern 60a, is created using the SADP emulation at step S420. Parameters related to the step processes of an SADP process are defined at step S422. The step processes of the SADP process may be sequentially performed to form an etching target layer and a first hard mask layer pattern on a substrate (refer to FIGS. 8A and 8B), conformally form a sacrificial layer on the first hard mask pattern and an exposed etching target layer (refer to FIG. 8C), form a second hard mask layer on the sacrificial layer (refer to FIG. 8D), planarize the second hard mask layer and the sacrificial layer so that the upper surface of the first hard mask pattern is exposed (refer to FIG. 8E) and the sacrificial layer exposed between the first hard mask layer pattern and the planarized second mask layer is removed (refer to FIGS. 8F, 8G and 8H), and remove the etching target layer exposed by the first hard mask layer pattern and the planarized second hard mask layer (refer to FIG. 8I). The parameters related to the step processes may include, for example, the deposition thickness of at least one layer formed through each of the step processes, or the degree of skew that occurs after each of the step processes is performed. The deposition thickness of at least one layer formed through each of the step processes may be exemplified by the deposition thickness of the first and/or second hard mask layer pattern, and the deposition thickness of the sacrificial layer. The skew that occurs after each of the step processes is performed may be exemplified by skew that occurs when the sacrificial layer and the etching target layer are removed using an etching process. Thereafter, each of the step processes is modeled using the defined parameters at step S424. For example, the modeling method may perform addition, subtraction, multiplication and division on the parameters that are defined to reveal the representative feature of each of the step processes. To exhibit the detailed features of each of the step processes, modeling may be performed to sufficiently reflect variation in patterns after each process step is performed while considering the distances between patterns in the second emulation pattern 60a and the characteristics of process equipment as well as the defined parameters. Such a modeling method may be modified in various ways according to exemplary embodiments of the present invention. Thereafter, each step process model, which is obtained through the modeling, is applied to the second emulation pattern (refer to reference numeral 60a (the dark dot pattern) of FIG. 3H). The third emulation pattern (refer to reference numeral 60b (the light dot pattern) of FIG. 3H) is thus created. Each of the created second emulation pattern 60a and the third emulation pattern 60b may have a hierarchical structure. In the image-based SADP emulation, the detailed shapes of the patterns are determined by modeling the SADP process, so that the emulation patterns 60a and 60b created using the image-based SADP emulation may be complicated. Accordingly, when the respective shapes of the patterns are stored in a storage medium coupled to an emulation tool, the amount of data to be stored or processed may be considerable. Accordingly, as shown in FIG. 4, the lowest layer (for example, layer 3) stores unit patterns, and a higher layer (for example, layer 1 or 2) stores only location information, which indicates where the unit patterns in the emulation patterns are located, so that the amount of data can be reduced. For example, location information 1 in layer 2 is used to form a pattern using a combination of unit patterns 1, 2 and 3, and location information ‘a’ in layer 3 is used to determine where the pattern is disposed in the emulation patterns 60a and 60b. With reference to FIGS. 1, 3I and 3J, a pattern, in which the second to fourth emulation patterns 60a, 60b and 70a overlap, and the first mask data pattern 10 are compared with each other at step S180. The second and third emulation patterns 60a and 60b are obtained from the second mask data pattern 20b, on which the OPC has been performed, and the fourth emulation pattern 70a is obtained from the third mask data pattern on which the OPC has been performed. Depending on the differences between the second to fourth emulation patterns 60a, 60b and 70a and the first mask data pattern 10, subsequent steps to be performed are changed. That is, the first mask data pattern 10 may be redesigned at step S110 if it is determined that the mask data pattern 10 has erroneously designed at a prior stage. The second mask data pattern 20 may be redesigned at step S120 if it is determined that the second mask data pattern 20 has been erroneously designed at a prior stage. The third mask data patterns 40 and 50 may be redesigned at step S150 if it is determined that the third mask data patterns 40 and 50 have been erroneously designed at a prior stage. If it is determined that it is not necessary to redesign the first to third mask data patterns 10, 20, 40 and 50, the first to third mask data patterns 10, 20, 40 and 50 may be partially modified. FIG. 3I shows the second and third emulation patterns 60a and 60b and the fourth emulation pattern 70a overlapping each other. With reference to FIG. 3I, it can be seen that a bridge is created between the second and third emulation patterns 60a and 60b by the fourth emulation pattern 70a. To remove such a bridge, the third mask data patterns 50, including the connection patterns (e.g., those indicated by reference numeral 51a in FIG. 3E) may be modified. The modified third mask data patterns 50 are shown in FIG. 3J. The connection patterns 51 of FIG. 3J are reduced in size compared to the connection patterns 51a of FIG. 3E. In an exemplary embodiment, the pattern, in which the second to fourth emulation patterns 60a, 60b and 70a overlap, and the first mask data pattern 10 may be compared with each other, and conflict points may be found. The conflict points may include, for example, the case where a notch occurs, the case where a bridge occurs, and the case where variation in Critical Dimension (CD) occurs, according to a classification reference. Such a classification reference may be modified according to exemplary embodiments of the present invention. Thereafter, OPC is performed on the newly modified third-mask data pattern 50 at step S160. Verification is performed at step S165. The image-based SADP emulation is performed on the second mask data pattern 20b and the third mask data patter, on which the OPC has been performed, at step 170. Comparison is performed at step S180. Thereafter, as a result of the image-based SADP emulation, if it is determined that the second mask data pattern 20b and the third mask data patterns, on which the OPC has been performed, have been normally manufactured, the first and second mask layers are manufactured at step S190. The first and second mask layers respectively correspond to the second mask data pattern 20b and the third mask data patterns, on which the OPC has been performed. FIG. 5 is a flowchart illustrating a method of manufacturing a mask according to an exemplary embodiment of the present invention. With reference to FIG. 5, a method of manufacturing a mask according to this exemplary embodiment of the present invention differs from the exemplary embodiment shown in FIG. 1 in that, after the comparison between the first emulation pattern and the first mask data pattern, the separate third mask data patterns are not redesigned but only the second mask data pattern is modified at step S151. Accordingly, the OPC is performed only on the modified second mask data pattern at step 161, and then a first mask layer, which corresponds to the second mask data pattern, on which the OPC has been performed, is manufactured at step S191. FIG. 6 is a flowchart for illustrating a method of manufacturing a mask according to an exemplary embodiment of the present invention. With reference to FIG. 6, the method of manufacturing a mask according to the this exemplary embodiment of the present invention differs from the exemplary embodiment shown in FIG. 1 in that, after the comparison between the first emulation pattern 10 and the first mask data pattern, the second mask data patterns are not modified but only the separate third mask data patterns are newly designed at step S152. Accordingly, the OPC is performed on the second and third mask data patterns at step S162. FIG. 7 is a flowchart illustrating a method of manufacturing a mask according to an exemplary embodiment of the present invention. With reference to FIG. 7, the method of manufacturing a mask according to the this exemplary embodiment of the present invention differs from the method of the exemplary embodiment shown in FIG. 1 in that the layout-based SADP emulation is not performed, but only the image-based SADP emulation is performed. In an exemplary embodiment of the present invention, the second mask data pattern may be modified, and separate third mask data patterns may be newly designed as a result of the comparison at step S426, as shown in FIG. 1. Furthermore, as a result of the comparison at step 426, only separate third mask data patterns may be newly designed, as shown in FIG. 6. With reference to FIGS. 8A to 8I, an SADP process is described according to an exemplary embodiment of the present invention. With reference to FIG. 8A, an etching target layer 220a, a first hard mask layer 230a, an anti-reflective film 240, and a first photoresist film pattern 310 are sequentially layered on a substrate 210. The etching target layer 220a is a layer that is to be etched, and may be a conductive layer for forming, for example, gate electrodes, bitlines, capacitor or storage nodes, or may be an insulation layer, such as an oxide layer or a nitride layer, in which contact holes are formed. The first hard mask layer 230a is formed of a layer having a high etching selectivity for the etching target layer 220a. When the etching target layer 220a is an insulation layer, the first hard mask layer 230a may be a polysilicon oxide layer. In an exemplary embodiment, the first photoresist film pattern 310 is formed using the first mask layer, which corresponds to the corrected second mask data pattern 20b. Accordingly, the values of pitches between the patterns of the first photoresist film pattern 310 may be equal to the first pitch P1. With reference to FIG. 8B, a first hard mask layer 230a exposed by the first photoresist film pattern 310 is patterned. Thus, a first hard mask layer pattern 230 having the first pitch P1 is formed. Thereafter, the first photoresist film pattern 310 is removed through an ashing/strip process. With reference to FIG. 8C, a sacrificial layer 250 is formed on the first hard mask layer pattern 230 and the exposed etching target layer 220a on the substrate 210 to a predetermined thickness. The sacrificial layer 250 has good step coverage to be conformally layered along the first hard mask layer pattern 230 and the exposed etching target layer 220a on the substrate 210. The sacrificial layer 250 may comprise a material having a high etching selectivity for the first hard mask layer pattern 230. A silicon oxide layer may be used as the sacrificial layer 250. Examples of the silicon oxide layer may include a Middle Temperature Oxide (MTO) layer, Undoped Silica Glass (USG), an O3-Tetra Ethyl Ortho Silicate (O3-TEOS) layer, and a High Density Plasma (HDP) layer. The silicon oxide layer may be formed using, for example, a thermal oxidation process or a Chemical Vapor Deposition (CVD) process. In an exemplary embodiment, the sacrificial layer 250 may have a thickness corresponding to about ⅓ of the first pitch P1. With reference to FIG. 8D, a second hard mask layer 260 is formed on the sacrificial layer 250. The second hard mask layer 260 can have a high etching selectivity for the sacrificial layer 250, and may comprise the same material as the first hard mask layer pattern 230. With reference to FIG. 8E, the second hard mask layer 260 and the sacrificial layer 250 are partially planarized to expose the upper surface of the first hard mask layer pattern 230. The planarization process may be performed using a Chemical Mechanical Polishing (CMP) process or an etch-back process. With reference to FIG. 8F, a second photoresist film pattern 320 is formed on the resultant product, and the second hard mask layer 260 exposed by the second photoresist film pattern 320 is removed using, for example, wet etching. In an exemplary embodiment, the second photoresist film pattern 320 is formed using the second mask layer, which corresponds to the third mask data pattern, on which the OPC has been performed, including trimming patterns. Thereafter, the second photoresist film pattern 320 is removed through an ashing process and/or a strip process. With reference to FIG. 8G, a third photoresist film pattern 330 is formed on the resultant product In an exemplary embodiment, the third photoresist film pattern 330 may be formed using the second mask layer. The second mask layer corresponds to the third mask data pattern, on which the OPC has been performed, including connection patterns. With reference to FIG. 8H, the sacrificial layer 250 exposed by the third photoresist film pattern 330 is removed using anisotropic etching, for example, a dry etching process or a reactive ion etching process. Accordingly, the first and second hard mask layers 230 and 260, formed using an anisotropic etching, have the first pitch P1, and the second pitch P2 less than the first pitch P1. With reference to FIG. 8I, the etching target layer 220a exposed by the first and second hard mask layers 230 and 260 is etched using an anisotropic etching. The hard mask layers 230 and 260 can be formed using an anisotropic etching. Thus the etching target layer pattern 220 is completed. The etching target layer pattern 220 has the first pitch P1, and the second pitch P2 less than the first pitch. In accordance with a method of manufacturing a mask according to an exemplary embodiment of the present invention, the mask data pattern for performing an SADP process having high reliability can be formed using the layout-based SADP emulation and the image-based SADP emulation. Although exemplary embodiments have been described with reference to the accompanying drawings, it is to be understood that the present invention is not limited to these precise embodiments but various changes and modifications can be made by one skilled in the art without departing from the spirit and scope of the present invention. All such changes and modifications are intended to be included within the scope of the invention as defined by the appended claims. |
|
055330789 | description | DETAILED DESCRIPTION OF THE INVENTION FIG. 1 represents a pressurized water reactor (PWR) nuclear fuel assembly 10 comprising a lower tie plate 12, guide tubes 14, fuel rods 18 which are spaced radially and supported by spacer grids 16a, 16b, 16c, 16d, 16e, and 16f which are spaced along the guide tubes, instrumentation tube 28, and upper tie plate 37 attached to the upper ends of the guide tubes. Although six spacers are shown for purposes of illustration, other fuel assembly designs can utilize more or less than that shown. Each fuel rod 18 generally includes nuclear fuel pellets 20 composed of fissionable material and an upper end plug 22 and lower end plug 24 which seal the fuel rod. Water as the coolant/moderator is pumped upwardly through the fuel assemblies thereby removing the heat generated by the fuel rods. Control rods which are used to assist in the controlling the fission reaction are disposed in the guide tubes, but are not shown in this view. Several control rods are grouped together and each control rod has a radial arm which interconnect with one another at a central cylindrical member to form a control rod cluster control mechanism for vertically lowering and raising the control rods in the cluster into and out of the guide tubes, and hence into and out of the fuel assembly. Referring to FIG. 2, a pressurized water reactor (PWR) nuclear fuel assembly 40 according to the invention is shown comprising a lower tie plate 42, guide tubes 14 the lower ends of which are connected to the lower tie plate (not shown in this view), extended fuel rods 48 which are spaced radially and supported along the guide tubes by spacer grids 16b, 16c, 16d and 16e, an instrumentation tube 28 (not shown in this view), and upper tie plate 46 which is attached to the upper ends of the guide tubes. Each extended fuel rod 48 includes nuclear fuel pellets 20 composed of fissionable material. Upper end plug 22 (not shown in FIG. 2) seals the upper end of the extended fuel rod. In order to decrease the pressure drop across the length of the fuel assembly and to thereby increase the amount of power which can be generated by the fuel assembly, the lowermost spacer (i.e. 16a of the prior art fuel assembly) is eliminated in the assembly 40 as shown in FIG. 2. However, the lowermost spacer of the prior art fuel assemblies functions not only to maintain rod-to-rod spacing between the fuel rods, but also to resists vibration induced fatigue of the lower end of the fuel rod which would occur if the lower ends of the fuel rods were not restrained against movement caused by coolant moderator which flows up through the fuel assembly. In accordance with the present invention, rather than secure the lower end of the fuel rods to the guide tube by either a spacer, as in the prior art, or any other means which attaches to the guide tube, extended fuel rods 48 extend down to the lower tie plate 42 where they are secured. Although the possibility of vibration induced fatigue of the lower ends of the fuel rods is reduced by extending the fuel rods down into and securing them within the lower tie plate, the possibility of flow induced vibration leading to fretting wear of the lower portion of the fuel rod positioned with the lower tie plate is increased. In accordance with the present invention, the extended fuel rods are secured within apertures in the lower tie plate by the use of a spring which exerts a lateral force on the fuel rod end plug to overcome the vibratory forces induced by the coolant flow thereby preventing lateral motion and possible fuel rod fretting, as well as vibration induced fatigue. Referring to FIG. 3 which is an enlarged partial sectional view of the lower portion of the fuel assembly 40 shown in FIG. 2 showing lower tie plate 42. Each extended fuel rod 48 has at its lower end a fuel rod lower end cap 49 which is positioned in a corresponding aperture 70 in lower tie plate 42. As shown in FIG. 4, which is an enlarged view of one fuel rod positioned within lower tie plate 42, within each aperture 70 is a bore 72 which accommodates spring 74 which exerts lateral forces against the fuel rod end cap 49 to restrain the fuel rod and overcome the vibratory forces induced by the coolant moderator flow thereby preventing lateral motion and possible fuel rod fretting as well as vibration induced fatigue. In order to further reduce the pressure drop across the fuel assembly and thereby obtain further increased power from the fuel assembly, the uppermost spacer (i.e. spacer 16f of the prior art fuel assemblies) is eliminated. However, as in the situation where the lowermost spacer of the prior art fuel assemblies was removed, vibration induced fatigue of the upper portion of the fuel rod can occur if the fuel rods are unrestrained. In accordance with a further aspect of the present invention, and as shown in FIG. 2, upper tie plate 46 extends down over the top of each fuel rod 48. The top of each fuel rod is secured within a fuel rod support housing which has a plurality of springs each of which exerts a lateral force on the top of the fuel rod to overcome the vibratory forces induced by the coolant flow thereby preventing lateral motion and possible fuel rod fretting. Referring to FIG. 5 which is an enlarged perspective view of a portion of upper tie plate 46 shown in FIG. 2 but with fuel rods and guide tubes removed, fuel rod support housing 50 is shown having bores 52 in which the upper ends of the extended fuel rods are positioned. Guide tube cells 60 (only one of which is shown in FIG. 5) receives guide tubes 14 through which the control rods move to increase or decrease the reactivity of the core. FIG. 6 is a perspective view looking up at the upper tie plate 46 and fuel rod support housing 50 showing the upper portions of extended fuel rods 48 positioned within each of their respective support locations. FIG. 8 is a partial sectional view of upper tie plate 46 taken along line 8--8 in FIG. 7 and shows the upper end of each of several fuel rods 48 positioned within fuel rod support housing 50. Fuel rod support housing 50 is adapted to have bores 52 in each of which is positioned a spring 54 which exerts a lateral force against the wall of fuel rod 48 to overcome the vibratory forces induced by the coolant flow thereby preventing lateral motion and possible fuel rod fretting. Coolant flow holes 59 allow coolant/moderator to pass through upper tie plate 46 and exit the top of the fuel assembly. Communicating with bore 52 is chamber 56 having a discharge passageway 58 to allow any coolant moderator which enters the opening for the fuel rods in the fuel rod support housing to discharge at the downstream side of the upper tie plate. As stated above, the fuel assembly of the present invention has several advantages. First, by eliminating the lowermost spacer, the pressure drop across the assembly is reduced and increased power is obtained. Second, by increasing the amount of fuel in each fuel rod by lengthening the active length of the fuel rods down to the lower tie plate, a further increase in power is obtained from the assembly. Third, by securing the lengthened fuel rods in the lower tie plate by the use of lateral restraint, vibration induced fatigue that would have resulted by the elimination of the lowermost spacer if the fuel rods were not restrained) is precluded, and fuel rod fretting resulting from possible lateral movement within the lower tie plate is also precluded. Fourth, by eliminating the uppermost spacer, the pressure drop across the assembly is again reduced and further increases in power is obtained. Fifth, by securing the upper end of the fuel rod in the upper tie plate by the use of lateral restraint, vibration induced fatigue that would have resulted by the elimination of the uppermost spacer (if the fuel rods were not restrained) is precluded, and fuel rod fretting resulting from lateral movement within the upper tie plate is precluded. The advantages of increased power, decreased pressure drop, and elimination of fuel rod fretting to the lower and upper ends of the fuel rods, all of which is achieved without changing either the length of the fuel assembly, or the fuel rod diameter, or fuel rod pitch, make the present invention particularly useful for all pressurized water reactors. While the foregoing description and drawings represent the preferred embodiments of the present invention, it will be apparent to those skilled in the art that various changes and modifications may be made therein without departing from the true spirit and scope of the present invention. |
summary | ||
046630864 | summary | BACKGROUND OF THE INVENTION The present invention relates to a conditioning process involving the coating in bitumen of radioactive waste constituted by ion exchange resins. The bitumen coating of radioactive waste is generally performed by mixing an aqueous suspension of the radioactive waste with fluidized bitumen, followed by the evaporation of the water of the suspension and pouring the thus obtained mixture into a container in order to solidify it. French Pat. No. 1,315,162, filed on Dec. 6, 1961 by the Commissariat a l'Energie Atomique, describes a bituminizing process of this type, according to which the suspension of the waste is mixed with fluidized bitumen in the presence of a surfactant, which facilitates the separation of the water contained in the separation. Most of this water is then separated either by decanting, or by means of a mechanical device and the product obtained after separating the water is then mixed at a temperature to ensure that it is sufficiently fluid and is finally poured or cast in such a way that cooling leads to solid blocks having an appropriate plasticity. French Pat. No. 2,052,093 filed on July 15, 1969 by the Commissariat a l'Energie Atomique described an installation for coating a suspension of radioactive products with bitumen using a film evaporator and a gear pump for extracting the coated products obtained at the bottom of the evaporator. French Pat. No. 2,356,246 describes an improvement to the processes for coating radioactive waste with bitumen consisting of subjecting the waste to a pretreatment by salts, such as calcium and barium chlorides, in order to improve the resistance to leaching of the solidified products obtained. In this case, the treated waste is constituted by evaporation concentrates and/or precipitation sediment comprising salts, such as sodium sulphates and carbonates. Installations for the continuous coating of radioactive waste suspensions are also known, which use a twin or four screw extruder, in which the waste is mixed with bitumen and the suspension is dried. Generally, before coating the suspension of the waste, the latter undergoes a neutralization treatment using soda. Thus, it is inadvisable to bitumen coat acid suspension, particularly free nitric acid, in order to obviate any risk of deterioration of the bitumen at the end of the operation when, after evaporating the water, the bitumen is in the presence of concentrated acid. Moreover, neutralization before bitumen coating limits corrosion to the installations. When the radioactive waste is formed by organic ion exchange resins, particularly of the anionic type in form OH.sup.- and/or Cl.sup.-, the performance of the bitumen coating processes does not make it possible to obtain an adequate treatment capacity of the bituminizing installation. Moreover, the bituminous coatings obtained suffer from the major disadvantages of swelling when subsequently immersed in water. In this case, the volume increase of the coatings can reach 20% and even exceed 100% in certain cases, which leads to a disintegration of the coated waste. Processes for conditioning ion exchange materials in thermosetting resins or in cement are also known of the type described in French Pat. No. 2,361,724 and Japanese Pat. No. 48-28899. According to these processes, in order to obtain crack-free conditioned products, it is necessary to pretreat the ion exchange resins to replace the H.sup.+ ions by other cations, in order to ensure that the resins do not fix certain of the reagents necessary for obtaining the setting of the thermosetting resin or the cement. German Pat. No. 3,102,473 also describes a process for pretreating a mixture of cation and anion exchange resins, which then makes it possible to separately condition the resins by incorporation in bitumen or cement. According to this patent, in order to separate the cation exchange resins from the anion exchange resins, the resin mixture is contacted with an aqueous solution of a salt, such as an alkali metal acetate, nitrate, chloride or sulphate, in order to replace the H.sup.+ and/or Na.sup.+ ions of the cationic resins by sodium and the OH.sup.- ions of the anionic resins by other anions. The resins can then be separated by means of a liquid, whose density is between that of the grains of one type of resin and that of the grains of the other type of resin. Thus, this patent does not deal with the problem of swelling in water on the part of the bituminous coatings. SUMMARY OF THE INVENTION The present invention relates to a process for conditioning with bitumen radioactive waste constituted by ion exchange resins, which comprises a pretreatment stage making it possible to solve the problems referred to hereinbefore. The present invention specifically relates to a process for conditioning by bituminizing radioactive waste constituted by cation and/or anion exchange resins wherein it comprises: (a) subjecting said resin or resins to a pretreatment using a salt for replacing the H.sup.+ and/or Na.sup.+ ions of the cation exchange resins by ions chosen from the group including Ca.sup.++, Sr.sup.++ and Ba.sup.++ and/or for replacing the OH.sup.- and/or Cl.sup.- ions of the anion exchange resins by an anion chosen from the group including NO.sub.3.sup.-, HCO.sub.2.sup.- and CH.sub.3 CO.sub.2.sup.-, (b) suspending the thus pretreated resin or resins in water, and (c) bituminizing said suspension. According to a feature of the invention, pretreatment is carried out by contacting, preferably accompanied by stirring, the said resins with an aqueous solution of a calcium, strontium or barium salt, said salt being chosen from the group including nitrates, acetates and formates. This way of performing the pretreatment is particularly suitable for the treatment of a mixture of cation and anion exchange resins, because it permits the simultaneous replacement of the OH.sup.- or Cl.sup.- ions of the anionic resins by nitrate, acetate or formate anions and also the replacement of the H.sup.+ ions of the cationic resins by calcium, strontium and barium. In this case, use is generally made of an aqueous solution of barium acetate or nitrate for carrying out the pretreatment. Thus, the use of barium nitrate is preferable, because the NO.sub.3.sup.- anion is generally present in effluence, particularly in the effluence of reprocessing centres and nuclear research centres. As a result of the pretreatment according to the invention, a modification takes place to the structure of the three-dimensional network of the ion exchange resins by introducing into the molecular chain ions such as Ba.sup.++ and/or NO.sub.3.sup.- or CH.sub.3 COO.sup.-, which occupy more space than the previously present H.sup.+, OH.sup.- and Cl.sup.- ions. This makes it possible to prevent the penetration of water into the denser, macromolecular network of the pretreated resin, thereby reducing the degree of hydration of the resin. Thus, ion, anion and cation exchange resins have the special feature of significantly inflating in water. Their volume increase is generally approximately 20% for anion resins in the form OH.sup.- and/or Cl.sup.- and exceeds 50% for cation resins in the form Na.sup.+. This swelling is due to the penetration of water into the three-dimensional skeleton of the resins and in particular hydrophilic in the case of anionic resins having quaternary or amino ammonium groups and in the case of cationic resins having sodium sulphonate groups, grafted on a polystyrene radical for both resin types. Therefore, the treatment capacity of the bituminizing installations is lower on treating resins of these types. Moreover, the coatings obtained at the end of the bituminizing treatment suffer from the disadvantage of swelling in water. However, when, in accordance with the invention, the ion exchange resins are pretreated, the hydration capacity of the macromolecular network of the resins is reduced, so that it is possible to increase the treatment capacity of the bituminizing installations and limit the swelling of the coatings obtained in contact with water. Thus, according to the process of the invention, the stage of pretreating ion exchange resins with a salt, such as barium nitrate, does not serve the same object as the pretreatment stage in the prior art processes of French Pat. No. 2,361,724 and Japanese Pat. No. 48-28899, because it is not intended to prevent the ion exchange resins from consuming certain of the reagents necessary for the coating matrix formation reaction. In the same way, the pretreatment stage according to the invention does not have the same objective as the pretreatment stage of French Pat. No. 2,356,246. Thus, in this patent, the waste is constituted by evaporation concentrates and/or sediments of radioactive materials such as chemical precipitation sediments containing salt and the problem to be solved is that of converting the salts such as sodium carbonate and sulphate into salts having a reduced tendency to trap molecules of water. In the present invention, the waste is constituted by ion exchange resins, which do not contain salts and the problem to be solved in that of reducing the hydration capacity of the macromolecular network of the ion exchange resins. For the bituminizing operation, the conventional procedure is adopted using e.g. the processes described in French Pat. Nos. 1,315,162 and 2,052,093, or U.S. Pat. No. 3,298,961. The means used are also of a conventional nature and can comprise film evaporators, twin or four screw extruders, etc. The process according to the invention is more particularly applied to the treatment of cationic and/or anionic resins with a polystyrene skeleton in ball and/or ground form. Examples of such resins are those in the form of balls, such as the resin sold under the trade mark DUOLITE by DIAPROSIM or resins marketed under the trade mark AMBERLITE by ROHM and HAAS. Examples of ground resins are those sold under the MICROIONEX mark by DIAPROSIM. When, according to the invention, mixed cationic and anionic resins are treated, it is possible to use any random proportion of anionic resins in the range 0 to 100%. For performing the process according to the invention, a pretreatment stage is preferably performed by introducing into a container a certain quantity of an aqueous solution of a salt of Ba.sup.++, Ca.sup.++ or Sr.sup.2+ containing NO.sub.3.sup.-, HCO.sub.2.sup.- or CH.sub.3 CO.sub.2.sup.- ions. The resins to be pretreated are suspended in the solution and stirring takes place for an adequate time, and, as a function of the salt concentration of the solution, this is chosen so as to obtain the desired saturation level of the ion exchange resins. Preferably, the salt concentration and the treatment time are chosen so as to obtain a saturation level of the ions NO.sub.3.sup.-, HCO.sub.2.sup.- or CH.sub.3 CO.sub.2.sup.- of the anion exchange resins close to 100%. As a result of the pretreatment of the ion exchange resins according to the process of the invention, the bimunizing process can be improved by: (1) increasing the treatment capacity of the bituminizing installation by 50%, because the volume of the resin suspended in the wafer is reduced and the evaporation capacity of the bituminizing device is increased; (2) eliminating from the distillates the products which deteriorate anionic resins, particularly NH.sub.3 and CH.sub.3 NH.sub.2, as a result of a chemical stabilization of the tertiary amine or quaternary ammonium functional groups, which makes it possible to recover neutral condensates instead of basic condensates; and (3) limiting the swelling process of the bituminous coatings in water, their volume increase not exceeding 5%, as will be shown hereinafter. |
abstract | In a boiling water reactor controlling each control rod hydraulically by driving a solenoid valve, the control system comprises an operation control means 41 having a duplicated data processing unit for generating sequence patterns based on the timing of the driving sequence based on the control information provided manually, a transmission control means 42 for creating a command word corresponding to each control rod being controlled based on said sequence pattern, and mutually communicating each command word between duplicated data processing units and computing the AND logic of said data within a predetermined time difference, and when the computed result coincide, transmitting the selected command word serially; a transmission unit 32 for receiving said command word and performing protocol conversion thereto, and transmitting the same to a plurality of transmission branch units positioned downstream as the control command; and a solenoid valve drive circuit 31 for driving the control rod drive unit corresponding to each rod branched from said transmission branch units. |
|
description | This application is based upon and claims the benefit of priority from the prior Japanese Patent Application No. 2005-268214, filed on Sep. 15, 2005, the entire contents of which are incorporated herein by reference. This invention relates to an apparatus for measuring the temperature of the coolant flowing in the reactor core of a nuclear reactor, a method for measuring the temperature of the coolant and an apparatus for monitoring a nuclear reactor. In the boiling water reactor, a neutron detector is inserted into the reactor core from the bottom. To calibrate the sensitivity of the neutron detector, it has been proposed to combine a γ-ray thermometer and a neutron detector to observe the thermal power distribution of the nuclear reactor of the boiling water reactor (see Japanese Patent Publication No. 3556409 (the entire contents of which are incorporated herein by reference)). A γ-ray thermometer is designed to measure the temperature difference within a detector due to the exothermic phenomena in the structural member attributable to γ-rays by a differential thermocouple. Since the amount of heat generation due to γ-rays is proportional to the thermal power of the nuclear reactor, it is possible to observe the axial thermal power distribution of the nuclear reactor by installing and arranging a plurality of detectors of γ-ray thermometers in the axial direction. Similarly, it is possible to observe the horizontal thermal power distribution of the nuclear reactor by installing and arranging a plurality of γ-ray thermometers at various horizontal positions. A normal γ-ray thermometer contains a calibration heater so as to calibrate a differential thermocouple of a γ-ray thermometer by providing electricity to the calibration heater. By one of the methods for calibrating sensitivity with a calibration heater, the sensitivity is determined from the amount of emitted heat generated by the calibration heater and the change in the output signal of the differential thermocouple. Meanwhile, although the temperature of the coolant flowing in the reactor core of a boiling water reactor is an important parameter for computationally evaluating the performance of the reactor core, the temperature indirectly observed at the coolant piping located outside the reactor is used for computationally evaluating the performance of the reactor core conventionally. Conventional boiling water reactors do not have any means for directly observing the temperature of the coolant flowing in the reactor core that is an important parameter for computationally evaluating the performance of the reactor core. This lack of direct observation means is an error factor for computationally evaluating the performance of the reactor core. However, to install an in-core thermometer in order to observe the temperature of the reactor core coolant, it is necessary to modify the design of the existing reactor vessel. So, the design modification of the existing reactor vessel causes large increase of cost and is not a realistic choice. Therefore, it is an object of the present invention to make it possible to directly observe the temperature of the coolant flowing in the reactor core in the nuclear reactor and monitor the nuclear reactor without modifying the design of the reactor vessel. The present invention has been made to solve the above problems, and has an object of making it possible to directly observe the temperature of the coolant flowing in the reactor core in the nuclear reactor and monitor the nuclear reactor without modifying the design of the reactor vessel. According to an aspect of the present invention, there is provided an apparatus for measuring a temperature of coolant flowing in a reactor core contained in a reactor vessel, the apparatus comprising: a γ-ray thermometer having a γ-ray heat emission detector installed in the reactor core and a thermocouple thermometer installed out of the reactor core in the reactor vessel; and an output signal processor receiving a signal output from the thermocouple thermometer and computationally determining a local temperature of the coolant at the thermocouple thermometer. According to another aspect of the present invention, there is provided an apparatus for measuring a temperature of coolant flowing in a reactor core contained in a reactor vessel, the apparatus comprising: a first temperature measuring apparatus having a delay in detection of a temperature change of the coolant less than or equal to an allowable delay criteria; a second temperature measuring apparatus having an error in measurement of a temperature of the coolant less than or equal to an allowable error criteria; and a signal processor regarding a first temperature computationally determined from a signal output from the first temperature measuring apparatus as the temperature of the coolant if a change per unit time of the first temperature exceeds the allowable delay criteria, and regarding a second temperature computationally determined from the signal output from the second temperature measuring apparatus as the temperature of the coolant if a change per unit time of the first temperature is less than or equal to the allowable delay criteria. According to yet another aspect of the present invention, there is provided a method for measuring a temperature of coolant flowing in a reactor core contained in a reactor vessel, the method comprising: measuring a temperature at a position inside a γ-ray thermometer but outside the reactor core in the reactor vessel, the γ-ray thermometer having a γ-ray heat generation detector installed in the reactor core; and determining computationally a local temperature of the coolant from the temperature measured in the step of measuring. According to yet another aspect of the present invention, there is provided a method for measuring a temperature of coolant flowing in a reactor core contained in a reactor vessel, the method comprising: regarding a first temperature computationally determined from a signal output from a temperature measuring apparatus having a delay in detection of a temperature change of the coolant less than or equal to an allowable delay criteria as the temperature of the coolant if a change per unit time of the first temperature exceeds the allowable delay criteria, and regarding a second temperature computationally determined from the signal output from a temperature measuring apparatus having an error in measurement of a temperature of the coolant less than or equal to an allowable error criteria as the temperature of the coolant if a change per unit time of the first temperature is less than or equal to the allowable delay criteria. According to yet another aspect of the present invention, there is provided a Apparatus for monitoring a reactor core contained in a reactor vessel, the apparatus comprising: γ-ray thermometers categorized into groups according to their positions, the γ-ray thermometer each having a γ-ray heat generation detector installed inside the reactor core and a thermocouple thermometer installed outside the reactor core; an output signal processor receiving a signal output from the thermocouple thermometers and determining computationally a local temperatures of coolant at the positions of the thermocouple thermometers and an average temperature of each of the groups; and a diagnosing device diagnosing the nuclear reactor as in an abnormal condition if at least one of the average temperature, the change per unit time of any of the local temperatures and the deviation of any of the local temperatures from the average temperature exceeds a corresponding criteria. Now, the present invention will be described in greater detail by referring to the accompanying drawings that illustrate preferred embodiments of the invention. FIG. 1 is a schematic illustration of an embodiment of coolant temperature measuring apparatus according to the present invention. Inside a reactor vessel 1, there is a reactor core 2 where a plurality of fuel assemblies are arranged cylindrical space extending vertically. The reactor core 2 is supported by a core support plate 34. A neutron detector assembly 3 is inserted into the reactor core 2 from a bottom of the reactor vessel 1. The neutron detector assembly 3 contains inside a neutron detector 5 and a γ-ray thermometer 4. Note that, while only a single neutron detector assembly 3 is shown in FIG. 1, a plurality of neutron detector assemblies 3 are installed and arranged at various positions in the reactor core. The γ-ray thermometer 4 includes a γ-ray heat generation detecting section 6 located in the reactor core 2 to measure a temperature difference due to the exothermic phenomenon attributable to γ-rays and a temperature measuring section 7 located lower than the bottom end of the reactor core 33 in the reactor vessel 1 where no exothermic phenomenon attributable to γ-rays takes place to measure the temperature of the coolant. The signals output from the neutron detector 5 and the γ-ray thermometer 4 are input to an output signal processing apparatus 10 through a multicore connector 8 and a multicore cable 9. FIG. 2 is an enlarged schematic longitudinal cross sectional view of the γ-ray thermometer 4 of the embodiment according to the present invention. The γ-ray thermometer 4 includes a cylindrical structural member 12 in the inside of the detector container 30, a differential thermocouple 13 having a contact point located in the reactor core 2 and a thermocouple thermometer 16 having a junction located lower than a calibration heater 18 and the bottom end of the reactor core 33. The differential thermocouple 13 has a hot junction 14 and a cold junction 15. The structural member 12 is provided with a heat insulator 11 located at the same vertical position as the hot junction 14 of the differential thermocouple 13 and the thermocouple thermometer 16. The calibration heater 18 is located substantially at the center of the horizontal cross section of the detector container 30. The heat generator of the calibration heater 18 extends from a lower position than the thermocouple thermometer 16 to a higher position than the hot junction 14 and the cold junction 15 of the differential thermocouple 13. FIG. 3 is a schematic illustration of an embodiment of nuclear reactor monitoring apparatus according to the present invention. The nuclear reactor monitoring apparatus includes a reactor condition evaluator 21, an output signal processing apparatus 10, a γ-ray thermometer 4, a device for measuring a core support plate differential pressure 31 and a coolant flow rate and temperature evaluator 32. The γ-ray thermometer 4 is connected to the output signal processing apparatus 10 and the output signal processing apparatus 10 is connected to the reactor condition evaluator 21. The device for measuring the core support plate 31 differential pressure is connected to the coolant flow rate and temperature evaluator 32 and the coolant flow rate and temperature evaluator 32 is connected to the reactor condition evaluator 21. Now, the method of measuring the coolant temperature by means of the embodiment of coolant temperature measuring apparatus and the method of monitoring the nuclear reactor by means of the embodiment of nuclear reactor condition monitoring apparatus will be described below. Firstly, the differential thermocouple 13 and the thermocouple thermometer 16 are calibrated by using the calibration heater 18. As electricity is provided to the calibration heater 18, the calibration heater 18 generates heat and raises an internal temperature of the γ-ray thermometer 4. Heat generated by the calibration heater 18 hardly diffuses to the coolant in the vicinity of the heat insulator 11, whereas heat generated by the calibration heater 18 diffuses more to the coolant in areas where the heat insulator 11 is not located. Therefore, the temperature of the hot junction 14 of the differential thermocouple 13 becomes higher relative to the temperature of the cold junction 15 of the differential thermocouple 13 so that it is possible to determine the sensitivity of the differential thermocouple 13 from the amount of heat generated by the calibration heater 18 and the change in the output signal of the differential thermocouple 13. The thermocouple thermometer 16 can be calibrated by using the calibration heater 18 like the calibration of the differential thermocouple 13. It is possible to determine the sensitivity of the thermocouple thermometer 16 from the amount of heat generated by the calibration heater 18 and the change in the output signal of the thermocouple thermometer 16. The heat insulator 11 is also located between the thermocouple thermometer 16 and coolant and heat generated by the calibration heater 18 hardly flows to the coolant so that the thermocouple thermometer 16 has a structure hardly influenced by the coolant during the calibration. Note, however, that it is possible to calibrate the thermocouple thermometer 16 even if there is no heat insulator 11 between the thermocouple thermometer 16 and coolant and hence the heat insulator 11 between the thermocouple thermometer 16 and coolant may be eliminated. If the heat insulator 11 between the thermocouple thermometer 16 and coolant is eliminated, the responsiveness to the temperature change of the coolant is improved, although the accuracy of calibration of the thermocouple thermometer 16 is degraded. The differential thermocouple 13 and the thermocouple thermometer 16 may be calibrated simultaneously or independently. During an operation of the nuclear reactor, the part of the structural member 12 in the detector container 30 that is located in the inside of the reactor core 2 generates heat due to γ-rays. Since the heat insulator 11 is located in the vicinity of the hot junction 14, heat generated in the structural member 12 due to γ-rays hardly flows to the coolant and hence the temperature of the vicinity of the hot junction 14 is higher than the temperature of the vicinity of the cold junction 15 where no heat insulator 11 is provided. Since the amount of heat generated in the structural member due to γ-rays is proportional to the thermal power of the nuclear reactor, it is possible to observe the axial distribution of the thermal power of the nuclear reactor by measuring the temperature difference by means of the plurality of differential thermocouples 13 located at various axial positions. In addition, it is possible to observe the horizontal distribution of the thermal power of the nuclear reactor by arranging a plurality of differential thermocouples 13, ie. neutron detector assemblies 3, at a various horizontal direction. The temperature of the vicinity of the thermocouple thermometer 16 is same as that of the coolant because it is not influenced by γ-rays. Therefore, it can be assumed that the temperature computed by the output signal processing apparatus 10 by using the signal from the thermocouple thermometer 16 is equal to the local coolant temperature at the reactor core inlet in the vicinity of the thermocouple thermometer 16. However, it should be noted that, when the coolant temperature changes, there is a time lag until the temperature in the vicinity of the thermocouple thermometer 16 becomes equal to the changed temperature because heat flows through the detector container 30 and the structural member 12. It is also possible to determine the coolant temperature at the reactor core inlet based on the thermal balance of the nuclear reactor by using the flow rate of the reactor core coolant determined from the differential pressure across the core support plate 34 that supports the reactor core. Then, the time lag will be smaller than that of the measurement of the thermocouple thermometer 16 but the error will be larger than that of the measurement of the coolant temperature by means of the thermocouple thermometer 16. Thus, the reactor condition evaluator 21 evaluates the coolant temperature at the reactor core inlet by combining the two methods of measuring the coolant temperature by means of the thermocouple temperature 16. One of these methods has a high accuracy but the responsiveness of it to a temperature change is not very good. The other is a method of measuring the coolant temperature from the differential pressure across the core support plate 34, of which the accuracy is not very high but the responsiveness is good. More specifically, the reactor condition evaluator 21 has a function to use the coolant temperature of the reactor core inlet determined from the core support plate differential pressure if the change per unit time of the coolant temperature at the reactor core inlet determined from the core support plate differential pressure exceeds a predetermined criteria, or the coolant temperature at the reactor core inlet determined from the thermocouple thermometer 16 if the change per unit time of the coolant temperature at the reactor core inlet determined from the core support plate differential pressure doesn't exceed the criteria. This function enables to measure with a relatively large error but a high responsiveness if the coolant temperature at the reactor core inlet changes quickly and to the measure with a small error if the coolant temperature at the reactor core inlet changes mildly. While the method of determining the coolant temperature at the reactor core inlet from the core support plate differential pressure is employed as a highly responsive method of measuring the coolant temperature and the method of measuring the coolant temperature by means of a thermocouple is employed as a highly accurate method of measuring the coolant temperature at the reactor core inlet in this embodiment, other measuring methods may alternatively be employed for the purpose of the present invention. For example, it is possible to increase the responsiveness of measuring the temperature of the coolant by contacting the thermocouple thermometer 16 with the container of the neutron detector assembly 3 and reducing the thickness of the container around the contact position. With this arrangement, the coolant temperature measured by means of the calibrated thermocouple thermometer 16 may have a large error because the influence of the coolant temperature is large during calibration of the thermocouple thermometer 16 but the change in the coolant temperature is transmitted to the thermocouple thermometer 16 more quickly. The reactor condition evaluator 21 also has a function of diagnosing the soundness of a thermocouple thermometer 16 from the output of a plurality of thermocouple thermometers 16. The reactor condition evaluator 21 diagnoses the thermocouple thermometer 16 as out of order, if the deviation of the local coolant temperature at the reactor core inlet determined from the output of a thermocouple thermometer 16 exceeds a predetermined criteria from the average value of the coolant temperatures at the reactor core inlet determined from the outputs of the plurality of thermocouple thermometers 16. The average of the coolant temperature at the reactor core inlet can be determined by averaging the coolant temperatures determined from the outputs of all the thermocouple thermometers 16, or by averaging the coolant temperatures in each region by dividing the thermocouple thermometers 16 into groups according to their positions. It is possible to appropriately diagnose the soundness of the thermocouple thermometer 16 by averaging the coolant temperatures of each of the regions even if the coolant temperatures of the regions are different from each other. When the nuclear reactor is operated in the normal condition, the coolant temperature at the reactor core inlet is not lower than the supplied water temperature and not higher than the saturated vapor temperature as determined from the pressure of the nuclear reactor. Therefore, a thermocouple thermometer 16 may be diagnosed as out of order if the difference between the local coolant temperature at the nuclear reactor inlet determined from the output of the thermocouple thermometer 16 and the above described average coolant temperature is greater than the difference of the saturated vapor temperature and the supplied water temperature. A thermocouple thermometer 16 may also be diagnosed as out of order if the coolant temperature determined from the output of the thermocouple thermometer 16 is lower than the supplied water temperature minus the measurement error of the thermocouple thermometer 16, or higher than the saturated vapor temperature plus the measurement error of the thermocouple thermometer 16. Generally, the reactor core is arranged symmetrically. For example, fuel having the same irradiation history may be arranged in every ¼ of the reactor core with the center disposed at the vertical axis of the reactor core so that the same fuel arrangement appears if the reactor core is rotated by 90° around the vertical axis. In addition, the coolant flowing in the reactor core includes the coolant sent to the steam turbine, subsequently condensed by the condenser, heated and then supplied to the reactor core, and the coolant without being sent to the steam turbine but re-circulated in the reactor vessel 1 and flown into the reactor core. Since their temperatures differ from each other to a certain extent, the temperature of the coolant that flows into the reactor core can be dispersed. The dispersion depends on the flow of coolant and hence typically on the arrangement of reactor internals. The flow of coolant can also be regarded as symmetrical. Therefore, if one of the thermocouple thermometers 16 goes out of order, the reactor condition evaluator 21 has a function to regard the local coolant temperature at the reactor core inlet determined from the output of the thermocouple thermometer 16 located at a position equivalent to the position of the thermocouple thermometer 16 that is out of order from the viewpoint of the above described symmetry (symmetrical position) as the local coolant temperature at the reactor core inlet of the thermocouple thermometer 16 that is out of order. Additionally, the reactor condition evaluator 21 has a function of monitoring the nuclear reactor. It can detect that the coolant supplied to the reactor vessel is not heated, ie. a loss of feed water heating. If a loss of feed water heating occurs, the temperature of supplied water decreases and the average temperature of the coolant flowing into the reactor core decreases. So, the reactor condition evaluator 21 diagnoses that a loss of feed water heating occurs if the average temperature of the coolant at the reactor core inlet determined from the outputs of the thermocouple thermometers 16 of any of the above described groups, all the thermocouple thermometers 16 falls below a predetermined criteria, or if the amount of the decrease per unit time of the average of the coolant temperature at the reactor core inlet exceeds a predetermined criteria. Furthermore, if a loss of feed water heating occurs, the difference between the temperature of the coolant sent to the steam turbine and subsequently supplied to the reactor vessel and the temperature of the coolant flowing into the reactor core without being sent to the steam turbine increases. Therefore, the reactor condition evaluator 21 so diagnoses that a loss of feed water heating occurs if the dispersion of the local coolant temperature at the reactor core inlet determined from the outputs of a plurality of thermocouple thermometers 16, for example, the maximum value of the deviation of the local coolant temperature at the reactor core inlet from the average of the coolant temperature at the reactor core inlet, exceeds a predetermined criteria. As described above, this embodiment of the present invention can measure the temperature of the reactor core coolant without modifying the design of any existing reactor vessel 1. Additionally, the output of the γ-ray heat generation detecting section 6 and that of the temperature measuring section 7 can be transmitted through one multicore connector 8 and one multicore cable 9 and can be sent to one output signal processing apparatus 10. Thus, it is possible to reduce the cost of hardware and the space for arranging hardware. Additionally, both the sensitivity of the γ-ray heat generation detecting section 6 and that of the temperature measuring section 7 can be calibrated by a single calibration heater 18. Thus, it is possible to reduce the cost of hardware and the number of measuring operations of operators. It is possible to measure the temperature of the coolant with an acceptable degree of both accuracy and responsiveness by using a measuring method of high responsiveness and low accuracy and a measuring method of low responsiveness and high accuracy so that the measuring method of high accuracy may be used in ordinary operating conditions but the measuring of high responsiveness may be used during the change of the operating conditions. It will be appreciated that the above described embodiments are only examples and the present invention can be embodied in various different ways. While the above embodiments are described in terms of a boiling water reactor, they can be applied to a nuclear reactor of any other type. Additionally, the thermocouple thermometers 16 may be replaced by resistance thermometers. |
|
abstract | The present application relates to an X-ray device having an X-ray radiation source (7), a collimator (6) and an X-ray detector (4). The reduction in the radiation dose to which a patient (5) is exposed, said reduction being generated by the collimator (6), is calculated by a data processing unit (2) from the current setting of the shield and filter elements of the collimator (6) and displayed on a display unit (3), for example, as an area shielding factor AA and/or a dose reduction factor DRF. The user can in this way detect the potential of the collimator that still exists for image improvement and for minimizing the exposure to radiation. |
|
046577238 | claims | 1. In a tokamak reactor having a toroidal field coil assembly enclosing a generally toroidal shaped region having a toroidal axis, said toroidal field coil assembly comprising a plurality of generally flat-washer shaped magnetic coils having generally flat opposing radial faces with respect to said toroidal axis, a cooling arrangement for distributing coolant in said toroidal field coil assembly, said cooling arrangement comprising, for each coil: (a) a plurality of coolant channels for flowing coolant in the same direction through an interior of said coil, said coolant channels embedded within said coil and having inlets and outlets opening on a face of said coil, (b) said inlet and oulet openings disposed adjacent opposite ends of said coil and arranged radially along said coil face with respect to the toroidal axis, (c) a single inlet header formed in part by said coil face containing said inlet openings and in part by a curved supply conduit having a longitudinal opening along the length thereof, (d) a single outlet header formed in part by said coil face containing said outlet openings and in part by a curved return conduit having a longitudinal opening along the length thereof, and (e) said supply and return conduits each having longitudinal edges along their longitudinal openings, said edges being secured to respective portions of said coil face containing said inlet and outlet openings, such as by welding or brazing. (a) a plurality of coolant channels for flowing coolant in the same direction through an interior of said coils, said coolant channels embedded within said coil and having inlets and outlets opening on the faces of said coils, (b) said inlet and outlet openings disposed adjacent said leading and trailing ends of said coils and arranged radially along the coil faces with respect to the toroidal axis, (c) inlet openings of any given pair of coils disposed adjacent to one another along opposite sides of the joint joining said pair, (d) outlet openings of any given pair of coils disposed adjacent to one another along opposite sides of the joint joining said pair, (e) a single inlet header for each coil pair, said inlet header formed in part by said coil faces containing said inlet openings of each pair and the joint connecting said pair, and in part by a curved supply conduit having a longitudinal opening along the length thereof, said curved supply conduit covering the inlet openings and connecting joint of each pair, (f) a single outlet header for each coil pair, said outlet header formed in part by said coil faces containing said outlet openings of each pair and the joint connecting said pair and in part by a curved return conduit having a longitudinal opening along the length thereof said curved return conduit covering the oulet openings and connecting joint of each pair, (g) said supply and return conduits each having longitudinal edges along their longitudinal openings, said edges being secured to respective portions of said coil faces containing said inlet and outlet openings such as by welding or brazing. (a) a first plurality of coolant channels for flowing coolant in a first direction through the interior of said coil, (b) a second plurality of coolant channels for flowing coolant in a second direction through the interior of said coil, said second direction opposite said first direction, (c) said first plurality of coolant channels having first openings on one face of said coil adjacent an end thereof, and second openings on the opposite face adjacent the other end thereof, (d) said second plurality of coolant channels having third openings on said opposite face of said coil adjacent the other end thereof, and fourth openings on said one face adjacent said opposite end thereof, (e) a first inlet header formed in part by said coil face containing said first openings and in part by a first curved supply conduit having a longitudinal opening along the length thereof, (f) a first outlet header formed in part by said coil face containing said second openings and in part by a first curved return conduit having a longitudinal opening along the length thereof, (g) a second inlet header having formed a header formed in part by said coil face containing said third openings and in part by a second curved supply conduit having a longitudinal opening along the length thereof, (h) a second outlet header formed in part by said coil face containing said fourth openings and in part by a second curved return conduit having a longitudinal opening along the length thereof, (i) said first and second supply conduits and said first and second return conduits each having longitudinal edges along their longitudinal openings, said edges being secured to respective portions of said coil faces containing said inlet and outlet openings, such as by welding or brazing. (a) a first plurality of coolant channels for flowing coolant in a first direction through an interior of said coils, (b) a second plurality of coolant channels for flowing coolant in a second, opposite direction through the interior of said coils, (c) each of said first and second plurality of coolant channels having inlet and outlet openings on the faces of said coils, (d) said inlet and outlet openings disposed adjacent said leading and trailing ends of said coils and arranged radially along the coil faces with respect to the toroidal axis, (e) inlet openings of any given pair of coils disposed adjacent to one another along opposite sides of the joint joining said pair, (f) outlet openings of any given pair of coils disposed adjacent to one another along opposite sides of the joint joining said pair, (g) a single inlet header for each coil pair, said inlet header formed in part by said coil faces containing said inlet openings of each pair and the joint connecting said pair, and in part by a curved supply conduit having a longitudinal opening along the length thereof, said curved supply conduit covering the inlet openings and connecting joint of each pair. (h) a single oulet header for each coil pair, said outlet header formed in part by said coil faces containing said oulet openings of each pair and the joint connecting said pair and in part by a curved return conduit having a longitudinal opening along the length thereof said curved return conduit covering the outlet openings and connecting joint of each pair, (i) said supply and return conduits each having longitudinal edges along their longitudinal openings, said edges being secured to respective portions of said coil faces containing said inlet and outlet openings such as by welding or brazing. 2. The cooling arrangement as recited in claim 1, wherein such supply header is tapered to decrease the cross-sectional flow area thereof with decreasing radial distance from said tokamak axis. 3. The cooling arrangement of claim 1, wherein said supply header is contoured to control the distribution of said coolant to said coolant inlets. 4. In a tokamak reactor having a toroidal field coil assembly enclosing a generally toroidal shaped region having a toroidal axis, said toroidal field coil assembly comprising a plurality of generally flat-washer shaped magnetic coils having generally flat opposing radial faces with respect to said toroidal axis, each coil having a leading and trailing end, the leading end of each coil connected to the trailing end of the immediately adjacent coil at a joint, a cooling arrangement for distributing coolant in said toroidal field coil assembly, said cooling arrangement comprising: 5. In a tokamak reactor having a toroidal field coil assembly enclosing a generally toroidal shaped region having a toroidal axis, said toroidal field coil assembly comprising a plurality of generally flat-washer shaped magnetic coils having generally flat opposing radial faces with respect to said toroidal axis, a cooling arrangement for distributing coolant in said toroidal field coil assembly, said cooling arrangement comprising: 6. The coil arrangement as recited in claim 5, wherein said first and second plurality of coolant channels are interlieved with one another such that adjacent channels in said coils carry fluid flowing in opposite directions. 7. In a tokamak reactor having a toroidal field coil assembly enclosing a generally toroidal shaped region having a toroidal axis, said toroidal field coil assembly comprising a plurality of generally flat-washer shaped magnetic coils having generally flat opposing radial faces with respect to said toroidal axis, each coil having a leading and trailing end, the end of each coil connected to the trailing end of the immediately adjacent coil at a joint, a cooling arrangement for distributing coolant in said toroidal field coil assembly, said cooling arrangement comprising: 8. A cooling arrangement as recited in claim 7, wherein said first and second plurality of coolant channels are interlieved with one another such that adjacent channels in said coils carry fluid flowing in opposite directions. |
description | 1. Field of the Invention The present invention relates generally to an image information detecting apparatus with an image information detecting unit that receives recording light representing image information, and records the image information by storing electric charges generated therein by the recording light. More specifically, the present invention relates to an image information detecting apparatus with a wireless transmission capability. 2. Description of the Related Art Today, in X-ray imaging for medical diagnosis or the like, various image information detecting devices with an image information detecting unit that receives recording light (X-ray, or the like) representing image information, and records the image information by storing electric charges generated therein by the recording light are proposed and put into practical use as described, for example, in U.S. Pat. No. 6,268,614. Most of these image information detecting devices include a charge readout means for reading out the electric charges stored therein as electrical signals. From the aspect of charge readout process, the optical-scan readout method in which the electric charges are read out by scanning the detecting unit with light beams, and electrical-scan readout method in which the electric charges are read out by electrically scan-driving TFTs (thin film transistors) or the like incorporated in the detecting unit, are widely known. Recently, development of cassette type image information detecting device accommodated in a case, which is mountable and usable with an existing X-ray imaging machine or the like, has been underway, and functional improvement is anticipated. Further, image information detecting devices that include a communication cable for outputting electrical signals readout from the image information detecting unit to an external image processing unit are also known. Further, image information detecting devices that include a wireless transmission means that wirelessly transmits electrical signals read out from the image information detecting unit to an external image processing device, thereby ensuring setting freedom of the device to an X-ray imaging machine without hindered by the communication cable, are also known as described, for example, in Japanese Unexamined Patent Publication Nos. 7(1995)-140255 and 2003-210444. The image information detecting device with a wireless transmission capability described in Japanese Unexamined Patent Publication Nos. 7 (1995)-140255 and 2003-210444, however, has only the wireless transmission means as a transmission means to an external device. Consequently, it has a problem that it is unable to transmit to the external device when wireless transmission is difficult, inappropriate, or the like. In view of the circumstances described above, it is an object of the present invention to provide an image information detecting apparatus with a wireless transmission capability, and further operable for transmission to an external device when wireless transmission is difficult, inappropriate, or the like. An image information detecting apparatus according to the present invention comprises: an image information detecting unit that receives recording light representing image information, and records the image information by storing electric charges generated therein by the recording light; a wireless communication means operable for transmission to an external device; a wire communication means operable for transmission to the external device through a detachable communication cable; a wire transmission setting means for setting the wire communication means operable for transmission; a wireless transmission prohibiting means for prohibiting wireless transmission from the wireless communication means when the wire communication means is set operable for transmission by the wire transmission setting means. The referent of “detachable communication cable” as used herein means a communication cable which is connectable to and detachable from the wire communication means. As for the transmission contents to be transmitted to the external device, various information items may be conceivable. More specifically, an image information recording ready signal, an image information recording completion signal, and, if an image information readout means is provided, the image information signals read out thereby may be included in the contents. Further, the transmission contents may include imaging information associated with the image information, and the like. The wireless transmission prohibiting means may be a means that prohibits the wireless transmission from the wireless communication means by terminating power supply thereto when the wire communication means is set operable for transmission by the wire transmission setting means. If a configuration is adopted in which a communication cable detection means for detecting connection of the communication cable to the wire communication means is further provided, the wire transmission setting means may be a means that sets the wire communication means operable for transmission when the connection of the communication cable is detected by the communication cable detection means. Further, if a configuration is adopted in which a receiving means operable for reception from external equipment is further provided, the wire transmission setting means may be a means that sets the wire communication means operable for transmission based on control through communication from the external equipment. Here, the “external equipment” may be a device identical with the “external device” described above. Still further, if a configuration is adopted in which a manual entry means such as a switch or a touch panel is provided on the body of the image information detecting apparatus, the wire communication setting means may be a means that sets the wire communication means operable for transmission by manual control through the manual entry means. The image information detecting apparatus of the present invention includes: a wireless communication means operable for transmission to an external device; a wire communication means operable for transmission to the external device through a detachable communication cable; and a wire transmission setting means for setting the wire communication means operable for transmission. This allows transmission to the external device through the wire communication means by setting the wire communication means operable for transmission by the wire transmission setting means and prohibiting wireless transmission from the wireless communication means when wireless transmission is difficult, inappropriate, or the like. This improves the convenience of the apparatus, and may prevent, in particular, accidental wireless transmission when the wireless transmission is inappropriate. If the wireless transmission prohibiting means is a means that prohibits the wireless transmission from the wireless communication means by terminating power supply thereto when the wire communication means is set operable for transmission by the wire transmission setting means, the power consumption of the apparatus may be reduced when wireless transmission is disabled. If a configuration is adopted in which a communication cable detection means for detecting connection of the communication cable to the wire communication means is further provided, and the wire transmission setting means sets the wire communication means operable for transmission when the connection of the communication cable is detected by the communication cable detection means, the wire communication means may be set operable for transmission by simply connecting the communication cable to the wire communication means, so that the operational procedure for wire transmission is simplified. If a configuration is adopted in which a receiving means operable for reception from external equipment is further provided, and the wire transmission setting means sets the wire communication means operable for transmission based on control through communication from the external equipment, the structure of the apparatus may be simplified without requiring any function for switching the communication means, such as a switch or the like, on the body of the apparatus. Hereinafter, embodiments of the present invention will be described in detail with reference to the accompanying drawings. FIG. 1 is a block diagram of a cassette type image information detecting apparatus according to a first embodiment of the present invention, illustrating the structure and circuit blocks of the relevant part thereof. The image information detecting apparatus 1 is used for medical X-ray imaging. It has a substantially rectangular solid shaped case 10 which includes therein: an image information detecting unit 12; a readout light emitting unit 13; a control unit 20; a battery 30 for supplying power to respective units or sections; a wireless communication unit 31; a wire communication unit 32; a connection terminal 33 that connects the wire communication unit 32; and a communication selection switch 34 for selecting wireless or wire transmission. The connection terminal 33 is connectable to a connection terminal 36 of a communication cable 35 which is connected to an external system controller 37. The system controller 37 is a controller capable of providing wireless and wire communication, acting as the external device or external equipment of the present invention. It may be provided on the X-ray imaging machine, or in a control room which is different from the X-ray imaging machine room, or the like. The image information detecting unit 12 receives X-ray exposure, which is the recording light transmitted through a subject, and records image information of the subject represented by the X-ray by storing electric charges generated therein by the X-ray as the latent image charges. The unit 12 includes the following layers in the order listed below: a first electrode layer that transmits the X-ray; a recording photoconductive layer that shows electrical conductivity when exposed to the X-ray; a charge transport layer that acts substantially as an insulator against the latent image charges and substantially a conductor for transport charges of the polarity opposite to that of the latent image charges; a readout photoconductive layer that shows electrical conductivity when exposed to readout light; and a second electrode layer that transmits the readout light. When recording image information, X-ray is irradiated on the image information detecting unit from the side of the first electrode layer with a high voltage being applied between the first and second electrode layers to provide an electric field therebetween, and an amount of electric charges corresponding to the dose of X-ray irradiated on the image information detecting unit is stored in a storage section formed at the interface between the recording photoconductive layer and charge transport layer as the latent image charges. The image information recorded in the image information detecting unit 12 is read out by scanning the image information detecting unit 12 with readout light emitted from the readout light emitting unit 13. The control unit 20 includes a recording/readout control section 21 for causing an electric field to be applied to the image information detecting unit 12 when recording image information, image signals to be read out according to the image information recorded in the image information detecting unit 12 through optical scanning of the image information detecting unit 12 with the readout light emitting unit 13, and image processing to be performed on the readout image signals; a memory 22 for storing the processed image signals; and a communication control section 23 for setting the wire communication unit 32 operable for transmission when wire communication is selected by the communication selection switch 34, and setting the wireless communication unit 31 operable for transmission when wireless communication is selected thereby. The communication control section 23 prohibits wireless transmission from the wireless communication unit 31 when the wire communication unit 32 is set operable for transmission, acting both as the wire transmission setting means and wireless transmission prohibiting means of the present invention. Prior to initiating X-ray imaging, the radiographer selects through the communication selection switch 34 either the wire communication unit 32 or wireless communication unit 31 for transmission. Normally, the wireless communication unit 31 is selected for transmission, since the communication cable 35 is not connected to the image information detecting apparatus 1 in order to ensure setting freedom thereof. The communication control section 23 sets the wireless communication unit 31 operable for reception and transmission when wireless transmission is selected by the communication selection switch 34. On the other hand, if wireless transmission is difficult, inappropriate, or the like, transmission using the wire communication unit 32 is selected. In this case, the connection terminal 36 of the communication cable 35 is connected to the connection terminal 33 in advance. The communication control section 23 sets the wire communication unit 32 operable for reception and transmission when wire transmission is selected by the communication selection switch 34, and disables the wireless transmission function of the wireless communication unit 31. Here, the wireless transmission function of the wireless communication unit 31 is disabled, but the receiving function thereof is maintained for normal operation. The operation of the image information detecting apparatus 1 will be described in detail with reference to an example case where wire communication is selected. Various data, including imaging menu, ID information, application voltage required for image recording, readout speed, and the like, are inputted to the control unit 20 from the system controller 37 through the communication cable 35, connection terminal 36, and connection terminal 33. Further, during the time period from the time when the X-ray is ready to be irradiated to the time just after the X-ray is irradiated, an X-ray ready signal is inputted to the control unit 20 from the system controller 37, and when the X-ray is irradiated, an X-ray irradiation signal is also inputted to the control unit 20 from the system controller 37. The control unit 20 also performs overall operation control of the image information detecting apparatus 1. Here, power is supplied to each unit or section from the battery 30 through a power supply wiring 38. In FIG. 1, however, the detailed wiring state of the power supply wiring 38 is omitted for clarity. Hereinafter, the operation of the image information detecting cassette 1 according to the present embodiment will be described. First, recording of image information and reading out of the image information will be described. Prior to initiating X-ray imaging, various data, including imaging menu, ID information, application voltage required for image recording, readout speed, and the like, are inputted to the control unit 20 from the system controller 37 through the communication cable 35, connection terminal 36, and connection terminal 33. The control unit 20 stores these data in the memory 22, and uses them as required by reading out from the memory 22. During the time period from the time when the X-ray is ready to be irradiated to the time just after the X-ray is irradiated, an X-ray ready signal is inputted to the control unit 20 from the system controller 37 through the communication cable 35 and wire communication unit 32. Further, when the X-ray is irradiated, an X-ray irradiation signal is inputted to the control unit 20. When the X-ray ready signal is inputted to the control unit 20, the control unit 20 controls the recording/readout control section 21 to cause a recording high voltage to be applied to the image information detecting unit 12. Then, X-ray transmitted through the subject is irradiated on the image information detecting unit 12, and an amount of electric charges corresponding to the dose of X-ray irradiated thereon is stored therein as the latent image charges. In reading out operation, the control unit 20 controls the recording/readout control section 21 to cause the image information detecting section 12 to be scanned with the readout light emitted from the readout light emitting unit 13 so that the image information recorded in the image information detecting section 12 is read out, which is processed and tentatively stored in the memory 22. Thereafter, the image information stored in the memory 22 is sent to the system controller 37 by the control unit 20 through the wire communication unit 32, connection terminal 33, connection terminal 36, and communication cable 35. As is clear from the above description, the image information detecting apparatus 1 includes the wireless communication unit 31 operable for transmission to the system controller which is an external device, and wire communication unit 32 operable for transmission to the system controller through the detachable communication cable 35. This allows transmission to the system controller through the wire communication unit 32 by setting the wire communication unit 32 operable for transmission and prohibiting wireless transmission from the wireless communication unit 31 when wireless transmission is difficult, inappropriate, or the like. This improves the convenience of the apparatus, and, in particular, accidental wireless transmission may be prevented when the wireless transmission is inappropriate. In the present embodiment, switching between wireless and wire transmission is implemented by the communication selection switch 34 provided on the body of the image information detecting apparatus 1. But, the present embodiment is not limited to this, and a configuration may be adopted in which the switching between the wireless and wire communication is implemented by the signal received by the wire communication unit 32 or wireless communication unit 31. In this case, the structure of the image information detecting apparatus 1 may be simplified without requiring the communication selection switch 34. Further, if the communication method is to be switched through the wire communication unit 32, switching of the communication unit may be implemented even if the transmission and receiving functions of the wireless communication unit 31 are disabled when the wire communication unit 32 is used. Hereinafter, an image information detecting apparatus 2 according to a second embodiment of the present invention will be described with reference to FIG. 2. In FIG. 2, components identical to those shown in FIG. 1 are given the same reference numerals, and will not be elaborated upon further here unless otherwise specifically required. FIG. 2 is a block diagram of the image information detecting apparatus 2, illustrating the structure and circuit blocks of the relevant part thereof. The image information detecting apparatus 2 is a cassette type apparatus used for medical X-ray imaging. It has the case 10 which includes therein: the image information detecting unit 12; the readout light emitting unit 13; a control unit 40; the battery 30 for supplying power to respective units or sections; a power supply control unit 51 for controlling power supply from the battery 30; the wireless communication unit 31; the wire communication unit 32; the connection terminal 33 that connects the wire communication unit 32; and a communication cable detection unit 50 for detecting if the connection terminal 36 of the communication cable 35, which is connected to the system controller 37, is connected to the connection terminal 33. The control unit 40 includes the recording/readout control section 21 for controlling the recording and reading out of image information; the memory 22; and a communication control section 41 for setting the wire communication unit 32 operable for transmission when connection of the connection terminal 36 of the communication cable 35, which is connected to the system controller 37, to the connection terminal 33 is detected by the communication cable detection unit 50, and setting the wireless communication unit 31 operable for transmission when the connection terminal 36 is detected not to be connected to the connection terminal 33. The communication control section 41 controls the power supply control unit 51 to terminate power supply to the wireless communication unit 31 when the wire communication unit 32 is set operable for transmission, acting both as the wire transmission setting means and wireless transmission prohibiting means of the present invention. When wireless communication is used, power is supplied to each unit or section, and when the wire communication is used, power is supplied to each unit or section other than the wireless communication unit 31 from the battery 30 through the power supply control unit 51 and power supply wiring 38. Prior to initiating X-ray imaging, the radiographer selects either the wire communication unit 32 or wireless communication unit 31 for transmission by determining whether to connect the connection terminal 36 of the communication cable 35 to the connection terminal 33 of the wire communication unit 32. Normally, the communication cable 35 is not connected to the wire communication unit 32 in order to ensure setting freedom of the image information detecting apparatus 1. Here, the connection terminal 36 of the communication cable 35 is detected not to be connected to the connection terminal 33 by the communication cable detection unit 50, and the wireless communication unit 31 is selected automatically for reception and transmission by the communication control section 41. On the other hand, if wireless transmission is difficult, inappropriate, or the like, the connection terminal 36 of the communication cable 35 is connected to the connection terminal 33 of the wire communication unit 32 by the radiographer. Here, the connection terminal 36 of the communication cable 35 is detected to be connected to the connection terminal 33 of the wire communication unit 32 by the communication cable detection unit 50, and transmission by the wire communication unit 32 is selected automatically for transmission by the communication control section 41, which at the same time controls the power supply control unit 51 to terminate power supply to the wireless communication unit 31. The operation of the image information detecting apparatus 2 is identical to that of the image information detecting apparatus 1, so that detailed description thereof will not be repeated here. As is clear from the above description, the image information detecting apparatus 2 includes the wireless communication unit 31 operable for transmission to the system controller 37, and wire communication unit 32 operable for transmission to the system controller 37 through the detachable communication cable 35. When the communication cable 35 is connected to the wire communication unit 32, the wire communication unit 32 is selected automatically, and power supply to the wireless communication unit 31 is terminated to prohibit wireless transmission. This allows transmission to the system controller 37 through the wire communication unit 32 by setting the wire communication unit 32 operable for transmission and prohibiting wireless transmission from the wireless communication unit 31 when wireless transmission is difficult, inappropriate, or the like. This improves the convenience of the apparatus, and, in particular, accidental wireless transmission may be prevented when the wireless transmission is inappropriate. Further, when the wire communication unit 32 is set operable for transmission, the power supply to the wireless communication unit 31 is terminated so that the power consumption of the apparatus 2 may be reduced when wireless communication is disabled. Further, the wire communication unit 32 is selected for transmission when the connection of the communication cable 35 is detected by the communication cable detection unit 50. Thus, the wire communication unit may be automatically set operable for transmission by simply connecting the communication cable 35 to the wire communication unit 32, so that the operational procedure for wire transmission is simplified. In the embodiments described above, a direct conversion/optical readout type image information detecting unit is used as the image information detecting unit 12. The present invention is not limited to this, and an indirect conversion type image information detecting unit, in which luminescence emitted from a phosphor by receiving recording light is irradiated on a recording photoconductive layer, and signal charges obtained by photoelectrically converting the irradiated luminescence are stored therein, may also be used. Still further, a TFT readout type image information detecting unit, in which electric charges generated in a photoconductive layer that shows electrical conductivity when exposed to recording light are read out by scan driving TFTs, or the like may also be used. |
|
039309428 | summary | This invention relates to installations for the storage or use of fluids which are harmful or noxious to the environment. It applies particularly to tanks which contain chemical products or fluids at a very low temperature, nuclear vessels in light water and sodium-cooled installations, and contaminated effluent tanks. In such tanks, inspection and small repair operations must be carried out quickly in order to minimise intervention times on equipment for which such intervention makes the equipment unavailable for operation. Such inspection must frequently be carried out under extreme operating temperature conditions (for example in cryogenic fluid tanks or in sodium reactors) which do not allow access by operators. It would be advantageous to provide automatic fast inspection means as a precaution. It would also be advantageous to provide automatic safety features for conditions where spacing and obstacles make access difficult. For safety reasons, the conventional system comprises constructing two fluid-tight barriers. The first barrier contains the dangerous or expensive fluid and the second is adapted to provide a seal in the event of rupture of the first during operation of the installation, and such an accident may be serious and result in prolonged or even complete stoppage of the installation. Disregarding rupture, even a minimal leakage may result in the installation being unavailable for operation, in which case the secondary barrier is then used simply for recovery of the product and for protective purposes. The object of this invention is to facilitate examination, maintenance and repair of the first barrier under difficult environmental conditions, and more particularly allow the use of inspection methods which can be applied easily and quickly, and means for logging the zone which is to be monitored. For reasons associated with the environment, and in view of the disorder and expense that may result from non-availability of the equipment, it is also advantageous to provide quick and automatic inspection equipment, fault locating equipment, and equipment for minor work, even work of a type which would not merit any special attention in less dangerous circumstances. To this end, the invention proposes systematically to promote free access to the primary barrier, to leave the outer surface free of any thermal insulation or any other protective layer, and transfer such insulation or layers to the secondary barrier, and to provide between the two barriers adequate space for the inspection and maintenance operations. Conveniently, locating members disposed on the surface of the barrier enable those zones which it is required to inspect from time to time to be accurately located. Advantageously, studs secured to the surface of the primary barrier enable the preceding aims to be accomplished and also facilitate the introduction and operation of automated control or working equipment. The above steps are not sufficient in every case of application of the invention. To enable equipment of all kinds to be readily used, the space between the primary and secondary barriers must be amply dimensioned. If no other steps were taken, the large space thus formed would be invaded by the fluid in the event of rupture of the primary barrier, and the level of fluid would drop abruptly inside said barrier. This may create extremely dangerous situations. This applies, for example, to nuclear reactors in which, for safety reasons combined with the operation of fission products, the reactor core must not be drained or exposed irrespective of the accidents occurring. The reason for this is that heating due to the gamma effect is sufficient to cause fusion of the core if the latter is not cooled by convection of the coolant fluid. To obviate this catastrophic situation, the invention proposes to fill the space between the two barriers by displaceable members, thus preventing any excessive fall in the level of liquid in the event of rupture of the primary barrier. In most cases, the space to be filled is in the form of a volume of revolution and the filling members or blocks are in the form of segments bounded by meridian planes and disposed consecutively in a circumferential series. They are placed on a circular track and an access door to the space fitted out in this way is provided in the secondary barrier. In the event of intervention, the door is opened, one of the filling segments is withdrawn if necessary to clear a space into which the inspection, monitoring, repair or, more generally, maintenance equipment is introduced. The assembly formed by the train of filling blocks and the equipment is moved along the outer circular surface of the primary barrier to bring the equipment to a predetermined zone, and work is continued by remotely controlled and remotely monitored operations. If necessary, the space liberated by removal of a segment may also be used for the introduction of personnel to enable them to reach the primary barrier. If required, the filling blocks may act as storage tanks which may, for example, be used to facilitate vessel emptying operations. On completion of the maintenance operations, the reverse sequence of events takes place, i.e. the train is set in motion to bring the equipment to the opening, the equipment is withdrawn and, where applicable, the missing segment is returned and the secondary barrier closed. The objects, characteristics and advantages of the invention will also be apparent from the following description of one exemplified embodiment illustrated in the accompanying drawings. The example selected, which has no limitative character, relates to a sodium-cooled nuclear reactor, but the invention applies equally to light water cooled reactor vessels and other tanks for dangerous products. |
claims | 1. A charged-particle beam irradiation device that irradiates an object to be irradiated with a charged-particle beam, the charged-particle beam irradiation device comprising:a scanning member that scans the object to be irradiated with the charged-particle beam;an irradiation amount setting unit that sets an irradiation amount of the charged-particle beam at a plurality of target scanning positions on a scanning line of the charged-particle beam with which the scanning member scans the object to be irradiated; anda scanning speed setting unit that sets a target scanning speed of the charged-particle beam at each of the target scanning positions on the basis of the irradiation amount set by the irradiation amount setting unit. 2. The charged-particle beam irradiation device according to claim 1, further comprising:a control unit that controls the scanning member according to the target scanning position and the target scanning speed so that the object to be irradiated is scanned with the charged-particle beam. 3. The charged-particle beam irradiation device according to claim 2, further comprising:a measurer that measures a scanning position and a scanning speed of the charged-particle beam with which the scanning member scans the object to be irradiated,wherein the control unit controls the scanning member by performing position/speed feedback control for adjusting a control input to the scanning member so that an error between the target scanning position and the scanning position measured by the measurer is eliminated and an error between the target scanning speed and the scanning speed measured by the measurer is eliminated. 4. The charged-particle beam irradiation device according to claim 1, further comprising:a beam intensity detector that detects intensity of the charged-particle beam,wherein the control unit adjusts scanning speed of the scanning member so as to offset the change of beam intensity when the beam intensity detected by the beam intensity detector is changed to the outside of a predetermined error range. 5. A charged-particle beam irradiation method that irradiates an object to be irradiated with a charged-particle beam, the charged-particle beam irradiation method comprising:an irradiation amount setting step of setting an irradiation amount of the charged-particle beam at a plurality of target scanning positions on a scanning line of the charged-particle beam with which a scanning member scans the object to be irradiated; anda scanning speed setting step of setting a target scanning speed of the charged-particle beam at each of the target scanning positions on the basis of the irradiation amount set in the irradiation amount setting step. 6. A non-transitory computer readable medium storing a program causing a computer to execute a process for irradiating an object with a charged particle beam, the process comprising:setting an irradiation amount of a charged-particle beam at a plurality of target scanning positions on a scanning line of the charged-particle beam with which a scanning member scans an object to be irradiated; andsetting a target scanning speed of the charged-particle beam at each of the target scanning positions on the basis of the irradiation amount set by the irradiation amount setting. |
|
abstract | Techniques and tools for rating computer products are described. For example, software ratings are based on subjective evaluations to determine computer system requirements for a positive user experience, while a computer running a capability tool rates a computer system's (or hardware component's) ability to run software. A capability rating for hardware is determined by comparing a set of features and performance results with capability rating requirements. In another aspect, a capability rating is communicated using a standardized presentation. In another aspect, capability rating level requirements are proposed (e.g., by a ratings board) and then finalized. A capability rating level is determined for computer products (e.g., by a testing organization) based on the finalized requirements and analysis of the products (e.g., by a computer running a capability tool). In another aspect, a software system includes an inventory module, a performance testing module, and an inventory and performance evaluator module. |
|
description | This application claims the benefit of U.S. Provisional Application No. 61/624,034 filed Apr. 13, 2012. U.S. Provisional Application No. 61/624,034 filed Apr. 15, 2012 is incorporated herein by reference in its entirety. The following relates to the nuclear power reactor arts, nuclear reaction control apparatus arts, control rod assembly arts, and related arts. In thermal nuclear power plants, a nuclear reactor core comprises a fissile material having size and composition selected to support a desired nuclear fission chain reaction. The core is disposed in a pressure vessel immersed in primary coolant water. It is further known to control or stop the reaction by inserting “control rods” comprising a neutron-absorbing material into guide tubes passing through the reactor core. When inserted, the control rods absorb neutrons so as to slow or stop the chain reaction. The control rods are operated by control rod drive mechanisms (CRDMs). In so-called “regulating” control rods, the insertion of the control rods is continuously adjustable so as to provide continuously adjustable reaction rate control. In so-called “shutdown” control rods, the insertion is either fully in or fully out. During normal operation the shutdown rods are fully retracted from the reactor core; during a SCRAM, the shutdown rods are rapidly fully inserted so as to rapidly stop the chain reaction. Control rods can also be designed to perform both regulating and shutdown rod functions. In some such dual function control rods, the control rod is configured to be detachable from the CRDM in the event of a SCRAM, such that the detached control rod falls into the reactor core under the influence of gravity. In some systems, such as naval systems, a hydraulic pressure or other positive force (other than gravity) is also provided to drive the detached control rod into the core. To complete the control system, a control rod/CRDM coupling is provided. A known coupling includes a connecting rod having a lower end at which a spider is secured. The upper portion of the connecting rod operatively connects with the CRDM. In regulating rods, this connection includes a lead screw or other incremental adjustment element. Conventionally, the lead screw scrams with the connecting rod, spider, and control rods as a translating assembly (also known as the “control rod assembly”). In some known approaches, however, the lead screw may be retained in the CRDM such that only the connecting rod scrams. See, e.g. U.S. Pub. No. 2010-0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety, and U.S. Pub. No. 2011-0222640 A1 published Sep. 15, 2011 which is incorporated herein by reference in its entirety. To reduce cost and overall system complexity, a single CRDM is typically connected with a plurality of control rods via a spider. In this arrangement, all the control rods coupled with a single spider together as a translating control rod assembly (CRA). In practice a number of CRDM units are provided, each of which is coupled with a plurality of control rods via a spider, so as to provide some redundancy. The spider extends laterally away from the lower end of the connecting rod to provide a large “surface area” for attachment of multiple control rods. The translating CRA (including the control rods, spider, connecting rod, and optionally also the lead screw) represents a substantial mass that falls under the force of gravity during a scram. It is advantageous for the translating CRA to have substantial mass in order to provide the driving force for the scram. In some designs, the translating CRA has a mass of a hundred pounds to a few hundred pounds, and may reach a terminal velocity of 10 feet per second or higher. Thus, consideration is given to the termination of the scram, that is, to the slowing and stopping of the downward falling of the translating assembly at the end of the scram event. Prior to termination of the scram, the descending control rod tips engage dashpot tubes of narrowed inner diameter that produce a slowing force via a piston effect. Alternatively, a dashpot can be located in the CRDM which engages with the descending connecting rod or lead screw. Although such dashpots can provide some cushioning, the ultimate “stop” for the scram is impact of the descending spider onto the top of the fuel assembly (or onto a structural plate located above the fuel assembly). To cushion this final impact, it is known to employ one or more helical springs disposed in the connecting rod and/or spider. However, it is difficult to insert long springs into the translating CRA, and shorter springs do not provide large energy absorption. As a consequence, a substantial portion of the kinetic energy of the translating CRA is ultimately absorbed by the impact of the spider onto the fuel assembly, which can lead to damage to these critical components. In one aspect of the disclosure, a control rod assembly comprises at least one movable control rod including a neutron absorbing material, a connecting rod configured for connection to an associated control rod drive mechanism (CRDM) for controlling movement of the at least one control rod, and a coupling operatively connecting the at least one control rod and the connecting rod. The coupling includes a spider engaged with the connecting rod and the at least one moveable control rod, and a kinetic energy absorbing element supported by the spider for absorbing kinetic energy during a SCRAM event, the kinetic energy absorbing element configured to act between the spider and an associated impact surface. The spider can include a casing having an upper surface and a lower surface, and the kinetic energy absorbing element can include a plunger mounted to the spider for reciprocating movement parallel to a SCRAM direction, and can further include at least one Belleville washer interposed between the casing and the plunger for biasing the plunger away from the top surface of the casing. The spider can be a heavy spider that is elongate in the SCRAM direction, and the at least one Belleville washer can be contained within the spider. The heavy spider can include a first portion comprising a first material having a first density and a second portion comprising a second material having a second density that is greater than the first density. The connecting rod can be detachably engagable with the associated CRDM such that detachment of the spider causes a translating assembly including at least the connecting rod, the spider and the at least one control rod to fall toward a reactor core disposed in a lower region of the nuclear reactor pressure vessel, and whereby the kinetic energy absorbing element is configured to absorb at least some of the kinetic energy of the translating assembly during impact with the reactor core. The coupling can include a J-Lock coupling for connecting the spider with a lower end of the connecting rod, the J-Lock coupling can be supported in a central bore of the spider extending between the upper and lower surface of the casing, the J-Lock coupling can include a spring, and the Belleville washer can be coaxially aligned with said spring of the J-Lock coupling. The spring of the J-Lock coupling and the Belleville washer can be axially coextensive along a portion of their respective lengths. In another aspect of the disclosure, an apparatus comprises a spider connectable with a lower end of an associated connecting rod of an associated CRDM and with upper ends of a plurality of associated control rods, and a kinetic energy absorbing element including at least one Belleville washer supported by the terminal element for absorbing kinetic energy during a SCRAM event. In another aspect of the disclosure, a control rod assembly comprises: a spider; a plurality of mutually parallel control rods connected with the spider, the control rods including neutron absorbing material; a connecting rod having an end connected with the spider and configured for detachable connection to an associated control rod drive mechanism (CRDM) wherein responsive to detachment of said connection to the associated CRDM the control rod assembly undergoes a scram event in which the control rod assembly descends under force of gravity; and a stack of Belleville washers disposed in at least one of the connecting rod and the spider, the stack of Belleville washers arranged to absorb kinetic energy of a descent terminating impact of the control rod assembly during a SCRAM event. In some such embodiments the stack of Belleville washers is disposed in a central bore of the spider. In another aspect of the disclosure, an apparatus comprises: a spider configured to support a plurality of mutually parallel control rods; and a kinetic energy absorbing element disposed at least partially in a central bore of the spider and including a plunger arranged to stop descent of the spider during a SCRAM event with kinetic energy developed during the SCRAM event being absorbed by the kinetic energy absorbing element disposed in the central bore of the spider. In some embodiments the kinetic energy absorbing element further comprises a stack of Belleville washers disposed in a central bore of the spider. In some such embodiments the stack of Belleville washers does not extend outside of the central bore of the spider. In existing control rod assemblies, a plurality of control rods connect with a lightweight, “spidery” spider having a minimal weight and surface area oriented broadside to the SCRAM direction. The spider is configured to provide a large “effective” area for attachment of control rods, but a small “actual” area contributing to hydraulic resistance during SCRAM. Both the spider and the connecting rod are stainless steel components so as to provide benefits such as strength and robustness, low cost, manufacturability, and compatibility with the reactor vessel environment. Disclosed herein are control rod assemblies that include one or both of the following aspects: (i) replacement of the conventional lightweight spider with a “heavy” spider that serves as a terminal weighting element, and/or (ii) replacement of a substantial portion of the stainless steel of the spider and/or connecting rod with a denser material such as tungsten (optionally in a powdered or granulated form), molybdenum, tantalum, or so forth. The disclosed control rod assemblies are substantially heavier than conventional control rod assemblies of the same vertical dimensions, which advantageously enhances the speed and reliability of gravitationally-induced SCRAM. In the case of control rod assemblies employing the disclosed heavy spider as a terminal weighting element, the increased weight provided by the heavy spider as compared with a conventional lightweight spider enables the heavy spider to optionally have a larger actual surface area broadside to the SCRAM direction (for example, in order to provide the additional weight) as compared with the conventional spider. In addition, the heavier than conventional control rod assemblies disclosed herein also include a kinetic energy absorbing element for absorbing kinetic energy during a scram event. The kinetic energy absorbing elements disclosed herein act to limit the impact of the spider with the “stopping” surface or component, such as a top plate fitting of a fuel assembly. As disclosed herein, special spring elements/arrangements are used to accommodate the kinetic load. In one example, a stack of Belleville washers is used. It will be appreciated that aspects of the disclosure relating to the kinetic energy absorbing element are also applicable to control rod assemblies with conventional light-weight spiders. For example, the disclosed kinetic energy absorbing element is useful with a conventional control rod assembly that uses a lightweight spider but includes relatively long (and hence massive) control rods and/or a long connecting rod. With reference to FIG. 1, a relevant portion of an illustrative nuclear reactor pressure vessel 10 includes a reactor core 12 located proximate to a bottom of the pressure vessel 10. The core 12 includes or contains radioactive material (not shown) such as, by way of illustrative example, enriched uranium oxide (that is, UO2 processed to have an elevated 235U/238U ratio). A control rod drive mechanism (CRDM) unit 14 is diagrammatically illustrated. The illustrative CRDM 14 is an internal CRDM that is disposed within the pressure vessel 10; alternatively, an external CRDM may be employed. FIG. 1 shows the single illustrated CRDM unit 14 as an illustrative example; however, more generally there are typically multiple CRDM units each coupled with a different plurality of control rods (although these additional CRDM units are not shown in FIG. 1, the pressure vessel 10 is drawn showing the space for such additional CRDM units). Below the CRDM unit 14 is a control rod guide frame 16, which in the perspective view of FIG. 1 blocks from view the control rod/CRDM coupling assembly (i.e., the spider and connecting rod, both not shown in FIG. 1). Extending below the guide frame 16 is a plurality of control rods 18. FIG. 1 shows the control rods 18 in their fully inserted position in which the control rods 18 are maximally inserted into the core 12. In the fully inserted position, the spider is located at a lower location 20 within the control rod guide frame 16 (and, again, hence not visible in FIG. 1). In the illustrative embodiment of FIG. 1, the CRDM unit 14 and the control rod guide frame 16 are spaced apart by a standoff 22 comprising a hollow tube having opposite ends coupled with the CRDM unit 14 and the guide frame 16, respectively, and through which the connecting rod (not shown in FIG. 1) passes. FIG. 1 shows only a lower portion of the illustrative pressure vessel 10. In an operating nuclear reactor, an open upper end 24 of the illustration is connected with one or more upper pressure vessel portions that together with the illustrated lower portion of the pressure vessel 10 forms an enclosed pressure volume containing the reactor core 12, the control rods 18, the guide frame 16, and the internal CRDM unit 14. In an alternative embodiment, the CRDM unit is external, located above the reactor pressure vessel. In such embodiments, the external CRDM is connected with the control rods by a control rod/CRDM coupling assembly in which the connecting rod extends through a portal in the upper portion of the pressure vessel. With reference to FIG. 2, the control assembly including the CRDM unit 14, the control rod guide frame 16, the intervening standoff 22, and the control rods 18 is illustrated isolated from the reactor pressure vessel. Again, the control rod/CRDM coupling assembly is hidden by the control rod guide frame 16 and the standoff 22 in the view of FIG. 2. With reference to FIG. 3, the control rod guide frame 16 and the standoff 22 is again illustrated, but with the CRDM unit removed so as to reveal an upper end of a connecting rod 30 extending upwardly above the standoff 22. If the CRDM unit has regulating rod functionality, then this illustrated upper end of the connecting rod 30 engages with the CRDM unit to enable the CRDM unit to raise or lower the control rod 30 and, hence, the attached control rods 18 (not shown in FIG. 3). If the CRDM unit has shutdown rod functionality, then this illustrated upper end is detachable from the CRDM unit during SCRAM. In each of FIGS. 1-4, a SCRAM direction S is indicated, which is the downward direction of acceleration of the falling control rods in the event of a SCRAM. With reference to FIG. 4, the control rods 18 and the connecting rod 30 are shown without any of the occluding components (e.g., without the guide frame, standoff, or CRDM unit). In the view of FIG. 4 an illustrative terminal weighting element or “heavy” spider 32 is visible, which provides connection of the plurality of control rods 18 with the lower end of the connecting rod 30. It will be noticed that, unlike a conventional spider, the heavy spider 32 has substantial elongation along the SCRAM direction S. The illustrated heavy spider 32 has the advantage of providing enhanced weight which facilitates rapid SCRAM; however, it is also contemplated to replace the illustrated terminal weighting element 32 with a conventional “spidery” spider. With reference to FIGS. 5 and 6, a perspective view and a side-sectional perspective view, respectively, of the heavy spider 32 is shown. The heavy spider 32 includes a substantially hollow casing 40 having upper and lower ends that are sealed off by upper and lower casing cover plates (surfaces) 42, 44. Four upper casing cover plates 42 are illustrated in FIG. 5 and two of the upper casing cover plates 42 are shown in the side-sectional perspective view of FIG. 6. The tilt of the perspective view of FIG. 5 occludes the lower cover plates from view, but two of the lower cover plates 44 are visible “on-edge” in the side-sectional view of FIG. 6. The illustrative heavy spider 32 includes four lower casing cover plates 44 arranged analogously to the four upper casing cover plates 42 illustrated in FIG. 5. Further visualization of the illustrative heavy spider 32 is provided by FIG. 7, which shows a top view of the hollow casing 40 with the cover plates omitted. As seen in FIG. 7, the hollow casing 40 is cylindrical having a cylinder axis parallel with the SCRAM direction S and a uniform cross-section transverse to the cylinder axis. That cross-section is complex, and defines a central passage 50 and four cavities 52 spaced radially at 90° intervals around the central passage 50. The cross-section of the hollow casing 40 also defines twenty-four small passages 54 (that is, small compared with the central passage 50), of which only some of the twenty-four small passages 54 are expressly labeled in FIG. 7. The four cavities 52 spaced radially at 90° intervals around the central passage 50 are next considered. The substantially hollow casing 40 and the upper and lower cover plates 42, 44 are suitably made of stainless steel, although other materials are also contemplated. The upper and lower cover plates 42, 44 seal the four cavities 52. As shown in the side-sectional view of FIG. 6, the four cavities 52 are filled with a filler 56 comprising a heavy material, where the term “heavy material” denotes a material that has a higher density than the stainless steel (or other material) that forms the hollow casing 40. For example, the filler 56 may comprise a heavy material such as tungsten (optionally in a powdered or granulated form), depleted uranium, molybdenum, or tantalum, by way of some illustrative examples. By way of illustrative example, stainless steel has a density of about 7.5-8.1 grams/cubic centimeter, while tungsten has a density of about 19.2 grams/cubic centimeter and tantalum has a density of about 16.6 grams per cubic centimeter. In some preferred embodiments, the heavy material comprising the filler 56 has a density that is at least twice the density of the material comprising the casing 40. In some preferred embodiments in which the casing 40 comprises stainless steel, the heavy material comprising the filler 56 preferably has a density that is at least 16.2 grams per cubic centimeter. (All quantitative densities specified herein are for room temperature.) In some embodiments, the filler 56 does not contribute to the structural strength or rigidity of the heavy spider 32. Accordingly, heavy material comprising the filler 56 can be selected without consideration of its mechanical properties. For the same reason, the filler 56 can be in the form of solid inserts sized and shaped to fit into the cavities 52, or the filler 56 can be a powder, granulation, or other constitution. The cover plates 42, 44 seal the cavities 52, and so it is also contemplated for the heavy material comprising the filler 56 to be a material that is not compatible with the primary coolant flowing in the pressure vessel 10. Alternatively, if the heavy material comprising the filler 56 is a material that is compatible with the primary coolant flowing in the pressure vessel 10, then it is contemplated to omit the upper cover plates 42, in which case the cavities 52 are not sealed. Indeed, if the filler 56 is a solid material securely held inside the cavities 52, then it is contemplated to omit both the upper cover plates 42 and the lower cover plates 44. With continuing reference to FIGS. 5-7 and with further reference to FIG. 8, the heavy spider 32 passes through the control rod guide frame 16 as the control rods 18 are raised or lowered by action of the CRDM unit 14. The cylindrical configuration with constant cross-section over the length of the heavy spider 32 along the SCRAM direction S simplifies this design aspect. Moreover, the control rod guide frame 16 should cam against each control rod 18 to provide the desired control rod guidance. Toward this end, the cross-section of the heavy spider 32 is designed with recesses 58 (some of which are labeled in FIG. 7). As shown in FIG. 8, into these recesses 58 fit mating extensions 60 of the control rod guide frame 16. A gap G also indicated in FIG. 8 provides a small tolerance between the outer surface of the heavy spider 32 and the proximate surface of the control rod guide frame 16. The twenty-four partial circular openings of the guide frame 16 which encompass the twenty-four small passages 54 of the heavy spider 32 are sized to cam against the control rods 18. For completeness, FIG. 8 also shows the connecting rod 30 disposed inside the central passage 50 of the heavy spider 32. FIGS. 5-7 show that providing space for the four cavities 52 substantially increases the actual cross-sectional area of the heavy spider 32 (that is, the area arranged broadside to the SCRAM direction S), as compared with the actual cross-sectional area that could be achieved without these four cavities 52. In some embodiments, the “fill factor” for the cross-section oriented broadside to the SCRAM direction S (including the area encompassed by the cover plates 42, 44) is at least 50%, and FIG. 7 demonstrates that the fill factor is substantially greater than 50% for the illustrative terminal weighting element. Thus, the design of the heavy spider 32 is distinct from the “spidery” design of a typical spider, which is optimized to minimize the actual surface area broadside to the SCRAM direction S and generally has a fill factor of substantially less than 50% in order to reduce hydraulic resistance. In general, the SCRAM force achieved by the weight of the heavy spider 32 more than offsets the increased hydraulic resistance of the greater actual broadside surface area imposed by the four cavities 52. Additional weight to overcome the hydraulic resistance and enhance SCRAM speed is obtained by elongating the heavy spider 32 in the SCRAM direction S as compared with a conventional spider. Said another way, a ratio of a length of the heavy spider 32 in the SCRAM direction S versus the largest dimension oriented broadside to the SCRAM direction S is optionally equal to or greater than one, and is more preferably equal to or greater than 1.2. The illustrative heavy spider 32 is not a generally planar element as per a typical spider, but rather is a volumetric component that provides substantial terminal weight to the lower end of the connecting rod 30. Another advantage of the elongation of the heavy spider 32 in the SCRAM direction S is that it optionally allows for streamlining the heavy spider 32 in the SCRAM direction S. This variation is not illustrated; however, it is contemplated to modify the configuration of FIG. 5 (by way of illustrative example) to have a narrower lower cross-section and a broader upper cross section, with a conical surface of increasing diameter running from the narrower lower cross-section to the broader upper cross section. The small passages 54 for securing the control rods would remain oriented precisely parallel with the SCRAM direction S (and, hence, would be shorter for control rods located at the outermost positions). Such streamlining represents a trade-off between hydraulic resistance (reduced by the streamlining) and weight reduction caused by the streamlining. The illustrative heavy spider 32 provides a desired weight by a combination of the filler 56 comprising a heavy material (which increases the average density of the heavy spider 32 to a value greater than the average density of stainless steel) and the elongation of the heavy spider 32 (which increases the total volume of the heavy spider 32). In some embodiments, it is contemplated to omit the filler material entirely, and instead to rely entirely upon elongation to provide the desired weight. For example, the illustrated heavy spider 32 can be modified by omitting the four cavities 52 and the filler 56. In this configuration the casing 40 can be replaced by a single solid stainless steel element having the same outer perimeter as the casing 40, with the top and bottom of the single solid stainless steel element defining (or perhaps better stated, replacing) the upper and lower casing cover plates 42, 44. Various embodiments of the disclosed heavy spiders use a stainless steel casing that does not compromise the primary function of providing a suitable structure for coupling the control rods to the lower end of the connecting rod. At the same time, the stainless steel casing leaves sufficient void or cavity volume to allow a filler comprising a heavy material to be inserted. Although stainless steel is referenced as a preferred material for the casing, it is to be understood that other materials having desired structural characteristics and reactor pressure vessel compatibility can also be used. The filler comprising heavy material is suitably tungsten, depleted uranium, or another suitably dense material. With reference to FIGS. 9, 10, and 11, various attachment configurations can be used for securing the connecting rod 30 in the attachment passage 50 of the casing 40 of the heavy spider 32. In an illustrative example of one such attachment configuration, the central passage 50 of the casing 40 houses a J-Lock female attachment assembly 70, which is suitably coaxially disposed inside the central passage 50 of the casing 40. FIG. 9 illustrates a side sectional view of the J-Lock female attachment assembly 70, while FIG. 10 shows a side view of the connected assembly and FIG. 11 shows a side sectional view of the connected assembly. With particular reference to FIG. 9, the illustrative J-Lock female attachment assembly 70 includes a hub 72 which in the illustrative embodiment comprises a round cylinder coaxially welded or otherwise secured in the central passage 50 of the casing 40. Alternatively, the hub may be integral with or defined by an inside surface of the central passage 50. The hub 72 serves as an interface between the casing 40 and the J-Lock female attachment components, which include three J-Lock pins 74 (two of which visible in the sectional view of FIG. 9) disposed inside of the hub 72. These pins 74 provide the connection points for a J-Lock male attachment assembly 80 (see FIG. 11) disposed at the lower end of the connecting rod 30. A J-Lock plunger 76 and a J-Lock spring 84 keeps the J-Lock male attachment assembly 80 of the connecting rod 30 in place once it has been engaged with the heavy spider 32. (Locked arrangement shown in FIG. 11). The illustrative J-Lock female attachment assembly 70 further includes a lower plunger 82, an inner spring 78, and a spring washer 86 which cooperate to absorb the impact of the lower translating assembly (that is, the translating combination of the control rods 18, the heavy spider 32, the connecting rod 30, and optionally a lead screw (not shown)) during a SCRAM. As will be appreciated, the lower plunger 82 is mounted within the attachment passage (bore) 50 of the casing 40 of the heavy spider 32, and protrudes from a bottom surface of the heavy spider 32 for engagement with another surface during a SCRAM event. The lower plunger 82 is supported for reciprocating movement within the attachment passage 50, and biased downwardly by spring 78 and/or J-Lock spring 84. Together, the plunger 82 and spring 78 and/or J-Lock spring 84 comprise a kinetic energy absorbing element supported by the heavy spider 32 for absorbing kinetic energy during a SCRAM event. The kinetic energy absorbing element can be configured to act between the terminal element 32 and an upper plate of an associated fuel assembly, for example, as will be described in more detail below. The illustrative J-Lock connection between the lower end of the connecting rod 30 and the heavy spider 32 is an example. More generally, substantially any type of connection, including another type of detachable connection or a permanently welded connection or an integral arrangement, is contemplated. The J-Lock arrangement has the advantage of enabling the connecting rod 30 to be detached from the heavy spider 32 (and, hence, from the control rods 18) by a simple “push-and-twist” operation. This allows the connecting rod 30 to be moved separately from the remainder of the translating assembly (that is, the heavy spider 32 and the attached control rods 18) during refueling of the nuclear reactor. With reference to FIGS. 10 and 11, additional weight for the translating assembly is additionally or alternatively obtained by enhancing the density of the connecting rod 30. Toward this end, the illustrative connecting rod 30 includes a hollow (or partially hollow) connecting rod tube 90 which (as seen in the sectional view of FIG. 11) contains a filler region 92 adapted to accept a filler comprising a heavy material. Thus, the connecting rod tube 90 serves the structural purpose analogous to the casing 40 of the terminal weighting element 32, while the filler comprising heavy material serves a weighting (or average density-enhancing) purpose analogous to the filler 56 of the terminal weighting element 32. In one suitable embodiment, the filler comprising heavy material is in the form of tungsten slugs each having a diameter substantially coinciding with an inner diameter of the connecting rod tube 90 and being stacked in the connecting rod tube 90, with the number of stacked tungsten slugs being selected to achieve the desired weight. If the number of tungsten slugs is insufficient to fill the interior volume of the connecting rod tube 90 and it is desired to avoid movement of these slugs, then optionally the filler is prevented from shifting by a suitable biasing arrangement or by filling the remaining space within the interior volume of the connecting rod tube 90 with a light weight material such as stainless steel slugs. In the illustrative example of FIG. 11, a biasing arrangement region 98 is employed, in which the interior volume of the connecting rod tube 90 is sealed off by upper and lower welded plugs 94, 96, and a biasing arrangement such as a compressed spring takes up any slack along the SCRAM direction S that may be introduced by incomplete filling of the interior volume of the connecting rod tube 90 by the filler. Instead of tungsten, the heavy material comprising the filler may be depleted uranium, molybdenum, tantalum, or so forth, by way of some other illustrative examples. The filler may comprise one or more solid slugs or rods, a powder, a granulation, or so forth. With continuing reference to FIGS. 10 and 11, the illustrative connecting rod 30 has an upper end that includes an annular groove 100 for securing with a latch of the CRDM unit 14 (latch not shown), and a magnet 102 for use in conjunction with a control rod position sensor (not shown). A suitable embodiment of the CRDM unit 14 including a motor/lead screw arrangement for continuous (regulating rod) adjustment and a separate latch for detaching the connecting rod 30 from the CRDM unit 14 (with the lead screw remaining operatively connected with the motor) is described in U.S. Pub. No. 2010-0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety and in U.S. Pub. No. 2011-0222640 A1 published Sep. 15, 2011 which is incorporated herein by reference in its entirety. Alternatively, in other embodiments a lead screw (not shown) is secured with or integral with the connecting rod tube 90, and the lead screw SCRAMs together with the connecting rod/terminal weighting element (or spider)/control rod (in other words, the lead screw forms part of the translating assembly during SCRAM). In some such alternative embodiments, the motor is suitably coupled with the lead screw by a separable ball nut that separates to release the lead screw and initiate SCRAM. If the interior volume of the hollow connecting rod tube 90 is only partially filled by the filler, then stainless steel rods or some other light weight filler (not shown) may be inserted into the remaining interior volume to fill complete the filling. The filler generally has a lower coefficient of thermal expansion than the stainless steel (or other material) of the hollow connecting rod tube 90. The connecting rod 30 is assembled at room temperature, and then heated to its operating temperature. For a connecting rod having a length of, e.g. 250 centimeters or greater, the thermal expansion will result in the rod tube 90 increasing by an amount of order a few centimeters or more. The lower coefficient of thermal expansion of the filler results in a substantially lower length increase of the filler. The spring in the biasing arrangement region 98 suitably compensates for this effect. Additionally, if the spring is located below the filler, then it can assist in dissipating the kinetic energy of the filler at the termination of the SCRAM drop. Turning now to FIGS. 12 and 13, and initially to FIG. 12, another exemplary control rod assembly is illustrated and identified generally by reference numeral 100. As will be appreciated, a typical reactor core will have a plurality of control rod assemblies 100. As shown in FIG. 12, the illustrative control rod assembly 100 consists of an array of twenty-three or twenty-four rodlets 104 attached to a spider 108, which may be a relatively massive tungsten-ballasted heavy spider as described herein and as illustrated in FIGS. 12 and 13, or alternatively may be a conventional lightweight spider. Each rod let 104 contains neutron poisons sealed in cladding tubes (e.g., 304L stainless steel) via welded end plugs of the same material. Each control rod assembly 100 is associated with a fuel assembly 110 into which the control rodlets 104 are moved to control operation of the reactor. In some core designs, the control rod assemblies in the standard control banks use full-length silver-indium-cadmium absorbers in twenty-three or twenty-four rodlets. The control rod assemblies in the shutdown banks have a row of fifteen or sixteen rodlets with B4C absorbers in the outboard positions surrounding eight inboard rodlets with standard silver-indium-cadmium absorbers. These are merely illustrative examples. The array of rodlets 104 are mechanically fastened to the arms 112 of the spider 108 by threading extensions on the top end plugs into tapped receiving holes in the bottom of the spider arms 112 and then lock welding them to the spider 108 so they cannot back out. Tapered “nuts” are threaded into similar tapped receiving holes in the top surface of the spider arms to serve as lead-ins when the control rods 104 are withdrawn from the core upward through the reactor internals guide structure. The “nuts” are also lock-welded to the spider to prevent them from backing off during operation. A reduced-diameter section in the top end plugs allows the rodlets 104 to flex relative to the spider 108 as necessary to accommodate any misalignment between a fuel assembly 110 and the rod guide structure. During operation of a reactor, the control rod assemblies 100 in the shutdown banks are typically fully withdrawn from the core, while the regulating control rod assemblies may be partially inserted into the core and their axial position changed with time. When a SCRAM occurs, the connecting rods (not shown in FIG. 12, but described above) are released and the control rod assemblies 104 fall into the core under gravity, reaching a terminal velocity of approximately 12 feet per second if they are fully withdrawn prior to the SCRAM. When a control rod assembly 104 is approximately 75% inserted into the corresponding fuel assembly 110 during a SCRAM, the tips of the rodlets 104 enter dashpot tubes (not shown) installed in the bottom of the control rod guide tubes of the fuel assembly 110. These small dashpot tubes have a significantly smaller inside diameter than the larger control rod guide tubes, greatly reducing the radial clearance with the control rodlets 104. This produces a piston effect as water is forced to flow through the resulting narrow annulus, slowing the falling control rod assemblies. However, the downward motion continues until eventually the spider 108 on the falling control rod assembly impacts the top of the fuel assembly 110. An impact limiter, also referred to as a kinetic energy absorbing element, accommodates the remaining kinetic energy. Turning to FIG. 13, the kinetic energy absorbing element includes a cup-like spider plunger 120 that is preloaded against the bottom of the spider central bore B by two concentric spring features. Plunger 120 includes a radially outwardly extending flange 121 that is configured to engage a shoulder 122 of central bore B to limit further downward axial movement of the plunger 120 from the position shown in FIG. 13. A relatively stronger, outer spring feature is comprised of a stack S of Belleville washers (also called Belleville springs) 124, such as age-hardened Inconel-718 Belleville washers. The stack S of Belleville washers 124 provides the majority of the preload on the plunger 120. The Belleville washer stack S reacts against a hub 128 which is welded into the top of the central bore B and contains a J-lock mechanism 136 that mates with a coupling mechanism on the bottom end of a connecting rod, (not shown). The relatively weaker inner spring 132 is a conventional helical compression spring made of, for example, age-hardened Inconel-718 wire, which reacts against a J-lock plunger 136. The Belleville washer stack S and the coil spring 132 are coaxially aligned and are axially coextensive along at least a portion of their respective lengths resulting in a compact arrangement. When the connecting rod engages the spider 108, the plunger 136 is depressed sufficiently to enable the coupling mechanism of the connecting rod to pass under J-lock pins 140 in the hub 128 and rotate to the locked position (e.g., as described above). During a scram, the kinetic energy absorbing element 120, S acts between the spider 32 and an associated impact surface, such as an upper plate of an associated fuel assembly, or an upper core plate (not shown) spanning the space above the fuel assemblies making up the reactor core. The Belleville washer stack S provides a higher load capability in a more compact arrangement than can be achieved using a conventional helical compression spring. This allows the kinetic energy absorbing element to absorb more energy than would otherwise be possible given the limited volume available in the spider hub. This high energy capability particularly well-suited to applications where a massive tungsten ballasted spider, such as that set forth above, is used, and more generally is well-suited to applications in which the overall translating control rod assembly is heavy, e.g. due to the use of long control rods, a long connecting rod, a heavy spider, various combinations thereof, and so forth. It will be appreciated that the amount of energy the impact limiter must absorb depends on the mass of the falling assembly and its velocity, which in turn depends on how far it falls and on the fluid resistance imparted by travel through the primary coolant. Everything else being equal, a longer control rod assembly will fall further from the fully withdrawn position, and impact the fuel assembly with greater energy. The capability of the kinetic energy absorbing element to absorb this energy can be adjusted by tailoring the Belleville washer stack S. Turning to FIGS. 14a-14f, various Belleville washer stacks S suitably used as the stack S of FIG. 13 are illustrated. The plunger stroke, stiffness, and maximum load capability of the kinetic energy absorbing element can be customized utilizing several different approaches. For example: Decreasing the number of Belleville washers in the stack S as illustrated in FIG. 14a, without changing their geometry or material of construction, will result in a kinetic energy absorbing element having a shorter allowable plunger stroke and a higher stiffness with the same maximum load capability. Conversely, increasing the number of Belleville washers in the stack S as illustrated in FIG. 14b, without changing their geometry or material of construction, will result in a kinetic energy absorbing element having a longer allowable plunger stroke and a lower stiffness with the same maximum load capability. Using thicker Belleville washers in the stack S as illustrated in FIG. 14c will result in an kinetic energy absorbing element having a shorter allowable plunger stroke, a higher stiffness, and a higher allowable load. Conversely, employing thinner Belleville washers in the stack S as illustrated in FIG. 14d will yield a design with a longer allowable plunger stroke, a lower stiffness, and a lower allowable load. It is also possible to tailor the kinetic energy absorbing element characteristics by changing the nesting arrangement of the Belleville washers. FIGS. 14a-14d all show a stacking arrangement in which the orientation of each Belleville washer is reversed relative to the Belleville washer above and below it. This alternating stacking arrangement maximizes the allowable stroke while minimizing the stiffness and allowable load of the kinetic energy absorbing element. If instead, the same Belleville washers are stacked in pairs as illustrated in FIG. 14e, with the washers in each pair nested together in the same orientation and each pair of washers then oriented the reverse of the neighboring pair above and below, the allowable stroke of the kinetic energy absorbing element will be cut in half, the stiffness will be doubled, and the allowable load will be doubled. If instead, the same Belleville washers are stacked in sets of three as illustrated in FIG. 14f, with the washers in each set again nested together in the same orientation and each set of three washers then oriented the reverse of the neighboring set of three washers above and below, the allowable stroke of the kinetic energy absorbing element will be cut by two thirds, the stiffness will be tripled, and the allowable load will be tripled. The same principle can be extended to Belleville washers arranged in sets of four, five, etc.; each increase in the number of washers nested together resulting in a proportional decrease in the allowable stroke of the kinetic energy absorbing element and proportional increases in the stiffness and allowable load. Although the kinetic energy absorbing elements disclosed herein have been described in the context of weighted terminal elements or elongated terminal elements (i.e., heavy spiders), it should be appreciated that aspects of the disclosure are applicable to conventional spider assemblies as well. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
|
abstract | A highly heat-resistant laminated component for a fusion reactor has at least of a plasma-facing area made of tungsten or a tungsten alloy, a heat-dissipating area of copper or a copper alloy with a mean grain size of more than 100 μm, and an interlayer of a refractory metal-copper-composite. The refractory metal-copper-composite has a macroscopically uniform copper and refractory-metal concentration progression and a refractory metal concentration of between 10 vol. % and 40 vol. % over its entire thickness. |
|
062597582 | claims | 1. A structure material for a cooling tube of a water-cooled nuclear reactor having a surface exposed to an aqueous cooling medium containing hydrogen peroxide, said surface coated with a coating comprising matter selected from the group consisting of manganese, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, the structure material being sufficient for effecting and causing decomposition of said hydrogen peroxide in said cooling medium. 2. The structure material as claimed in claim 1 wherein said alloy is selected from the group consisting of carbon steel, zirconium, alloy steel, stainless steel, nickel-based alloys, and cobalt-based alloys. 3. The structure material as claimed in claim 1 wherein said matter of said coating is capable of adsorbing said hydrogen peroxide. 4. The structure material as claimed in claim 3 wherein the coating is deposited by a process selected from the group consisting of plasma spraying, chemical vapour deposition, physical vapour deposition, welding, electroless plating and electrolytic plating. 5. A structure material for a water-cooled nuclear reactor comprising metal piping, said metal piping having a surface exposed to an aqueous liquid containing hydrogen peroxide, the structure material comprising a coating, the coating comprising matter selected from the group consisting of manganese, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, said coating being sufficient for effecting and causing decomposition of said hydrogen peroxide in said aqueous liquid. 6. The structure material for a nuclear reactor as claimed in claim 5 wherein said metal piping is comprised of alloys selected from the group consisting of carbon steel, alloy steel, zirconium, stainless steel, nickel-based alloys, cobalt-based alloys and mixtures thereof. 7. The reactor as claimed in claim 6 wherein said matter of said coating is capable of adsorbing said hydrogen peroxide. 8. The reactor as claimed in claim 7 wherein said piping surface coating is deposited by a process selected from the group consisting of plasma spraying, chemical vapour deposition, physical vapour deposition, welding, electroless plating and electrolytic plating. 9. A method for lowering the electrochemical corrosion potential of a metal alloy for use in a cooling tube in a water-cooled nuclear reactor having a surface exposed to an aqueous liquid containing hydrogen peroxide, comprising the step of coating said surface with matter selected from the group consisting manganese, copper, cadmium, oxides thereof, chemical compounds thereof and mixtures thereof, said coating being sufficient for effecting and causing decomposition of said hydrogen peroxide. 10. The method as claimed in claim 9 wherein said metal alloy is selected from the group consisting of carbon steel, alloy steel, stainless steel, zirconium nickel-based alloys, and cobalt-based alloys and mixtures thereof. 11. The method as claimed in claim 10 wherein said matter is capable of adsorbing said hydrogen peroxide in said liquid. 12. The method as claimed in claim 11 wherein said matter is deposited by a process selected from the group consisting of plasma spraying, chemical vapour deposition, physical vapour deposition, welding, electroless plating, electrolytic plating and mixtures thereof. |
claims | 1. A system for imaging a patient's breast with x-rays comprising:a flat panel digital x-ray receptor having a proximal edge;an x-ray source selectively emitting a collimated x-ray beam toward the receptor,a breast platform for supporting a patient's breast for x-ray imaging:a compression paddle removably mounted for selective movement at least (a) along a direction of said proximal edge before compressing a patient's breast against the breast platform and (b) along a direction of said beam to thereafter compress the breast against the breast platform before imaging the breast with said x-ray beam and to release the breast after said imaging;wherein at least for selected breast x-ray protocols a patient's breast is positioned off-center relative to said proximal edge and said paddle also is positioned off-center relative the proximal edge, according to said off-center position of the breast, to compress the breast against the breast platform and toward the receptor for x-ray imaging; andan anti-scatter grid mounted between the breast platform and the x-ray receptor for selective movement relative to the x-ray receptor and the breast platform between a first position in which imaging x-rays reaching the receptor after passing through the breast also pass through the grid and a second position in which said imaging x-rays do not pass through the grid. 2. A system as in claim 1 in which said anti-scatter grid is mounted for movement in a direction transverse to the direction of said proximal edge of the receptor. 3. A system as in claim 1 in which said anti-scatter grid is in said first position for a first imaging mode and in said second position for a second imaging mode. 4. A system as in claim 3 in which said breast platform is spaced from the x-ray receptor by a greater distance in said second imaging mode than in said first imaging mode. 5. A system for imaging a patient's breast with x-rays comprising:a flat panel digital x-ray receptor having a proximal edge;an x-ray source selectively emitting a collimated x-ray beam toward the receptor;a breast platform for supporting a patient's breast for x-ray imaging;a compression paddle removably mounted for selective movement relative to the breast platform in at least two directions, said compression paddle compressing a patient's breast against the breast platform before imaging the breast with said x-ray beam and releasing the breast from compression after said imaging, and at least a portion of said movement in each of said at least two directions taking place before the breast is in a compressed state for imaging; andan anti-scatter grid mounted between the breast platform and the x-ray receptor For selective movement relative to the x-ray receptor and the breast platform between a first position in which imaging x-rays reaching the receptor after passing through the breast also pass through the grid and a second position in which said imaging x-rays do not pass through the grid;said anti-scatter grid being in said first position for imaging the breast with x-rays in a first imaging mode and being in said second position for imaging the breast with x-rays in a second imaging mode that produces x-ray images of the breast different from those produced in the first imaging mode. 6. A system as in claim 5 in which said at least two directions of motion of the compression paddle comprise a first direction and a second direction that is transverse to the first direction. 7. A system as in claim 6 in which said first direction of motion of the compression paddle is in a direction generally parallel to an imaging plane of said x-ray receptor. 8. A system as in claim 6 in which said first direction is generally along the direction of said proximal edge of said x-ray receptor. 9. A system for imaging a patient's breast with x-rays comprising:a flat panel digital x-ray receptor having a proximal edge;an x-ray source selectively emitting a collimated x-ray beam toward the receptor;a breast platform for supporting a patient's breast for x-ray imaging;a compression paddle removably mounted for selective movement relative to the breast platform, said compression paddle compressing a patient's breast against the breast platform before imaging the breast with said x-ray beam and releasing the breast from compression after said imaging; andan anti-scatter grid movably mounted between the breast platform and the x-ray receptor and having a first position in which imaging x-rays reaching the receptor after passing through the breast also pass through the grid;said anti-scatter grid being in said first position for imaging the breast with x-rays in a first imaging mode but being out of said first position, such that said imaging x-rays do not pass through the grid, for imaging the breast with x-rays in a second imaging mode that produces x-ray images of the breast different from those produced in the first imaging mode. 10. A system as in claim 9 in which said compression paddle is mounted for selective motion toward and away from the breast platform to compress and release the patient 's breast and in a direction generally parallel to said proximal edge to offset the paddle relative to the breast platform and the x-ray receptor before compressing the patient's breast. 11. A system as in claim 9 in which said compression paddle has a breast engaging surface that is substantially smaller in area than said imaging surface of the receptor. 12. A system as in claim 11 in which said breast engaging surface of the compression paddle is generally parallel to the imaging surface of said x-ray receptor. 13. A system as in claim 11 in which said compression paddle moves in a direction generally parallel to said proximal edge between an end position in which a lateral edge thereof is generally aligned with an edge of the receptor and a position in which the compression paddle is centered relative to the receptor. |
|
summary | ||
061335789 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 illustrates the scanner unit 10 which has a pair of cross beams 11 perpendicular to the machine direction (MD) (also designated y) 12 or the moving web 13. In this particular case this is a paper sheet being manufactured by a paper making machine. Mounted to travel across the sheet in a cross direction (CD), also designated x, is a pair of measuring heads or transducers 14a and 14b (see also FIG. 2). The top transducer 14a includes a radiation source 16 and the bottom transducer 14b a radiation detector 17. As is well-known in the art, the amount of radiation passing through the moving web 13 is indicative of a characteristic of the web such as basis weight. Referring to FIGS. 3 and 4 radiation detector 17 includes a circular plate 21 mounted to the transducer 14b with a open aperture 22 which has a 4 segment detector array 1, 2, 3, and 4 in a planar format. Such array is also shown in FIG. 5 and how it is aligned with the machine direction (MD) (or y direction) and cross direction (CD) (or x). Each of the segments of the array is a silicon photo voltaic cell with a uniform radiation response having good dynamic range. Each of the elements or segments 1, 2, 3 and 4 is symmetrically arranged around a center detector axis 23. This axis is nominally coincident with the axis of the radiation source (or the center axis of the aperture over which it is superimposed) assuming there is no misalignment. Each detector segment 1, 2, 3, and 4 is nominally square in configuration with the diagonals of elements 1 and 2 being co-linear and passing through the center detector axis 23. The center detector 23 is also nominally coincident with the radiation axis. This arrangement of detectors will provide compensation for the lateral misalignment that is commonly caused by the drive belt arrangement that moves the source and detector across the sheet in a cross direction. As illustrated in FIG. 5 each square detector has, of course, a center indicated with the centers of detectors 3 and 4 being offset in the y direction a distance designated "a"; similarly the detectors 1 and 2 have their centers offset in the x direction a distance "a." As illustrated in FIG. 6, radiation received by the each of the detectors is in the form of the four currents I.sub.1, I.sub.2, I.sub.3 and I.sub.4. These are input into a processor 26 to provide a measurement such as basis weight (B.W.) which has been compensated for any misalignment of the source and detector. Now referring specifically (FIGS. 7 and 8) to the top transducer head 14a a circular plate 30 which faces the moving web 13 includes an aperture 31 which has a center axis 32. FIG. 8 is the opposite side of plate 30 showing mounted on it an encapsulated radiation source 33 (Promethium 147) which is mounted for rotation on an axis 34 to move along the locus 35 from a stowed position where the radiation from the nuclear source is shielded by the plate 30 to a first position where the radiation source and its radiation axis is actually coincident with the center axis 32. Similarly also illustrated in FIG. 8 is clear transparent radiation standardization flag 36, for example made of a thin sheet of clear plastic, which is mounted for rotation on an axis 37 along from a stowed position and along a locus of movement 38 so that flag may be rotated between the radiation source and the moving web for standardization purposes. Thus referring back to FIG. 7 both the offset axis 34 of rotation of the encapsulated radiation source and flag axis 37 are on the line 39 which is a diameter of aperture 31 and of course passes through the center axis 32 of the aperture. FIGS. 9 and 11 illustrate the rotating parts and the mounting for encapsulated radiation source 33 which rotates in the direction shown by the arrow on the axis 34 and the standardization flag 36 rotates in the direction shown around the axis 37. FIGS. 10 and 12 are side views of the source and standardization flag, respectively. Referring to FIG. 10, the radiation source includes a substantially large area of radiation emission, for example 15.6 mm, which radiates in a substantially parabolic pattern 41 having a radiation axis 40. Mounted to provide movement to the radiation source and the standardization flag are ball-type connectors 42 and 43, respectively. And referring to FIG. 13 these connectors are again shown connected to pistons 44 and 45 to rotate the radiation source 33 and standardization flag 36 from a stowed positions to an active positions. With the mounting of the radiation source 33 for movement in a plane parallel to that of the web or perpendicular to that of the radiation axis even though a fairly large area of radiation is present necessitating a relatively larger encapsulated radiation source 33, effective shielding (or shuttering) is provided for the radiation source in the stowed position illustrated in FIG. 8. Here the open-ended encapsulated radiation source faces the circular disk 30. At the same time, there is adequate spacing 51 (FIG. 10) between the circular disk 30 and the bottom of the transducer 33 to allow space for standardization flag 36 to slide there between. Although the detector ideally is shown as having 4 segments (this is because the error correction process is believed to be relatively simple in this case) three segments in the form of 120.degree. wedges could be used. All that is necessary is that a planar array of three or more segments be used which are symmetrically arranged around a center axis. Referring now to FIGS. 5 and 6 and the set of equations following below, as discussed above the four detectors arranged symmetrically around the center 23 provides compensation for lateral misalignment caused by the drive belt arrangement which moves the source and detector across the sheet. As illustrated and discussed above two of the detectors 1 and 2 have their centers on the x axis and detectors 3 and 4 have their centers on the y axis. These centers are all at the nominal distance "a" from the center detector axis 23. When the detector and the source are perfectly aligned individual detector signals I.sub.1 through I.sub.4 can be modeled by the equations (1) through (4) as a function of misalignment in the x-y plane. Referring specifically to these equations, x and y are the misalignment coordinates, a is the distance of the individual detectors from the center detector axis 23, S is a signal from an any one individual detector when the center such detector is coincident with the radiation axis of the source, and k is a factor that depends on the size and distance of the source, and to some degree on the basis weight to be measured. Thus, for example, referring to equation (1) if the radiation axis is centered on the center of the detector 1, then x=a and y=0 and I.sub.1 is equal to S. However, when it moves to the center axis 23, as shown illustrated in FIG. 10, the radiation source has an emission in the shape of a parabola and therefore the equations (1) through (4) are in the form of a parabolic function. Equation (5) is the sum of equations (1) through (4) and the sum is indicated as I.sub.T. It is apparent that if x and y are equal to 0 (that is there is no misalignment) that equation (5) provides accurate measure of received radiation which is the term 4S(1-ka.sup.2). In other words, the actual radiation received and the constants k and a. However, where there is an alignment, x and y must be taken into account. Since there is no direct way of measuring x and y, the error must be eliminated by a mathematical manipulation of the various currents I.sub.1 through I.sub.4. From an inspection of FIG. 5 it can be seen that detectors 1 and 2 and their equations represent errors in the cross direction (the most significant misalignment) and that detectors 3 and 4 and their equations (3) and (4) errors in the machine direction. And it is also obvious that the error term is, as illustrated in equation (5), as an x.sup.2 plus y.sup.2 type of factor. Thus the mathematical entity is created of equation (6) which matches the error term k(x.sup.2 +y.sup.2). And in addition this error correcting entity must also eliminate the S term. It has been found that this can be accomplished as indicated in equation (6) by taking the square of the cross directional signals that is (I.sub.1 -I.sub.2).sup.2, the square of the difference of the machine direction signals (I.sub.3 -I.sub.4).sup.2 and dividing it by the square of the I.sub.T. Referring to the result of that computation, because of the division by I.sub.T.sup.2 no S is in the result. Furthermore in the denominator of equation (6) all of the latter terms have less than 1% effect on the total value of the expression which may be therefore reduced to the expression of equation (7). Therefore, the error term in equation (5) containing the x.sup.2 +y.sup.2 can be rewritten as shown in equation (8). What has been done is that the k(x.sup.2 +y.sup.2) term of equation (7) which is the other half of equation (6), has been solved and has been substituted in equation (8). Then the right side of the equation (8) is substituted in equation (5) to produce a misalignment corrected signal I.sub.T. Noted that equation (9) is a re-arrangement of equation (5) where the term 4S(1-ka.sup.2) is solved for and thus the correction term is all in the denominator. The constant term (viz. ka.sup.2) of equation (9) as illustrated in the left-hand side of the equation (10) depends slightly on the amount of mass between the source and detector due to scattering of the radiation beam. Experiments have shown one possible way to model this term is as illustrated in the right side of equation (10) where p and q are constants and I.sub.T0 is the value of I with no web in the measuring gap. When it is taken into account that denominator of equation (9) is almost one the final correction algorithm is equation (11). This corrected signal is processed by the processor 26 illustrated in FIG. 6. The calculation of basis weight is done in the same manner as uncompensated signals have been used previously. That is a corrected form of Beer's laws. Thus to summarize the processing means for eliminating the misalignment error takes the sum of the signals from the four detectors and the square of difference of the pairs of machine direction and cross directional signals. And this directly eliminates the error term. Thus an improved nuclear gauge for measuring a characteristic of moving sheet material and alignment compensation has been provided. Equations EQU I.sub.1 =S{1-k[(x-a).sup.2 +y.sup.2 ]} (1) EQU I.sub.2 =S-{1-k[(x+a).sup.2 +y.sup.2 ]} (2) EQU I.sub.3 =S{1-k[x.sup.2 +(y-a).sup.2 ]} (3) EQU I.sub.4 =S{1-k[x.sup.2 +(y+a).sup.2 ]} (4) EQU I.sub.T =4S(1-ka.sup.2)[1-k(x.sup.2 +y.sup.2)/(1-ka.sup.2)](5) EQU [(I.sub.1 -I.sub.2).sup.2 +(I.sub.3 -I.sub.4).sup.2 ]/I.sub.T.sup.2 =k(x.sup.2 +y.sup.2)/{(1-ka.sup.2)[(1-ka.sup.2)/ka.sup.2 -2(x.sup.2 +y.sup.2)/a.sup.2 +(ka.sup.2 /(1-ka.sup.2))((x.sup.2 +y.sup.2)/a.sup.2).sup.2 ]} (6) ##EQU1## EQU [1-k(x.sup.2 +y.sup.2)/(1-ka.sup.2)]=1-[(1-ka.sup.2)/ka.sup.2 ][(I.sub.1 -I.sub.2).sup.2 +(I.sub.3 -I.sub.4).sup.2 ]/I.sub.T.sup.2 (8) EQU I'.sub.T =I.sub.T /{1-[(1-ka.sup.2)/ka.sup.2 ][(I.sub.1 -I.sub.3).sup.2 +(I.sub.3 -I.sub.4).sup.2 ]/I.sub.T.sup.2 }=4S(1-ka.sup.2)(9) EQU (1-ka.sup.2)/ka.sup.2 =p+q(I.sub.T /I.sub.T.sbsb.0) (10) EQU I'.sub.T =I.sub.T {1+[p+q(I.sub.T /I.sub.T.sbsb.0)][(I.sub.1 -I.sub.2).sup.2 +(I.sub.3 -I.sub.4).sup.2 ]/I.sub.T.sup.2 }(11) |
042343844 | summary | FIELD OF THE INVENTION The present invention relates to a support structure for the core of a high capacity gas cooled high temperature reactor surrounded by an annular side reflector. More particularly, the invention relates to a core formed by a bed of spherical fuel elements with a number of pebble retracting tubes passing through the support structure. BACKGROUND OF THE PRIOR ART Supporting structures for gas cooled core reactors built up from vertical columns are known. In such structures, either a number of supporting plates of polygonal configuration are arranged with lateral spacing adjacent to each other in a plane. Each of the plates are designed to carry a limited number of moderator columns, or else each moderator column is supported by a metal cylinder. At the same time, cooling gas channels present in both the moderator columns and the metal cylinders are aligned with each other. An example of a supporting structure of the type described in the foregoing is shown in the West German published application No. 11 77 751. In this patent application, seven moderator columns, equipped with central borings, are associated with every supporting plate. The borings are coaxial with borings in the supporting plates. Only the column arranged in the center of the group of moderator columns is connected with the supporting plate. In this manner, the supporting plate is capable of expansion around the axis of this column. In West German published application No. 1 122,641, a supporting structure is illustrated in which, as previously described, for every moderator column a supporting element in the form of a metal cylinder is provided. The moderator columns and the metal cylinders rest on each other on spherical bearing surfaces in the manner of a ball joint. It is known from West German published application No. 1 194 071, that support of the moderator structure is made of a solid material of a nuclear reactor on a flat supporting surface. The supporting surface is composed of several parts, which are rigid in themselves. Each part of the supporting structure is carried by a certain number of support posts, arranged symmetrically around the axis of the moderator structure. The supporting surfaces surround a center part of the plate and several concentric annular frames which may have a small amount of free radical play between them. According to the West German patent application No. 1 614,684, the supporting structure for a reactor block may consist of a cell structure composed of tubular elements, connected with each other at their ends by two horizontally arranged disks. Each of the tubular elements, which have hexagonal cross sections, is aligned with one of the columns of the reactor block. The supporting structure also serves as a biological shield, for which purpose at least two layers of graphite inserts are provided in the tubular elements. Another supporting structure for the core of a gas cooled nuclear reactor is described in the West German patent application No. 1 956 266, in which the core rests on the bottom of the reactor pressure vessel upon refractory material. Cooling gas is admitted to the core through channels formed in the refractory material and connected with gas channels in the core and the gas space outside the core. The state of the art also includes a support floor for a pebble bed reactor consisting of pebbles of a high temperature material and a supporting structure for the weight of the pebble support layer and the fuel pebbles, the pebbles being piled directly onto the pebble support layer. The supporting structure and the pebble support layer are separated by a layer of tiles resistant to elevated temperatures. Several vertical tubular stacks are provided at regular intervals for the constant venting of the fuel pebbles through the supporting structure and the pebble bed layer. These tubular stacks determine the least thickness and the average thickness of the supporting pebble bed. In another pebble bed reactor, the THTR-300 MWe, the supporting floor for the bed of fuel pebbles consist of a plurality of hexagonal graphite blocks arranged into freely movable columns and having axial borings for the cooling gas. The columns formed by the graphite blocks are individually supported by one round column each, said round columns being attached into the floor and consists of graphite plates. The fixed point of the round columns is represented by the central pebble vent tube. By reducing the nominal dimensions of the hexagonal graphite blocks, expansion gaps are created which permit unhindered thermal expansion within the supporting floor without exceeding its overall dimensions. Under certain non-stationary operating conditions, e.g., in the case of accidental disruption, the gaps may add up and lead to relatively large single gaps. The closing of such gaps by relocating the graphite blocks requires high relocating forces which, however, in view of the dimensions of THTR-300 MWe are of minor importance. If the capacity of the reactor is increased, together with the dimensions of the reactor core and its installations, supporting floors designed on the principle described in the foregoing cannot be utilized without further changes, because the expansion gap and the relocating forces in the hexagonal arrangement of the blocks assume orders of magnitude, as a result of the substantial variation of parameters in the presence of which the supporting floor can no longer perform its function (vertical support of fuel elements, venting of fuel elements from the core, conduct of gas, shielding). A decisive factor in this respect is the flow behavior of fuel pebbles through the reactor core. A balanced flow behavior requires the availability of several pebble removal tubes, for which conical pebble inlets must be provided. Other parameters to be considered are stresses generated by dead weight, pressure gradients and the forces of absorber rods, also thermal expansions occurring with large dimensions of the core. All of these must be controlled. SUMMARY OF THE INVENTION Beginning with the so-called state-of-the-art, the present invention provides a support structure for the core of the high temperature gas cooled reactor, which guarantees the capability of the support structure to perform the functions of said structure listed in the foregoing independently of the performance of the structure of the core. The present invention also furnishes solutions of the thermal and mechanical problems encountered in may prior art devices. According to the invention, the support structure consists of several layers of prismatic graphite blocks arranged over each other. The layers are constructed as closed units without expansion gaps and the blocks of one layer are keyed together with the blocks of the adjacent layers. The upper layers are composed of a plurality of preferably hexagonal graphite blocks equipped with passages for the cooling gas, while the bottom layer is formed by a number of support structures, each consisting of several support segments fitted together preferably into a hexagonal cross section, with each support unit resting at its central section on a column head of a round column and carrying a limited number of the hexagonal graphite blocks and that cooling gas channels are provided at the locations of the bottom layer where three support units meet. The support device of the invention represents a stable and rigid support structure which satisfactorily performs all functions, such as supporting, pebble removal, gas conduction and shielding. It forms a closed unit without expansion gaps which adapts itself to the thermal cycles of the reactor core. As the result of the interkeying of the prismatic graphite blocks in the form of a layer structure, thermal movements are in the support structure, and are passed on to the adjacent blocks so that summation of gaps can occur. Correspondingly, the entire support structure and the annular lateral reflector expand in the radial direction. Because of the interkeying of the prismatic blocks, no relocating forces are necessary to equalize the bottom configuration. The support structure is, therefore, capable of adapting itself to any reactor state. Thermal expansion and supporting forces in the radial direction are equalized by way of a suitable lateral support of the side reflector. In the design of the prismatic graphite blocks, material properties are taken specifically into consideration, so that even in the case of an external disruption, e.g., by an earthquake, no tensile stresses can be transmitted in the support structure. The support structure according to the invention is largely independent of the capacity and dimensions of the core. It may be applied with special advantage to high temperature reactors with spherical fuel elements. However, with slight modification, it may also be used in high temperature reactors with block-shaped fuel elements. In the process, the structure facilitates particularly the solution of thermal problems. The support structure according to the invention offers an additional advantage by the fact that it renders the solution of problems associated with design difficulties in the bottom area of high temperature reactors and with the support of the side reflector. Thus, no roller bearings are required for the bottom area and only slight forces, or none at all, are generated in the side reflector. The upper layers of the prismatic graphite blocks are designed with respect to their height so that a conical inlet for the spherical fuel elements is formed for every pebble removal tube. Support units located in the bottom layer are designed in the area of the pebble removal tubes with respect to their cross section so that they surround the pebble removal tubes without gaps. There is a suitable adaptation in the boundary zone toward the side reflector. The support structure is preferably constructed of three layers. In the upper and intermediate layer, a central graphite block is surrounded by six graphite blocks and is aligned with one of the round columns. The peripheral graphite blocks are coordinated in the process always with three different central graphite blocks, i.e., each peripheral graphite block simultaneously belongs to the border of three central graphite blocks. Conveniently, in all of the hexagonal graphite blocks of the upper layer, a plurality of small vertical borings is provided for the cooling gas, said borings being connected with the collector spaces found in the graphite blocks of the intermediate layer. The collector spaces of the peripheral graphite blocks are designed in the form of continuous borings, aligned with the cooling gas channels present in the bottom layer. The collector spaces provided in the central graphite blocks, representing sack-like borings. They are connected with the continuous borings in the corresponding peripheral graphite blocks through several connecting borings. Further, gas flow takes place through the cooling gas channels of the bottom layer in a hot gas collector space located underneath the support structure. Underneath the side reflector, the hot gas collector space widens, which renders connection with radial gas conduits of a suitable diameter possible. Between the support structure and the side reflector, a continuous vertical separating gap may be provided advantageously, so that differential thermal expansion (because of temperature differences) in the vertical direction is possible. |
abstract | Techniques and compositions are provided for shielding radioactive energy. The composition includes a hydrocarbon component and a radiation shielding and absorbing material or additive. The composition may be applied to substrates or to radioactive materials. Moreover, the composition may be mixed with raw materials of products. |
|
050776851 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The detailed study of the radioactive ray control system disclosed in the above-mentioned Japasese Patent Application, JP-A-60-120295 (1985), by the inventors has revealed that it is not possible to prepare procedures for the target operation which satisfies a determined value of the radioactivity exposure amount by the control system. That is, in the Patent Application, when the procedures of the target operation for this time is different from the procedures of the previous operation, the radioactivity exposure amount is estimated taking the prepared evaluation factors into account. In such an estimation of the radioactivity exposure amount, it is necessary to decide the evaluation factors of work operation appropriately. However, it is difficult to decide the evaluation factors of operation appropriately. Thus the expected total radioactivity exposure amount which has been estimated contains a large amount of errors and it is not easy to prepare an appropriate procedures for the target operation. Accordingly, inventors investigated several kinds of methods for obtaining proper work operation procedures, and arrived to the present invention. Hereinbelow, embodiments of the present invention are explained. An apparatus for supporting the operation under radioactive rays according to one embodiment of this invention will be explained. FIG. 2 shows the constitution of the apparatus for supporting the operation under radioactive rays according to this embodiment. An operation processing arrangement 5 has an operating portion 5a, a processing procedure storing portion 5b, input portions 5c and 5e, an output portion 5d and 5g, and storing portion as an internal memory 5f. The processing procedure storing portion 5b stores the processing procedure shown in FIG. 1. An input unit (keyboard) 4 is connected with the input portion 5c. A storage unit 6 is connected with the input portion 5c and output portion 5g. The storage unit 6 stores the graphic data of a state, in which the equipments and pipings of a nuclear power plant are laid out, in a certain storing area. The storage unit 6 stores the surface radioactivity rates of the above-mentioned equipments and piping in another storing area. The layout graphic data and the data of the surface radioactivity rates may be stored in separate storage units. An image data storage unit 3 stores the information such as graphic information, etc., outputted from the output portion 5d of the operation processing arrangement 5. An image displaying control unit 2 displays the information stored in the image data storage unit 3 on displaying means (display) 1. The processing procedures in FIG. 1 stored in the processing procedure storing portion 5b of FIG. 2 are successively called into the operating portion 5a and executed. The content of the processing in this embodiment will be explained below referring to the processing procedure of FIG. 1. In the first place, the input of layout data (step 10) is carried out. the detailed content of the processing of step 10 is shown in FIG. 3. The operator input data of an area within the nuclear power plant, in which the operation requiring for studying the radioactivity exposure amount (target operation) is carried out, from the input unit 4. For example, the name of the nuclear power plant "X", the name of the plant housing "R" and the name of the area "R5B" are inputted. These data are entered into the operating portion 5a (step 10A). The inner part of the housing in which the nuclear power plant is disposed is an area with controlled radioactive rays. The graphic data for the area concerning to these inputted data (area with controlled radioactive rays), in which a plurality of structures (referred to in general for equipments, pipings, walls of the housing, shields from radioactive rays, heat insulators and supports, etc.) are laid out, is searched from the storage unit 6 and entered (step 10B). The structure means the constituting elements of the plant dealing with radioactive substances, the constituting elements of the housing containing the plant, and the objects provided in the housing along with the installation of the plant. These layout graphic data are generated by use of a CAD system not shown. The generation of layout graphic data is made e.g. as described in Japanese Patent Application (kokai) 62-114063 (from p.24, upper-right column, line 9 to p.30, upper-right column, line 10). In step 10A, the structures to which the target operation is applied or the name of the target work operation may be inputted, instead of the area with controlled radioactive rays in which the target work operation is carried out. In this case, in step 10B, the layout graphic data of the area with controlled radioactive rays containing the designated structures or the structures contained in carrying out the designated operation is searched. Thus anyway in step 10, the designated layout graphic data is searched. In step 11, the surface radioactivity rates of the structures are calculated. The detailed processing in this step is shown in FIG. 5. The surface radioactivity rates for the structures disposed in the area with controlled radioactive rays (designated area) are searched from the storage unit 6 and entered (step 11A). In step 11B, it is determined as to whether there is a request for the calculation of the surface radioactivity rates based on the measured values by a radioactive ray detector or not. This request is entered by the operator via the input unit 4 after the processing in step 11A has been completed. Of course, the guidance page is displayed for deciding as to whether the request is needed or not. If there is no such request, the processing in step 11C is initiated. If there is such a request, the processing in 11D is initiated. In step 11C, the increased amounts of the surface radioactivity rates of the structures are calculated on the basis of the time which elapsed after the previous regular inspection of the nuclear power plant had been completed. The increased amounts of the surface radioactivity rates vs. time that elapsed can be obtained according to the empirical formula based on the statistic operation of the data in the past. The reason of taking the previous regular inspection as the basis is that the data of the surface radioactivity rates in the storage unit 6 are corrected by the measured values of radioactive rays at the time when the regular inspection has been completed. If such correction is not carried out, the surface radioactivity rates are obtained by the empirical formula based on the tendency of the data in the past on the nuclear power plant operation experience. In step 11D, the radioactivity amount measured by a radioactive ray detector disposed in the designated area is entered. In the next place, the surface radioactivity rates of the equipments and pipings at the present time are calculated (step 11E). If the processing in the step 11C has been made, the surface radioactivity rates at the present time R.sub.p are obtained by adding the increased amounts of surface radioactivity rates obtained in step 11D to the surface radioactivity rates entered in step 11A. If the processing in step 11D has been made, the surface radioactivity rates R.sub.p at the present time are obtained by use of the measured values of the radioactivity amounts. The surface radioactivity rates R.sub.p thus obtained are entered into the storage unit 5f (step 11F). Step 12 is executed after step 11 has been completed. FIG. 6 shows the detail of step 12. The graphic data entered in step 10B are converted into graphic information for displaying (step 12A). The color information corresponding to the levels of the surface radioactivity rates R.sub.p for the equipments or pipings obtained in step 11 is selected. The selected color information is added to the graphic information of the corresponding equipments or pipings (step 12B). The color information is predetermined depending on each level of the surface radioactivity rate. For example, green is selected for a surface radioactivity rate R.sub.p of low level, yellow for a surface radioactivity rate R.sub.p of middle level and red for a surface radioactivity rate R.sub.p of high level. The graphic information added with color information is outputted into the image data storage unit 3 (step 12C). This graphic information is outputted into the display unit 1 through the image display and control unit 2 and displayed on the display unit 1. FIG. 7 illustrates an example of the image displayed on the display unit 1. Here, the equipment A.sub.1 is displayed with green lines, because the surface radioactivity rate R.sub.p is of low level. The wiring A.sub.2, the surface radioactivity rate of which is of middle level, is displayed with yellow level. Besides, A.sub.3 is a wall. The graphic of the paths, along which workers can go, are also displayed, although they are not depicted in the picture of FIG. 7. The operator can know the surface radioactivity rate for each portion of the actual nuclear power plant by seeing the graphics of the nuclear power plant displayed on the display unit 1 as shown in FIG. 7. Particularly, it is described in Japanese Patent Applications Kokai (Laid-Open) Nos. 59-54985 (1984) and 61-212782 (1986) that the surface to be measured is divided into small sections based on the measured value by a radiation detector and the color information corresponding to the strength of radioactive ray is alloted for each small section. However, because the layout graphic data of the structures in the housing are used in this embodiment, the radioactivity amount of each structure, which is laid out actually, can be seen. In the next place, the generation of simulation data of movement is carried out (step 13). The details of step 13 are shown in FIGS. 8 and 9. The name of the target work operation carried out in the area designated in step 10A is inputted (step 13A). For example, "Strainer exchange operation of the system so-and-so" is inputted as the name of the target operation. When the name of the target operation is inputted in step 10A, the processing in step 13A is not required. The work operation area in which the target operation is carried out is inputted (step 13B). The work operation area is a particular narrow part contained within the above-mentioned designated area. An access route (passage), along which workers move in order to reach from the entrance of the housing R to the work operation area, is inputted (step 13C). The input of this access route is made by successively displaying the graphic data containing the passage (searched from the storage unit 6), along which workers move in order to reach from the entrance of the building R to the operation area, on the display unit 1. This input of the work access route will be explained on the basis of FIG. 10. FIG. 10 shows the graphic data displayed on the display unit 1, a part of the data which are read out from the storage unit 6. In FIG. 10, R.sub.0 is the path along which workers move and B is the work operation area designated in step 13B. In this, the graphics of equipment and pipings are omitted in FIG. 10. The operator inputs the access route of the workers, e.g. an access route P.sub.1 .fwdarw.P.sub.2 .fwdarw.P.sub.3 .fwdarw.P.sub.4 .fwdarw.P.sub.5 .fwdarw.P.sub.6, from the input unit 1 (or with a light pen), looking at the screen on which the graphics of FIG. 10 are displayed. If there is another access route to be considered (P.sub.1 .fwdarw.P.sub.2 .fwdarw.P.sub.7 .fwdarw.P.sub.4 .fwdarw.P.sub.5 .fwdarw.P.sub.6, in FIG. 10), also the access route is inputted. Because the information of the surface radioactivity rates of equipment and pipings is added to the graphic information, the operator can grasp generally an access route with a higher radioactivity exposure amount and an access route with a lower radioactivity exposure amount, looking at the screen of the display unit 1. Therefore setting (inputting) of an access route with increasing radioactivity exposure amount can be avoided. This allows the time required for simulation of the radioactivity exposure amount when there are a plurality of access routes to be reduced. In the next place, the procedures and instruction for the target work operation are inputted (step 13D). The procedures and instruction for the exchange operation of the strainer entered in step 13A are shown in FIG. 11. The target system by such a work operation is shown in FIG. 12. In FIG. 12, 50 is a tank, 51 is a strainer, 52 is a valve and 53 is a pump. the strainer 51 has flanges at both ends. These flanges are detachably secured to the flange of the pump 53 and to the flange attached to the valve 52 by means of bolts respectively. The valve 52 is connected with the piping attached to the tank 50. Although not shown in FIG. 11, the action of going to the work operation area along the above-mentioned access route in order to carry out the exchange operation of the strainer shown in FIG. 11 and the action of returning from the work operation area to the entrance of the housing after having finished the determined operation are ones included in the operation procedures, respectively. A few of representative configuration graphic data of the equipments and pipings which are a target of the work operations during detaching and attaching operation are entered (step 13E). For the strainer exchange operation, the configuration graphic data of the state in which the strainer 51 is taken away from the graphic data of FIG. 12 (which is prepared by a CAD system and stored in the storage unit 6) is inputted. In the case of disassembly of equipment, the data denoting a few of representative figuration graphics of the equipments in the midway of disassembly and after completed disassembly are inputted. When equipments for inspection (for example, a supersonic flaw detector) are to be attached to the equipments or pipings, a few of the figuration graphic data in the midway of attachment are inputted. The input of each data in steps 13A.about.13D is urged upon the operator by displaying the guidance on the display unit 1 requiring the data input before each step. These data are entered by use of the input unit 4. Data of displacement state of objects at the time of their detachment or attachment, which are contained in an installation (for example, a plant, a system) and are required to be detached or attached (hereinafter, referred to as target displacement portions, concretely e.g. the strainer 51) are generated (step 13F). That is, the data of displacement of the strainer 51, which moves corresponding to the proceeding of the operation, is generated. It is determined as to whether auxiliary units are necessary or not by use of the data of the procedures or instruction of the work operation (step 13G). The auxiliary units are instruments and devices used when the objects required to be detached or attached are moved. In the example of FIG. 11, the chain block and pulling cart serve as auxiliary units. If the determination in step 13G is "YES", step 13H and step 13I are executed. The data of the representative figuration graphics of the auxiliary units are inputted (step 13h). The data of the representative figuration graphics of the chain block and pulling cart to be inputted by the operator through the input unit 4 are inputted. This data input is also urged by the above-mentioned displaying of the guidance requiring the data input on the display unit 1. In step 13I, the data of displacement of the auxiliary units during work operation are generated by utilizing the data inputted in step 13H (step 13I). That is, the data of displacement of the chain block and pulling cart are generated. In the next place, translation to step 13J is made. If the determination in step 13G is "NO", step 13J is executed. In step 13J, the approximation graphic of a worker is generated. The approximation graphic 20 is one as shown in FIG. 13. The approximation graphic of a worker is shown by 35 in FIG. 30 of U.S. patent application Ser. No. 929,894 (filed on Nov. 13, 1986), and as an example the coordinate such as the formula (4) on p.27, of the above U.S. patent application is given. The approximation graphics of workers are generated by the number of workers necessary for carrying out the target work operation. The number of necessary workers is determined by the operator on the basis of the procedures and instruction for the target work operation. In the next place, the movement simulation data for approximating the movement of workers are generated for respective worker approximation graphics (step 13K). When the target work operation is to be carried out by a plurality of workers, the tasks contained in the target work operation are assigned separately and the task assigned is decided for each worker. Consequently, the movement simulation data of each worker approximation graphic is generated based on the movement of the corresponding worker carrying out the assigned task. The movement simulation data for one worker approximation graphic includes the figuration data of the worker approximation graphic approximating the movement of a worker during his movement along a certain access route R.sub.0 and the figuration data of the worker approximation graphic approximating the movement of a worker while he is carrying out the task assigned within the target work operation. If the procedures and instruction of the work operation in FIG. 11 are carried out, the figuration data of the worker approximation graphic while he carries out the task assigned within the target work operation is generated. The processings in steps 13F, 13I, 13J and 13K are carried out interactively between the apparatus for assisting the operation under radioactive rays according to this embodiment and the operator via the display unit 1 and input unit 4. The data generated and inputted in each of the above-mentioned steps are stored in the storage portion 5f (step 13L). The location of the radioactivity exposure monitor within the each worker approximation graphic is inputted (step 13M). This set location is designated by the operator with the input unit 4. The set location may be of a plural number for one worker approximation graphic. The existence of a plurality of locations means that the radioactivity exposure monitors are disposed at respective designated locations, e.g., at the head, breast and waist part, etc. In step 13N, the respective data obtained in steps 13F and 13K are combined and the simulation data S.sub.0, which moves the respective worker approximation graphics and the graphics of target displacing portion (containing the graphics of auxiliary units, if they are used) and which includes a series of movements through the whole target work operation, are generated. The simulation data S.sub.0 are generated by combining the respective data of the movement of a worker when he goes from the entrance of the housing R to the work operation area B through the path R.sub.0, the movement of the worker in carrying out the task assigned within the target work operation, the movements of the target displacing portion and auxiliary units in relation to the task assigned, and the movement of the worker when he returns from the work operation area B to the entrance of the housing R, taking the procedures and instruction of the target work operation into account. The data of the velocities of displacement and movement of each worker approximation graphics at the time of simulation are generated (step 13P). These velocity data are generated by use of each velocity data for an actual worker inputted by the operator with the input unit 4. The displacing velocity data of the above-mentioned target portion to be displaced and auxiliary units are also generated by use of the velocity data at the time of above-mentioned associating simulation. In order to reduce the time for simulation, each velocity at the time of simulation is remarkably short compared with each velocity of an actual worker. For example, each velocity is decided so that the simulation may be completed in 20.about.30 minutes. However, it is also possible to decide each velocity so that the time for simulation may be shorter, or longer, than the above-mentioned time. If there are a plurality of workers for target operation, in order to reduce the waiting time within the work operation area B, the start time from the entrance of the housing R (the movement start time of the worker approximation graphic) is decided in different times corresponding to the tasks assigned to the workers. This decision is made by use of the input unit 4. The data generated and inputted in steps 13M, 13N and 13P are stored in the storage portion 5f (step 13Q). The processing in step 13 is completed by the heretofore mentioned. In step 14, the simulation of the radioactivity exposure amount is carried out. In step 14, the radioactivity exposure amount is calculated at the location of the radioactivity monitor inputted in step 13M. The details of step 14 are shown in FIGS. 14 and 15. In the first procedure of step 14, the data of the worker approximation graphic, the surface radioactivity rate R.sub.p and the simulation data for each worker approximation graphic S.sub.0 are read from the storage portion 5f (step 14A). In step 14B, the worker approximation graphic is converted into the graphic information on the basis of the data inputted in step 14A for displaying graphics moving at the velocity generated in step 13P. This graphic information is outputted into the image data storage unit 3 along with the graphic information for equipments, pipings, walls of housings, shields from radioactive rays and paths R.sub.0 generated in step 12A. Consequently, the state in which a worker approximation graphic moves along a path, a determined access route at the above-mentioned displacement velocity is displayed on the screen of the display unit 4. The worker approximation graphic, of which the start time from the entrance of the building R is different, starts movement when the time determined in step 13P has come. At the time of moving along the access route, the radioactivity exposure amount ED.sub.1 of a worker is obtained for each part of the access route (step 14C). Each part of the access route means each small section obtained by subdividing the access route into sections of a determined width along its longitudinal direction. Because the moving velocity of each worker approximation graphic moving along the access route is identical, the radioactivity exposure amount ED.sub.1 is obtained by use of a certain worker approximation graphic. That is, the radioactivity exposure amount ED.sub.1 is one at the decided location entered in step 13M and is obtained for each small section on the basis of the surface radioactivity rate R.sub.p of equipments and pipings. This radioactivity exposure amount ED.sub.1 has a relation also with the time T.sub.1 required by a worker approximation graphic for moving through a small section. The longer the time T.sub.1 is, the larger the radioactivity exposure amount ED.sub.1 becomes, and adversely the shorter the time T.sub.2 is, the smaller it becomes. The time T.sub.1 is obtained according to the moving velocity of the worker approximation graphic during simulation. The time T.sub.2 required by an actual worker for moving through a small section is extremely long compared with the time T.sub.1. Consequently, the radioactivity exposure amount ED.sub.1 is calculated as corresponding to the time T.sub.2 using the equation relating the time T.sub.1 and the time T.sub.2. Further, if there is a wall of the housing R or a shield from radioactive rays between the location of a worker approximation graphic and the equipment or pipings giving radioactive effects on the location, the radioactivity exposure amount ED.sub.1 is calculated taking the shielding effect from radioactive rays according to this into account. It is determined easily whether there is a wall or shield or not, because the data of the laid out graphics created by a CAD system are entered in step 10B. After the radioactivity exposure amount ED.sub.1 have been calculated for all the small sections along an access route, the total radioactivity exposure amount ED.sub.2 during the movement along the access route is calculated (step 14D). That is, the total radioactivity exposure amount ED.sub.2 corresponds to the sum of the radioactivity exposure amount ED.sub.1 of all the small sections. The time T.sub.0 required by a worker for moving from the entrance of the housing R to arrival at the work operation area B is obtained (step 14E). The time T.sub.0 is obtained by summing the time T.sub.2 for all the small sections. In step 14F, it is determined whether there is another access route decided in step 13C. If this determination is "YES", one of the other access routes is selected and the processings in steps 14B.about.14F are repeated. When the determination in step 14F is "NO", the processing in step 14G is executed. In step 14G, the access route for which the total radioactivity exposure amount ED.sub.2 is the minimum is selected. Each worker approximation graphic is converted into graphic information for displaying the figuration in which the task assigned is performed at the movement velocity created in step 13P on the basis of the data entered in 14A (step 14H). Further, in step 14H, the graphics of target parts to be moved and auxiliary units are also converted into graphic information for displaying on the basis of data entered in step 14A. Here, those obtained graphic information is outputted along with the graphic information of equipments, etc., similarly as in step 14B. According to this, the movement of each worker, the movements of the target parts to be moved and auxiliary units are displayed as movements of each worker approximation graphic, the approximation graphics of the target parts to be moved and approximation units according to the order of the work procedures of the operation on the screen of the display unit 1. A series of movements of the respective approximation graphics according to the procedures of work operation when the target work operation is carried out is generated because the necessary data are created on the basis of the corresponding procedures and instruction of work operation in step 13N. That is, because the data are created by rearranging the movements of the respective worker approximation graphics for the tasks assigned along with the steps of the procedures of work operation (FIG. 11, No. 1.about.7), a series of movements in the target operation can be approximated. The steps in the procedures of work operation is called work operation procedure steps. The radioactivity exposure amount ED.sub.3 of each worker for each work operation procedure step is calculated within the task assigned of the corresponding worker (step 14I). The calculation of the radioactivity exposure amount ED.sub.3 is made similarly as the calculation of the radioactivity exposure amount ED.sub.1. Of course, the effects of the walls of housing and the shields for radioactivity are taken into account. However, the time T.sub.4 required actually by a worker for carrying out a task related with a work operation procedure step within the task assigned is longer than the time T.sub.3 required for the simulation carrying out the task. Consequently, the radioactivity exposure amount ED.sub.3 is calculated as corresponding to the time T.sub.4 using the equation relating the time T.sub.3 and the time T.sub.4. The time T.sub.3 is obtained according to the movement velocity decided in step 13P. The time T.sub.5 required in each work operation procedures step during simulation is also obtained according to the above-mentioned movement velocity. The time T.sub.1 required actually for the work operation procedure step is obtained as a function of time T.sub.5. If the task assigned for a worker includes two work operation procedure steps between which there is a work operation procedure step to be carried out by another worker, the time for which the other worker is carrying out the work operation procedure step of his own task assigned results in waiting time for the former worker until he begins to carry out the next task assigned. In step 14J, the radioactivity exposure amount ED.sub.4 in such a waiting time is calculated. The waiting time for the work operation is obtained from the equation relating the time T.sub.3 and the time T.sub.4 using the time T.sub.3 required by the other worker for carrying out the task. The total radioactivity exposure amount ED.sub.5 during the carrying out of the target work operation is calculated for each worker (step 14K). The total radioactivity exposure amount ED.sub.5 amounts to the sum of all the radioactivity exposure amounts ED.sub.3 and ED.sub.4 for a worker. In step 14L, time relating to performing a target work operation is obtained. Specifically, step 14L has two steps of processings. In the first step of processing, the time T.sub.s during which a worker remains within the work operation area B until the task assigned is completed is obtained for each worker. This time T.sub.s calculated by summing all the time T.sub.4 used for performing the task assigned within the operation area B and the waiting time for work operation for each worker. In the second processing step, the time T.sub.M elapsing from the start of the target work operation to its completion is calculated. This is calculated by summing the time T.sub.i required for each work operation procedure step. The sum ED.sub.6 of the radioactivity exposure amount received by a worker in coming and going along the access route and the radioactivity exposure amount received by him in carrying out the target work operation is calculated for each worker (step 14M). The latter radioactivity exposure amount is the radioactivity exposure amount ED.sub.5. The former radioactivity exposure amount is twice the radioactivity exposure amount ED.sub.2 at the time when movement through the access route with the radioactivity exposure amount ED.sub.2 being minimum. This is because the radioactivity exposure amount are equal for coming and going paths. Each processing in step 14 heretofore described corresponds, as it were, to the measurement of the radioactivity exposure amount when a worker carries out the target work operation by use of a radioactivity exposure amount monitor provided at the location designated in step 13M. That is, the radioactivity exposure amount monitor measures the radioactivity exposure amount while the worker approximation graphic with the radioactivity exposure amount monitor provided at the set location is displacing and moving in order to carry out the target work operation. The data of each radioactivity exposure amount and each time obtained in each step heretofore mentioned are stored into the storage unit 5f. Further, the data of each radioactivity exposure amount, access route with the minimum radioactivity exposure amount and each time obtained in each of heretofore mentioned steps are outputted into the image data storage unit 3 (step 14N). The information for these is displayed on the display unit 1. According to this, the operator can know the access route with the minimum radioactivity exposure amount, the radioactivity exposure amount in carrying out the target work operation and each required time. These information is extremely useful in planning the schedule of work operation under radioactive rays. That is, the planning of the schedule of work operation under radioactive rays (process charts of work operation, man-Rem stack chart and instruction of work operation procedures) becomes extremely easy. Because the movement of each worker is displayed on the display unit 1 at the same time with the simulation of the movement of each worker in the target work operation and the calculation of the radioactivity exposure amount, the operator can know a series of movements of each worker in the target work operation while the radioactivity exposure amount is calculated. Further, because the radioactivity exposure amount is obtained by simulating the actual displacement and movement of a worker within the area with controlled radioactive rays, the obtained radioactivity exposure amount is with extremely high precision. Particularly, because the radioactivity exposure amount received by a worker approximation graphic is calculated by use of the figuration data of the structures laid out in the area with controlled radioactive rays and the surface radioactivity rates of the structures, the actual distances between the worker and structures in the actual area with controlled radioactive rays and the actual relation between their locations can be reflected, so that the calculated radioactivity exposure amount are of higher precision. It is determined whether the result obtained by the processing in up to step 14 satisfies the condition or not (step 15). The detail of this step is shown in FIG. 16. In step 15, it is determined whether the target work operation will be completed within a predetermined working hours (step 15A), whether the man-Rem of the target work operation is OK' d or not (step 15D), and whether the radioactivity exposure amount of each worker is below the maximum allowable radioactivity exposure amount or not (step 15G). In this, each worker leaves the work operation area at the time when the last of his own tasks assigned within the work operation area has been completed. That is, if there is no task to be carried out in another area, each worker returns to the entrance of the housing R through the access route with the minimum radioactivity exposure amount. In step 15A, the time required for coming and going along the access route (twice of T.sub.0) and the time T.sub.M are summed in the first place. In the next place, it is determined whether the time obtained by this summation is within a predetermined working hours (e.g. eight hours a day) or not. If the determination is "NO", translation to step 15B is made. In step 15B, the guidance indicating to decide the range of the target work operation to be completed within the predetermined working hours is outputted into the image data storage unit 3. This guidance is displayed on the display unit 1. In step 15B, the required times T.sub.i and T.sub.0 for each work operation procedure step are outputted along with this guidance. These times are displayed on the display unit 4. The operator enters the range of the target work operation through the input unit 4, seeing the guidance. This range of the target work operation is entered into the operating portion 5a (step 15C). The operator can easily decide the range of the target work operation to be completed within the predetermined hours, as the times T.sub.1 and T.sub.0 are displayed on the display unit 1. That is, after determining the time obtained by subtracting the time required for coming and going the access route from the predetermined working hours it becomes extremely easy to select the work operation procedure steps allowed to be carried out in a day on the basis of the time Ti required for work operation procedure step. Further, because the operator can know the time required for coming and going along the access route and the time T.sub.i required for each work operation procedure step, it is possible to calculate approximately how many days are required for carrying out the target work operation. Consequently, the range of work operation can be divided for each day so that the work operation is completed within the approximated days. These ranges of the target work operation are inputted in step 15C. Thereafter, the processings in steps 13.about.14 and 15A, which are applied to the decided rang of target work operation, are repeated. If the determination in step 15A is "YES", translation to step 15D is made. In step 15D, the value of man-Rem is calculated which can be obtained by further summing the radioactivity exposure amount ED.sub.6 of respective workers calculated in step 14M. The man-Rem means the summed value of the radioactivity exposure amount (referred to as total received radiactivity) which all the (a plurality of) workers receives while they are in the area with controlled radioactive rays (e.g. the housing of a nuclear power plant). As mentioned above, because the precision of the radioactivity exposure amount calculated in step 14 is high, the total radioactivity exposure amount obtained for the target work operation is also with high precision. The total radioactivity exposure amount includes also the radioactivity exposure amount during the movement along the access route. When the value of the calculated total radioactivity exposure amount satisfies a predetermined condition, translation to step 15G is made. When the value does not satisfy the predetermined condition, translation to step 15E is made. There are two ways, as follows, for determining whether the calculated value of the total radioactivity exposure amount (man-Rem) is OK' d or not. In this embodiment, either of the ways may be employed. That is, in the first way, a predetermined value of man-Rem and the calculated value of man-Rem are compared, and it determined as OK' d when the latter value is below the former value. In the second way, the value of man-Rem is not determined, the calculated value of man-Rem is displayed on the display unit 1 and it is determined that the apparatus according to this embodiment is OK' d, that is, the calculated value of man-Rem satisfies a predetermined value, in response to the signal of "OK' d" inputted by the operator with the input unit 4. In step 15E, the factor increasing the calculated value of man-Rem, that is, the information of the locations, for which radioactivity prevention countermeasures should be taken, is created and the information is outputted into the image data storage unit 3. The information is displayed on the display unit 1 along with the information outputted in step 12C. The information created in step 15E is created by combining the radioactivity exposure amounts up to the predetermined order (for example, from the radioactivity exposure amount with the maximum value to one with the third value) out of the radioactivity exposure amounts for respective work operation procedure steps (including also the procedure in which a worker moves along the access route when he goes to the work operation area and when he returns from the work operation area) with the corresponding work operation procedure. These radioactivity exposure amount have been obtained in step 14D and step 14I. The operator can know the operation procedure increasing the total radioactivity exposure amount (man-Rem) due to carrying-out of the target work operation by looking at the screen of the display unit 1 on which the information created in step 14S is displayed. In this, when the radioactivity exposure amount are obtained in step 14I not only for respective work operation procedures but also for respective small sections into which the area in which a worker moves for carrying out the work operation procedures is divided as in step 14C, the information including the radioactivity exposure amount in the predetermined number of small sections and the corresponding small sections is created in step 15E. It is a matter of course that the information for the radioactivity exposure amount in the small sections of the access route calculated in step 14C is also created. If the information employing the radioactivity exposure amount in the small sections is to be created in step 15E, it may be created as follows. That is, the color information corresponding to the levels of the calculated radioactivity exposure amount is added to the above-mentioned small sections of the graphic information obtained in step 12A. The levels are provided, for example, in ten grades, and individual colors are assigned for each level. Such color information of radioactivity exposure amount is added to the graphic information obtained in step 12A and displayed along with this on the display unit 1. Also the color information obtained in step 12B is displayed on the display unit 1. According to this, the operator can easily know the small sections, as man-Rem is increased and also can recognize easily the radioactive ray sources with high radioactivity rates (structures) affecting thereon. Consequently, the operator can easily find countermeasures to decrease man-Rem. It is also possible to display the information of the above-mentioned radioactivity exposure amount of the work operation procedures along with the information in steps 12A, 12B. It is also allowed to display the information on radioactivity exposure amounts of all the work operation procedures or of all the small sections. The operator devises countermeasures for decreasing man-Rem seeing the information created in step 15E. Then the operator enters the countermeasures through the input unit 4. The countermeasures are entered into the operating portion 5e (step 15F). Thereafter, the processings after step 13 reflecting the countermeasures are repeated. Countermeasures and processing based on the countermeasures will be explained below. For example, supposing that each work operation procedure in Nos. 2.about.7 of FIG. 11 is inputted as the exchange operation of the strainer in step 13D, in the first place. Each processing in steps 13E.about.13Q, 14A.about.14N, 15A and 15D is executed. In step 15D, it is determined as "NO". In step 15E, the radioactivity exposure amount of a worker in the procedures of Nos. 3.about.6 becomes higher, because the tank 50 is filled with liquid generating high radioactivity. Because the displaying on the display unit 1 is made by use of a color indicating that the tank 50 is with a high level of surface radioactivity rate, the operator recognizes that the tank 50 affects the radioactivity exposure amount in the procedures of Nos. 3.about.6. For this, the operator conceives carrying out the work operation procedures of No. 1 in FIG. 11 before the work operation procedures of No. 2 as countermeasures, and the countermeasures are entered in step 15F. The radioactivity exposure amount is simulated in the state reflecting the countermeasures. Because the radioactive liquid within the tank 50 is discharged by the work operation procedure of No. 1, the radioactivity exposure amount in the work operation procedure of Nos. 3.about.6 is decreased below the value obtained at the previous time. Supposing even in this case that displaying is made again that the radioactivity exposure amount in the work operation procedures Nos. 2.about.7 is high. The operator conceives provision of a portable shield from radioactive rays temporarily located at the side facing to the strainer 51 of the tank 50 as countermeasures against the remaining radioactive liquid in the tank 50, seeing the graphics in FIG. 12 displayed on the display unit 1. For this purpose, the operator enters a work operation procedures of provision of a portable shield from radioactive rays at the side facing to the strainer 51 of the tank 50 between the procedures of No. 2 and No. 3 in FIG. 11 through the input unit 4. The operating unit 5a repeats each of the above-mentioned processings for the further countermeasures. When the determination in step 15D has become "YES" due to the execution of these processings, translation to step 15G is made. In this, when the determination of "YES" is made in step 15D, all the operation procedures of the target work operation in which the determination could be obtained are stored into a predetermined area of the storage unit 6. This stored information is searched and entered in step 13D when the work operation of the countermeasures has been entered in step 13A at the time of the next inspection. In step 15G, it is determined whether the radioactivity exposure amount of each worker received during carrying out of the target work operation is below the maximum allowable radioactivity exposure amount or not. The maximum allowable radioactivity exposure amount are decided for a day, a week, three months and a year. If the condition is not satisfied for any of the maximum allowable radioactivity exposure amount, the determination in step 15G becomes "NO". A new worker, for whom the condition for the maximum allowable radioactivity exposure amount is satisfied even when the radioactivity exposure amount of another worker determined as "NO" in step 15G is added, is selected (step 15H). The radioactivity exposure amount of each worker is stored in storage unit 6. If the determination in step 15G is "YES", step 16 is executed. In step 16, all the work operation procedures of the target work operation which satisfy the predetermined values of man-Rem in step 15D are outputted into the image data storage unit 3. These operation procedures are displayed on the display unit 1. Further, these operation procedures are printed by the printer 7 according to the request from the operator. The printer 7 is also a kind of display means. Also other information obtained by the operating portion 5a can be outputted through the printer 7 according to the request from the operator. Further, in step 16, the task assigned carried out by each worker can be outputted when the predetermined value of man-Rem has satisfied in the target work operation. Also this is displayed on the display unit 1. According to this embodiment as mentioned heretofore, it is easy to create operation procedures in which the radioactivity exposure amount (particularly the total radioactivity exposure amount) satisfies the predetermined value in carrying out the target work operation. In particular, because the precision of the value of the calculated radioactivity exposure amount is high, appropriate operation procedures satisfying the predetermined value of the radioactivity exposure amount can be created. It is remarkably useful in planning for the number of workers for every kind of job required for carrying out the target work operation that a series of appropriate work operation procedures satisfying the value of the predetermined radioactivity exposure amount can be created for the target work operation. In this, if there are a plurality of target work operations carried out in the area with controlled radioactive rays, at least the processings in steps 13.about.16 are executed for each target work operation. FIG. 17 represents functionally the operation processing unit 5 executing the processing procedure in FIG. 1. It can be said that the operation processing unit 5 has means for generating structure body graphic information 5A, means for generating radioactivity information of structure body 5B, means for generating displacement data of target moving part 5C, means for generating work craftman movement data 5D, means for generating target work operation simulation information 5E, means for calculating radioactivity exposure amount 5F and means for deciding radioactivity exposure amount 5G. The means for generating structure body graphic information 5A executes the processings in steps 10A, 10B and 12A and generates the graphic information for displaying each structure disposed in the area with controlled radioactive rays designated by the input unit 4. This graphic information is outputted into the image data storage unit 3. The means for generating radioactivity information of structure body 5B, into which the layout graphic data entered by the means for generating structure graphic information in step 10 is inputted, and generates the radioactive ray information related to the structures contained in this layout graphic data. This creation of information is executed in steps 11A.about.11E, 12B and 12C. The means for generating displacement data of target moving part 5C executes the processings in steps 13A, 13B, 13D.about.13I. The displacement data generation means 5C generates the displacement data of the target moving part corresponding to the target work operation and, when required, of the auxiliary units through the processings in steps 13F and 13I, as mentioned above, on the basis of the layout graphic data outputted from the structure body graphic information generation means 5A and each work operation procedures for the target work operation. The workcraftman movement data generation means 15D executes the processings in steps 13A.about.13D, 13J and 13K. The target work operation simulation information generation means 5E executes the processings in steps 13M, 13N, 13P, 14B and 14H. That is, the target work operation simulation information generation means 5E, into which the graphic information obtained in structure body graphic information generation means 5A, the data obtained in the data generation means 5D and 5E, and the location of the radioactivity exposure amount monitor and each velocity data inputted from the input unit 4 are inputted, generates the graphic information indicating a series of movements of a worker within the area where the structures for carrying out the target work operation are laid out and the graphic information indicating the displacement of the target moving part, etc. Such information is outputted into the image data storage unit 3 along with the graphic information generated by the generation means 5A. The radioactivity exposure amount calculating means 5F executes processings steps 14B.about.14G and 14I.about.14N. The radioactivity exposure amount calculating means 5F calculates the radioactivity exposure amount at the location of the designated radioactivity exposure amount monitor, on the basis of the radioactive ray information of each structure in the designated area with controlled radioactive rays obtained by the radioactivity information generation means 5B and the graphic information obtained by the target work operation simulation information generation means 5E. Each of the calculated radioactivity exposure amount is outputted into the image data storage unit 3 and the radioactivity exposure amount deciding means 5G. The radioactivity exposure amount deciding means executes the processing in steps 15A.about.15F and 16. When the determination in step 15D is "YES", all the work operation procedures related to the target work operation used in data generation of the work craftman movement data generation means 5D by the processing in step 16 are outputted into the image data display unit 3. The data and information entered into the image data display unit 3 is displayed on the display unit 1, as explained for the processing procedures in FIG. 1. FIG. 18 illustrates another embodiment of the processing procedures in step 11 shown in FIG. 5 in detail. In the processing procedures in this embodiment, step 11F is removed from the processing procedures shown in FIG. 5 and step 11G and 11H are added anew following to step 11E. Only these different parts will be explained. In step 11G, the surface radioactivity rate R.sub.s is calculated for each small partitioned section obtained by subdividing the designated area, using the obtained surface radioactivity rate R.sub.p of the structures (equipments, etc.). In the next place, each of the obtained radioactivity rates R.sub.s is stored in the storing portion 5f (step 11H). When such a processing procedure as shown in FIG. 18 is used as step 11 shown in FIG. 1 instead of the processing procedure shown in FIG. 5, the surface radioactivity rates R.sub.p used in step 12 and thereafter are substituted by the radioactivity rates R.sub.s. Also when the processing procedure in FIG. 18 is used instead of step 11, similar effects as in the above-mentioned embodiment can be obtained. According to the first characteristic of this invention, appropriate operation procedures satisfying the predetermined radioactivity exposure amount can be obtained easily for the target work operation. According to the second characteristic of this invention, the operator can know easily the operation procedure or area for which receiving of radioactivity exposure should be restrained, when the predetermined radioactivity exposure amount is not satisfied. According to this, countermeasures for restraining the radioactivity exposure can be conceived simply. According to the third characteristic of this invention, the radioactive ray of each structure located in the area with controlled radioactive rays in which the target work operation is carried out can be known easily. In particular, because radioactivity amount information is displayed along with the actual wiring situation of each structure based on the layout graphic data, the radioactivity amount distribution in the actual located situation of the structures can be known. According to the fourth characteristic of this invention, the movement of the worker carrying out each work operation procedure of the target work operation can be recognized, while the radioactivity exposure amount of the worker in the target work operation is calculated. |
056493231 | summary | BACKGROUND OF THE INVENTION This invention relates to a composition and process for disposing of radioactive, hazardous and mixed wastes, and, in particular, to encapsulation and stabilization of low-level radioactive, hazardous and mixed wastes. Industrial wastes which are not economical to recycle must be disposed of in the environment. Some of this waste material can be rendered harmless and then disposed of in a convenient manner. Other wastes, however, such as heavy metals, for example, mercury, lead, antimony, arsenic, and radioactive substances, cannot be rendered harmless. Consequently, the disposal of these materials into the environment must be made in a manner in which they are stabilized against dispersion into the environment. Large volumes of low-level radioactive wastes are routinely generated through the operation of defense-related and commercial nuclear facilities. As a result of its defense and research activities, the Department of Energy (DOE) generates not only low level radioactive waste, but also hazardous and mixed waste. Hazardous and mixed wastes at DOE facilities include a broad range of waste types, such as evaporator concentrates, salts, blowdown slurries, sludges, filter materials, ion exchange resins, and incinerator ash, that encompass diverse physical and chemical properties. Many of these wastes have been identified as problem wastes because they are difficult to encapsulate using conventional technologies and/or produce waste forms of poor quality that do not successfully retain hazardous constituents in the disposal environment. Due to reductions in volume resulting from incineration of contaminated combustible materials, remaining ash residues may contain sufficient quantities of hazardous elements, including heavy metals, that they meet the Environment Protection Agency (EPA) definitions for hazardous waste as well as DOE definitions for low-level radioactive waste. According to EPA's guidelines for delisting of hazardous and mixed wastes, before disposal at approved low-level waste disposal sites, such wastes must first be treated to immobilize the hazardous constituents. Attempts have been made in the past to render radioactive, hazardous and mixed wastes harmless by immobilizing the wastes against dispersion by ecological forces. In one method, the mixed wastes are sealed into metal or plastic containers which are then stored underground or in the ocean. In another method, the wastes are incorporated into a matrix of materials, such as inorganic cements and polymers while in their fluid or molten state followed by solidification. The high viscosity of molten plastics generally has limited the quantity of waste which can be loaded into the plastic matrix. Frequently, incorporation of wastes in a plastic mixture is limited by the inability of the matrix to isolate the waste from the environment. Highly loaded matrices having over 30 percent loadings have been unsatisfactory because of leaching. Thus, one disadvantage of known processes used to encapsulate waste is the tendency of waste to become mobilized. Other disadvantages of conventional hydraulic cement and other thermosetting polymer processes include low efficiency of waste encapsulation, a requirement to cure the matrix by adding chemicals and/or increasing the temperature, steps which result in increased operating costs. Accordingly, there is still a need in the art of waste disposal to provide a composition and method for encapsulating and stabilizing radioactive, hazardous and mixed wastes. It is, therefore, an object of the present invention to provide a composition and process for encapsulating and stabilizing radioactive, hazardous and mixed wastes in a multi-barrier process. Another object of this invention is to provide durable waste forms which can withstand mobilization by ecological forces. Yet, another object of the present invention is to develop an encapsulating process which has no curing requirements for solidification, thus providing significant cost savings. SUMMARY OF THE INVENTION The present invention, which addresses the needs of the prior art, provides a composition and a process for the disposal and immobilization of radioactive, hazardous and mixed wastes. More specifically, it has now been found that by heating and mixing in an extruder, substantially simultaneously, (i) dry waste powder including radioactive, hazardous and mixed wastes; PA1 (ii) a non-biodegradable thermoplastic polymer; and PA1 (iii) an anhydrous additive capable of forming a precipitate with the radioactive and/or toxic materials found in the dry waste powder; a molten homogenous waste matrix is obtained which is conveyed into a substantially pure or "clean" polyethylene container, wherein the molten matrix cools slowly to form a solid waste form which has an exterior layer of substantially pure or "clean" polyethylene. The anhydrous additives used to form insoluble precipitates with the radioactive and toxic materials include calcium hydroxide, sodium hydroxide, sodium sulfide, calcium oxide, magnesium oxide or a mixture thereof. Loss-in-weight feeders are used to provide the components of the molten waste mixture to the extruder in precise pre-determined ratios. Pretreating steps include drying and comminuting all waste to a dry waste powder having a particle size not greater than 3,000 microns. Waste particles which are less than 50 microns are pre-mixed with polyethylene binder by a high-speed kinetic batch mixer to form a homogenous waste mixture prior to charging into the extruder. As a result of the present invention a composition and process for disposal of radioactive, hazardous and mixed waste are provided, wherein radioactive and toxic wastes are stabilized by forming insoluble precipitates with chemical additives, all enclosed in a monolithic waste form having at least a second barrier to leaching of waste components into the environment. In contrast to hydraulic cement and thermosetting polymer processes, the present invention does not require chemical reactions for solidification, so that waste-binder interactions are minimized. The present invention allows a broader range of waste types to be encapsulated and results in better waste loading efficiencies, i.e., more waste per container. Preferred formulations prepared in accordance with the present invention can contain as much as 60% by weight, while still maintaining leachability of toxic metals below allowable EPA criteria. In contrast, maximum loadings using conventional Portland cement are limited to about 16% by weight ash. Other improvements which the present invention provides over the prior art will be identified as a result of the following description which sets forth the preferred embodiments of the present invention. The description is not in any way intended to limit the scope of the present invention, but rather only to provide a working example of the present preferred embodiments. The scope of the present invention will be pointed out in the appended claims. |
summary | ||
abstract | A charged particle irradiation system that positions the beam at a target position to avoid irradiation of normal tissue includes an acceleration system 6 for extracting a charged particle beam, scanning magnets 24 and 25, and charged particle beam position monitors 26 and 27. On the basis of signals received from the charged particle beam position monitors 26 and 27, the control unit 70 calculates a beam position at a target position and then controls the scanning magnets 24 and 25 so that the charged particle beam is moved to a desired irradiation position at the target position. The control unit 70 corrects the value of an excitation current applied to each of the scanning magnets 24 and 25 on a specified cycle basis on the basis of information about the position and the angle of the charged particle beam. |
|
054769892 | summary | FIELD OF THE INVENTION This invention relates to an adsorbent useful for the adsorption of radioactive nuclides which generate, for instance, in the course of reprocessing steps for the separation and recovery of valuable substances such as uranium, plutonium and the like from nuclear fuel used in a nuclear reactor, and to a process for the volume-reduction treatment of radioactive waste that contains radioactive nuclides. BACKGROUND OF THE INVENTION Various types of liquid waste accumulated at reprocessing facilities after treatment of spent nuclear fuel discharged from nuclear power stations contain many radioactive nuclides including long-lived .beta. and .gamma. nuclides of cesium and the like and transuranium elements such as uranium, plutonium and the like. For the treatment of radioactive liquid waste, it is necessary to reduce the amount of radiation by separating and removing radioactive nuclides from the liquid waste in order to reduce radiation exposure. In general, radioactive liquid waste is treated by means of evaporation concentration, coagulating sedimentation, ion exchanging and the like. In the evaporation concentration process, liquid waste to be treated is put in an evaporator and heated under atmospheric or reduced pressure to allow only moisture to evaporate, thereby concentrating the radioactive liquid waste to a reduced volume. The evaporated moisture is recovered using a condenser. On the other hand, the thus concentrated liquid waste is subjected to further treatment such as bituminization or the like depending on the radioactive nuclides present in the waste. The evaporation concentration process, however, has disadvantages in that because corrosion-resistant materials are required, the decontamination factor (DF) decreases due to evaporation of radioactive nuclides which also occurs and the volume-reducing effect is not sufficient. In the coagulating sedimentation process, radioactive nuclides in the liquid waste are removed after their coagulation and precipitation. Radioactive nuclide-including sludge in which the radioactive nuclides are incorporated is subjected to dehydration treatment, and the resulting residue is treated as solid waste, and the supernatant fluid is treated as low-level liquid waste. In the coagulating sedimentation process, however, the sludge formed has a high moisture content which causes a difficulty in carrying out the dehydration treatment, thus entailing a disadvantage in that the volume-reducing effect is not sufficient. In the ion exchanging process, ions of interest are removed from the liquid waste by conducting ion exchange using an ion exchange resin. The spent resin containing the ions of interest is treated as solid waste, and the liquid portion after the treatment is treated as low-level liquid waste. A chelate resin may be used instead of an ion exchange resin in a process similar to the ion exchange process. However, when an ion exchange or chelate resin which is commonly used for the removal of metals from general liquid waste is applied to the treatment of radioactive liquid waste, it is difficult to use such a resin because of the tendency toward deterioration of such an organic polymer resin due to the action of radiation. Even where such an application could be effected, a problem of selectivity occurs. For example, virtually nothing is known about an adsorbent useful for the selective separation and removal of transuranium elements such as plutonium and the like which are present in a small amount in liquid waste of high uranium content. Inorganic adsorbents may have radioactive resistance, but nothing is known on an inorganic adsorbent which has excellent adsorptivity. On the other hand, an adsorbent in which a ferrocyanate as an inorganic functional group is supported on an acrylic fiber as an organic support is disclosed in JP-B-63-24415 (the term "P-B" as used herein means an "examined Japanese patent publication"). This adsorbent, however, has disadvantages in that the functional group is not selective for the adsorption of transuranium elements and the support, being organic, has poor durability. Also, JP-B-60-51491 discloses a phenol-based chelate resin which has an aminomethylphosphonic acid-type functional group, and which is described as having excellent uranium-adsorbing ability. This resin, however, is not capable of selectively adsorbing plutonium present in radioactive liquid waste. In addition to the above described disadvantages, the ion exchanging and chelate resin processes have other problems in that each of these processes generates a large quantity of secondary wastes such as incombustible or flame retardant spent resin, liquid waste after resin washing and the like, and insufficient volume-reducing effect arises. Although the volume of spent organic ion exchange resin may be reduced by incineration, generation of harmful gas, formation of smoke dust and scattering of radioactive nuclides all occur. In addition to such problems, the resin cannot be incinerated completely, leaving a soft charcoal residue which causes another problem by scattering atomized dust containing radioactive nuclides when the residue is treated. As a result, incineration of this type of resin is practically impossible. Thus, as has been described above, the prior art adsorbents have common problems in that they have poor durability against radiation and transuranium elements in radioactive liquid waste are not adsorbed selectively. In addition, the prior art volume-reduction treatment methods have problems in that secondary wastes are generated in a large quantity, insufficient volume-reducing effect arises and the facility cost becomes high because of the necessity to use corrosion-resistant materials. SUMMARY OF THE INVENTION In view of the above, therefore, an object of the present invention is to provide an adsorbent for radioactive nuclides, which is durable against radiation and is capable of adsorbing transuranium elements selectively. Another object of the present invention is to provide a process for the volume-reduction treatment of radioactive waste, which enables significant volume reduction of radioactive nuclides-adsorbed waste and scattering of radioactive nuclides at the time of incineration does not occur. To achieve these objects, the inventors of the present invention have conducted intensive studies and they have found that an adsorbent of a fibrous active carbon system with an inorganic framework, especially with a specified equilibrium moisture regain, adsorbs radioactive nuclides excellently, that plutonium present in plutonium-bearing liquid waste can be separated and removed selectively and securely by using such an adsorbent and that, by incinerating the fibrous active carbon with excellent adsorptivity at a temperature higher than its ignition point, the fibrous active carbon alone can be gasified and scattered substantially completely while preventing adsorbed radioactive nuclides from being scattered. The present invention has been accomplished on the basis of these findings. Particularly, the present invention provides an adsorbent useful for the adsorption of radioactive nuclides which comprises fibrous active carbon having a specific surface area of 1,000 m.sup.2 /g or more and an equilibrium moisture regain of 10% or more at a relative humidity of 45%. The present invention further provides a process for the volume-reduction treatment of radioactive liquid waste which comprises subjecting radioactive liquid waste containing radioactive nuclides to an adsorption treatment using an adsorbent that comprises fibrous active carbon having a specific surface area of 1,000 m.sup.2 /g or more, and subsequently incinerating the adsorbent at a temperature higher than the ignition point of the fibrous active carbon. In this instance, the specific surface area is calculated by the so-called BET method based on nitrogen gas adsorption isotherm at liquid nitrogen temperature. These and other objects and advantages of the present invention will become apparent as the description progresses. DETAILED DESCRIPTION OF THE INVENTION Though not particularly limited, the material of fibrous active carbon which constitutes the adsorbent of the present invention may be selected, for example, from coal pitch, petroleum pitch, rayon, phenol fiber, acrylic fiber and the like. The fibrous active carbon may be produced by utilizing the conventional process as disclosed, for example, in U.S. Pat. No. 4,808,202, herein incorporated by reference. The specific surface area of the fibrous active carbon is not particularly limited, provided that the fibrous material has enough pores to adsorb the radioactive nuclides. However, since the adsorbed quantity of radioactive nuclides increases as the specific surface area increases, the fibrous active carbon has a specific surface area of preferably 1,000 m.sup.2 /g or more, more preferably from 1,500 to 2,500 m.sup.2 /g. The specific surface area may be controlled by changing the temperature or time with regard to the activating treatment. That is, the activating treatment at a higher temperature or for a longer period causes increase of the specific surface area. Also, the fibrous active carbon forming the adsorbent has an equilibrium moisture regain of preferably 10% or more, more preferably 15% or more, at a relative humidity of 45%. The equilibrium moisture regain at a relative humidity of 45% is measured in accordance with Kagakubinran, Kisohen, II, p. 143, edited by The Chemical Society of Japan, published by Maruzen Co. That is, a saturated KNO.sub.2 solution is placed in a sealed vessel and left at 20.degree. C. until the space of the vessel has a constant humidity (relative humidity: 45%). Then, a sample of fibrous active carbon is placed in the space and left until the amount of water adsorption is saturated. The moisture regain is measured, and the value thus measured is taken as the equilibrium moisture regain at a relative humidity of 45%. Since active carbon is mainly made of carbon, active carbon is generally non-polar and, therefore, its surface is hydrophobic. Accordingly, active carbon hardly adsorbs moisture at a low relative humidity of about 45%, thus showing a low equilibrium moisture regain. If active carbon is subjected to a carbonization treatment at a high temperature of from 500.degree.to 1,000.degree. C. during production, polar groups such as carboxyl, carbonyl, hydroxyl and the like groups are formed on the surface of the active carbon. This means that the active carbon becomes hydrophilic although its degree varies depending on the production process. Also, active carbon having high hydrophilic property can be produced by applying an addition treatment to commonly used active carbon as described later. The inventors of the present invention have conducted adsorption tests of radioactive nuclides using various fibrous active carbon preparations with different hydrophilic properties obtained in this method. As a result, they have found that the adsorptivity of fibrous active carbon becomes high as the hydrophilic degree of its surface increases. In other words, it is desirable for the fibrous active carbon to have an equilibrium moisture regain of 10% or more at a relative humidity of 45%. Although the reason for this is not clear while not desiring to be found, it appears that the capacity of fibrous active carbon to adsorb radioactive nuclides is improved when its surface is hydrophilic, because radioactive nuclides form complex compounds in radioactive liquid waste. A fibrous active carbon preparation having an equilibrium moisture regain of 10% or more at a relative humidity of 45% may be obtained by subjecting conventional fibrous active carbon to air oxidation, ozone oxidation or liquid phase oxidation, or by adding hydrophilic functional groups to the fibrous active carbon. The air oxidation treatment may be effected in the atmosphere at a temperature of from 300.degree.to 700.degree. C., preferably from 350.degree.to 600.degree. C. If the temperature is lower than 300.degree. C., a prolonged period of time necessary for the oxidation reaction is required, and if the temperature is higher than 700.degree. C., excess burning occurs reducing the volume of the fibrous active carbon. The oxidation time, although it varies depending on the heating temperature, is generally in the range of from 10 minutes to 5 hours. Alternatively, the air oxidation treatment may be carried out in the stream of heated air. The ozone oxidation may be effected by oxidizing fibrous active carbon in an ozone-containing atmosphere. The ozone concentration can be in the range of preferably from 100 to 500 ppm, more preferably from 250 to 450 ppm. If the ozone concentration is lower than 100 ppm, a prolonged period of time is necessary for the oxidation reaction, and if the concentration is higher than 500 ppm, the volume of the fibrous active carbon is reduced. The liquid phase oxidation may be effected by soaking fibrous active carbon in an oxidant solution for a period of from several hours to several days, followed by filtration, washing and drying in that order. The oxidizing agent may be selected from permanganates, chromates, hypochlorites, persulfates, bromic acid ion, chlorine, dilute nitric acid, concentrated nitric acid, hydrogen peroxide and the like. The functional groups to be added to fibrous active carbon are not particularly limited provided that they are hydrophilic in nature. Examples of suitable functional groups include hydroxyl, carboxyl, carbonyl, primary amino, secondary amino, tertiary amino, quaternary ammonium, sulfonic, phosphonic, ester, amide, nitroso, nitro, thiol, silanol, selenol and the like. Among these groups, hydroxyl, carboxyl, primary amino, secondary amino, tertiary amino, sulfonic, and phosphonic are preferable. These functional groups may be present alone or as a mixture of two or more. These functional groups can be added to fibrous active carbon for example: by adding a functional group-containing low or high molecular weight compound; by first adding a functional group-containing low molecular weight compound and then converting the thus obtained compound into a high molecular weight compound; by first adding a compound without any functional groups and then adding functional groups to the compound obtained; or by adding functional groups directly to the fibrous active carbon. When functional groups are added to the fibrous active carbon, the fibrous active carbon may have any optional shape and may therefore be molded into the form of web, sheet, cartridge or the like. Although the specific surface area of the fibrous active carbon decreases due to by the addition of functional groups, it is desirable to maintain the specific surface area at 1,000 m.sup.2 /g or more even after the addition reaction. According to the present invention, the fibrous active carbon which is used for the volume-reduction by incineration after adsorption of radioactive nuclides may have a composition in which the total amount of carbon, oxygen and hydrogen is 60% or more, preferably 80% or more. When a total of these components is less than 60%, flameless burning is not achieved appropriately at the time of incineration, and thus a harmful gas is sometimes generated. According to the present invention, liquid waste which contains radioactive nuclides may be subjected to an adsorption treatment using the adsorbent comprising fibrous active carbon with no pretreatment of the liquid waste, or after improving adsorptivity of the radioactive nuclides to the adsorbent by adding a complexing agent to the radioactive nuclides-bearing liquid waste to form complex compounds with the radioactive nuclides in the liquid waste. Examples of complexing agents which can be used include ethylenediaminetetraacetic acid (EDTA), tributyl phosphate, bis-(2-ethylhexyl) phosphate, mono-2-ethylhexyl 2-ethylhexylphosphonate, triethylamine, trioctylamine, phthalocyanine and the like. Among these agents, tributyl phosphate and bis-(2-ethylhexyl) phosphate are preferable. Some radioactive nuclides may change their ionic states and dispersion conditions depending on the acid concentration, thus changing their adsorptivity to the fibrous active carbon. For the purpose of improving the adsorptivity, the adsorption treatment may be carried out after adjusting the acid concentration to an appropriate level with the addition of an alkali or acid such as NaOH, HCl, HNO.sub.3 or the like, preferably with a 0.01-3N nitric acid solution. According to the present invention, practical treatment of radioactive liquid waste may be effected by employing any of the prior art means such as a batch method in an adsorption tank, a column process through an adsorption column and a combined means thereof. Alternatively, the fibrous active carbon may be molded into a sheet, a cartridge or the like, and a column process may be carried out using the molded product as an adsorbent. When the shape of adsorbent cannot be produced easily from the fibrous active carbon alone, an inorganic or organic binder may be blended with the fibrous active carbon. Also, the adsorbent may be formed by blending the fibrous active carbon with heat adhesive fibers and thermally fusing them at the time of molding. Examples of such fibers include copolymerized polyester fibers, polyolefin fibers made of polyethylene, polypropylene and the like, low melting point nylon and conjugated fibers made of polyester as the core and polyolefin as the sheath. These binders and fibers may be blended in an amount of 40% by weight or less, because amounts exceeding this range sometimes causes poor incineration. In the column process, radioactive liquid waste is passed through an adsorption column which has been packed with the adsorbent. The packed adsorbent layer may have a thickness of 200 mm or more, preferably in the range of from 500 to 2,000 mm. If the thickness of the packed layer is less than 200 mm, leakage occurs. Although the liquid passing rate varies depending on the properties of liquid waste to be treated, it is generally 0.5 hr.sup.-1 or more, preferably in the range of from 1 to 10 hr.sup.-1, as a space velocity (SV). In the cartridge process, liquid waste is passed through a cartridge of the fibrous active carbon which has been molded into a cylindrical or columnar form and set in a housing. The use of such a cartridge is quite effective especially when radioactive materials are handled, because it can be installed and detached easily and disposal of the spent cartridge is easy. Such a cartridge may be obtained, for example, by wet-molding the fibrous active carbon, or by firstly molding the fibrous active carbon into a sheet by a paper making process or a dry process and then tightly rolling the molded sheet. When molded into a cartridge, the fibrous active carbon may be blended with a small amount of inorganic and organic binders. According to the present invention, radioactive liquid waste, which is difficult to treat using prior art processes, can be processed easily, because plutonium can be adsorbed and removed selectively from plutonium-bearing liquid waste, by the use of an adsorbent which comprises fibrous active carbon having excellent selective adsorptivity and durability, especially having a specific surface area of 1,000 m.sup.2 /g or more and an equilibrium moisture regain of 10% or more at a relative humidity of 45%. According to the present invention, incineration of the radioactive nuclides-bearing adsorbent after the adsorption treatment can be carried out at a temperature which is equal to or higher than the ignition point of the adsorbent-constituting fibrous active carbon. The term "ignition point" as used herein means the temperature at which the temperature starts to increase sharply when a sample is heated in accordance with the ignition point measurement procedure of JIS-K-1474 (granulated active carbon test methods). When fibrous active carbon is heated at a temperature equal to or higher than its ignition point, it becomes red, starts flameless burning and its volume is reduced. Such a volume reduction seems to occur due to scattering of carbon converted into the form of carbon dioxide, because most portion of the fibrous active carbon is comprises carbon. As described above, incineration of the adsorbent is carried out at a temperature which is equal to or higher than the ignition point of the adsorbent-constituting fibrous active carbon, but preferably with an upper limit temperature of "ignition point +600.degree. C.". When fibrous active carbon is placed in a temperature atmosphere which is above a temperature of "ignition point +600.degree. C.", carbon bonds are severed by thermal decomposition simultaneously with the burning, thus burning with flames occurs. When fibrous active carbon burns with the emission of flames, some of radioactive nuclides scatter together with the flames, thus an additional gas-adsorbing filter is required. The flameless burning of fibrous activated carbon seems to occur due to its small apparent density (0.05 to 0.3 g/cm.sup.3) and high air content necessary for burning, because each fiber has a considerably small diameter (10 to 30 .mu.m) and innumerable pores. According to the present invention, the radioactive nuclides-bearing adsorbent removed after the adsorption treatment may be subjected to the incineration step as it is, but preferably after dehydration and drying treatments. As has been described above, according to the present invention, fibrous active carbon is used as an adsorbent and the radioactive nuclides-bearing waste removed after the adsorption treatment is subjected to an incineration treatment, thus making possible a marked reduction in the waste and prevention of the scattering of radioactive nuclides at the time of incineration, without bringing up a problem of the need for extra storage space due to increased amounts of waste. In addition, since only a burned residue, mainly containing non-volatile radioactive nuclides and co-present metal components, remains as a secondary waste in a very small amount, and the burned residue can be applied to any prior art processing, the process of the present invention does not require any additional special facility and therefore is quite effective from the economic point of view. |
050193236 | description | DISCLOSURE OF THE PREFERRED EMBODIMENT A. Preparation of Tellurium-124 Targets In synthesizing Iodine-124, a copper metal plate is first milled and uniformly lapped to the dimensional specifications required for ultimately placing the target matrix in the accelerated deuteron particle path of a nuclear accelerator apparatus, such as a cyclotron. The surface of the copper plate is sanded, washed with distilled water, and dried. The copper plate is then placed in a nickel plating solution prepared from a salt, such as nickel sulfate hexahydrate, and is then electroplated using a platinum electrode as the anode. The nickel plated, copper plate is then placed in a tellurium plating solution comprising isotopically enriched Tellurium-124 dioxide dissolved in a solution of potassium hydroxide. The tellurium is electroplated onto the nickel plated copper plate using a platinum electrode while the plate is water cooled. The target thickness of the Tellurium-124 is typically 10 to 14 mg/cm.sup.2 for routine production targets, and generally must be at least 0.1 mg/cm.sup.2 Enriched Tellurium-124 is most preferably plated in the quantity of about 13 mg/cm.sup.2 on the nickel plated copper target. These targets are then irradiated in the internal beam line of a Cyclotron. The current is varied from 25 to 80 micro-amperes, and the irradiation time is varied from 4 to 8 hours. B. Radiochemical Processing the Iodine-124 After irradiating the Tellurium-124 with deuterons by means of the internal beam line of a cyclotron, the irradiated Tellurium-124 is dissolved from the copper plate by means of a sodium hydroxide solution, with equal volumes of 30% H.sub.2 O.sub.2, and 5 molar NaOH, plus sufficient deionized H.sub.2 O to cover the electrodeposited Tellurium-124. The dissolution converts most of the .sup.124 I to .sup.124 IO.sub.3 and .sup.124 IO.sub.4 ; whereas, most .sup.124 Te was converted to .sup.124 Te+4 (Eqs.1-2). EQU (Eq. 1) .sup.124 Te.degree.+2NaOH+3H.sub.2 O.sub.2 =.sup.124 TeO.sub.4 +4H.sub.2 O+2Na+ EQU (Eq. 2) .sup.124 I.sup.- +2NaOH+H.sub.2 O.sub.2 =.sup.124 IO.sub.3 +2Na++H.sub.2 O Following dissolution, the solution and water rinse was transferred to a 250 ML round bottom flask containing 250 mg of Al Powder. The Al promotes the Iodine-124 to be converted to the iodide (I.sup.-) form which is required for subsequent medical uses. The flask was gently heated until the H.sub.2 O.sub.2 was decomposed and Tellurium-124 precipitated. EQU (Eq.3) .sup.124 IO.sub.3 +2Al+OH.sup.- =.sup.124 I.sup.- +2AlO.sub.3 +H.sub.2 EQU (Eq.4) .sup.124 TeO.sub.4 +3Al+2OH=.sup.124 Te.degree.+3AlO.sub.2 +H.sub.2 EQU (Eq.5) .sup.124 Te.degree.+Al+2OH.sup.- =.sup.124 Te.sup.-2 +AlO.sup.2 +H.sub.2 Occasionally, a dark violet coloration was observed due to the formation of telluride (Eq. 5). A five minute purge of air through the solution oxidizes telluride to Te.degree.. A purge of CO.sub.2 for five minutes converts sodium aluminate to aluminum hydroxide (Eqs. 6-7). The final volume was adjusted to specific needs before the Iodine-124 solution was passed through a fine glass filter. The precipitated Tellurium-124 and sodium aluminate was retained in the filter. EQU (Eq. 6) 2.sup.124 Te.sup.-2 +O.sub.2 +2H.sub.2 O=2Te.degree.+40H.sup.- EQU (Eq. 7) AlO.sub.2 +CO.sub.2 +2H.sub.2 O=Al(OH).sub.3 +HCO.sub.3 The solution was predominantly .sup.124 I.sup.- at pH.about.8.5 buffered by bicarbonate formed during the process of CO.sub.2 addition to form the soluble Al(OH).sub.3. Table 1 illustrates Iodine-124 production yields and levels of Iodine-126 impurity 48 hours after irradiation of the Tellurium-124 target of greater than 95% isotopic enrichment. TABLE I ______________________________________ Iodine-124 Production Yield Data .sup.124 I (%) Dose, Fluence, .sup.124 I mCi .sup.126 I mCi (after mAh A mCi/Ah (EOB) (EOB) 48 hours) ______________________________________ 100 25 0.56 57.0 0.5 99.1 200 25 0.60 119.0 0.6 99.5 240 40 0.52 124.0 0.7 99.4 250 50 0.57 143.0 0.5 99.6 300 60 0.59 150.0 0.6 99.6 500 75 0.48 260.0 <1.2 99.5 210 70 0.44 93.7 <0.65 99.3 325 65 0.62 210.0 <1.3 99.3 500 80 0.54 269.66 <1.83 99.3 ______________________________________ Radioanalysis by gamma-ray spectrometry was performed to assess the radionuclidic purity and to identify the impurities. Irradiation conditions in the examples range from 25 to 85 microampere deuteron beam current, with irradiation doses ranging from 100 to 550 microampere hours. The production yield of Iodine-124 was directly proportional to the dose. The current could be increased to 85 micro-amperes without damaging the target. Iodine-124 was prepared in quantities of greater than 100 mCi by 15 MeV deuteron irradiation of isotopically enriched Tellurium-124 and the Tellurium-124 (d,2n) Iodine-124 nuclear reaction. The threshold deuteron energy for the nuclear reaction is about 6.5 MeV. For synthesis of labelled organic molecules, the Iodine-124 and iodine mixture was passed through a cation-exchange column to remove salts and trace metals. C. Removal of Salts from Iodine-124 Solution It is important in order to label many organic compounds such as proteins, monoclonal antibodies, and natural products, that the labeling solutions be as chemically pure as possible. Only the radiochemically pure form (I.sup.-) of Iodide is used in the labeling of radiopharmaceuticals. The presence of salts and reducing agents interferes with labeling methods used by those skilled in the art. It should be noted that with the method of the present invention, the use of reducing agents is not required. However, removal of deleterious salts greatly decreases the rate of autoradiolytic decomposition of the solutions. If the salts are not removed, the radiochemical composition of the high specific activity Iodine-124 solutions changes with time. In FIG. 4, the upper line 17 shows the rate of autoradiolytic decomposition of Iodine-124 and the lower lines 18, 20 and 22 show the increasing presence of unwanted iodate radiochemical forms of Iodine-124, IO.sup.-.sub.6, IO.sup.-.sub.3, and UI, UI being an unidentifiable radioactive species FIG. 5 shows a slower rate of decomposition of Iodine-124 after removal of salts from the solution Line 17a represents the improved rate of decomposition of Iodine-124, and line 18a shows the decreased rate of formation of the products of decomposition. An effective method for removal of the deleterious salts from solutions containing Iodine-124 has been found without the reasons for its effectiveness being completely understood. The procedure is as follows: Fill a column (about 1.5.times.50 cm) with a Chelex-100 resin to a level of about 20 cm. The resin should be prepared beforehand by placing it in a beaker and covering it with water for at least 12 hours. The column should then be washed with water. Then, 150 ml. of 7.0 M HCl should be passed through the column, followed by a water washing such that the eluant has a pH of approximately 7.0. It is important to maintain a neutral pH, and the pH should be checked. Excess water should be drained and the column should be closed to prevent drying of the resin. In a beaker, place 1 ml. of 0.1M NaOH, and place it under the column containing the resin. Then, pour the iodine solution containing salts into the column and allow it to drain through the column. The column used should be washed to remove most of the radioactivity. The solution is then heated to remove excess water so that its concentration is about 15 to 120 mCi per 1.0 ml. D. Production of (I.sup.124)-m-IBG As way of example, the following is a discussion of Iodine-124 being incorporated in meta-iodobenzylguanidine (m-IBG). (I.sup.124)-m-IBG is one example of a radiopharmaceutical which can be used to provide either diagnosis or therapy using PET instrumentation. Non-radioactive m-IBG was synthesized by the method of Wieland, disclosed in Wieland et al., 21 Journal of Nuclear Medicine 349 (1980). The m-IBG was characterized on a Nicolet Model 5DX IR: (KBr) showed broad peaks between 3100 to 3448 (NH.sub.2,NH), 1600(C=N), 1590 (aromatic C=C), 772 & 687 cm .sup.-1 (m-disubstituted phenyl). Mass spectroscopy analysis (direct probe insertion) was performed on a Finnegan MAT Model-311 : molecular ion (M+) and a base peak (rel. intensity 100%) at m/z 276, a peak (relative intensity 060%) at m/z 233 (M-43) representing the split of the --C group. Proton NMR analysis was performed on a Varian Model T-60A : (DMSO-d6); delta 7 to 7.8 (m,4H aromatic), the benzylic CH.sub.2 group is overmasked by the water peak at delta 3.4. Melting point: 167.3 degrees centigrade (corrected); literature: 167.0 degrees centigrade (uncorrected). The exchange reaction to prepare (I.sup.124)-m-IBG was adopted with a modification from the method of Van Doremalen, et al., 96 J. Radioanal. Nucl. Chem., Letters., 97 (1985). In a 10 ml borosilicate serum vial 2.7 micro-moles of "cold" metaiodobenzylgaunidine sulphate was mixed with 6.2 micro-moles of Cu(NO.sub.3).sub.2. Iodine-124 (5 to 20 mCi) was added. The total volume was brought to approximately 0 8 ml. with water, and the mixture was then adjusted to pH 5. The vessel was stoppered and heated to 150 degrees centigrade in an oil bath for 45 minutes Upon cooling, 1.5 mL of 2.45% sodium biphosphate buffer solution was added to precipitate copper. Copper phosphate precipitate was then removed by filtering through 0.22 micron millipore filter. The filtrate was passed through 100 to 200 mesh Bio-Rad AGI-X8 anion-exchange resin to remove the unreacted iodide. Incorporation of Iodine-124 into m-IBG was accomplished in a radiochemical yield of 70 to 90%. HPLC analysis of the filtrate after removal of copper phosphate precipitate (see FIG. 1) indicated the presence of unreacted iodide and occasionally the unprecipitated copper. However, a careful passage of the filtrate through a Bio-Rad anion-exchange resin completely removed the Iodine-124, rendering the filtrate greater than 95% radiochemically pure (see Table 1 above). Complete precipitation of copper was ensured by adjustment of the pH and concentration of the phosphate buffer. With this modification, the final preparation contained less than 1 micro-gram/ml of copper by wet chemical analysis. Alternatively, m-IBG labeled with Iodine-124 can be produced when Iodine-124 having high chemical purity is used in a procedure whereby reaction mixtures are passed through a Waters octadecyl "Sep-Pak" cartridge while the cartridge is purged with water to remove the inorganic chemical forms of sulphate and Iodine-124. The labelled m-IBG is removed from the cartridge with 5 ml. of ethanol followed by rapid concentration in a stream of air. Reconstitution for injection follows by the addition of isotonic saline. Another example of the use of radionuclidicly pure Iodine-124 is in the labeling of the B-HCG polyclonal antibody, which can be used to locate choriocarcinoma, which is very difficult to diagnose by other conventional methods. E. Chromatography Chromatographic and radiochemical procedures were applied to obtain greater than 95% Iodine-124 activity in iodide anion form for further radiochemical synthesis. Sodium Iodine-124 solution was analyzed by thin layer chromatography (TLC) using SG ITLC, available from Gelman Instrument Co., USA. The developing solvent was the organic phased prepared by mixing 3 ml. of NH.sub.4 OH in 12 ml. of 1-butanol; R.sub.f values: I.sup.- : 0.8; IO.sub.3.sup.- and other radiochemical impurities: 0.0 to 0.15. (I.sup.124)-m-IBG was analyzed by TLC on silica gel plates with ethylacetate: ethanol: H.sub.2 O (20:20:1) as the developing solvent (Rf, I.sup.- : 0.75; I.sup.124 -mIBG:0.00). High pressure liquid chromatography (HPLC) analysis of m-IBG was performed on a Varian 5000 HPLC System. Column effluent was passed first through a variable UV detector (254nm), and then through a radioactivity detector (NaI) connected in series with the UV detector as shown in FIG. 1. Analysis of m-IBG was performed on an Altech C-18, 10 micron column. The column was eluted with an eluate composed of 60% 0.05M NH.sub.4 H.sub.2 PO.sub.4 and 40% CH.sub.3 CN, at a flow rate of 2mL/minute. Retention time for m-IBG was 4.68 minutes while the unbound Iodine-124 and/or Cu.sup.+2 eluted with the solvent front. Analysis of Iodine-124 iodide (I.sup.-) was performed on RP 18 Lichorsorb 4.6.times.250 mm column. The eluting buffer was composed of 0.05 M phosphate and 0.002 M tetrabutylammonium hydroxide in 5% methyl cyanide, pH 7.0 at a flow rate of 1 ml/min. The K' values for IO.sub.3.sup.-, IO.sub.6.sup.-, I.sup.- and IO.sub.4.sup.- under the conditions are 0, 1.57, 1.92 and 11.67, respectively. F. Iodine-124 As a Radiopharmaceutical The Iodine-124 can be produced as sodium-Iodine-124 and orally administered to a patient diagnosed with thyroid carcinoma. A 1 mCi to 5 mCi oral dose of (I.sup.124)-iodide may be administered for tomographic imaging of the thyroid gland. Positron camera imaging can be used to evaluate the therapy. Other gamma radiation associated with the radioactive decay of Iodine-124 does not cause appreciable interference with imaging the positron annihilation photons. Imaging of the thyroid can be accomplished at 4 to 24 hours after administration of the Iodine-124 iodide dose. If biopsy, or other clinical data indicate, a 100 to 200 mCi internal therapeutic dose of Iodine-131 could be administered for the purpose of destroying residual thyroid tissue after surgery. The low dose of Iodine-124 PET study aids in accurate estimation of thyroid function, and the anatomical and morphological structure involved. This allows more accurate dosing than is afforded by conventional imaging methods. Alternatively, a therapeutic dose of Iodine-124 could be used instead of Iodine-131. The absorbed radiation dosimetry for Iodine-124 is approximately 69% of that for a comparable quantity of Iodine-131. Therefore, Iodine-124 should be used primarily in patients suspected of having a diseased condition which, if clinically confirmed, would be subsequently treated with a radiotherapeutic dose of Iodine-124 or Iodine-131. Another feature of the use of Iodine-124 is that a diagnostic PET study could follow internal radioiodine therapy. Iodine-124 remaining in the patient can be used to tomographically establish the presence, if any, of residual thyroid tissue intended to be removed by surgery or internal radiation treatment. Large quantities (150 mCi) of Iodine-124 can be routinely produced by the Tellurium-124 (d,2n) Iodine-124 reaction. The production yield data is given in Table 1. Iodine-124 can be produced in higher yields and final product purity by using this reaction rather than by the Tellurium-124 (p,n) Iodine-124 reaction. The yields for the Tellurium-124 (d,2n) Iodine-124 was 0.57 mCi per microampere hour, compared to 0.093 mCi per microampere hour for the Tellurium-124(p,n) Iodine-124 reaction reported in Kondo et al, 28 Int. J. App. Rad. and Isotopes, 765, (1977). Another example of the use of Iodine-124 is in the diagnosis of tuberculoma. Isonicotinic acid hydrazide (INH) has been one of the most effective agents in tuberculosis therapy since 1952. The aromatic nucleus of INH can be labeled with I-124 to be used as a radiotracer for differential diagnosis of tuberculoma. 2-iodoisonicotinic acid (1.8 mg) was suspended in was (200 1) and 5N sodium hydroxide solution (100 1) was added and the vial was capped tightly, and heated at 140.degree. C. for 2h. Then the solution was acidified with dilute hydrochloric acid until a faint precipitate appears. The solvents were removed with the aid of a steam of nitrogen and the resultant material was extracted with methanol (3.times.500) 1). This methanol solution was treated with diazomethane until a persistent yellow color appears. The solvents were evaporated and the residue was dissolved in ethanol (100 1): and heated to boil. Then hydrazine hydrate (20 1) was added and after 1 minute, the reaction mixture was analyzed by HPLC using carbon-18 reverse phase column, with acetonitrile: water (40v:60v) as the eluent. Retention time of free .sup.124 I-Iiodide, .sup.124 I-2-iodo-methylisonicotinate and .sup.124 I-2-iodoisonicotinic acid hydrazide were 2.34, 12.32, and 3.01 minutes respectively. Overall radiochemical yield was 16% and the time spent for chemical manipulations was 3.5 hours. Biodistribution studies can then be conducted. Iodine-124 of the present invention is intended for use in a wide variety of radiopharmaceutical applications. Such uses include synthesis of organic compounds with Iodine-124. The purity of the Iodine-124 solutions of the present invention greatly facilitate the production of such products. For example, one can make biologically active cell-specific or receptor-specific compounds that are selectively sequestered at desired tissue sites without rise of in vivo release of the radionuclide from its carrier. Iodine-124 can be incorporated into a cell-specific binding agent, as an unsaturated organic linker, and may be used directly as a radiopharmaceutical or may be covalently bonded to monoclonal antibodies Iodine-124 prepared in accordance with this invention may be provided in a kit usable by a physician, pharmacist, or researcher to prepare radiopharmaceuticals to their own specifications. Specific examples of possible uses include incorporating Iodine-124 into a steroid group, an aryl group, a substituted aryl group, a vinyl group, or an aryl group capable of coupling with antibodies. Similarly, Iodine-124 can be incorporated into an aromatic amine, an aromatic isocyanate, an aromatic carboxylic acid, and aromatic isothiocyanate, benzoic acid, a substituted benzoic acid group, or a vinylestradial group. Alternatively, the Iodine-124 can be combined with non-cell selective compounds, such as styrenes or styrene polymers that can be formed into a colloidal dispersion or particulate form and then used for radiation synovectomy in the treatment of rheumatoid arthritis. In addition, the Iodine-124 produced in accordance with the present invention can be incorporated into iodinated organic compounds such as steroids, cholesterol and estrogen derivatives and hormones. In summary, this invention is a reliable method for obtaining greater than 100 millicurie quantities of Iodine-124 in greater than 99.5% radionuclide purity via bombardment of isotopically enriched Tellurium-124. The Iodine-124 has physical properties that are useful for diagnostic and therapeutic radiopharmaceuticals, particularly when used in conjunction with positron emission tomography. Furthermore, there is also interest in using Iodine-124 as a radioactive standard. The foregoing detailed description has been given for illustration purposes only. A wide range of changes and modifications can be made to the preferred embodiment described above. It should, therefore, be understood that it is the following claims, including all equivalents, which are intended to define the scope of this invention. |
description | The present invention relates to a technique of inspecting a circuit pattern of an electronic device. A recent electronic device has been refined and multilayered, furthermore, its logic has become complicated, and therefore the manufacturing process has become complicated. Accordingly, a defect due to the manufacturing process is generated, and therefore it is expected to efficiently and accurately inspect a circuit pattern of the electronic device. Especially, it is important to accurately inspect a circuit pattern with high and low points such as a hole having a high aspect ratio of the depth and bore diameter and a circuit having a multilayer structure. For inspection of such a circuit pattern, apparatuses such as a CD-SEM (Critical Dimension Scanning Electron Microscope) and DR-SEM (Defect Review Scanning Electron Microscope) are used. These apparatuses send a charged particle radiation such as an electron beam to a circuit pattern formed on a silicon wafer or reticle (mask), convert a secondary electron released from the circuit pattern into an image signal (hereinafter, referred to as “secondary electron image”), analyze the secondary electron image and inspect the circuit pattern. To accurately inspect a circuit pattern, a secondary electron image that accurately reflects an actual circuit pattern image is required. However, in the above-noted circuit pattern with high and low points, it is difficult to complement a secondary electron released from a circuit pattern in a lower position. Therefore, the observation image contrast of a circuit pattern in a higher position may be degraded. The following Patent Literature 1 suggests a method of correcting the contrast of a secondary electron image. The following Patent Literatures 2 and 3 suggest a pattern matching method using a position determination image registered in advance, design data or data generated by a wafer manufacturing process simulation, as a template, in order to accurately specify a circuit pattern of an inspection target on a secondary electron image. Patent Literature 1: JP Patent Publication (Kokai) No. 2002-319366A Patent Literature 2: JP Patent Publication (Kokai) No. H05-101166A (1993) Patent Literature 3: JP Patent Publication (Kokai) No. 2002-328015A In the technique disclosed in Patent Literature 1 described above, the user needs to designate an image region used to analyze the contrast, which causes a burden. Also, in the case of inspecting many circuit patterns, it is difficult for the user to designate the above-noted image region to each circuit pattern. In the techniques disclosed in Patent Literatures 2 and 3 described above, a circuit pattern having a multilayer structure is inspected using a secondary electron image. However, it is generally difficult to accurately distinguish between a circuit pattern on the upper side and a circuit pattern on the lower side, which are included in the secondary electron image. Therefore, it is difficult to accurately find an inspection position. In addition to the above methods, it may be possible to correct the contrast of a secondary electron image based on brightness distribution of the entire image. However, since the circuit pattern on the upper side has a high contrast from the beginning, if the contrast is further corrected, the brightness becomes extremely high, which may cause the circuit pattern not to be identified. The present invention is made to solve the above-noted problem and it is an object of the present invention to provide a technique of accurately inspecting a circuit pattern in which the contract of an observation image is not clear, like a circuit pattern having a multilayer structure. A pattern inspection method according to the present invention divides a circuit pattern using the brightness of a reflection electron image and associates the region in the reflection electron image belonging to each division with a region in a secondary electron image. According to the pattern inspection method of the present invention, by dividing brightness values of a reflection electron image, it is possible to divide circuit patterns on the reflection electron image. Since the reflection electron image reflects a height direction shape of a multilayer structure well, it is considered that the contrast related to the height direction is better than a secondary electron image, and therefore it is considered that the above-described division represents a height direction shape of a circuit pattern well. By associating this reflection electron image and the secondary electron image, it is possible to accurately identify circuit patterns on the secondary electron image, so that it is possible to accurately inspect circuit patterns having a multilayer structure. FIG. 1 is a configuration diagram of an electronic device inspection system 1000 according to Embodiment 1 of the present invention. The electronic device inspection system 1000 denotes a system of inspecting a circuit pattern of an electronic device such as a semiconductor device and has an electron optical system 200, a computing unit 215, an imaging recipe creation unit 225 and a design system 230. In the following, each component of the electronic device inspection system 1000 will be explained. It should be noted that, although a silicon wafer 201 is used as an example of an electronic device in the following explanation, it does not limit the invention. The electron optical system 200 has an electron gun 203, a capacitor lens 205, a deflector 206, an ExB deflector 207, an objective lens 208, a secondary electron detector 209 and reflection electron detectors 210 and 211. The electron gun 203 generates a charged particle radiation such as an electron beam (i.e., primary electron) 204. In the following, a case will be explained where the electron beam 204 is generated. The capacitor lens 205 causes the electron beam 204 generated from the electron gun 203 to converge. The deflector 206 deflects the convergent electron beam 204. The ExB deflector 207 deflects a secondary electron to the secondary electron detector 209. The objective lens 208 images the converged electron beam 204 on the silicon wafer 201. The silicon wafer 201 is placed on an XY stage 217. The deflector 206 and the objective lens 208 control an irradiation position and diaphragm of the electron beam 204 such that the electron beam 204 is focused and irradiated in an arbitrary position on the silicon wafer 201 placed on the XY stage 217. The XY stage 217 is configured such that it is possible to shifting the silicon wafer 201 and takes a photograph of an image in an arbitrary position of the silicon wafer 201. To change an observation position by the XY stage 217 refers to as “stage shift,” and to deflect the electron beam 204 by the deflector 206 to change an observation position refers to as “beam shift.” A secondary electron and reflection electron are released from the silicon wafer 201 to which the electron beam 204 is irradiated, and the secondary electron is detected from the secondary electron detector 209. The reflection electron is detected by the reflection electron detectors 210 and 211. The reflection electron detectors 210 and 211 are placed in different positions, for example, in upper left and lower right positions of the silicon wafer 201 on the XY plane. The secondary electron and the reflection electron detected in the secondary electron detector 209 and the reflection electron detectors 210 and 211 are converted into digital signals in A/D converters 212, 213 and 214. The computing unit 215 controls operations of the above-described units. Also, it receives the detection result of each detector converted into a digital signal and stores the detection result in an image memory 252. In addition, it has a function as an observation image acquisition unit to generate a reflection electron image and a secondary electron image based on the detection result of each detector. For example, a CPU (Central Processing Unit) 251 and image processing hardware 253 perform image processing based on an inspection object and inspects an electronic device. Data stored in the image memory 252 may be stored in an external storage apparatus 223 anew. The computing unit 215 has a GUI (Graphical User Interface) to display, for example, an observation image or an inspection result for the user using a display 216 having an input unit. Although FIG. 1 shows a configuration example providing two detectors for reflection electron images, the number of detectors may be three or more. Also, part or all of the control in the computing unit 215 may be processed after being assigned to, for example, an electronic computer having a CPU or a memory that can accumulate an image. Further, the computing unit 215 is connected to the imaging recipe creation unit 225 via a network. The imaging recipe creation unit 225 creates imaging recipe data including a silicon wafer inspection coordinate required for inspection; a pattern matching template used to determine an inspection position; and an imaging condition. In the case of using design data as a pattern matching template, the imaging recipe creation unit 225 is connected to a design system 230 via a network or the like in order to acquire the design data. The design system 230 denotes a system to perform an operation of designing an electronic device such as an EDA (Electronic Design Automation) tool. The configuration of the electronic device inspection system 1000 has been described above. Next, a method of imaging a signal acquired by irradiating the electron beam 204 to the silicon wafer 201 will be explained. FIG. 2 is a diagram illustrating relationships between positions to which the electron beam 204 is irradiated on the silicon wafer 201 and an observation image. For example, as shown in FIG. 2(a), the electron beam 204 is scanned and irradiated as shown in irradiation directions 1001 to 1003 along the “x” direction or irradiation directions 1004 to 1006 along the “y” direction. By changing the deflection direction of the electron beam 204, it is possible to change a scanning direction. In FIG. 2, H1 to H3 represent positions in which the electron beam 204 scanned in the “x” directions 1001 to 1003 is irradiated on the silicon wafer 201. Similarly, H4 to H6 represent positions in which the electron beam 204 scanned in the “y” directions 1004 to 1006 is irradiated on the silicon wafer 201. The signal amount of secondary electrons released in H1 to H6 each is converted to a pixel brightness value through the secondary electron detector 209 and an AD converter 212. Also, the signal amount of reflection electrons is similarly converted to a pixel brightness value via the reflection electron detectors 210 and 211 and the AD converters 213 and 214. FIG. 2(b) is a diagram showing a state where each irradiation position of the electron beam 204 is associated with XY coordinates. Detection signals of the secondary electrons and reflection electrons are converted to brightness values of pixels H1 to H6. SE (Secondary Electron) images are generated from the secondary electron signal amounts and BSE (Back Scattered Electron) images are generated from the reflection electron signal amounts. A method of imaging signals acquired by irradiating the electron beam 204 to the silicon wafer 201 has been described above. Next, operations of the electronic device inspection system 1000 will be explained. FIG. 3 is a diagram showing an operation flow of the electronic device inspection system 1000. In the following, each step of FIG. 3 will be explained. It should be noted that each step will be explained anew in detail using figures described below. (FIG. 3: Step S301) The computing unit 215 acquires detection results of the reflection electron detectors 210 and 211 and generates a BSE image for each detector. The number of reflection electron detectors is two in Embodiment 1, and therefore two BSE images are generated in the present step. A state of generating two BSE images is shown anew in FIGS. 4 and 5 described below. (FIG. 3: Step S302) The computing unit 215 combines the two BSE images generated in step S301. A state of generating the synthetic BSE image in the present step will be shown anew in FIG. 6 described later. (FIG. 3: Step S303) The computing unit 215 divides the pixel brightness values of the synthetic BSE image into two or more brightness ranges. To divide the brightness values means to express the brightness intensity by level such that, for example, when the brightness has a minimum value of 0 and a maximum value of 255, brightness values 0 to 85 belong to division 1, brightness values 86 to 170 belong to division 2 and brightness values 171 to 255 belong to division 3. The present step will be shown anew in detail in FIG. 7 described below. It should be noted that the above-described division and the brightness values are just an example for explanation. (FIG. 3: Step S304) The computing unit 215 divides the regions in the synthetic BSE image based on the result of step S303. For example, in the case of the example described in the above-described step S303, regions having the brightness values corresponding to the brightness value ranges of divisions 1 to 3 are divided on the synthetic BSE image. As a result, the synthetic BSE image is divided into three kinds of regions. Next, the computing unit 215 replaces the brightness value of each region in the synthetic BSE image with a representative brightness value (described later) and equalizes it. An example of an image (i.e., region identification image) generated in the present step will be shown anew in FIG. 8 described later. (FIG. 3: Step S305) The computing unit 215 corrects the contrast of SE images based on the result of step S304. To be more specific, the regions of synthetic BSE image divided as a result of step S304 are associated with the regions of SE image and the brightness of each region is corrected such that the boundary between the regions of the SE image can be easily identified. The present step will be explained anew in detail in FIGS. 9 and 10 described later. (FIG. 3: Step S306) The computing unit 215 performs pattern matching using template data such as design data and specifies an inspection position on the SE image. (FIG. 3: Step S307) The computing unit 215 inspects whether an expected circuit pattern is acquired in an inspection position on the SE image specified in step S306, by comparing circuit pattern shapes of the design data and SE image, for example. FIG. 4 is a diagram of the silicon wafer 201 and reflection electron detectors 210 and 211 seen from above. The silicon wafer 201 has a multilayer structure. An upper part 201a is positioned above a lower part 201b (i.e., on a side closer to the reflection electron detectors). Also, a background part, that is, a pattern unformed part of the silicon wafer 201 is indicated by “201c”. The reflection electron detectors 210 and 211 are arranged in different positions on the XY plane of the silicon wafer 201 seen from the above. In the example of FIG. 4, the reflection electron detectors 210 and 211 are arranged in positions observed from the upper left and upper right on the XY plane of the silicon wafer 201. In step S301, the BSE image is acquired via each detector. FIG. 5 shows examples of BSE images and SE image. FIG. 5(a) shows an example of a BSE image acquired via the reflection electron detector 210, FIG. 5(b) shows an example of a BSE image acquired via the reflection electron detector 211 and FIG. 5(c) shows an example of an SE image acquired via the secondary electron detector 209. The BES images generated via the reflection electron detectors 210 and 211 generally have a higher brightness value in a pattern pixel in a higher position and a lower brightness value in a pattern pixel in a lower position. Therefore, in the BSE images shown in FIGS. 5(a) and 5(b), the image of the upper part 201a becomes bright and the image of the lower part 201b becomes dark. The image of the background part 201c becomes darker. The reflection electron detector 210 detects a reflection electron from the upper left direction (on the XY plane) as shown in FIG. 4 so that it is possible to better detect the contrast of observation images seen from the left and above in concavity and convexity patterns of the silicon wafer 201. However, observation images seen from the right and below are behind the concavity and convexity patterns, and therefore the contrast of the BSE images tends to be lower. Similarly, the reflection electron detector 211 detects a reflection electron from the lower right direction (on the XY plane) so that it is possible to better detect the contrast of observation images seen from the right and below in the concavity and convexity patterns of the silicon wafer 201, but observation images seen from the left and above become lower. Thus, taken into account a characteristic that a BSE image denotes an observation image generated by detecting a reflection electron, it can be said that it reflects the concavity and convexity of a circuit pattern well. One SE image is made by collecting secondary electrons released from the surface of the silicon wafer 201, which are acquired by irradiating the electron beam 204 to the silicon wafer 201, using an electrical field caused by the power voltage applied to the secondary electron detector 209, and imaging them. Therefore, it is possible to image the all-around edge used to inspect a circuit pattern without being influenced by pattern concavity and convexity unlike a BSE image. However, in the case of taking a photograph of a circuit pattern with high and low points like a multilayer structure circuit, it is difficult to capture a secondary electron released from a circuit pattern in the lower position. Therefore, the SE image of the circuit pattern in the lower position has a lower contrast than that of the SE image of the circuit pattern in the higher position. FIG. 6 is a diagram showing an example of the synthetic BSE image generated in step S302. The computing unit 215 combines a BSE image acquired via the reflection electron detector 210 and a BSE image acquired via the reflection electron detector 211 using, for example, the following method to generate a synthetic BSE image. By this means, it is possible to acquire a synthetic BSE image in which the convexity part of a circuit pattern has a high brightness value and the brightness of the concavity part is low. (Synthetic BSE Image Generation: Method 1) The computing unit 215 calculates a brightness average value of pixels in the same position of BSE images and provides this average value as a brightness value of a pixel in the same position of a synthetic BSE image. (Synthetic BSE Image Generation: Method 2) The computing unit 215 compares the brightness of pixels in the same position of BSE images to acquire the highest brightness value and provides this highest brightness value as a brightness value of a pixel in the same position of the synthetic BSE image. FIG. 7 is a diagram showing an example of a histogram indicating the brightness value of each pixel of a BSE image and its appearance frequency. Since the BSE image includes pixels having brightness corresponding to concavity and convexity patterns, frequency peaks corresponding to these patterns occur on the histogram. In the cases of the BSE images exemplified in FIGS. 4 to 6, a frequency peak 701 corresponding to the upper part 201a, a frequency peak 702 corresponding to the lower part 201b and a frequency peak 703 corresponding to the background part 201c are illustrated. The same applies to the synthetic BSE image. As shown in FIG. 7, the frequency distribution trend of the histogram in which the frequency peak corresponding to the higher position, the frequency peak corresponding to the lower position and the frequency peak corresponding to the background occur, is a common feature in all BSE images acquired by imaging a pattern with high and low points. In step S304, using the histogram as shown in FIG. 7, the computing unit 215 divides the brightness values of the synthetic BSE image into brightness ranges corresponding to the upper part 201a, the lower part 201b and the background part 201. In FIG. 7, it is assumed that the brightness value corresponding to the boundary between the upper part 201a and the lower part 201b is “704” and the brightness value corresponding to the boundary between the lower part 201b and the background part 201c is “705.” To be more specific, for example, the brightness value of the lowest frequency between the frequency peak 701 and the frequency peak 702 can be regarded as a brightness threshold 704 and the brightness value of the lowest frequency between the frequency peak 702 and the frequency peak 703 can be regarded as a brightness threshold 705. In step S304, using these brightness values 704 and 705 as the threshold, the computing unit 215 divides each pixel of the synthetic BSE image into one of the divisions. Next, the computing unit 215 sets the brightness representative value of each brightness range. For example, it is assumed that the brightness value of the upper pattern 201a is 255, the brightness value of the lower pattern 201b is 128 and the brightness value of the background part 201c is 0. As these representative brightness values, for example, it is possible to use a brightness value of the highest appearance frequency in each division. Next, the computing unit 215 generates an image in which the brightness value of each pixel of the synthetic BSE image is replaced with representative brightness values (hereinafter also called “region identification image”). By this means, the region identification image becomes an image having only three brightness values. Since an actual BSE image includes noise through imaging process, even if the brightness of each pixel is replaced with the representative brightness value as described above, the noise may remain. In this case, by passing the region identification image through a smoothing filter and applying noise reduction processing, it is possible to suppress the noise influence. Also, before it is determined that in which division each pixel of the synthetic BSE image is included, the same noise reduction processing may be applied. FIG. 8 is a diagram showing a state where the contrast of a synthetic BSE image is corrected. In step S304, the computing unit 215 replaces the brightness of pixels belonging to the regions explained in FIG. 7 with representative brightness values and generates a region identification image having only the representative brightness values. By this means, the synthetic BSE image is an image having only three brightness values corresponding to the upper part 201a, the lower part 201b and the background part 201c. FIG. 9 is a diagram showing a detailed flow of step S305 in FIG. 3. Here, although a step of correcting the contrast of the lower part 201b and the background 201c is shown, steps of contrast correction in the other parts can similarly be performed. Each step in FIG. 9 will be explained below. (FIG. 9: Step S901) The computing unit 215 reads the region identification image generated in step S304. (FIG. 9: Step S902) The computing unit 215 overlaps the SE image and the region identification image and identifies the lower part 201b and the background 201c of the SE image. Since the region identification image shows the upper part 201a, the lower part 201b and the background part 201c with good contrast, by overlapping the region identification image and the SE image, it is easily possible to identify each region in the SE image. Regarding position correction in a case where the imaging positions of the SE image and the region identification image are shifted, it will be described later in Embodiment 2. (FIG. 9: Step S903) The computing unit 215 acquires the maximum brightness value and the minimum brightness value in each of the lower part 201b and the background part 201c in the SE image, and uses these values as contrast correction parameters. (FIG. 9: Step S904) The computing unit 215 corrects the maximum brightness values and minimum brightness values of the lower part 201b and the background part 201c in the SE image to “255” and “0,” respectively, to clear the contrast. For example, it is possible to correct the brightness using a correction equation as shown in the following Equation 1. [ Expression 1 ] Dst = 255 ( Src - min ) max - min ( Equation 1 ) whereSrc: brightness value before the SE image is corrected;Dst: brightness value after the SE image is corrected;max: maximum brightness value in the region to which Src pixel belongs; andmin: minimum brightness value in the region to which Src pixel belongs. It should be noted that various contrast correction methods based on image brightness are proposed in addition to the above-noted method and a contrast correction method is not limited to Equation 1. Also, contrast correction parameters for each region may be designated by the user. FIG. 10 is a diagram showing a state where the contrast of the SE image is corrected in step S904. FIG. 10(a) shows the SE image before correction, FIG. 10(b) shows the region identification image and FIG. 10(c) shows the SE image after correction. In the corrected SE image, the upper part 201a, the lower part 201b, the background part 201c and the contrast of the boundaries between these parts are clear. As described above, in Embodiment 1, the computing unit 215 identifies the upper part 201a, the lower part 201b and the background part 201c of the circuit pattern using the BSE image and uses this identification result to correct the contrast of the SE image. Since the BSE image reflects the height of the circuit pattern well, compared to the case where the SE image is used as it is, it is possible to clarify concavity and convexity patterns of the circuit pattern better. Therefore, it is possible to correct the contrast such that the SE image shows the concavity and convexity of the circuit pattern more clearly and accurately so as to make it easier to inspect the circuit pattern using the SE image. Also, in Embodiment 1, in step S304, the computing unit 215 divides brightness values into two or more brightness ranges (in Embodiment 1, three brightness ranges) based on the brightness appearance frequency in the BSE image, and equalizes the brightness by replacing the image region brightness included in each division with a representative brightness value (for example, the most frequent brightness value). By this means, the contrast of the image region included in each brightness range becomes clear, which makes it easier to identify the concavity and convexity of the circuit pattern. Also, in Embodiment 1, using the region identification image generated by dividing the brightness range of the BSE image, the computing unit 215 identifies, for example, the upper part 201a in the SE image and further corrects the contrast such that the concavity and convexity patterns in the SE image becomes clear. By this means, it is possible to accurately inspect a circuit pattern with high and low points using the SE image. Also, in Embodiment 1, the computing unit 215 synthesizes BSE images acquired via two or more reflection electron detectors and generate a synthetic BSE image. By this means, it is possible to clarify a BSE image of a shady part seen from the reflection electron detector. (Step 1) The computing unit 215 performs secondary-differentiation edge detection processing on the region identification image as shown in FIG. 8. (Step 2) The computing unit 215 extracts an edge part shown on an SE image. (Step 3) By overlapping the edge part of the region identification image and the edge part of the SE image and, for example, adding a brightness value, the computing unit 215 evaluates the overlapping degree of the edge parts. (Step 4) The computing unit 215 performs the same processing several times while changing the region size of the region identification image or the brightness value, specifies a position in which the overlapping degree of the edge parts is the largest, and corrects the position of the region identification image or the position of the SE image. (Step 1) The computing unit 215 performs secondary-differentiation edge detection processing on the region identification image as shown in FIG. 8. (Step 2) The computing unit 215 extracts an edge part shown on an SE image. (Step 3) The computing unit 215 creates contour data of an edge part of the region identification image. (Step 4) The computing unit 215 superposes the contour data on the SE image and uses an edge search method such as a snake method to search for a position in which the edge part on the SE image and the contour data overlap with each other, while moving or elongating and contracting the contour data.(Step 5) The computing unit 215 corrects a position of the region identification image or a position of the SE image to the position in which the edge part on the SE image and the contour data overlap with each other. As explained in Embodiment 1, in the case of dividing the brightness range of a BSE image and equalizing the brightness, the brightness of an edge part on the BSE is corrected and a position of the edge part may be shifted. However, in a case where a circuit pattern or a brightness correction method is known, the shift length of the edge part may also be known. Therefore, the shift length is stored in advance in an arbitrary storage apparatus. The computing unit 215 can read the shift length and correct a shift of the edge position of the region identification image or the SE image. As described above, according to Embodiment 2, even in a case where the imaging positions of a region identification image and an SE image are shifted, it is possible to correct these shifts, precisely correlate the region identification image and the SE image and accurately perform contrast correction of the SE image. In the SE image, only an edge part of a circuit pattern is imaged as a pattern shape. Therefore, especially, in the case of inspecting a circuit pattern configured with only a simple line pattern and space pattern, it is difficult to judge whether the edge part of the SE image is the line pattern or it is the space pattern. Meanwhile, in a BSE image, as described above, the concavity and convexity of a pattern are clearly shown as a luminance difference, and therefore a brightness histogram of a synthetic BSE image forms brightness distribution peaks for a line pattern and space pattern. By dividing these two peaks to generate a region identification image and performing pattern matching between an image generated by extracting a region determined as a convexity region and an image acquired by filling the line pattern of the SE image, it is possible to accurately specify an inspection position. In Embodiment 4 of the present invention, a method of inspecting an outcome of a circuit pattern using an SE image will be explained. This corresponds to step S307 in Embodiment 1. Regarding steps S301 to S306, the same method as in Embodiments 1 to 3 can be adopted. A configuration of the electronic device inspection system 1000 is the same as in Embodiments 1 to 3. In the following, differences from Embodiments 1 to 3 will be mainly explained. FIG. 13 is a diagram showing process of inspecting an outcome of a circuit pattern. FIG. 13(a) shows a region identification image of an inspection target pattern, FIG. 13(b) shows an image in which a contour line of FIG. 13(a) has been extracted, and FIG. 13(c) shows a state where the SE image and the contour line are compared. In step S307, the computing unit 215 extracts a contour line showing the outline by an arbitrary method, from the region identification image of the circuit pattern of the inspection target. This contour line is represented by reference numeral “1302” in FIG. 13(b). Next, the computing unit 215 superposes the contour line 1302 on the SE image. At this time, since the pattern matching in step S306 is completed and the inspection position is specified, the contour line 1302 and a circuit pattern 1301 on the SE image should ideally overlap. However, due to differences in the imaging conditions between the BSE image and the SE image, for example, a slight position shift may actually be caused at individual circuit pattern levels. Therefore, the computing unit 215 searches for the circuit pattern 1301 on the SE image within a range in which the position shift in individual circuit patterns can be caused, using the contour line 1302 generated from the region identification image shown in FIG. 13(b) as the starting point. By this means, even at individual circuit pattern levels, it is possible to mitigate influences of the lower part 201b and the background part 201c and accurately specify an inspection position. After specifying the circuit pattern 1301 on the SE image, the computing unit 215 compares a shape of the circuit patter 1301 and a shape of the circuit pattern on the design data to evaluate an outcome of the circuit pattern 1301. As an evaluation method, an arbitrary method can be used. For example, a method of comparing an area of the circuit pattern on the SE image and an area of the circuit pattern on the design data or a method of measuring an interval between the circuit pattern on the SE image and the circuit pattern on the design data is possible. As described above, according to Embodiment 4, by specifying the contour line 1302 of a circuit pattern using a region identification image, even at individual circuit pattern levels, it is possible to mitigate influences of the lower part 201b and the background part 201c and accurately specify an inspection position. By this means, even at individual circuit pattern levels, it is possible to improve inspection accuracy. Each processing flow, which was explained in Embodiments 1 to 4 and performed by the computing unit 215, can be mounted to hardware such as a circuit device that realizes the processing or implemented using software defining a computing apparatus such as a CPU 251 and a microcomputer and its operations. 200: Electron optical system 201: Silicon wafer 201a: Upper part 201b: Lower part 201c: Background part 203: Electron gun 204: Electron beam 205: Capacitor lens 206: Deflector 207: ExB deflector 208: Objective lens 209: Secondary electron detector 210 and 211: Reflection electron detector 212 to 214: A/D converter 215: Computing unit 217: XY stage 223: Storage apparatus 225: Imaging recipe creation unit 230: Design system 251: CPU 252: Image memory 253: Image processing hardware 701 to 703: Frequency peak 704 and 705: Brightness value 1000: Electronic device inspection system 1001 to 1006: Scanning direction 1301: Circuit pattern 1302: Contour line |
|
042254671 | claims | 1. A neutron absorbing article which comprises boron carbide particles in which the boron carbide content is at least 90% by weight and which are substantially all of a size to pass through a No. 20 U.S. Sieve Series screen and a solid, irreversibly cured phenol aldehyde condensation polymer cured to a continuous matrix about the boron carbide particles, with the proportion of boron carbide in the article being such that it contains at least 6% by weight of B.sup.10 from the boron carbide content thereof. 2. A neutron absorbing article according to claim 1, for use in storage racks for spent nuclear fuel, which is operable over a temperature range at which the spent nuclear fuel is stored, withstands thermal cycling from repeated spent fuel insertions and removals and withstands radiation from said spent nuclear fuel for long periods of time without losing desirable neutron absorbing and physical properties, is sufficiently chemically inert in water so as to retain neutron absorbing properties in the event of a leak allowing the entry of water into an enclosure for the neutron absorbing article in a storage rack for spent nuclear fuel and into contact with it, does not galvanically corrode and is sufficiently flexible so as to withstand operational basis earthquake and safe shutdown earthquake seismic events without loss of neutron absorbing capability and other desirable physical properties when installed in such storage rack, in which the born carbide content of the boron carbide particles is at least 94%, such particles contain no more than 2% of B.sub.2 O.sub.3 and in which article the B.sup.10 content is at least 7%. 3. A neutron absorbing article according to claim 2 in plate form, in which the phenol aldehyde condensation polymer is a phenol formaldehyde type polymer, the boron carbide particles contain at least 13% of B.sup.10 and the plate contains at least 8% thereof, the plate contains 60 to 80% of boron carbide particles and 20 to 40% of irreversibly cured phenol formaldehyde type polymer and the density of the plate is in the range of about 1.2 g./cc. to about 2.3 g./cc. 4. A neutron absorbing plate according to claim 3 wherein the boron carbide particle content is from 65 to 80%, the phenol formaldehyde type polymer content is from 20 to 35%, the B.sup.10 content is from 8.5 to 11.5%, the phenol formaldehyde type polymer continuously covers the boron carbide particles at the plate surfaces and fills openings between such particles and the density of the plate is from 1.6 to 2.1 g./cc. 5. A neutron absorbing article according to claim 2, in plate form, in which the thickness is from 0.2 to 1 cm., the width is from 10 to 100 times the thickness and the length is from 20 to 500 times the thickness, the modulus of rupture (flexural) is at least 100 kg./sq. cm. at room temperature, 38.degree. C. and 149.degree. C., the crush strength is at least 750 kg./sq. cm. at 38.degree. C. and 149.degree. C., the modulus of elasticity is less than 3.times.10.sup.5 kg./sq. cm. at 38.degree. C., and the coefficient of thermal expansion at 66.degree. C. is less than 1.5.times.10.sup.-5 cm./cm. .degree. C. 6. A neutron absorbing plate according to claim 5 wherein the phenol formaldehyde type polymer is substantially free of halogens, mercury, lead and sulfur and the boron carbide particles contain no more than 1% of B.sub.2 O.sub.3. 7. A neutron absorbing plate according to claim 6 wherein over 50% of the phenol formaldehyde type polymer is trimethylol phenol formaldehyde polymer, the boron carbide particles contain no more than 2% of iron and the plate is substantially free of filler, plasticizer and solvent. 8. A neutron absorbing plate according to claim 7 wherein the boron carbide particles are of particle sizes such that at least 95% thereof passes through a No. 60 U.S. Sieve Series screen and at least 50% thereof passes through a No. 120 U.S. Sieve Series screen, the boron carbide particle content of the plate is from 65 to 80%, the phenol formaldehyde type polymer content is from 20 to 35%, the B.sup.10 content is from 8.5 to 11.5%, the phenol formladehyde type polymer continuously covers the boron carbide particles at the plate surface and fills openings between such particles and the density of the plate is from 1.6 to 2.1 g./cc. 9. A neutron absorbing plate according to claim 8 consisting essentially of the described boron carbide particles and phenol formaldehyde type polymer. 10. An assembly of a plurality of neutron absorbing articles according to claim 1 in a container for nuclear material positioned at such locations as to absorb neutrons emitted by such nuclear material. 11. An assembly according to claim 10 wherein the neutron absorbing articles are installed in a container for spent nuclear fuel between locations therein where such spent nuclear fuel is stored and out of contact with said fuel. 12. An assembly according to claim 11 of a plurality of neutron absorbing plates according to claim 3 in a storage rack for spent nuclear fuel which is to be stored in elongated vertical volumes in an aqueous medium, with the neutron absorbing plates being positioned between such locations of the volumes of spent nuclear fuel and being kept out of contact with locations for said aqueous medium by enclosures for said neutron absorbing plates. 13. An assembly according to claim 11 of a plurality of neutron absorbing plates according to claim 9 in a storage rack for spent nuclear fuel which is to be stored in elongated vertical volumes in an aqueous medium of acidic or neutral pH, with the neutron absorbing plates being positioned between such locations of the volumes of spent nuclear fuel and being kept out of contact with locations for said aqueous medium by enclosures for said neutron absorbing plates. 14. A method of absorbing neutrons from nuclear material which comprises interposing between such material and its surroundings a neutron absorbing article of boron carbide particles and a solid irreversibly cured phenolic polymer matrix about said particles, with the proportion of boron carbide in the structure being such that it contains at least 6% of B.sup.10 from the boron carbide content thereof. 15. A method according to claim 14 wherein the neutron absorbing article is in plate form, a plurality of such plates is present in a container for nuclear material and the plates contain at least 7% of B.sup.10 from the boron carbide content thereof. 16. A neutron absorbing article which comprises boron carbide particles of a B.sub.2 O.sub.3 content no more than 2% and a solid, irreversible phenolic polymer cured to a continuous matrix about the boron carbide particles. |
abstract | Disclosed is an electromagnetic wave EMI/RFI shielding resin composite material that includes a thermoplastic polymer resin, an electrically conductive filler having a polyhedral shape or being capable of forming a polyhedral shape, and a low-melting point metal, and a molded product made using the EMI/RFI shielding resin composite material. |
|
summary | ||
046559981 | summary | BACKGROUND OF THE INVENTION This invention relates to nuclear reactors, and in particular to liquid metal cooled nuclear reactors. Such reactors are generally of the fast breeder type. A liquid metal cooled nuclear reactor is usually housed in a steel vessel which is itself enclosed within a concrete containment vault, and in order to prevent damage to the concrete due to heat radiated from the vessel containing the reactor core it is usual to interpose thermal insulation between vessel and vault. It is an object of the present invention to provide a construction of thermal insulation capable of protecting the concrete of the vault from heat radiation from the reactor vessel, even under fault conditions. FEATURES AND ASPECTS OF THE INVENTION According to the invention, in a liquid metal cooled nuclear reactor comprising a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault and having thermal insulation interposed between the vessel and the vault, the thermal insulation consists of a plurality of adjoining units each hung from metal structure secured to and projecting from the concrete of the vault, each unit including a pack of thermal insulating material and also including a contained void co-extensive with the said pack, such void being situated between the pack and the concrete and being connected to the void of at least one adjoining unit so as to form a continuous duct or ducts for a fluid coolant. Adjoining units of the thermal insulation are advantageously interleaved or overlapped in order to present a barrier to radiation passing between the units and to minimise leakage of fluid coolant from said duct or ducts. The said metal structure is preferably formed so that the said units are themselves spaced from the concrete of the vault in order to provide a volume in which the environmental atmosphere remains generally static, thereby to assist in thermal insulation. The said metal structure preferably consists of stanchions secured to spaced pads embedded in the concrete of the vault, there being pegs projecting from the stanchions at convenient intervals and from which the said units are hung with a keyhole or other suitable engagement. Conveniently there are two pegs for each unit arranged symmetrically. Each unit is preferably of sheet metal construction and defining the void, with the pack of heat insulation material mounted on that void boundary which is remote from the concrete, and the edges of both metal sheets defining the void and the thermal insulation being arranged so that there is interleaving or over-lapping between adjoining units after assembly. |
046474250 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention: The present invention relates to a process for vacuum degassing and refilling a reactor coolant system and more particularly to a method for reducing radiogas and non-radiogas concentrations in reactor coolant to low residual levels--radiogas referring to the radioactive gas in the reactor system. 2. Background of the Invention: During pressurized water reactor (PWR) plant shutdowns, it is a common practice to draindown the reactor coolant system past the reactor vessel flange to the midplane of the reactor vessel nozzles. That midplane coincides with the midplane of the connecting "hot leg" piping leading to the steam generators. This draindown permits inspection, testing and maintenance, during shutdown, of pumps, steam generators, support structure and the like. During reactor operation, some fission gasses (radiogas) created by fission reactions in the nuclear fuel, will enter the reactor coolant system (RCS) and become dissolved in the reactor coolant. Subsequent to shutdown but before refueling and maintenance operations commence, the radiogas concentration must be reduced to avoid excessive radiation exposure to plant maintenance and inspection personnel. Reactor coolant has previously been degassed using a volume control tank (VCT) connected to the RCS. As used herein the RCS primarily includes such nuclear steam supply system (NSSS) components as the reactor vessel, the steam generators, the reactor coolant pumps and the connecting pipes. The VCT is part of a system known as the chemical and volume control system (CVCS) which operates in the degassing mode by flashing radiogas out of the reactor coolant and into the vapor space of the VCT. An example of such a system is illustrated in the accompanying Figure. Typically, a relatively small flow of reactor coolant is diverted from the RCS and through the CVCS. This stream is first cooled in a letdown heat exchanger to prevent steam from forming when the reactor coolant is subsequently depressurized. The stream is then purified in a mixed bed demineralizer and filtered to remove dissolved ionic or suspended particulate material and passed to the VCT. In the VCT the stream is subjected to a spraying action to remove any dissolved radiogas from solution. The radiogas then collects in the vapor space of the VCT as free gas and is purged to a waste gas system for further processing. Finally the degassed reactor coolant is returned to the RCS with a high pressure charging pump which dilutes the remaining reactor coolant with respect to radiogas. This process is continued until the radiogas concentration of the reactor coolant is compatible with shutdown operations. This method of degassing is undesirably time consuming requiring up to two days for accomplishing the degassing operation. This degassing method also involves an intricate start-up procedure including filling, venting and jogging the reactor coolant pumps multiple times until the coolant level in the pressure vessel is restored to its normal operating level. This is due to the need to continuously vent the pressure vessel as the coolant level is increased so as to prevent undesired gases from being forced into solution with the coolant water and to avoid pumping in two phases through the reactor coolant pumps. An improvement over this CVCS procedure is a vacuum degassing system in which a reactor coolant system is drained approximately to the middle of a hot leg connecting the pressure vessel to the steam generator. This draining is typically accomplished over a slight nitrogen pressure, introduced through a pressurizer, to avoid introducing air and therefore oxygen into the system. A reactor coolant draindown pump as illustrated in the accompanying Figure is generally used for this purpose. After the coolant level is lowered to the midplane of the hot leg nozzle, a vacuum is drawn on the system by removing the nitrogen until the cooling system saturation pressure is reached. This results in boiling of the reactor coolant left in the system causing it to degas. After degassing, refueling and maintenance operations are preformed. Prior to start-up, the vacuum system is used to refill the reactor coolant system under vacuum, thus eliminating the need for the fill-vent-jog cycle of the reactor coolant pumps as described above. This simplified refill procedure is possible as a result of the presence of a vacuum in the system which permits the reactor coolant level to raise without trapping a significant gas bubble in the vapor space. Therefore, there is no need to periodically vent the reactor vessel during refilling. Shen et al, in U.S. Pat. No. 4,187,146 discloses a method and apparatus for reducing radioactive emissions from a nuclear reactor plant which result from leakages of reactor coolant into the secondary liquid in steam generators. One aspect of the invention relates to condensing and decontaminating blowdown tank vapors instead of venting then to the atmosphere. Kausz et al, in U.S. Pat. No. 4,043,865 discloses a PWR coolant treatment system which controls the boron content of the coolant and degasses the coolant during reactor operation. Boron control is effected in a rectification column and degasification is periodically effected as required by a conventional degasifier. Gross et al, in U.S. Pat. No. 3,932,212 discloses a method and apparatus for depressurizing and degassing the condensates of boiling water reactors (BWR). Secondary condensate (from a feedwater preheater) is directed to a relatively high pressure, high temperature chamber and then fed into the primary condensate flow (from the main condenser) whereby the secondary condensate vaporizes in the primary condensate to degas the primary condensate. Kausz et al, in U.S. Pat. No. 3,964,965 discloses a conventional PWR coolant radiogas disposal system which utilizes a conventional degasser and a separator for separating noble gases which can then be stored. Goeldner, in U.S. Pat. No. 3,480,515 discloses a system for the concentration of radioactive materials from reactor coolant. The system disclosed is basically a vapor compression still system. Peake et al, in U.S. Pat. No. 3,210,912 discloses a method and apparatus for removing highly soluble gases such as ammonia from a liquid such as steam generator feedwater. Other non-reactor degassers are disclosed in U.S. Pat. No. 3,342,020 to Ross. Maldague, in U.S. Pat. No. 3,222,255 discloses a method for purifying reactor coolant during reactor operation by separating a small stream of reactor coolant from the RCS and distilling the stream at a pressure substantially the same as the reactor operating pressure to form a vapor of primary fluid and a liquid residue. The vapor is returned to the RCS and the residue discarded. None to the prior art discloses a simple, fast and effective method for degassing reactor coolant after reactor shutdown and which makes use of many existing NSSS components. SUMMARY OF THE INVENTION It is therefore a primary object of the present invention to provide a simple and fast method for degassing and refilling an RCS. It is a further object of the present invention to provide a vacuum degassing procedure which will safely reduce the radiogas and non-radiogas concentration in the reactor coolant to low residual levels in a short period of time. It is a still further object of the present invention to provide a vacuum degassing procedure which utilizes existing reactor equipment and which also enhances residual heat removal (RHR). It is a further object of the present invention to provide a reactor coolant refill procedure which is rapid and which minimizes the need for oxygen scavenging chemicals such as hydrazine to be added to the reactor coolant during refill operations to remove dissolved oxygen. To achieve the foregoing and other objects and in accordance with the purpose of the present invention, as embodied and broadly described herein, the invention may comprise a method for vacuum degassing a reactor coolant system having reactor coolant in a reactor pressure vessel connected to at least one steam generator by a hot leg. The method comprises draining down the reactor coolant system to approximately the midpoint of the hot leg and maintaining the reactor coolant system in an unvented condition during the draindown operation. Any flashed reactor coolant in the primary side of the steam generator is then refluxed. As used herein, flashed reactor coolant means liquid coolant which flashes into the steam phase as a result of lowered ambient pressure. The bulk of the reactor coolant as well as the refluxed reactor coolant is circulated through a residual heat removal system to cool the reactor coolant. A vacuum is drawn on the reactor coolant system to evacuate any gas stripped from the reactor coolant. Preferably, the step of draining the coolant system further comprises using a two phase pump to establish a partial vacuum in the unvented reactor coolant system during draindown. The partial vacuum should be sufficient to cause the reactor coolant to boil at the prevailing temperatures in the reactor coolant system whereby degassing occurs during the draindown step. Preferably, the step of refluxed comprises flowing a secondary coolant through a secondary side of the steam generator so that any flashed reactor coolant in the primary side of the steam generator is condensed back into a liquid and any non-condensible gases may be stripped away by the vacuum system. Preferably, the heat removal system used utilizes preexisting heat removal equipment located in the secondary or steam side of the plant. Preferably, the step of drawing a vacuum is performed simultaneous to draining down the reactor coolant system and after the heat removal step is operating. The step of drawing a vacuum may be performed utilizing an existing reactor waste gas removal system or may involve the use of a dedicated waste gas system. It is also preferred that the circulating reactor coolant in the residual heat removal system be sampled and that the vacuum be maintained until a predetermined level of gas concentration is detected during said sampling. After the proper level of gas concentration is detected, the vacuum is preferably broken by admitting air into the circulating system. The oxygen in the air dissolves in the reactor coolant, thus facilitating the solubilization of radioactive material that may subsequently be removed by ion-exchange in a CVCS demineralizer. Removal of the radioactive material at this point by deliberate aeration advantageously prevents later delays in operations should aeration occur in an uncontrolled manner. Preferably, when the vacuum is broken and the steam generators cease operation as reflux condensers, the circulation through the residual heat removal pump is increased in order to support the increased heat load. It is also preferred that after degassing, the reactor coolant system be refilled under vacuum conditions to eliminate the time consuming operation of jogging the reactor coolant pumps and venting the reactor coolant system multiple times. Vacuum refilling has the additional significant advantage in that the amount of oxygen which must be removed by the addition of hydrazine to the coolant is reduced thereby requiring less hydrazine, saving time and cost. |
040509877 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Throughout the description which follows, like reference characters indicate like elements of various figures of the drawings. FIG. 1 of the drawing illustrates a typical nuclear reactor system which can employ the principles of this invention. An elongated nuclear reactor pressure vessel 10 contains a quantity of reactor coolant such as liquid sodium graphically illustrated and designated as the Numeral 12. The top of the pressure vessel 10 is sealed by a nuclear reactor pressure vessel closure head 14. In the space formed by the pressure vessel 10 and the closure head 14 and above the level of liquid coolant 12 is installed an inert gas known generally as cover gas 16. The head 14 has a stationary outer ring 18 which is bolted to a flange 20 of the pressure vessel 10. The head 14 has a plurality of generally planar, cylindrical rotating members such as plugs 22, 24 and 26. The plug 22, of the largest diameter, is coaxial with the vessel 10. The stationary member, ring 18, has an opening 19 in which the large plug 22 is positioned. The large plug 22 has an opening 23, excentric to the large plug axis 37 in which the intermediate plug 24 is positioned. The intermediate plug 24 has an opening 25, excentric to the intermediate plug axis 39 in which the small plug 26 is positioned. These rotating plugs 22, 24, 26 position the fuel and control handling equipment over all desired vessel locations. The plug 26 supports an in-vessel transfer machine plug 28 excentrically. The plug 22 supports the ex-vessel transfer machine plug 30 excentrically. The plug 24 supports columns 32 which support the upper internals (not shown), the control rod assembly mechanisms 34, and one or more sealed surveillance ports 36. By rotating the plugs 22, 24, 26, the in-vessel transfer machine plug 28 can be positioned over the various components in the vessel, and over the ex-vessel transfer machine plug 30. The small plug 26 is connected to and supported by plug 24 through the load structure 42. The intermediate plug 24 is connected to and supported by plug 22 through the load structure 40. The large plug 22 is connected to and supported by the stationary outer ring 18 through the load structure 38. FIG. 2, a top view of the closure head 14, illustrates the rotation of these plugs 22, 24 and 26. The smallest diameter plug 26 can rotate about its axis 41. The intermediate plug 24 can rotate around its axis 39 and the small plug 26 remains in its position on plug 24. The large plug 22 can rotate around its axis 37 while both plugs 24 and 26 remain in their position on plug 22. In this manner, the small plug 26 and its in-vessel transfer machine plug 28 can be positioned over any desired vessel locations. Reference is now made to FIG. 3, which shows a detailed view of the load structure 38 between the stationary outer ring 18 and the largest diameter plug 22. This load structure 38 is illustrated for descriptive purposes, and it would be obvious to one skilled in the art that a similar structure may be employed as load structures 40 and 42. Likewise, the asimilar structure may be used in conjunction with any rotating member which penetrates the closure head 14. The stationary ring 18 has an annular supporting structure 44 rising vertically above the top edge 46 of the stationary ring 18. The supporting structure 44 is connected to the inner race 48 of the bearing 50. The inner race 48 and its associated bearing 50 are secured to the supporting structure 44 by means of a ring 52. The inner side 54 of the annular supporting structure 44, contains cavities 56 into which maintenance seals (not shown) can be inserted during maintenance operations. The large plug 22 has an annular support riser 58 rising vertically above the top edge 60 of the large ring 22. The support riser 58 has an extension 59 comprised of an upper flange 62 which fits over the top side 64 of the supporting member 44 of the stationary ring 18, and an annular sealing structure 66 which is hermetically secured to the flange 62 of the support riser 58 and extends vertically downward from the upper flange 62 adjacent to the outer side 68 of the supporting structure 44. The support riser 58 and the extension 59 form a generally U-shaped first annular space 63 in which the support structure 44 is positioned. The outer side 68 of the supporting structure 44 and the edge 70 of the sealing structure 66 (part of the extension 59) form a second annular space 84. Three elements, the support riser 58, the upper flange 62, and the sealing structure 66 form the support structure 61 of the large plug 22. While support structure 61 can be manufactured as one piece, for ease of installation and maintenance the preferred method is to use the three separate elements 58, 62, and 66. Throughout the following description, it will be assumed that the support structure 61 is comprised of the three elements 58, 62 and 66, although the invention is equally applicable to a one-piece structure. The edge 70 of the sealing structure 66 has cavities 72 above the elevation of the bearing 50 into which sealing means such as inflatable annular seals 74 are placed. Lubricating means 76 are connected on one side to the inflatable seals 74, and are accessible to the exterior 78 of the sealing structure 66. The sealing structure 66 is secured to, and supported by, the outer race 80 of the bearing 50. A lubricant collector 82 is located beneath the bearing 50, the inner race 48 and the outer race 80. This lubricant collector 82 is shown in the drawing as being connected to outer race 80, although alternate locations would have it connected to the inner race 48 or the support structure 44. During rotation of the plug 22, the support structure 61, shown as support riser 58 flange 62, and sealing structure 66, the outer race 80, the seals 74 and the lubricant collector 82, if so connected, rotate around the large plug axis 37. The stationary ring 18 with its supporting structure 44, the inner race 48, and the ring 52 remain stationary. The plug 22 is supported above the reactor vessel 10 during this rotation by the stationary ring 18 through the bearing 50. During lubrication, lubricant is inserted through the lubricating means 76 to the seals 74. The lubricant then flows down the annulur space 84 between the external edge 68 of the supporting structure 44 and the edge 70 of the sealing structure 66 to the bearing 50. Any excess lubrication is then caught in the lubricant collector 82. If the lubricant collector 82 becomes full, the lubricant could be removed by means of a drain hole (not shown). Although not shown on the drawing, if means for lubricating the bearing 50 directly are desired, such means can be inserted through the outer race 80. Any excess lubricant from this lubrication would also be caught in the lubricant collector 82. During maintenance operations, maintenance seals (not shown) are inserted into the cavities 56 provided for them in the supporting structure 44 of the stationary ring 18. These seals prevent the escape of any gases from the reactor cavity 16. Then the sealing structure 66 and the bearing 50 are removed from their positions. The seals 74 and the bearing 50 can then be replaced. Thus, it can be seen that this invention teaches a system by which seals and bearings for rotating members above the reactor vessel of nuclear reactors can be lubricated without risk of having lubricant drain into the reactor vessel. |
claims | 1. A control rod motion monitoring system for a reactor comprising:a plurality of neutron detector assemblies arranged in a radial direction of the core, each assembly including a plurality of neutron detectors, wherein the neutron detectors of each of the different neutron detector assemblies are located at a plurality of predetermined heights in the axial direction of the core, the plurality of predetermined heights being substantially the same for each of the different neutron detector assemblies and including a first predetermined height and a plurality of other predetermined heights;a signal processing device which calculates average values of neutron fluxes continuously measured by the neutron detectors for each of the plurality of predetermined heights; andan arithmetic device which outputs a signal based on the average values calculated by the signal processing device,whereinthe arithmetic device calculates deviations of the calculated average values between the first predetermined height and the plurality of other predetermined heights and outputs the signal when a maximum value of the deviations exceeds a predetermined set point, andthe signal processing device monitors all of the neutron detector assemblies in the core and constantly monitors control rod insertion throughout the core to detect abnormalities. 2. The control rod motion monitoring system according to claim 1, whereinthe arithmetic device is configured to transmit the signal when the average values calculated by the signal processing device exceed another predetermined set point. 3. The control rod motion monitoring system according to claim 1, whereina signal from a neutron detector assembly, of the plurality of neutron detector assemblies, at an outermost peripheral portion of the core is excluded from the average values calculated by the signal processing device. 4. A control rod motion monitoring system for a reactor comprising:a plurality of neutron detector assemblies arranged in a radial direction of the core, each assembly including a plurality of neutron detectors, wherein the neutron detectors of each of the different neutron detector assemblies are located at a plurality of predetermined heights in the axial direction of the core, the plurality of predetermined heights being substantially the same height for each of the different neutron detector assemblies and including a first predetermined height and a plurality of other predetermined heights;a signal processing device which calculates average values of neutron fluxes continuously measured by the neutron detectors for each of the plurality of predetermined heights; andan arithmetic device which outputs a signal based on the average values calculated by the signal processing device, whereinthe arithmetic device calculates a ratio between the calculated average value for the first predetermined height relative to the calculated average values for the plurality of other predetermined heights and outputs the signal when the ratio exceeds a predetermined set point, andthe signal processing device monitors all of the neutron detector assemblies in the core and constantly monitors control rod insertion throughout the core to detect abnormalities. 5. The control rod motion monitoring system according to claim 4, wherein the arithmetic device is configured to transmit the signal when the average values calculated by the signal processing device exceed another predetermined set point. 6. The control rod motion monitoring system according to claim 4, whereina signal from a neutron detector assembly, of the plurality of neutron detector assemblies, at an outermost peripheral portion of the core is excluded from the average values calculated by the signal processing device. |
|
039716970 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawing there is shown a target assembly which extends into a beam duct 12 of a high energy accelerator. It is contemplated the target assembly 10 may be used with the LOS ALAMOS MESON FACILITY, known as LAMF. Another high energy accelerator located in Canada, called TRIUMF, would be a suitable source of high energy particles. The beam in the duct 12 from such accelerators is characterized by high energy between 200 MeV and 2000 MeV and high current. The target assembly 10 is in the form of a heat pipe which comprises a tubular container 14 having a supply 16 of cesium-133 in the end which extends into the beam duct 12. The opposite end of the tube 14 which protrudes from the duct 12 is surrounded by cooling coils 18. A suitable cooling fluid, such as water, is circulated through the coils 18. A porous metal plug 20 is mounted in the container 14 adjacent to the cooling coils 18. A wick 22 extends along the inner wall of the tube 14 between the plug 20 and the cesium 16. A tube 24 connects the inside of the container 14 to a cold trap 26 through a valve 28. Tygon tubing has been satisfactory for the tubular conduit 24. The cold trap 26 comprises a U-tube 30 immersed in a coolant in an insulated container 32. A one-fourth inch copper U-tube surrounded by solid CO.sub.2 in a Dewar has been satisfactory. The dry ice maintains a trap 26 at a temperature of -79.degree.C. A valve 34 connects the dry ice trap 26 to a second cold trap 36. A one-fourth inch copper U-tube 38 immersed in liquid nitrogen in a Dewar 40 has been satisfactory for the cold trap 36. The liquid nitrogen maintains the trap 36 at a temperature of -196.degree.C. A valve 42 is used to isolate the trap 36 or connect it to a vacuum pump 44. In operation, a beam of high energy protons identified by the arrow in the duct 12 penetrates the tubular container 14 and strikes the cesium-133 in the target assembly 10. The beam penetrating the cesium causes what is known in nuclear physics as a spallation reaction which produces .sup.123 Xe according to the reaction .sup.133 Cs (p,p 10n).sup.123 Xe. This is only one of a number of reactions that lead to significant impurities. Some of these reactions are .sup.133 Cs (p, 2p 8n).sup.124 I, .sup.133 Cs (p, 2p 7n).sup.125 I, .sup.133 Cs(p, 2p 6n).sup.126 I, .sup.133 Cs (p, 3p 8n).sup.123 Te, .sup.133 Cs (p, 4p 6n).sup.124 Sb. To produce these spallation reactions the incident proton must have an energy greater than 200 MeV. The first three reactions produce the radioactive iodines .sup.124 I, .sup.125 I and .sup.126 I. These radioactive iodines would seriously contaminate the desired .sup.123 I because they cannot be chemically separated. The other impurities formed by the above listed reactions could be separated chemically because they are different elements. However, this is not necessary in the heat pipe device because all of the impurities have a lower vapor pressure than the desired .sup.123 Xe and can be collected on cool surfaces at the heat rejection end of the heat pipe. Radioactive isotopes of iodine, tellurium, antimony, tin, indium, and cesium are all contaminants. All of the isotopes of hydrogen and helium could be accelerated to hundreds of MeV and produce the desired spallation reaction. An example would be .sup.133 Cs(.alpha., 3p 11n).sup.123 Xe. The same contaminants as produced by proton bombardment would also be produced. The beam penetrating the cesium-133 deposits energy in the cesium that heats this target to the point where it vaporizes. All charged particle beams lose energy by ionizing and exciting electrons on the atoms of the stopping material, such as cesium. The individual particles that make up the beam actually lose velocity in a continuous manner. This energy lost by the beam appears as heat, and if the beam current is large enough this heat will melt metallic cesium and vaporize it. The vapor is transported to the end of the tubular chamber 14 where it is cooled by the cooling coils 18. The temperature at the hot end of the heat pipe is 670.degree.C which is the boiling point of cesium. The temperature at the cold end is above the melting point and below the boiling point of cesium. This temperature must be above the melting point for the apparatus to be suitable for its intended use because the cesium must condense as a liquid thus giving up the heat of vaporization. Then as a liquid the cesium flows back to the hot end of the heat pipe. It is not difficult to achieve this condition in practice because the cesium vapor column and the cesium condensed on the wall of the heat pipe make the column nearly isothermal. In one embodiment it is possible to have a sharp temperature drop when the column is run as a two component heat pipe where one phase is a non-condensible gas. Even in this embodiment it is not difficult to achieve heat pipe operation although there may be some solid cesium on the walls. The cesium condenses at this cool end of the tube. The beam power that was deposited in the cesium is rejected to the coolant that flows in the coils 18. Some small amount of cesium vapor may be transported through the plug. However, most of the cesium vapor will be collected on the cool walls because the vapor has a greater opportunity to contact the cool walls than the plug. The .sup.123 Xe, .sup.125 Xe, .sup.123 I, .sup.124 I, .sup.125 I, .sup.126 I, and .sup.129 I will pass through the plug plus smaller amounts of other contaminants. All these elements passing the plug will be subsequently stopped by cooler surfaces except for xenon. Xenon will not be collected until it is pure and free of contaminants. The .sup.123 Xe and other volatile contaminants also travel to the cool end of the tube 14 where they pass through the tube 24 and valve 28 to the cold trap 26. The porous metal plug 20 prevents accidental transport of liquid cesium into the trap 26 which might take place where boiling occurs. The vapors of radioactive contaminants condense in the trap 26. The .sup.123 Xe is still a vapor and passes to the trap 36 at liquid nitrogen temperature. The .sup.123 Xe condenses in the trap 26. The removal of the xenon from the vapor phase produces a pumping action that causes almost all the xenon that is produced to be transported to the trap 36. The cesium vapors that condense in the heat pipe at the coils 18 are transported back to the target area by capillary action of the wick 22. It is also contemplated that grooves in or on the inner wall of the tubular container 14 may be used to transport the condensed cesium vapors back to the target area. It is apparent the target assembly 10 uses cesium as both a heat pipe working material and as a target material for the production of radioisotopes by high energy charged particles. Cesium -133 is the preferred working material because of the uniqueness of its heat-transfer properties, such as enthalpy, melting point, boiling point and vapor pressure-temperature curve. No other element has these properties together with the nuclear requirements as required by the target material. Although the preferred embodiment has been shown and described it is contemplated that various structural modifications may be made to the disclosed apparatus without departing from the spirit of the invention or the scope of the subjoined claims. By way of example, it is contemplated that different trap configurations could be utilized. The trap 26 could be a chemical trap, such as hot silver, to remove radioiodine impurities. The .sup.123 Xe trap 36 could work on the absorption principle. The trap 34 could also contain a pharmaceutical compound that would become tagged or labeled when the xenon decays to iodine. It is generally desirable to select a pharmaceutical that goes to a specific organ where it is involved in some metabolic process. Hippuran goes to the kidney and has been tagged with .sup.123 I by placing it on the walls of a container and condensing .sup.123 Xe on these same walls. Other pharmaceuticals that could be labeled are human serum albumin, cholesterol, dopamine and fibrinogen. |
abstract | An apparatus for processing a radionuclide including a parent radionuclide that decays over time into a daughter radionuclide, a separation column that separates the daughter radionuclide from the parent radionuclide, a plurality of valves and at least one pump that operate to separate the daughter radionuclide from the parent radionuclide and deliver the daughter radionuclide into the daughter radionuclide container by alternately connecting at least two of the parent radionuclide container, the daughter radionuclide container, the separation column container and the plurality of processing containers, a plurality of RFID tags including an RFID tag of the plurality of RFID tags affixed to each of the daughter radionuclide container and the separation column and a programmed processor that reads an identifier of each of the plurality of RFID tags, an identifier and position of each of the plurality of valves and pump and saves the identifiers, positions and operations into a tracking file. |
|
summary | ||
claims | 1. An illumination system, particularly for microlithography with wavelengths xe2x89xa6193 nm, comprising: a primary light source; a first optical component; a second optical component; an image plane; and an exit pupil, wherein said first optical component transforms said primary light source into a plurality of secondary light sources that are imaged by said second optical component in said exit pupil, wherein said first optical component includes a first optical element having a plurality of first raster elements that are imaged into said image plane, producing a plurality of images being superimposed, at least partially, on a field in said image plane, wherein said plurality of first raster elements deflect a plurality of incoming ray bundles to produce a plurality of deflected ray bundles with first deflection angles, and wherein at least two of said first deflection angles are different from one another. 2. The illumination system according to claim 1 , claim 1 wherein said plurality of incoming ray bundles has a plurality of incoming centroid rays, wherein said plurality of deflected ray bundles has a plurality of deflected centroid rays, wherein each one of said plurality of deflected centroid rays corresponds to one of said plurality of incoming centroid rays thus defining a plurality of planes of incidence, and wherein at least two of said plurality of planes of incidence intersect each other. 3. The illumination system according to claim 1 , wherein each of said plurality of first raster elements corresponds to one of said plurality of secondary light sources, and wherein each of said plurality of first raster elements deflects one of said plurality of incoming ray bundles to said corresponding one of said plurality of secondary light sources. claim 1 4. The illumination system according to claim 1 , wherein said plurality of first raster elements are mirrors being tilted to generate said first deflection angles. claim 1 5. The illumination system according to claim 1 , wherein said plurality of first raster elements are lenses comprising a prismatic optical power to generate said first deflection angles. claim 1 6. The illumination system according to claim 1 , wherein said first optical component comprises a collector unit for collecting a plurality of rays generated by said primary light source and for directing said plurality of rays to said first optical element. claim 1 7. The illumination system according to claim 6 , wherein said collector unit has positive optical power to generate an intermediate image of said primary light source between said collector unit and said first optical element. claim 6 8. The illumination system according to claim 6 , wherein said collector unit has positive optical power to generate a converging ray bundle between said collector unit and said first optical element. claim 6 9. The illumination system according to claim 8 , further comprising: claim 8 a collector object plane of said collector unit in which said primary light source is located; a collector image plane conjugated to said collector object plane; an image-side principal plane of said collector unit; a first optical distance between said image-side principal plane and said collector image plane; a second optical distance between said image-side principal plane and said first optical element; and a third optical distance between said first optical element and a plane with said secondary light sources, wherein said first optical distance is approximately a sum of said second and third optical distances. 10. The illumination system according to claim 1 , wherein said plurality of first raster elements are plane mirrors. claim 1 11. The illumination system according to claim 1 , wherein said plurality of first raster elements are prisms. claim 1 12. The illumination system according to claim 1 , wherein said plurality of first raster elements are concave mirrors. claim 1 13. The illumination system according to claim 1 , wherein said plurality of first raster elements are convex mirrors. claim 1 14. The illumination system according to claim 1 , wherein said plurality of first raster elements are lenses having a positive optical power. claim 1 15. The illumination system according to claim 1 , wherein said plurality of first raster elements are lenses having a negative optical power. claim 1 16. The illumination system according to claim 1 , claim 1 wherein said plurality of first raster elements are arranged in a plurality of rows, wherein each of said plurality of rows includes at least one of said plurality of first raster elements, and wherein at least one of said plurality of rows is displaced relative to an adjacent row. 17. The illumination system according to claim 1 , claim 1 wherein said plurality of first raster elements are arranged in a two-dimensional array having an area being illuminated, and wherein 90% of said plurality of first raster elements are arranged completely inside said area. 18. The illumination system according to claim 1 , claim 1 wherein said first optical component further comprises a second optical element having a plurality of second raster elements, wherein one of said plurality of first raster elements corresponds to one of said plurality of second raster elements, and wherein said one of said plurality of first raster element deflects one of said plurality of incoming ray bundles to said corresponding one of said plurality of second raster elements. 19. The illumination system according to claim 18 , wherein said plurality of second raster elements are located at a distance from said plurality of secondary light sources ranging from 0% to 10% of a distance between said plurality of first raster elements and said plurality of second raster elements. claim 18 20. The illumination system according to claim 18 , wherein said plurality of second raster elements and said second optical component image said corresponding first raster elements into said image plane. claim 18 21. The illumination system according to claim 18 , wherein said plurality of second raster elements are concave mirrors. claim 18 22. The illumination system according to claim 18 , wherein said plurality of second raster elements are lenses with positive optical power. claim 18 23. The illumination system according to claim 18 , wherein said plurality of second raster elements deflects said plurality of incoming ray bundles with second deflection angles to superimpose said plurality of images, at least partially, on said field. claim 18 24. The illumination system according to claim 23 , wherein at least two of said second deflection angles are different from one another. claim 23 25. The illumination system according to claim 23 , wherein said plurality of second raster elements are tilted planar mirrors. claim 23 26. The illumination system according to claim 23 , wherein said plurality of second raster elements are tilted concave mirrors. claim 23 27. The illumination system according to claim 23 , wherein said plurality of second raster elements are prisms. claim 23 28. The illumination system according to claim 23 , wherein said plurality of second raster elements are lenses having a prismatic optical power and a positive optical power. claim 23 29. The illumination system according to claim 18 , wherein at least two of said plurality of first raster elements are adjacent to one another and have two corresponding second raster elements, and wherein at least another one of said plurality of second raster elements is arranged between said two corresponding second raster elements. claim 18 30. The illumination system according to claim 18 , wherein a distance between individuals of said plurality of second raster elements is irregular. claim 18 31. The illumination system according to claim 18 , claim 18 wherein said plurality of first raster elements are arranged in a first two-dimensional array, wherein said plurality of second raster elements are arranged in a second two-dimensional array, wherein said first array has a first extent in a direction, wherein said second array has a second extent in said direction, and wherein said first and second extents are equal within a range of xc2x115%. 32. The illumination system according to claim 1 , further comprising a masking unit for changing an illumination mode, wherein said masking unit is arranged at said plurality of secondary light sources. claim 1 33. The illumination system according to claim 1 , wherein said field is a segment of an annulus. claim 1 34. The illumination system according to claim 1 , wherein said plurality of first raster elements are rectangular. claim 1 35. The illumination system according to claim 34 , wherein said rectangular first raster elements have an aspect ratio greater than 5:1. claim 34 36. The illumination system according to claim 33 , wherein said second optical component comprises a first field mirror for shaping said field to said segment of said annulus. claim 33 37. Illumination system for microlithography with wavelengths xe2x89xa6193 nm, comprising: a primary light source; a first optical component; a second optical component; an image plane; and an exit pupil, wherein said first optical component transforms said primary light source into a plurality of secondary light sources that are imaged by said second optical component in said exit pupil, wherein said first optical component includes a first optical element having a plurality of first raster elements that are imaged into said image plane, producing a plurality of images being superimposed at least partially, on a field in said image plane, wherein said plurality of first raster elements are rectangular, wherein said field is a segment of an annulus, and wherein said second optical component includes a first field mirror for shaping said field to said segment of said annulus. 38. The illumination system according to claim 37 , wherein said first field mirror has negative optical power. claim 37 39. The illumination system according to claim 37 , wherein said first field mirror is an off-axis segment of a rotational symmetric reflective surface. claim 37 40. The illumination system according to claim 37 , wherein said first field mirror is an on-axis segment of a toroidal reflective surface. claim 37 41. The illumination system according to claim 37 , wherein each of a plurality of rays intersects said first field mirror with an incidence angle of greater than 70xc2x0. claim 37 42. The illumination system according to claim 37 , wherein said second optical component comprises a second field mirror with positive optical power. claim 37 43. The illumination system according to claim 42 , wherein said second field mirror is an off-axis segment of a rotational symmetric reflective surface. claim 42 44. The illumination system according to claim 42 , wherein said second field mirror is an on-axis segment of a toroidal reflective surface. claim 42 45. The illumination system according to claim 42 , wherein each of a plurality of rays intersects said second field mirror with an incidence angle of less than 25xc2x0. claim 42 46. The illumination system according to claim 42 , wherein said second optical component comprises a third field mirror. claim 42 47. The illumination system according to claim 46 , wherein said third field mirror has negative optical power. claim 46 48. The illumination system according to claim 47 , wherein said first, second and third field mirrors form (a) a telescope objective with a tele-object plane at said plurality of secondary light sources, (b) a tele-pupil plane at said image plane of said illumination system, and (c) a tele-image plane at said exit pupil. claim 47 49. The illumination system according to claim 46 , wherein said third field mirror has positive optical power. claim 46 50. The illumination system according to claim 49 , claim 49 wherein said third field mirror images said plurality of secondary light sources in a plane between said third field mirror and said second field mirror forming a plurality of tertiary light sources, and wherein said second field mirror and said first field mirror images said plurality of tertiary light sources in said exit pupil. 51. The illumination system according to claim 50 , further comprising a masking unit for changing an illumination mode, wherein said masking unit is arranged at said plurality of tertiary light sources. claim 50 52. The illumination system according to claim 46 , wherein said third field mirror is an off-axis segment of a rotational symmetric reflective surface. claim 46 53. The illumination system according to claim 46 , wherein said third field mirror is an on-axis segment of a toroidal reflective surface. claim 46 54. The illumination system according to claim 46 , wherein each of a plurality of rays intersects said third field mirror with an incidence angle of less than 25xc2x0. claim 46 55. The illumination system according to claim 46 , wherein said first, second and third field mirrors form a non-centered system. claim 46 56. The illumination system according to claim 1 , claim 1 wherein said light source produces a beam cone oriented in a first direction, wherein said image plane has a surface normal that is substantially perpendicular to said first direction, wherein said first optical component comprises at least one first mirror, and wherein said second optical component comprises at least one second mirror, said illumination system having a beam path between said primary light source and said image plane that is bent with said at least one first mirror and said at least one second mirror. 57. The illumination system according to claim 1 , claim 1 wherein said first optical component comprises a collector unit and a second optical element having a plurality of second raster elements, said illumination system further comprising: a first beam path between said collector unit and said first optical element, wherein said first optical element is reflective; a second beam path between said first optical element and said second optical element, wherein said second optical element is reflective; and a third beam path between said second optical element and said second optical component, wherein said first and second optical elements are tilted to cause a crossing of said third beam path and said first beam path. 58. The illumination system according to claim 1 , further comprising: claim 1 a straight line from a center of said field in said image plane to a center of said exit pupil; and an angle between said straight line and a surface normal of said image plane, wherein said angle is between 3xc2x0 and 10xc2x0. 59. The illumination system according to claim 1 , claim 1 wherein said first optical component and said second optical component comprise only mirrors, and wherein each of a plurality of rays intersects said mirrors with incidence angles of greater than 65xc2x0 or less than 25xc2x0. 60. The illumination system according to claim 1 , wherein said second optical component comprises an even number of normal incidence mirrors having incidence angles of less than 25xc2x0. claim 1 61. The illumination system according to claim 1 , wherein said first optical element is arranged in a divergent beam path. claim 1 62. The illumination system according to claim 1 , wherein said first optical element is arranged in a convergent beam path. claim 1 63. A projection exposure apparatus for microlithography comprising: the illumination system of claim 1 ; claim 1 a reticle being located at said image plane; a light-sensitive object on a support system; and a projection objective to image said reticle onto said light-sensitive object. 64. The projection exposure apparatus of claim 63 , further comprising: claim 63 an illumination beam path between said primary light source and said reticle that passes through said first optical component and said second optical component; and a projection beam path between said reticle and said light-sensitive object that passes through said projection objective, wherein said illumination beam path and said projection beam path do not cross. 65. The projection exposure apparatus of claim 63 , further comprising: claim 63 a projection beam path between said reticle and a first imaging element of said projection objective, wherein said reticle is reflective, and wherein said projection beam path converges towards an optical axis of said projection objective. |
|
abstract | A method for fabricating a collimator includes mixing an x-ray absorbent material with at least one of a temporary binder and a temporary gel, and extruding the mixed x-ray absorbent material through a die to form a unitary collimator structure that is at least one of substantially honeycomb in shape and substantially rectangular in shape. |
|
summary | ||
claims | 1. A method of monitoring delamination of a film from a target comprising:projecting light at said target while the target is being processed, wherein processing of the target comprises ion implanting;detecting said light reflected from said target and measuring the reflectivity of said target;comparing said measured reflectivity to a first expected reflectivity;determining presence of bubbles forming in said target based on said comparison to said first expected reflectivity;continuing to implant ions after said determining of said presence of bubbles;comparing said measured reflectivity to a second expected reflectivity; anddetermining presence of delamination of said film from said target during the processing of the target based on said comparison to said second expected reflectivity. 2. The method of claim 1, wherein said projected light comprises a plurality of wavelengths. 3. The method of claim 2, wherein said measuring and comparing steps are performed for a plurality of wavelengths. 4. The method of claim 1, wherein said comparing steps comprise detecting an increase or decrease in amplitude at one or more frequencies. 5. The method of claim 1, further comprising repeating said projecting, measuring, comparing and determining a plurality of times. 6. The method of claim 1, wherein said processing is terminated in response to said determining presence of delamination of said film step. 7. The method of claim 1, wherein said processing is ion implanting into said target during said measuring, comparing and determining steps. 8. The method of claim 7, wherein said ion implanting step is terminated in response to said determining presence of delamination of said film step. 9. The method of claim 1, wherein said target comprises a semiconductor wafer. 10. The method of claim 1, further comprising scanning light in at least one direction. 11. A method for processing a substrate, the method comprising:performing a first process on the substrate, wherein the first process comprises implanting ions into said substrate so as to form at least one gas bubble between two portions of the substrate;directing radiation toward the substrate during the performing the first process;receiving radiation directed toward the substrate during the performing the first process and reflected from the substrate;detecting, based on said reflected radiation, that one of said portions has delaminated from a second of said portions; andterminating the first process based upon said detection. 12. The method of claim 11, further comprising performing a second process on the substrate, wherein the second process is performed after detecting the endpoint of the first process. 13. The method of claim 12, wherein the second process comprises separating the two portions of the substrate. 14. The method of claim 11, further comprising:determining, before said detection, based on said reflected radiation, that said at least one gas bubble has formed; andcontinuing to implant ions into said substrate after said determination and before said detection. |
|
summary | ||
044977700 | claims | 1. A storage structure for storing waste of nuclear plants comprising a plurality of tubular waste storage members, each of said plurality of tubular members being adapted to form one or a plurality of waste storing chambers wherein (a) each of said plurality of tubular storage members is provided near the bottom thereof with at least three horizontal bolts for locking and supporting the storage members in a fixed position for storing waste of all nuclear plants the longitudinal axes of the bolts intersecting at a common point of said axes; wherein (b) a plurality of support plates for each of said plurality of tubular storage members is provided at the lower end of each of said plurality of storage members; wherein (c) each of said plurality of support plates has a plurality of upstanding portions equal in number to the number of bolts on one of said plurality of storage tubes; wherein (d) each of said plurality of upstanding portions of said plurality of support plates is provided with an open recess for insertion into one of said plurality of bolts; wherein (e) each recess is wide at the open end thereof and bounded by downward slanting surfaces of decreasing spacing; and wherein (f) each recess has a lower closed end which forms a substantially circular bolt-bearing surface coaxial with one of said plurality of bolts (g) wherein the longitudinal axis of each bolt intersects the center longitudinal axis of said tubular storage member, and wherein each bolt is disposed at a corner of said tubular storage member. (a) each of said plurality of tubular storage members is provided near the bottom thereof with four bolts; and wherein (b) each of said plurality of support plates has four sides of equal length and four upstanding portions of equal length of which each upstanding portion is provided with one open recess. (a) each recess is asymmetrical; wherein (b) each recess is formed by a relatively long portion and by a relatively short portion of one of said plurality of upstanding portions of one of said plurality of support plates; and wherein (c) said relatively short portion projects at least to the center of said substantially circular bolt-bearing surface to lock one of said plurality of bolts in position. (a) said tubular member is provided near the bottom thereof with at least three horizontal bolts the longitudinal axes of al the bolts intersecting at a point common to said longitudinal axes; wherein (b) said tubular member is supported by a first circular plate having a plurality of angularly displaced upstanding portions each for supporting one of said plurality of bolts; and wherein (c) each of said upstanding portions is provided on the upper edge thereof with an open recess, said recess including a relatively wide slanting entrance portion for inserting one of said at least three bolts therein, and said recess further including a substantially circular relatively narrow bottom portion coaxial with one of said plurality of horizontal bolts; and wherein (d) the longitudinal axis of each bolt intersects the center longitudinal axis of said tubular member, and wherein each bolt is disposed at a corner of said tubular member. (a) said tubular member is closed at the bottom thereof by a second plate parallel to said first circular plate, said second plate being perforated for the admission of a cooling medium to said chambers; and wherein (b) spacing means are interposed between said first plate and said second plate. 2. A storage structure as specified in claim 1 wherein 3. A storage structure as specified in claim 1 wherein the bottom of each of said plurality of tubular storage members is closed by a perforated plate for admission of cooling fluid to said chambers, and wherein said bolts are affixed with one end thereof to said perforated plate. 4. A storage structure as specified in claim 1 wherein 5. A storage structure as specified in claim 1 wherein each of said plurality of tubular storage members is rectangular in cross-section and wherein each of said plurality of horizontal bolts is arranged in the direction of the diagonals of one of said plurality of tubular storage members. 6. A storage structure as specified in claim 1 wherein the surface, which is bounding said open recess, has a nose-like projection that extends at least to the vertical median plane of one of said plurality of bolts to lock said one of said plurality of bolts in position against the action of upward directed forces. 7. A storage structure for waste of nuclear plants comprising a tubular member adapted to form one or more waste receiving chambers wherein 8. A storage structure as specified in claim 7 wherein |
description | This application is a continuation of co-pending U.S. patent application Ser. No. 13/626,203, filed Sep. 25, 2012, entitled “Method and Mold for Casting Thin Metal Objects,” the entire contents of which are incorporated by reference herein. The U.S. Government has rights to this invention pursuant to contract number DE-AC05-00OR22800 between the U.S. Department of Energy and Babcock & Wilcox Technical Services Y-12, LLC. This disclosure relates to the field of metal casting. More particularly, this disclosure relates to casting thin metal sheets. The production of metal coupons and plates is necessary for many industrial applications. One important application is the manufacture of plates of fissile material that are to be subjected to rolling operations in order to create foils for use as nuclear reactor fuel elements. Traditional methods for manufacturing such metal plates include the steps of: casting an ingot, applying a thermal (softening) treatment, rolling the ingot with a break-down mill (to a thickness suitable for feeding a foil mill), annealing the rolled plate, and cutting the plate into coupons that can be used to feed a foil mill. In such processes, it is desirable to begin with very thin castings in order to minimize the amount of rolling operations that must be performed on the castings. For many applications it would be desirable to cast metals as sheets that have a thickness that is less than 0.020 inches. Historically, the casting of such thin metal sheets has been difficult. When traditional casting processes are applied to the casting of thin metal sheets, the resulting sheets often have unacceptable quality defects. For example, the resultant castings often have excessive porosity, and/or have incomplete extension (i.e., the molten metal solidifies before filling the mold cavity), and/or have cold shunts (i.e., areas where two or more portions of the molten metal have flowed together but did not fuse together before solidifying). When thicker castings are used, rolling operations often result in “alligatoring” (sometimes referred to as “fish-mouthing”), which refers to a splitting of an edge of a rolled metal slab in which the plane of the split is parallel to the rolled surface. What is needed therefore are improved methods for casting thin metal plates and sheets that may be more productively used in subsequent rolling operations. In one embodiment, the present disclosure provides a first mold backing structure that has a first mold backing structure plate. The first mold backing structure plate forms a first mold cavity first planar surface. The first mold backing structure also has a first mold cavity first framing portion that forms a first mold backing structure first surface and a first mold cavity first edge portion. Typically, at least a portion of the first mold cavity first edge portion is a first beveled edge. Such embodiments typically further include a first mold facing structure that has a first mold cavity second planar surface. Typically, the first mold backing structure first surface is removably attachable to the first mold cavity second planar surface. Such embodiments further typically include a means for removably attaching the first mold backing structure first surface to the first mold cavity second planar surface. Generally, when the first mold backing structure first surface is attached to the first mold cavity second planar surface, then the combination of the first mold cavity first planar surface, the first mold cavity second planar surface, and the first mold cavity first edge portion form a first mold cavity and a first horizontal open edge for providing a flow of a molten metal into the first mold cavity. In another embodiment, a metal casting system has a first mold backing structure, and the first mold backing structure includes a first mold backing structure plate that forms a first mold cavity first planar surface. In certain embodiments, the first mold backing structure also has a first mold cavity first framing portion that forms a first mold backing structure first surface and a first mold cavity first edge portion. Generally, such embodiments further include a first means for removably attaching the first mold backing structure first surface to a first mold cavity second planar surface of a first mold facing structure. These embodiments generally further include a distributor that is removably attachable to the first mold backing structure and the first mold facing structure. Further, a second means is provided for removably attaching the distributor to the first mold backing structure and the first mold facing structure. In these embodiments, when the first mold backing structure first surface is attached to the first mold cavity second planar surface, and when the distributor is attached to the first mold backing structure and to the first mold facing structure, the distributor forms a first mold cavity second edge portion. Furthermore, the combination of the first mold cavity first planar surface, the first mold cavity first edge portion, the first mold cavity second planar surface, and the first mold cavity second edge portion form a first mold cavity. Then a first vertical feed orifice is provided through the distributor for providing a flow of a molten metal into the first mold cavity. Further disclosed are methods for forming a cast metal sheet having a thickness ratio between about 0.000035 in./sq. in. and about 0.005 in./sq. in. Typically, methods include melting a solid metal having a melting temperature to form a molten metal, and preheating a mold having a cavity with a thickness ratio between about 0.000035 in./sq. in. and about 0.005 in./sq. in. to a temperature greater than the melting temperature of the metal. Then, typically, the molten metal is flowed into the cavity. The method proceeds with cooling the molten metal to a temperature below the melting temperature of the metal to form the cast metal sheet and then removing the cast metal sheet from the mold. In the following detailed description of the preferred and other embodiments, reference is made to the accompanying drawings, which form a part hereof, and within which are shown by way of illustration the practice of specific embodiments of metal casting systems. It is to be understood that other embodiments may be utilized, and that structural changes may be made and processes may vary in other embodiments. Disclosed herein are various embodiments of casting systems that may be used to cast thin metal plates and sheets. Plates and sheets are materials that have widths and lengths that are at least an order of magnitude greater than their thickness. For purposes of discussion herein, the distinction between what is considered to be a “plate” and what is considered to be a “sheet” is that plates are considered to be materials that have a thickness of at least 0.020 inches whereas sheets are materials that have a thickness that is less than 0.020 inches. Typically the manufacturing tolerance on these parts is about +/−0.002 inches, so when a thickness of a plate or a sheet is stated as being “about” a particular dimensional value, that tolerance is implied. In preferred embodiments the casting systems are used in a furnace such that an appropriate thermal profile may be applied throughout the casting system. For example, the casting system may be fabricated from structures that include susceptors of microwaves, such that the casting system may be disposed in a microwave furnace and microwave energy may be used to heat the casting system. A casting system fabricated from graphite for use in a microwave furnace is an example of such a system. Typically, various embodiments disclosed herein will reduce or potentially eliminate the need for break-down rolling, thermal treatments, and cutting of metal plates or sheets. The reduction or elimination of these steps greatly reduces the complexity, size, and processing time required to prepare plates or sheets for foil rolling. The casting of thin plates or sheets also increases manufacturing efficiency by reducing scrap generation. Using embodiments disclosed herein, metal test coupons have been cast over a range between thicknesses from about 0.005 inches thick to about 0.2 inches thick, although smaller thicknesses are believed possible. FIG. 1A illustrates a metal casting system 10. The metal casting system 10 has a first mold cavity 14 and a second mold cavity 18. The first mold cavity 14 has a first open edge 16 and the second mold cavity 18 has a second open edge 20. As shown, the open edges 16 and 20 are typically horizontal. In the embodiment of FIG. 1A, the first horizontal open edge 16 and the second horizontal open edge 20 each have the shape of a trapezium (truncated triangle). Molten metal may be poured into the first mold cavity 14 through the first horizontal open edge 16 and molten metal may be poured into the second mold cavity 18 through the second horizontal open edge 20. An optional well 22 is provided around the first horizontal open edge 16 and the second horizontal open edge 20 to facilitate the pouring of molten metal into the first horizontal open edge 16 and the second horizontal open edge 20. The first mold cavity 14 has a thickness 24 and the second mold cavity 18 has a thickness 26. Typically, the thicknesses 24 and 26 are each in a range between about 0.005 inches and about 0.2 inches. The first mold cavity 14 and the second mold cavity 18 each have a width 28. Typically, the width 28 is in a range between about four inches to about twelve inches. The metal casting system 10 includes a first mold backing structure 30, a first mold facing structure 70, and a second mold backing structure 110. A metal band 114 is provided to attach together the first mold backing structure 30, the first mold facing structure 70, and the second mold backing structure 110. The view of the metal casting system 10 in FIG. 1A is taken from the perspective of a corner 126 as shown in FIG. 1B. FIG. 1B illustrates an exploded view of certain elements of the metal casting system 10. The first mold backing structure 30 includes a first mold backing structure plate 34. The first mold backing structure plate 34 forms a first mold cavity first planar surface 38. The first mold backing structure 30 also includes a first mold cavity first framing portion 40. The first mold backing structure 30, the first mold cavity first framing portion 40, the first mold facing structure 70, and the second mold backing structure 110 may be fabricated from stock plate having the required thickness or may be fabricated from sections sliced from a block of material. In the embodiment depicted in FIGS. 1A and 1B (and 1C), the first mold backing structure plate 34 and the first mold cavity first framing portion 40 are formed as separate removably attachable elements. As used herein, the term “removably attachable” refers to a configuration of elements that are configured both to be held together in contact with each other and to be separated from each other. Various means may be used to configure elements as removably attachable to each other, such as bands (e.g., the band 114 of FIG. 1A), clamps, screws, bolts and nuts, wires, straps, frames, casings, dissolvable adhesives, weights applying sufficient pressure to opposing ends, and similar mechanisms. When a mold backing structure plate (such as the first mold backing structure plate 34) and a mold cavity framing portion (such as the first mold cavity first framing portion 40) are removably attachable to each other, the attachment mechanism used to attach together the first mold backing structure 30, the first mold facing structure 70, and the second mold backing structure 110 is typically also used to attach the mold backing structure plate to the mold cavity framing portion. Forming a mold backing structure plate (e.g., the first mold backing structure plate 34) and a cavity first framing portion (e.g., the first mold cavity first framing portion 40) as separate removably attachable elements may provide some advantages in removing thin plates or sheets that have been cast in a metal casting system. In FIG. 1B, the first mold backing structure plate 34 and the first mold cavity first framing portion 40 are depicted in their attached configuration. The first mold cavity first framing portion 40 has a height 42 that establishes the height of parts cast in the first mold cavity 14. Typically, the height 42 is in a range between about four inches to about twelve inches. Continuing with FIG. 1B, the first mold cavity first framing portion 40 forms a first mold backing structure first surface 46. The first mold cavity first framing portion 40 also forms a first mold cavity first edge portion 50. In the embodiment of FIG. 1B the first mold cavity first edge portion 50 is a beveled edge. As used herein the term “beveled edge” refers to an edge that is angled at an angle between 120° and 150° with respect to an adjoining surface (for example, in this case, with respect to the first mold cavity first planar surface 38 or with respect to the first mold backing structure first surface 46). The use of a beveled edge in embodiments of casting systems for metal plates provides a significant reduction in alligatoring when such cast plates are processed through a rolling mill. The formation of a beveled edge is not as important when processing cast sheets. FIG. 1B further illustrates a corner 62 of the first mold cavity first edge portion 50 that is formed as a square corner. In many embodiments a corner of the first mold cavity first edge portion 50 is formed as a rounded corner because it generally reduces machining costs. FIG. 1C illustrates the metal casting system 10 as viewed from the perspective of corner 150, which is diagonally opposed from the corner 126. FIG. 1C illustrates that the first mold facing structure 70 has a first mold cavity second planar surface 166. The first mold cavity second planar surface 166 is removably attachable to the first mold backing structure first surface 46 of the first mold cavity first framing portion 40 (depicted in FIG. 1B). When the first mold backing structure first surface 46 is attached to the first mold cavity second planar surface 166, the first mold cavity first planar surface 38, the first mold cavity second planar surface 166, and the first mold cavity first edge portion 50 (depicted in FIG. 1B) form the first mold cavity 14 (FIG. 1A) and the first horizontal open edge 16 (FIGS. 1A and 1C). FIG. 1C further illustrates that the first mold backing structure plate 34 forms a second mold cavity second planar surface 170 opposing the first mold cavity first planar surface 38. FIG. 1C also identifies a second mold backing surface 118 of the second mold backing structure 110. In some embodiments, the metal casting system 10 employs a second mold facing structure (similar to the first mold facing structure 70) that is disposed adjacent the second mold backing surface 118. Returning to FIG. 1B, further illustrated are certain details of the second mold backing structure 110. The second mold backing structure 110 includes a second mold backing structure plate 174. The second mold backing structure plate 174 forms a second mold cavity first planar surface 178. The second mold backing structure 110 also includes a second mold cavity first framing portion 182. In certain embodiments, including the embodiment depicted in FIGS. 1A and 1B (and 1C), the second mold backing structure plate 174 and the second mold cavity first framing portion 182 are formed as an integral element. The second mold cavity first framing portion 182 has a height 184 that establishes the height of parts cast in the second mold casting cavity 18. Typically the height 184 is in a range between about four inches to about twelve inches. As previously noted, the typical width of a casting cavity is about four inches to about twelve inches. With the typical height of a casting cavity being between about four inches and about twelve inches, the typical surface area of a cast part is between about 16 square inches and about 144 square inches. Also, as previously noted, the typical thickness of a casting cavity is between about 0.005 inches and about 0.2 inches. These dimensions establish a “thickness ratio,” which is defined as the thickness of a part divided by its surface area. Thus, the thickness ratio of cast parts provided by embodiments disclosed herein may vary between 0.005 in./144 sq. in., which is about 0.000035 in./sq. in., and 0.2 in./16 sq. in., which is about 0.013 in./sq. in. For sheets, which as previously indicated have a thickness that is less than 0.020 inches, the upper limit for the thickness ratio is about 0.020 in/16 sq. in., or about 0.00125 in./sq. in. The second mold cavity first framing portion 182 forms a second mold backing structure first surface 186 and a second mold cavity first edge portion 190. In the embodiment of FIG. 1B, the second mold cavity first edge portion 190 is a beveled edge. In the embodiment of FIG. 1A, the first horizontal open edge 16 and the second horizontal open edge 20 each have the shape of a truncated triangle. The second mold cavity second flat planar surface 170 is removably attachable to the first mold backing structure first surface 46 (depicted in FIG. 1B). When the second mold backing structure first surface 186 is attached to the second mold cavity second planar surface 170, the second mold cavity first planar surface 178, the second mold cavity second planar surface 170 (FIG. 1C), and the second mold cavity first edge portion 190 form the second mold cavity 18 (FIG. 1A) and the second horizontal open edge 20 (FIGS. 1A and 1C). FIG. 1D illustrates a metal casting system 710 shown from the same perspective point of view as the metal casting system 10 in FIG. 1C. The metal casting system 710 has the second mold backing structure 110 that is used in the metal casting system 10 of FIGS. 1A, 1B, and 1C, including the second horizontal open edge 20 as depicted in FIGS. 1A and 1C and the second mold cavity first framing portion 182 depicted in FIG. 1B. In the metal casting system 710, there is a first mold cavity second framing portion 712 that has a third horizontal open edge 714, a first mold cavity second edge portion 718, and a first mold cavity second planar surface 722 that have been formed in an alternate first mold facing structure 726. Also, in the metal casting system 710 there is a second mold cavity second framing portion 730 that has a fourth horizontal open edge 734, a second mold cavity second edge portion 738, and a second mold cavity second planar surface 742 that have been formed in an alternate first mold backing structure 746. Also formed in the alternate first mold backing structure 746 is the first horizontal open edge 16 depicted in FIGS. 1A and 1C and the first mold cavity first framing portion 40 depicted in FIG. 1A. The first mold cavity first framing portion 40 and the first mold cavity second framing portion 712 each have a height 750 that establishes the height of parts cast in an alternate first mold cavity 758 (depicted in FIG. 1E). The second mold cavity first framing portion 182 and the second mold cavity second framing portion 730 each have a height 754 that establishes the height of parts cast in an alternate second mold cavity 762 (depicted in FIG. 1E). Typically the heights 750 and 754 are in a range between about four inches to about twelve inches. FIG. 1E illustrates the metal casting system 710 assembled together with a metal band 766. The alternate first mold facing structure 726, the alternate first mold backing structure 746 and the second mold backing structure 110 form the alternate first mold cavity 758 and the alternate second mold cavity 762. The alternate first mold cavity 758 has the third horizontal open edge 714 and the alternate second mold cavity 762 has the second horizontal open edge 20 (as also depicted in FIGS. 1A and 1C). In the alternate first mold cavity 758 depicted in FIG. 1E, the third horizontal open edge 714 in combination with the first horizontal open edge 16 forms a first hexagon shape. In the alternate second mold cavity 762, a second hexagon shape is formed by the fourth horizontal open edge 734 in combination with the second horizontal open edge 20. The hexagonal shape is particularly beneficial for reducing alligatoring during rolling operations performed on plates or sheets that have been cast in the metal casting system 710. The alternate first mold cavity 758 has a thickness 774 and the alternate second mold cavity 762 has a thickness 778. Typically, the thicknesses 774 and 778 are each in a range between about 0.005 inches and about 0.2 inches. The alternate first mold cavity 758 and the alternate second mold cavity 762 each have a width 782. Typically, the width 782 is in a range between about four inches to about twelve inches. FIG. 1F illustrates a modified framing structure portion 786 that has a necked opening 788. In embodiments with a necked opening (e.g., the necked opening 788), a horizontal open edge 790 for providing a flow of a molten metal into a modified mold cavity portion 794 is defined below the necked opening 788. In embodiments with a necked opening, a width 796 that establishes the width of parts cast in the modified framing structure portion 786 is defined below the necked opening (e.g., the necked opening 788). Typically, the width 796 is in a range between about four inches to about twelve inches. The height of parts that are cast in the modified framing structure portion 786 is defined by a height 798 of the modified framing structure portion 786 below the necked opening 788. Typically, the height 798 is in a range between about four inches to about twelve inches. FIG. 2 illustrates metal casting system 210 that includes the first mold backing structure 30, the first mold facing structure 70, and the second mold backing structure 110 as previously described herein. The metal casting system 210 further includes a distributor 214 that is removably attachable to the first mold backing structure 30, the first mold facing structure 70, and the second mold backing structure 110. The view of the metal casting system 210 in FIG. 2 is taken from the perspective of a corner 218 of the distributor 214, and an adjoining corner 222 of the distributor 214 is also depicted. FIG. 3 depicts the distributor 214 in a perspective where it is inverted from the view shown in FIG. 2. When the distributor 214 is attached to the first mold backing structure 30, the first mold facing structure 70, and the second mold backing structure 110, the distributor 214 forms a first mold cavity second edge portion 226 and a second mold cavity second edge portion 230. Thus, in the embodiment of FIG. 2, the first mold cavity first planar surface 38 (FIG. 1B), the first mold cavity first edge portion 50 (FIG. 1B), the first mold cavity second planar surface 166 (FIG. 1C), and the first mold cavity second edge portion 230 (FIG. 3) form a first mold cavity (analogous to the first mold cavity 14 of FIG. 1A with the addition of a top edge). Further, in the embodiment of FIG. 2, the second mold cavity first planar surface 178 (FIG. 1B), the second mold cavity first edge portion 190 (FIG. 1B), the second mold cavity second planar surface 170 (FIG. 1C), and the second mold cavity second edge portion 230 (FIG. 3) form a second mold cavity (analogous to the second mold cavity 18 of FIG. 1A with the addition of a top edge). Returning to FIG. 2, the distributor 214 has at least one orifice that provides a conduit for flowing molten metal into the first mold cavity 14 and the second mold cavity 18. In preferred embodiments, as shown in FIG. 2, the distributor 214 includes a first vertical feed orifice 242, a second vertical feed orifice 246, a third vertical feed orifice 250, and a fourth vertical feed orifice 254. Feed orifices may be cylindrical, slot-shaped, or formed as openings having other geometries. The first vertical feed orifice 242 and the second vertical feed orifice 246 provide a conduit for flowing molten metal into the first mold cavity 14 (FIG. 1A). The third vertical feed orifice 250 and the fourth vertical feed orifice 254 provide a conduit for flowing molten metal into the second mold cavity 18 (FIG. 1A). The distributor 214 further includes a well 224. The well 224 facilitates the flowing of a molten metal into the first vertical feed orifice 242, the second vertical feed orifice 246, the third vertical feed orifice 250, and the fourth vertical feed orifice 254. Embodiments that employ the distributor 214 with the well 224 typically do not employ the optional well 22 that is depicted in FIG. 1A. The reason that the optional well 22 is not used in such configurations is that using the distributor 214 provides the advantage of reducing the size of the spruest that are connected to molded plates or sheets to just small cylinders, slots, holes, etc. that are formed in the vertical feed orifices (242, 246, 250, and 254). In contrast, if the optional well 22 is used, much larger spruest may be formed in the optional well 22 (and connected to molded plates or sheets). Reducing the size of the sprues to small cylinders also facilitates trimming of the molded plates or sheets to a final shape for subsequent rolling operations. FIG. 4 illustrates a metal casting system 300. The metal casting system has ten casting cavities, the casting cavities preferably being formed substantially as described above with respect to the mold facing structures and mold backing structures of the embodiments of FIGS. 1A-1F, disposed underneath and parallel to reference line 304. In other embodiments fewer or more than ten casting cavities may be employed. A distributor 308 provides four orifices (such as orifices 312, 316, 320 and 324) for each mold cavity. The orifices (e.g., orifices 312, 316, 320, and 324) have a funnel-shape countersink formed at the top and are disposed in a well 328 in order to facilitate the flow of molten metal through the orifices into the ten casting cavities. Such a countersink is optional. Bolts 340 are provided to attach together the components that define the ten casting cavities, and four holes 344 are provided to secure the distributor 308 to the components that define the ten casting cavities. In some embodiments, two or more cavities may be formed in the same plate. Sometimes the cavities may be front and back instead of all facing the same direction. In a typical operation, the planar casting surfaces and the edges of the mold cavities are often treated with a mold release material prior to being assembled into a casting system. Such materials as yttrium oxide, aluminum oxide, and erbium oxide, preferably in a finely powdered form, are effective mold release agents for casting uranium and uranium alloys. Typically, prior to flowing the molten metal into the cavity(ies) of a mold, the mold is preheated to a temperature greater than the melting temperature of the metal that is being cast. In some embodiments, the mold is preheated to a temperature of at least 100° C. greater than the melting temperature of the metal, and in some embodiments the mold is preheated to a temperature of at least 200° C. greater than the melting temperature of the metal. Often, a crucible is secured to the top of the casting system to provide the source of molten metal into the mold cavities. With such embodiments, a solid metal may be disposed in the crucible, and the crucible with the casting system underneath may then be disposed in a furnace (such as, but not limited to, a microwave casting furnace) where the solid metal melts and flows into the mold cavities. Such configurations provide penetration heating that helps ensure that all mold components sustain heat above the metal melt temperature for a period of time after pouring. Such heating reduces or eliminates porosity, incomplete extensions, and cold shunts. Typically, it is useful to maintain the temperature of the mold with the molten metal in the cavity at a temperature above the melting temperature of the metal (and sometimes to a temperature that is at least 100° C. or at least 200° C. above the melting temperature of the metal) for at least one minute prior to cooling the molten metal to a temperature below the melting temperature of the metal. The higher temperatures (i.e., at least 100° C. or at least 200° C. above the melting temperature) tend to facilitate the flow of the molten metal into the mold cavities. Without being bound by any scientific theory, it is believed that these higher temperatures may either reduce the viscosity of the molten metal or may reduce adherence of the metal to the walls of the cavity, or both. The minimum thickness of a sheet that may be cast may also be dependent upon surface tension properties as a function of temperature. Surface tension of uranium is estimated to be around 1500 dynes/cm at 1200° C. whereas the surface tension of mercury is 478 dynes/cm at 15° C. FIG. 5 illustrates an embodiment of a metal casting system 400 that employs a crucible 404. A threaded plug 408 in the bottom of the crucible 404 includes a knockout butterfly plug 412. When a solid metal is entirely melted in the crucible 404, a pour rod 420 is used to break the knockout butterfly plug 412 to flow the molten metal into a plurality of mold cavities 430 that are formed by mold backing structures 434 in a mold 438. This arrangement provides a pressure head of molten metal into the mold cavities 430, which facilitates the complete filling of the mold cavities. FIG. 5 further illustrates that the mold 438 has a distributor 450 for flowing the molten metal into the mold cavities 430. A distributor cap 454 is provided between the crucible 404 and the mold 438. The distributor cap 454 is removably attachable to the distributor 450 by short bolts 462, and the crucible 404 is removably attachable to the distributor cap by the threaded plug 408, which is screwed into the distributor cap 454. The distributor cap 454 provides thermal conductivity between the crucible 404 and the mold 438. The mold backing structures 434 are disposed between a first mold facing structure 470 and a second mold facing structure 474. Long bolts 478 extending from the second mold facing structure 474 into the first mold facing structure 470 are used in part to hold the first mold facing structure 470, the mold backing structures 434, and the second mold facing structure 474 together. A clamp 482 further holds the first mold facing structure 470, the mold backing structures 434, and the second mold facing structure 474 together. FIG. 6 illustrates an embodiment of a metal casting system 500 that employs many of the same components as the metal casting system 400 of FIG. 5. A difference is that a distributor cap 554 is configured as a clamp to help hold the first mold facing structure 470, the mold backing structures 434, and the second mold facing structure 474 together. FIG. 7 illustrates a crucible 600 that may be used in various embodiments of a metal casting system. The crucible 600 has an opening 608, and a plug portion 612 of a pour rod 616 is disposed in the opening 608. When molten metal is disposed in the crucible 600, the pour rod 616 may be raised vertically to permit the molten metal to pass through the opening 608 into a mold such as the mold 438 depicted in FIGS. 5 and 6. A further crucible and pouring mechanism that may be used in various embodiments of a metal casting system is described in U.S. Pat. No. 5,286,009—“Device for Controlling the Pouring of Molten Metals,” issued Feb. 15, 1994 to Moore et al. Typically, various embodiments described herein are used to cast plates or sheets having a thickness that is in a range from about 0.005 inches to about 0.1 inches. In some embodiments the mold cavities may be of varying thicknesses, heights, and widths to accommodate the casting of comparable plates or sheets in a single pour. Besides the casting of plates or sheets, the casting of rods and objects having a variety of regular or irregular geometries may be accommodated, even in a single pour. In such embodiments the previously described planar surfaces may be replaced with curvilinear planar surfaces or irregular surfaces. Such cast objects may be of near-net (final product) shape. Various embodiments described herein may be used to fabricate cast metal sheets having a thickness of about 0.005 inches. It was previously unexpected that such paper-thin metal sheets could be formed by a casting process. The foregoing descriptions of embodiments have been presented for purposes of illustration and exposition. They are not intended to be exhaustive or to limit the embodiments to the precise forms disclosed. Obvious modifications or variations are possible in light of the above teachings. The embodiments are chosen and described in an effort to provide the best illustrations of principles and practical applications, and to thereby enable one of ordinary skill in the art to utilize the various embodiments as described and with various modifications as are suited to the particular use contemplated. All such modifications and variations are within the scope of the appended claims when interpreted in accordance with the breadth to which they are fairly, legally, and equitably entitled. |
|
049869523 | claims | 1. A protection system for protecting a nuclear reactor in the event of a reaction-inhibiting element falling, said system being applicable to a reactor having a vertical axis core which is the site of a nuclear reaction, said reaction being accompanied by a neutron flux which is distributed, at least angularly, around said axis, said core having vertical hollows distributed at least angularly around said axis and being provided with controllable reaction-inhibiting elements capable of descending in said hollows in order to absorb said neutron flux, thereby controlling said nuclear reaction, such that an accidental fall of one of said reaction-inhibiting elements into one of said hollows locally absorbs said neutron flux and disturbs the neutron flux distribution around said axis, and such that continuing said nuclear reaction may then damage said core, said system including not less than three separate protection chains, each of which comprises: a neutron flux detector (D1) disposed at a distance from said axis (A) for measuring a neutron flux which, in the event of one of said reaction-inhibiting elements falling, is subjected to a reduction, with the reduction being larger the nearer said detector is to said hollow in which said fall has taken place; and a primary treatment circuit (R1, S1) associated with said detector (D1) for providing a primary fall signal solely when the neutron flux measured by said detector is subjected to a reduction whose rate of decrease exceeds a predetermined speed threshold; said detectors being angularly distributed around said axis; said system further including a secondary treatment circuit (6) receiving the fall signals output by said protection chains and providing a secondary fall signal solely on receiving not less than two of said fall signals from two respective chains, thereby reducing the risk of such a secondary fall signal being provided when one of said reaction-inhibiting elements has not fallen; and means for limiting said nuclear reaction in the event that said secondary fall signal is provided; said system being characterized by the fact that each of said protection chains includes: at least two of said neutron flux detectors (D1A, D1B) angularly separated by more than 90.degree. about said axis (A); two of said primary treatment circuits (R1A, S1A and RIB, S1B) associated respectively with said two detectors for providing said primary fall signals; and an intermediate treatment circuit (P1) for receiving said primary fall signals and for providing an intermediate fall signal on receiving at least one primary fall signal from at least one of said two primary treatment circuits; said fall signals provided at the outputs from said protection chains and received by said secondary treatment circuit (6) being constituted by said intermediate fall signals, such that said secondary fall signal is provided in the event of one said reaction-inhibiting elements falling even if the resulting reductions in neutron flux received by a plurality of said detectors which are relatively far away from said hollow receiving said falling reaction-inhibiting element do not exceed said predetermined variation speed threshold, and even in the event of one of said chains being faulty so that it prevents one of said primary fall signals from a detector relatively close to said hollow being generated. the number of said protection chains (DIA, R1A, S1A, D1B, R1B, S1B, P1) being four; and said detectors (D1A, D4B, D3A, D1B, D2A, D3B, D4A, D2B) being angularly distributed around said axis at an angular pitch close to 45.degree.. 2. A system according to claim 1, said system being applicable to a reactor whose core (2) has the same symmetry about said axis (A) as a square about its center; 3. A system according to claim 2, characterized by the fact that each of said protection chains (DIA, R1A, S1A, D1B, R1B, S1B, P1) includes two of said detectors (D1A, D1B) which are separated by an angle of 135.degree. about said axis (A). 4. A system according to claim 2, characterized by the fact that each of said protection chains (D1C, R1C, S1C, D1D, R1D, S1D, P1) includes two of said detectors (D1C, D1D) which are separated by an angle of 180.degree. about said axis (A). |
abstract | Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI2, CuI, or Bi5O7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425° C. to 550° C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500° C. (below the silver iodide sublimation temperature of 500° C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon. |
|
claims | 1. A drawing apparatus for performing drawing on a substrate with a charged particle beam, the apparatus comprising:a first member in which an aperture, through which the charged particle beam passes, is formed;a chamber including a first space and a second space which are partitioned by the first member; anda removing device including a first supply device configured to supply a first gas containing unsaturated hydrocarbon to the first space and a second supply device configured to supply a second gas containing ozone to the second space, and configured to remove contamination on the first member by active species generated by reaction of the first gas with the second gas. 2. The apparatus according to claim 1, further comprising:a regulator configured to regulate a gas pressure in at least one of the first space and the second space; anda controller configured to control the regulator so as to form a gas pressure difference between the first space and the second space. 3. The apparatus according to claim 2, further comprising a stage disposed in one of the first space and the second space, and configured to hold the substrate and to be moved,wherein the controller is configured to control the regulator so that a gas pressure in the other of the first space and the second space is higher than that of the one of the first space and the second space. 4. The apparatus according to claim 2, wherein the controller is configured to control the regulator so that the gas pressure difference changes with time. 5. The apparatus according to claim 4, wherein the controller is configured to control the regulator so that a magnitude relation between a gas pressure in the first space and a gas pressure in the second space is inverted with time. 6. The apparatus according to claim 1, wherein the first member includes at least one of an electrode member included in a charged particle optical system, an aperture array member for dividing a charged particle beam into a plurality of charged particle beams, and an aperture member for blanking a charged particle beam. 7. The apparatus according to claim 2, further comprising a charged particle source disposed in one of the first space and the second space, and configured to generate the charged particle beam,wherein the controller is configured to control the regulator so that a gas pressure in the other of the first space and the second space is higher than that of the one of the first space and the second space. 8. The apparatus according to claim 1, further comprising a second member in which an aperture, through which the charged particle beam passes, is formed,wherein the first supply device is configured to supply the first gas to the first space via a third space partitioned by the second member. 9. The apparatus according to claim 1, further comprising a second member in which an aperture, through which the charged particle beam passes, is formed,wherein the second supply device is configured to supply the second gas to the second space via a third space partitioned by the second member. 10. The apparatus according to claim 1, further comprising a second member in which an aperture, through which the charged particle beam passes, is formed,wherein the removing device includes a supply device configured to supply a third gas containing an inert gas to the first space via a third space partitioned by the second member. 11. The apparatus according to claim 1, further comprising a second member in which an aperture, through which the charged particle beam passes, is formed,wherein the removing device includes a supply device configured to supply a third gas containing an inert gas to the second space via a third space partitioned by the second member. 12. The apparatus according to claim 1, further comprising a third member disposed in one of the first space and the second space, and configured to divide the one of the first space and the second space into a fourth space, which neighbors the other of the first space and the second space, and a fifth space outside the fourth space,wherein the removing device includes a supply device configured to supply a gas, corresponding to the one of the first space and the second space, of the first gas and the second gas to the fourth space. 13. The apparatus according to claim 12, wherein the third member is a movable object. 14. The apparatus according to claim 13, wherein the movable object is a stage. 15. The apparatus according to claim 13, wherein the third member includes a sensor configured to measure the charged particle beam. 16. A method of manufacturing an article, the method comprising:performing drawing on a substrate using a drawing apparatus;developing the substrate on which the drawing has been performed; andprocessing the developed substrate to manufacture the article,wherein the drawing apparatus performs drawing on the substrate with a charged particle beam, and includes:a first member in which an aperture, through which the charged particle beam passes, is formed;a chamber including a first space and a second space which are partitioned by the first member; anda removing device including a first supply device configured to supply a first gas containing unsaturated hydrocarbon to the first space and a second supply device configured to supply a second gas containing ozone to the second space, and configured to remove contamination on the first member by active species generated by reaction of the first gas with the second gas. |
|
description | This application is a Continuation-In-Part application of U.S. application Ser. No. 12/577,789, entitled WIRELESS TRANSMISSION OF NUCLEAR INSTRUMENTATION SIGNALS, filed on Oct. 13, 2009. The present invention is directed to monitoring of a nuclear reactor pressure vessel. More particularly, the present invention is directed to a system and method for wirelessly monitoring a condition of a nuclear reactor pressure vessel. The mechanical movement (i.e. insertion, withdrawal) and the associated monitoring of the position of the control rods are necessary functions for the operation of a nuclear reactor. Each of the instruments that perform this function typically is terminated with a power cable and one or two position indication cables that transmit signals from the instrument back to processing units, typically located in a control room. As used herein, the term instrument may also include a sensor or sensing device. Known rod position indicator cable systems, such as the one depicted in FIGS. 1 and 2, typically include multi pin connector disconnect points 10 located at the top of the nuclear reactor vessel head 12 and at the reactor cavity wall 14 poolside. Additional disconnect points 10 may also be located at other points between the vessel head 12 and the cavity wall 14. The multi pin connector disconnect points 10 allow each of the interconnecting cable sections 16 to be removed from corresponding sensing instruments 18 to allow for the disassembly of the reactor vessel 8 for refueling. The typical reactor vessel 8 includes on the order of magnitude of 100 or more of these cable assemblies. The removal and installation of the cable sections 16 is generally part of the “critical path” schedule for a refueling outage and generally requires the services of a specially trained crew of technicians during both the initial and concluding stages of the refueling outage in order to complete the work. Typically, such work can take up to an entire shift to complete. In total, the manipulation of the signal cable sections 16 may occupy an entire day of a 30 day outage. Given an estimated cost of $20,000 to $25,000 per hour of lost critical path time, this one day period would represent a cost of approximately $500,000 per refueling outage without even taking into consideration the cost of the trained work crew. Additionally, the repeated manipulation of the signal cables increases the potential for damage, leading to the need to repair and/or replace the cables and/or the related hardware. Furthermore, the manipulation of the signal cables must be carried out in a radiation area located above the reactor vessel. Elimination of this work scope would thus eliminate the radiation exposure associated with this work activity. Accordingly, there exists room for improvement in the system and method for monitoring the position of the control rods and other reactor conditions. In accordance with an embodiment of the invention, a method of monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment is provided. The method comprises: sensing a condition of the reactor pressure vessel with an instrument, transmitting a signal indicative of the condition of the reactor pressure vessel from the instrument to a powered wireless transmitting modem disposed in the radioactive environment, wirelessly transmitting a signal indicative of the condition of the reactor pressure vessel from the transmitting modem to a receiving modem in the line of sight of the transmitting modem, transmitting a signal indicative of the condition of the reactor pressure vessel from the receiving modem to a signal processing unit, and determining the condition of the reactor pressure vessel from the wirelessly transmitted signal. The wireless transmission may comprise an infrared transmission. The condition of the reactor pressure vessel may be sensed by a plurality of instruments operatively connected with a plurality of transmitting modems. The plurality of wireless transmitting modems may transmit signals to a plurality of receiving modems operatively connected with the signal processing unit for determining the condition of the reactor pressure vessel. The condition of the reactor pressure vessel may be sensed by a plurality of instruments operatively connected with a transmitting modem and the transmitting modem may transmit a signal to a receiving modem operatively connected with the signal processing unit for determining the condition of the reactor pressure vessel. The condition of the reactor pressure vessel may be determined during power generation operations. The condition of the reactor pressure vessel may be determined while the reactor pressure vessel is disassembled. The powered transmitting modem may be bridged with a second powered transmitting modem so that the second transmitting modem will continue to function should its power source fail. The transmitting modem may be powered by a regenerative battery. The transmitting modem may be externally powered. The transmitting modem may be powered parasitically from a power cable associated with a control rod drive mechanism. The condition monitored may be one of: control rod position, coolant water bulk temperature, coolant water level, radiation level, and ion chamber level. In accordance with another embodiment of the invention, a system for monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment is provided which comprises an instrument structured to monitor a condition of the nuclear reactor pressure vessel, a powered wireless transmitting modem disposed in the radioactive environment, a receiving modem in the line of sight of the transmitting modem, and a signal processing unit electrically coupled to the receiving modem. The wireless transmitting modem is electrically coupled to the instrument. The receiving modem is in wireless communication with the transmitting modem. The signal processing unit is structured to determine the condition of the nuclear reactor pressure vessel from the instrument. The condition of the reactor pressure vessel may be sensed by a plurality of instruments operatively connected with a plurality of transmitting modems. The plurality of wireless transmitting modems may transmit signals to a plurality of receiving modems operatively connected with the signal processing unit for determining the condition of the reactor pressure vessel. The condition of the reactor pressure vessel may be monitored during power generation operations. The condition of the reactor pressure vessel may be monitored while the reactor pressure vessel is disassembled. The powered transmitting modem may be bridged with a second powered transmitting modem so that the second transmitting modem will continue to function should the powered transmitting modem fail. The transmitting modem may be powered by a regenerative battery. The transmitting modem may be externally powered. The transmitting modem may be powered parasitically from a power cable associated with a control rod drive mechanism. The transmitting modem may be powered by a thermocouple disposed in or on the reactor pressure vessel. The transmitting modem may be directly powered by a regenerative battery which is charged by the thermocouple disposed in or on the reactor pressure vessel. The transmitting modem may be powered by one or both of a regenerative battery and the thermocouple electrically coupled to the transmitting modem in a parallel arrangement. The condition monitored may be one of control rod position, coolant water bulk temperature, coolant water level, radiation level, and ion chamber level. The instrument may comprise a plurality of sensing instruments, each instrument being structured to monitor a condition of the nuclear reactor pressure vessel. The present invention will now be described more fully hereinafter with reference to the accompanying drawings, in which examples of the invention are shown. The invention may, however, be embodied in many different forms and should not be construed as limited to the examples set forth herein. Rather, these examples are provided so that this disclosure will be thorough and complete, and will fully convey the scope of the invention to those skilled in the art. Like numbers refer to like elements throughout. FIGS. 3 and 4 illustrate an example improved monitoring system 20 in accordance with the present invention that provides for the monitoring of one or more conditions of a nuclear reactor pressure vessel 8 without the need of cable sections 16 (such as those shown in FIGS. 1 and 2). As shown in FIG. 4, the multi pin connectors 10 (shown in FIGS. 1 and 2) at both ends of the former cable section 16 are replaced with wireless modems 22,26, with each modem 22,26 having a respective integral mating connector assembly 24,28. The modem 22 at the instrument 18 end (i.e. at the reactor vessel 8, FIG. 4) is a wireless transmitter electrically coupled to the instrument 18, while the modem 26 at the reactor cavity wall 14 (FIG. 3) is a wireless receiver. The term “modem”, as used herein, shall be used to refer to a suitable electrical device capable of at least one of sending and receiving wireless transmission signals such as, for example, without limitation, via infrared transmission. The remainder of the rod control position and other instrument signal hardware is not changed from that shown in FIGS. 1 and 2. Referring to FIG. 4, the transmitter modem 22 is parasitically powered from the power cable 17 for the associated rod control drive mechanisms. This is accomplished by insertion of a double-ended multi pin connector assembly 30 in series with the existing power cable 17 at the multi pin connector on the instrument assembly 18. This double-ended multi pin connector assembly 30 also includes an appropriately sized parasitic bleed power cable 32 that is mated to, and powers, the neighboring transmitter modem 22 and instrument 18. Each modem typically only requires a fraction of a watt to power which has a generally transparent effect on the capacity of the comparably massive power cable 17. Likewise, the voltage potential required to operate the level indication probe (not numbered) could be reduced, if needed, whereas an overly amplified signal is not required or desired at the front end of the modem 22. Therefore, sufficient source power is available without significant additional modification to an existing power cable circuit. Additionally, expected power supply interruptions from the power cables as a result of instrument operations (i.e. mechanical control rod movement and static retention mode) may be bridged and conditioned within either the modem 22 and/or the double ended connector assembly 30 in the power cable circuit. Although not a preferred embodiment, it can also be appreciated that the power for each of the modems 22 could be supplied from alternate power sources found within the reactor vessel assembly (e.g., without limitation, power sources for thermocouples, solenoid operated devices, fan motors, switches, lighting) and/or dedicated power sources (e.g., without limitation, regenerative batteries). FIG. 5 illustrates some alternative embodiments of the present invention in which the transmitter modem 22 is powered by voltage produced by a thermocouple 40 (shown schematically) disposed in or on the reactor vessel. As shown in the arrangement in the left portion of FIG. 5, the thermocouple 40 may be directly electrically connected to the transmitter modem 22. As shown in the arrangement in the middle of FIG. 5, the thermocouple 40 may be electrically connected to a regenerative battery 42 which is electrically connected to the transmitter modem 22. As shown in the arrangement in the right portion of FIG. 5, the thermocouple 40 and regenerative battery 42 may both be directly electrically connected to the transmitter modem 22 in a parallel arrangement. Although shown generally as being electrically connected to only one thermocouple 40, it is to be appreciated that multiple thermocouples 40 and/or regenerative batteries 42 could also be employed to power one or more transmitting modems 22. Such an arrangement provides a) redundancy in the unlikely event that a thermocouple fails and b) excess power which may be utilized to power a more powerful transmitter modem and/or provide faster recharging of the regenerative battery or batteries. In another example, one or more of the transmitting modems 22 may be bridged with a second transmitting modem (not shown) powered by a different power source. Such redundant arrangement would provide for the second transmitting modem to continue to function should the power source of the transmitting modem 22 fail. In yet another example, one or more of the transmitting modems 22 may be electrically coupled to a plurality of instruments 18 for detecting and transmitting one or more conditions of the reactor. Such arrangement may be employed to reduce the number of modems 22 needed. Such an arrangement may also be employed to provide redundancy by electrically coupling one instrument to multiple modems 22 (and thus having each modem electrically coupled to multiple instruments 18). Regardless of the transmitting modem arrangement employed, the receiver modem 26 is installed on the “abandoned” end (i.e. electrical connector 10) of the existing rod position indication cables located at the reactor cavity wall 14 (FIG. 3). Preferably each of the receiver modems 26 is installed within the line of sight of each of the corresponding transmitting modems 22. Such line of sight transmission generally minimizes power requirements and the possibility of interference with the transmitted signals. Each of the receiver modems 26 may be supplied power either from the existing source voltage of the existing rod position indication cable system, parasitically from the associated power cable system, or from an alternate or dedicated power source(s). In a preferred embodiment, each existing signal cable end is assigned a discrete modem 26. That is, if there were fifty signal cables, there would be fifty transmitter modems 22 and fifty receiver modems 26. It is to be appreciated that the modems 22,26 could be combined into a lesser number of larger modems. It is also to be appreciated that the present invention may be incorporated into other in-containment cable instrumentation systems that would benefit from elimination of interconnecting cable assemblies (e.g., without limitation, reactor vessel level indication, containment area radiation monitors and ion chambers). In installations with two or more independent level indication instruments for a particular mechanism, one instrument could be outfitted according to the present invention while the second or others could remain unchanged. Such arrangement would provide additional system redundancy using independent hardware. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of invention which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
|
claims | 1. A method of interrogating a container, comprising the steps of:a) irradiating the container for approximately 30 seconds with neutrons having energies between about 2.45 and 14 MeV or with a gamma ray beam having an energy of at least 10 MeV;b) stopping the irradiating;c) after stopping the irradiating, counting β-delayed gamma rays having an energy range between about 3 and 6 MeV for approximately 30 seconds;d) making a first plot comprising number of β-delayed gamma rays counted in step c as a function of β-delayed gamma ray energy to produce an observed energy spectrum;e) making a second plot of the total number of β-delayed gamma rays counted in a portion of the energy range between about 4 and 6 MeV as a function of time in order to determine an effective half-life;f) comparing the observed energy spectrum with known energy spectra produced by fission products of Pu-239; andg) concluding that there is Pu-239 in the container when:the observed energy spectrum and the known energy spectra have the same overall shape; andthe observed effective half-life is approximately 20 to 30 seconds. 2. The method of claim 1 wherein the neutrons comprise D-D neutrons. 3. The method of claim 1 wherein the neutrons comprise D-T neutrons. 4. The method of claim 1 wherein a plastic or liquid scintillation detector is used for the counting step. 5. The method of claim 1 wherein, in step g, the observed energy spectrum and the known energy spectra having the same overall shape comprises having the same overall shape wherein the number of β-delayed gamma rays decreases as the energy increases at energies greater than approximately 3 MeV. 6. The method of claim 1 wherein:the container has dimensions of approximately 8 feet by 40 feet by 8.5 feet and is made of steel;the container holds at least 500 grams of Pu-239;the neutrons have an energy of approximately 14 MeV;the neutrons have a flux of approximately 3.8×104 neutrons/cm2 sec at a distance of approximately 15 feet from the container;detectors surrounding the container on at least three sides are used for the counting; andat least 350 β-delayed gamma rays with energies above 3 MeV are counted. |
|
050230450 | claims | 1. A plant malfunction diagnostic method comprising: a step of determining by simulation a change in a plant state variable at the time of a malfunction and forming a pattern among plant state variables obtained by autoregressive analysis of the change in plant state variable; a step of inserting the formed pattern among the plant state variables in a neural network comprising an input layer, a middle layer and an output layer, and performing learning within a preset precision to decide a model of the neural network as well as connection weight between processing elements (neurons) in the input layer and middle layer and between the middle layer and the output layer; and a step of identifying the cause of the malfunction by inputting a pattern, which indicates the pattern among plant state variables formed from the plant state variable, after detection of an actual plant malfunction. 2. The method according to claim 1, wherein the pattern among said plant state variable is a coherency function indicating a correlation of the frequency spectrum of each plant state variable. |
description | This application claims priority, under Section 371 and/or as a continuation under Section 120, to PCT Application No. PCT/IB2011/001565. filed on Jun. 2, 2011. The present invention generally relates to systems, methods and containers for storing hazardous waste material and, more particularly, to systems, methods and containers for storing nuclear waste material. 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. A container for storing hazardous waste material, according to some embodiments of the present invention, includes a container body, a filling port configured to couple with a filling nozzle and a filling plug, and an evacuation port having a filter, the evacuation port configured to couple with an evacuation nozzle and an evacuation plug. In some embodiments, the evacuation plug is configured to allow air and/or gas to pass through the filter and between the evacuation plug and the evacuation port in a filling configuration. In some embodiments, the evacuation plug closes the evacuation port in a closed configuration. In some embodiments, the evacuation port and the filling port each extend axially from a top surface of the container body. In some embodiments, the container further includes a gasket disposed between the evacuation plug and the evacuation port. In some embodiments, the gasket is comprised of one or more of metal, ceramic or graphite. In some embodiments, the evacuation plug is threadably coupled with the evacuation port. In some embodiments, the evacuation plug and the evacuation port are configured to provide a hermetic seal in a closed configuration. In further embodiments, the evacuation plug and the evacuation plug are configured to be subsequently welded distally to the hermetic seal with respect to the container body in the closed configuration. In some embodiments, the container further includes a lifting member. In some embodiments, the lifting member is substantially co-axial to a longitudinal axis of the container body. In some embodiments, the lifting member includes a projection extending axially from the container body, the projection having a circumferentially extending groove. In some embodiments, the container further includes an evacuation plug. In some embodiments, the evacuation plug includes a thread and the evacuation port is configured to receive the thread of the evacuation plug. In some embodiments, the container body is configured to be hot isostatic pressed. In some embodiments, the container body comprises a vessel configured to be reduced in volume by applying a vacuum to an internal volume of the container body. In some embodiments, the filter is comprised of sinterized material. In some embodiments, the filter is configured to substantially prevent particles having a diameter of at least 10 microns from exiting through the evacuation port. In some embodiments, the filter is welded to the evacuation port. In some embodiments, the filter is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. In some embodiments, the evacuation plug includes a socket. In some embodiments, the evacuation plug and the filing plug each include an inner surface, the inner surfaces each decreasing in diameter in a direction toward the container body. In some embodiments, the inner surfaces are each stepped. A container for storing hazardous waste material according to another embodiment of the present invention includes a container body, a port configured to sealingly couple with a filling nozzle, and a plug including a filter and configured to couple to the port, the plug configured to allow air and/or gas to pass through the filter and between the plug and the port in a filling configuration, the plug closing the port in a closed configuration. In some embodiments, the port is substantially co-axial to a longitudinal axis of the container body. In some embodiments, the port extends axially from a top surface of the container body. In some embodiments, the container further includes a gasket disposed between the plug and the port. In some embodiments, the gasket is comprised of one or more of metal, ceramic or graphite. In some embodiments, the plug is threadably coupled with the port. In some embodiments, the plug includes a thread and the port is configured to receive the thread of the plug. In some embodiments, the plug and the port are configured to provide a hermetic seal. In some embodiments, the plug and the port are configured to be subsequently welded distally to the hermetic seal with respect to the container body in the closed configuration. In some embodiments, the container body is configured to be hot isostatic pressed. In some embodiments, the container body comprises a vessel configured to be reduced in volume by applying a vacuum to an internal volume of the container body. In some embodiments, the filter is comprised of sinterized material. In some embodiments, the filter is configured to substantially prevent particles having a diameter of at least 10 microns from exiting through the evacuation port. In some embodiments, the filter is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. In some embodiments, the filter is coupled to a distal end of the plug. In some embodiments, the plug includes a socket. In some embodiments, the plug includes an inner surface, the inner surface decreasing in diameter in a direction toward the container body. In some embodiments, the inner surface of the plug is stepped. A method of storing hazardous waste material, according to some embodiments of the present invention, includes adding hazardous waste material via a filling nozzle sealingly coupled to a port of a container configured to sealingly contain the hazardous waste material, evacuating the container during adding of the hazardous waste material via a first evacuation nozzle sealingly coupled to the container, heating the container, evacuating the container during heating of the container via a second evacuation nozzle sealingly coupled to the container, inserting a plug into the port, and hot isostatically pressing the container. In some embodiments of the method, the port includes a filling port and the container includes an evacuation port configured to sealingly couple with the first and second evacuation nozzles. In some embodiments, the method further includes welding a filling plug to the filling port to seal the filling port. In some embodiments, the filling plug is welded to the filling port using an orbital welder. In some embodiments of the method, the evacuation port includes an evacuation plug threadably coupled to the evacuation port and allowing air and/or gas to pass through a filter and between the evacuation plug and the evacuation port in a filling configuration and a heating configuration, and wherein the evacuation plug closes the evacuation port in a closed configuration. In some embodiments, the method further includes closing the evacuation plug following heating of the container and welding the evacuation plug to the evacuation port. In some embodiments of the method, the evacuation plug is closed between adding the hazardous waste material and heating the container. In some embodiments, the evacuation plug is closed while the evacuation nozzle is coupled to the evacuation nozzle. In some embodiments of the method, the evacuation plug is welded to the evacuation port using an orbital welder. In some embodiments, the method further includes maintaining a vacuum on the container via the second evacuation nozzle for a period of time following heating. In some embodiments, the method further includes verifying that the vacuum is maintained. In some embodiments of the method, the hazardous waste material is added to the container in a first cell. In some embodiments, the method further includes closing the port in the first cell. In some embodiments, the method includes moving the container to an air interlock between the first cell and a second cell, and moving the container to the second cell. In some embodiments, the first cell is configured to not exchange air with the second cell while at least the container is being filled. In some embodiments, the container is heated in the second cell. In some embodiments of the method, the port includes a filling port and the container includes an evacuation port configured to sealingly couple with the first and second evacuation nozzles. In some embodiments, the method further includes closing the evacuation port using an evacuation plug after adding the hazardous waste material into the container, at least partially opening the evacuation port before heating the container, attaching an evacuation nozzle to the evacuation port before heating the container, closing the evacuation port using the evacuation plug after heating the container, and sealing the evacuation plug to the evacuation port. In some embodiments of the method, the container includes an evacuation port having a filter. In some embodiments, the filter of the evacuation port is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. In some embodiments of the method, the first evacuation nozzle includes a filter. In some embodiments of the method, the hazardous waste material includes calcined material. In some embodiments, the method further includes adding secondary hazardous waste via the filling nozzle into the container. In some embodiments, the secondary hazardous waste includes mercury evacuated from previous containers. In some embodiments, the secondary hazardous waste includes an evacuation filter used during evacuation of previous containers. There is disclosed herein a container for storing hazardous waste material, the container comprising: a container body; a filling port configured to couple with a Idling nozzle and a filling plug; and an evacuation pen having a filter, the evacuation port configured to couple with an evacuation nozzle and an evacuation plug. Preferably, the evacuation plug is configured to allow air and/or gas to pass through the filter and between the evacuation plug and the evacuation port in a filling configuration, the evacuation plug closing the evacuation port in a closed configuration. Preferably, the container further comprises: a gasket disposed between the evacuation plug and the evacuation port. Preferably, the gasket is comprised of one or more of metal, ceramic or graphite. Preferably, the evacuation plug is threadably coupled with the evacuation port. Preferably, the evacuation plug and the evacuation port are configured to provide a hermetic seal in a closed configuration. Preferably, the evacuation plug and the evacuation plug are configured to be subsequently welded distally to the hermetic seal with respect to the container body in the closed configuration. Preferably, the container further comprises; a lifting member. Preferably, the lifting member is substantially coaxial to a longitudinal axis of the container body. Preferably, the lifting member includes a projection extending axially from the container body, the projection haying a circumferentially extending groove. Preferably, the container further comprises: an evacuation ping. Preferably, the evacuation plug includes a thread and the evacuation port is configured to receive the thread of the evacuation plug. Preferably, the container body is configured isostatic pressed. Preferably, the container body comprises a vessel configured to be reduced in volume by applying a vacuum to an internal volume of the container body. Preferably, the filter is comprised of sinterized material. Preferably, the filter is configured to substantially prevent particles having a diameter of at least 10 microns from exiting through the evacuation port. Preferably, the filter is welded to the evacuation port. Preferably, the filter, is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. Preferably, the evacuation port and the filling port each extend axially from a top surface of the container body. Preferably, the evacuation plug icludes a socket. Preferably, the evacuation plug and the filing plug each include an inner surface, the inner surfaces each decreasing in diameter in a direction toward the container body. Preferably, the inner surfaces are each stepped. There is further disclosed herein a container for storing hazardous waste material, the container comprising: a container body; a port configured to sealingly couple with a filling nozzle; and a plug including a filter and configured to couple to the port, the plug configured to allow air and/or gas to pass through the filter and between the plug and the port in a filling configuration, the plug closing the port in a closed configuration. Preferably, the container further comprises: a gasket disposed between the plug and the pot. Preferably, the gasket is comprised of one or more of metal to ceramic or graphite. Preferably, the plug is threadably coupled with the port. Preferably, the plug and the port are configured to provide a hermetic sea. Preferably, the plug and the port are configured to be subsequently welded distally to the hermetic seal with respect to the container body in the closed configuration. Preferably, the plug includes a thread, and the port is configured to receive the thread of the plug. Preferably, the inner surface of the plug is stepped. Preferably, the port is substantially co-axial to a longitudinal axis of the container Linc Preferably, the container body is configured to be hot isostatic pressed. Preferably, the container body comprises a vessel configured to be reduced in volume by applying a vacuum to an internal volume of the container body. Preferably, the filter is comprised of sinterized material. Preferably, the filter is configured to substantially prevent particles having a diameter of at least 10 microns from exiting through the evacuation port. Preferably, the filter is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. Preferably tne port extends axially from a top surface of the container body. Preferably, the plug includes a socket. Preferably, the filter is coupled to a distal end of the plug. Preferably, the plug includes an inner surface, the inner surface decreasing in diameter in a direction toward the container body. There is still further disclosed herein a method of storing hazardous waste material, the method comprising; adding hazardous waste material via a filling nozzle sealingly coupled to a port of a container configured to sealingly contain the hazardous waste material; evacuating the container during adding of the hazardous waste material via a first evacuation nozzle sealingly coupled to the container; heating the container; evacuating the container during heating of the container via a second evacuation nozzle sealingly coupled to the container; inserting a plug into the port; and hot isostatically pressing the container. Preferably, the port includes a filling port and the container includes an evacuation port configured to sealingly couple with the first and second evacuation nozzles. Preferably, the method further comprises: welding a filling plug to the filling port to seal the filling port. Preferably, to filling plug is welded to the filling port using an orbital welder. Preferably, the evacuation port includes an evacuation plug threadably coupled to the evacuation port and allowing air and/or gas to pass through a filter and between the evacuation plug and the evacuation port in a filling configuration and a heating configuration, and wherein the evacuation plug closes the evacuation port in a closed configuration Preferably, the method further comprises: closing the evacuation plug following heating of the container; and welding the evacuation plug to the evacuation port. Preferably, the evacuation plug is welded to the evacuation port using an orbital welder. Preferably, the evacuation plug is closed between adding the hazardous waste material and heating the container. Preferably, the evacuation plug is closed while the evacuation nozzle is coupled to the evacuation nozzle. Preferably, the method further comprises: maintaining a vacuum on the container via second evacuation nozzle for a period of time following heating. Preferably, the method further comprises: verifying that the vacuum is maintained. Preferably, the hazardous waste material is added to the container in a first cell and the method further comprises: closing the port in the first cell; moving the container to an air interlock between the first cell and a second cell; moving the container to a second cell, the first cell configured to not exchange air with the second cell while at least the container is being filled, the container being heated in the second cell, Preferably, the port includes a filling port and the container includes an evacuation port configured to sealingly couple with the first and second evacuation nozzles, the method further comprising: closing the evacuation an evacuation plug after adding the hazardous waste material into the container; at least partially opening the evacuation port before heating the container; attaching an evacuation nozzle to the evacuation port before heating the container; closing the evacuation port using the evacuation plug after heating the container; and sealing the evacuation plug to the evacuation port. Preferably, the evacuation port includes a filter. Preferably, the filter is porous at a first temperature and non-porous at a second temperature, the second temperature being higher than the first temperature. Preferably, the first evacuation nozzle includes a filter. Preferably, the hazardous waste material includes calcined material. Preferably, the method further comprises: adding secondary hazardous waste via the filling nozzle into the container. Preferably, secondary hazardous waste includes mercury evacuated from previous containers. Preferably, the secondary hazardous waste includes an evacuation filter used during evacuation of previous containers. 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 sealing 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 sealed 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 562 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 Dg. 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. A2SiO5) 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 560 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 288 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 292, 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 sealed) 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 600EC). 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 scaling. 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. |
|
062051954 | abstract | This invention provides coded aperture imaging apparatus and methods for the detection and imaging of radiation which results from nuclear interrogation of a target object. The apparatus includes: 1) a radiation detector for detecting at least a portion of the radiation emitted by the object in response to nuclear excitation and for producing detection signals responsive to the radiation; 2) a coded aperture disposed between the detector and the object such that emitted radiation is detected by the detector after passage through the coded aperture; and 3) a data processor for characterizing the object based upon the detection signals from the detector and upon the configuration of the coded aperture. The method includes the steps of: 1) disposing a coded aperture in selected proximity to the object; 2) bombarding the object with a interrogation beam from a source of excitation energy; 3) detecting, with a detector, at least a portion of the radiation emitted in response to the interrogation beam, the detector producing detection signals responsive to the radiation, the detector being disposed so that the coded aperture is between the detector and the object and such that emitted radiation is detected by the detector after passage through the coded aperture; and 4) processing the detection signals to characterize the object based upon radiation detected by the detector after passage through the coded aperture, and based upon the configuration of the coded aperture. |
039420230 | description | The mortar-based protective screen for radiological purposes is shown in FIG. 1 as comprising a jacket 1 made of plastics material of the polyamide or elastomer type, or of any other suitable material which is leakproof and inert towards the human body, and is intended to be placed, directly or indirectly, on the patient's skin. Thus it is also possible to use a plasticised fabric to produce the jacket. This jacket 1 comprises two walls 1a and 1b which extend parallel to one another and have a surface area which is very large relative to that of the side wall 1c joining those two parallel walls 1a and 1b. As shown in FIG. 1, the cross-section of the jacket comprising walls 1a and 1b and the side wall 1c is, for example, of rectangular shape. This jacket 1 is filled with (a) a thermosetting resin, for example of the epoxy, polyester or phenolic type, and with (b) a fine particulate filler of at least one substance which absorbs medical radiation. This fine particulate filler is evenly mixed with the said resin and consists of barium sulphate, antimony oxide, silica or lead oxide, or preferably a mixture of at least two of these components. The diameter of the fine particles forming the filler is of the order of a few microns to a few hundred microns. The mixture of thermosetting resin and filler is referenced 2 in FIGS. 1 to 4. A window 3, permeable to medical radiation, is provided preferably in the centre of the jacket and extends between the two parallel walls 1a and 1b of the jacket 1. It is advantageous to produce this window 3 by means of an elongate solid portion which is made of a material such as polyethylene which is permeable to medical radiation, and the window comprises two parallel end faces one of which is preferably firmly fixed, for example by welding or gluing, to one of the parallel walls 1a and 1b of the jacket 1. The height of this elongate portion 3 is substantially the same as the width of the side wall 1c. Furthermore, it is possible to use the height of this elongate portion 3 as a guide in order to determine the appropriate width of the side wall 1c. At its side, the jacket 1 carries a tubular portion 1d, which, on one side, communicates with the inside of the jacket, and which, at its free end, is generally sealed. When the sealed end of the portion has been cut, it is possible to use this portion to introduce the necessary amount of curing agent inside the jacket. Of course, while the contents of the jacket are being mixed, the resulting aperture must be closed, for example by means of a surgical clamp which presses against the two folded-over branches of the tubular portion. It is also possible to place the curing agent too, beforehand, inside the jacket 1. In this case, in order to avoid premature reaction, the curing agent can be contained within a pouch or sachet which can be opened by rupturing and is located inside the jacket. This rupturable pouch or sachet has been indicated diagrammatically at 5 in FIGS. 1 to 4. In a first embodiment, the curing agent 6 is contained in an elongate sachet 5, placed either directly inside the jacket or inside the tubular extension 1d of this jacket 1. The rupturable sachet 5 is partially fixed to the inside face of the jacket 1 or of the tubular extension. When the sachet 5 is placed in the tubular extension 1d, the join between the sachet and the tubular extension is made in the sealing zone of this tubular extension and this sealing can, for example, be effected by welding or gluing. In the embodiment represented in FIGS. 1 and 2, the rupturable sachet consists partially of a rupturable sheet 5a covering the curing agent 6, and partially of a part of the jacket 1 to which the edge of the rupturable sheet 5a is welded. The jacket of the protective screen may, for example, measure 30 cm. by 30 cm. in plan view, and may have a height of the order of a few centimetres, for example 3 cm. A constituent of the composition 2 may be polyepoxides which are organic compounds containing more than one ##EQU1## group. Such polyepoxides can be saturated or unsaturated; aliphatic, cycloaliphatic, aromatic or heterocyclic; can be substituted, if so desired, by substituents such as chlorine atoms, hydroxyl groups, ether radicals and the like; and can also be monomeric or polymeric. In addition to the polyepoxides described above, diluents or elasticising agents, containing at least 10 and preferably at least 12 carbon atoms, may be added to the compositions. Examples of these agents include, amongst others, pine oil, pine oil distillates, tar, bitumens, polythiopolymercaptans, polyamides, aromatic chlorinated compounds, polyesters, monomeric phthalate esters, long chain acids and long chain compounds containing epoxy groups, and their mixtures. The composition 2 hardens under the action of a curing agent. In some cases, the elasticising agent may contain active hydrogen and can also serve as a curing agent. In other cases, it may be necessary to add an elasticising agent to the curing agent. Suitable curing agents containing epoxy groups may be acidic, neutral or alkaline. Examples of these agents are, amongst others, alkalis, carboxylic acids or anhydrides, Friedel-Crafts halogenated compounds, amino compounds, for example ethylene-diamine, addition products of amines and epoxides, and amide derivatives. The proportions, relative to the binder, vary greatly as a function of the curing agent used; for example, quantities from a few % to 300 or 400% by weight can be employed. The unsaturated polyesters to be added to the composition 2 are organic compounds prepared in a manner which is in itself known from unsaturated .alpha.,.beta.-dicarboxylic acids or their anhydrides, or optionally from saturated dicarboxylic acids, and from polyols, or mixed with a solution of an unsaturated polyester in vinyl and/or allyl monomers. It is also known that it is possible to prepare polyesters from polyols and from acids or their esterifiable derivatives by using, as the acid components, benzene-1-amino or 1-alkylamino-3,5-dicarboxylic acids or their lower alkyl esters. Either of the two reactions takes place equally well, for example, in the presence of 10 to 25% by weight of styrene. The copolymerisation of unsaturated esters with vinyl compounds and mainly styrene is carried out in the presence of catalysts which form free radicals. Peroxides, for example benzoyl peroxide, lauryl peroxide, cumene hydroperoxide and the like, and certain aldehyde, ketone, diketone or amine compounds are generally used as catalysts which form free radicals. It is also possible to use polymerisation initiators based on metal salts or amines. These catalysts are used in amounts of the order of 0.01 to 5% by weight. The proportion of radiologically inert particles present in the composition 2 must be at least 25% by weight of the total mixture of the binder and elasticising agents, and preferably between 50 and 1,000% by weight or, even better, between 100 and 400% by weight of the said total mixture. In a first embodiment the jacket contains, as constituents for the synthetic mortar, the following components: polyester resin of type No. 8,000 100 g. polyester resin of type No. 8,130 20 g. precipitated barium sulphate 100 g. antimony oxide as a fine powder 20 g. silica as a fine powder 10 g. curing agent, methyl ethyl ketone peroxide 2 g. In a second embodiment the jacket contains, as constituents for the synthetic mortar, the following components: epoxy resin 100 g. diglycidyl-ethyl 10 g. pine oil 10 g. curing agent, diethylene-triamine 9 g. precipitated barium sulphate 100 g. lead oxide 20 g. silica as a fine powder 10 g. It is advantageous not to use jackets which are thicker than 6 cm. and, in the case where the protective screen must have a greater thickness, it is preferable to superpose several jackets each having a height of less than 6 cm. Where the synthetic mortar composition hardens exothermically, and thus produces a temperature gradient which is too high for the skin of the patient to tolerate, a flexible and heat insulating mass, for example a sheet 7 of foam rubber may advantageously be provided on the wall 1b of the jacket facing the patient. This sheet 7 can be positioned outside or inside the wall of the jacket 1. If it is positioned inside the jacket 1, at least its edge must then adhere, in a leak-proof manner, to the wall 1b. On the other hand, if the sheet 7 is provided on the outer face of the jacket 1, it can then be fixed to the wall 1b by only a few welding or gluing points, and this further improves the heat insulation. It should be understood that the above description is given by way of example only and the scope of the invention should not be considered as being restricted to the specific details given. Modifications can readily be incorporated without departing from the scope of the invention as defined in the following claims. |
040070859 | summary | BACKGROUND OF THE INVENTION This invention relates to nuclear reactor fuel elements. It has been proposed that nuclear reactor fuel elements of the kind in which nuclear fuel material is enclosed in a protective sheath should be individually identifiable by providing within each sheath an insert which is unique to that fuel element. Such inserts may conveniently be provided with markings whereby a digital output train of pulses can be produced, on passage of the fuel element through an activated eddy current head, for example, and de-coding equipment can readily be devised to accept these sequential pulses. However this is a time dependent function and if the position of the markings is to be correctly identified it is essential to control strictly the relative movement of insert and marking detector. This may be difficult. SUMMARY OF THE INVENTION According to the present invention an insert for a nuclear reactor fuel element whereby the fuel element may be individually identified bears an array of markings at spaced positions to produce a series of signals on presentation of the insert to a detector responsive to each marking and further markings at some only of said positions to vary the series of signals produced on presentation to the detector. The insert can thus provide both clock pulses from the array of markings at spaced positions and numerical identification pulses from the further markings to ensure uncritical time dependence when the insert is passed through a detector. Also according to the invention is a nuclear reactor fuel element comprising nuclear fuel material within a protective sheath and including within the sheath an insert in accordance with the preceding paragraph. Preferably the insert is a bar of metal in which all the markings are holes extending through the bar, the further markings being at right angles to the array at spaced positions. As an alternative to holes the markings may be grooves with the further markings provided by deepening the grooves. The bar may be hollow, that is tubular in form. For the further markings which are to provide the individual identification a binary coded decimal system is preferred, that is one in which four markings are provided for each decade to represent 8, 4, 2, and 1. A further marking may also be provided similarly to provide a lead-in pulse for each decade. |
claims | 1. A device for closing and opening a beam path of electromagnetic and/or ionizing radiation, comprising:at least one part of a shutter body which is permanently situated in the beam path and rotatable about a longitudinal axis situated essentially transversely with respect to the beam path, and which contains a material that is opaque to the radiation and blocks the beam path when the shutter body is in a closed rotary position, and which defines a passage that is transparent to the radiation when in an open rotary position, and wherein the at least one part of the shutter body hermetically seals the beam path;a magnetic drive which is coupled to the shutter body for rotation of same about the longitudinal axis between the rotary positions the magnetic drive configured for moving the shutter body between the rotary positions, wherein at least one of the rotary positions corresponds to a stable position of the magnetic drive which maintains the magnetic drive without current, and wherein the magnetic drive comprises a bistable electric solenoid drive having two stable end positions; andat least two permanent magnets to hold the bistable electrical solenoid drive in a predetermined position without current, wherein the magnetic drive is operated in an overload range to achieve maximum acceleration. 2. The shutter device according to claim 1, wherein the magnetic drive is a solenoid drive or a linear magnetic drive. 3. The shutter device according to claim 1, wherein the shutter body is situated in a device for shaping the radiation, such as a collimator, or is situated on a device for generating the radiation, such as a X-ray tube. 4. The shutter device according to claim 1, wherein in the open rotary position, inner surfaces of the passage which are directed toward the beam path are designed in such a way that the inner surfaces are aligned with housing surfaces which delimit the beam path, do not limit the free cross section of the beam path, or define the free cross section of the beam path. 5. The shutter device according to claim 1 wherein the shutter body, at least in the area that is permanently situated in the beam path, has the shape of a half-cylinder or cylindrical section, or has the shape of a solid cylinder with the passage extending essentially radially through the shutter body. 6. The shutter device according to claim 5, wherein the passage extending essentially radially through the shutter body comprises a rectangular cross section. 7. The shutter device according to claim 1, wherein end stops associated with the two rotary positions are provided on the shutter body or the magnetic drive in such a way that the shutter body or the magnetic drive is movably only in a range defined by the stops, which essentially corresponds to a 90° rotation of the shutter body. 8. A method for opening and closing a beam path for electromagnetic and/or ionized radiation, comprising the following steps:rotating a part of a shutter body which is permanently situated in the beam path and rotatable about a longitudinal axis situated essentially transversely with respect to the beam path, and which is made of a material that is opaque to the radiation, into an open rotary position, so that a passage which is formed in the shutter body and is transparent to the radiation is brought into alignment with the beam path, and wherein the part of the shutter body hermetically seals the beam path;rotating the shutter body situated in the beam path into a closed rotary position, so that the beam path is closed by material of the shutter body which is opaque to the radiation; andcarrying out the particular rotary motions of the shutter body between the rotary positions via a magnetic drive, and holding at least one of the rotary positions of the magnetic drive, without current, by means of a permanent magnet associated with this rotary position, and wherein the magnetic drive comprises a bistable electric solenoid drive having two stable end positions, wherein at least two permanent magnets hold the bistable electrical solenoid drive in a predetermined position without current, wherein the magnetic drive is operated in an overload range to achieve maximum acceleration. 9. An X-ray inspection system comprising an X-ray source, a shutter device according to claim 1, and a control device which is operatively connected to the shutter device and is configured for controlling the shutter device using the method according to claim 8. |
|
043022944 | summary | The invention relates to improvements in nuclear reactor fuel assemblies. Fuel assemblies, in particular those which are used in water-cooled reactors, generally consist of a bundle of parallel fuel rods held apart by means of cross-pieces which extend transversely relative to the rods. Tubes for supporting the assembly are substituted for some of the fuel rods, and these support tubes, which are longer than the rods, are themselves joined to end plates which make it possible to achieve a good mechanical rigidity of the assembly and to hold the fuel rods in the longitudinal direction. When the nuclear reactor core is recharged, it is extremely useful to be able to remove defective fuel rods and to replace them, so as to be able to continue using the whole assembly, with the removal, from the reactor circuit, of the contamination caused by fission products originating from the defective fuel rods. This ability to remove rods provides the advantage that it economizes in terms of fuel since, once the assembly has been reconstituted, it can be recharged and re-used, after the defective rods have been removed and replaced. Furthermore, replacement of defective rods enables the reactors to be run with greater safety for the installations and the personnel by reducing the contamination risks. However, changing defective rods is an operation which requires considerable precautions; the irradiated element can ony be repaired under radiological protection. In general, this operation is carried out under a certain depth of water in a swimming pool adjacent the reactor. Furthermore, for certain more specific reasons, for example in order to carry out examinations of some of the rods or to perform mechanical or physical tests on these rods, it can be of value for the user to be able to dismantle a fuel assembly in order to recover those bars on which the tests or examinations are to be carried out. In order to have access to these fuel rods and to be able to separate them from the remainder of the assembly, it is necessary to dismantle the end plates; this dismantling operation in the swimming pool presents certain difficulties because the assembly is only accessible by remote control and because the dismantling operation is performed on an immersed assembly. For the purpose of facilitating the operation for dismantling and reassembling the fuel rods in the assembly inside the swimming pool, it has been proposed to detachably fix the end plates to the support tubes so that the end plates can be easily dismantled by remote control when the assembly is inside the swimming pool. To enable this detachable fixing of the end plates to the support tubes, screw-threaded sockets, for example, have been proposed which sockets are screwed inside the support tubes. Each socket has a shaped enlarged portion which fits in a housing, of corresponding shape, provided on the inner face of the end plate. Rotation of the sockets is prevented by expanding part of the sockets inside the housings provided in the end plate. This proposal makes it possible to dismantle the end plate from the support tubes rapidly and easily; however, when the end plates have been dismantled, the ends of the support tubes are no longer held in the transverse direction relative to one another, and, when reassembling, it is necessary to realign the end plate with the set of support tubes which must be inserted in the housings provided in the end plate. According to the invention there is provided a nuclear reactor fuel assembly comprising a bundle of parallel fuel rods held apart by cross-pieces which extend transversely relative to said rods, and tubes, for supporting the assembly, which are substituted for some of said fuel rods, said support tubes being longer than said rods and ensuring, in cooperation with transverse end plates, the rigid assembling of the whole, said plates also ensuring the longitudinal support of said rods, wherein, at one or both of their end, said support tubes are fixed to a grid extending transversely relative to said support tubes, said grid forming a network of cells which approximately correspond, in size and position, to said fuel rods, so that a said rod can pass by longitudinal displacement through a said cell of said grid, said grid being detachably fixed to the corresponding said end plate which is positioned outwardly of said grid, by cylindrical sockets which extend through passages provided in said end plate and through some of said cells of said grid, each said socket comprising: bearing surfaces at the level of the outer face of said end plate and the inner face of said grid, means at the outer end of said socket for cooperation with a tool for rotating said socket between a first position in which said socket can be freely inserted into said cell of said grid, and a second position in which said bearing surface at the level of said inner face of said grid comes into contact with said inner face of said grid to lock said grid to said end plate, and a deformable part which is expandable into one or more housings provided at the level of said passage in said end plate or of said cell of said grid to prevent rotation of said socket relative to said grid. |
summary | ||
055764681 | claims | 1. A resin encapsulated waste agglomerate, comprising: a substantially rigid agglomerate formed by mixing waste material with a thermosetting binder and subjecting the mixture to heat and pressure; an agglomerate enclosing coat of resin formed from a thermoplastic sheet fused to said agglomerate by application of heat and pressure; and, an enclosing jacket of resin formed from a thermoplastic particulate resin approximately one centimeter thick fused to said coat by application of heat and pressure. a substantially rigid agglomerate formed by mixing waste material with a thermosetting binder and subjecting the mixture to heat and pressure; an agglomerate enclosing coat of resin formed from a thermoplastic sheet fused to said agglomerate by application of heat and pressure; and, an enclosing jacket of resin formed from a thermoplastic particulate resin approximately two millimeters to two and one-half centimeters thick fused to said coat by application of heat and pressure. 2. The encapsulated agglomerate of claim 1 wherein said thermosetting binder comprises atactic 1,2-polybutadiene. 3. A resin encapsulated waste agglomerate, comprising: 4. The encapsulated agglomerate of claim 3 wherein said thermosetting binder comprises atactic 1,2-polybutadiene. |
abstract | A charged particle apparatus is equipped with a third stigmator positioned between the objective lens and a detector system, as a result of which a third degree of freedom is created for reducing the linear distortion. Further, a method of using said three stigmators, comprises exciting the first stigmator to reduce astigmatism when imaging the sample, exciting the second stigmator to reduce astigmatism when imaging the diffraction plane, and exciting the third stigmator to reduce the linear distortion. |
|
048329037 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The storage arrangement comprises a chamber 1 having walls, floor and ceiling of reinforced concrete. The ceiling 2 of the chamber is pierced by a matrix of openings 3, the upper ends of which have a slightly greater diameter than the lower ends as shown more clearly in FIG. 2, ledge 4 joining the parts of greater and lesser diameter. Each of the openings 3 has, located within it, the upper end of a storage tube 5 of steel, the top section 6 of the part of the tube within the opening 3 being of greater diameter than the remainder and being connected to it by an annular shoulder 7. The stepped annular gap so formed provides attenuation to radiation from the tube contents. The tubes 5 are supported from the floor 8 of the chamber 1 and are free to expand thermally upwards through openings 3. The tubes 5, which are closed at the bottom, serve to store irradiated nuclear fuel as at 9, and their upper ends are closed by plugs 10 which are supported within the tubes by the inner surfaces of the inclined shoulders 7. A double seal 11 is provided between the top of each plug and the tube, and each tube/plug assembly is surmounted by a removable tile 21, the tiles together forming the floor of a charge hall 12 (FIG. 3), the walls and roof of which are also of reinforced concrete, and in which there is located a charge machine and gantry 13, of any convenient construction, for introducing fuel into and removing it from the storage tubes 5. The floor of the chamber 1 has an opening 14 which communicates, via an inlet duct 15, with a louvred air inlet 16 (FIG. 3), and the ceiling 2 of the chamber has a further opening 17 leading into a discharge stact 18 extending upwards to a louvred air outlet 19 disposed a distance above the air inlet 16. The opening 17 is located at the opposite side of the chamber to the inlet opening 14, and in use decay heat from spent fuel 9 within the storage tubes 5 is transferred to the walls of the tubes by conduction, convection and radiation. The heat is removed from the tubes 5 to the atmosphere by a natural thermosyphon process, the heated air rising within the outlet stack 18 by convection, and being replaced by cooler air drawn into the chamber 1 through the inlet opening 14. The disposition of the inlet and outlet openings 14, 17 at opposite sides of the chamber 1 ensures that there is a flow of air between the tubes in a direction transverse to the tube axes, as well as vertically, as indicated by the arrows in FIG. 1, which gives rise to optimum cooling. It will be seen that the amount of air flow is governed by the heat generated within the store, and the arrangement is designed to be sure that the cooling is adequate to maintain the fuel within the tubes at a safe temperature, consistent with the gas used within the tubes. In accordance with the invention, the interior of each tube 5 communicates, by means of a pipe 20, connected to the side of the tube between the seals 11, to manifolds 22 common to a plurality of tubes, and by which the gas within the tubes can be changed and the pressure controlled. A hole in the plug 10 leads from the space between the seals 11 to a cavity in the base of the plug 10. The cavity is closed by a porous metal filter 12 that limits the passage of radioactive particles passing into the pipe 20. A suitable service point valve 23 associated with each tube allows the pipe 20 to be connected to a manifold 22 leading to an air extraction system 24 or to a second manifold 22 leading to an alternative gas system 25. The service point valve 23 also allows individual storage tubes to be isolated from the manifolds 22, if desired, in order to permit a rapid segregation of tubes should a fault condition occur, thereby enabling the fault position to be speedily traced. The air system 24 incorporates a suitable filter 26 followed by a flow measuring device 27, a one-direction valve 28 permits the outflow of gas from the air system 24. An exhauster 31 connected after the one-direction valve 28 allows gas to be drawn from the system until a depression is established in the tubes 5. The exhauster 31 is operated when leak checking of the connected tubes 5 are required or when leaks have been established because of unexpected faults. The discharge from the exhauster 31 passes to atmosphere. The flow measuring device 27 in conjunction with the operation of the exhauster 31 provides a measure of the leak tightness of the tubes 5. The depression that is maintained within the storage tubes 5 supplements the high integrity sealed enclosure for the fuel created by the tube 5 and the sealed plug 10, as any leakage that occurs will be into the tubes 5. Similarly, any leakage that occurs at the seals 11 will also be inwards. The use of an exhauster and a flow measuring device connected to pipes to the individual storage tubes has the practical advantage of enabling any fault conditions to be more rapidly detected. A gas sampling point 29 allows the radioactive and moisture content of the gas in the system 24 to be measured by suitable instruments. Air drying equipment 30 supplies dry air to the system 24 when the pressure in the tubes 5 falls below a level set by suitable valves in the drying equipment 30. The natural atmospheric temperature variations that act on the tubes 5 via the thermosyphon cooling system cause the air within the tubes to expand or contract. Expanding air passes from the system to the atmosphere via the filter 26 and one-way valve 28, carrying with it some water vapor. The inflow of air required as the air cools and contracts is supplied from the drying equipment 30. The alternative gas system 25 incorporates a suitable flow measuring device 32, a source of low pressure gas 33 and a suitable pressure relief device 34 that vents to the air system upstream of the filter 26. The gas source 33 supplies gas at the required pressure to the tubes 5 when connected via the service point valve 23. The maxium pressure within the tubes is set by the pressure relief device 34 that allows excess gas to pass from the tubes 5 to atmosphere via the filter 26. The flow of gas measured with the flow measuring device 32 provides a measure of the leak tightness of the tubes 5. If desired, portable monitors may also be used to check the conditions within the tubes 5 periodically by connection to tapping points on the individual pipes or to separate pipes communicating with the interiors of the respective tubes either through the tube walls or through the respective plugs 10. The storage arrangement described has the advantage that it can be used to store uncanned fuel, enabling inspection and monitoring to be readily carried out. Moreover, removal of fuel from the store, should this become necessary, can be speedily effected, without the need to interfere with neighboring storage tubes, simply by removing the respective tile and plug. Re-use of the storage tubes after fuel removal is an operational option. Moreover, the manner of supporting the storage tubes 5 enables a tube to be readily withdrawn upwards into the charge hall for examination or replacement should this be required. Although the tiles 21 have been shown resting on the tops of the tubes 5, the skirt portions 27 of the tiles can be extended, if desired, so that these rest upon the surface of the concrete, giving a more even floor surface to the charge hall 12. The storage tubes 5 will normally be of circular cross-section, but this is not essential, and other shapes may alternatively be employed. In some cases, the tubes may carry external cooling fins of any convenient configuration to enhance the cooling effect of the air flow. FIG. 3 shows a plurality of independent storage arrangement modules as described above, sections of which are shown at C, combined to form a nuclear storage structure having, in this case, a common receipt/dispatch building A for spent fuel or vitrified waste, and associated with a storage submodule for initially storing waste fuel before transfer to the charge hall and storage tubes. The charge hall 12 is common to all the storage modules, as in the gantry 13. |
Subsets and Splits
No community queries yet
The top public SQL queries from the community will appear here once available.