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claims | 1. A spacer grid for placing and supporting a plurality of longitudinal fuel rods in a nuclear reactor fuel assembly, comprising a plurality of inner strips arranged while intersecting each other at right angles prior to being encircled with four perimeter strips, thus forming an egg-crate pattern, said inner and perimeter strips each being fabricated with a plurality of unit strips arranged in parallel, and forming a plurality of unit rod cells for receiving and supporting the fuel rods therein while isolating the fuel rods from each other, each unit strip of said inner strips having one grid spring formed on a central portion of the unit strip while projecting in a direction, and two dimples formed on the unit strip at positions above and under the grid spring while projecting in a direction opposite to the grid spring, and each unit strip of said perimeter strips having one grid spring formed on a central portion of the unit strip while projecting in a direction, wherein said grid spring comprises: a vertical support part including: an opening formed at the central portion of the unit strip; upper and lower base parts extending downward and upward from central portions of top and bottom edges of said opening; and two bridge parts branched from the upper base part and extending symmetrically downward until they are united at the lower base part; and a fuel rod support part including: a conformal support part brought into surface contact with an external surface of a fuel rod; and two transverse connection parts extending outward from opposite outside edge of said conformal support part while being specifically bent, and integrated with the central portions of inside edges of said two bridge parts into a single structure, thus connecting the fuel rod support part to the vertical support part wherein said vertical support part including said upper and lower base parts and said two bridge parts has a strength which is lower than that of said fuel rod support part having said conformal support part wherein the vertical support part is elastically deformed before deformation of the conformal support part so as to allow the conformal support part to maintain a conformal contact with the fuel rod upon application of a force from the fuel rod to the grid spring. 2. The spacer grid according to claim 1 , wherein said vertical support part has a plurality of bent portions and projects from the unit strip toward the fuel rod, thus elastically supporting load applied from the fuel rod thereto through the conformal support part 45 coming into direct contact with the fuel rod. claim 1 3. The spacer grid according to claim 1 , wherein said conformal support part has the same radius of curvature as that of the fuel rod, thus being brought into surface contact with the external surface of the fuel rod, said conformal support part also has a circular or elliptical profile at its contact surface, thus enlarging a surface contact area thereof relative to the fuel rod, in addition to accomplishing a uniform contact pressure distribution and reducing a peak stress thereof. claim 1 4. The spacer grid according to claim 1 , wherein said vertical support part includes the upper and lower base parts, two bridge parts and a plurality of bent portions. claim 1 5. The spacer grid according to claim 1 , wherein said vertical support part is bent at a plurality of portions so as to have an enlarge elastic range and elastically supporting the fuel rod regardless of a variation in fuel rod support conditions in the nuclear reactor, and maintaining a spring force capable of effectively supporting high load that is applied from the fuel rod and exceeds the expected maximum load and the elastic limit of the grid spring. claim 1 6. The spacer grid according to claim 1 , wherein said base parts and bridge parts of the vertical support part are changeable in their widths and positions of their bent portions in accordance with design objects, thus providing optimized spring characteristic curves. claim 1 7. The spacer grid according to claim 1 , wherein said grid spring have upper and lower opening surrounded by the vertical support part and the fuel rod support part, and further include upper and lower extending part each extending from center of upper/lower base part of the vertical support part and partially enclosing said upper and lower opening. claim 1 |
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claims | 1. Particle storing apparatus comprising:a container in which an electrostatic field, and substantially no magnetic field, is restraining charged particles moving in an ultra-high vacuum from striking an internal surface of the container for a half-life of at least 1 hour, said electric field configured to synthesize a piecewise integrable optical system such that said restraining is caused by movement of the charged particles through the electrostatic field. 2. The apparatus of claim 1, wherein motion of the charged particles within the electrostatic field provides three-dimensional focusing that constrains the particles from striking the surface. 3. The apparatus of claim 1, wherein the electrostatic field is shaped by at least one electrode. 4. The apparatus of claim 1, wherein the apparatus provides storage for the charged particles with the electrostatic field being generated by electrode voltage that is constant. 5. The apparatus of claim 1, wherein the apparatus provides storage for the charged particles with the electrostatic field being generated by electrode voltage that is modified based on measured properties of the charged particles. 6. The apparatus of claim 1, wherein apparatus modifies the electrostatic field, based on measured properties of the charged particles, to allow injection of charged particles into the container. 7. The apparatus of claim 1, wherein the apparatus modifies the electrostatic field, based on measured properties of the charged particles, to allow extraction of charged particles from the container. 8. The apparatus of claim 1, wherein the electrostatic field is axisymmetric along a direction of particle motion. 9. The apparatus of claim 1, wherein the electrostatic field is partially comprised of quadrupole fields that provide some of the focusing of the particles. 10. The apparatus of claim 1, further including a linear array of electrodes that maintain motion of the particles in a direction of the linear array. 11. The apparatus of claim 1, further including a toroidal array of electrodes that maintain motion of the particles in an azimuthal direction. 12. The apparatus of claim 9, further including a linear array of electrodes that sustain motion of the particles in a direction of the linear array. 13. The apparatus of claim 9, further including a toroidal array of electrodes that sustain motion of the particles in an azimuthal direction. 14. The apparatus of claim 1, wherein the apparatus provides for modifying the electrostatic field, based on measured properties of the charged particles, during storage of charged particles to maximize their lifetime. 15. The apparatus of claim 9, wherein the apparatus provides for modifying the electrostatic field, based on measured properties of the charged particles, during storage of the particles to maximize their lifetime. 16. The apparatus of claim 1, wherein at least some of said inside surface is comprised of titanium. 17. The apparatus of claim 1, wherein the apparatus shapes the electrostatic field to enable storage of the particles for a half-life exceeding one day. 18. The apparatus of claim 1, wherein the apparatus shapes the electrostatic field to enable storage of the particles for a half-life exceeding one week. 19. The apparatus of claim 1, wherein the surface is comprised of a material that getters residual gas molecules. 20. The apparatus of claim 1, wherein the vacuum is maintained by using ion-sputter pumping. 21. The apparatus of claim 1, wherein the vacuum is maintained by surrounding the vacuum with a vacuum envelope that limits hydrogen diffusion. 22. The apparatus of claim 1, wherein the vacuum is sufficient to enable storage of the particles for a half-life exceeding one hour. 23. The apparatus of claim 1, wherein the vacuum is sufficient to enable storage of the particles for a half-life exceeding one day. 24. The apparatus of claim 1, wherein the vacuum is sufficient to enable storage of the particles for a half-life exceeding one week. 25. The apparatus of claim 1, further including a means for controlling the temperature of the charged particles. 26. The apparatus of claim 25, wherein the means for controlling the temperature includes a stochastic cooling means. 27. The apparatus of claim 25, wherein the means for controlling the temperature includes an electron cooling means. 28. The apparatus of claim 1, further including a portable power source to maintain the electric field during transportation. 29. The apparatus of claim 21, further including a portable power source to maintain pumping on the vacuum envelope. 30. The apparatus of claim 1, wherein the particles include antiprotons. 31. The apparatus of claim 30, further including an injector of a gas. 32. The apparatus of claim 31, wherein the gas is hydrogen. 33. The apparatus of claim 31, wherein the apparatus produces gamma-rays by the stored antiprotons annihilating with the injected gas. 34. The apparatus of claim 31, wherein the apparatus produces pi-mesons by the stored antiprotons annihilating with the injected gas. 35. The apparatus of claim 31, wherein the apparatus produces neutrons by the stored antiprotons annihilating with the injected gas. 36. The apparatus of claim 31, wherein the apparatus produces secondary elementary particles by the stored antiprotons annihilating with the injected gas. 37. The apparatus of claim 3, further including at least one additional electrode. 38. A method of storing particles, the method comprising:generating an electrostatic field configured to synthesize a piecewise integrable optical system within a container so as to restrain, while substantially no magnetic field restrains, charged particles moving with respect to said electrostatic field from striking an internal surface of the container; andwhile the charged particles are restrained, maintaining an ultrahigh vacuum in said electrostatic field wherein the charged particles are movingly restrained. 39. The method of claim 38, wherein motion of the charged particles within the electrostatic field provides three-dimensional focusing that constrains the particles from striking the surface. 40. The method of claim 38, wherein the electrostatic field is shaped by at least one electrode. 41. The method of claim 38, wherein the electrostatic field being generated by electrode voltage that is constant when storing particles. 42. The method of claim 38, wherein the electrostatic field being generated by electrode voltage that is modified, based on measured properties of the charged particles, when storing the particles. 43. The method of claim 38, wherein the electrostatic field is modified, based on measured properties of the charged particles, to allow injection of charged particles into the container. 44. The method of claim 38, wherein the electrostatic field is modified, based on measured properties of the charged particles, to allow extraction of charged particles from the container. 45. The method of claim 38, wherein the electrostatic field is axisymmetric along a direction of particle motion. 46. The method of claim 39, wherein the electrostatic field is at least partially comprised of quadrupole fields that provide some of the focusing. 47. The method of claim 38, wherein the electrostatic field is modified, based on measured properties of the charged particles, during storage of the particles to maximize their lifetime. 48. The method of claim 46, wherein the electrostatic field is modified, based on measured properties of the charged particles, during storage of the particles in order to maximize their lifetime. 49. The method of claim 38, wherein the electrostatic field is shaped to allow storage of the particles for a half-life exceeding one hour. 50. The method of claim 38, wherein the electrostatic field is shaped to allow storage of the particles for a half-life exceeding one day. 51. The method of claim 38, wherein the electrostatic field is shaped to allow storage of the particles for a half-life exceeding one week. 52. The method of claim 38, wherein the vacuum is maintained by surrounding the vacuum system with a secondary vacuum envelope. 53. The method of claim 38, wherein the vacuum allows storage of the particles for a half-life exceeding one hour. 54. The method of claim 38, wherein the vacuum allows storage of the particles for a half-life exceeding one day. 55. The method of claim 38, wherein the vacuum allows storage of the particles for a half-life exceeding one week. 56. The method of claim 38, further including controlling temperature of the particles. 57. The method of claim 56, wherein the controlling includes stochastic cooling. 58. The method of claim 56, wherein the controlling includes electron cooling. 59. The method of claim 38, wherein the particles include antiprotons. 60. The method of claim 38, further including extracting the particles. 61. The method of claim 60, further including injecting the particles into a particle accelerator. 62. The method of claim 60, further including producing, with the particles, a therapeutic treatment of a medical condition. 63. The method of claim 60, further including detecting, with the particles, an isotope. 64. The method of claim 60, further including producing, with the particles, an isotope. 65. The method of claim 60, further including inducing, with the particles, nuclear fission. 66. The method of claim 60, further including inducing nuclear fusion. 67. The method of claim 60, further including producing, with the particles, an image. 68. The method of claim 60, further including catalyzing, with the particles, a chemical reaction. 69. The method of claim 38, further including transporting the container with the particles. 70. The method of claim 69, further including cushioning the container from a transportation vehicle to reduce deleterious effects of vibration during the transporting. 71. The method of claim 69, wherein the transporting includes transporting on a motorized vehicle. 72. The method of claim 69, wherein the transporting includes transporting on a rocket. 73. The method of claim 69, wherein the transporting includes transporting by a person. 74. A method for transporting antiprotons to a point of use, the method comprising:providing an antiproton confinement region devoid of a confining magnetic field;maintaining said antiprotons confinement region at an ultra-low pressure;establishing a controllable electrostatic field configured to synthesize a piecewise integrable optical system in said antiproton confinement region;modifying said electrostatic field, based on measured properties of the charged particles, in restraining said antiprotons in said antiproton confinement region, wherein said antiprotons are restrained by travelling through the electrostatic field;transporting said antiprotons to a point of use while maintaining said antiproton confinement region at an ultra-low pressure; andmodifying said electrostatic field, based on measured properties of the charged particles, to urge said antiprotons from said antiproton confinement region. 75. The method of claim 74, wherein the said antiproton confinement region is maintained above cryogenic temperatures. 76. An apparatus comprising:a means for injecting antiprotons into a container suitable for transporting antiprotons, said container comprisingan electrostatic field configured to synthesize a piecewise integrable optical system, and essentially no magnetic field, restraining the antiprotons from striking an inside surface of said container, wherein said restraining is caused by the antiprotons travelling through the electrostatic field;a means for maintaining an ultrahigh vacuum within the container in the region of the antiprotons;a means of for injecting a gas into the container to cause annihilations of said antiprotons, creating secondary particles used to interrogate a shielded container for nuclear materials. 77. An apparatus for transporting antiprotons, the apparatus comprising;a substantially evacuated cavity in a container not comprising a dewar;at least one antiproton trap within said cavity, said trap utilizing an electrostatic field, configured to synthesize a piecewise integrable optical system, coupled with motion of antiprotons within said electrostatic field to provide antiproton confinement; anda sealable cavity access port providing access to said substantially evacuated cavity for selective introduction into and removal from the cavity of said antiprotons. 78. A method of transporting a plurality of antiprotons to a desired location, the method comprising:fastening a container to a source of antiprotons to receive a plurality of antiprotons;providing an electrostatic field to cause said plurality of antiprotons to move into a passageway integrally formed within said container;using said electrostatic field, configured to synthesize a piecewise integrable optical system, coupled with antiproton motion within said electrostatic field, but without using substantially any magnetic field, to trap said plurality of antiprotons;detecting aid antiprotons;cooling said antiprotons;delivering the container to the desired location. 79. Particle transporting method including:restraining, with an electrostatic field and substantially without a confining magnetic field, charged particles from striking a container surfaces, wherein said restraining is caused by the charged particles travelling through the electrostatic field which is configured to synthesize a piecewise integrable optical system;maintaining an ultra-high vacuum with said container; andtransporting the container. 80. The method of claim 79, wherein the restraining includes restraining with only one electrostatic field. 81. The method of claim 79, wherein the restraining is devoid of more than one electrostatic field. 82. The method of claim 79, wherein the restraining produces a half-life of at least 1 hour. 83. The method of claim 79, wherein the surface is not a cryogenically cold surface. 84. The method of claim 79, wherein the restraining is devoid of more than one controllable electric field. 85. The method of claim 79, wherein the container is not a dewar. 86. The apparatus of claim 1, further comprising a battery to provide power for portable operation of said apparatus, wherein a power level of 10 Watts is sufficient for said restraining and maintaining said vacuum. |
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description | This application is a U.S. national phase filing of Int'l Application Ser. No. PCT/US15/37550, with an international filing date of Jun. 24, 2015, which claims the benefit of and priority to U.S. Provisional Application Ser. No. 62/018,030, filed Jun. 27, 2014, and to U.S. Provisional Application Ser. No. 62/025,776, filed Jul. 17, 2014, each of which is entitled “RADIATION THERAPY WITH SEGMENTED BEAMS OF PROTONS AND OTHER IONS,” the entirety of each of which is hereby incorporated herein by reference thereto. The present invention relates to methods for performing minibeam radiation therapy for treating tumors, neurological targets, and other diseases, and, particularly, to methods of delivering therapeutic segmented beams of protons and other ions, particularly, light ions. Proton therapy has become a significant radiation therapy around the world with more than ten facilities currently operating in the United States alone and several more in the making. Furthermore, although other light ions, e.g., deuterons, tritons, He-3 and He-4 ions, Li-6 and Li-7 ions, and ions of beryllium and boron, have not been used for radiation therapy, several types of accelerators currently used for proton and carbon therapy are capable of accelerating such light ions. Finally, radiation therapy with carbon ions, although not in clinical use in the United States yet, has been in clinical use in Japan and Germany for close to 15 years. As used herein, the term “charged particles” refers generally to ions of elements from the periodic table of elements, of any atomic number. In addition, the term “particle therapy,” as used herein, refers to radiation therapy using any charged particle or ion. While the term “light ions” may be used to refer to any ions or charged particles from protons to neon (from Z=1 to 10 inclusive), the methods of the present disclosure are particularly suited for ions and charged particles from protons to boron. The advantages of light ions for radiation therapy over x-rays and gamma rays used in conventional radiation therapy are mostly their Bragg peak feature of dose deposition in tissues that allows better confinement of the dose to the target, as depicted in FIG. 2. FIG. 1 compares the dose deposition 10 with depth in tissues 12 of protons and carbon ion beams with that of high energy x-rays called MV x-rays because they are produced by MV electrons linacs. Proton therapy and carbon therapy of tumors are implemented by spreading the Bragg peak 14 to produce a flat dose 16 over the length of the tumor with only little exit dose (FIG. 2). Representative plots showing the Bragg peak for different energies 14 of carbon ions are shown as an example in FIG. 2, along with the resultant flat dose 16 resulting from Bragg-peak spreading across the energies. However, protons, carbons, and other charged heavy particles have a major disadvantage over MV x-rays and gamma rays used in Gamma Knife and that is that they do not have the sparing effect of the shallow tissues of the MV x-rays and gamma rays, an effect loosely called “skin-sparing effect”, which is demonstrated in FIG. 1 as a large dip in the entrance dose of MV x-rays. Therefore, despite the fact that the Bragg-peak-spread dose protons and carbons deliver to the target is larger than their entrance doses, their entrance dose is still much higher than that of MV x-rays and gamma rays. This limitation of protons, carbon ions, and particle beams in general, combined with the very high radiosensitivity of the skin and certain other shallow tissues, such as the brain's frontal cortex, as described below, limits the entrance dose in each therapy session from these particles as compared to possible entrance doses from MV x-rays and gamma rays. Therefore, in contrast to MV x-rays and gamma rays that, depending on the size of the target, can be given in a single session (called single dose fraction) or just a few sessions, proton therapy is mostly administered in 20 to 30 dose fractions, which would be four to six weeks of five treatments each week. Although the number of dose fractions used in carbon therapy is generally smaller because of the high relative biological effectiveness (RBE) of carbon ions, which is mostly at their Bragg Peak, it still ranges from 5 to 15 because of their above lack of sparing effect of shallow tissues. The sparing effect in shallow tissues by MV x-rays and gamma rays is well known, and is a result of the mechanism by which the dose is deposited. Being neutral particles, the x-rays and gamma rays deposit a dose by setting electrons in motion via either the photoelectric effect, Compton scattering, or pair-production, which in turn deposit the dose in tissue. For MV x-rays, Compton scattering is the dominant mode of interaction with tissues. The population of the electrons set in motion by the incident x-rays is built up only gradually, and typically it takes a centimeter or so for the built-up electron density to reach the equilibrium state. The depth and the shape of the “tissue sparing curve” of MV x-rays depend on the energy of the MV x-rays or gamma rays. It ranges from several millimeters for 6 MV linacs to about 15 mm for the 18 MV ones (FIG. 1). In that regard the curve in FIG. 1, made for 21 MeV x-rays, is highly atypical and is therefore exaggerated for our discussion. On the other hand, protons, being charged particles, start depositing their energy into tissue immediately as they enter it. Accordingly, there is no such shallow tissue-sparing effect as there is for MV x-rays and gamma rays. This limits the entrance dose from these particles in comparison to possible entrance doses of MV x-rays and gamma rays. Accordingly, there is a need for an effective radiation therapy using protons or other ions, particularly light ions, which advantageously confines the radiation dosage to the target, and that also offers shallow tissue and skin-sparing. Features of the disclosure will become apparent from the following detailed description considered in conjunction with the accompanying drawings. It is to be understood, however, that the drawings are designed as an illustration only and not as a definition of the limits of this disclosure. The present disclosure relates to a system and methods for providing an effective radiation dose, using proton or other light ions, to a confined target volume, with a shallow tissue-sparing effect that allows for a higher entrance dose and consequently a higher therapeutic dose at the target. The present disclosure is also directed to a method for delivering therapeutic light ion radiation to a target volume of a subject, wherein the target volume is located at a predetermined depth, the predetermined depth being measured from an irradiated portion of the surface of the skin of the subject. The method includes selecting a species of light ions for forming an array of minibeams directed at the target volume based on the predetermined depth. The method further includes selecting a predetermined energy of the selected species of light ions for confining the therapeutic radiation within the target volume such that the Bragg peak corresponding to the predetermined energy of the species is at a distal side of the target volume. The therapeutic radiation is delivered to the target volume by forming the array of minibeams, which are comprised of the species of light ions at the predetermined energy, and directing the array at the target volume. A portion of the surface of the skin is irradiated with the array of minibeams. The minibeams are arranged as parallel, spatially distinct minibeams at the surface of the skin in an amount and spatially arranged and sized to maintain a tissue-sparing effect from the surface of the skin to a proximal edge of the target volume and to merge into a solid beam at the proximal edge of the target volume. The species of light ions is selected such that the minibeams broaden and merge into the solid beam at the proximal edge of the target volume to deliver a therapeutic dose of radiation to at least a portion of the target volume. Forming the array further includes selecting a gap between adjacent minibeams in the array to maintain the solid beam at the predetermined energy of the selected light ions at the proximal edge of the target volume. In one aspect, the step of delivering the therapeutic dose further includes spreading the Bragg-peak of the light ions forming the minibeams by stepwise adjusting the predetermined energy of the light ions across a range of energies to produce a uniform dose distribution throughout the target volume. The step of selecting the gap includes selecting the gap for which the solid beam is maintained at the proximal edge for each of the energies across the range of energies. In some aspects, the light ions forming the minibeams are protons. In aspects, the array of light ion minibeams is a two-dimensional array of pencil minibeams. In additional aspects, the method includes shaping a cross-section of the two-dimensional array to substantially match a cross-sectional shape of the target volume. In yet additional aspects, the species of light ions for forming the minibeams are selected from the group consisting of protons, deuterons and ions of helium, lithium, beryllium, and boron. In other aspects, the species of light ions for forming the minibeams are selected from the group consisting of deuterons and ions of helium, lithium, beryllium, and boron. A cross-sectional profile of at least one of the light ion minibeams, in various aspects, has one of a circular, square, rectangular, elliptical, and polygonal shape. In yet additional aspects, the cross-sectional profile of each of the light ion minibeams has a substantially radially symmetrical shape. In aspects, the method further includes providing a light ion source and a collimator downstream of the light ion source for forming the array. In further aspects, the collimator is spaced apart from the surface of the skin for forming the array of light ion minibeams. In various aspects, a width of each of the light ion minibeams at the surface is between 0.1 mm and 0.6 mm. In some aspects, the width of each minibeam at the surface is about 0.3 mm. In some additional aspects, the width of each minibeam at the surface is between about 0.1 mm and 1.0 mm. In additional aspects, a gap between the minibeams at the surface is between about 0.1 mm and about 3.0 mm. In some aspects, the gap between the minibeams at the surface is between about 0.1 mm and about 1.0 mm. In another aspect, the light ions forming the minibeams have energies between 10 MeV per nucleon and 1000 MeV per nucleon. In still other aspects, the array of light ion minibeams is a one-dimensional array of planar minibeams. Some aspects of the method further include performing the additional steps of selecting a species of light ions, selecting a predetermined energy, and delivering the therapeutic radiation from a second direction, a second portion of the surface of the skin being irradiated from the second direction, the predetermined depth of the target volume being measured from the second portion of the skin. The step of delivering the therapeutic dose from the second direction further includes, in aspects, spreading the Bragg-peak of the selected light ions forming the minibeams on the second portion of the skin by stepwise lowering the predetermined energy across a range of energies to produce a uniform dose distribution throughout the target volume. The gap between adjacent minibeams of an array irradiating the second portion is selected so that a solid beam is maintained at a proximal edge relative to the second direction for each of the energies across the range of energies. The present disclosure is also directed to a method for delivering therapeutic light ion radiation to a target volume of a subject, wherein the target volume is located at a predetermined depth. The predetermined depth is measured from an irradiated portion of the skin of the subject. The method includes irradiating a portion of a surface of the skin with an array of light ion minibeams comprising parallel, spatially distinct minibeams at the surface in an amount and spatially arranged and sized to maintain a tissue-sparing effect from the surface of the skin to a proximal side of the target volume, and to merge into a solid beam at the proximal side of the target volume. A gap between adjacent parallel, spatially distinct minibeams at the surface and a species of light ions forming the minibeams are selected based on a depth of the target volume from the surface. In aspects, a species of light ions forming the light ion minibeams is selected from the group consisting of protons, deuterons, and ions of helium, lithium, beryllium, and boron. In additional aspects, the method includes spreading the Bragg-peak of a predetermined energy of the species of light ions forming the minibeams by stepwise adjusting the predetermined energy of the light ions across a range of energies to produce a uniform dose distribution throughout the target volume. The species of light ions and the gap are selected so that the minibeams broaden and merge into the solid beam at the proximal side for each of the energies across the range of energies. In additional aspects, the method can include raster-scanning an incident light ion radiation beam to form the array of light ion minibeams for irradiating the target volume. In addition to the above aspects of the present disclosure, additional aspects, objects, features and advantages will be apparent from the embodiments presented in the following description and in connection with the accompanying drawings. The following sections describe exemplary embodiments of the present invention. It should be apparent to those skilled in the art that the described embodiments of the present invention provided herein are illustrative only and not limiting, having been presented by way of example only. All features disclosed in this description may be replaced by alternative features serving the same or similar purpose, unless expressly stated otherwise. Therefore, numerous other embodiments of the modifications thereof are contemplated as filling within the scope of the present invention as defined herein and equivalents thereto. It is noted that protons, and proton beam therapy, are treated separately from other radiation therapies in the prior art. To date, radiation beam therapy for other light ions is not known in the prior art. The present disclosure is directed to radiation therapies using both protons and other light ions. For simplicity, the terms “light ions” and “species of light ions” as used in the present disclosure include protons as well as deuterons and ions of helium, lithium, beryllium, and boron. The present disclosure relates to a new solution to the problem of the lack of the sparing effect of skin and other shallow tissues in protons and other ions, particularly, light ions. Referring to FIG. 3, for example, in one embodiment, to deliver a therapeutic dose of radiation to a target volume which encompasses, for example, a tumor, incident, parallel beams 20 of ions of a predetermined species are formed, for example, by segmenting an incident radiation beam using a collimator 24 positioned on, or in front of, the surface 26 of the skin. The application of such an array spares the skin and the shallow tissues until they go above 0.7 mm size and/or merge with each other to produce a solid beam. In some embodiments, the incident radiation can be segmented into an array of nearly parallel, small pencil beams 20 on the surface of the skin. In some embodiments, the minibeams formed in accordance with the present disclosure have a width (or diameter, in the case of circular pencil minibeams) between about 0.1 mm and about 0.6 mm. The cross-section of the pencil beams can be, but is not limited to, a circular, or nearly circular shape. In other embodiments, one or more of the pencil beams can be square, rectangular, elliptical, or polygonal in shape. One of ordinary skill in the art will appreciate that numerous other cross-sectional shapes can also be used. Referring also to FIG. 4, in some embodiments, the incident radiation can be segmented into an array of parallel, or nearly parallel, narrow planar minibeams 25 on the surface of the skin. In some embodiments, the planar minibeams have a width or thickness 27 of between about 0.1 mm and about 0.6-mm-diameter. In particular embodiments, the minibeams have a width 27, or a diameter in the case of pencil beams, of about 0.3 mm. The minibeams formed in accordance with the present disclosure are spaced sufficiently on the surface of the skin to spare the tissue between the minibeams as well as to form a solid or continuous beam at the desired target depth. The spacing of the minibeams is described herein in terms of a gap between the edges of the minibeams. An on-center spacing between the minibeams may also be specified. One of ordinary skill in the art will appreciate that in such cases, the gap between the minibeams is determined by both the on-center spacing and the width or thickness (FWHM) of the minibeams. In some embodiments, a gap 29 between the minibeams formed in accordance with the present disclosure is between about 0.1 mm and about 3.0 mm. In other embodiments, a gap between the minibeams formed in accordance with the present disclosure is between about 0.1 mm and about 1.0 mm. In still other embodiments, a gap between the minibeams formed in accordance with the present disclosure is between about 0.5 mm and about 0.8 mm. In additional embodiments, a gap between the minibeams formed in accordance with the present disclosure is about 0.7 mm. The application of the arrays of minibeams of the present disclosure, which can be pencil beams or planar beams, spares the skin and the shallow tissues until they go above 0.7 mm size and/or merge with each other to produce a solid beam. The minibeams 20 shown in FIG. 3 can be pencil beams or planar beams. The size and spacing of the beams at the surface of the skin advantageously promote shallow tissue sparing from the skin to a proximal side of the target volume. As these individual beams penetrate the tissues they gradually broaden. For a known depth of the target volume from the surface, the gap between the minibeams and the particular species of ions forming the minibeams can be selected so that the minibeams merge with their neighbors to form a solid radiation field 28 at the target volume. It is noted that the target volume is usually not confined only to a tumor, for example, but can also include a certain amount of tissue surrounding the tumor. Referring to FIG. 4, in some embodiments, the array of parallel light ion minibeams of the present disclosure is a one-dimensional array of planar minibeams 25. In some embodiments, a multislit collimator, for example, a tungsten multislit collimator, is used to produce an array of minibeams. For example, a tungsten multislit collimator was used to produce an array of 100 MeV planar proton minibeams with 0.3 mm width and 0.7 mm gaps between the planar minibeams. FIG. 4 shows the resulting pattern captured on a chromographic film, which was positioned at the downstream end of the collimator. In other embodiments, the array of parallel light ion minibeams of the present disclosure is a one-dimensional array of pencil minibeams formed in accordance with the present disclosure. Referring to FIG. 3, in other embodiments, the array of parallel light ion minibeams of the present disclosure is a two-dimensional array of pencil minibeams formed in accordance with the present disclosure. In some embodiments, the depth at which the minibeams of the present disclosure merge is about 1 cm to about 3 cm. In embodiments, the methods of the present disclosure are implemented to deliver therapeutic doses of radiation to brain tumors, including pediatric brain tumors. The medical significance of the present method's sparing of the shallow tissues can be divided in the following categories. First, the effect spares the skin, which is a highly radiosensitive organ. This allows the use of higher incident particle doses than those possible today. As a result the dose given in each session (called dose fraction) can be increased and therefore the total number of dose fractions can be reduced, making the treatment easier on the patient. This reduction is called “hypofractionation.” Second, the method will spare the brain's frontal cortex. Sparing of the frontal cortex is vital in reducing late cognitive effects in children and also in adults because it is a major site of generation of the brain's actively dividing neural stem cell (NSCs) that turn into glia, particularly oligodendrocytes (SP Rodgers et al., Neural Plasticity 2013; 698528). Oligodendrocytes are the cells that produce myelin, the coating of the axons. The process, called gliogenesis, introduces plasticity in the brain, particularly the pediatric brain. The need for new oligodendrocytes and new myelin are the most important feature of the brain particular the growing brain. This process also involves angiogenesis in the cortex. The preponderance of dividing cells makes the process highly radiosensitive. As a result, radiation damage to the pediatric frontal cortex produces a long-term decrement in cell proliferation. This cell decrement, together with the direct radiation effect on the neural cells, produces a neural environment that is hostile to plasticity. Furthermore, long-term suppression of cell proliferation deprives the brain of the raw materials needed for optimum cognitive performance, e.g., new glia in frontal cortex, while chronic inflammation and dearth of trophic substances, such as the above, limit neuroplastic potential in existing circuitry. Finally, the grey-matter layer of the cerebral cortex, which is rich with neurons and their dendrites, is also radiosensitive and is therefore another organ whose sparing by the present methods will reduce the depth of the neurological deficits caused by radiation in children and adults. There is no known prior attempt to add skin-sparing effect to protons or light ions. One attempt to add skin-sparing effect to heavier ions, such as carbon ions, is described in U.S. Pat. No. 8,269,198 to Dilmanian et al. (the “'198 patent”), the entirety of which is incorporated herein by reference. In the '198 patent, some skin-sparing effect is achieved by interleaving carbon minibeams. The method has shown certain success in pre-clinical studies. However, it requires the tissues to be completely immobilized. Also, it requires much more expensive facilities than those of proton therapy. The heavy ion beams of the '198 patent, while exhibiting some broadening, are able to be sufficiently shaped and controlled to form a solid beam at a desired target using the interleaving methods of the '198 patent. On the contrary, light ion beams, including proton beams, broaden too excessively to be suitable for interleaving, except for very shallow and very small targets. In the present disclosure, the intersection of broadening light ion minibeams in an array forms a solid or continuous minibeam at a predetermined depth by carefully controlling the minibeam widths and the gap between arrays of minibeams. Accordingly, in contrast to prior methods known in the art, each array of minibeams formed in accordance with the present disclosure independently forms a solid or continuous beam of radiation at the target. No interleaving of multiple arrays is required to form the solid beam. In addition, because a solid beam of radiation can be delivered by each independent array, without interleaving arrays, the target volume can be irradiated with arrays of the minibeams formed in accordance with the present disclosure from any number of different directions, each aimed to strike the target volume. It should be understood that while the embodiments described herein are directed to proton and light ion minibeams, the methods are not limited thereto. Accordingly, in some embodiments, the methods may include irradiation of a shallow target volume using two-dimensional arrays of heavier ion parallel pencil beams formed in accordance with the present disclosure. In some embodiments, the incident radiation comprises one or more of protons, deuterons, and ions of helium, lithium, beryllium, and boron. In other embodiments, the incident radiation may comprise carbon ions. Referring again to the embodiment of FIG. 3, in one embodiment, the incident beam 22 comprises protons and is converted into an array of nearly parallel proton pencil beams of 0.1 to 0.6 mm incident diameter using a multi-hole heavy-metal collimator 24 positioned on the patient's skin 26. The gaps between the holes in the collimator, depending on the clinical need, can be as small as 0.1 mm. In other embodiments, the incident beam comprises one or more species of light ions. It is estimated that the tissue-sparing effect of arrays of proton minibeams formed in accordance with the present disclosure starts to diminish when they are about ⅔ of the way to merge with each other. This is because just before they merge the gaps between them start to diminish, producing some “valley” dose between the minibeams. It has also been estimated that the full tissue-sparing effect will be about 10-fold above the tolerance of the tissue to solid beams of protons. As the individual proton minibeams penetrate the tissue, they gradually broaden by multiple scattering off the electrons in their paths, and at some tissue depth, depending on the gap (or on-center spacing and beam width) between the beams, they merge with their neighbors to produce a solid proton beam. Also, as the individual minibeams broaden, they gradually lose their tissue sparing effect, with the nominal threshold for losing all their tissue-sparing effect having been found to be about 0.7 mm for the proton and light ion minibeams of the present disclosure. For this reason, the gap between the minibeam holes in the collimator is preferably adjusted so that the adjacent minibeams for a particular species of light ions merge with their neighbors either when they reach 0.7 mm in diameter, which would be a depth of about 30 mm from the skin, or before that, depending upon whether the target to be treated is farther from the skin than that or closer. In this way the target volume, which includes the tumor, receives a solid beam of protons, while the skin and the shallower tissues receive tissue-sparing beams. In the system and methods of the present disclosure, a two-dimensional array of parallel minibeams of protons formed in accordance with the present disclosure, for example, can be arranged in any pattern to form a cross-sectional shape and size to match the cross-sectional shape and size of, for example, a tumor. As shown in FIG. 3, for example, the array of minibeams form a circular cross-section at the target volume. Accordingly, the present method is not limited to treatment of very small objects. In one embodiment, at the surface of the skin where the array is formed, each minibeam has a width of ˜0.3 mm. This is considerably smaller than the 0.7 mm beam diameter where the method's tissue-sparing effect starts to diminish. In some embodiments, the width of each minibeam is between about 0.15 and 0.25 mm. In other embodiments, the depth of the tissue in which the minibeams for a particular species of light ions merge is selected in accordance with the depth of the proximal side of a target volume that encompasses the tumor by adjusting the gap between the minibeams. Because of the approximate 0.7 mm diameter of the minibeams at which the tissue tolerance starts to diminish for a given proton (or other species of ion forming the minibeams) beam energy, the tissue depth for beam merging is preferably selected for depths at which the minibeams grow beyond 0.7-mm in diameter. As illustrated in FIG. 6, an embodiment of the method of the present disclosure may include selecting a species of light ions, at 45, such that minibeams of a particular width (0.3 mm, for example) will merge at the proximal side of a target volume that encompasses the tumor. In embodiments, the gap between the minibeams is also set such that the beams merge at the same depth at which the minibeams broaden to about 0.7 mm width. As illustrated in FIG. 10, the ion species are characterized by different rates of broadening, so that the optimum selection is based on the depth of the proximal target location. In particular, the species of light ions is selected that provides an adequate rate of minibeam broadening so that minibeams, with proper beam spacing, merge by the time they reach the proximal side of the target, yet still provide tissue sparing along the path of the array from the skin surface to the proximal side of the target. In embodiments of a method of the present disclosure, an energy of the light ions in the incident minibeams (“beam energy”) irradiating the target volume is also selected based on a known depth within the target volume at which the light ions forming the minibeams will stop traveling. This depth can be calculated as the position of the well-known Bragg-peak, the depth at which the ions lose all their energy and at which the highest radiation dose is delivered. In the embodiment of FIG. 6, the method includes selecting a beam energy, at 47, for producing the array such that the Bragg peak occurs at a distal side of the target volume to confine the therapeutic dose of radiation to the target volume. For the predetermined energy, a gap between the adjacent minibeams is also selected, at 48, such that the solid beam is maintained at the proximal side of the target. The gap between minibeams, the beam energy of the light ions, the species of the light ions, and the width of the minibeams, may each be adjusted within the various parameters described herein, to select the optimum parameters for forming an array of minibeams on the surface of the skin, at 49, which is directed at the target volume, to deliver a therapeutic solid (continuous) beam of radiation, at 51, substantially only to the target volume, and not to the surrounding tissue. Embodiments of the method also include spreading the Bragg-peak of the light ions, at 48, by any means known in the art, to deliver a uniform dose of therapeutic radiation across the target volume. The maximum beam energy selected corresponds to the Bragg peak at the distal side of the target. By stepwise lowering the energy from the maximum energy, so that successive Bragg peaks occur from the distal to the proximal side of the target volume, a uniform dose distribution is provided throughout the depth of the target. Any known method of producing this Bragg-peak spreading can be used to produce a uniform dose distribution along the known depth of the target. Accordingly, the method may further include, at 48, selecting the optimal gap between the minibeams for the species of light ions selected, such that the minibeams forming the array, at 49, merge into a solid beam at the proximal side of the target volume for every beam energy in the range of beam energies used for Bragg-peak spreading. In some instances, a patient may be treated with radiation administered during the same session, or during different sessions, where each successive radiation treatment may be delivered to the tumor from a different direction. In embodiments of the present method, for each irradiation direction used to treat a patient, the target volume is preferably irradiated across the range of predetermined energies, as described supra, to produce a uniform dose by Bragg-peak spreading. FIG. 5A shows a schematic view of a brain tumor 34 being irradiated by a method formed in accordance with the present disclosure, including irradiating the brain tumor from three directions with minibeam arrays 36, with a collimator 38 appropriately positioned on the surface of the skin 40 for each irradiation. Each irradiation is performed across different beam energies selected in accordance with the present disclosure to confine the radiation dose within the target volume and to deliver a uniform dose across the target, using the Bragg-spreading effect. The distance of the target volume from the irradiated surface of the skin differs for each of the radiation angles shown. In accordance with the method of the present disclosure, for each of the different radiation angles, the selected range of beam energies for producing a uniform dose, and the corresponding optimal gap and species of light ions are adjusted in accordance with the distance of the proximal and distal side of the target volume from the irradiated surface of the skin. FIG. 5B shows the magnified view 42 of one of the arrays of minibeams providing one of the three irradiation exposures of FIG. 5A. The target volume can be irradiated with arrays of the minibeams formed in accordance with the present disclosure from any number of different directions, each aimed to strike the target volume. It should be noted that, for each direction, the ion species and/or the gap between the minibeams should be adjusted so that the minibeams merge into a solid or nearly solid beam near the proximal side of a therapy target. The tissue depth of the proximal side of the target from the surface of the skin, and thus the depth at which the minibeams should merge, will, of course, most likely differ for each irradiation angle. As shown in FIGS. 5A and 5B, the minibeams, which were about 0.3 mm in beam width (FWHM) at the surface of the skin, preferably merge together at the proximal side 44 of the target volume encompassing a tumor, for example (before or at the proximal side of a tumor), and at a tissue depth 46 where the minibeams become 0.7 mm or larger in diameter, which is at ˜2.5 cm. The merging should not be deeper than 2-cm because this would mean that the minibeam spacing on-center will be larger than 0.7 mm, which will reduce the rate of delivered dose to the target. For 0.7-mm on center spacing, the ratio between the dose rate reaching the target from a solid proton beam and that from minibeams for the same incident beam intensity is 7:1, which is calculated by dividing the area of a square with 0.7 mm side to that of a circle with 0.3 mm diameter. For example, for 0.9 mm on-center spacing that ratio becomes 11.5:1, which means much beam is wasted. In fact the on-center spacing is preferably smaller than 0.7 mm for a target closer than ˜2.5 cm to the surface. Finally, the subject should be completely immobilized through the irradiation times. In this regard, it is noted that the brain's cardio-synchronous brain pulsation, which can be 0.3 mm or larger in adults, is expected to be smaller in children. In an embodiment of a system of the present disclosure, a quadrupole magnetic lens is used to focus raster-scanned beams of light ions produced by a source, such as a synchrotron source, to a very narrow width only at the patient surface and simultaneously introduce angular spread in the beams so that they broaden as they approach the target volume, e.g., the cancer target depth. In another embodiment of a system formed in accordance with the present disclosure, a scattering foil is placed on the downstream surface of a pinhole collimator to introduce angular confusion in the individual minibeams, so that the beams of light ions have very narrow width at the patient surface and broaden to converge into a broad beam, as they approach the target volume encompassing, for example, a cancerous tumor. In another embodiment of a system and method of the present disclosure, a gap is introduced between the collimator producing the array of parallel minibeams and the skin. The frame used for this purpose can push on the skin instead of the collimator for patient immobilization. The introduction of a 5-cm gap between a collimator producing an array of planar proton minibeams of 0.3 mm FWHM thickness, with 1.0 mm spacing on-center, was shown to reduce the skin dose from neutrons produced in the collimator by 7 fold compared to the skin dose that occurs when the collimator is resting on the skin. In embodiments, the collimator producing the array of parallel minibeams of the present disclosure is spaced from the skin by about 5 cm, or between about 2.5 cm and about 5 cm from the skin. In other embodiments, the collimator is spaced from the skin by between about 2.0 cm and 6.0 cm of the skin. Measuring the Divergence of Pencil Minibeams Formed in Accordance with the Present Disclosure Produced with a 150-MeV Proton Minibeam: A 0.3 mm collimator 50 was positioned in the way of a 150-MeV proton beam. Next, a stack of chromographic films 52 was positioned in front of the beam with 2-mm plastic sheets between the adjacent films. The experimental set-up is depicted in FIG. 7, while the results are presented in FIG. 8. Quantitative analysis of the results of FIG. 8 shows that the 0.3 mm proton minibeams broaden to ˜1.3 mm FWHM at 50 mm plastic depth. This is close (namely 1.24 mm) to what was observed with Monte Carlo simulations using Monte Carlo N-Particle extended (MCNPX) code. Measuring and Simulating the Beam Divergence of Pencil Minibeams of 109 MeV Protons: The above measurement of FIGS. 7 and 8 were repeated using 109 MeV protons, which is closer to the beam energy used for treating brain tumors. FIG. 9 compares the experimental results (circles) 54 with those from MCNPX simulations in water (x marks) 56. The match is very good. The results show that the water depth where the beam reaches 0.7 mm FWHM is about 2.5 cm. Simulating the Beam Divergence of the Pencil Minibeams for Several Light Ions Compared to Protons: The code MCNPX was also used to calculate the divergence of 0.3-mm pencil minibeams made of protons 58 and several other light ions formed in accordance with the present disclosure, namely H-2 60, H-3 62, He-3 64, He-4 66, Li-6 68, and Li-7 70 in water (FIG. 10). The results clearly show the trend of smaller beam divergence for particles of larger mass and larger atomic number. This is a most significant effect because it indicates that the heavier particles reach the 0.7 mm diameter (FWHM) at larger tissue depth than the approximate 26 mm depth measured for protons (see FIG. 10, plot 58). As a result, in some embodiments formed in accordance with the present disclosure arrays of parallel minibeams made of these light ions can be used to spare a larger depth of the shallow tissues. According to FIG. 10, for minibeams of protons and the light ions listed above, these tissue depths are approximately 26, 34, 39, 41, 43, 49, and 51 mm, respectively. For arrays of planar minibeams of 0.3 mm FWHM, the protons and light ions were found to reach the 0.7 mm FWHM at depths of 1 to 2 mm longer. The beam energies used for these particles, the beam energies being chosen for producing Bragg peaks at approximately 10 cm water depth, were 109, 157, 188, 410, 462, 873, and 931 MeV, respectively. Ten million particle histories were tracked for each ion species. Simulating the Beam Divergence for Entire Arrays of Proton and Li-7 Pencil Minibeams: Referring to FIGS. 11A and 111B, Monte Carlo simulations of absorbed dose from 2-dimensional pencil minibeam arrays composed of 116-MeV proton beams 72 and 931-MeV lithium-7 beams 74, producing Bragg peaks at 10 cm water depth, were performed. The arrays consisted of 21×21 minibeams, spaced on a 1-mm grid (1.0 mm on-center spacing), each having an initial circular cross section with 0.3-mm diameter (FWHM). No Bragg-peak spreading was carried out. These data support the above concept that heavier light ions allow sparing of deeper proximal tissues to the target than protons because of their smaller divergence. This makes the technique of selecting the right light-ion-species for each individual patient treatment, in addition to selecting the beam geometry including the thickness and separation gap, a most powerful aspect of this method. Dosimetry Results for Various Beam Geometries and Ion Species: As described herein, the ion species and beam geometries (spacing and thickness of planar or pencil beams) are optimized for a particular beam energy to enhance tissue sparing while delivering appropriate dose to a tumor in accordance with the present disclosure. The results of a MCNPX simulation for three (3) different target geometries using 0.3-mm planar beams are shown in Tables 1 and 2 below. The ion species was chosen on the basis of the depth of the target's proximal edge, although the larger RBE of heavier ions could be also a consideration for treating radioresistant tumors. Table 2 shows the minibeam merging depth in water for planar minibeams of 10-cm range; for pencil minibeams, the depths are 1 to 2 mm longer. TABLE 1Dosimetric ConsiderationsFWHMTarget'sTarget'sof beamRecommendedProximalDistalBeamat target'sbeam spacingdepthdepthRecommendedEnergyproximalon-centerCollimator's(cm)(cm)Ion Species(MeV)edge(mm)(mm)transmission48H1021.60.743%412H1291.20.743%He-45120.60.650%Li-71,0320.50.560%816Li-71,2141.10.743% TABLE 2Depth of merging in water for planar beams of 10-cm rangeBeam spacingIncident beamMerging depth for H-1, He-4, andon-center (mm)thickness (mm)Li-7, respectively (mm)0.50.321, 35, 410.70.325, 41, 491.20.337, 60, 701.20.537, 59, 69Comparison Between X-Ray Grid Therapy and Proton Minibeam Therapy As indicated above in the discussion of the mechanistic bases for the tissue-sparing effect of minibeams and minibeams, the skin-sparing effect of Grid Therapy was based on the dose-volume effect. The method, used with orthovoltage machines, involved positioning of a metal grid with 2-20 mm sized holes on the patient's chest to avoid severe skin damage during radiation therapy. However, for two reasons the tissue-sparing effect of proton minibeams formed in accordance with the present disclosure is much more substantial than that of x-ray Grid Therapy. First, it uses much smaller minibeams, namely 0.3 mm instead of 2-20 mm. Second, the proton minibeams stay minibeams for several centimeters depth in the body and spare all that tissue. This is different from the grid therapy segmented x-ray beams that not only have only a mild skin sparing effect caused by the dose-volume effect, but also quickly broaden further with tissue depth because of the large penumbra effect of the x-ray source, and therefore has no sizable sparing effect in tissue depth. Again, as indicated in the discussion of the mechanisms of the minibeams' tissue-spring effect the effect caused not only by the dose-volume effect but also by the “prompt microscopic biological repair” of very small beams, and this is why it is so robust. On the other hand, Grid therapy spares only the skin and that is only because of the dose-volume effect. Finally, as indicated above the robust tissue-sparing of the proton minibeams for the depth of ˜2.5 cm should lead to the sparing of the frontal cortex as well as that of the cortex's gray matter layer; such sparing of the central nervous system will have major ramifications in terms of the brain's function and the patient's cognitive and neurological well-being. Some of the clinical advantages of the methods formed in accordance with the present disclosure include the following. Sparing the skin. Sparing the cortex, including the frontal cortex. Reducing the number of dose fractions from 30 in the conventional proton therapy to below five or six. This reduction in the number of dose fractions, called hypo-fractionation, occurs because the patient can be administered with a higher dose in each treatment session because proton minibeams increase the tolerance of the skin and shallow tissues to the proton dose. Treating larger and/or deeper tumors in the chest and abdomen that respond better to hypofractionation. Because large elapsed times between radiation sessions can help larger tumors to recover to some extent, using larger dose fractions in large tumors is always beneficial. Because treatment of large and/or deep tumors require higher entrance doses, the method can be beneficial because it allows the skin and shallow normal tissues to better tolerate the larger incident doses required to treat these tumors. Treating tumors of the head and neck. These tumors are difficult to treat not only because they are often radioresistant, requiring large doses to be controlled, but they are often residing near radiosensitive organs such as the parotid glands. The proposed method will be beneficial for both these effects. First, it allows the delivery of much higher dose to the tumor in each session. Second, in treating tumors residing behind shallow, radiosensitive organs, which are not thicker than ˜2.5 cm, such as the parotid glands, proton minibeams can go through that organ without damaging it. While the invention has been particularly shown and described with reference to specific embodiments, it should be apparent to those skilled in the art that the foregoing is illustrative only and not limiting, having been presented by way of example only. Various changes in form and detail may be made therein without departing from the spirit and scope of the invention. Therefore, numerous other embodiments are contemplated as falling within the scope of the present invention as defined by the accompanying claims and equivalents thereto. |
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053533208 | abstract | A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. |
abstract | A radiation imaging apparatus includes a radiation detection panel configured to detect radiation irradiated by a radiation generation unit, a first member and second member arranged on the incident direction side of the radiation, and a third member and fourth member arranged on a side opposite to the incident direction of the radiation. The second member is arranged between the first member and the radiation detection panel, and the third member is arranged between the radiation detection panel and the fourth member. The second member and the third member are lower in elastic modulus than the first member and the fourth member, and the elastic modulus of the second member is equal to or lower than that of the third member. |
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abstract | A position measurement system for measuring a position of an object is described, the system including: a first incremental measurement unit for measuring a first number of first distance steps in a distance between a reference frame and the object, wherein the first number equals a first integer value plus a first fraction, and a second incremental measurement unit for measuring a second number of second distance steps in a distance between the reference frame and the object, wherein the second number equals a second integer value plus a second fraction, wherein the position measurement system is constructed and arranged to initialize the second incremental measurement unit on the basis of the first number and the second fraction. |
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claims | 1. A passive auxiliary condensing apparatus of a nuclear power plant, the passive auxiliary condensing apparatus comprising:a steam generation unit generating steam by a heat produced when operating a nuclear reactor;a cooling tank disposed outside the steam generation unit and storing cooling water therein;a heat exchanger installed inside the cooling tank and condensing the steam provided from the steam generation unit by the cooling water of the cooling tank, the heat exchanger having an inlet and an outlet;a steam pipe connecting an upper portion of the steam generation unit and the inlet of the heat exchanger such that the steam from the steam generation unit flows into the heat exchanger through the steam pipe;a steam-water separation tank disposed side by side with the cooling tank separately from each other, wherein the steam-water separation tank separates a mixture of water and steam provided from the heat exchanger into the water and the steam respectively and includesan inlet hole formed in a side wall of the steam-water separation tank,an outlet hole formed in a bottom wall of the steam-water separation tank,an inlet pipe having a first end and a second end, wherein the first end of the inlet pipe is connected to the inlet hole of the steam-water separation tank and the second end of the inlet pipe is connected to the outlet of the heat exchanger such that the mixture from the heat exchanger flows into the steam-water separation tank through the inlet pipe, andan outlet pipe having a first end and a second end, wherein the first end of the outlet pipe is connected to the outlet hole of the steam-water separation tank and the second end of the outlet pipe is connected to a lower portion of the steam generation unit such that the water from the steam-water separation tank flows into the steam generation unit; anda bypass pipe including a check valve and connecting the inlet of the heat exchanger and an upper portion of the steam-water separation tank such that the steam flows from the steam-water separation tank to the heat exchanger without passing through the steam generation unit,wherein the steam provided from the steam generation unit circulates through the steam generation unit, the steam pipe, the heat exchanger, the inlet pipe of the steam-water separation tank, the steam-water separation tank, the outlet pipe of the steam-water separation tank and the steam generation unit. 2. The passive auxiliary condensing apparatus according to claim 1, wherein the steam-water separation tank has a cylindrical shape elongated in a longitudinal direction. 3. The passive auxiliary condensing apparatus according to claim 1, wherein the bypass pipe further includes a control valve. |
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059011920 | abstract | Core spray line riser repair apparatus and methods for axially restraining a reactor core spray line riser to a T-box are described. In one embodiment, the repair apparatus includes a sleeve, a threaded draw bolt extending through the axial bore, and a threaded block. The sleeve includes a substantially cylindrical main body having an axial bore extending therethrough and a flange at one end of the main body. The sleeve is configured to be inserted into the T-box until the flange abuts against the T-box. The draw bolt is configured to extend through the sleeve flange, the sleeve bore, and an opening in the core spray line riser. The draw bolt engages the block to draw the block in tight engagement with the core spray line riser. |
047599022 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Analysis of the radiation build-up data base led to the surprising discovery that the long-term (equilibrium) dose rate is dictated by the early build-up rate, which is determined by current plant conditions and is related to the growth and morphology of the oxide film. Therefore, a physical parameter that is sensitive to the oxide film growth and morphology will provide early indication of what the long-term radiation build-up and dose levels are likely to be. I have discovered that the electrochemical potential of an unprefilmed corroding metal surface is a strong function of the thickness and character of the corrosion film. Therefore, I conclude that the rate of increase in the electrochemical potential, measured with an unprefilmed measuring electrode over a short-term period, should provide a measurement of the morphology of the oxide films and, hence, of radiation build-up rates and the eventual long-term radiation levels from the cooling system of a nuclear power plant. If the curves in FIGS. 8 and 9 are compared, it is apparent that the Hatch-2 data show the most rapid electrochemical potential increase and the earliest leveling off of the electrochemical potential; and the Hatch-2 plant has the lowest radiation build-up rate and the lowest long-term (equilibrium) radiation levels. Conversely, the Vermont Yankee data show the slowest electrochemical potential increase and latest leveling off of the electrochemical potential, and Vermont Yankee plant has had the most rapid radiation build-up rates and the highest long-term radiation levels. The normal chemistry data (no hydrogen addition) from Dresden-2 lie between the other two plants in regard to both electrochemical potential change and the long-term radiation levels. However, the radiation levels at Dresden-2 have begun to increase since hydrogen has been routinely added (initiated at about 7.6 effective full-power years), and the shape of the hydrogen water chemistry electrochemical potential curve from Dresden-2 seems to resemble the Vermont Yankee electrochemical potential curve. The discovery of these qualitative relationships prompted me to develop a more quantitative method for predicting long-term radiation dose rates from short-term electrochemical measurements. I discovered that if the electrochemical potential data are normalized by dividing by the measured or interpolated value after a short exposure time and if these normalized electrochemical potential factions are plotted versus the logarithm of time (in hours), an approximately linear relationship is produced for each of the three data sets (see FIG. 10). Any short-time value can be used, as long as the same time is used for all data sets. I further discovered that when I plotted the concentration of Co-60 in the cooling water divided by the negative slopes of these lines versus the long-term dose rates for these plants, I had prepared a standard curve (FIG. 11) that can be used to predict long-term dose rates from any properly determined set of electrochemical measurements that have been normalized to the same short-time value (a two-hour value was used in FIGS. 10 and 11). This standard curve is shown in FIG. 11 with the three points for the three plants where the necessary electrochemical potential data, long-term dose rate data, and Co-60 concentrations are available in the published literature. The process of data treatment that is a part of this invention can be illustrated by using the short-term electrochemical potential data for hydrogen injection at Dresden-2. These data have been normalized to the two-hour value and plotted versus the logarithm of time in FIG. 12. The slope of the line is -0.325. The average Co-60 concentration at Dresden-2 near the time the electrochemical potential measurements were made was about 0.15 uCi/liter (8). Using the standard curve of FIG. 11, the quotient of these two numbers (0.46 uCi/1) predicts a long-term dose rate for Dresden-2 with hydrogen injection of about 320 mR/h as shown in FIG. 13. Unfortunately, after only 1.7 effective full-power years of operation with hydrogen injection, the Dresden-2 recirculation lines were decontaminated chemically so the long-term result will never be known. However, at that time of the decontamination, the Dresden-2 dose rate had risen to about 300 mR/h from about 230 mR/h just before hydrogen injection was started. Since the build-up rate is logarithmic with time, it is likely the long-term (equilibrium) dose rate would have been about 320 mR/h. REFRENCES TO PRIOR ART 1. C. J. Wood, "Recent Developments in LWR Radiation Field Control," Progress in Nuclear Energy, Electric Power Research Institute, June 1985. PA1 2. R. A. Shaw and M. D. Naughton, "Radiation Control in Light Water Reactors," Proceedings of an International Conference on Water Chemistry of Nuclear Reactor System 2, Oct. 1980 (page 32). PA1 3. L. D. Anstine and M. Naughton, "Radiation Level Assessment and Control for Boiling Water Reactors," ibid (paper 50). PA1 4. W. E. Berry and R. B. Diegle, "Survey of Corrosion Product Generation, Transport, and Deposition in Light Water Reactors," Electric Power Research Institute, Mar. 1979 (EPRI NP-522). PA1 5. L. D. Anstine, "BWR Radiation Assessment and Control Program: Assessment and Control of BWR Radiation Fields," Electric Power Research Institute, May 1983 (EPRI NP-3114, vol. 2). PA1 6. L. D. Anstine, J. J. Zimmer and T. L. Wong, "BWR Corrosion-Product Transport Survey," Electric Power Research Institute, Sept. 1984 (EPRI NP-3681). PA1 7. W. Marble, "Control of Radiation-Field Buildup in BWRs," Electric Power Research Institute, June 1985 (NP-4072). PA1 8. "BWR Radiation Control Handouts from EPRI Contractors Meeting," Electric Power Research Institute, Nov. 1985. PA1 9. R. S. Greeley, M. H. Lietzke, W. T. Smith, and R. W. Stoughton, "Electromotive Force Studies in Aqueous Solutions at Elevated Temperatures. I. The Standard Potential of the Silver-Silver Chloride Electrode," Journal of Physical Chemistry, vol. 64, p. 652, 1980. PA1 10. M. E. Indig and J. E. Weber, "Electrochemical Potential Measurements in a Boiling Water Reactor," Electric Power Research Institute, Nov. 1983 (EPRI NP-3362). PA1 11. J. Leibovitz, W. R. Kassen, W. L. Pearl and S. G. Sawochka, Draft-BWR "In-Plant Measurements of Electrochemical Potentials," Electric Power Research Institute, May 1983 (EPRI NP-3521. PA1 12. E. L. Burley, "Oxygen Suppression in Boiling Water Reactors--Phase 2," General Electric Company, Oct. 1982 (NEDC-23856-7). As can be seen from the preceding descriptions and discussion, the invention provides an effective method for predicting the long-term radiation level changes that will be associated with the corrosion and films on the interior surfaces of the coolant system piping of a nuclear power plant. Having thus described the invention, what is believed to be new and novel and sought to be protected by letters patent of the United States is as follows: |
summary | ||
062263412 | claims | 1. A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, a shield surrounding the active portion of the reactor, means to control a neutronic chain reaction within the reactor, comprising a safety device and a regulating device, the safety device including means defining a vertical channel extending into the reactor from an aperture in the shield, a rod containing neutron-absorbing material slidably disposed within the channel, and means to introduce bodies of neutron-absorbing materials into the channel comprising a hopper having a hollow sleeve disposed within the channel and extending through the aperture in the shield, a spiral fin disposed about the sleeve in the portion of the channel extending through the shield, said hopper having an inner and outer wall defining a compartment adapted to contain bodies of neutron-absorbing materials, a floor disposed between said inner and outer walls and a door disposed within the floor opening into the portion of the channel containing the spiraled fins. 2. A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, the safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted on a level above the aperture in the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper. 3. A neutronic reactor comprising the elements of claim 2 wherein the regulating device comprises a rod containing neutron-absorbing materials translatably disposed at least partially within the active portion of the reactor. 4. A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, the safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including an electro-magnetic clutch for releasing said rod on deactuation thereof, a hopper mounted on a level above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on deactuation of the clutch to enter the active portion of the reactor for opening the door in the hopper. 5. A neutronic reactor comprising the elements of claim 2 wherein the means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper comprise a section of magnetized magnetic material disposed within a safety rod at a point within the shield of the reactor when the rod is fully inserted into the reactor, a coil of wire disposed about the channel adjacent to the shield of the reactor above the aperture therein, electrical means for releasing the safety rod and opening the door in the hopper including means to interject a time delay between actuation and opening of the door in the hopper, and means responsive to the generation of a pulse in the coil resulting from the release of the safety rod for preventing the opening of the door in the hopper. 6. A neutronic reactor comprising the elements of claim 4 wherein said hopper is embedded within the shield of the reactor. 7. A neutronic reactor comprising the elements of claim 4 wherein the hopper is mounted above the shield of the reactor. 8. A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper. |
claims | 1. A method for managing a plurality of instrumented function components, the method comprising:receiving, from an online safety availability application, operating information about the plurality of instrumented function components;updating, within an asset management database, status information for the plurality of instrumented function components based upon the operating information;determining, by a computer, a probability of failure on demand, at time of receiving the operating information, for the instrumented-function based upon the operating information;comparing, by a computer, the probability of failure on demand with a designed probability of failure on demand for the instrumented function to establish a variance; andmanaging the plurality of instrumented function components based on the status information and at least in part on the variance. 2. The method of claim 1 including:sending the status information to the online safety availability application; andcalculating, via the online safety availability application, a probability of failure on demand for a safety instrumented function, and wherein the safety instrumented function comprises at least a portion of the plurality of instrumented function components. 3. The method of claim 2 wherein the sending includes sending an indication that at least one of the plurality of instrumented function components has been replaced. 4. The method of claim 2 wherein the probability of failure on demand is calculated as an average probability of failure on demand. 5. The method of claim 2 wherein the probability of failure on demand is calculated as an instantaneous probability of failure on demand. 6. The method of claim 1 wherein the operating information about the plurality of instrumented function components includes an indication that at least one of the plurality of instrumented function components has failed. 7. The method of claim 6 wherein the managing includes initiating procurement of at least one new instrumented function component to replace the at least one of the plurality of instrumented function components that has failed. 8. The method of claim 1 wherein the receiving includes receiving test information, the test information indicating whether a test performed on an instrumented function component was successful. 9. The method of claim 1 further comprising receiving one or more environmental inputs associated with the plurality of instrumented function components, wherein said updating status information includes updating environmental information associated with said plurality of instrumented function components based at least in part on said environmental inputs. 10. The method of claim 9 wherein the environmental information includes temperature information. 11. The method of claim 9 wherein the environmental information includes pressure information. 12. The method of claim 9 wherein the environmental information includes humidity information. 13. The method of claim 1 including:sending the status information to the online safety availability application. 14. The method of claim 1 wherein the probability of failure on demand is calculated as an average failure of probability on demand. 15. The method of claim 1 wherein the probability of failure on demand is calculated as an instantaneous failure of probability on demand. 16. The method of claim 1 wherein the probability of failure on demand is calculated in part based on an environmental input associated with the one or more instrumented-function components. 17. The method of claim 1 wherein the managing comprises providing an alarm responsive to the variance. 18. The method of claim 1 wherein the managing comprises providing historical, on-line and/or predictive reporting of failure of probability on demand information for the instrumented function. 19. The method of claim 1 wherein the managing comprises electronically initiating the procurement of one or more replacement components for one or more of the instrumented-function components responsive to the variance. 20. A system for managing a plurality of instrumented function components, the system comprising:means for receiving, from an online safety availability application, operating information about the plurality of instrumented function components;means for updating, within an asset management database, status information for the plurality of instrumented function components based upon the received operating information;means for determining a probability of failure on demand, at time of receiving the operating information, for the instrumented-function based upon the operating information;means for comparing the probability of failure on demand with a designed probability of failure on demand for the instrumented function to establish a variance; andmeans for managing the plurality of instrumented function components based on the updated status information and at least part upon the variance. 21. The system of claim 20 including means for sending status information to the online safety availability application wherein the online safety availability application is configured to calculate a probability of failure on demand for a safety instrumented function, and wherein the safety instrumented function comprises at least a portion of the plurality of instrumented function components. 22. The system of claim 21 wherein the means for managing includes means for initiating procurement of at least one new instrumented function component to replace the at least one of the plurality of instrumented function components that has failed. 23. The system of claim 21 wherein the means for sending includes means for sending an indication that at least one of the plurality of instrumented function components has been replaced. 24. The system of claim 20 wherein the operating information about the plurality of instrumented function components includes an indication that at least one of the plurality of instrumented function components has failed. 25. The system of claim 20 wherein the means for receiving includes means for receiving test information, the test information indicating whether a test performed on an instrumented function component was successful. 26. A non-transitory processor-readable medium encoded with instructions for managing a plurality of instrumented function components, the medium including instructions for:receiving, from an online safety availability application, operating information about the plurality of instrumented function components;updating, within an asset management database, status information for the plurality of instrumented function components based upon the operating information;determining a probability of failure on demand, at time of receiving the operating information, for the instrumented-function based upon the operating information;comparing the probability of failure on demand with a designed probability of failure on demand for the instrumented function to establish a variance; andmanaging the plurality of instrumented function components based on the status information and at least in part upon the variance. 27. The non-transitory processor-readable medium of claim 26 including:sending the status information to the online safety availability application; andcalculating, via the online safety availability application, a probability of failure on demand for a safety instrumented function, and wherein the safety instrumented function comprises at least a portion of the plurality of instrumented function components. 28. The non-transitory processor-readable medium of claim 27 wherein the sending includes sending an indication that at least one of the plurality of instrumented function components has been replaced. 29. The non-transitory processor-readable medium of claim 26 wherein the operating information about the plurality of instrumented function components includes an indication that at least one of the plurality of instrumented function components has failed. 30. The non-transitory processor-readable medium of claim 29 wherein the managing includes initiating procurement of at least one new instrumented function component to replace the at least one of the plurality of instrumented function components that has failed. 31. The non-transitory processor-readable medium of claim 26 wherein the receiving includes receiving test information, the test information indicating whether a test performed on an instrumented function component was successful. 32. A computer-assisted method for managing a plurality of instrumented-function components which collectively implement an instrumented function, the method comprising:receiving, from an online safety availability application, operating information about the plurality of instrumented-function components;determining, by a computer, a probability of failure on demand, at time of receiving the operating information, for the instrumented-function based on the operating information;comparing, by a computer, the probability of failure on demand with a designed probability of failure on demand for the instrumented function to establish a variance; andmanaging the plurality of instrumented function components based at least in part upon the variance. 33. The method of claim 32 wherein the failure of probability on demand is calculated as an average probability of failure on demand. 34. The method of claim 32 wherein the failure of probability on demand is calculated as an instantaneous probability of failure on demand. 35. The method of claim 32 wherein the probability of failure on demand is calculated in part based on an environmental input associated with the one or more instrumented-function components. 36. The method of claim 32 wherein the managing comprises providing an alarm responsive to the variance. 37. The method of claim 32 wherein the managing comprises providing historical, on-line and/or predictive reporting of probability of failure on demand information for the instrumented function. 38. The method of claim 32 wherein the managing comprises electronically initiating the procurement of one or more replacement components for one or more of the instrumented-function components responsive to the variance. 39. A system for managing a plurality of instrumented function components, comprising:a processor;an input/output (I/O) module; anda memory containing instructions for execution by the processor to:receive operating information about the plurality of instrumented function components from the I/O module;provide, to an asset management database, status information for the plurality of instrumented function components based upon the operating information;determine a probability of failure on demand, at time of receiving the operating information, for the instrumented-function based upon the operating information;compare the probability of failure on demand with a designed probability of failure on demand for the instrumented function to establish a variance; andmanage the plurality of instrumented function components based on the status information and at least in part on the variance. |
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summary | ||
description | Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a reactor pressure vessel 1 situated beneath a surface 2 of cooling water in a boiling water reactor, while a handling machine 4 can be displaced over the surface 2 on a bridge or ramp 3. The handling machine 4 bears a mast 5 in the form of a telescopic arm, and on end of the mast 5 a hood 6 is disposed. In the reactor pressure vessel 1, fuel assemblies 7, 70 of the reactor are positioned on a lower fuel assembly grid 8. The hood 6, in a working position of the mast 5, being fitted over top fittings of the fuel assemblies 7 belonging to one division X, while the other fuel assemblies 70, belonging to a second division and other divisions, are divided. In accordance with FIG. 2, the hood 6 is divided by side walls 9 into individual cells 12, the side walls 9 resting on bars 10 of an upper core grid. Each mesh of the core grid contains a group of in each case four fuel assemblies, the hood 6 being divided by the side walls 9 into four cells, under each of which there is a group of fuel assemblies belonging to the division X. In FIG. 2, in each case only the first fuel assemblies A, Axe2x80x2 and the second fuel assemblies B, Bxe2x80x2 of the division X, which belong to a first group I and a second group II, are shown. The other two fuel assemblies C, D of the first group I and the fuel assemblies Cxe2x80x2, Dxe2x80x2 of the group II, as well as the remaining groups III and IV belonging to the division X, cannot be seen in FIG. 2. The hood 6 is disposed in a frame 13 that is positioned on the top fittings of the fuel assemblies 70 that are spatially adjacent to the division X. The frame 13 bears video cameras 14 that are directed toward an outlet end of filling-level test lines 15. The filling-level test lines 15 are initially configured as flexible hoses leading to a connection on a height-adjustment device 15a, vent tubes 15b being connected to the connections. The vent tubes 15b are attached to suction tubes 16a, which branch in the form of a two-pronged fork and form an end of flexible extraction lines 16 which are guided out of the water as a bundle. For reasons of clarity, the filling-level test lines 15a are only shown for the fuel assemblies A and Bxe2x80x2 in FIG. 2, but it should be noted that corresponding vent tubes are also disposed on the suction tubes which lead into all the other fuel assemblies. The vent tubes 15b and the suction tubes 16a are jointly lowered by the height-adjustment devices 15a until they are positioned precisely on the top fitting of the corresponding fuel assembly. The positioning may be effected, for example, as a result of the branching point of the fork-shaped suction tubes 16a resting on a bow 110 of a fuel-assembly top fitting 11 or the ends of the suction tubes 16a resting on upper rod-holding plate 112 of the corresponding fuel assembly. If gas is then pumped into the cells 12 (e.g. via the extraction line 16) beneath the hood 6, a water level 17a falls in the cells 12 until it reaches a lower opening of the vent tubes 15b. The vent tube 15b of a cell whose end is highest determines the water level 17a beneath the gas cushion formed in the cell 12, since further gas then escapes through the corresponding vent tube 15b, generating gas bubbles at the other end of the vent tubes 15b, onto which the television camera 14 is directed. Adjustment of the vent tubes 15b and the extraction tubes 16, which are attached to one another, ensures that the extraction tubes always extract water below the water level that has been set in this way. Furthermore, an edge 23 of the fuel assembly channel of each fuel assembly generally projects a relatively long way above its upper rod-holding plate 112. It is therefore generally possible, by adjusting the position of the extraction tubes 16 and the vent tubes 15b with respect to one another, to ensure that even in the case of fuel assemblies which (on account of different production dimensions or different radiation-induced growth) do not all end at the same level, the fuel assembly channels of one group always project slightly above the water level 17a below the gas cushion of the corresponding cell. Fission products that enter the water of a fuel assembly therefore cannot pass through the water to the extraction tube 16 of the other fuel assembly. Therefore, they are also unable to pass into the gas cushion, which is common to a plurality of fuel assemblies, since the water of the fuel assembly is extracted while the pressure difference is still being built up-during heating, and therefore the fission products are initially driven gradually out of a defective fuel rod. FIG. 3 illustrates the first fuel assemblies A, Axe2x80x2 and the second fuel assemblies B, Bxe2x80x2 belonging to groups I and II, which have already been shown in FIG. 2, and also the two remaining fuel assemblies belonging to the groups and the corresponding fuel assemblies belonging to two further groups III, IV of the first division X. Furthermore, there are the further fuel assemblies 70 which belong to another, second division Y. Otherwise, FIG. 3 shows only those components of the apparatus according to the invention that project above the water level of the reactor well. These include four degassing devices 17, which are connected to the extraction lines 16 via a corresponding extraction device 18. The gases that are released in the degassing devices 17 are each removed from an assembly 19 which includes a gas-delivery device and a detector device which is configured to record radioactivity in gases. The measurement signals from the detector devices are fed via line channels 20 to an electronic appliance 21 which evaluates the measured values, displays them on a screen 22 in the form of measurement curves and inputs them to a programmed control unit 25 via an output line 24. Control lines 26 lead from the control unit 25 and are used to control the extraction devices 18 and the delivery devices, by which the removal of the gases released in the degassing devices 17 and the recording of the radioactivity of these gases in the assemblies 19 are controlled. FIG. 3 shows the configuration that is activated during the preliminary testing in groups of the fuel assemblies in the first division X. FIG. 3 indicates that the extraction lines 16 of each extraction device 18 branch a number of times, with a shut-off valve 28, 28xe2x80x2, which is likewise actuated by the control lines 26 of the control device 25, being located in each branch. It can be seen from FIG. 3 that the water from the first fuel assembly A belonging to group I together with the water from the second fuel assembly B and the further fuel assemblies belonging to group I is fed via the extraction lines 16 to one of the four extraction devices 18, namely a device 181, connected to a degassing device 171 and an assembly 191 (detector device), the shut-off valves 28 in these lines being set to pass (i.e. being open). FIG. 3 also shows that the same extraction device 18 which is connected to the first fuel assembly A belonging to group I is also connected to three further extraction lines 16xe2x80x2, the shut-off valves 28xe2x80x2 of which are in a blocking position. The extraction lines 16xe2x80x2 are provided in order, in the configuration shown in FIG. 4, to switch the corresponding extraction device in each case to a first fuel assembly belonging to one of the other groups II, III and IV as desired. This is because if neither the detector device of the assembly 191 nor any of the other detector devices on any of the channels reveals any significant increase in the radioactivity, the testing of all the fuel assemblies belonging to the first group I has ended. Then, the extraction hood 6 is raised by the handling machine, all the shut-off valves are opened by the control device 25 and all the lines are purged with pool water via the extraction devices 18. The extraction hood 6 can then be positioned at a new position above a further division of fuel assemblies, and the testing of the fuel assemblies belonging to a new division is commenced by purging the degassing devices 17 with fresh air via the extraction devices 18 and initially delivering fresh air beneath the hood 6 which has been fixed at the new position via the extraction lines 16. However, if a measurement channel, for example for group III, records a significant increase in the radioactivity, there is a leak in a fuel assembly in group III associated with the channel, and the leak has to be identified. For this purpose, as shown in FIG. 4, the devices which have already been used during the preliminary test are now assigned to the fuel assemblies in a different way by their shut-off valves 28 and 28xe2x80x2. For example, the extraction device 181, which in FIG. 3 was connected to the further fuel assemblies belonging to the first group I, is switched over so that it is only connected to the first fuel assembly belonging to group III which has the significant radioactivity, and likewise in each case one of the further extraction devices is switched over to in each case one fuel assembly belonging to the significant group III. In this way, each individual fuel assembly can be individually tested by extraction, degassing and detection, the corresponding measurement curve in the display 22 now allowing the defective fuel assembly to be identified. The extraction devices are assigned to the individual fuel assemblies in a program-controlled manner via the control device 25 through actuation of the shut-off valves 28. FIG. 5 shows a water circuit above the level of water 40 in which the fuel assemblies to be tested are positioned. A water circuit feeds the water 40, which has been sucked out beneath one or more hoods 6 in accordance with FIGS. 3 and 4, and returns the water 40 via the overflow line 16a after the degassing, to a collection vessel of the degassing device 171 via the extraction line 16 and an extraction pump 181. A line 41 that is used to suck in ambient air, for example via a pump 42 and an air filter 43, opens into the collection vessel of the degassing device 171. However, a throttle valve can also be used to introduce nitrogen or any other non-radioactive gas out of a pressure vessel into the collection vessel of the degassing device 171. The carrier gas bubbles through the collected water in the collection vessel of the degassing device 171 and collects all the gases that are dissolved and released in the extracted water. In this example, the release of the gases is facilitated by a pump 191a that, together with a corresponding outlet pump 181a, generates a vacuum in the collection vessel. The pumps, compressors and similar delivery devices 181, 181a, 191a and 42 illustrated here are controlled synchronously with an evaluation device 21 of a detector device 191b by the control device and are in this case only symbolically illustrated for corresponding devices which the person skilled in the art will provide at any time in order to ensure that, during the testing of a fuel assembly, suitable pressure conditions for extracting the water from the fuel assembly and for transferring the fission products which are released and are mixed with the carrier gas are ensured in the collection vessel 171. In the situation illustrated, the detector device 191b contains a xcex2 counter 19a and a xcex3 counter 19b, the sensitivity spectrum of which is specifically adapted to the energy spectrum of the fission products that are most frequently produced. The person skilled in the art selects the number and type of detectors in accordance with those isotopes that are primarily expected to form in the interior of the fuel rods. A disposal line 44 on the one hand connects the inlet of the detector device 191b to the outlet of the degassing device 171 and on the other hand connects the outlet of the detector device 191b to an outgoing-air duct of the nuclear power plant. Before water samples from the fuel assemblies are tested, the configuration is purged with water that, although it originates from the same pool, is not removed beneath the hood that has already been fitted over a fuel assembly that is in the heat-up phase. At this time, therefore, only fission products that are already present in the extracted water and form a constant xe2x80x9cradioactivity backgroundxe2x80x9d for the subsequent measurement are released in the extracted water. If a fuel assembly that is beneath the hood connected to the line 16 is now heated and the radioactivity recorded in the detector configuration rises, the rise is attributable to fission products escaping from the fuel assembly which is in the heat-up phase, and the measured values rise by the extent to which the discharge of fission gases increases. As a result, defects in the claddings of the fuel rods are detected at an early stage, and it is possible to start measuring further fuel assemblies at an early stage. The invention is not restricted to boiling water fuel assemblies and positions in the reactor pressure vessel. Rather, the function of the fuel assembly channel of an assembly of this type may also be performed by other containers or, for example, by corresponding shafts of a storage rack for pressurized water fuel assemblies. |
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claims | 1. A neutron flux mapping system for a nuclear reactor comprising:drivers each including a geared motor, a helical gear driven by the geared motor, and a storage reel adapted to supply, to the helical gear, a detector cable carrying a detector; anda double indexing path selector unit includinga body including upper and lower fixed plates, and tie rods connecting the upper and lower fixed plates,a fixed shaft fixedly mounted at a central portion of the body,an outer path selector arranged to be rotatable about the fixed shaft,the outer path selector including an upper rotating plate arranged to be rotatable about the fixed shaft while carrying a drive unit for rotating the outer path selector about the fixed shaft, and a control unit for controlling the drive unit, and a lower rotating plate arranged to be symmetrical with the upper rotating plate, and connected to the upper rotating plate to rotate along with the upper rotating plate, andinner path selectors each including a hollow rotating shaft rotatably mounted between the upper and lower rotating plates of the outer path selector, a path select tubing connected, at an upper end thereof, to an upper end of the rotating shaft in the interior of the rotating shaft while extending downwardly and radially outwardly from the rotating shaft through a hole formed at the rotating shaft, and a disc mounted to a lower end of the rotating shaft, and provided with a plurality of circumferentially-arranged paths. 2. The neutron flux mapping system according to claim 1, wherein the geared motor comprises an induction motor adapted to be controlled by an inverter. 3. The neutron flux mapping system according to claim 1, wherein:each driver further includes means for bring the helical gear into close contact with the detector cable; andthe means comprises at least one idle gear. 4. The neutron flux mapping system according to claim 3, wherein the at least one idle gear comprises:a first idle gear arranged to face the helical gear in a horizontal direction; anda second idle gear arranged to face the helical gear in a vertical direction. 5. The neutron flux mapping system according to claim 4, wherein the second idle gear is arranged toward the detector cable outlet in advance of a vertical center line of the helical gear by a predetermined angle. 6. The neutron flux mapping system according to claim 1, wherein each driver further includes an AC torque motor adapted to drive the storage reel. 7. The neutron flux mapping system according to claim 1, wherein each driver further includes an acoustic vibration sensor. 8. The neutron flux mapping system according to claim 1, wherein:the upper fixed plate is provided with stop plates; andthe control unit of the upper rotating plate is provided with limit switches, which selectively come into contact with the stop plates, respectively. 9. The neutron flux mapping system according to claim 8, wherein the stop plates are arranged to allow the upper rotating plate to rotate through a predetermined angle in a reciprocating manner. 10. The neutron flux mapping system according to claim 8, wherein:the control unit of the upper rotating plate is provided with a latch; andthe upper fixed plate of the selector unit body is provided with latch rods engagable with the latch. 11. The neutron flux mapping system according to claim 10, wherein the control unit further includes a plurality of limit switches adapted to detect a rotated position of the upper rotating plate in cooperation with the latch rods. 12. The neutron flux mapping system according to any one of claims 8 to 11, wherein the control unit further includes an acoustic vibration sensor. 13. The neutron flux mapping system according to any one of claims 1 to 8, wherein the driving unit and the control unit are mounted on an upper surface of the upper rotating plate. 14. The neutron flux mapping system according to claim 1, wherein:each inner path selector further includes an indexing mechanism adapted to rotate the rotating shaft such that the path select tubing is aligned with a selected one of the paths;the indexing mechanism includes a plurality of path select switches arranged around the disc to correspond to the paths, respectively, each of the path select switches sensing alignment of a corresponding one of the paths with the path select tubing, thereby generating a sensing signal; andthe indexing mechanism is driven by a geared motor controlled by the sensing signal. 15. The neutron flux mapping system according to claim 14, wherein the plurality of path select switches are connected in the form of a matrix arrangement while being coupled to an input module included in a control system. 16. The neutron flux mapping system according to claim 15, wherein the control system controls the geared motor of the indexing mechanism, based on signals received from the path select switches. 17. The neutron flux mapping system according to claim 16, wherein:the indexing mechanism further includes a cam switch adapted to perform an ON/OFF operation every time the indexing mechanism rotates a predetermined unit angle; andthe control system checks, based on a signal generated in accordance with the ON/OFF operation of the cam switch, whether or not the indexing mechanism and the path select switches operate normally. 18. The neutron flux mapping system according to claim 17, wherein, when the index mechanism or one of the path select switches operates abnormally, the control system performs a control operation for treating, as being in a failure state, the inner path selector associated with abnormally-operating the index mechanism or path select switch, withdrawing the associated detector to a predetermined position when the detector has been inserted into the inner path selector, and preventing use of the inner path selector. 19. The neutron flux mapping system according to claim 1, wherein the double indexing path selector unit further includes tubing anti-twister means. 20. The neutron flux mapping system according to claim 19, wherein the tubing anti-twister means comprises:an anti-twister frame; anda rotating plate rotatably mounted in the anti-twister frame. 21. The neutron flux mapping system according to claim 20, wherein:the rotating plate is provided with a stop pin; andthe anti-twister frame is provided with stop plates engagable with the stop pin. 22. The neutron flux mapping system according to claim 21, wherein the stop plates are respectively mounted to diametrically-opposed ends of the anti-twister frame. 23. The neutron flux mapping system according to any one of claims 20 to 22, wherein the rotating plate is further provided with rotors. 24. The neutron flux mapping system according to claim 19, wherein the tubing ant-twister means further comprises withdraw limit switches each adapted to detect passage of the detector cable, associated therewith, through the double indexing path selector unit. 25. The neutron flux mapping system according to claim 24, wherein each of the withdraw limit switches is a non-contact type proximity reed switch. 26. The neutron flux mapping system according to claim 1, further comprising:detector storage guiders each arranged between an associated one of the drivers and the double indexing path selector unit; anda detector storage area adapted to store the detector guided by an associated one of the detector storage guiders. |
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062332996 | description | PREFERRED EMBODIMENTS OF THE INVENTION FIG. 1 is a transverse sectional view of an assembly for transmutation (ie., transmutation assembly) of a long-lived radioactive material according to an embodiment of the present invention and an example of a FP pin loaded into the transmutation assembly. The FP pin 20 has a structure in which a single wire-type member 22 of the long-lived radioactive material composed of metals, alloys or compounds including long-lived fission product (LLFP) nuclides is disposed at a center of the FP pin and the wire-type member 22 is surrounded by a moderator material 24 to form a pellet-like or rod-like structure and loaded in a cladding tube 26. A plurality of such FP pins 20 are prepared and arranged in a bundle-like structure and located in a hexagonal wrapper tube 28 to thereby form the transmutation assembly 30 of the long-lived radioactive material of the present invention. The long-lived fission product (LLFP) nuclides include, for example, technetium-99 and iodine-129 and are used in the form of a metal, an alloy or a compound. In the case of technetium, for example, Tc (metal), TcO.sub.2 and so forth are used. In the case of iodine, AgI, NaI, PdI.sub.2, CeI.sub.3 and so forth can be used. The wire-type member 22 of the long-lived radioactive material is preferably formed such that it has a diameter of about 1 to 2 mm. This will permit moderator to slow down neutrons in a suitable manner and restrict as much as possible the self-shielding effect of the neutrons (that is, the effect of preventing the neutrons entering deep into the FP) so that a high transmutation rate can be achieved. The wire-type member 22 having a diameter of 1 mm or more can be produced relatively easily. As the moderator material 24, zirconium hydride or beryllium oxide, for example, can be used. The pellet or rod, which is formed by surrounding the wire-type member 22 of the long-lived radioactive material by means of the moderator material 24 as described above, is inserted into the cladding tube 26 along a substantially entire length thereof and sealed at its upper and lower ends by end plugs (not shown) in a manner similar to the case of general fuel pins. The wrapper tube 28 in which a plurality of FP pins 20, and nothing else, are located in a regular arrangement has a similar structure to that of a general fuel assembly and has an entrance nozzle (not shown) at its lower portion and a handling head (not shown) at an upper portion for facilitating the loading work by the use of a fuel loading/unloading machine so that a coolant can flow inside the wrapper tube 28. FIG. 2 shows another embodiment of the FP pin. In this embodiment, a plurality of wire-type members (seven wire-type members in the illustrated embodiment) 32 are disposed in a dispersed arrangement and each of the wire-type members 32 is surrounded by a moderator material 34 to form a pellet or rod structure, and loaded into a cladding tube 36. Installation of a plurality of wire-type members 32 of the long-lived radioactive material can increase the amount of transmutation. Similar to the case of the previous embodiment, it is desirable that the diameter of the wire-type member 32 of the long-lived radioactive material is selected from a range from about 1 mm to 2 mm. It is appreciated that the number of the wire-type members 32 and the location thereof are optional and can be selected as desired in accordance with requirements. FIG. 3A, FIG. 3B and FIG. 3C show further embodiments of the FP pin, in which a thin ring-type or thin-wall tubular member 42 of a long-lived radioactive material is used. In the embodiment of FIG. 3A, a thin ring-type member 42 of the long-lived radioactive material has a relatively small diameter and is positioned at the center of the pin so that the thin ring-type member 42 is surrounded at both its inner and outer surfaces by a moderator material to form a pellet-like or rod-like structure and loaded into the cladding tube 46. In the embodiment of FIG. 3B, a thin ring-type member 48 of the long-lived radioactive material having a relatively large diameter is provided so that the thin ring-type member 48 is surrounded at it both inner and outer surfaces by the moderator material 44 to form a pellet-like or rod-like structure and then loaded into the cladding tube 46. In the embodiment of FIG. 3C, the above-described thin ring-type member 42 of a small diameter and the thin ring-type member 48 of a large diameter are concentrically located and surrounded at their inner and outer surfaces by the moderator material 44 to form a pellet-like or rod-like structure, and loaded into the cladding tube 46. This structure having a plurality of thin ring-type members 42 and 48 of the long-lived radioactive material can increase its transmutation performance. In the embodiments of FIGS. 3A, 3B and 3C, it is desirable that each of the thin ring-type members 42, 48 has a thickness of about 1 to 2 mm. The FP pins shown in FIGS. 3A, 3B and 3C are suitable for decreasing or lowering the self-shielding effect of neutrons and increasing the transmutation rate. The thin ring-type members each having a thickness of 1 mm or more can be produced relatively easily. The transmutation assembly of a long-lived radioactive material can be formed by using one of the types of those FP pins described above and, therefore, production and inspection of the transmutation assembly can be performed quite simply, with the result of a reduction in costs. The transmutation assemblies of the present invention can be loaded selectively and partly into a core region, a blanket region or a shield region of a reactor core in a fast reactor. When the transmutation assemblies are loaded into the blanket region, all of the blanket assemblies may be replaced by the transmutation assemblies. In order to effectively use the excess of neutrons and restrict an influence upon the reactor core characteristics, it is optimal that the transmutation assemblies are loaded in the blanket region. EXAMPLE An experiment was made with reference to a fast reactor which has transmutation assemblies of a long-lived radioactive material loaded in the blanket region and detailed analysis was made by using a Monte Carlo Code. Table 1 below shows the results. The transmutation assembly which is the scope of the present invention (Invention 1 and Invention 2, below) had a structure as shown by FIG. 1, in which long-lived technetium-99 was formed into a thin metal wire and surrounded by a moderator material of zirconium hydride to form pellets, and then the pellets were loaded into a cladding tube to form a FP pin. A plurality of FP pins surrounded by a wrapper tube form a transmutation assembly. The transmutation assemblies were loaded into a blanket region of a fast reactor of 1 million kWe as shown in FIGS. 4 and 5 and a one-year term transmutation rate was obtained. FIG. 4 shows the core structure of the fast reactor and the loading position of the transmutation assemblies and FIG. 5 shows the dimensions of the reactor core. In this structure, it is seen that the transmutation assemblies are loaded in the position of a radial blanket. For comparison purposes, analysis was made for the conventional transmutation assembly of a long-lived radioactive material shown in FIG. 6, as shown by "Conventional Example (Conv. Ex.)1" and "Conventional Example (Conv. Ex. 2)" in Table 1 below. TABLE 1 No. of No. Diameter Trans- Trans- Loading pins in of of Loaded muted mutation method of assem- FP FP pins amount amount rate FP pins bly pins (mm) (kg) (kg/year) (%/year) Conv. Ex. 127 37 10 3750 67.5 1.8 1 Conv. Ex. 127 22 10 1883 45.8 2.5 2 Invention 127 127 1.3 183 17.9 9.8 1 Invention 217 217 1.3 313 28.5 9.1 2 In Table 1, the "No. of FP pins" of the conventional examples (Conv. Ex. 1 and Conv. Ex. 2) represent the FP pins into which only a long-lived radioactive material is loaded and the "No. of FP pins" of the present invention ("Invention 1" and "Invention 2") represent the FP pins loaded with pellets which are composed of wire-type long-lived radioactive material surrounded by the moderator material. Accordingly, the radius of the conventional FP material is coincident with the pin radius, and the radius of FP material of the invention is equal to the radius of the wire type material and the radius of the actual pins (FP pins) is 5 mm, which is the same as that of the conventional ones. As shown by Table 1, the transmutation rate is low (that is, 1.8 to 2.5 %/year) in the case where the conventional transmutation assembly is used and is therefore not so effective. By contrast, use of the transmutation assembly for long-lived radioactive material of the present invention successfully achieved a high transmutation rate of about 9 to 10%/year, which is 4-5 times as high as the conventional structure, regardless of the number of pins used in the assembly. From the analysis, the number of pins in the transmutation assembly in the present invention can be extended to about 127-271 as shown in Table 1. In the example, the wire-type long-lived radioactive material had a diameter of 1.3 mm but any wire-type long-lived material can be used if it has a diameter in a range between about 1 mm and about 2 mm. According to the present invention, the transmutation assembly is composed solely of FP-containing pins each of which is formed of a cladding tube and wire-type or thin ring-type members of a long-lived radioactive material surrounded by a moderator material; this allows the transmutation rate of the long-lived FP nuclides to achieve an extraordinarily high level. Further, the transmutation assembly can be formed with FP pins of the same type, with a consequent simplification of production and inspection and reduction of costs. |
claims | 1. A charged particle beam writing apparatus for deflecting a charged particle beam by a main deflector and a sub deflector to write a pattern sample, comprising:a shot data generator for generating shot data from write data in which the shape and position of each graphic pattern are defined; anda deflection controller for generating deflection data for controlling the main deflector and the sub deflector from the shot data,wherein the shot data generator comprises shot dividing means for dividing the graphic pattern defined in the write data into graphics expressed in shot units, and means for distributing the respective graphics divided by the shot dividing means to subfield areas capable of being deflected by the main deflector, such that a number of shots over all subfield areas is constant regardless of a presence or absence of subfield boundaries,wherein a plurality of subfield areas are configured in such a manner that the subfield areas adjacent to one another overlap one another in widths greater than or equal to a maximum shot size, andwherein when part of each of the graphics divided by the shot dividing means is located in an overlapped portion of each of the subfield areas, the distributing means distributes the corresponding graphic to any one of the subfield areas, based on the position of a standard point of the graphic. 2. The charged particle beam writing apparatus according to claim 1, further comprising:a first shaping aperture having an opening for allowing a charged particle beam irradiated from a charged particle gun to be penetrated therethrough; anda second shaping aperture having an opening for allowing the charged particle beam penetrated through the opening of the first shaping aperture to be penetrated therethrough,wherein the shot data generator further comprises write data modifying means for executing a write data modifying process for moving the position of each of the graphics divided by the shot dividing means on write data, based on which portion of the opening of the second shaping aperture is a graphic corresponding to each pattern written by the penetrated charged particle beam, andwherein the graphics whose positions have been moved on the write data by the write data modifying means, are distributed by the distributing means. 3. A charged particle beam writing method for writing each pattern onto a sample using a charged particle beam deflected by a main deflector and a sub deflector, comprising the steps of:generating shot data from write data in which the shape and position of each graphic pattern are defined; andgenerating deflection data for controlling the main deflector and the sub deflector from the shot data,wherein the shot data generating step includes a shot dividing step for dividing the graphic pattern defined in the write data into a plurality of graphics expressed in shot units, and a distributing step for distributing the divided graphics to their corresponding subfield areas capable of being deflected by the main deflector, such that a number of shots over all subfield areas is constant regardless of a presence or absence of subfield boundaries,wherein a plurality of subfield areas are configured in such a manner that the subfield areas adjacent to one another overlap one another in widths greater than or equal to a maximum shot size, andwherein, in the distributing step, when part of each of the graphics divided at the dividing step is located in an overlapped portion of each of the subfield areas, the corresponding graphic is distributed to any one of the subfield areas, based on the position of a standard point of the graphic. 4. The charged particle beam writing method according to claim 3, wherein after the completion of the shot dividing step, the distributing step is executed while a process for developing positional information defined in the write data is being performed. 5. The charged particle beam writing method according to claim 3, wherein after the completion of the shot dividing step, the distributing step is executed while a process for developing positional information defined in the write data is being performed. 6. The charged particle beam writing method according to claim 3, wherein the charged particle beam is further penetrated through an opening of a second shaping aperture after having been penetrated through an opening of a first shaping aperture, wherein the shot data generating step further comprises a write data modifying step for executing a write data modifying process for moving the position of each of the graphics divided at the shot dividing step on write data, based on which portion of the opening of the second shaping aperture is a graphic corresponding to each pattern written by the penetrated charged particle beam and wherein the graphics whose positions have been moved on the write data in accordance with the write data modifying step, are distributed at the distributing step. |
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052672859 | summary | BACKGROUND OF THE INVENTION This invention relates generally to cooling systems for nuclear reactors, and, more particularly, to an apparatus and method for suppressing the formation of vortices in coolant fluid circulating within such reactors. Generally, nuclear reactors include a cylindrically reactor vessel having a hemispherical lower end. Within the reactor vessel is the reactor core, supported by a main core support. The main core support is connected to the interior walls of the reactor vessel at or adjacent to the area where the cylindrical and hemispherical portions of the reactor vessel meet. Below the main core support, the hemispherical vessel defines a lower head, or lower plenum. A generally cylindrical downcomer surrounds the reactor core. Coolant fluid, typically water, is pumped into the downcomer. The coolant fluid circulates downwardly into the lower plenum. The hemispherical shape of the lower plenum assists in evenly circulating the coolant fluid therein. A plurality of reactor core coolant inlet openings are located on the underside of the main core support. Coolant flows from the lower plenum, into the core coolant inlet openings and upwardly into the core to cool the fuel assemblies. In order to maintain adequate and uniform cooling throughout the core, it is important that a uniform coolant flow and pressure be maintained across all of the reactor core coolant inlet openings. Non-uniform coolant pressure or flow causes uneven coolant flow into the core, which results in uneven cooling of the fuel assemblies of the core. Uneven fuel assembly cooling may force the entire core to be derated to accommodate "hot assembly" locations. Non-uniform coolant flow and pressure may result if vortices or other flow disruptions form in the coolant fluid circulating in the lower plenum. It is desirable to provide core monitoring instrumentation within the core of a nuclear reactor. Traditionally, the leads connecting such instrumentation to the exterior of the reactor exit the reactor vessel through a central portion of the hemispherical portion of the reactor vessel. A plurality of conduits carry the instrumentation lines through the lower plenum. The presence of the conduits in the lower plenum assists in maintaining even coolant flow within the lower plenum and disrupting the formation of vortices in the circulating coolant fluid. Such vortices disrupt coolant flow and produce low pressure areas at the core coolant inlets which they intersect. In newer reactors, it has become desirable for any instrumentation conduits to exit the reactor vessel other than through the lower plenum. It has been found that the absence of instrumentation conduits from the lower plenum permits vortices to form in the circulating coolant in the lower plenum. Accordingly, there is a need for a simple and inexpensive apparatus and method for effectively suppressing the formation of vortices in the coolant fluid circulating within the lower plenum of a nuclear reactor where instrumentation conduits are not present in the lower plenum. SUMMARY OF THE INVENTION The present invention has met the above-described needs. This invention provides a generally planar plate which is suspended in the lower plenum of a reactor vessel below and generally parallel to the main core support. The plate has a plurality of openings therein which allow coolant fluid to flow therethrough. A plurality of support columns connect the upper surface of the plate with the lower surface of the main core support. The plate is positioned within the lower plenum so as to optimize its vortex suppression effect. The plate's distance below the bottom of the main core support, its distance from the interior side walls of the reactor vessel and its distance above the center of the interior of the hemispherical portion of the reactor vessel may be selected so as to maximize vortex suppression. Preferably, the plate will be positioned so as to intersect any potential vortex at or near the center line of such vortex. Modelling techniques may be utilized to predict the locations where vortices are likely to form. The plate's diameter, its thickness and the size of the openings therein may also be varied to obtain maximum vortex suppression. The vortex suppression plate may also be utilized to brace secondary core support columns which extend generally downwardly from the main core support. A secondary core support plate is typically disposed near the bottom of the center of the lower plenum. It is an object of this invention to provide an apparatus and method for suppressing the formation of vortices in coolant fluid circulating in the lower plenum of a nuclear reactor. It is another object of this invention to provide an apparatus and method for suppressing the formation of such vortices in reactors in which core monitoring instrumentation conduits are absent from the lower plenum thereof. It is a further object of this invention to provide an apparatus and method for suppressing formation of such vortices which assist in maintaining uniform coolant flow and pressure to the reactor coolant inlets. It is yet another object of this invention to provide an apparatus and method for suppressing formation of such vortices which produce minimal contribution to the pressure drop of the coolant fluid circulating within the lower plenum. It is another object of this invention to provide an apparatus and method for suppressing formation of such vortices which are simple and relatively inexpensive. It is a further object of this invention to provide an apparatus and method for suppressing formation of such vortices which are readily realized using existing hardware design and fabrication techniques. These and other objects of the present invention will be more fully understood from the following description of the preferred embodiment of the invention with reference to the drawings appended hereto. |
description | This application is a continuation-in-part (CIP) application based upon the International Application PCT/JP2009/004384, the International Filing Date of which is Sep. 4, 2009, the entire content of which is incorporated herein by reference, and claims the benefit of priority from the prior Japanese Patent Application No. 2008-255573, filed in the Japanese Patent Office on Sep. 30, 2008, the entire content of which is incorporated herein by reference. Embodiment described herein relate generally to a pressurized water reactor plant. In general, a pressurized water reactor (PWR) in a commercial nuclear plant for use in power generation or hydrogen production needs to satisfy national safety standards, and the minimum number of loops between steam generators and reactor coolant system loops is set to two. In a typical conventional two-loop PWR, two reactor coolant system loops are disposed symmetrically with respect to a reactor vessel. A reactor coolant pump and a steam generator are disposed in each of the reactor coolant system loops, and the steam generator and the reactor vessel are connected to each other by a hot leg pipe and a cold leg pipe. The reactor coolant pump is disposed on the cold leg pipe. Further, two separate emergency core cooling systems (ECCS) each inject cooling water through an injection nozzle disposed on the cold leg pipe. Two separate ECCS injection pipes are connected to each other by a tie line and are configured to be able to inject water into any of the cold leg pipes. The reason that two reactor coolant system loops are required in terms of safety as described above is as follows. (1) Responding to One-Pump Trip Transient In one-pump trip transient, another reactor coolant pump continues to work to ensure a core flow rate required for cooling core fuel, guaranteeing integrity and reusability of the fuel. If the second reactor coolant pump is not provided, the one-pump trip transient can be a serious accident event entirely equivalent to all-pump trip, making it impossible to satisfy safety standards for a transient event. (2) Responding to One-Pump Seizer Accident At the time of a pump seizer accident where a rotor of one reactor coolant pump is suddenly locked during operation, the another reactor coolant pump coasts down according to its inertia to ensure the minimum core flow rate required for cooling a reactor core to thereby reduce the failure of the core fuel to the minimum level. Further, overpressure in a reactor pressure boundary is prevented so as to satisfy safety standards at the accident time. If the second reactor coolant pump is not provided, the core flow rate immediately runs short due to the pump seizer accident, resulting in occurrence of a serious failure of the core fuel and overpressure in the reactor pressure boundary. (3) Responding to Loss-of-Coolant Accident If one cold leg pipe is ruptured at the time of loss-of-coolant accident (LOCA), one emergency core cooling system for injecting cooling water into the cold leg pipe may be disabled. Another emergency core cooling system is assumed to be disabled according to a single-failure criterion that one emergency core cooling system is disabled. Even in such a case, the intact emergency core cooling system injects cooling water into the intact cold leg pipe through a tie line so as to cool the core fuel. If the second intact cold leg pipe does not exist, both the two separate emergency core cooling systems may be disabled. (4) Responding to Steam Generator Tube Rupture Accident Upon occurrence of a steam generator tube rupture accident (SGTR), the intact steam generator is used to perform primary system depressurization to equalize the pressure of a primary system to the pressure of a secondary system, thereby stopping outflow of nuclear reactor coolant from the ruptured steam generator to the secondary system. If the second steam generator is not provided, depressurization can be achieved only by means of a pilot operated relief valve (PORV) or a pressurizer spray, resulting in a prolonged outflow of the primary coolant to the secondary system. The prolonged outflow of the primary coolant leads to a prolonged discharge of the primary coolant from the relief valve of the steam generator to environment. For the above reasons, the minimum required number of the reactor coolant system loops of the conventional PWR is set to two. The two-loop PWR is the minimum constituent unit in the conventional PWRs and generally generates a power of 300 MWe to 600 MWe. That is, the two-loop PWR is categorized as a small-sized reactor as commercial reactors. Basically, design concepts of more than 50 years ago are used for the two-loop PWR, and many active components including pumps, etc., are used for a safety system such as the ECCS. Therefore, there was a problem that the safety of a nuclear reactor cannot be maintained when a prolonged station blackout (SBO) occurs. The number of the reactor coolant system loops has been increased to three or four in order to increase the output power of the PWR. A three-loop PWR generally generates a power of 800 MWe to 900 MWe as a middle-sized reactor. A four-loop PWR generally generates a power of 1100 MWe class or more as a large-sized reactor. In the conventional four-loop PWR, four reactor coolant system loops are disposed around the rector vessel. As in the two-loop PWR, one reactor coolant pump and one steam generator are disposed in each reactor coolant system loop. In recent years, supersized four-loop reactors capable of generating a power of 1600 MWe class or more are built or planned. The steam generator in such a supersized reactor has a height of as high as about 20 m, and the volume of a containment vessel for housing four such steam generators reaches up to as extremely great as about 80,000 m3. In such a way the reactor output power of a large PWR is increased by simply increasing the number of loops, and its deign is based on the same design concepts as those of more than 50 years ago for the two-loop PWR. Thus, a full passive safety system without any motor-driven pumps has not been adopted. On the contrary, there is an AP1000 as an example of a passive safety PWR that is capable of increasing the reactor output power without increasing the number of the reactor coolant system loops from two and capable of satisfying the safety standards for accidents with only passive safety systems in which any motor-driven pumps are not used (Refer to, e.g., IAEA-TECDOC-1391, “Status of advanced light water reactor designs 2004”, IAEA, May 2004, p207-p231, p279-p306, the entire content of which is incorporated herein by reference). A description will be given on the AP1000 below with reference to FIGS. 5 to 9. FIG. 5 is a plan view illustrating a configuration of the reactor coolant system loop (two-loop structure) in a conventional passive safety PWR (AP1000). FIG. 6 is an elevation view illustrating the steam generator and reactor coolant pump of FIG. 5. FIG. 7 is a side view of the steam generator and reactor coolant pump of FIG. 6, which illustrates the inside of the steam generator in a sectional manner. FIG. 8 is an elevation cross-sectional view of a containment vessel used in the passive safety PWR of FIG. 5 and inside thereof. FIG. 9 is a block system diagram of a reactor pressure boundary and a passive cooling and depressurization system (PCDS) used in the passive safety PWR of FIG. 5. In FIG. 5, a reactor core 1 is housed in a rector vessel 2. Two reactor coolant system loops 50a and 50b are disposed symmetrically with respect to the rector vessel 2. Steam generators 3a and 3b are disposed in their respective reactor coolant system loops. The steam generators 3a, 3b and rector vessel 2 are connected by hot leg pipes 5a, 5b and cold leg pipes 4a, 4b, 4c, 4d. Two reactor coolant pumps 6a and 6b are directly connected to the lower portion of the steam generator 3a, and two reactor coolant pumps 6c and 6d are directly connected to the lower portion of the steam generator 3b. Two separate emergency core cooling systems (ECCS) inject cooling water into the reactor vessel 2 through direct vessel injection nozzles 58a and 58b. Therefore, even if a loss-of-coolant accident in which a cold leg pipe or the like is ruptured occurs, the ECCS is not disabled. The AP1000 generates a power of about 1117 MWe and thus belongs to four-loop large-sized PWR class in the conventional classification. Thus, four cold leg pipes and four reactor coolant pumps are provided. However, the volume of each steam generator is increased to reduce the numbers of the steam generators and hot leg pipes to two, respectively, thereby succeeding in reducing the number of loops from four to two. This significantly improves the layout efficiency in the containment vessel, thereby succeeding in reducing the volume of the containment vessel to as small as about 58,000 m3. The above advantages are brought about by the improvement of the configuration of the reactor coolant pump directly connected to the steam generator. FIGS. 6 and 7 are structural views each illustrating an installation method of the steam generator and reactor coolant pump of the AP1000 as a conventional passive safety reactor. The two steam generators have the same configuration, and thus only the steam generator 3a will be described hereinafter. The two reactor coolant pumps 6a and 6b are directly connected to a channel head 91 disposed at the lower part of the steam generator 3a. FIG. 7 illustrates a connecting state between the steam generator 3a and the reactor coolant pump 6a as viewed from the direction perpendicular to the direction of FIG. 6. Since the two reactor coolant pumps are overlapped in this point of view, only the reactor coolant pump 6a is illustrated. Further, the internal structure of the steam generator is also illustrated. A large number of tubes 92 which are heat exchange pipes having a reverse U-shape are disposed inside a barrel portion 22 of the steam generator 3a. In FIG. 7, only one tube is illustrated. The tube 92 is disposed on a tube sheet 93, and the inside of the steam generator is separated into a primary side 94 and a secondary side 95 by tube sheet 93 and tubes 92. The inside of the tubes 92 and the space below the tube sheet 93 are referred to as a primary side of the steam generator. The outside of the tubes 92 above the tube sheet 93 is referred to as a secondary side of the steam generator. The primary sides of the steam generators and a system connected to the primary sides of the steam generators are collectively referred to as a primary system. Conversely, the secondary sides of the steam generators and a system (not illustrated) connected to the secondary sides of the steam generators are collectively referred to as a secondary system. A water plenum 96 which is a primary side component is located below the tube sheet 93. The water plenum 96 is divided into an inlet side and an outlet side by a divider plate 97. An inlet nozzle 98 is located on the inlet side, and the hot leg pipe 5a is connected to the inlet nozzle. The two reactor coolant pumps 6a and 6b are connected to the outlet side of the water plenum. Coolant is sucked by the two reactor coolant pumps and discharged from outlet nozzles 99 of the reactor coolant pumps. The cold leg pipes 4a and 4b are connected to the outlet nozzles 99, respectively. The two reactor coolant pumps are connected to one steam generator, and thus the two cold leg pipes are connected to one steam generator. In FIG. 8, the reactor core 1 is housed inside the reactor vessel 2. The reactor vessel 2 is connected to the two steam generators 3a and 3b by the cold leg pipes 4 (4a, 4b, 4c, 4d) and hot leg pipes 5 (5a, 5b). Further, the reactor coolant pumps 6 (6a, 6b, 6c, 6d) are directly connected to the lower portions of the steam generators 3a and 3b. The components and pipes constituting the reactor pressure boundary are housed inside a containment vessel (CV) 7. The containment vessel 7 of the AP1000 is the most typical containment vessel, called “large dry CV”, for use in PWRs. The containment vessel 7 is made of steel, because it is designed to be cooled with the external air in case of an accident. Inside the containment vessel 7, an in-containment refueling water storage tank 8 is disposed. The in-containment refueling water storage tank 8 works as a gravity-driven cooling system (GDCS) if a loss-of-coolant accident (LOCA) in which the cold leg pipe 4 or the like is ruptured occurs. This gravity-driven cooling system cooperating with other passive ECCS submerges the lower part of the containment vessel up to a higher level than the cold leg pipe 4. After that, it is designed so that a recirculation screen (not illustrated) is opened, introducing water always into the reactor vessel 2 to cool the fuel in the reactor core safely. Once the water introduced into the reactor vessel 2 is heated by the decay heat of the fuel in the reactor core, steam is generated and the steam fills the gas phase of the containment vessel 7, resulting in a rise of the temperature and pressure in the containment vessel 7. A shield building 71 is built outside the containment vessel 7. A cooling water pool 72 of a passive containment cooling system (PCS) is disposed on the top of the shield building 71. The cooling water pool 72 is filled with PCS pool water 73. In case of a loss-of-coolant accident, the PCS pool water 73 drains onto the containment vessel 7. Air flows into the shield building 71 through an external air inlet 74 and then a natural circulation force raises the air through the gap between an air baffle 75 and the wall of the containment vessel 7 until the air is discharged outside through a heated air discharge 76 formed at the top of the shield building 71. The combination of the drainage of the PCS pool water 73 and the natural convection of air cools the containment vessel 7 safely. The shield building 71 including its side wall and ceiling portions has a structure endurable against a large plane crash. In this way, AP1000 can cool the reactor core 1 and containment vessel 7 with an extremely high reliability only by the passive safety systems requiring no external AC power source. However, the plant output power of the AP1000 is as large as 1117 MWe, and the decay heat after an accident is significantly high, so that the PCS pool water 73 depletes in about three days after the accident. Thereafter, the PCS pool water 73 needs to be replenished. That is, the cooling cannot be achieved only by external cooling air. In FIG. 9, the reactor pressure boundary of the AP1000 is constituted by one rector vessel 2, primary sides of the two steam generators 3a and 3b, two hot leg pipes 5a and 5b connecting therein, four cold leg pipes 4a, 4b, 4c, and 4d, and one pressurizer 80. The cold leg pipes 4a, 4b, 4c, and 4d circulate coolant cooled by the steam generators 3a and 3b into the reactor vessel 2 by means of the driving force of the reactor coolant pumps 6a, 6b, 6c, and 6d. In FIG. 9, only the cold leg pipes 4a and 4c of the four cold leg pipes are illustrated, and reactor coolant pumps 6a and 6b of the four reactor coolant pumps are illustrated. The pressurizer 80 is connected to the hot leg pipe 5a by a surge piping 81. A passive residual heat removal system 60 (passive RHR) of the AP1000 includes a passive RHR heat exchanger 61. The passive RHR heat exchanger 61 is disposed so as to be submerged in refueling water 66 stored in an in-containment refueling water storage tank 8. The in-containment refueling water storage tank 8 is disposed below an operating deck 90. The passive RHR heat exchanger 61 is connected to the hot leg pipe 5a through a coolant supply piping 62. An inlet valve 63 is disposed in the middle of the coolant supply piping 62. The passive RHR heat exchanger 61 is connected to the cold leg pipe 4a around the outlet of the steam generator 3a through a coolant return piping 65. An outlet valve 64 is disposed in the middle of the coolant return piping 65. At the normal operation time, the inlet valve 63 is always opened, allowing coolant to be always supplied to the passive RHR heat exchanger 61 through the coolant supply piping 62. Further, at the normal operation time, the outlet valve 64 is always closed. During the normal operation time of the plant, the outlet valve 64 is closed, preventing the coolant in the passive RHR heat exchanger 61 from passing inside the cold leg pipe 4a for circulation into the reactor vessel 2. However, when water feed to the steam generators 3a and 3b is stopped due to occurrence of a transient such as loss of offsite electric power or loss of feedwater, the outlet valve 64 is automatically opened by a low water-level signal of the steam generators 3a and 3b. As a result, the primary coolant in the passive RHR heat exchanger 61 passes through the coolant return piping 65 and the cold leg pipe 4a to be circulated into the reactor vessel 2. The drive source for the above circulation is the natural circulation force of the primary coolant given by the decay heat generated in the reactor core 1, and an active drive sources such as pumps are not required for the natural circulation in this configuration. In the case where a steam generator tube rupture accident (SGTR) occurs, the primary coolant outflows from a ruptured location, and the outlet valve 64 of the passive RHR is automatically opened by a low water-level signal of the pressurizer 80. As a result, the primary coolant in the passive RHR heat exchanger 61 passes through the coolant return piping 65 and the cold leg pipe 4a to be circulated into the reactor vessel 2. However, the depressurization of the primary system by the passive RHR is slow and, actually, the emergency core cooling system (ECCS) is automatically activated at the same time to inject cooling water into the reactor vessel 2 for rapid depressurization. Because decay heat removal after the depressurization is performed by the passive RHR smoothly, the ECCS is manually stopped by an operator and the accident is terminated. Actually, when a steam generator tube rupture accident (SGTR) occurs, a normal operating chemical and volume control system functions to make up for reduction of the water level in the pressurizer 80. This adversely causes a delay of generation of the pressurizer low water-level signal, which may result in an increase in the outflow of the primary coolant to the secondary system. Actually, occurrence of a steam generator tube rupture accident is obvious from rise of secondary system pressure and water level in the ruptured steam generator and thus it is expected that manual depressurization of the primary system can be performed by an operator at an early stage. However, at this stage, the outflow of the primary coolant is so small that the operator cannot inject ECCS water into the reactor vessel 2 to depressurize the primary system. Further, the depressurization of the primary system using the passive RHR is slow because it only removes the decay heat. Therefore, the operator uses the intact steam generator having a higher cooling and depressurization function to perform the primary system depressurization. The steam generator has heat removal capacity as about 50 times as the decay heat, and thus the primary system depressurization by the intact steam generator is achieved at high-speed. As a result, actually, the accident can be terminated at an earlier stage. As described above, in the conventional AP1000, it has been necessary to provide two steam generators for terminating the steam generator tube rupture accident at an earlier stage. In the AP1000, an automatic depressurization system (ADS) is provided for the purpose of achieving the primary system depressurization at high speed upon occurrence of a loss-of-coolant accident (LOCA) and a station blackout (SBO). The automatic depressurization system has four stages: first to third stages 51 to 53 and a fourth stage 68. The first to third stages 51 to 53 are disposed on the pressurizer 80. The fourth stage 68 of automatic depressurization system is disposed at the same location as the branch position of the coolant supply piping 62 connected to the hot leg pipe 5a. Once the automatic depressurization system starts operating, all the stages up to the final fourth stage 68 operate. When the final fourth stage 68 operates, the containment vessel 7 is submerged up to the position of the cold leg pipes 4a and 4b, leading to damage of plant property. As a result the plant cannot be restarted for a long period. In a steam generator tube rupture accident (SGTR), a damaged location is limited only to the inside of the steam generator although it is an accident. Thus, simply by repairing the tube 92 of the steam generator or replacing it with new one, it is possible to restart the plant in a short period. Therefore, for a steam generator tube rupture accident, it is not allowed to use the automatic pressurization system to depressurize the primary system. It is intended to avoid ADS actuation not only for safety but also property protection both in the primary system depressurization using the intact steam generator by the operator and safety systems of the ECCS and the passive RHR. Along with global warming and increase in crude oil price, expectations for nuclear power generation plant have increased recently on a global basis. In countries with economic power, construction rush of large nuclear reactors of 1000 MWe class or more is about to start. On the other hand, in developing countries, there is a stronger need for small nuclear reactors of 500 MWe or less in terms of relationship between power demand and the scale of a power network corresponding to the power demand. This trend may increase in the future. However, the small nuclear plants are economically inefficient for their scale disadvantage in terms of unit construction cost. Further, unlike the large nuclear reactors, the small nuclear plants have unique designs so as to make it difficult to prove such unique elemental technologies. Further, siting conditions are often worse than those in the economic powers, so that it is necessary to dispose higher safety than that for the large nuclear plants built in the economic powers. Under the circumstances, demanded is a small PWR capable of increasing economic efficiency by simplification, enhancing safety by a passive safety system, and ensuring reliability by proven elemental technologies common with large nuclear reactors. The minimum number of loops in the conventional PWR was set to two. However, as the plant output power is increased, the number of loops is increased to three and four, and the structure of the primary system becomes complicated. The AP1000 incorporates simplification by the passive safety system has also a two-loop structure. For further simplification, it is desirable to reduce the number of loops to one. In this case, however, the AP1000 needs to be configured to cope with each of the following events with only one reactor coolant loop: one-reactor coolant pump trip, all-pump trip accident, pump seizer accident, loss-of-coolant accident (LOCA), and steam generator tube rupture accident (SGTR). Further, in the AP1000, although the containment vessel can passively be cooled by a passive containment cooling system (PCS), it is necessary to replenish cooling water after three days. In the worldwide view, as to the siting conditions of the small nuclear reactors, there exist areas where the small nuclear reactors need to be constructed at sites in the inner portions of a continent and along a river. In the entire operating period, e.g., 60 years, of the plant, shortage of river water can be anticipated to occur. Thus, to cope with such problems with siting conditions, it is necessary to provide a small nuclear reactor provided with a passive containment cooling system capable of ensuring safety of the nuclear reactor without necessity of replenishing cooling water in case of an accident. Further, more severe natural conditions can be anticipated worldwidely as the siting conditions of the small nuclear reactors. Examples of these include giant cyclones in South-East Asia, the massive earthquake that occurred in Sichuan province of China, and the big Tsunami in the Indian Ocean. Occurrence of a station blackout (SBO) due to a sever natural disaster such as a giant cyclone may prevent a recovery work from being started for a long period of time. The cases of Hurricane Katrina in the United States and the giant cyclone in Myanmar suggest the possibility of such a situation. Similarly, the cases of the massive earthquake in Sichuan province of China and the big Tsunami in the Indian Ocean suggest the possibility of such a situation. Thus, it is necessary to provide a small nuclear reactor capable of performing cooling of the reactor core and containment vessel in a continuous manner even when such a prolonged station blackout occurs. To this end, it is necessary to provide a small nuclear reactor capable of naturally ensuring safety forever without supporting actions such as accident management even if the station blackout continues forever. In the case of a small nuclear reactor of 500 MWe class or less, thorough simplification needs to be conducted so as to overcome the scale disadvantage. This thorough simplification results in adoption of peculiar and less proven new elemental technologies which is possible only in the individual small nuclear reactor. Most of these new elemental technologies have not been adopted at all and will never be adopted in the future in large nuclear reactors that will surely be constructed. Thus, it is impossible to remove risk of occurrence of defect if a small reactor based on such a new peculiar technology is actually constructed and operated. An object of the present invention is therefore to use proven elemental technologies and device components of large PWRs or those of large PWRs that will surely be constructed in the future so as to remove the risk associated with new construction and thereby to realize a pressurized water reactor plant which is more reliable, better proved, more simplified, and improved in passive safety. According to one embodiment, a pressurized water reactor plant is provided. The pressurized water reactor plant includes a single reactor vessel housing a reactor core cooled by a high-pressure primary coolant and a single steam generator for generating steam of a secondary coolant. The steam generator has a plurality of tubes in which the primary coolant heated in the reactor core and discharged outside the reactor vessel is circulated, and one barrel portion housing the plurality of tubes and configured so as to guide the secondary coolant, which is lower in pressure than the primary coolant and higher in pressure than atmospheric pressure around the tubes. The pressurized water reactor further includes a hot leg pipe that guides the primary coolant heated in the reactor core from the reactor vessel to the tubes of the steam generator, at least two mutually parallel cold leg pipes for returning the primary coolant that has been passed through the plurality of tubes of the steam generator to the reactor vessel, at least two reactor coolant pumps that feed the primary coolant in the at least two cold leg pipes to the reactor vessel, a pressurizer provided so as to communicate with a reactor pressure boundary in which the primary coolant flows, having a liquid surface therein, and that pressurizes the reactor pressure boundary, a containment vessel containing the reactor vessel, the steam generator, the hot leg pipe, the at least two cold leg pipes, the reactor coolant pumps, and the pressurizer; and a primary system depressurization device for equalizing a pressure of a primary system where the primary coolant flows to a pressure of a secondary system where the secondary coolant flows when an accident has occurred in which a part of the plurality of tubes in the steam generator is ruptured. The present invention allows pressurized water reactor plants to benefit from reduced risks by employing elemental technologies and device components from large pressurized water reactor plants, including those yet to be constructed. Thus, pressurized water reactor plants which are more reliable, better-proven and have greater passive safety can be realized. Furthermore, the risks associated with new construction can be mitigated, and greater simplicity can be attained. According to an embodiment, in order to achieve the object, there is provided a pressurized water reactor plant comprising: a single reactor vessel housing a reactor core cooled by high-pressure primary coolant; a single steam generator for generating steam of secondary coolant; the steam generator having: a plurality of tubes in which the primary coolant heated in the reactor core and discharged outside the reactor vessel is circulated, and one barrel portion housing the plurality of tubes and configured so as to guide the secondary coolant lower in pressure than the primary coolant and higher in pressure than atmosphere pressure around the tubes; a hot leg pipe that guides the primary coolant heated in the reactor core from the reactor vessel to the tubes of the steam generator; at least two mutually parallel cold leg pipes for returning the primary coolant that has been passed through the tubes of the steam generator to the reactor vessel; at least two reactor coolant pumps that feeds the primary coolant in the at least two cold leg pipes to the reactor vessel; a pressurizer provided so as to communicate with a reactor pressure boundary in which the primary coolant flows, having a liquid surface therein, and pressurizing the reactor pressure boundary; a containment vessel containing the reactor vessel, the steam generator, the hot leg pipe, the cold leg pipes, the reactor coolant pumps, and the pressurizer; and a primary system depressurization means for equalizing pressure of a primary system where the primary coolant flows to pressure of a secondary system where the secondary coolant flows, when an accident has occurred in which a part of the tubes in the steam generator is ruptured. A first embodiment of a pressurized water reactor plant according to the present invention will be described with reference to FIGS. 1 to 3. The same reference numerals are assigned to the same or similar parts as those in the conventional example, and redundant descriptions are omitted. Further, even if only one valve is illustrated as representative of various types of valves in each of the following drawings for simplification, a plurality of valves may be actually arranged in parallel or in series for ensuring of reliability. FIG. 1 is a plan view illustrating a configuration of a reactor coolant system loop in the first embodiment of the pressurized water reactor plant according to the present invention. FIG. 2 is a block system diagram of a reactor pressure boundary and a passive cooling and depressurization system in the first embodiment of the pressurized water reactor plant according to the present invention. FIG. 3 is an elevation cross-sectional view of a containment vessel and its internal components in the first embodiment of the pressurized water reactor plant according to the present invention. The pressurized water reactor plant according to the present invention has a reactor core 1 and a reactor vessel 2 housing the reactor core 1. The reactor vessel 2 is connected to one steam generator 3 by two cold leg pipes 4a, 4b and one hot leg pipe 5. The positions of the two cold leg pipes 4a, 4b and connection nozzles of the reactor vessel 2 are not limited to those illustrated in FIG. 1 but may be arbitrarily determined. For example, the two cold leg pipes 4a, 4b and connection nozzles may be arranged so as to face each other at 180 degrees intervals, respectively. Two reactor coolant pumps 6a and 6b for circulating primary coolant in the reactor core 1 and the steam generator 3 are directly connected to the lower portion of the steam generator 3. These primary system components are disposed inside a containment vessel 7 (refer to FIG. 3). The configuration of the steam generator 3 is the same as that of the conventional steam generator 3a illustrated in FIGS. 6 and 7. The present embodiment employs a configuration in which only one steam generator 3 is disposed in the plant. The plant output power is about 550 MWe, and the pressurized water reactor plant of the present embodiment is of a small type. At the normal operation time of the pressurized water reactor plant according to the present invention, primary coolant is heated by heat generated in the reactor core 1, and the heated high-temperature primary coolant is supplied from the reactor vessel 2 to tubes 92 of the steam generator 3 through the hot leg pipe 5. Where, the heat of the primary coolant is transmitted to secondary coolant flowing in a secondary side 95 (outside of the tubes 92) of a barrel portion 22 of the steam generator 3, and steam of the secondary coolant is generated. The generated steam of the secondary coolant is fed to a steam turbine (not illustrated) outside the containment vessel 7 and used for power generation. The primary coolant whose heat has been transmitted to the secondary coolant in the tubes 92 is boosted by the reactor coolant pumps 6a and 6b, led through the cold leg pipes 4a and 4b, and returned to the reactor vessel 2. The primary coolant is higher in pressure than the pressure of the secondary coolant, and both the primary coolant and secondary coolant are higher in pressure than atmospheric pressure. The reactor coolant pumps 6a and 6b are motor-driven pumps, and their electric power sources are connected to different buses (not illustrated). Therefore, in a single bus failure, the two pumps do not lose their electric power simultaneously. The buses are connected to a generator (not illustrated) during the normal operation of the plant. When a generator trip occurs, the buses are supplied from the offsite power, and the pumps continue to operate. Still, it is anticipated that one-pump trip transient in which one reactor coolant pump is stopped occur due to a single bus failure. Even in this case, the remaining one reactor coolant pump continues to operate and provides required coolant flow, thereby ensuring integrity and reusability of the core fuel. In the case where a whole station bus failure and loss of offsite power occur simultaneously to cause a double pump trip accident in which two reactor coolant pumps lose their electric power simultaneously, the reactor coolant pumps 6a and 6b coast down according to their inertia to allow the coolant flow rate required for cooling of the core fuel at the accident to be maintained for a certain amount of time period. In the case where the pump seizer accident where the shaft of one reactor coolant pump locks occurs, it is necessary to assume loss of offsite power according to safety criteria. In this case, another reactor coolant pump coasts down according to its inertia to secure cooling of the core fuel and to prevent overpressure of the reactor pressure boundary after the accident. Two direct vessel injection nozzles (DVI nozzles) 58a and 58b are disposed in the reactor vessel 2. To the direct vessel injection nozzles 58a and 58b, injection pipes of two separate emergency core cooling systems (not illustrated) are connected. Thus, even when the loss-of-coolant accident where, e.g., the cold leg pipe is ruptured, cooling water can directly be injected into the reactor vessel 2 by the two separate emergency core cooling systems. As illustrated in FIG. 2, the reactor coolant pumps 6a and 6b are directly connected to the lower portion of the steam generator 3. In FIG. 2, only the reactor coolant pump 6a is illustrated. Although the cold leg pipe 4b is depicted to extend in the opposite direction to the cold leg pipe 4a as an example, the other end of the cold leg pipe 4b is connected to the reactor coolant pump 6b. A pressurizer 80 is connected to the hot leg pipe 5 by a surge piping 81. A sensor 21 for detecting the steam generator tube rupture accident (SGTR) is attached to the secondary side 95 of the steam generator 3. The sensor 21 is preferably, e.g., a pressure gauge for detecting the pressure of the secondary side of the steam generator 3, a water level gauge for detecting the water level of the secondary side of the steam generator 3, a radioactivity detector for detecting the radioactivity level of the secondary side of the steam generator 3, or combination thereof. In this case, based on one or more of signals indicating the pressure, water level, and radioactivity level of the secondary side 95 of the steam generator 3, leakage of primary coolant to the secondary side 95 due to occurrence of the steam generator tube rupture accident (SGTR) can be detected. A primary system depressurization device in the present embodiment at the time of occurrence of the steam generator tube rupture accident (SGTR) includes a passive residual heat removal system (passive RHR) 60 and a passive cooling and depressurization system (PCDS) 30. The passive residual heat removal system (passive RHR) 60 of the present embodiment includes an in-containment refueling water storage tank 8, a passive residual heat removal system heat exchanger (passive RHR heat exchanger) 61, a coolant supply piping 62, an inlet valve 63, a coolant return piping 65, and an outlet valve 64. The passive RHR heat exchanger 61 is disposed inside the in-containment refueling water storage tank 8. During normal operation of the plant, the inlet valve 63 is always opened, allowing coolant to be always supplied to the passive RHR heat exchanger 61 through the coolant supply piping 62. Further, during normal operation, the outlet valve 64 is always closed. In the case where the steam generator tube rupture accident (SGTR) occurs, the primary coolant outflows from a ruptured location, and the outlet valve 64 of the passive RHR is automatically opened by a low water-level signal of the pressurizer 80. As a result, the primary coolant in the passive RHR heat exchanger 61 passes through the coolant return piping 65 and cold leg pipe 4a to be circulated into the reactor vessel 2. The low water-level signal of the pressurizer 80 also activates the emergency core cooling system (ECCS) to inject cooling water into the reactor vessel 2 for rapid depressurization. Decay heat removal after the depressurization is smoothly performed by the passive RHR, so that the ECCS is manually stopped by an operator for termination of the accident. Actually, upon occurrence of the steam generator tube rupture accident (SGTR), a normally operating chemical and volume control system (not illustrated) makes up for reduction in the water level of the pressurizer. This adversely causes a delay of generation of the pressurizer low water-level signal, which may result in an increase in the outflow of the primary coolant to the secondary system and environment. In this case, occurrence of the steam generator tube rupture accident (SGTR) is obvious from rapid rise of secondary system pressure and water level of the rupture-side steam generator 3 and thus it is expected that manual depressurization of the primary system can be performed by an operator at an earlier stage. In this case, although the depressurization of the primary system by the passive RHR 60 is slow, the operator can perform the primary system depressurization by using the passive cooling and depressurization system 30 having a larger capacity for primary system depressurization function. The passive cooling and depressurization system (PCDS) 30 includes a passive cooling and depressurization system pool (PCDS pool) 35 and a passive cooling and depressurization system heat exchanger (PCDS heat exchanger) 37. PCDS pool water 36 is stored in the PCDS pool 35. The PCDS heat exchanger 37 is disposed inside the PCDS pool 35. That is, the PCDS pool 35 is a cooling water pool capable of storing cooling water used for heat exchange in the PCDS heat exchanger 37. A steam supply piping 32 extends from the steam phase filled with saturated steam 83 of the pressurizer 80 to the PCDS heat exchanger 37. Parallel branched pipings are formed in the middle of the steam supply piping 32, and a steam supply valve 33 and a depressurization valve 34 are disposed in the middle of each of the parallel branched pipings. A condensate return piping 45 extends from the PCDS heat exchanger 37 to the cold leg pipe 4b constituting the reactor pressure boundary. An injection valve 46 is disposed in the middle of the condensate return piping 45. A steam discharge piping 47 has one end connected to the gas phase of the PCDS pool 35 and the other end opened to the ambient air. A steam discharge isolation valve 48 is disposed in the middle of the steam discharge piping 47. The present embodiment is so designed as to guide saturated steam 83 in the pressurizer 80 to the PCDS heat exchanger 37, so that the installation position of the PCDS heat exchanger 37 is not restricted by the position of the pressurizer 80. Since steam is a gas, it exhibits a significant higher flow mobility, and the steam easily ascends upward, exceeding the potential energy, by using the differential pressure between the pressurizer 80 and the PCDS heat exchanger 37 as a drive force. Thus, the PCDS pool 35 can be located above an operating deck 90. The vertical interval between the PCDS heat exchanger 37 and the reactor core 1 can be set as large as about 20 m, so that condensate can smoothly be guided to the reactor core 1 by the potential energy. The steam fed to the PCDS heat exchanger 37 is cooled by the external PCDS pool water 36 to be condensed, so that the inside of the PCDS heat exchanger 37 is always maintained in a depressurized state, that is, maintained in a lower pressure state than the pressure of the pressurizer 80. As a result, the saturated steam 83 in the pressurizer 80 can smoothly be guided to the PCDS heat exchanger 37. This enables higher speed primary system depressurization than the passive RHR 60. At the time of occurrence of the steam generator tube rupture accident (SGTR), the depressurization valve 34 and the injection valve 46 are rapidly opened by the operator's manual operation, thereby achieving the primary system depressurization at high speed. As a result, equalization between the primary system pressure and the secondary system pressure is performed, thereby rapidly stopping outflow of the primary coolant from ruptured steam generator tube. Only insignificant amount of the primary coolant outflows at the time of occurrence of the steam generator tube rupture accident (SGTR). Thus, the water level of the pressurizer is not lowered, and the ECCS is not activated. As described above, in the present embodiment, if the steam generator tube rupture accident (SGTR) occurs, it is possible to minimize or completely prevent the outflow of the primary coolant to the environment by using the passive cooling and depressurization system 30. Further, it is possible to avoid unnecessary activation of the emergency core cooling system (ECCS). The occurrence of the steam generator tube rupture accident (SGTR) can be detected not only based on the low water level signal of the pressurizer 80 as described above but also by using, e.g., the sensor 21 attached to the secondary side of the steam generator 3. When the occurrence of the steam generator tube rupture accident (SGTR) is detected in the manner as mentioned above, the passive cooling and depressurization system (PCDS) 30 can be activated automatically or manually by the operator. When the passive cooling and depressurization system (PCDS) 30 is activated at the time of occurrence of the steam generator tube rupture accident (SGTR), the saturated steam 83 in the pressurizer 80 is guided to the PCDS heat exchanger 37 to heat the PCDS pool water 36, resulting in generation of steam in the PCDS pool 35. The generated steam is passed through the steam discharge piping 47 and then discharged to the ambient air. The PCDS pool water 36 is clean water that does not include radioactivity at all, so that the discharge of the steam generated in the PCDS pool 35 to the ambient air does not adversely affect the environment. As described above, at the time of occurrence of the steam generator tube rupture accident (SGTR), the passive cooling and depressurization system (PCDS) 30 is activated to perform the primary system high-speed depressurization so as to prevent outflow of the primary coolant including radioactivity from the ruptured steam generator tube to the secondary system and environment and, instead, clean steam generated from clean water in the PCDS pool water 36 that does not include radioactivity at all is discharged to the environment. As a result, it is possible to ensure both safety of the nuclear reactor and public safety. Although not illustrated, a vent pipe is connected to the header portion of the heat exchanger through a vent valve so as to cope with the case where noncondensable gas such as nitrogen is accumulated inside the PCDS heat exchanger 37. The other end of the vent pipe is guided to, e.g., the inside of the in-containment refueling water storage tank 8. Further, a steam generator tube rupture accident detector (SGTR detector) that detects one or more of the pressure “high”, the water level “high”, and the radioactivity level “high” using the sensor 21 (pressure gauge, water level gauge, radioactivity detector, etc.) disposed in the secondary side 95 of the steam generator to generate a steam generator tube rupture accident (SGTR) occurrence signal is provided. Additionally, a passive cooling and depressurization system automatic activation device that automatically activates the passive cooling and depressurization system (PCDS) 30 in response to the steam generator tube rupture accident (SGTR) occurrence signal is also provided. With this configuration, it is possible to terminate the accident with rapidity and high reliability without relying on operator intervention, thereby minimizing adverse environmental impact. As illustrated in FIG. 2, an accumulator 84 is disposed outside the reactor vessel 2, and cooling water 85 is accumulated therein. High-pressure nitrogen gas is accumulated in the upper portion of the accumulator 84. The lower portion of the accumulator 84 is connected to the reactor vessel 2 by piping having an injection valve 87. Further, the lower portion of the in-containment refueling water storage tank 8 is connected to the reactor vessel 2 by piping having an injection valve 89. This configuration allows injection of the coolant into the reactor vessel 2 in a time of emergency. In FIG. 3, only the reactor coolant pump 6a of the two reactor coolant pumps 6a and 6b is illustrated. Further, for the sake of convenience of illustration, the cold leg pipe 4b in FIG. 3 is illustrated so as to extend in the opposite direction to the cold leg pipe 4a. These components (the reactor vessel 2, the steam generator 3, the pressurizer 80, the reactor coolant pumps 6a and 6b, etc.) and pipes (cold leg pipes 4a and 4b, hot leg pipe 5, etc.) constituting the reactor pressure boundary are all housed inside the containment vessel 7. The PCDS pool 35 of the passive cooling and depressurization system 30 in the present embodiment is disposed inside the containment vessel 7. The steam discharge piping 47 penetrates the side wall of the containment vessel 7 and opened to the ambient air. The containment vessel 7 of the present embodiment is the most typical containment vessel, called “large dry CV”, for use in PWRs. The containment vessel 7 is made of steel, because it is designed to be cooled with the external air in case of an accident. The in-containment refueling water storage tank 8 is disposed inside the containment vessel 7. The in-containment refueling water storage tank 8 functions as a gravity-driven cooling system ECCS and, cooperating with other passive ECCS, submerges the lower part of the containment vessel to a level above the cold leg pipes 4a and 4b. After that, it is designed so that a recirculation screen (not illustrated) is opened, introducing water always into the reactor vessel 2 to cool the fuel in the reactor core safely. If the water introduced into the reactor vessel 2 is heated by the decay heat of the fuel in the reactor core, steam is generated and the steam fills the gas phase of the containment vessel 7, resulting in a rise of the temperature and pressure in the containment vessel 7. A shield building 71 is built outside the containment vessel 7. A cooling water pool 72 of the passive containment cooling system is disposed on the top of the shield building 71. The cooling water pool 72 is filled with PCS pool water 73. In case of a loss-of-coolant accident, the PCS pool water 73 drains onto the containment vessel 7. Air flows into the shield building 71 through an external air inlet 74 and then a natural circulation force raises the air through the gap between an air baffle 75 and the wall of the containment vessel 7 until the air is discharged outside through a heated air discharge 76 formed at the top of the shield building 71. The drainage of the PCS pool water 73 and the natural convection of air serve to cool the containment vessel 7 in safety. A polar crane 18 is disposed in the upper portion of the containment vessel 7. In the present embodiment, the plant output power is as small as about 550 MWe, and the decay heat after accident is low, so that the drainage of the PCS pool water 73 is not essential for safety. Nevertheless, in order to maintain the internal pressure of the containment vessel after an accident at a lower level, a configuration allowing drainage of the PCS pool water 73 is adopted. The shield building 71 including its side wall and ceiling portions has a structure endurable against a large plane crash. In this way, the pressurized water reactor plant of the present embodiment can cool the reactor core 1 and the containment vessel 7 with an extremely high reliability only by the passive safety systems requiring no external AC power source. Further, the plant output power is as small as about 550 MWe and therefore the decay heat after accident is low. Thus, although the PCS pool water 73 dries up about seven days after the accident, the PCS pool water 73 need not be replenished afterward. That is, the cooling of the reactor core 1 and the containment vessel 7 can be achieved forever only by external cooling air. As a result, it is possible to construct the pressurized water reactor plant of the present embodiment even in an area, such as inner portions of a continent, where cooling water is difficult to be ensured at the accident time. In the case where a considerably severe natural phenomenon occurs to cause a prolonged station blackout, the reactor core 1 is cooled by the passive cooling and depressurization system 30 in the seven days after the occurrence of the SBO. The generated steam is discharged to the environment from the steam discharge piping 47, so that the containment vessel 7 is not heated. Thus, in this period of time, cooling of the containment vessel 7 is not required. After eight days after the SBO, the reactor core 1 is cooled by the passive RHR 60 only if the PCDS pool water 36 cannot be replenished. The containment vessel 7 heated by the steam generated during operation of the passive RHR 60 is cooled by the passive containment cooling system (PCS). The cooling of the containment vessel 7 by the passive containment cooling system (PCS) of the present embodiment can be continued forever only by external cooling air even after the PCS pool water 73 depletion. The condensate obtained by condensation of the steam in the containment vessel 7 flows back to the in-containment refueling water storage tank 8 and is used as the cooling water for the passive RHR 60 once again. Therefore, in the pressurized water reactor plant of the present embodiment, it is possible to ensure safety of the reactor core 1 and containment vessel 7 against an indefinite station blackout (SBO). A second embodiment of the pressurized water reactor plant according to the present invention will be described with reference to FIG. 4. The same reference numerals are assigned to the same or similar parts as those in the first embodiment, and redundant descriptions are omitted. FIG. 4 is an elevation cross-sectional view of a containment vessel and its internal components in a second embodiment of the pressurized water reactor plant according to the present invention. The configurations illustrated in FIGS. 1 and 2 are the same as those of the second embodiment. In the present embodiment, the PCDS pool 35 of the passive cooling and depressurization system 30 is disposed outside the containment vessel 7. The PCDS pool water 36 is stored in the PCDS pool 35. The PCDS heat exchanger 37 is submerged in the PCDS pool water 36. Since the PCDS pool 35 is disposed outside the containment vessel 7, it is possible to eliminate the need for the PCDS pool 35 to have the same degree of pressure resistance and air tightness as the containment vessel 7 and to prevent influence on the layout of components in the containment vessel 7. The shield building 71 is built outside the containment vessel 7. Air flows into the shield building 71 through an external air inlet 74 and then a natural circulation force raises the air through the gap between an air baffle 75 and the wall of the containment vessel 7 until the air is discharged outside through the heated air discharge 76 formed at the top of the shield building 71. Since the decay heat is low, the natural convection of air serves to cool the containment vessel 7 safely. The shield building 71 including its side wall and ceiling portions has a structure endurable against a large plane crash. A protective grating 79 is disposed inside the heated air discharge 76 to thereby ensure endurance against a large plane crash. In this way, the pressurized water reactor plant of the present embodiment can cool the reactor core 1 and containment vessel 7 with an extremely high reliability only by the passive safety systems requiring no external AC power source. Further, the plant output power is as small as about 550 MWe and therefore the decay heat after accident is low. Thus, the cooling of the reactor core 1 and containment vessel 7 can be achieved forever only by external cooling air. As a result, it is possible to construct the pressurized water reactor plant of the present embodiment even in an area, such as inner portions of a continent, where cooling water is difficult to be ensured at the accident time. Further, in the present embodiment, it is not necessary to provide the PCS cooling water pool 72 and PCS pool water 73 of the first embodiment (FIG. 3), which simplifies the structure of a ceiling portion 78 and reduces the weight thereof. Thus, the pressurized water reactor plant of the present embodiment is excellent in earthquake resistance and can be constructed at a site at which a major earthquake is likely to occur. In a case where a considerably severe natural phenomenon occurs to cause a prolonged station blackout, the reactor core 1 is cooled by the passive cooling and depressurization system 30 in seven days after the SBO. The generated steam is discharged to the environment from the steam discharge piping 47, so that the containment vessel 7 is not heated. Thus, in this period of time, cooling of the containment vessel 7 is not required. After eight days after the accident, the reactor core 1 is cooled by the passive residual heat removal system (passive RHR) 60 only if the PCDS pool water 36 cannot be replenished. The containment vessel 7 heated by the steam generated during operation of the passive RHR 60 is cooled by the passive containment cooling system (PCS) using external cooling air. The cooling of the containment vessel 7 by the passive containment cooling system (PCS) can be continued forever. The condensate obtained by condensation of the steam in the containment vessel 7 flows back to the in-containment refueling water storage tank 8 and is used as the cooling water for the passive RHR 60 once again. Therefore, in the pressurized water reactor plant of the present embodiment, it is possible to ensure safety of the reactor core 1 and containment vessel 7 against an indefinite station blackout. The configuration of the first embodiment in which the PCDS pool 35 and the like are disposed inside the containment vessel 7 and the configuration of the second embodiment in which not the PCS cooling water pool 72 but the protective grating 79 is disposed in the upper portion of the containment vessel 7 may be combined. Conversely, the configuration of the second embodiment in which the PCDS pool 35 and the like are disposed outside the containment vessel 7 and configuration of the first embodiment in which the PCDS coolant pool 72 is disposed in the upper portion of the containment vessel 7 may be combined. While certain embodiments have been described, these embodiments have been presented by way of example only, and are not intended to limit the scope of the inventions. Indeed, the novel embodiments described herein may be embodied in a variety of other forms; furthermore, various omissions, substitutions and changes in the form of the embodiments described herein may be made without departing from the spirit of the inventions. The accompanying claims and their equivalents are intended to cover such forms or modifications as would fall within the scope and spirit of the inventions. |
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048254550 | claims | 1. A device for generating collimated X-ray beams, said device comprising: an X-ray source for emitting an input X-ray beam; a cone arranged in the input X-ray beam to receive the entire input X-ray beam, said cone blocking a portion of the input X-ray beam such that a limited output X-ray beam emerges from the cone, said output X-ray beam being smaller than the input X-ray beam; a diaphragm arranged in the cone for blocking a portion of the input X-ray beam such that the output X-ray beam is divided into first and second separate portions, said diaphragm being pivotable about a pivot axis such that the first portion of the output X-ray beam does not change when the diaphragm is pivoted, and the second portion of the output X-ray beam changes when the diaphragm is pivoted. 2. A device as claimed in claim 1, characterized in that the pivot axis is at an edge of the diaphragm. 3. A device as claimed in claim 2, characterized in that the diaphragm is flat. 4. A device as claimed in claim 3, characterized in that the diaphragm has an L-shaped cross-section in a plane perpendicular to the pivot axis. 5. A device as claimed in claim 3, characterized in that the diaphragm has an edge opposite the pivot axis which edge is parallel to the pivot axis. 6. A device as claimed in claim 3, characterized in that the diaphragm has an edge opposite the pivot axis, which edge is concave. 7. A device as claimed in claim 1, characterized in that the pivot axis defines an edge of the first portion of the output X-ray beam. |
claims | 1. An emissions calculation system for calculating emissions resulting from transporting a shipment from an origin address to a destination address through a carrier's transportation network, the emissions calculation system comprising at least one computer processor and a memory, said at least one computer processor configured for:receiving at least a portion of shipment parameters associated with a particular shipment, the received shipment parameters comprising an origin address and a destination address for the particular shipment;identifying a transportation route along which the particular shipment is expected to travel from the origin address to the destination address;retrieving one or more emissions factors related to at least one of the shipment parameters, each of the one or more emissions factors being based at least in part on an amount of fuel used during a particular time period by the carrier to transport previously shipped packages along at least a portion of the transportation route; andestimating an amount of emissions resulting from transporting the particular shipment along the at least a portion of the transportation route based at least in part on the one or more retrieved emissions factors and at least a portion of the shipment parameters. 2. The emissions calculation system of claim 1, wherein the at least one computer processor is further configured for estimating the amount of fuel used during the particular time period to transport packages along the at least a portion of the transportation route based at least in part on an expected amount of time for transporting the packages along the at least a portion of the transportation route, the expected amount of time based on historical shipment data of the carrier. 3. The emissions calculation system of claim 2, wherein the expected amount of time is associated with a distance range that includes a distance of the at least a portion of the transportation route, and the amount of fuel is further based on the number of packages transported within the distance range by the carrier during the particular time period. 4. The emissions calculation system of claim 3, wherein the expected amount of time is further based on a service level associated with the packages, the service level in part indicating the transportation route along which the packages travel, one or more modes of transportation for transporting the packages along the transportation route, the expected amount of time for transporting the packages, and an average volume per package transported via the service level. 5. The emissions calculation system of claim 2, wherein the transportation route comprises one or more route legs, and wherein the expected amount of time is associated with transporting packages along a particular route leg, and the amount of fuel is further based on the number of packages transported along the route leg by the carrier during the particular time period. 6. The emissions calculation system of claim 5, wherein the expected amount of time is further based on a service level associated with the packages, the service level in part indicating the transportation route along which the packages travel, a mode of transportation for transporting the packages along each route leg, the expected amount of time for transporting the packages along each route leg, and an average volume per package transported via the service level. 7. The emissions calculation system of claim 1, wherein:the transportation route comprises a first route leg from the origin address to a first carrier facility, a second route leg between the first carrier facility and a second carrier facility, and a third route leg from the second carrier facility to the destination address,a transport operational activity comprises transporting the particular shipment along the second route leg, and a transport emissions factor of the one or more retrieved emissions factors is associated with the transport operational activity,a pickup operational activity comprises transporting the particular shipment along the first route leg, and a pickup emissions factor of the one or more retrieved emissions factors is associated with the pickup operational activity, anda delivery operational activity comprises transporting the particular shipment along the second route leg, and a delivery emissions factor of the one or more retrieved emissions factors is associated with the delivery operational activity. 8. The emissions calculation system of claim 7, wherein the transport operational activity further comprises transporting the particular shipment through at least one intermediate carrier facility located between the first carrier facility and the second carrier facility. 9. The emissions calculation system of claim 7, wherein the particular shipment is associated with a first service product, and the pickup emissions factor is based at least in part on the amount of fuel per stop previously used by the carrier for picking up packages associated with the first service product. 10. The emissions calculation system of claim 7, wherein the particular shipment is associated with a first service product, and the delivery emissions factor is based at least in part on the amount of fuel per stop previously used by the carrier for delivering packages associated with the first service product. 11. The emissions calculation system of claim 7, wherein the one or more retrieved emissions factors further comprises a stationary operational activities emissions factor associated with one or more stationary operational activities. 12. The emissions calculation system of claim 1, wherein estimating the amount of emissions comprises estimating an amount of carbon dioxide. |
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047708174 | description | DESCRIPTION OF THE INVENTION This invention utilizes a non-particulate transition alumina. The alumina usually has a delta structure, though other crystalline structures could also be used as long as they are not alpha. The alumina has a porosity of at least 40%, and preferably the porosity exceeds 60%. Preparation of the preferred, high porosity alumina is described in U.S. Pat. No. 3,941,719, herein incorporated by reference. Additional descriptions of the preferred alumina can be found in an article by Bulent E. Yoldas entitled "A Transparent Porous Alumina," which appeared in The American Ceramic Soc. Bulletin Vol. 54, No. 3, March, 1975 pp. 286-289 and "Alumina Gels That Form Porous Transport Al.sub.2 O.sub.3," in The American Ceramic Society Bulletin, Vol. 54, No. 3, March, 1975, pp. 289-290, also herein incorporated by reference, and an article by Bulent E. Yoldas entitled "Alumina Sol Preparation from Alkoxides," which appeared in the Journal of Materials Science, Vol. 10, 1975, pp. 1856-1860. In the preferred alumina, all the porosity is open and consist of channels around 100.ANG. in diameter. Another unusual property of this Al.sub.2 O.sub.3 is that it goes under crystalline transformations at 1200.degree. C., during which the open structure non-destructively collapses to a dense and virtually pore free .alpha.-Al.sub.2 O.sub.3 . The alumina may be used as a solid block material but it is preferably prepared as gravel-sized pieces because the absorption into the alumina of the solution containing the dissolved or colloidal solids is faster when smaller pieces are used. A solution or colloid is prepared of the solid material one wishes to entrap in the alumina. The solids may be radioactive waste materials, poisons, corrosive substances, or other types of solids. If a colloid is prepared, the colloidal solids must be smaller than the pore sizes of the alumina. The solids in the solution or colloid should thermally decompose to insoluble stable components, such as to oxides, upon heating and should have a low vapor pressure below 1200.degree. C. so that they are not vaporized when the pores are sealed. The liquid used to form the solution or colloid must have a surface tension which is low enough to wet the alumina so that the liquid flows into the pores of the alumina. The liquid must also be capable of either dissolving the solid material or else of forming a colloid with it. It is preferable that the liquid be inexpensive and non-toxic to hold down material and processing costs. A liquid with a low heat of vaporization is also desirable to reduce the amount of energy needed to evaporate it. Suitable liquids include water, alcohols, and various organic solvents. Water is a desirable liquid because it is inexpensive and many solids are soluble in it. Alcohols to C.sub.4 are desirable because of their fluidity and wetting characteristics. The solution may be prepared at almost any concentration even though saturated solutions are desirable, melts of 100% waste products should be avoided as some shrinkage of the alumina is needed to seal its pores. This is usually not a problem, however, as most solutions become saturated at concentrations considerably below 100% and many solids decompose to give off gases at temperatures below the alumina transformation temperature, which reduces the volume of solids remaining. It is preferable to soak the alumina in the solution or colloid under vacuum in order to remove entrapped air from the alumina. It is also very helpful to heat the solution as this reduces the surface tension of the solution and produces a more rapid and complete penetration of the solution into the pores of the alumina. Once the pores of the alumina have been filled with the solution or colloid, the carrier liquid or solvent is evaporated by drying, leaving the waste material deposited in the Al.sub.2 O.sub.3 pores. The alumina is then further heated to its transformation temperature, which is usually about 1200.degree. to about 1250.degree. C., which converts the alumina to .alpha.-alumina with collapsing of the entire porosity. This shrinks the alumina and seals its pores, trapping the solid material inside the closed pores. Because the surface nucleation takes place throughout the matrix, the shrinkage does not result in the cracking of the alumina. The conversion to .alpha.-alumina is readily observed because the alumina goes from a translucient or transparent state to an opaque, white china color, and up to 20% shrinkage may occur. After the alumina has cooled, it is desirable to wash the surface to remove any solids which have not been trapped in the pores. These solids can then be added to the solution or colloid used in the next batch. The following examples further illustrate this invention. EXAMPLE In this example, 10 gram, one-piece samples of 64% porous .delta.-alumina prepared according to U.S. Pat. No. 3,941,719 were soaked overnight in aqueous saturated solutions of various salts. The samples were then removed from these solution, surface dried with a tissue, and heated to 130.degree. C. until dry. The samples were weighed, then heated to 1200.degree. C. for a few minutes until they changed from the translucient or transparent .delta.-alumina to the opaque .alpha.-alumina. The samples were then cooled, washed, dried, and weighed a second time. The following table gives the salts which were used and the percent weight gain of the alumina before and after conversion to .alpha.-alumina. ______________________________________ Weight Gain After Impregnating After Impregnation Conversion Sample Salt and Drying at 130.degree. C. to .alpha.-alumina ______________________________________ 1 NaNO.sub.3 26.8 6.6 2 NaOH 12.5 3.3 3 NaCl 22.6 7.1 4 Zn (C.sub.2 H.sub.3 O.sub.2).sub.2 14.8 4.3 5 CuSO.sub.4 23.9 4.9 6 KH.sub.2 PO.sub.4 41.0 34.0 ______________________________________ The solution containing the potassium dihydrogen phosphate had been heated to about 50.degree. C. during impregnation which indicates that larger amounts of solids may be contained in the alumina if the solutions are heated. |
055641040 | abstract | This invention relates to the processing of liquid radioactive waste containing radioactively labeled biological molecules. More specifically, this invention relates to the use of solid phase binders to remove radioactively labeled biological molecules from liquid radioactive waste solutions. |
claims | 1. A method comprising:determining a neutron production coefficient of a controllably movable rod in a nuclear fission reactor, the controllably movable rod including fertile nuclear fission fuel material and neutron absorbing material;determining a neutron absorption coefficient of the controllably movable rod in the nuclear fission reactor;comparing a first combination of the determined neutron production coefficient and the determined neutron absorption coefficient with a target;based on the comparing of the first combination, determining a first application of the controllably movable rod to be a nuclear fission fuel rod when the combination is at least the target;again determining the neutron production coefficient of the controllably movable rod in the nuclear fission reactor after determining the first application of the controllably movable rod;again determining the neutron absorption coefficient of the controllably movable rod in the nuclear fission reactor after determining the first application of the controllably movable rod;comparing a second combination of the again-determined neutron production coefficient and the again-determined neutron absorption coefficient with the target; andbased on the comparing of the second combination, determining a second application of the controllably movable rod to be a reactivity control rod when the combination is less than the target. 2. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod are based on neutron exposure history of the controllably movable rod. 3. The method of claim 1, before comparing the first combination and before comparing the second combination, the method further comprising:yet again determining the neutron production coefficient of the controllably movable rod in the nuclear fission reactor; andyet again determining the neutron absorption coefficient of the controllably movable rod in the nuclear fission reactor. 4. The method of claim 3, further comprising:comparing a third combination of the yet-again-determined neutron production coefficient and the yet-again-determined neutron absorption coefficient with the target; andbased on the comparing of the third combination, determining a third application of the controllably movable rod to be a nuclear fission fuel rod when the combination is at least the target, and determining the third application to be a reactivity control rod when the combination is less than the target. 5. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod are based on a property of fertile nuclear fission fuel material of the controllably movable rod. 6. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod are based on a property of fissile nuclear fission fuel material of the controllably movable rod. 7. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod are based on a property of neutron absorbing poison of the controllably movable rod. 8. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod are based on a property of fission products of the controllably movable rod. 9. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod include monitoring at least one reactivity parameter of the controllably movable rod in the nuclear fission reactor. 10. The method of claim 1, wherein determining the neutron production coefficient, determining the neutron absorption coefficient, again determining the neutron production coefficient, and again determining the neutron absorption coefficient of the controllably movable rod includes predicting at least one of the neutron production coefficient, the neutron absorption coefficient, the again-determined neutron production coefficient, and the again-determined neutron absorption coefficient of the controllably movable rod in the nuclear fission reactor. 11. The method of claim 10, wherein predicting at least one of the neutron production coefficient, the neutron absorption coefficient, the again-determined neutron production coefficient, and the again-determined neutron absorption coefficient includes calculating at least one of the neutron production coefficient, the neutron absorption coefficient, the again-determined neutron production coefficient, and the again-determined neutron absorption coefficient of the controllably movable rod in the nuclear fission reactor. |
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description | Any and all applications for which a foreign or domestic priority claim is identified in the Application Data Sheet as filed with the present application are hereby incorporated by reference under 37 CFR 1.57. This application is a divisional of U.S. application Ser. No. 14/154,633, filed Jan. 14, 2014, which in turn claims priority to and the benefit of U.S. provisional application 61/753,851, entitled HETEROGENEOUS CORE DESIGNS AND THORIUM BASED FUELS FOR HEAVY WATER REACTORS, and filed Jan. 17, 2013. The entire contents of all priority applications are hereby incorporated by reference. The invention relates to core designs for thorium based fuels for heavy water reactors and more specifically to heterogeneous core designs for thorium based seed fuel and blanket fuel for channel-type heavy water reactors as well as thorium based fuel bundles for a heterogeneous core design. Research into the use of thorium as a new primary energy source has recently been explored. Thorium-232 (Th-232) is a naturally occurring isotope and is substantially more abundant than uranium. Although not fissile, upon absorbing a neutron will transmute to uranium-233 (U-233), which is an excellent fissile fuel material. Thorium fuel concepts therefore require that Th-232 is first irradiated in a reactor to provide the necessary neutron dosing. The U-233 that is produced can either be chemically separated from the parent thorium fuel and recycled into new fuel, or the U-233 may be usable in-situ in the same fuel form. Thorium fuels therefore require a fissile material as a driver so that a chain reaction (and thus supply of surplus neutrons) may be maintained. Fissile driver options are U-233, U-235 or Pu-239. It is possible, although difficult, to design thorium fuels that produce more U-233 in thermal reactors than the fissile material they consume (this is referred to as having a fissile conversion ratio of more than 1.0 and is also called breeding). Thermal breeding with thorium is possible using U-233 as the fissile driver, and to achieve this the neutron economy in the reactor has to be very good (i.e., low neutron loss through escape or parasitic absorption). The possibility to breed fissile material in slow neutron systems is a unique feature for thorium-based fuels. Another distinct option for using thorium is as a ‘fertile matrix’ for fuels containing transuranic elements such as plutonium. No new plutonium is produced from the thorium component, unlike for uranium fuels, and so the level of net consumption of this metal is rather high. In fresh thorium fuel, all of the fissions (thus power and neutrons) derive from the driver component. As the fuel operates the U-233 content gradually increases and it contributes more and more to the power output of the fuel. The ultimate energy output from U-233, and hence indirectly thorium, depends on numerous fuel design parameters, including: fuel burnup attained, fuel arrangement, neutron energy spectrum and neutron flux. The fission of a U-233 nucleus releases about the same amount of energy (200 MeV) as that of U-235. An important principle in the design of thorium fuel is that of fuel arrangements in which a high fissile (and therefore higher power) fuel zone referred to as the seed region is physically separated from the fertile (low or zero power) thorium part of the fuel referred to as the blanket region. Such an arrangement is far better for supplying surplus neutrons to thorium nuclei so they can convert to fissile U-233. Previous heavy water reactor core designs and associated fuel for channel-type heavy water reactors using thorium-based fuels have not been able to achieve simultaneously high fuel burnup, high fissile utilization and high conversion ratios, while also meeting design goals of high core-average power densities, meeting goals of operating limits on bundle power and maximum linear element ratings while keeping reactivity coefficients, such as for example coolant void reactivity, within desired values to enhance safety characteristics. Previous research in heavy water reactors have tended to focus on the design of homogeneous cores and heterogeneous fuel bundle designs that use neutron absorbing poisons to reduce void reactivity and has neglected to consider alternative design options. A thorium fuel based core design and/or a fuel bundle design that mitigates one or more various shortcomings is therefore in need. Thorium is an attractive fuel option to improve the sustainability of the nuclear fuel cycle, given the limited and unevenly distributed uranium reserves. As natural thorium does not contain a fissile isotope, implementation of thorium fuels in a reactor must involve a fissile component, generally either plutonium or uranium. The physical separation of a lower fissile blanket fuel and a higher fissile seed fuel into separate adjacent regions in a heterogeneous reactor core allows for the potential to improve the fissile utilization and increase the sustainability of the thorium fuel cycle. In one embodiment of the invention, there is provided a channel type heterogeneous reactor core for a heavy water reactor for burnup of thorium based fuel, the heterogeneous reactor core comprising at least one seed fuel channel region comprising seed fuel channels for receiving seed fuel bundles of thorium based fuel; and at least one blanket fuel channel region comprising blanket fuel channels for receiving blanket fuel bundles of thorium based fuel; wherein the seed fuel bundles have a higher percentage content of fissile fuel than the blanket fuel bundles. In an additional embodiment to that outlined above, the at least one seed fuel channel region and the at least one blanket fuel channel region are set out in a checkerboard pattern within the heterogeneous reactor core. In an additional embodiment to that outlined above, the at least one seed fuel channel region and the at least one blanket fuel channel region are set out in an annular pattern within the heterogeneous reactor core. In an additional embodiment to that outlined above, the seed fuel bundle comprises 35% or more UO2 and 65% or less ThO2. In an additional embodiment to that outlined above, the seed fuel bundle comprises 3% or more PuO2 and 97% or less ThO2. In an additional embodiment to that outlined above, the blanket fuel bundle comprises 30% or less UO2 and 70% or more ThO2. In an additional embodiment to that outlined above, the blanket fuel bundle comprises 2% or less PuO2 and 98% or more ThO2. In another embodiment of the invention, there is provided a fuel bundle for use in a channel type heterogeneous reactor core of a heavy water reactor, the fuel bundle comprising a central displacement tube; and a plurality of thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, the central displacement tube is filled with ZrO2, MgO, BeO, graphite or stagnant D2O coolant. In an additional embodiment to that outlined above, there are 21 radially positioned thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, there are 35 radially positioned thorium based fuel pins surrounding the central displacement tube. In an additional embodiment to that outlined above, the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 3% or higher. In an additional embodiment to that outlined above, the fuel bundle is a seed fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2+ThO2) with a UO2 content of 35% or higher. In an additional embodiment to that outlined above, the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (PuO2+ThO2) with a PuO2 content of 2% or less. In an additional embodiment to that outlined above, the fuel bundle is a blanket fuel bundle and the plurality of thorium based fuel pins comprises a homogeneous mixture of (UO2+ThO2) with a UO2 content of 30% or less. In an additional embodiment, the present invention provides for the use of a fuel bundle such as those embodiments outlined above in channel type heterogeneous reactor core of a heavy water reactor for burnup of thorium based fuel. A heterogeneous reactor core for a channel type heavy water reactor is provided. A channel-type heavy water reactor, similar to what is being currently used in nuclear power generation may be used as the initial basis for the design. The heterogeneous core comprises a lattice of channels for receiving seed or blanket fuel bundles in the channels as will be discussed below with reference to FIGS. 2-7. The core may contain from 25% to 84% seed fuel channels while the balance are blanket fuel channels. The nuclear fuel is in the form of short, (˜50 cm) or longer (>50 cm) fuel bundles made generally with one or two rings of fuel pins. It has been determined that to help minimize coolant void reactivity while maximizing fuel burnup and fissile utilization, the fuel bundle is designed to have only one or two rings of fuel pins, with a central displacer tube filled with stagnant coolant, or a solid moderator, for example graphite, or material with a low neutron scattering and low neutron absorption cross section, for example ZrO2 or MgO. The fuel bundles will be discussed in more detail below with reference to FIGS. 1A-1E. The nuclear fuel bundles are made from thorium, mixed with either plutonium or uranium, generally in oxide, carbine, silicide or a metallic-alloy form. As depicted in various non-limiting embodiments in FIGS. 2-7, an embodiment of the reactor core of the is a heterogeneous design with physically separate regions of seed fuel channels and blanket fuel channels arranged in a lattice. In FIGS. 2-7, seed channels are represented by an S and blanket channels are represented by a B. Seed fuel is made with higher concentrations of fissile fuel mixed with thorium and is used primarily to generate power and excess neutrons to drive blanket fuel. The blanket fuel is made with lower concentrations of fissile fuel mixed with thorium and used primarily to convert fertile thorium fuel into fissile fuel. There is some power generation by the blanket fuel. A seed channel, in one embodiment, is for seed bundles only while a blanket channel is for blanket bundles only. In the embodiments shown, each channel of the heterogeneous core has 12 bundles (either seed or blanket). As shown, the core may contain from 25% to 84% seed fuel channels, while the balance are blanket fuel channels. The core may have a lattice in a checkerboard-type arrangement of seed and blanket fuel channels such as those shown in FIGS. 6 and 7. Alternatively, the core may have a lattice in an annular arrangement of seed and blanket fuel channel regions with the outermost ring of the fuel channels adjacent to the radial reflector (not shown) of the core filled with blanket fuel channels such as those shown in FIGS. 2, 3, 4 and 5. It will be appreciated that there are several different permutations of heterogeneous seed/blanket core layouts which may be used or implemented and those shown in FIGS. 2 to 7 are not intended to be limited but rather illustrative of various embodiments of the concept of heterogeneous cores of the invention. The heterogeneous core allows for different and dynamic refueling strategies as the blanket fuel regions and the seed fuel regions can be refueled at different rates to achieve desirable burnup levels and core power distributions. Refueling strategies will be discussed in more detail below. The reactor core may be similar to current reactor cores such as the CANDU-6/EC-6 reactor which has 380 fuel channels with a square lattice pitch of 28.575 cm. Each channel thereof contains 12 fuel bundles, each approximately 50 cm long. Current CANDU cores use a homogeneous core of natural uranium (NU). Some more advanced designs use a single type of fuel and are still considered homogeneous. Shown in FIGS. 1A-1E are embodiments of fuel bundles for use in the channels of the heterogeneous core. As can be seen in the Figures, the fuel bundles include a central displacer tube to replace the central 8 fuel pins in a 43-element bundle, leaving outer rings of 14 and 21 fuel pins (FIGS. 1B and 1C). A further design, shown in FIGS. 1D and 1E includes a larger central displacer tube to replace the central 22 pins in a 43-element bundle leaving an outer ring of 21 fuel pins. Without wishing to be limited, the central displacer tube may be filled with ZrO2, MgO, BeO, graphite or stagnant D2O coolant. The purpose of the central displacer tube is to reduce coolant void reactivity (CVR). An advantage of the central displacer tube is that it helps to reduce the CVR, improving the safety characteristics of the lattice and the reactor during a postulated accident scenario, where there is a loss of coolant. The fuel pins of either the 21-element bundle or the 35-element bundle may be a combination of plutonium and thorium or low enriched uranium and thorium depending on whether the bundle is for use in a seed fuel region or blanket fuel region. In FIG. 1A, the inner 8 fuel pins are all the same, namely ThO2 and the outer 35 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2). In the fuel bundles shown in FIGS. 1B and 1C, the 35 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2) wherein the fuel bundle of FIG. 1B has central Zr-4 displacement tube filled with stagnant D2O coolant and the fuel bundle of FIG. 1C has a central Zr-4 displacement tube filled with ZrO2. In the fuel bundles shown in FIGS. 1D and 1E, the 21 fuel pins are all the same, namely a homogeneous mixture of (PuO2+ThO2) or (UO2+ThO2) wherein the fuel bundle of FIG. 1D has central Zr-4 displacement tube filled with stagnant D2O coolant and the fuel bundle of FIG. 1E has a central Zr-4 displacement tube filled with ZrO2. In the embodiments of fuel bundles wherein PuO2 is mixed with ThO2, the Pu is “reactor grade” Pu. In embodiments of fuel bundles wherein UO2 is mixed with ThO2, then the U is LEU (low enriched uranium), with a fissile content of about 5 wt % U-235/U in one non-limiting embodiment. The volume fraction of PuO2 in (Pu+Th)O2 may range from 1% to 13% in various non-limiting embodiments. The volume fraction of UO2 in (U+Th)O2 may range from 5% to 70% in various non-limiting embodiments. It will be appreciated that the mixture (volume fractions of either PuO2 or UO2 in (Pu+Th)O2 or (U+Th)O2) is dependent on whether the fuel is “seed” or “blanket” fuel. Seed fuel has a higher volume fraction of PuO2 or UO2 than blanket fuel. Typically, seed fuel contains fuel with 3% or higher PuO2 in (Pu,Th)O2, or 35% or higher UO2 in (U,Th)O2. The choice of LEU (in the non-limiting embodiment shown, 5 wt % U-235/U) for mixing with thorium (Th) is generally based on practical and economic considerations. 5 wt % U-235/U is readily available from existing enrichment facilities throughout the world as is therefore more commonly used. The choice of reactor grade Pu (generally about 0.67 wt % fissile Pu (Pu-239+Pu-241)) for mixing with Th is generally based on the assumption that most of the Pu inventory available in the world today is found in the spent fuel from light water reactors (LWRs). It is conceivable that one might use Pu from other sources, such as spent CANDU reactor natural uranium fuel, or Magnox reactor natural uranium fuel, or plutonium obtained from nuclear weapons stockpiles, or from a fast breeder reactor. In these other potential sources of plutonium, the fissile content will be different, probably higher. In principle, the plutonium from these alternative sources may be used in the heterogeneous reactor design as well, but given the assumption that the fissile plutonium content is higher, then the volume fraction of PuO2 in (Pu,Th)O2 would likely be lower to achieve the same level of burnup. Generally, a typical seed fuel will contain 35% UO2 (or more) and 65% ThO2 (or less), or it will contain 3% PuO2 (or more) and 97% (or less) of ThO2. Whereas a typical blanket fuel will contain 30% UO2 (or less) and 70% ThO2 (or more), or it will contain 2% PuO2 (or less) and 98% (or more) of ThO2. The fraction of the core's fuel channels that are seed channels can range from about 25% to about 84%. In most designs, the fraction is approximately 50% seed fuel channels and 50% blanket fuel channels as shown for example in FIGS. 2, 4 and 7. The core layout shown in FIG. 5 includes approximately 84% seed channels (320 channels) and 16% blanket channels (60 channels). An advantage of using more seed channels is that one can generate more power and achieve higher burnup while maintaining core reactivity. In addition, by using more seed the reactor may be operated at a higher power level, with a higher core-average power density. Typically, most of the previous CANDU core designs involving thorium based fuels have assumed a homogeneous core with one fuel type. The refuelling rates (and the core-average burnup of the fuel) depend on the choice of the fuel used (its initial enrichment), the desired radial and axial power distribution in the core, and the refuelling scheme. One refuelling scheme is a simple two-bundle shift, with bi-directional fuelling in alternating channels. Bundles are inserted from one side of the reactor, and are progressively moved to the other side until they reach the desired burnup. The objective in adjusting the exit burnup in each channel (and hence the refuelling rate) is to ensure that the maximum bundle power stays below ˜750 kW, and that the maximum channel power stays below ˜6,500 kW. However, it is also ideal to make the radial and axial power distribution as flat as possible, in order to maximize the power generated in the core, for economic advantage. The initial core designs used 35-element Pu/Th seed fuel that would achieve an approximate discharge burnup of 20 MWd/kg to 40 MWd/kg burnup. In most of the cases studied that meant using (3 wt % PuO2/97 wt % ThO2) for the seed to achieve a burnup of ˜20 MWd/kg. For core-average burnups closer to 40 MWd/kg, this means using (4 wt % PuO2/96 wt % ThO2). Most of the blanket fuel was either (2 wt % PuO2/98 wt % ThO2), burned to ˜20 MWd/kg, or (1 wt % PuO2/99 wt % ThO2) burned to 40 MWd/kg. Heterogeneous cores with LEU/Th fuel have not been tested yet, but they would use the same methods that were used in the analysis of the cores with Pu/Th fuel. There are two additional refuelling strategies to further improve the performance of the heterogeneous seed/blanket core, although these have not yet been tested: 1) To carry out axial shuffling of the fuel bundles in a given channel to help flatten the axial power distribution. This could be particularly useful in cores using seed fuel with higher levels of fissile enrichment (such as 5 wt % PuO2/95 wt % ThO2) and higher burnups (greater than 40 MWd/kg). The use of axial shuffling has been considered in the past by AECL in studies of CANDU reactor cores using SEU fuels (1.2 to 3 wt % U-235/U). 2) To send high enrichment, high-burnup seed fuel through a core twice or three times, somewhat analogous to what is done with batch refuelling in light water reactors. This is what would be called a 2TT (2 times through thorium) or 3TT (3 times through thorium) fuel cycle. For example, a seed fuel bundle which is estimated to have enough reactivity (and initial fissile content) to achieve a large discharge burnup will go through the CANDU core in three passes in three different channels. In addition, for example, a 35-element bundle might be made of (5 wt % PuO2/95 wt % ThO2) and lattice physics calculations indicate that it could achieve a final burnup of ˜54 MWd/kg. Instead of pushing the burnup of the fuel bundle from 0 to 54 MWd/kg in a single pass through the core, it can be divided up into two or three passes through the core. If divided into 3 passes, then the fuel would be burned from 0 to 18 MWd/kg in the first pass in one channel, 18 to 36 MWd/kg in the 2nd pass in another channel, and finally 36 to 54 MWd/kg in the third pass through another channel. A smaller change in the burnup between the inlet and exit of a given fuel channel will help flatten the axial power distribution, and permit a higher core power density, while staying within limits of peak bundle power and peak channel power. This type of refuelling scheme combines the on-line, bi-directional, continuous refuelling features of a CANDU reactor with the multi-batch zone refuelling schemes of a light water reactor (such as a PWR). Shown in FIGS. 8 to 16 are Tables 1 to 9 which set out geometry specifications and material specifications of the different fuel designs. Table 1 in FIG. 8 shows an embodiment wherein the reactor grade plutonium contains ˜52 wt % Pu-239 and ˜15 wt % Pu-241, giving a total fissile content of ˜67 wt % Pu-fissile/Pu. Table 2 in FIG. 9 shows the isotopic composition of LEU in oxide form. Thus, the fissile content is ˜5 wt % U-235/U and the balance of uranium is U-238 and U-234. Table 3 in FIG. 10 shows a description of different lattices tested. There are 10 different lattice designs, which are differentiated by geometry (5 geometry types) and fuel type (two fuel types, either (U,Th)O2 or (Pu,Th)O2) in the outer 35 or 21 pins. Only bundle designs 1 and 6 have 8 central ThO2 pins. All other pins are a mixture of either (U,Th)O2 or (Pu,Th)O2. Table 4 in FIG. 11 shows the dimensions of components for various lattices tested. The dimensions are given for a fuel pellet made of (Pu,Th)O2 or (U,Th)O2, or ThO2, the radius of the clad for the fuel element, the inner and outer radius for the central displacer tube, the inner and outer radius for the pressure tube (PT), the inner and outer radius for the calandria tube (CT). Table 5 in FIG. 12 shows the number of fuel pins and the pitch circle and radius, and the angular offset for the first fuel pin in the bundle. Note: bundle design 1a is the only one that has 4 rings of fuel pins (1+7+14+21). Bundle designs 1b and 1c do not have a central pin or an inner ring of fuel pins, only two outer rings of fuel pins (14+21). Bundle designs 1d and 1e have only a single outer ring of 21 fuel pins. Table 6 in FIG. 13 shows the material specifications for key components for various lattices tested. The type of material, its nominal operating temperature, and its nominal material mass density are given. The nominal purity of the heavy water moderator and the heavy water coolant are also specified. However, it should be pointed out that the purity of the heavy water in both the moderator and the coolant could be increased. Table 7 in FIG. 14 shows the value of the mass fractions for Pu-fissile (Pu-239+Pu-241) Pu, Th, and O in (Pu,Th)O2 for various volume fractions of PuO2 in (Pu,Th)O2. The fuels containing low volume fractions of PuO2 (e.g., 2% or less) are considered blanket fuel, while the fuels containing higher volume fractions of PuO2 (e.g. 3% or higher) are considered seed fuel. Also shown below is a sample set of core calculation results for two cores (1S-1B, and 84% Seed/16% Blanket) with different combinations of Seed and Blanket fuels. The data for the 1S-1B core design is shown in Table 8/FIG. 15. This shows the various performance characteristics of 5 different core designs, which differ in the type of seed and blanket fuel used. The data for the 84%-Seed/16% blanket core design is shown in Table 9/FIG. 16. This shows the various performance characteristics of 4 different core designs, which differ in the type of seed and blanket fuel used. The above described heterogeneous reactor core and fuel bundles are intended to be illustrative of the invention and are not intended to be limiting in any way. It will be appreciated that modifications and alterations to the design, function or use of the heterogeneous reactor core and fuel bundles may be made which are within the sphere of the invention contemplated and are within the scope of the claims. |
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claims | 1. A passive reactor cavity cooling system, comprising:a reactor cavity formed between a reactor vessel and a containment structure enclosing the reactor vessel; a first cooling system to control external air to sequentially pass through an air falling pipe and an air rising pipe provided in the reactor cavity, so that residual heat of a core transferred to the reactor cavity is discharged to an atmosphere or removed;a second cooling system having a water cooling pipe disposed in an inner space of the containment structure or in a wall of the containment structure to discharge the residual heat of the core transferred to the reactor cavity to outside the atmosphere or removed; anda functional conductor having an insulating property in a normal operation temperature range of a reactor and a heat transfer property in an accident occurrence temperature range of the reactor which is a higher temperature environment than the normal operation temperature range, wherein the air falling pipe and the water cooling pipe are disposed behind the air rising pipe with respect to a direction viewed from the reactor vessel, and the functional conductor is disposed between the air falling pipe and the air rising pipe,wherein the functional conductor comprises:a first plate disposed close to the air rising pipe;a second plate disposed close to the air falling pipe, wherein the first plate and the second plate are disposed to face each other at spaced positions, so that a fluid is filled between the first plate and the second plate;lattices disposed between the first plate and the second plate, and wherein a space to be filled with the fluid is formed by the first plate, the second plate, and the lattices; andwherein the first plate, the second plate, the fluid, and the lattices form a unit structure of the functional conductor, and the functional conductor is formed by an assembly of the unit structures, thereby the plurality of plates is sequentially disposed and spaced apart from one another in a manner of interposing the lattices therebetween, and the fluid is filled between the plurality of plates. 2. The system of claim 1, further comprising a water tank provided outside the containment structure, wherein the water cooling pipe comprises:a water falling part connected to the water tank; and a water rising part extending along an inner space of the air falling pipe or along inside of the wall of the containment structure, wherein the water falling part and the water rising part are connected through the containment structure. 3. The system of claim 1, wherein the first cooling system is formed along an inner circumferential surface of the containment structure and surrounds the reactor vessel at a position spaced apart from the reactor vessel. 4. The system of claim 1, wherein the reactor cavity is cooled by the first cooling system in a normal operation temperature range of the reactor, and wherein the reactor cavity is cooled by the second cooling system when a function of the first cooling system is lost in an accident occurrence temperature range of the reactor. 5. The system of claim 1, further comprising a water tank provided outside the containment structure, wherein the water cooling pipe is connected to a lower portion of the water tank. 6. The system of claim 1, wherein the water cooling pipe extends along an inner space of the air falling pipe through the containment structure or along the inside of the wall of the containment structure. 7. The system of claim 1, wherein the functional conductor has effective thermal conductivity in a range that water passing through the water cooling pipe is maintained in a liquid state in the normal operation temperature range of the reactor, and wherein the functional conductor has effective thermal conductivity in a range that the water passing through the water cooling pipe is boiled in the accident occurrence temperature range of the reactor. 8. The system of claim 1, wherein the fluid suppresses heat transfer between the first plate and the second plate in the normal operation temperature range of the reactor, and heat transfer through radiation is performed from one plate of the first and second plates to another plate of the first and second plates in the accident occurrence temperature range of the reactor. 9. The system of claim 1, wherein each of the first plate and the second plate is provided with a first surface and a second surface opposite to each other on each of the first plate and the second plate, so that heat conduction is performed from one surface of the first and the second surfaces to another surface of the first and second surfaces. 10. The system of claim 9, wherein thermal conductivity from one surface of the first and second surfaces to another surface of the first and second surfaces is 1W/m·K or 5 more. 11. The system of claim 1, wherein a surface of the first plate and a surface of the second plate have emissivity of 0.60 to 0.95. 12. The system of claim 1, wherein the first plate and the second plate are formed of a metal and have a thickness of 0.1 mm to 5 mm. 13. The system of claim 1, wherein the fluid contains at least one selected from a group including air, helium, nitrogen, and water. 14. The system of claim 1, wherein the lattices are formed of ceramic and have thermal conductivity of 0.1 W/m·K to 1.0 W/m·K. 15. The system of claim 1, wherein a length of each of the first plate and the second plate in a vertical direction is greater than a thickness of the lattice, and wherein a thickness ratio of the lattice to the vertical length of each of the first plate and the second plate is 0.2 or less. 16. The system of claim 1, wherein the unit structures are repeatedly arranged in a vertical direction. |
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claims | 1. A treatment method for a used ion exchange resin, comprising:bringing a used ion exchange resin into contact with a reaction solution, the used ion exchange resin having an ion exchange group with at least a radionuclide or a heavy metal element adsorbed thereon, and the reaction solution containing an iron compound, hydrogen peroxide, and ozone;separating at least a part of the reaction solution in contact with the used ion exchange resin from the used ion exchange resin; anddecomposing an organic component contained in the reaction solution separated from the used ion exchange resin. 2. The treatment method for the used ion exchange resin according to claim 1,wherein the reaction solution is added with an iron compound, hydrogen peroxide, and ozone during decomposing the organic component contained in the reaction solution. 3. The treatment method for the used ion exchange resin according to claim 1,wherein at least the part of the reaction solution in contact with the used ion exchange resin is separated from the used ion exchange resin after total organic carbon concentration in the reaction solution becomes 1.7 mol/L or more. 4. The treatment method for the used ion exchange resin according to claim 1,wherein at least the part of the reaction solution in contact with the used ion exchange resin is separated from the used ion exchange resin after concentration of the radionuclide or concentration of the heavy metal element in the reaction solution becomes a predetermined value or more. 5. The treatment method for the used ion exchange resin according to claim 1,wherein at least the part of the reaction solution in contact with the used ion exchange resin is separated from the used ion exchange resin after adding an aggregating agent that aggregates the used ion exchange resin to the reaction solution. 6. The treatment method for the used ion exchange resin according to claim 1, further comprising recovering the ozone during bringing the used ion exchange resin into contact with the reaction solution. 7. The treatment method for the used ion exchange resin according to claim 1,wherein:the used ion exchange resin is brought into contact with the reaction solution in a reaction tank, andthe organic component contained in the reaction solution is decomposed in a regeneration tank different from the reaction tank. |
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042467519 | abstract | 1. A nuclear engine and nozzle arrangement for a nuclear rocket, said arrangement comprising a cluster of elongated fissile fuel bearing and high temperature capacity modules suitably supported in a pressure vessel, said modules each having a plurality of coolant-propellant channels extending therethrough, a convergent-divergent nozzle structure of fixed cross-sectional dimensions secured to the end portion of each of said modules said modules, a divergent-only unitary skirt member connected directly to the propellant exit end of said modular cluster in series with and diverging from the divergent ends of said convergent-divergent nozzle structures, said modules being formed to conduct a compressible propellant therethrough at sub-sonic velocities, said nozzle structures being formed to develop supersonic velocities of the propellant and said divergent-only skirt being formed to develop further the supersonic velocities of said propellant. |
claims | 1. A spacecraft collision avoidance propulsion system for use in the presence of an ambient flux of cosmic rays, comprising:a supply of deuterium-containing particle fuel material;a store of pressurized propellant, with valves responsive to a received indication of a potential collision with a space object, for projecting the deuterium-containing particle fuel material in a specified direction outward from a spacecraft, the projected material interacting with the ambient flux of cosmic rays to generate products having kinetic energy,the spacecraft receiving at least some portion of the generated kinetic-energy-containing products to produce thrust upon the spacecraft to provide a change of trajectory of at least a dimension of the spacecraft to avoid the indicated potential collision. 2. The propulsion system as in claim 1, wherein the spacecraft is in an orbit around a planet or moon. 3. The propulsion system as in claim 1, wherein the space object is orbiting debris. 4. The propulsion system as in claim 1, wherein the space object is another spacecraft. 5. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material comprises Li6D. 6. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material comprises D2O. 7. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material comprises D2. 8. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material is in solid powder form. 9. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material is in pellet form. 10. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material is in frozen form. 11. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material is in liquid droplet form. 12. The propulsion system as in claim 1, wherein the deuterium-containing particle fuel material also contains up to 20% by weight of added non-fuel powder or dust particles. 13. A method of spacecraft propulsion for use in the presence of an ambient flux of cosmic rays for collision avoidance, comprising:projecting with pressurized propellant, through valves responsive to a received indication of a potential collision with a space object, deuterium-containing particle fuel material in a specified direction outward from a spacecraft, the projected material interacting with the ambient flux of cosmic rays to generate products having kinetic energy; andreceiving on the spacecraft at least some portion of the generated kinetic-energy-containing products to produce thrust upon the spacecraft to provide a change of trajectory of at least a dimension of the spacecraft to avoid the indicated potential collision. 14. The method as in claim 13, wherein the spacecraft is in an orbit around a planet or moon. 15. The method as in claim 13, wherein the space object is orbiting debris. 16. The method as in claim 13, wherein the space object is another spacecraft. 17. The method as in claim 13, wherein the deuterium-containing particle fuel material is Li6D. 18. The method as in claim 13, wherein the deuterium-containing particle fuel material is D2O. 19. The method as in claim 13, wherein the deuterium-containing particle fuel is D2. 20. The method as in claim 13, wherein the deuterium-containing particle fuel material is in solid powder form. 21. The method as in claim 13, wherein the deuterium-containing particle fuel material is in pellet form. 22. The method as in claim 13, wherein the deuterium-containing particle fuel material is in frozen form. 23. The method as in claim 13, wherein the deuterium-containing particle fuel material is in liquid droplet form. 24. The method as in claim 13, wherein the deuterium-containing particle fuel material also contains up to 20% by weight of added non-fuel powder or dust particles. |
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047175346 | summary | FIELD OF THE INVENTION The present invention relates to an improved nuclear fuel cladding and fuel element formed with such a cladding wherein the cladding contains a boron-containing burnable absorber integral therewith. BACKGROUND OF THE INVENTION It is well-known that the incorporation, in various manners, of a burnable absorber with nuclear fuel rods, which enables the use of excessive amounts of fuel in a reactor during the initial life of the fuel, can extend the life of the fuel rods. In some instances, the burnable absorber is mixed directly with the fuel and integrated therewith, while in other instances, a burnable absorber coating may be applied to the surface of fuel pellets, or discrete forms of a burnable absorber may be interspersed between conventional fuel pellets, or otherwise located within the cladding for the nuclear fuel. In U.S. Pat. No. 3,427,222, for example, a fuel rod is comprised of a tubular cladding that contains fuel pellets which have a fusion-bonded coating on the surface of each pellet, the coating comprised of a boron-containing material that functions as a burnable absorber. It has also been proposed to provide cladding materials such as zirconium-based alloys that have various coatings or barrier means on the inside wall of the tubular cladding to protect the cladding from attack by constituents released from the nuclear fuel during operation of a reactor containing the fuel rod. As examples of such coatings or barrier means which cover the full surface of the cladding, U.S. Pat. No. 4,022,662 describes a cladding in connection with a separate unattached metal liner comprised of stainless steel, copper, copper alloys, nickel or nickel alloys, the liner disposed between the cladding and the fuel material. A diffusion barrier of chromium or chromium alloy is also disposed between the cladding and the metal liner. In U.S. Pat. No. 4,045,288, a composite fuel element cladding is described which comprises a zirconium or zirconium alloy substrate having a metal barrier of 1-4 percent of the wall thickness formed from niobium, aluminum, copper, nickel, stainless steel and iron, and an inner layer of stainless steel, zirconium, or a zirconium alloy metallurgically bonded on the inner surface of the metal barrier. Also, it has been proposed to provide a burnable absorber, such as a boron-containing compound, directly in connection with the cladding material. U.S. Pat. No. 3,019,176, for example, discloses a fuel element that has a mixture of particulate fissionable material and a metal hydride moderator disposed in a matrix of a radiation resistant metal which is encased in a container. The matrix is bonded to the metal hydride and to the container to form an integral fuel element. U.S. Pat. No. 3,103,476 discloses the incorporation of a burnable absorber, such as boron, into the cladding of a nuclear fuel element. The boron is added to the cladding, which is preferably stainless steel, but may be zirconium or other material, in an amount of 200-1000 parts of natural boron per million parts of cladding material and homogeneously dispersed throughout the cladding. U.S. Pat. No. 3,625,821 describes a nuclear fuel element that has a zirconium or zircaloy cladding tube, with the inner surface of the tube coated with boron which is a burnable absorber. The boron is dispersed, as finely dispersed particles, in a matrix of nickel or other retaining metal. In Canadian Pat. No. 682,057, there is also described a fuel element where a cladding contains a burnable absorber. An outer layer of the tube may be of a corrosion resistant metal or alloy, and an inner layer is preferably comprised of the same material but contains, in addition, boron that may be a mixture or dispersed in the metal. The stated advantage over cladding containing dispersed boron is that boron, as part of the inner layer, is not contacted by coolant. It is an object of the present invention to provide a composite nuclear fuel cladding that has a burnable absorber integrally incorporated therein. SUMMARY OF THE INVENTION A nuclear fuel cladding having a burnable absorber integrally incorporated therein is formed as a hollow composite tube which has an outer tubular layer of a zirconium alloy, an intermediate layer, of a thickness less than the outer layer, of a zirconium alloy having admixed therewith a boron-containing burnable absorber, and an inner layer, of a thickness less than the intermediate layer, of zirconium metal. The layers of the composite are bonded together as an integral unit and is used in nuclear fuel rods. The zirconium alloys are preferably Zircaloys, while the preferred boron-containing burnable absorber is zirconium boride. Also, in the preferred embodiment, the outer tubular layer has a thickness of about 15 mils, the intermediate layer a thickness of about 3-5 mils, and the inner layer a thickness of about 1-2 mils. |
043127053 | summary | The invention relates to a spacer for fuel rods assembled into a fuel assembly for nuclear reactors formed of a grid structure of sheetmetal webs passing edgewise perpendicularly through one another and having resilient and rigid contact elements fastened thereto. A great number of constructions have become known heretofore for such spacers; in particular, reference may be had to German Pat. No. 1,589,051 wherein a so-called three-point support is proposed in which a fuel rod bears on one side against two rigid contact elements disposed one above the other and is pressed against them by a resilient contact element which engages the fuel rod on the opposite side thereof at half the height between the two rigid contact elements. Rigid and resilient contact elements are always thus located in the middle of the mesh walls of the spacer grid i.e. at the narrowest location between the fuel rod and the wall of the mesh, so that especially unfavorable cooling conditions prevail at this location. It has already been proposed heretofore also to place these contact elements in part at the intersection points of the spacer webs (note German Published Non-Prosecuted Application DE-OS Nos. 1 764 396 and 1 930 981). In these cases, the resilient contact elements, of themselves, form complete parts which are suspended into the web grid. The rigid contact elements, however, continue to remain at the narrowest location between the fuel rod and the wall of the mesh. This construction makes it possible to reduce the neutron absorption of such spacer grids with respect to previous constructions by making the web grid per se of a material having a low neutron absorption cross section, such as Zircaloy, for example, and only the spring contact elements per se of appropriately spring-hardened material, such as, Inconel, for example. The invention of the instant application is also based upon this principle. The two last-mentioned proposals for spacer construction provide one spring contact element and four rigid contact elements for each spacer mesh. The rigid contact elements are located in the zone of the narrowest coolant gap and reduce heat removal by the coolant at these locations. It is further possible that, with these constructions, the fuel rod can be pushed out of the centered position thereof if strong lateral forces should occur, such as in the event of an earthquake, for example, and the fuel rod will then directly contact the wall of the mesh which, due to the degradation of the heat removal connected therewith, can lead to clodding tube defects. It is accordingly an object of the invention to provide a spacer of the foregoing general type wherein cooling conditions are made more uniform over the entire circumference of the fuel rods at the height of the spacer grids and, at the same time, the pressure drop due to the latter is reduced. It is a further object of the invention to provide such a spacer wherein excursions of the fuel rods from the centered position thereof due to strong lateral forces are prevented as much as possible, while providing a construction of adequate strength with reduced material expenditure. With the foregoing and other objects in view, there is provided, in accordance with the invention, a spacer for fuel rods assembled into a fuel assembly for nuclear reactors formed of a grid structure of sheetmetal webs passing edgewise perpendicularly through one another comprising resilient and rigid contact elements fastened to the sheetmetal webs, each of the fuel rods being supportable by two diagonally opposing three-point contact systems, each of the systems being formed of one of the resilient contact elements and two of the rigid contact elements, the one resilient contact element being centrally disposed, as viewed in axial direction, and the two rigid contact elements being oppositely disposed and being ring-shaped. In accordance with another feature of the invention, the sheetmetal webs pass through one another at intersecting points and are formed with respective slots at the upper and lower edges thereof adjacent the intersecting points, the ring-shaped rigid contact elements being located at the intersecting points at the upper and lower edges of the sheetmetal webs and being engaged in the respective slots and welded to the sheetmetal webs, the ring-shaped rigid contact elements being formed of short tube sections having a diameter smaller than the diameter of the fuel rods to be supported in the spacer, at least the sheetmetal webs passing in the one direction through the other sheetmetal webs being formed with cutouts, members having a substantially triangular cross section respectively suspended in the cutouts and being held in position by the other sheetmetal webs, the members having two of the resilient contact elements extending in axial direction and projecting diagonally into two adjacent meshes of the grid structure. In accordance with a further feature of the invention, the two resilient contact elements of the members, respectively, project laterally into the mesh beyond the respective ring-shaped rigid contact elements at the upper and lower edges of the sheetmetal webs so as to contact a respective fuel rod, whereas the ring-shaped rigid contact elements are spaced from the respective fuel rod. In accordance with an additional feature of the invention, the grid structure has a multiplicity of meshes respectively defined by wall portions of the sheetmetal webs, the wall portions, respectively, being formed with large-area openings. In accordance with a concomitant feature of the invention, the members suspended in the cutouts are formed of one-piece stamped-out blanks of resilient material having respective head and base parts formed by bends into a substantially triangular cross section, the head and base parts being formed with slots for guidance into cutouts and engaging a mesh wall forming part of the respective sheetmetal webs. 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 spacer for nuclear reactor fuel assemblies, 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. |
044568274 | claims | 1. A transportation and/or storage container for radioactive substances, especially for irradiated fuel elements from nuclear reactors, comprising: a thick-walled open-top closed-bottom metallic cylindrical body for shielding gamma radiation from radioactive substances carried therein, said body having a circular interior cavity and a counterbore in its top; a detachable lining for said cavity of corrosion resistant material, said lining having a top peripheral flange overlying the bottom of said counterbore; a thick metallic cover for said lining disposed in said counterbore for shielding gamma radiation from radioactive substances carried in said lining; sealing means interposed between the opposed surfaces of said flange and said lining cover inwardly of the periphery of said flange; fastening means for said lining extending through said flange into said container body; and bolts extending through said cover into tapped sockets inside said fastening means for detachably fastening said cover to said body independently of the fastening of said lining to said body. an outside cover disposed in said second counterbore; sealing means interposed between the opposed surfaces of said outside cover and the bottom of said second counterbore inwardly of the periphery of said outside cover; means detachably securing said outside cover to the body outwardly of said sealing means; and means defining a test connection having a removable closure therefor and extending through said outside cover for monitoring the interspace between the interior enclosure formed by said lining and its cover and the outer enclosure formed by said body and said outside cover. 2. The structure defined in claim 1 including: a second counterbore in the top of said body outwardly of the first-mentioned counterbore; |
claims | 1. A system for controlling a nuclear power plant including at least one nuclear reactor, the system comprising:a set of sensors and actuators associated with the at least one nuclear reactor;a plurality of electronic control units, each of the electronic control units being configured to perform at least one action from among acquiring a value measured by a corresponding sensor and controlling an actuator, the electronic control units and/or the at least one sensor and/or the at least one actuator being according to several different nuclear safety classes; andat least one information processing unit formed by a processor and a memory associated with the processor, the at least one information processing unit configured for managing the display of data on at least one display screen, the data being associated with the electronic control units, the at least one information processing unit being connected to the plurality of electronic control units, the at least one information processing unit comprising:a set of overlay creation electronic modules for creating overlays, the set of overlay creation electronic modules being configured to create several separate overlays, each of the overlays containing information to be displayed for a respective one of the nuclear safety classes, associated with one or several of the electronic control units according to the nuclear safety class; andan electronic generating module configured to generate at least one data page to be displayed, each of the data pages being obtained by superimposing a plurality of the separate overlays. 2. The system according to claim 1, wherein the at least one information processing unit further comprises at least one module for creating the at least one overlay for each respective nuclear safety class, the overlay creation modules being separated from one nuclear safety class to the next. 3. The system according to claim 1, wherein the at least one information processing unit comprises a single electronic generating module, the electronic generating module being the single electronic generating module. 4. The system according to claim 1, wherein the electronic generating module is according to a highest nuclear safety class from among the different nuclear safety classes. 5. The system according to claim 1, wherein the overlays include an overlay associated with a higher nuclear safety class and an overlay associated with a lower nuclear safety class, the electronic generating module being configured to generate each of the pages by superimposing the overlay associated with the higher nuclear safety class on top of the overlay associated with the lower nuclear safety class, to favor the information from the higher nuclear safety class in case of conflict during the superposition of the overlays. 6. The system according to claim 1, wherein the created overlays are separated from one nuclear safety class to the next. 7. The system according to claim 1, wherein each nuclear safety class is a safety class according to a standard chosen from among the group consisting of: standard IEC 61513, standard IEC 61226, standard IAEA, the United States of America nuclear safety standard, the European nuclear safety standard, the French N4 nuclear safety standard, the Japanese nuclear safety standard, the nuclear safety standard of the Republic of Korea, the Russian nuclear safety standard, the Swiss nuclear safety standard and the British nuclear safety standard. 8. A method for managing the display of data on at least one display screen to control a nuclear power plant including at least one nuclear reactor, the data coming from a plurality of electronic control units, each of the electronic control units being configured to perform at least one action from among acquiring a value measured by a sensor and controlling an actuator, the at least one sensor and/or the at least one actuator being associated with the at least one nuclear reactor, the electronic control units and/or the at least one sensor and/or the at least one actuator being according to several different nuclear safety classes, the method being carried out by at least one information processing unit configured to be connected to the plurality of electronic control units, the at least one information processing unit being formed by a processor and a memory associated with the processor, the method comprising:creating, for each page to be displayed on at least one display screen, several separate overlays, each of the overlays containing information to be displayed for a respective one of the nuclear safety classes, associated with one or several of the electronic control units according to the nuclear safety class,generating at least one data page to be displayed, each of the at least one pages being obtained by superimposing a plurality of the separate overlays, anddisplaying the at least one data page on the at least one display screen. 9. The method according to claim 8, wherein, during the creating of the overlays, the created overlays are separated from one nuclear safety class to the next. 10. The method according to claim 8, wherein the overlays include an overlay associated with a higher nuclear safety class and an overlay associated with a lower nuclear safety class, during the generating of the at least one page, the overlay associated with the higher nuclear safety class is superimposed on top of the overlay associated with the lower nuclear safety class, to favor the information from the higher nuclear safety class in case of conflict during the superposition of the overlays. 11. A non-transitory computer readable medium including instructions which, when executed by a computer, carry out all of the steps of the method according to claim 8. |
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summary | ||
abstract | A radiation source is configured to generate radiation. The radiation source includes a fuel droplet generator constructed and arranged to generate a stream of droplets of fuel that are directed to a plasma generation site; a laser constructed and arranged to generate a laser beam that is directed to the plasma generation site, an angle between the direction of movement of the stream of droplets and the direction of the laser beam being less than about 90°; and a collector constructed and arranged to collect radiation generated by a plasma formed at the plasma formation site when the beam of radiation and a droplet collide. The collector is configured to reflect the radiation substantially along an optical axis of the radiation source. The laser beam is directed to the plasma generation site through an aperture provided in the collector. |
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description | This application claims priority to U.S. Provisional Patent Application No. 62/434,053, filed Dec. 14, 2016, entitled “Dynamic Three Dimensional Compensator,” by E. Bowman et al., and incorporated herein by reference in its entirety. The use of radiation therapy (radiotherapy) to treat cancer is well known. Typically, radiotherapy involves directing a treatment beam of high energy therapeutic radiation (e.g., an electron, photon, ion, or proton beam) into a target or target volume (e.g., a tumor or lesion) in a patient. Before the patient is treated with radiation, a treatment plan specific to the patient is developed. Among other things, the treatment plan includes a dose distribution map and settings for controlling the radiotherapy system in order to achieve that dose distribution. Multi-leaf collimators (MLCs) are commonly used in radiotherapy systems to shape the treatment beam. Typically, an MLC includes two sets of independently adjustable leaves. The first set of leaves is positioned on one side of the beam's path, and the second set is positioned on the other side of the beam's path. Each leaf is thick enough to either completely block or attenuate a beam of radiation. The leaves can be positioned independently of one another to form an aperture so that the shape of the beam corresponds to the shape of the target. In this manner, the target is exposed to the beam while irradiation of tissue surrounding the target is blocked or reduced. The MLC can be used to shape the beam topography at each treatment field through dynamic leaf movements. That is, by extending and retracting the leaves during the time period in which the beam is on and irradiating the target, the shape of the beam can be changed to provide the dose distribution prescribed in the treatment plan. In effect, while the beam is on, the dose is painted across the target by moving the leaves to expose each location (area) in the target to the beam for the length of time specified in the treatment plan in order to achieve the prescribed dose at that location. A recent radiobiology study has demonstrated the effectiveness of delivering an entire, relatively high radiation dose to a target within a single, short period of time (within a fraction of a second, e.g., half a second). This type of treatment is referred to generally herein as FLASH radiation therapy (FLASH RT). Evidence to date suggests that FLASH RT advantageously spares normal, healthy tissue from damage when that tissue is exposed to only a single irradiation for only a very short period of time. A high dose that can be delivered in a short period of time may be referred to herein as a “shot.” With FLASH RT, the dose is delivered so quickly that MLC leaves cannot be used to distribute the dose across the target as described above. The leaves cannot be moved fast enough to change the shape of the beam during a shot. Instead of using the multi-leaf collimator (MLC) leaves to shape the beam, a preformed compensator (e.g., for proton therapy) or collimator (e.g., for electron therapy and for photon therapy) milled specifically for the patient can be placed in the beam path before each FLASH shot to shape the topography of the shot. However, this solution has disadvantages. For example, a compensator or collimator must be uniquely formed for each patient. Also, once milled, the compensator or collimator cannot be changed to accommodate, for example, changes in the treatment plan or changes in the patient's position, and also limits the options for changing the angle of the beam emitter (e.g., nozzle or target assembly) or beam delivery system (e.g., gantry). Embodiments according to the invention overcome the shortcomings associated with using MLC leaves and milled compensators or collimators in FLASH radiotherapy (FLASH RT). Embodiments according to the invention include different dynamic three-dimensional beam modifier designs that can be used to shape the distribution of a dose delivered to a target by a radiation beam emitted from a nozzle or target assembly of a radiotherapy system, particularly a beam that delivers a high radiation dose within a single, short period of time (e.g., less than a second). The beam modifier may be a compensator or a collimator depending on the type of radiation beam. Elements of the beam modifier include material that can block or attenuate the beam. The elements can be dynamically and quickly configured to form an opening or a transparent area through which a portion of the beam can pass unimpeded, and to present different thicknesses of material to completely block other portions of the beam or attenuate other portions of the beam to different degrees, in this manner shaping the dose distribution at the target while protecting surrounding tissue. In some embodiments, the beam modifier includes groups of opposing rods that are at different angles relative to an incident radiation beam and at different elevations relative to each other. Each of the groups includes multiple layers of rods. The beam modifier can be configured by selectively and independently moving the rods to positions that intersect the beam. Rods can be positioned to create an opening in the beam modifier through which the beam can pass. Rods at different elevations can overlap other rods to attenuate or block portions of the beam outside the opening. For example, to completely block a portion of the beam, a sufficient number of rods can be inserted from different directions to overlap in that portion of the beam. For example, to attenuate a portion of the beam, a number of overlapping rods can be inserted in that portion, where the degree of attenuation is determined by the number of overlapping rods. The rods can be quickly reconfigured between shots. In other embodiments, the beam modifier includes layers of blocks (shielding blocks) that block or attenuate the beam and other blocks that are transparent to the beam. The beam modifier can be configured by selectively arranging the blocks. Blocks that are transparent to the beam can be stacked to create the equivalent of an opening through the beam modifier. Shielding blocks can be stacked to attenuate or block the beam outside the transparent portion. For example, to completely block a portion of the beam, a sufficient number of shielding blocks can be stacked in that portion of the beam. For example, to attenuate a portion of the beam, a number of shielding blocks can be stacked in that portion, where the degree of attenuation is determined by the number of stacked shielding blocks. The blocks can be quickly rearranged between shots. In yet other embodiments, the beam modifier includes columns (channels) that can be filled to different levels with liquid metal that can block or attenuate the beam. The beam modifier can be configured by selectively adding or removing liquid metal from the channels. The equivalent of an opening through the beam modifier can be formed by removing liquid metal from some of the channels. Other channels can be partially filled to attenuate the beam or filled to the level necessary to block the beam. For example, to completely block a portion of the beam, a sufficient amount of liquid metal is added to the channels in that portion of the beam. For example, to attenuate a portion of the beam, less liquid metal is included in the channels in that portion, where the degree of attenuation is determined by the amount of liquid metal in the channels. The amount of liquid metal in each channel can be quickly changed between shots. Embodiments according to the invention can thus be described as dynamic three-dimensional beam modifiers because they can be configured and reconfigured in three dimensions—across the path of the beam (two dimensions) and in the direction of the beam. In summary, embodiments according to the invention reduce setup times between dose deliveries, thereby shortening the treatment time and therefore also shortening the total amount of time that the patient needs to remain in position while the treatments are completed. Also, the patient does not need to be moved while reconfiguring the beam modifier, which improves the repeatability and accuracy of the dosage distribution delivered to the target. In addition, embodiments according to the invention provide greater flexibility with respect to patient position, beam delivery system (e.g., gantry) angles, and beam emitter (e.g., nozzle or target assembly) angles, and thus increase the number of treatment options. Significantly, embodiments according to the invention can be used to control dose distribution during FLASH RT. In particular, the dose distribution of each single FLASH shot can be controlled. Even though a shot may last less than a second, the desired dose distribution can be delivered across a target during the shot. In general, embodiments according to the invention remove time as a variable when distributing doses to the target. These and other objects and advantages of embodiments according to the present invention will be recognized by one skilled in the art after having read the following detailed description, which are illustrated in the various drawing figures. This summary is provided to introduce a selection of concepts in a simplified form that is further described below in the detailed description that follows. This summary is not intended to identify key features or essential features of the claimed subject matter, nor is it intended to be used to limit the scope of the claimed subject matter. Reference will now be made in detail to the various embodiments of the present disclosure, examples of which are illustrated in the accompanying drawings. While described in conjunction with these embodiments, it will be understood that they are not intended to limit the disclosure to these embodiments. On the contrary, the disclosure is intended to cover alternatives, modifications and equivalents, which may be included within the spirit and scope of the disclosure as defined by the appended claims. Furthermore, in the following detailed description of the present disclosure, numerous specific details are set forth in order to provide a thorough understanding of the present disclosure. However, it will be understood that the present disclosure may be practiced without these specific details. In other instances, well-known methods, procedures, components, and circuits have not been described in detail so as not to unnecessarily obscure aspects of the present disclosure. Portions of the detailed description that follows are presented and discussed in terms of a method. Although steps and sequencing thereof are disclosed in figures herein (e.g., FIG. 17) describing the operations of this method, such steps and sequencing are exemplary. Embodiments are well suited to performing various other steps or variations of the steps recited in the flowchart of the figure herein, and in a sequence other than that depicted and described herein. FIG. 1 is a block diagram showing selected components of a radiation treatment (radiotherapy) system 100 upon which embodiments according to the present invention can be implemented. In the example of FIG. 1, the system 100 includes an accelerator and beam transport system 104 that generates and/or accelerates and/or delivers a radiation beam 101. In embodiments according to the invention, the system 100 can be a system that can generate and deliver proton beams, electron beams, neutron beams, photon beams, ion beams, or atomic nuclei beams (e.g., beams using elements such as carbon, helium, or lithium). The operations and parameters of the accelerator and beam transport system 104 are controlled so that the intensity, energy, size, and/or shape of the beam are dynamically modulated or controlled during treatment of a patient according to a radiation treatment plan. For FLASH radiotherapy (RT), the accelerator and beam transport system 104 can generate beams that can deliver at least four grays (Gy) in less than one second, and may deliver as much as 20 Gy or 50 Gy or more in less than one second. The beam emitter 106 is used to aim the beam toward various locations (a target) within a patient supported on the patient support device 108 (e.g., a chair, couch, or table) in a treatment room. The beam emitter 106 may be a nozzle or a target assembly depending on the type of beam. For example, a nozzle is used for proton applications, and a target assembly is used for electron or photon applications. A target may be an organ, a portion of an organ (e.g., a volume or region within the organ), a tumor, diseased tissue, or a patient outline. The beam emitter 106 may be mounted on or may be a part of a beam delivery system 202 (FIG. 2; e.g., a gantry) that can be moved relative to the patient support device 108, which may also be moveable. In embodiments, the accelerator and beam transport system 104 is also mounted on or is a part of the beam delivery system 202; in another embodiment, the accelerator and beam transport system is separate from (but in communication with) the beam delivery system. Continuing with reference to FIG. 1, a beam modifier 107 is situated between the beam emitter 106 and the patient support 108. That is, the beam modifier 107 is positioned outside the beam emitter 106, between the beam emitter and the target. The beam modifier 107 may be a compensator or a collimator depending on the type of radiation beam. For example, a compensator is used for proton applications, and a collimator is used for electron or photon applications. The beam modifier 107 and its function(s) are described further below. The beam 101 is thus generated by the accelerator and beam transport system 104 and is emitted from the beam emitter 106 to and through the beam modifier 107, then to the target. The control system 110 receives and implements a prescribed treatment plan. In embodiments, the control system 110 includes a computing system having a processor, memory, an input device (e.g., a keyboard), and perhaps a display. The system 1800 of FIG. 18 is an example of a platform for the control system 110. The control system 110 can receive data regarding operation of the radiotherapy system 100. The control system 110 can control parameters of the accelerator and beam transport system 104, beam emitter 106, beam modifier 107, beam delivery system 202 (FIG. 2) and patient support device 108, including parameters such as the energy, intensity, size, and/or shape of the beam, direction of the beam emitter, configuration of the beam modifier, and position of the patient support device (and the patient) relative to the beam emitter, according to data the control system 110 receives and according to the radiation treatment plan. FIG. 2 illustrates elements of a radiotherapy system 200 for treating a patient 204 in embodiments according to the invention. The system 200 is an example of an implementation of the radiotherapy system 100 of FIG. 1. In embodiments, the beam delivery system 202 and beam emitter 106 can be moved up and down the length of the patient 204 and/or around the patient, and the beam delivery system and beam emitter can move independently of one another. In embodiments, the patient support device 108 can be moved to different positions relative to the beam delivery system 202 and beam emitter 106, rotated about its longitudinal axis, rotated about a central (normal) axis, and/or tilted back and forth about a transverse axis. While the patient 204 is supine in the example of FIG. 2, the invention is not so limited. For example, the patient 204 can instead be seated in a chair. As noted, the radiotherapy system 200 also includes a beam modifier 107 that is located outside the beam emitter 106. The beam modifier 107 can be attached to the beam emitter 106, it can be incorporated into the beam emitter, or it can be held in place downstream of the beam emitter in some other manner. In general, the beam modifier 107 can be moved in concert with the beam emitter 106 so that a radiation beam emitted from the beam emitter will pass through the beam modifier before the beam reaches its target. This provides greater flexibility with respect to patient position and beam delivery system/beam emitter angles and thus increases the number of treatment options. As will be explained further below, the beam modifier 107 can be configured and reconfigured depending, for example, on the treatment plan, in order to achieve the dose distribution prescribed in the treatment plan. In embodiments, the beam modifier 107 can be configured and reconfigured while in place in the radiotherapy system 200. In other embodiments, the beam modifier 107 can be removed from the radiotherapy system 200, reconfigured, and then replaced into the system. Embodiments according to the invention include different dynamic three-dimensional beam modifier designs that can be used to shape the distribution of a dose delivered to a target by a radiation beam emitted from a beam emitter of a radiotherapy system, particularly a beam that delivers a high radiation dose within a single, short period of time (e.g., less than a second). The beam modifier designs are more specifically described below. Rod-Based Beam Modifier FIG. 3 illustrates a beam modifier 300 in embodiments according to the invention. The beam modifier 300 is an example of the beam modifier 107 of FIGS. 1 and 2. The beam modifier 300 includes a number of individual rods exemplified by the rod 303. In embodiments, each rod is solid along its entire length; however, the invention is not so limited. For example, portions of the rods that will not be inserted into the path of an incident radiation beam may not be solid. In the FIG. 3 embodiments, the beam modifier 300 includes a first group 301 of the rods and a second group 302 of the rods. Each rod can be moved (extended and retracted) independently of the other rods, and the distance each rod is moved can be independently controlled. More specifically, the first group 301 includes a number of opposing rods that can each move independently along a respective axis of travel that is parallel to a first axis 305, and the second group 302 also includes a number of opposing rods that can each move independently along a respective axis of travel that is parallel to a second axis 306. Each rod can be moved any distance; that is, a rod may be all the way inserted or all the way withdrawn or at any point between these extremes. Opposing rods are rods that are parallel to the same axis and that have ends that face each other; the end of a rod is the surface of the rod that intersects the longitudinal axis of the rod (the rod's axis of travel). Each of the rods in the beam modifier 300 can include material that completely blocks a radiation beam or material that attenuates a radiation beam. As mentioned above, the beam may be, for example, a proton beam, electron beam, neutron beam, photon beam, ion beam, or atomic nuclei beam. The rod material can be, for example, a high density metal such as, but not limited to, tungsten, tungsten alloys, tantalum, tantalum alloys, lead, lead alloys, brass alloys, or copper alloys. For proton beams, the rod material can be graphite, for example. In embodiments, the same material is used along the entire length of each rod. However, the invention is not so limited. For example, a different material may be used in the portions of the rods that will not be inserted into the path of an incident radiation beam. In embodiments, as described below in conjunction with FIG. 7, some rods may include a portion that is transparent to the beam. Continuing with reference to FIG. 3, the first axis 305 and the second axis 306 are not parallel to each other. In the example of FIG. 3, the first axis 305 and the second axis 306 are orthogonal to each other; however the invention is not so limited. In embodiments, the first group 301 of rods and/or the second group 302 of rods can be moved to change the angle between the first axis 305 and the second axis 306. More specifically, the first group 301 of rods and/or the second group 302 of rods can be rotated collectively relative to each other, so that the angle between them (and between the axes 305 and 306) can be adjusted to facilitate implementation of the treatment plan. The first group 301 of rods includes a first subgroup 311 of rods disposed in multiple first layers and a second subgroup 312 of opposing rods disposed in multiple second layers (e.g., the layer 440 of FIG. 4). The second group 302 of rods includes a third subgroup 313 of rods disposed in multiple third layers and a fourth subgroup 314 of opposing rods disposed in multiple fourth layers. Each of the first, second, third, and fourth layers includes multiple rods. In an embodiment, there are ten layers per subgroup and ten rods per layer (that is, ten rows and ten columns of rods per subgroup). In an embodiment, the height of a rod is one centimeter, and the width is also one centimeter. The rods can be of any practical length, width, and height. In embodiments, the first subgroup 311 and the second subgroup 312 are aligned, and each of the rods in one of these subgroups is aligned with a corresponding rod in the other of these subgroups. Similarly, in embodiments, the third subgroup 313 and the fourth subgroup 314 are aligned, and each of the rods in one of these subgroups is aligned with a corresponding rod in the other of these subgroups. In other words, for example, when a rod in the subgroup 311 is moved toward the center of the beam modifier 300, and the corresponding opposing rod in the subgroup 312 is moved toward the center of the beam modifier, then those rods will both lie on the same axis and their ends will touch if the rods are moved to their fullest extent. In these embodiments, the number of rods in each subgroup is the same. However, the invention is not so limited. For example, in some embodiments, the number of rods in the subgroups 311 and 312 is the same, and the number of rods in the subgroups 313 and 314 is the same but is different from the number of rods in the subgroups 311 and 312. FIG. 4 illustrates another view of the beam modifier 300 in embodiments according to the invention. As can be seen in the figure, the first and second groups of rods 301 and 302 are at different elevations in the direction parallel to the axis 407 (which is orthogonal to the axes 305 and 306). That is, one of the subgroups is closer to the source of the radiation beam 101 (FIG. 1) than the other. The axis 407 corresponds to the axis of the incident radiation beam 101. Accordingly, if rods in the subgroup 311 and/or subgroup 312 are moved far enough toward the beam axis 407 (e.g., into the radiation beam) and rods in the subgroup 313 and/or the subgroup 314 are also moved far enough toward the beam axis, then rods in the first and second groups 301 and 302 will overlap and can be used to block and/or attenuate some portions of the radiation beam. FIG. 5A illustrates a top-down view of the beam modifier 300 in embodiments according to the invention. Only a single layer (e.g., the top layer) of rods is inserted in the example of FIG. 5A. In this example, some of the rods in the subgroups 311 and 312 are only partially inserted toward the center of the beam modifier, and some of the rods in the subgroups 311 and 312 are fully inserted until they are in contact with opposing rods. The rods in the subgroups 313 and 314 are not inserted in the example of FIG. 5A; however, they could be inserted various distances. In FIG. 5A, while only five rods per subgroup are illustrated, the invention is not so limited. The gaps between the rods form an opening 510. The shape of the opening 510 depends on how many of the rods are inserted and how far they are inserted. As shown in the example of FIG. 4 above, there can be multiple layers of rods per subgroup. Thus, in the example of FIG. 5A, there may be layers of rods between the layer of rods of the first and second subgroups 311 and 312 and the layer of rods of the third and fourth subgroups 313 and 314. More specifically, as mentioned above, the subgroups 311, 312, 313, and 314 include multiple layers of rods, the rods can be individually moved different distances, and the rods in the subgroups 311 and 312 can overlap the rods in the subgroups 313 and 314 (and vice versa) depending on how far they are moved. In an embodiment, there are ten layers per subgroup. Thus, depending on how the rods in the beam modifier 300 are configured, the shape of the opening 510 can be different at each layer of the beam modifier. Also, the thickness of the material around the opening 510, measured in the z-direction of FIG. 5A (into the page), can be varied in the x-direction and in the y-direction of FIG. 5A depending on the number of rods that are inserted. This is illustrated in FIG. 5B. As described above, the third and fourth subgroups of rods 313 and 314 are at a higher elevation (closer to the source of the beam 101) than the first and second subgroups of rods 311 and 312. At an areal location (x1,y1) in the beam 101, two rods (e.g., adjacent rods) from two layers of the subgroup 311 may be inserted and two rods (e.g., adjacent rods) from two layers of the subgroup 313 may be inserted so that the rods from each subgroup overlap in the beam, in which case the material in the z-direction of FIG. 5B will be four rods thick at the first location. In effect, the four overlapping rods cast a shadow on the target corresponding to the location (x1,y1) where the rods intersect the beam 101, where the depth of the shadow (the amount of blockage or attenuation) depends on whether the rods attenuate the beam or completely block the beam at that location. As mentioned above, there are multiple rows and columns (layers) of rods in each subgroup. Accordingly, there are many possible arrangements of overlapping rods that can be formed in the path of the beam 101, so that each area of the beam can be attenuated or blocked at the same time to achieve a desired dose distribution. In operation, a radiation beam enters the beam modifier 300 in the z-direction of FIG. 5A and will pass unimpeded through the opening 510. As mentioned above, each of the rods in the beam modifier 300 can include material that completely blocks the beam or material that attenuates the beam. Accordingly, as just described above, different numbers and different lengths of rods can be independently inserted from different directions, layer by layer, to shape the opening 510 and to block portions of the beam completely and/or attenuate portions of the beam by different amounts outside the opening. The overlapping of rods at multiple elevations within the beam creates the desired dosage gradient. The amount of attenuation in a portion of the beam depends on the number of overlapping rods that portion of the beam passes through. In this manner, the beam can be shaped by the beam modifier 300, meaning that the beam modifier both shapes the area of the target covered by the beam and shapes the intensity of the beam that reaches that area, and hence shapes the dose distribution across that area. This is represented in FIG. 6, which illustrates the relative intensity of the beam 101 (FIG. 1) after the beam has passed through the beam modifier 300 of FIG. 3, where the beam modifier has been configured to form an opening 610 and has also been configured with overlapping rods that attenuate portions of the beam to different degrees and overlapping rods that completely block portions of the beam. In FIG. 6, the unshaded areas indicate areas where the beam passed through the opening 610 in the beam modifier 300. Lighter-shaded areas indicate areas where the beam was attenuated by the beam modifier 300 and/or areas that are within the beam penumbra. Darker-shaded areas indicate areas where the beam was further attenuated, and the darkest areas indicate areas where the beam was completely blocked. FIG. 6 thus also represents the dose distribution across the target, where the unshaded areas indicate areas that receive the highest dose, and so on. The beam can be a FLASH shot—a beam that delivers an entire, relatively high radiation dose to the target within a single, short period of time (within a fraction of a second, e.g., half a second). Thus, the beam modifier 300 can control the dose distribution of a FLASH shot. After the shot is delivered, the beam modifier 300 can be rapidly reconfigured by moving rods from their current positions to new positions as needed to change the shape of the opening 510 and/or to change which portions of the beam are to be blocked or attenuated and the amount of attenuation. In embodiments, the beam modifier 300 is remotely configured using the control system 110 of FIG. 1. Once reconfigured, another shot can be delivered. This process can be repeated until the treatments prescribed by the treatment plan are delivered. Specifically, this process is repeated until the prescribed dose distribution across and through the target is achieved. The beam modifier 300 thus reduces setup time between treatments/shots, which shortens the treatment time and therefore also shortens the total amount of time that the patient needs to be in place on the patient support 108 (FIG. 1). The patient does not need to be moved while the beam modifier 300 is reconfigured, which improves the repeatability and accuracy of the dose distribution delivered to the target. In embodiments, some of the rods in the beam modifier 300 have a portion or volume that is transparent to the radiation beam. FIG. 7 illustrates a partially transparent rod 702 that includes a transparent portion 704 and a portion 706 that blocks or attenuates the beam, in embodiments according to the invention. There may be multiple partially transparent rods in each of the subgroups 311, 312, 313, and 314 of FIG. 3. The length of the transparent portion 704 may or may not be the same in each of the partially transparent rods. Partially transparent rods like the rod 702 can be used, for example, to form a region in the opening 510 that blocks or attenuates the beam. This is illustrated in FIG. 8, which shows the opening 510 formed by a layer of rods as described above. In the example of FIG. 8, a layer of partially transparent rods 702 (FIG. 7) is inserted as part of or under the layer of rods that form the opening 510. In this manner, the beam is blocked or attenuated in the areas 804 around the perimeter of the opening 510, and is also blocked or attenuated in the area 806 within the opening 510. FIGS. 9A, 9B, and 9C illustrate examples of cross-sectional shapes of the rods in the beam modifier 300 of FIG. 3. In an embodiment, the height H of a rod is one centimeter, and the width W is also one centimeter. In the example of FIG. 9A, the cross-section is squarish; that is, the height and width of a rod are equal. In the example of FIG. 9B, the cross-section is rectangular; that is, the height is less than the width. The beam modifier 300 can include a combination of both squarish and rectangular rods. For example, the beam modifier can include one or more layers of squarish rods and one or more layers of rectangular rods. Rods with squarish or rectangular cross-sections are advantageous because they can provide better control over the size of the penumbra and greater flexibility in how they can be configured. In the example of FIG. 9C, the cross-section is hexagonal. Other cross-sectional shapes can be utilized. FIG. 10A illustrates the ends of a pair of opposing rods 1001 and 1002 of the beam modifier 300 (FIG. 3) in embodiments according to the invention. For instance, the rod 1001 may be in the subgroup 311, and the rod 1002 may be in the subgroup 312. In general, the respective ends 1011 and 1012 of the opposing rods 1001 and 1002 are complementary in shape, allowing the two ends to fit against each other when the rods are brought into contact. The complementary shape of the ends precludes or reduces the amount of radiation leakage between opposing rod end pairs (e.g., between the two rods 1001 and 1002) when the rods are brought into contact with each other in the beam path. In the example of FIG. 10A, the ends 1011 and 1012 are flat but have step-like surfaces, in which the end 1011 is indented where the end 1012 protrudes in complementary fashion, and vice versa. However, the invention is not so limited. For example, the flat surfaces of the ends 1011 and 1012 can instead be rounded. For example, one of the ends can have a concave surface or surfaces, and the other end can have a complementary convex surface or surfaces. In the examples presented in some of the above figures, the surfaces of the sides of the rods are flat. However, in some embodiments, the surfaces of the sides of the rods are not flat, as shown in the example of FIG. 10B. FIG. 10B is a cross-sectional view showing ends of rods in the beam modifier 300. In the example of FIG. 10B, the rods include protrusions 1024 and indentations 1026 along their lengths (their tops, bottoms, and sides). For instance, the positions of the indentations in the rod 1020 correspond to the positions of the protrusions in the adjacent rods, and vice versa, so that the protrusions fit into the indentations. This type of interlocking rod structure keeps the rods in proper alignment as the rods are moved back-and-forth along their axes of travel. Thus, the rods in the beam modifier 300 can support each other and, therefore, a channel surrounding each rod is not needed. Also, when the rods are inserted into the beam path, the non-flat surfaces preclude or reduce the amount of leakage of radiation through the gaps between the moveable rods. FIG. 11 illustrates a side view of a rod 1101 connected to a drive 1102 in the beam modifier 300 in embodiments according to the invention. In embodiments, each rod in the beam modifier 300 is driven by its own drive 1102. For example, there is a motor per rod. The drives also help hold the rods in place and maintain their alignment. In embodiments, a primary encoder 1105 (e.g., a linear encoder) on each drive 1102 provides an indication that the rod 1101 is properly positioned, and a secondary encoder 1106 (e.g., another linear encoder) mounted on a printed circuit board adjacent to the rod 1101 verifies that the rod is properly positioned. Thus, each rod possesses redundant feedback to ensure its location in the radiation field to comply with applicable safety standards. Block-Based Beam Modifier FIG. 12 illustrates a beam modifier 1200 in embodiments according to the invention. The beam modifier 1200 is an example of the beam modifier 107 of FIGS. 1 and 2. The beam modifier 1200 includes a number of individual blocks exemplified by the block 1203. Each block in the beam modifier 1200 is free-standing; that is, the blocks are not connected to each other when in place in the beam modifier 1200, but are stacked on top of and/or side-by-side with other blocks. The blocks are held in place by a container (not shown) that is transparent to the radiation beam 101 (FIG. 1). However, in embodiments, each block can include a mechanism that allows it to be affixed to an adjacent block. In embodiments, the container is open at the top, but a cover can be used to close that opening. In an embodiment, the blocks in the beam modifier 1200 are arranged in ten rows, ten columns, and ten layers (e.g., a ten-by-ten-by-ten cube of blocks). In an embodiment, each block is cubic. In an embodiment, each block is one cubic centimeter in size. The invention is not limited to all blocks being the same size or shape. For example, one layer may include blocks of a different size then an adjacent layer, so that the edges of the blocks in the first layer are offset from the edges of the blocks in the adjacent layer. For example, one layer may include cubic blocks and an adjacent layer may include rectangular prisms (rectangular cuboids), or each layer may include a combination of cubic blocks and rectangular prisms. As such, the edges of the blocks in one layer would be offset from the edges of the blocks in the adjacent layer. In this manner, leakage of radiation between blocks is precluded or reduced. In embodiments, some of the blocks (shielding blocks) in the beam modifier 1200 completely block or attenuate an incident radiation beam and other blocks (transparent blocks) in the beam modifier are transparent to the beam. FIG. 13 illustrates an example of the beam modifier 1200 that includes a combination of shielding blocks 1302 and transparent blocks 1304. In FIG. 13, the shielding blocks 1302 and transparent blocks 1304 are separated. In operation, the transparent blocks 1304 are combined with the shielding blocks to form the beam modifier 1200 in the shape of a cube. That is, the shape formed by the transparent blocks 1304 and the shape formed by the shielding blocks 1302 have complementary surfaces. In the example of FIG. 13, the shielding blocks 1302 are grouped and the transparent blocks 1304 are grouped, with the transparent blocks located toward the top of the beam modifier 1200. However, the invention is not so limited. For example, shielding blocks and transparent blocks can be intermixed in the beam modifier 1200 in any pattern. FIG. 14 illustrates a top-down view of the beam modifier 1200 in embodiments according to the invention. Only a single layer (e.g., the top layer) of blocks is shown in FIG. 14. In the example of FIG. 14, the shielding blocks 1302 (which are shaded in the figure) and the transparent blocks 1304 are arranged to shape the equivalent of an opening (transparent portion 1406) in the beam modifier 1200. One or more columns of transparent blocks 1304 extending all the way through the beam modifier 1200 can be formed so that the equivalent of an opening exists all the way through the beam modifier. In operation, an incident radiation beam will pass unimpeded through the transparent portion 1406 (through the transparent blocks 1304) but will be blocked or attenuated by the surrounding shielding blocks 1302. As mentioned above, there are multiple (e.g., ten) layers of blocks in the beam modifier 1200. Each layer can be configured independently and differently from the other layers. As such, depending on how the blocks in the beam modifier 1200 are configured, the shape of the transparent portion 1406 can be different at each layer of the beam modifier 1200, and the thickness of the shielding material around the transparent portion 1406 can be varied in the z-direction of FIG. 14. Also, more than one transparent portion like the transparent portion 1406 can be readily formed at different locations in the beam modifier 1200. Accordingly, there are many possible arrangements of blocks that can be formed in the path of the beam 101 (FIG. 1), so that each area of the beam can be attenuated or blocked at the same time to achieve a desired dose distribution. Thus, in operation and depending on how the blocks in the beam modifier 1200 are arranged, some portions of a radiation beam can be blocked by the layers of shielding blocks 1302 and/or some portions of the beam can be attenuated by the shielding blocks, while other portions of the beam are not blocked or attenuated. The amount of attenuation in a portion of the beam depends on the number of overlapping (stacked) shielding blocks that portion of the beam passes through. In this manner, the beam can be shaped by the beam modifier 1200, meaning that the beam modifier both shapes the area of the target covered by the beam and shapes the intensity of the beam that reaches that area, and hence shapes the dose distribution across that area, similar to the example shown previously herein in FIG. 6. The beam modifier 1200 of FIG. 12 can control the dose distribution of a FLASH shot. After the shot is delivered, the beam modifier 1200 can be rapidly reconfigured by removing some or all of the blocks and then rearranging them for the next shot. This can be quickly achieved using, for example, a programmed robot that can quickly identify, select (e.g., pick up), and position each block under control of the control system 110 (FIG. 1). Alternatively, the beam modifier 1200 can be removed and replaced with another similar type of beam modifier that has already been configured for the next shot. Once the beam modifier is reconfigured or replaced, another shot can be delivered. This process can be repeated until the treatments prescribed by the treatment plan are delivered. Specifically, this process is repeated until the prescribed dose distribution across and through the target is achieved. The beam modifier 1200 of FIG. 12 can reduce setup time between treatments/shots, which shortens the treatment time and therefore also shortens the total amount of time that the patient needs to remain on the patient support 108 (FIG. 1). The patient does not need to be moved while the beam modifier 1200 is replaced or reconfigured, which improves the repeatability and accuracy of the dosage distribution delivered to the target. Liquid Shielding-Based Beam Modifier FIG. 15 illustrates a cross-sectional view of a beam modifier 1500 in embodiments according to the invention. The beam modifier 1500 is an example of the beam modifier 107 of FIGS. 1 and 2. The beam modifier 1500 includes a number of channels exemplified by the channel 1503. Each channel includes a portion (shielding portion; e.g., portion 1504) that lies within the path of the radiation beam 101 and a portion (reservoir portion; e.g., portion 1505) that lies outside the path of the beam. In an embodiment, the beam modifier 1500 includes a ten-by-ten arrangement of channels. However, the invention is not so limited. In the FIG. 15 embodiments, each channel includes an actuator 1510, a piston 1511, a plunger 1512, and a spring 1513. Between the piston 1511 and the plunger 1512, each channel is filled with a heavy liquid or high density liquefied metal (e.g., lead) that is suitable for blocking or attenuating the radiation beam. Each channel includes a channel wall (e.g., the channel wall 1520) that surrounds the channel. In the FIG. 15 embodiments, the channel wall is illustrated as being straight. However, the invention is not so limited. FIG. 16 is a top-down view of the beam modifier 1500 showing the arrangement of the channels in the beam modifier in embodiments according to the invention. In the FIG. 16 embodiments, the channels (e.g., the channel 1503) have a squarish cross-section. However, the invention is not so limited. Also, the channels can have the same cross-sectional areas or they can have different cross-sectional areas. In the FIG. 16 embodiments, the channels are illustrated as being uniformly arranged in rows and columns. However, the invention is not so limited, and the channels can be offset from one another. By offsetting (staggering) the channels, a more uniform dose distribution at the target may be achieved. With reference again to FIG. 15, the levels of the liquid metal in the channels in the shielding portion 1504 of the beam modifier 1500 can be independently controlled by the control system 110 (FIG. 1). Each channel may include no liquid metal in the shielding portion 1504, or it may be filled to its maximum level with liquid metal, or it may be filled to a level somewhere between the minimum level and the maximum level, independent of the level of liquid metal in the other channels. To raise the level of liquid metal in, for example, the channel 1503, the actuator 1510 is used to depress the piston 1511, thereby forcing the liquid metal from the reservoir portion 1505 into the shielding portion 1504. This will raise the level of the liquid metal in the shielding portion 1504 of the channel 1503, thereby applying pressure to the plunger 1512 and compressing the spring 1513. To lower the level of liquid metal in the channel 1503, the actuator raises the piston 1511, drawing liquid metal from the shielding portion 1504 of the channel into the reservoir portion 1505 with the aid of pressure applied by the spring 1513 on the plunger 1512. With reference again to FIG. 16, the equivalent of an opening (transparent portion 1610) in the beam modifier 1500 is formed by channels in the beam modifier that do not include any liquid metal in the shielding portion 1504. In the example of FIG. 16, the channels that are not shaded (e.g., the channel 1611) do not include any liquid metal in the shielding portion 1504, and so those channels form the transparent portion 1610 that extends all the way through the beam modifier 1500. In operation, an incident radiation beam will pass unimpeded through the transparent portion 1610 (through the channel 1611 and the like) but will be blocked or attenuated by the liquid metal in the surrounding channels (the shaded channels, e.g., the channel 1612). The shape of the transparent portion 1610 can be changed by adding liquid metal to one or more of the channels in the beam modifier 1500 and/or removing liquid metal from one or more of the adjacent channels. Also, more than one transparent portion like the transparent portion 1610 can be readily formed at different locations in the beam modifier 1500. As mentioned above, each channel can be configured independently and differently from the other channels. Accordingly, there are many possible channel configurations that can be formed in the path of the beam 101 (FIG. 1), so that each area of the beam can be attenuated or blocked at the same time to achieve a desired dose distribution. In operation and depending on how the channels in the beam modifier 1500 of FIG. 15 are arranged (how much liquid metal is contained in the shielding portion 1504 of each channel), some portions of a radiation beam can be blocked and/or some portions of the beam can be attenuated, while other portions of the beam are not blocked or attenuated. The amount of attenuation in a portion of the beam depends on the amount of liquid metal that portion of the beam passes through. In this manner, the beam can be shaped by the beam modifier 1500, meaning that the beam modifier both shapes the area of the target covered by the beam and shapes the intensity of the beam that reaches that area, and hence shapes the dose distribution across that area, similar to the example shown previously herein in FIG. 6. The beam modifier 1500 can control the dose distribution of a FLASH shot. After the shot is delivered, the beam modifier 1500 can be rapidly reconfigured by changing the level of the liquid metal in any of the channels as needed. In embodiments, the beam modifier 1500 is remotely configured using the control system 110 of FIG. 1. Once the beam modifier 1500 is reconfigured, another shot can be delivered. This process can be repeated until the treatments prescribed by the treatment plan are delivered. Specifically, this process is repeated until the prescribed dose distribution across and through the target is achieved. The beam modifier 1500 of FIG. 15 can reduce setup time between treatments/shots, which shortens the treatment time and therefore also shortens the total amount of time that the patient needs to remain on the patient support 108 (FIG. 1). The patient does not need to be moved while the beam modifier 1500 is reconfigured, which improves the repeatability and accuracy of the dosage distribution delivered to the target. Radiation Therapy Method FIG. 17 is a flowchart 1700 of a radiation therapy method in embodiments according to the invention. Aspects of the operations presented below can be performed using the computing system 1800 of FIG. 18 to implement the control system 110 of FIG. 1. In block 1702 of FIG. 17, a beam modifier (e.g., the beam modifier 107 of FIG. 1) is configured to intersect an incident radiation beam. In embodiments utilizing the beam modifier 300 of FIG. 3, the configuring is performed by selectively moving rods of the beam modifier to positions that intersect the beam as described above. In embodiments utilizing the beam modifier 1200 of FIG. 12, the configuring is performed by selectively arranging blocks in the beam modifier as described above. In embodiments utilizing the beam modifier 1500 of FIG. 15, the configuring is performed by selectively adding or removing liquid metal from shielding portions of channels as described above. In block 1704 of FIG. 17, the radiation beam is emitted from a beam emitter into the beam modifier, then to a target. FIG. 18 shows a block diagram of an example of a computing system 1800 with which the embodiments described herein may be implemented. In its most basic configuration, the system 1800 includes at least one processing unit 1802 and memory 1804. This most basic configuration is illustrated in FIG. 18 by dashed line 1806. The system 1800 may also have additional features and/or functionality. For example, the system 1800 may also include additional storage (removable and/or non-removable) including, but not limited to, magnetic or optical disks or tape. Such additional storage is illustrated in FIG. 18 by removable storage 1808 and non-removable storage 1820. The system 1800 may also contain communications connection(s) 1822 that allow the device to communicate with other devices, e.g., in a networked environment using logical connections to one or more remote computers. The system 1800 also includes input device(s) 1824 such as keyboard, mouse, pen, voice input device, touch input device, etc. Output device(s) 1826 such as a display device, speakers, printer, etc., are also included. In the example of FIG. 18, the memory 1804 includes computer-readable instructions, data structures, program modules, and the like. Depending on how it is to be used, the system 1800—by executing the appropriate instructions or the like—can be used as a control system to implement a radiation therapy method. For example, the instructions and the like can be used to remotely configure and reconfigure the beam modifier 107 (FIG. 1) and to remotely trigger the radiation beam 101 (FIG. 1) once the beam modifier is ready. In summary, embodiments according to the invention provide greater flexibility with respect to patient position, beam delivery system angles, and beam emitter angles, and thus increase the number of treatment options. Embodiments according to the invention also reduce setup times between dose deliveries, thereby shortening the treatment time and therefore also shortening the total amount of time that the patient needs to remain in position while the treatments are completed. Also, the patient does not need to be moved while reconfiguring the beam modifier, which improves the repeatability and accuracy of the dosage distribution delivered to the target. Significantly, embodiments according to the invention can be used to control dose distribution in FLASH RT, in particular during each single FLASH shot. Even though a shot may last less than a second, the desired dose distribution can be delivered across a target. In general, embodiments according to the invention remove time as a variable when delivering doses to the target. Although the subject matter has been described in language specific to structural features and/or methodological acts, it is to be understood that the subject matter defined in the appended claims is not necessarily limited to the specific features or acts described above. Rather, the specific features and acts described above are disclosed as example forms of implementing the claims. |
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052271285 | summary | The present invention relates generally to nuclear reactors, and, more specifically, to an assembly for controlling nuclear reactivity in a fuel bundle. BACKGROUND OF THE INVENTION In a nuclear reactor such as a boiling water reactor (BWR), a reactor core containing nuclear fuel rods is provided for heating water to be used as a power source for a steam turbine-generator, for example. The fuel rods are typically grouped together in fuel bundles or assemblies, having a square matrix for example, with upper and lower tie plates being used to maintain a predetermined lateral spacing between the adjacent fuel rods. In the BWR, a recirculating coolant, or water, is suitably channeled through the lower tie plate and upwardly between the fuel rods for cooling the fuel rods during operation, with the coolant having an increasing steam void fraction as it rises upwardly along the fuel rods with the resulting liberated steam being suitably channeled to the steam turbine. In order to control reactivity of the fuel rods, conventional solid control rods are selectively translatable upwardly and downwardly between the fuel rods for selectively absorbing neutrons emitted therefrom during operation. The control rods may be in the conventional form of a cruciform disposed between adjacent fuel bundles, or may be in the form of a plurality of finger-type rods insertable in the fuel bundles between selected fuel rods thereof. In both examples, suitable control rod drives (CRDs) are also provided which may be located below the lower head of the pressure vessel or above the upper head of the pressure vessel depending upon the particular design. In both designs, however, the CRDs are disposed outside the pressure vessel and require suitable penetrations of the pressure vessel for the translatable plungers thereof to translate the control rods. Concepts for controlling reactivity of the reactor core using a liquid neutron absorber in hollow tubes are known in the literature, with the level of the liquid absorber being selectively varied. This is analogous to the degree of insertion of the solid control rods into the reactor core. However, a liquid neutron absorber reactivity control system has practical problems associated with the installation of the required many tubes and reservoirs for the liquid absorber in the pressure vessel. Furthermore, the ability to replace components of such a system is also required which imposes even further practical problems for dealing with the substantial number of tubes and connections therebetween which require individual replacement with suitable leak tight connections. The liquid absorber must also be suitably separated from the circulating coolant in the reactor core to prevent the adulteration thereof which would adversely affect operation of the reactor core. SUMMARY OF THE INVENTION A reactor core removable fuel assembly includes upper and lower tie plates having pluralities of fuel rods and hollow control rods extending therebetween. The lower tie plate includes a lower manifold therein joined in flow communication with a reservoir containing a neutron absorbing control liquid, with the reservoir being removable from the reactor core together with the fuel assembly. The control liquid is selectively pumped from the reservoir through the lower manifold and into the control rods for selectively varying the level of the control liquid therein for controlling reactivity. |
abstract | A device for accelerating ions between a potential towards a central point in space is disclosed. The device can be used to accelerate ions along a collision path with other accelerated ions or other present particles resulting in a nuclear fusion reaction. The device improves upon the prior art by using a plasma as one of the electrodes forming the potential. |
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043274439 | description | More particularly, FIG. 1 illustrates a typical cross section normal to the longitudinal axes of the capillary fuel elements comprising the fuel components which are constructed as integral units from moderator material and are arranged to form channels for coolant circulation. 1 is the fuel component which contains horizontal capillary troughs, slightly inclined to vertical, which are nonadhesive to the fuel and which have longitudinal openings that approximately equal twice the capillary constant of the fuel at the environment of the core. 2 is the liquid fuel which flows under gravity through the capillary troughs and which forms a meniscus that has an approximately circular cross section and projects above the trough edge. The combination of moderator construction material and fuel forms a critical mass which generates heat through a nuclear chain reaction. Heat is removed from the core by a coolant which circulates through the channel 3. FIG. 2 shows, in exaggeration, the helical flow direction of fuel through a fuel element of a cylindrical fuel component. Fuel is fed to the trough from an upper fuel reservoir and flows at a rate predetermined by the pitch of the fuel element helix down to the lower fuel reservoir. FIG. 3 is a longitudinal cross section through a cylindrical reactor. Cylindrical fuel components 1, constructed as integral units from moderator material are arranged concentrically to form a core with coolant channels 3. The core is surrounded by a neutron reflector 4. Fuel 2 from the upper fuel reservoir 5 is distributed through the distribution means 6 to the fuel components 1 through which it flows into the lower fuel reservoir 7 from where a pump 8 returns it through conduit 9, which is the inner surface of the center fuel component, to the upper fuel reservoir. Coolant is admitted through inlet 10, circulates through the core channels 3 and exists through outlet 11. The core is surrounded by a neutron reflector 4 and is enclosed in containment vessel 12. Because the reactor has a negative temperature coefficient, it will tend to operate at the same average moderator temperature at all power levels. This temperature is maintained by controlling the uranium fuel concentration. A more particular example is a cylindrical reactor, illustrated by FIG. 3, which employs 4.6 percent enriched uranium dissolved in molten bismuth to form a 8.5 percent solution as fuel at an average 850.degree. C. (1560.degree. F.) temperature. The fuel capillary constant at 850.degree. C. is 0.28 cm (0.11 in). The fuel component 1, a magnified cross section of which is illustrated by FIG. 1, are 25 concentric, 2.8 cm (1.1 in) thick, 180 cm (70.9 in) high, graphite cylinders constructed as an integral unit of capillary troughs and moderator material. The distance between the cylinders, i.e., the coolant channel 3 width, is 0.75 cm (0.30 in); and the width of the cylinder lips which contain the fuel 2, is 0.90 cm (0.35 in). There are 183 equally spaced lips per vertical surface of a cylinder. The outside radius of the outermost cylinder is 90 cm (35.4 in), and the inside radius of the innermost cylinder is 5.0 cm (2.0 in). The neutron reflector 4, constructed of graphite, is 30 cm (11.8 in) thick. The coolant is methane at 134 atmospheres (2000 psi) pressure. The physical parameters of the bare example reactor have been determined to be as follows: Regeneration factor (.eta.) 1.89 PA0 Thermal utilization factor (f): 0.980 PA0 Infinite multiplication factor (k.infin.): 1.85 PA0 Diffusion area (corrected for coolant channel voids) (L.sup.2): 156.9 cm.sup.2 PA0 Neutron age (.tau.): 486 cm.sup.2 PA0 Geometric buckling factor (B.sup.2): 0.00098/cm.sup.2 PA0 Critical mass: 28.6 kg (62.9 lb) uranium-253 dissolved in 5883 kg (12,943 lb) of molten bismuth. PA0 Core thermal power: 50 Mw PA0 Neutron energy: thermal PA0 Fuel (clean): 3900 ppm uranium-235 in molten bismuth; 8.5 at. % of 4.6% enriched uranium in molten bismuth. PA0 Fuel capillary constant: 0.28 cm (0.11 in) at 850.degree. C. (1560.degree. F.) PA0 Moderator: graphite (1.9 gm/cm.sup.3) as an integral part of fuel component. PA0 Coolant: Methane at 134 atm (2000 psi) PA0 Core height: 180 cm (70.9 in) PA0 Core radius: 90 cm (35.4 in) PA0 Total number of fuel components: 25 concentric cylinders PA0 Total number of capillary troughs per cylinder vertical surface: 183 equally spaced. PA0 Coolant channel width (distance between cylinder surfaces): 0.75 cm (1.90 in) PA0 Fuel volume: 0.65 m.sup.3 (22.9 ft.sup.3) PA0 Moderator volume: 2.15 m.sup.3 (75.7 ft.sup.3) PA0 Coolant (void) volume: 1.78 m.sup.3 (62.7 ft.sup.3) PA0 Fuel volume fraction: 14.7% PA0 Moderator volume fraction: 46.8% PA0 Coolant volume fraction: 38.9% PA0 Maximum fuel temperature: 1100.degree. C. (2012.degree. F.) PA0 Maximum moderator temperature: 1000.degree. C. (1832.degree. F.) PA0 Coolant inlet temperature: 600.degree. C. (1112.degree. F.) PA0 Coolant outlet temperature: 900.degree. C. (1652.degree. F.) PA0 Coolant flow rate: 129,102 kg/hr (284,024 lb/hr) PA0 Coolant core channel velocity: 2.7 m/sec (9.0 ft/sec) PA0 Coolant flow frontal area: 0.52 m.sup.2 (5.8 ft.sup.2) PA0 Total heat transfer surface: 821 m.sup.2 (8834 ft.sup.2) The operating characteristics of the example reactor are as follows: Although the invention had been described with a certain degree of particularity, especially in the use of a nuclear reactor that utilizes fuel components which are constructed as integral units from moderator materials, it is to be understood that the present disclosure is made by way of example and illustration only and that numerous changes in application, in construction details, and in the arrangement and combination of parts may be made without departing from the spirit and scope of the invention hereinafter claimed. |
claims | 1. A non-resonance photo-neutralizer for neutral beam injectors comprisingfirst and second mirrors having opposing mirror surfaces forming a photon trap, wherein the mirror surface of the first mirror is concave and the mirror surface of the second mirror is flat, wherein the first mirror comprises a mirror assembly including a central mirror and first and second outer mirrors coupled to the central mirror. 2. The photo-neutralizer of claim 1 wherein the photon trap comprises a confinement region adjacent a family of normals common to the mirror surfaces of the first and second mirrors. 3. The photo-neutralizer of claim 1 wherein the central mirror is cylindrically shaped and the outer mirrors are conically shaped. 4. A negative ion based neutral beam injector comprisinga negative ion source, anda non-resonance photo-neutralizer co-axially positioned with the negative ion source, wherein the photo-neutralizer including first and second mirrors having opposing mirror surfaces forming a photon trap, wherein the mirror surface of the first mirror is concave and the mirror surface of the second mirror is flat, wherein the first mirror comprises a mirror assembly including a central mirror and first and second outer mirrors coupled to the central mirror. 5. The neutral beam injector of claim 4 wherein the photon trap comprises a confinement region adjacent a family of normals common to the first and second mirror surfaces. 6. The neutral beam injector of claim 4 wherein the central mirror is cylindrically shaped and the outer mirrors are conically shaped. |
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059237201 | summary | BACKGROUND OF THE INVENTION 1. Field of Invention The present invention is directed toward the field of x-ray diffraction as a versatile tool to determine the structure of atomic and superlattice systems with preferred orientation along at least one dimension. The invention may be configured for the determination of structure in lipid membranes, in-situ thickness measurements of thin films during growth, and determination of lattice mismatch in epitaxial crystalline films. 2. Description of the Prior Art X-Ray diffraction has been used to measure in situ thickness of thin films during deposition (Luken, et. al., SPIE Vol. 2253:327 (1994)). Luken et al. describe an angle dispersive x-ray reflectometer which employs a Johansson-type (T. Johansson, Zeit. Physik, 82:507 (1933)) curved crystal monochromator to focus and wavelength-select X-radiation, with a convergence angle of 2.5.degree. (4.4.times.10.sup.-1 radians), down to a silicon substrate surface on which a W/Si multilayer is grown. The Johansson-type crystal is one in which the reflecting surface is ground to a radius of curvature 2R and the crystal is subsequently bent to a radius of curvature R. The low angle x-ray reflectivity is monitored from the Si substrate simultaneously between 0.degree. and 2.5.degree. using a linear position sensitive charge-coupled device (CCD) detector. The authors used the instrument to monitor the growth of the multilayer in-situ during evaporative deposition. While in principle, the Johansson crystal provides "perfect" point-to-point focusing, there are limitations to using Johansson crystals. For example, the size of the beam at the focus is approximately the same size as the source. For a fine focus x-ray source, with a target source size of 0.4.times.8 mm.sup.2, this dimension at a 6.degree. (0.10 radians) takeoff angle is defined by (0.4 mm)Sin(0.10) and has a value of about 42 .mu.m in the focusing plane. To further reduce the focus, the effective source size would need to be reduced with a slit to block part of the radiation. Not only would the use of a slit diminish the intensity of the x-ray beam, but alignment is now made considerably more difficult, since the monochromator and the sample need to be positioned to within microns with respect to the source in order to take advantage of the small focus. Furthermore, because the crystal monochromator surface must be ground and bent to a very specific curvature, there is, for practical purposes, no forgiveness built into the design to compensate for misalignments or bending error. Thus, the requirement that the surface be ground and then bent makes the fabrication expensive. Small angle x-ray scattering has been used to measure structure in oriented lipid bilayers (Mason and Trumbore, Biochemical Pharmacology 51:653 (1996)). Using small-angle x-ray spectrometry, Mason and Trumbore report the sensitivity of the method to indicate the incorporation and location of antioxidants into the lipid matrix. To achieve the orientation, multilayer stacks of the lipid bilayers are centrifuged down onto a flat substrate from vesicles suspended in an aqueous medium. The lipids are found to align spontaneously with the stacking axis normal to the substrate surface. The substrate is made from a bendable sheet of aluminum which is subsequently mounted on a curved glass surface (radius of curvature c.a. 20 mm). An incident x-ray beam is then focused with a bent grazing incidence mirror to illuminate the curved substrate with an intense beam of small, but unspecified, angular divergence. Different parts of the incident beam intersect the curved surface at different angles of incidence, and the scattering from the entire beam is measured on a position-sensitive x-ray detector which measures the intensity as a function of linear position along the detector axis. The discrete diffraction peak intensities are then Fourier transformed to determine the electron density profile within the lipid bilayer. Mason and Trumbore report the difference in electron density between the oriented lipid lamellar stack incubated in the vesicle state without antioxidant and the same lipid incubated with the target antioxidant in the vesicular suspension before centrifugation. While this method is able to capture all the relevant diffraction information, the technique suffers from the time-consuming step of gluing the aluminum substrate to the curved glass surface. Furthermore, the x-ray beam is not monochromatic, but is simply filtered to significantly reduce the K.sub..beta. radiation. The dominant radiation which diffracts from the sample is the K.sub..alpha.1 /K.sub..alpha.2 doublet. The continuous brehmstrahlung background radiation, particularly at energies between 4 and 8 keV, remains. This continuous spectrum radiation increases the background signal on top of which the diffraction peaks are observed and this subsequently diminishes the ability to observe weak diffraction lines and accurately determine integral peak areas. High resolution, wide angle x-ray scattering is commonly used to determine the lattice parameters in epitaxially grown films (in particular, strained-layer superlattices) with respect to the lattice of the single crystal substrate. The typical approach is to employ a two-crystal spectrometer (monochromator and sample) and measure a rocking curve of the sample in the vicinity of a Bragg diffraction angle from the sample substrate. These angles are typically in the range of 30.degree. to 50.degree. and the rocking curve scan is performed over a range of several degrees. During the rocking scan, the diffraction intensities are measured using a scintillation detector with an entrance slit large enough to accept diffraction over an angular range of several degrees. The wide detector slit precludes the ability to know the diffraction angle precisely. As a result, satellite peaks and orientation of reciprocal lattice vector in strained-layer superlattices are not readily discernible. Picreaux et al. (Semiconductors and Semimetals, 33:139 (1991)) employ a linear position-sensitive x-ray detector (PSD) to measure diffraction intensities from epitaxial films in rocking curve scans with a high resolution two-crystal x-ray spectrometer. While the use of the PSD provides information to allow reciprocal space mapping of the epitaxial layers, the method still requires illumination of the substrate with a highly collimated, monochromatic beam and then measuring the diffraction intensities while step scanning the sample tilt one angle at a time; this approach is both complicated and time-consuming. Using a high resolution, two-crystal x-ray spectrometer, Tsuchiya et al. (Proc. 4th Indium Phosphide and Related Materials Conf., Newport, R.I., 1992) describe feedback control used to adjust the growth conditions during deposition of a vapor phase epitaxial grown film of InGaAs on a single crystal substrate, InP. A scintillation detector with a wide slit was used and the entire x-ray source and monochromator optics were rotated about the sample in order to perform the rocking curve scans. While this method demonstrates the feasibility of using x-ray diffraction for deposition feedback, rotation of the x-ray source about the sample is cumbersome and limits the amount of space available for the deposition system. Furthermore, the method is impractical for faster deposition, since the incident angles are stepped one at a time. X-rays may be simultaneously focused and monochromatized by reflecting a divergent x-ray beam from a curved single crystal such that incident beam intersects the crystal at the Bragg diffraction angle for the desired wavelength. An ideal shape for such a focusing x-ray monochromator is for the crystal curvature to be identical with a logarithmic spiral. DeWolff (Selected Topics on X-Ray Crystallography, Ch. 3, ed. J. Bouman, North-Holland, Amsterdam, 1951) describes a four-point crystal bender to approximate the ideal logarithmic spiral form for a focusing monochromator crystal to second order with respect to the local crystal curvature. This monochromator design has been employed for almost half a decade in powder diffraction spectrometers. The bending design is simple, robust, and in contrast to the Johansson-type focusing, the logarithmic spiral does not require a true point x-ray source. The main disadvantage of this type of focusing monochromator is that the focusing quality can not be improved beyond that already accomplished with the conventional four-point bending apparatus. This inherent limitation is due to the difference in functional form between the ideal logarithmic spiral and the shape that the monochromator can assume in a mechanical, four-point bending apparatus. X-rays are totally reflected from smooth mirror surfaces when the x-rays illuminate the mirror below a grazing incident critical angle. For hard x-rays (>1 keV), this angle is typically on the order of a few tenths of a degree. Underwood and Turner (SPIE, 106:125 (1977)) describe how bent, nondiffracting mirror surfaces can made to focus or collimate x-rays more efficiently by grinding the sides of the mirror such that the width of the reflecting surface varies as function of the length. This shaping procedure is used to "tune" the moment of inertia as a function of length, and allows a bending system to more accurately define the shape of the mirror to the ideal parabola or ellipse. The authors intended this design to be used in astrophysical applications for x-ray telescopes; and these mirror focusing elements differ significantly from diffracting crystal optics. Thus, there remains a need for an x-ray spectrometer with a curved crystal monochromator which can provide improved point focusing of the x-ray source and micron scale scanning of the sample surface. There is a further need for methods of preparing curved crystals having a curvature of a logarithmic spiral which overcome the inherent limitations of the prior art. SUMMARY OF THE INVENTION It is the object of the present invention to provide an x-ray spectrometer which provides superior point focusing of a source x-ray beam from a real extended source. It is a further object of the invention to provide a curved single crystal or other dynamically diffracting element for use as an x-ray monochromator having a surface curvature which most nearly approximates a logarithmic spiral. Dynamical diffraction includes diffraction from perfect crystals, like silicon germanium and lithium fluoride, as well a reflection from synthetic multilayers, such as W/Si alternating film stacks. It is yet a further object of the invention to provide an x-ray reflectometer which permits data accumulation over a range of angles incident to the sample surface in a single measurement. It is yet a further object of the invention to provide a method of measuring x-ray reflectance or x-ray diffraction over a range of incident angles to the sample surface in a single measurement. It is yet a further object of the present invention to provide an improved and more efficient method of determining electron density profiles in lipid layers. It is yet a further object of the present invention to provide an improved and more efficient method of determining epitaxial film structure and growth. These and other objects of the invention are accomplished by use of the x-ray spectrometer described herein. Unique features of the invention are intended to make it possible to reduce the data collection times several orders of magnitude and perform diffraction scanning with resolution on the order of microns in the scanning direction. By "logarithmic spiral", it is meant a mathematically defined surface curvature that is often occurring in nature. Diverse objects such as snail shells, fiddlehead ferns and other naturally occurring elements follow the shape of the logarithmic spiral; while the curvatures described in this disclosure are of the logarithmic spiral form, the radius of curvature is orders of magnitude larger than those found in nature. An advantage of the present invention is that the x-ray source need not be a point in order to obtain high resolution. This is an advantage as most real laboratory x-ray sources are extended and not point sources. In addition, due to the ability to simultaneously collect date over a range of incident angles, data collection is much more rapid than in conventional spectrometers. In one aspect of the invention, an x-ray spectrometer is provided which includes an x-ray source; a curved crystal monochromator which focuses a monochromatic x-ray beam onto a sample surface, the curved crystal monochromator comprising a shape which is substantially identical to a logarithmic spiral; and a position-sensitive x-ray detector positioned to receive x-ray beams diffracted or reflected from a sample surface. Whether or not the source x-ray beam is diffracted or reflected is a function of the angle of incidence of the x-ray beam on the sample. In preferred embodiments, the width of the curved crystal monochromator is linearly tapered along an arclength s of the crystal. In other preferred embodiments, width of the the curved crystal monochromator has a taper selected to minimize a third order difference in s between an ideal logarithmic spiral curve and the curved crystal. The monochromator crystal may be shaped along its length in a form other than rectangular. The taper of the crystal monochromator may be a taper less than about 100 milliradians and preferably about 20 milliradians. In other preferred embodiments, a linear position-sensitive proportional detector is used. The position-sensitive detector may be a linear photodiode array, a linear charge coupled device, a two-dimensional proportional x-ray detector, or a two-dimensional charge coupled device. The spectrometer may also include at least one single slit is positioned between the x-ray source and the monochromator and a plurality of slits may be positioned in front of the monochromator. In other preferred embodiments of the invention, the x-ray spectrometer includes a sample which may be an epitaxially grown layer, an evaporated layer, an epitaxially grown multilayer or superlattice, or a multilayer deposited by evaporation. The x-ray spectrometer may be attached as an accessory to a larger film deposition system and the film parameters determined by the spectrometer may be used to control the deposition of a film. In other preferred embodiments, the x-ray spectrometer may be adapted oriented to measure x-ray reflectivity from horizontal surfaces (i.e., liquids) or for scanning in the lateral direction. The x-ray spectrometer may additionally include a second focusing device, such as a curved mirror, positioned so as to focus in the plane substantially perpendicular to the curved crystal monochromator. The second focusing device may be a curved crystal monochromator. In another aspect of the invention, a curved crystal monochromator is provided which focuses a monochromatic x-ray beam onto a sample surface, wherein the curved crystal monochromator comprises a shape substantially identical to a logarithmic spiral as measured along the crystal monochromator length, wherein the width of the curved crystal monochromator is linearly tapered along an arclength s of the crystal. In preferred embodiments, the curved crystal monochromator has a taper selected to minimize a third order difference in s between an ideal logarithmic spiral curve and the curved crystal. The monochromator crystal may be shaped along its length in a form other than rectangular. The monochromator crystal may have a taper less than about 0.10 radians, and preferably less than 0.02 radians. In another aspect of the invention, a method for measuring electron density in a lipid layer is provided by providing an x-ray spectrometer comprising an X-ray source; a curved crystal monochromator which focuses a monochromatic x-ray beam onto a sample surface, the curved monochromator comprising the shape of a logarithmic spiral; and a position-sensitive x-ray detector; and providing a sample comprising single or multilamellar lipid layers deposited on a flat substrate; exposing the sample to the focused x-ray beam of the x-ray spectrometer and measuring the diffraction intensity at the position sensitive detector. In preferred embodiments, the sample may be natural or synthetic lipids, natural or synthetic lipids deposited by centrifugation from solution or suspension, a lipid deposited by Langmuir-Blodgett deposition, or a lipid deposited by self-assembly from solution. In yet another aspect of the invention, a method of measuring diffraction intensities from oriented samples in real time is provided by providing an x-ray spectrometer comprising an X-ray source; a curved crystal monochromator which focuses a monochromatic x-ray beam onto a sample surface, the curved monochromator comprising the shape of a logarithmic spiral; and a position-sensitive x-ray detector; and providing a crystallographically oriented sample; exposing the sample to the focused x-ray beam of the x-ray spectrometer; and measuring the diffraction intensity at the position-sensitive detector. |
description | 1. Field of the Invention The invention is in the field of particle sources and more specifically in the field of particle sources configured for medical applications. 2. Related Art It has been shown that high-energy particles can be advantageously used for medical treatment of cancer. These high-energy particles typically have energies greater than 20 MeV (million electron volts). For example, protons with energies between 70 MeV and 250 MeV can be used to deposit energy at a very precise depth within a human body. High-energy protons are generated in a particle accelerator and delivered to a patient at a treatment station. A typical treatment station includes an adjustable gurney or chair configured to position the patient relative to a fixed proton beam. In some instances, the output of the particle accelerator is directed through two alternative paths using particle transport optics such as magnets and electric fields. For example, in one instance a first set of particle transport optics is used to direct protons from above a patient and a second set of particle transport optics is used to direct protons toward the patient from the side at an angle 90 degrees from the first set of particle transport optics. One limitation of this arrangement is that each separate path requires a separate set of expensive particle transport optics and a separate particle beam nozzle. During a treatment the depth of proton penetration and the position of the proton beam may be varied in order to treat a three dimensional volume within a patient. Depth control is achieved by varying the energy of the protons. This variation can be achieved by passing the protons through varying lengths of an energy adsorbing material or by using a particle source capable of generating particles at selectable energies. The proton beam may be applied over an area perpendicular to the depth dimension by either scanning or scattering the proton beam. Scattering or energy variation of the proton beam is optionally performed in more than one stage. For example, a first scattering step may be applied as the protons leave the particle accelerator and a second scattering step may be applied after the protons pass through final beam steering elements. Two steps are required when the final beam steering elements cannot handle a desired final spatial or energetic distribution. A path through which particles are transported typically includes a proton nozzle. Proton nozzles can be designed for special purposes, for example, double scattering nozzles, single scattering nozzles, scanning nozzles, and other specialized nozzles known in the art. Different medical treatments require the use of different proton nozzles each weighing one thousand or more pounds and costing hundreds of thousands of dollars. Changing nozzles is a time consuming and labor intensive process that limits the flexibility of treatments particularly between successive patients and causes system downtime. For the various reasons discussed above, and additional reasons, there is a need for improved sources of high-energy particles. Some embodiments include a particle source coupled to three or more alternative beam paths. These alternative beam paths are configured to direct protons or other nuclei toward a patient from a variety of different directions. These different directions may be significantly greater than or less than 90 degrees apart and may be disposed in different planes. Some embodiments include automated systems and methods of changing particle beam nozzles in a particle beam path and/or changing particle beam nozzles between particle beam paths. For example, various embodiments include a rail system configured to move particle beam nozzles from a first particle beam path to a second particle beam path, from a storage location to a particle beam path, and/or from a first treatment station to a second treatment station. Particle beam nozzles can be moved into position relative to a particle beam path automatically during a treatment session that includes more than one separate particle dosing of a single patient. Thus, a patient can be treated using more than one type of nozzle during a single treatment session. Further, the same nozzle can be used in more than one particle beam path during the same treatment session. Although examples discussed herein are related to proton beams, the illustrated embodiments can be applied to other particle beams such as Helium and Carbon beams, etc. Various embodiments include a system comprising a treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam, and three or more particle beam paths through which the particle beam can be delivered to the patient at the treatment station, the three or more particle beam paths including at least two particle beam paths significantly greater than 90 degrees apart. Various embodiments include a system comprising a treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam for treatment of the patient, and three or more particle beam paths through which the particle beam can be delivered to the patient at the treatment station, the three or more particle beam paths configured such that a first particle beam path is located outside of a plane including a second particle beam path and a third particle beam path. Various embodiments include a method of treating a patient, the method comprising generating a particle beam of high-energy particles, directing the particle beam of high-energy particles along a first beam path, treating the patient using the particle beam of high-energy particles directed along the first particle beam path, selecting a second particle beam path from among a plurality of alternative particle beam paths different from the first particle beam path, at least one of the plurality of alternative particle beam paths being disposed in part beneath the patient or the first particle beam path laying outside of a plane defined by two of the plurality of alternative particle beam paths, and directing the particle beam of high-energy particles along the second particle beam path. Various embodiments include a system comprising a first treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam, a first particle beam path along which the particle beam can be delivered to the patient, a second particle beam path along which the particle beam can be delivered to the patient, a particle beam nozzle configured to modify the particle beam, and a transport system configured to automatically move the particle beam nozzle from the first particle beam path to the second particle beam path. Various embodiments include a method comprising generating a first particle beam of high-energy particles, directing the first particle beam of high-energy particles along a first particle beam path having first particle beam transport optics, modifying the first particle beam of high-energy particles using a particle beam nozzle, treating a first patient using the first particle beam of high-energy particles modified using the particle beam nozzle, selecting a second particle beam path having second particle beam transport optics, moving the particle beam nozzle from the first particle beam path to the second particle beam path under control of a processing unit, generating a second particle beam of high-energy particles, directing the second particle beam of high-energy particles along the second particle beam path, modifying the second particle beam of high energy particles using the particle beam nozzle, and treating the first patient or a second patient using the second particle beam of high-energy particles modified using the particle beam nozzle. Various embodiments include a system comprising a treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam, a first particle beam path through which the particle beam can be delivered to the patient, and a transport system configured to automatically move a first of a plurality of different particle beam nozzles to the first particle beam path, and to separately move at least a second of the plurality of different particle beam nozzles to the first particle beam path. Various embodiments include a method comprising generating a first particle beam of high-energy particles, directing the first particle beam of high-energy particles along a particle beam path, modifying the first particle beam of high-energy particles using a first particle beam nozzle, treating a first patient using the first particle beam of high-energy particles modified using the first particle beam nozzle, exchanging the first particle beam nozzle for a second particle beam nozzle under control of a processing unit, generating a second particle beam of high-energy particles, directing the second particle beam of high-energy particles along the particle beam path, modifying the second particle beam of high energy particles using the second particle beam nozzle, and treating the first patient or a second patient using the second particle beam of high-energy particles modified using the second particle beam nozzle. Various embodiments include a system comprising a first particle beam nozzle configured for use in a first treatment type, a second particle beam nozzle configured for use in a second treatment type, a transport system configured to alternatively position under control of a processing unit the first particle beam nozzle and the second particle beam nozzle between a particle accelerator and a treatment station. Various embodiments include a particle beam nozzle comprising, a mount configured to alternatively hold the particle beam nozzle along each of a plurality of alternative particle beam paths, a positioner configured to automatically position the particle beam nozzle relative to a treatment station or one of the plurality of alternative particle beam paths, and an energy modifier configured to vary an energy of high-energy particles within each of the plurality of alternative particle beam paths. Various embodiments include a particle beam nozzle comprising a conveyance configured to automatically move the particle beam nozzle to a first particle beam path, a coupler configured to hold the particle beam nozzle relative to the first particle beam path, and a beam scanner configured to scan a particle beam of high-energy particles from the first particle beam path. Various embodiments include a system comprising a treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam, and three or more particle beam paths through which the particle beam can be delivered to the patient at the treatment station. Various embodiments include a system comprising a treatment station for particle beam treatment of a patient, a particle accelerator configured to generate a particle beam, and a first particle beam path configured to deliver the particle beam to the patient from beneath the patient. Some embodiments include three or more alternative particle beam paths through which a particle beam can be delivered to a particular treatment station. At least part of each particle beam path typically includes a separate set of particle transport optics such as magnets or electric fields. The three or more particle beam paths may lie in a single plane or in two or more different planes. The three or more particle beam paths may also be configured to arrive at the treatment station at a variety of different angular separations. In some embodiments, the complexity of having multiple particle beam paths is reduced by the inclusion of particle beam nozzles that can be moved from one particle beam path to another automatically, e.g., under control of a processing unit or under the control of a device configured to operate without human intervention. For example, a particular particle beam nozzle may be moved from a first particle beam path to a second particle beam path. This can reduce the number of particle beam nozzles required to support the multiple particle beam paths and thus reduce costs. The movable particle beam nozzles also allow exchange of particle beam nozzles in a particular particle beam path, which exchange may be automated. For example, in some embodiments, a particle beam nozzle preferred for one type of treatment can easily and quickly be exchanged for a particle beam nozzle preferred for another type of treatment. By making this exchange under control of a processing unit, e.g., under the control of a processor, microprocessor, computer, electronic circuit, electronic controller, and/or the like, with or without software running thereon, the exchange can be performed with minimized downtime and even during a treatment session of a particular patient. As is described further herein, example mechanisms used to transport particle beam nozzles may include support, coupling and positioning elements, as well as a rail system, conveyance, gantry, carrier, belt, carrier, carriage, and/or the like. These mechanisms are typically automated, e.g., some or all of their operations are performed without the need for human intervention. Normally, automated mechanisms operate under the control of a processing unit. The processing unit may include logic configured for selecting a specific particle beam nozzle, selecting a specific particle beam path, controlling movement of the particle beam nozzle, positioning the particle beam nozzle, responding to an interlock, opening or closing a shutter, responding to a collision avoidance parameter, receiving data from a treatment plan, accessing a database including patient information, and/or the like. In various embodiments, the mechanisms used to transport particle beam nozzles are configured to move a particle beam nozzle between particle beam paths or exchange particle beam nozzles at a particle beam path in less than 24 hours, 12 hours, 6 hours, 2 hours, 1 hour, 30 minutes, 15 minutes, 10 minutes, or 5 minutes. Various embodiments include the use of movable particle beam nozzles at treatment stations having one, two, three or more particle beam paths. Automated mechanisms for manipulating particle beam nozzles may be contrasted with manual approaches in which changing of a nozzle could take hours or days, and result in system shutdown over an extended period. FIG. 1 is a block diagram of a multi-beam path Treatment System 100, according to various embodiments. Treatment System 100 includes at least a Particle Accelerator 110, a First Beam Path 120A, a First Nozzle 130A, and a Treatment Station 140. In various embodiments, Treatment System 100 includes further particle beam paths, such as a Second Beam Path 120B and a Third Beam Path 120C, and/or additional particle beam nozzles, such as a Second Nozzle 130B and a Third Nozzle 130C. First Nozzle 130A, Second Nozzle 130B and Third Nozzle 130C are optionally movable using a Transport System 150 and/or stored in a Nozzle Storage 160. Particle Accelerator 110 is a source of high-energy particles such as protons, Helium, Carbon, Neon, Argon, and/or some other stable or unstable elemental particle. For example Particle Accelerator 110 can include a cyclotron, synchrotron, linear accelerator, or any other device configured to accelerate particles. In various embodiments, these particles have energy greater than 20, 50, 70, 100, 250 or 500 MeV/u (MeV per nucleon). For example, in one embodiment Particle Accelerator 110 is configured to generate protons with energies between 70 and 250 MeV. These protons are generated in a particle beam having a cross-section as small as 1.0 millimeter (mm), and a kinetic energy distribution as narrow as 1%, 2%, 5%, 20% or 50% of the average particle energy. Such small cross-sections and narrow energy distributions are useful when the particle beam is to be turned or focused. For example, if magnets are used to turn the particle beam, a specific set of magnets will result in a turning radius that is a function of the kinetic energy, particle beams having greater kinetic energy distributions being more difficult to turn without particle loss. However, as discussed further herein, a greater kinetic energy distribution may be desirable when using the particle beam for medical treatment. Therefore, Particle Accelerator 110 optionally includes an energy broadener (e.g., range shifter) and/or a particle beam defocuser (e.g., scatterer) configured to vary the kinetic energy or increase the cross-section of the particle beam, respectively. Such particle beam broadeners and defocusers are known in the art. First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C are each particle beam paths through which the particle beam generated using Particle Accelerator 110 may travel to reach an intersection zone at Treatment Station 140. First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C may each have separate particle beam transport optics. The intersection zone is a zone in which a patient may be placed for treatment and may be a point, area or volume. As is described further herein, First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C may be configured to direct the particle beam to the intersection zone from a variety of different directions and at a variety of different angles. In some embodiments, these different directions and angles advantageously add flexibility to the treatment of a patient at Treatment Station 140. Each of First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C can include a variety of steering magnets, collimating elements, or the like. In some embodiments, each of First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C each include an interface configured to couple with First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C. First Nozzle 130A, Second Nozzle 130B and Third Nozzle 130C are configured to be disposed along First Beam Path 120, Second Beam Path 120B and/or Third Beam Path 120C. In some embodiments, First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C are each configured to be moved from along one particle beam path to along another particle beam path using Transport System 150. For example, Second Nozzle 130B may be moved from along Second Beam Path 120B to along Third Beam Path 120C using Transport System 150. Alternatively, Second Nozzle 130B may be exchanged for First Nozzle 130A along First Beam Path 120A using Transport System 150. Thus, the particle beam nozzles can be moved between particle beam paths and/or a single particle beam path may receive different particle beam nozzles. First Nozzle 130A, Second Nozzle 130B and Third Nozzle 130C are configured to modify the particle beam generated using Particle Accelerator 110, optionally in different ways. These modifications can include, for example, scattering, kinetic energy variation, and/or scanning the particle beam. Different particle beam nozzles may be configured to broaden the kinetic energy distribution or otherwise vary the kinetic energy by different amounts and, thus, control a volume within a patient in which treatment is directed. In various embodiments, First Nozzle 130A, Second Nozzle 130B and Third Nozzle 130C include double scattering nozzles, single scattering nozzles, scanning nozzles, or the like. Treatment Station 140 is configured for treating a patient using the particle beam generated by Particle Accelerator 110 and optionally modified by one of First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C. In some embodiments, Treatment Station 140 includes a patient support such as a platform, harness, chair, gurney, or the like. This patient support typically includes multiple degrees of freedom to position the patient and may be robotic. In some embodiments, Treatment Station 140 includes one or more openings (e.g., removable panels or panels including holes) configured for the particle beam to pass through from beneath the patient. Some embodiments include more than one of Treatment Station 140. These Treatment Station 140 are optionally located in different rooms. Transport System 150 is configured to move First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C, typically under control of a processing unit. Transport System 150 can include, for example, a processing unit configured to receive information from an encoder, mechanical contact, or other position sensor, a data input configured to receive instructions regarding where a particle beam nozzle should be moved, and logic configured to control the movement of a particle beam nozzle from one particle beam path to another, or the like. As described further herein, different particle beam paths may be associated with (e.g., directed toward) different treatment stations. Thus, Transport System 150 is optionally configured for moving First Nozzle 130A from a position relative to a first Treatment Station 140 to a position relative to a second Treatment Station 140. These first and second Treatment Station 140 can optionally be in different rooms. Transport System 150 further includes mechanisms for moving a particle beam nozzle. These mechanisms may include, for example, one or more of, a gantry, a system of one or more rails, a motor, a belt, a screw drive, a chain drive, a carriage, hydraulics, a conveyance such as a conveyor or carriage, and/or the like. For example, in some embodiments, Transport System 150 includes a gantry coupled to one or more of First Nozzle 130A, Second Nozzle 130B and Third Nozzle 130C, and configured to move these particle beam nozzles to positions along First Beam Path 120A, Second Beam Path 120B and/or Third Beam Path 120C. The motion of the gantry is optionally circular, in which case positions for the particle beam nozzles along the various particle beam paths may be distributed in a circular fashion around an intersection zone. In some embodiments, Transport System 150 further includes mechanisms for moving a particle beam nozzle in and out of the gantry. In other embodiments, Transport System 150 includes a rail system comprising one, two or more rails configured to support a carriage. The carriage includes a position sensor configured to determine the position of a particle beam nozzle and optionally a positioner configured to make fine adjustments in the position of the particle beam nozzle relative to a particle beam and/or Treatment Station 140. In various embodiments, Transport System 150 is configured to move a particle beam nozzle from along one particle beam path to along a second particle beam path, or from Nozzle Storage 160 to along a particle beam path in less than 15, 10, 5, 3, or 1 minutes, or less than 45, 30 or 15 seconds. For example, in one specific example, Transport system 150 is configured to move First Nozzle 130A from along First Beam Path 120A to along Second Beam Path 120B, or from Nozzle Storage 160 to along Third Beam Path 120C in less than 15, 10, 5, 3, or 1 minutes, or less than 45, 30 or 15 seconds. Movements between or to other particle beam paths may be accomplished in similar times. Some of these movement times are facilitated by a processing unit included in Transport System 150 and, as such, they may be achieved through automatic movement under processing unit control. Some of these movement times are used to move a particle beam nozzle to more than one particle beam path during a single treatment session without significant downtime. Transport System 150 optionally includes collision avoidance features. For example, in some embodiments Transport System 150 includes a sensor configured to halt movement of a particle beam nozzle when contact is made between the particle beam nozzle and an unexpected object (e.g., a patient or another particle beam nozzle). This sensor may be electrostatic, mechanical, electromagnetic, optical, or the like. Some or all of these collision avoidance features may be included in a particle beam nozzle. In some embodiments these collision avoidance features are configured to halt or otherwise change movement of the particle beam nozzle prior to an undesirable contact. In some embodiments, Transport System 150 includes a robot configured to move First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C to specific positions relative to one or more particle beam path. For example, Transport System 150 may include a robotic manipulator arm configured to move particle beam nozzles from one position to another. This robotic manipulator arm is optionally disposed on a movable support. Optional Nozzle Storage 160 is configured to store First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C when these particle beam nozzles are not disposed along a particle beam path. In some embodiments, Nozzle Storage 160 includes a controlled environment including, for example, a positive pressure atmosphere, or the like. Nozzle Storage 160 optionally includes access for replacement or maintenance of First Nozzle 130A, Second Nozzle 130B and/or Third Nozzle 130C. Nozzle Storage 160 is optionally configured to store particle beam nozzles configured for use at more than one of Treatment Station 140. While FIG. 1 illustrates three particle beam paths and three particle beam nozzles, alternative embodiments can include greater or fewer numbers of each of these features. For example, some embodiments include only one or two particle beam paths, while some embodiments include four or more alternative particle beam paths. Some embodiments include a single particle beam path configured to include several alternative particle beam nozzles and some embodiments include a single particle beam nozzle configured to be included in different particle beam paths. FIG. 2 illustrates a physical layout of Treatment System 100, according to various embodiments. This illustrated physical layout includes up to five alternative particle beam paths including, for example, First Beam Path 120A, Second Beam Path 120B, Third Beam Path 120C, a Fourth Beam Path 120D and a Fifth Beam Path 120E. Each of these particle beam paths optionally include a particle beam nozzle such as First Nozzle 130A, Second Nozzle 130B, Third Nozzle 130C, a Fourth Nozzle 130D and/or a Fifth Nozzle 130E. Fourth Beam Path 120D and Fifth Beam Path 120E include features and characteristics similar to those of First Beam Path 120A. Likewise, Fourth Nozzle 130D and Fifth Nozzle 130E include features and characteristics similar to those of First Nozzle 130A. Alternative embodiments include more than five alternative particle beam paths. The particle beam paths illustrated in FIG. 2 are optionally each configured to be coupled to the particle beam nozzles using an Interface 210. In some embodiments, Interface 210 is configured for the attachment of different particle beam nozzles. For example, in some embodiments, Interface 210 includes one or more guide pins configured for the alignment of a particle beam nozzle, such as First Nozzle 130A. In some embodiments, Interface 210 includes a mechanical, electronic or optical encoder or other position sensor configured for determining the position of a particle beam nozzle, such as First Nozzle 130A. In some embodiments, Interface 210 includes a particle transparent vacuum interface configured for the maintenance of a pressure differential between part of a particle beam path closer to Particle Accelerator 110 and part of the particle beam path closer to Treatment Station 140. Such particle transparent vacuum interfaces are known in the art. Interface 210 optionally includes a shutter configured to be closed to protect the particle transparent vacuum interface when a particle beam nozzle is not disposed in front of a particular Interface 210 and to be opened when a particle beam nozzle is disposed in front of the Interface 210. In these embodiments, each particle transparent vacuum interface is typically protected by either a closed shutter or by a particle beam nozzle. The shutter is optionally automatically opened and closed by the movement of a particle beam nozzle or by Transport System 150. The embodiments of Transport System 100 illustrated in FIG. 2 further include an optional Beam Conditioner 220 configured to modify the diameter and/or kinetic energy of the particle beam generated by Particle Accelerator 110. Beam Conditioner 220 may be disposed before or after the particle beam paths separate. For example, Beam Conditioner 220 may be disposed along all particle beam paths (as illustrated) or may be disposed such that it is only along First Beam Path 120A and/or Second Beam Path 120B. FIG. 2 illustrates one Trifurcation 230 and two Bifurcations 240 of particle beam paths. In alternative embodiments, different patterns of Trifurcations 230, Bifurcations 240 and/or greater divisions are used to generate 3, 4, 5 or more separate particle beam paths. Three or more of First Beam Path 120A, Second Beam Path 120B, Third Beam Path 120C, Fourth Beam Path 120D and Fifth Beam Path 120E are optionally coplanar. For example, in some embodiments all five of these particle beam paths lie in the same plane. In alternative embodiments, two, three or four of these particle beam paths lie in the same plane. For the purposes of this discussion, the plane in which a particle beam path lies, angles between particle beam paths, or other aspects of particle beam path orientation are defined by considering those parts of the particle beam paths between Interface 210 and an Intersection Zone 250 disposed at Treatment Station 140. In some embodiments, two particle beam paths may be approximately collinear and arrive at Treatment Station 140 from opposite directions. Specifically, as illustrated in FIG. 2, First Beam Path 120A and Fifth Beam Path 120E are approximately collinear but arrive at Intersection Zone 250 from opposite directions. In some embodiments, particle beam paths arrive at Intersection Zone 250 at separations of substantially greater than or less than 90 degrees. For example, both Fourth Beam Path 120D and Fifth Beam Path 120E arrive at Intersection Zone 250 at an angle substantially greater than 90 degrees from First Beam Path 120A, while First Beam Path 120A and Second Beam Path 120B arrive at Intersection Zone 250 at an angle substantially less than 90 degrees. Substantially less than 90 degrees includes less than approximately 80 degrees in some embodiments, less than 70 degrees in further embodiments, and less than 60 degrees in still further embodiments. Substantially greater than 90 degrees includes more than 100 degrees in some embodiments, more than 110 degrees in further embodiments, and more than 120 degrees in still further embodiments. For example, in various embodiments these particle beam paths arrive at angles of at least 100, 110, 120 or 135 degrees. In various embodiments, these particle beam paths arrive at angles of less than 35, 50, 60, 70 or 80 degrees. In some embodiments, pairs of particle beam paths arrive at Intersection Zone 250 with different angular separations. For example, First Beam Path 120A and Third Beam Path 120C arrive at Intersection Zone 250 with an angular separation of approximately 45 degrees, while Second Beam Path 120B and Fifth Beam Path 120E arrive at Intersection Zone 250 with an angular separation of approximately 135 degrees. In some embodiments, three different particle beam paths arrive at Intersection Zone 250 spaced approximately 120 degrees from each other. In some embodiments, particle beam paths arrive at Intersection Zone 250 from both below and above a patient. For example, First Beam Path 120A arrives at Intersection Zone 250 from above while Fourth Beam path 120D and Fifth Beam Path 120E arrive at Intersection Zone 250 from below. Part of Fifth Beam Path 120E is, thus, disposed beneath the patient. In some embodiments, Treatment Station 140 includes openings to allow passage of a particle beam from below Treatment Station 140. FIG. 3 illustrates an alternative physical layout of Treatment System 100 in a perspective view, according to various embodiments. In these embodiments, three or more particle beam paths arrive at Intersection Zone 250 from along at least two different planes. Specifically, Third Beam Path 120C is not within a plane defined by First Beam Path 120A and Second Beam Path 120B. In some embodiments, First Beam Path 120A, Second Beam Path 120B and Third Beam Path 120C each arrive at Intersection Zone 250 at approximately 90 degrees of each other, as illustrated in FIG. 3. FIG. 4 illustrates an alternative layout of Treatment System 100, according to various embodiments. These embodiments include a Rail System 410 configured to position First Nozzle 130A, Second Nozzle 130B, Third Nozzle 130C, Fourth Nozzle 130D and/or Fifth Nozzle 130E relative to First Beam Path 120A, Second Beam Path 120B, Third Beam Path 120C, Fourth Beam Path 120D and/or Fifth Beam Path 120E. Rail System 410 is optionally further configured to move particle beam nozzles to and from Nozzle Storage 160. In various embodiments, Rail System 410 is included in Transport System 150 and includes a conveyance, a track, a gantry, a system of one or more rails, a motor, a belt, a screw drive, a chain drive, a carriage, hydraulics, and/or the like FIG. 5 illustrates a physical layout of Treatment System 100 including more than one Treatment Station 140, according to various embodiments of the invention. In these embodiments, different particle beam paths are optionally associated with different treatment stations. For example, as illustrated in FIG. 5, First Beam Path 120A is configured for treating a patient at a first Treatment Station 140A, while Second Beam Path 120B and Third Beam Path 120C are configured for treating a patient at a second Treatment Station 140B. A Barrier 520, such as a radiation shield or wall, optionally separates Treatment Station 140A and Treatment Station 140B. Thus, Treatment Station 140A and Treatment Station 140B may be in different rooms. Rail System 410 is optionally configured for moving particle beam nozzles between particle beam paths associated with different treatment stations and/or between different rooms. Each of these rooms optionally include one, two, three or more particle beam paths. Rail System 410 optionally includes a Switch 530 configured for moving particle beam nozzles to alternative paths of Transport System 150. FIG. 6 is a block diagram of a Particle Beam Nozzle 600, according to various embodiments. Particle Beam Nozzle 600 may be included in embodiments of particle beam nozzles, such as First Nozzle 130A, Second Nozzle 130B, Third Nozzle 130C, Fourth Nozzle 130D and/or Fifth Nozzle 130E. Particle Beam Nozzle 600 includes one or more of a Mount 610, an optional Conveyance 620, an optional Positioner 630, and an optional Coupler 640. These features are used for moving, supporting and positioning Particle Beam Nozzle 600 relative to a particle beam path or Treatment Station 140. Mount 610 is a supporting structure of Particle Beam Nozzle 600 that connects Particle Beam Nozzle 600 to Transport System 150. For example, Mount 610 may include a flange, bolting hardware, guide pins, or the like. In some embodiments, Mount 610 is connected directly to Transport System 150, and in other embodiments, Mount 610 is coupled to Transport System 150 via Conveyance 620. Conveyance 620 is a vehicle, carriage, cart, trolley, movable platform, or the like, configured to move along Transport System 150. For example, in some embodiments, Conveyance 620 is a self-propelled rail car configured to be coupled to one or more rails of Transport System 150. Conveyance 620 can include a position sensor configured to determine its position along Transport System 150. In some embodiments, Conveyance 620 is part of Transport System 150 rather than Particle Beam Nozzle 600. Positioner 630 is configured for making fine adjustments in the position of Particle Beam Nozzle 600 relative to a particle beam path or Treatment Station 140. In various embodiments, Positioner 630 is configured to position Particle Beam Nozzle 600 to an accuracy of 0.005, 0.01, 0.1, 0.2, 0.5 or 1.0 mm. Positioner 630 may include stepper motors, hydraulics, piezoelectric devices (PZTs) or the like. For example, in some embodiments, Positioner 630 is configured to move Particle Beam Nozzle 600 using a combination of stepper motors and hydraulics. In some embodiments, Particle Beam Nozzle 600 is first moved along a particle beam path using Conveyance 620 and then more precisely positioned using Positioner 630. Coupler 640 is configured to attach Particle Beam Nozzle 600 to Interface 210. For example, Coupler 640 may include locking mechanisms, clamps, guide pins, bolts, or the like. In some embodiments, Coupler 640 is configured to assure that Particle Beam Nozzle 600 is precisely positioned. Coupler 640 is optional, for example, in embodiments wherein Particle Beam Nozzle 600 does not make physical contact with Interface 210 or parts of a particle beam path. In some embodiments, Coupler 640 is configured to be moved relative to other parts of Particle Beam Nozzle 600. For example, in one embodiment, Coupler 640 is configured to move relative to Conveyance 620. As such, Conveyance 620 may be used to move Particle Beam Nozzle 600 close to Interface 210 and then while Conveyance 620 is held in a fixed position, Coupler 640 may be moved to attach to Interface 210. By moving Coupler 640 independently from Conveyance 620, Coupler 640 has the freedom of movement to respond to guide pins or other alignment features when attaching to Interface 210. Particle Beam Nozzle 600 optionally includes an Environmental Control 650 configured for controlling an environment within part of Particle Beam Nozzle 600. Environmental Control 650 may be configured to maintain part of Particle Beam Nozzle 600 at a reduced pressure, to maintain part of Particle Beam Nozzle 600 in a Helium atmosphere, or the like. Typically, Environmental Control 650 is moved between particle beam paths along with other parts of Particle Beam Nozzle 600. Particle Beam Nozzle 600 optionally includes a Shutter Control 660 configured to open and close a shutter included in Interface 210. For example, Shutter Control 660 may be configured to open a shutter when Particle Beam Nozzle 600 is moved along a particle beam path and to close the shutter when Particle Beam Nozzle 600 is moved out of the particle beam path. Shutter Control 660 can be a mechanical, electrical or optical mechanism. For example, Shutter Control 660 may include a protrusion configured to physically move the shutter when Coupler 640 is connected to Interface 210. Shutter Control 660 may include an electrical connection, a radio frequency identification (RFID) tag or bar code detectable by Interface 210 and configured to cause Interface 210 to move the shutter. Shutter Control 660 is optionally configured to be responsive to an interlock discussed elsewhere herein. Particle Beam Nozzle 600 optionally includes a Collimator 670. Collimator 670 may be a multi-leaf collimator, a micro multi-leaf collimator a fixed collimator, or the like. In some embodiments, Transport System 150 is configured to move Collimator 670 independently from other parts of Particle Beam Nozzle 600. Thus, Transport System 150 is optionally configured to move Collimator 670 into and out of a particle beam path separately from Particle Beam Nozzle 600. In some embodiments, Nozzle Storage 160 is configured for insertion of Collimator 670 into a particle beam nozzle. Particle Beam Nozzle 600 optionally includes a Beam Scanner 680 configured to scan a particle beam in a zone close to Treatment Station 140. Particle Beam Nozzle 600 optionally includes an Energy Modifier 690 configured to vary the kinetic energy of a particle beam. Energy Modifier 690 may be configured to vary the kinetic energy in a spatial and/or time dependent manner. For example, in some embodiments, Energy Modifier 690 includes a bolus configured for controlling the energy of particles in a spatial manner. Collimator 670, Beam Scanner 680 and/or Energy Modifier 690 are optionally movable independently of other parts of Particle Beam Nozzle 600 using Transport System 150. As such, they may be automatically added to or removed from Particle Beam Nozzle 600. In some embodiments, Nozzle Storage 160 includes features configured for performing this automatic addition or removal. In some embodiments, Particle Beam Nozzle 600 includes an Interlock 695 configured to ensure that the proper particle beam nozzle is disposed between Particle Accelerator 110 and Interaction Zone 250, and/or configured to ensure that the particle beam nozzle is properly positioned relative to a particle beam path. Interlock 695 can be mechanical, electrical, magnetic, optical, and/or the like. In some embodiments, Interlock 695 includes identifying features configured to identify the particle beam nozzle. This identifying feature may include a bar code, radio frequency identifying tag, electronic circuit, electronic characteristic, magnetic characteristic, optical characteristic, and/or the like. In some embodiments, all or part of Interlock 695 is included in Transport System 150, Interface 210 and/or other parts of Treatment System 100. In some embodiments, Interlock 695 is configured to assure that a correct component, such as Collimator 670, Beam Scanner 680, and/or Energy Modifier 690, is within a particle beam nozzle as required by a treatment plan for a specific patient. Interlock 695 optionally uses patient identity information for this purpose. For example, Interlock 695 may be configured to receive data from a barcode or radio frequency identification tag worn by a patient or from a database system before applying a particle beam to the patient. Particle Beam Nozzle 600 optionally includes a plurality of interlocks such as Interlock 695. For example, a first Interlock 695 configured to assure that a shutter is closed before moving a particle beam nozzle, a second Interlock 695 configured to assure that a correct particle beam nozzle is placed along a particle beam path, and a third Interlock 695 configured to assure that the proper patient is positioned at Treatment Station 140. Particle Beam Nozzle 600 optionally includes a Collision Avoidance Feature 698 discussed elsewhere herein. FIG. 7 illustrates a method of operating a treatment system including a plurality of alternative particle beam paths, according to various embodiments. In this method, a beam of high-energy particles is directed through a plurality of alternative particle beam paths in order to treat one or more patients. In a Generate Beam Step 710, a particle beam is generated using Particle Accelerator 110. This particle beam may include protons at high energies, e.g., greater than 20 MeV. Alternatively, this particle beam may include Helium, Carbon or other types of nuclei. In a Direct Beam Step 720, the particle beam generated in Generate Beam Step 710 is directed along a first particle beam path, such as First Beam Path 120A. The first particle beam path typically includes electric or magnetic fields and/or other particle transport optics configured to steer the particle beam toward Treatment Station 140. In a Treat Patient Step 730, the particle beam directed along the first particle beam path is used to treat a patient at the treatment station. In a Select Second Path Step 740, a second particle beam path is selected from among a plurality of alternative particle beam paths different from the first particle beam path. The plurality of alternative particle beam paths may include, for example, Second Particle Beam Path 120B, Third Particle Beam Path 120C, and/or other particle beam paths discussed herein. One of the alternative particle beam paths is optionally significantly more that 90 degrees from the first particle beam path. For example, in one embodiment one of the alternative particle beam paths is 110 degrees or greater from the first particle beam path. One of the alternative particle beam paths is optionally configured to arrive at the treatment station from below the patient. In a Direct Beam Step 750, the particle beam generated in Generate Beam Step 710 is directed through the particle beam path selected in Select Second Path Step 740. In a Treat Patient Step 760, the patient is treated using the particle beam directed through the selected particle beam path. The patient is optionally moved between Treat Patient Step 730 and Treat Patient Step 760. In some embodiments, different patients are treated in Treat Patient Step 730 and Treat Patient Step 760 FIG. 8 illustrates a method of operating a treatment system including a particle beam nozzle configured to be moved between alternative particle beam paths, according to various embodiments. In this method, a particle beam nozzle is moved from one particle beam path to another particle beam path for the treatment of a patient. In a Generate Beam Step 810, a particle beam is generated. Generate Beam Step 810 is an embodiment of Generate Beam Step 710. In a Direct Beam Step 820, the particle beam generated in Generated Beam Step 810 is directed through a first particle beam path such as First Beam Path 120A. In a Modify Beam Step 830, the particle beam directed through the first particle beam path in Direct Beam Step 820 is modified using a particle beam nozzle, such as First Nozzle 130A. This modification can include changes in direction, kinetic energy, dispersion, beam diameter, or the like. For example, in one embodiment, the modification includes changing the direction of the particle beam in order to scan the particle beam over a treatment zone. In a Treat Patient Step 840, the particle beam modified in Modify Beam Step 830 is used to treat a patient at a treatment station. In a Select Second Beam Path Step 850, a second particle beam path is selected. This particle beam path may be directed at the same treatment station as the first particle beam path, or at a different treatment station. The second particle beam path may be, for example, Second Beam Path 120B or Third Beam Path 120C, or other particle beam paths discussed herein. In a Move Nozzle Step 860, the particle beam nozzle used to modify the particle beam in Modify Beam Step 830 is moved to the second particle beam path using Transport System 150. This movement is optionally performed using a processing unit. For example, a processing unit may be used to control the movement using the transport system and/or the processing unit may be used to position the particle beam nozzle precisely relative to the second particle beam path. The particle beam nozzle is optionally moved from one room to another room. The particle beam nozzle is optionally moved from above a patient below a patient. In a repeat of Generate Beam Step 810, the particle beam is again generated. In some embodiments, a single particle beam is generated continuously throughout the steps illustrated in FIG. 8. Thus, the repeat of Generate Beam Step 810 may be a continuation of the first Generate Beam Step 810. In other embodiments, the generation of a particle beam is halted during at least Move Nozzle Step 860, and then the particle beam is again generated in the repeat of Generate Beam Step 810. In a Direct Beam Step 870, the particle beam is directed through the second particle beam path. In a Modify Beam Step 880, the particle beam directed through the second particle beam path is modified using the particle beam nozzle. This modification is optionally the same as the modification of Modify Beam Step 830 and is made using the particle beam nozzle moved in Move Nozzle Step 860. In a Treat Patient Step 890, a patient is treated using the particle beam modified in Modify Beam Step 880. This patient may be the same patient treated in Treat Patient Step 840 or a different patient. If the same patient, then the patient is optionally moved between Treat Patient Step 840 and Treat Patient Step 890. In various embodiments, Treat Patient Step 840 and Treat patient Step 890 occur within 15, 10, 5 or 2 minutes of each other. FIG. 9 illustrates a method of operating a treatment system including a particle beam path configured to receive a plurality of alternative particle beam nozzles. In this method, two different particle beam nozzles are used to modify a particle beam passed through a single particle beam path. In a Generate Beam Step 910, a first particle beam is generated. Generate Beam Step 910 is an embodiment of Generate Beam Step 710. In a Direct Beam Step 920, the first particle beam generated in Generate Beam Step 910 is directed through a particle beam path, such as First Beam Path 120A. In a Modify Beam Step 930, the particle beam directed through a particle beam path in Direct Beam Step 920 is modified using a first particle beam nozzle, such as First Nozzle 130A. This modification may include changes in direction, dispersion, kinetic energy, beam diameter, or the like. For example, in one embodiment, the modification includes changing the kinetic energy in order to control a depth of treatment. In a Treat Patient Step 940, a patient is treated using the particle beam modified in Modify Beam Step 930. In an Exchange Nozzle Step 950, the first particle beam nozzle is exchanged for a second particle beam nozzle, such as Second Nozzle 130B, using Transport System 150. In some embodiments, this exchange is made while the patient is at Treatment Station 140. In various embodiments, this exchange is made in less than 15, 10, 5, or 2 minutes. The second particle beam nozzle is typically configured to modify the particle beam in a different manner or to a different degree than the first particle beam nozzle. Exchange Nozzle Step 950 is optionally performed under control of a processing unit. For example, in some embodiments, a processing unit is used to remove the first particle beam nozzle from the particle beam path and/or a processing unit is used to assure that the second particle beam nozzle is positioned correctly in the particle beam path. In a repetition of Generate Beam Step 910, a second particle beam is generated using Particle Accelerator 110. As with the repeat of Generate Beam Step 810, the repetition of Generate Beam Step 910 may be an interrupted or uninterrupted continuation of the first Generate Beam Step 910. In a Direct Beam Step 960, the second particle beam is directed through the particle beam path. In a Modify Beam Step 970, the second particle beam is modified using the second particle beam nozzle. This modification may include changes in direction, dispersion, kinetic energy, beam diameter, or the like. In a Treat Patient Step 980, the particle beam modified in Modify Beam Step 970 is used to treat a patient. This patient may be the same patient treated in Treat Patient Step 940 or a different patient. If the same patient, the patient is optionally moved between Treat Patient Step 940 and Treat Patient Step 980. Several embodiments are specifically illustrated and/or described herein. However, it will be appreciated that modifications and variations are covered by the above teachings and within the scope of the appended claims without departing from the spirit and intended scope thereof. For example, while processing unit control of Transport System 150 is discussed herein, all or part of Transport System 150 may be manual. Further, the labeling of particle beam paths and particle beam nozzles within the figures is for illustrative purposes only. Thus, attributes applied to one particle beam path or one particle beam nozzle may be applied to other particle beam paths or other particle beam nozzles. For example, while First Beam Path 120A is shown as coming from above Treatment Station 140 and Beam Path 120E is shown as coming from below Treatment Station 140, these labels and/or positions may be reversed. The embodiments discussed herein are illustrative of the present invention. As these embodiments of the present invention are described with reference to illustrations, various modifications or adaptations of the methods and or specific structures described may become apparent to those skilled in the art. All such modifications, adaptations, or variations that rely upon the teachings of the present invention, and through which these teachings have advanced the art, are considered to be within the spirit and scope of the present invention. Hence, these descriptions and drawings should not be considered in a limiting sense, as it is understood that the present invention is in no way limited to only the embodiments illustrated. In general, features or aspects shown or discussed in relation to one embodiment are not limited to that embodiment and can be used in different embodiments, and each embodiment need not contain each feature shown or described in relation to that embodiment. |
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claims | 1. Apparatus for imaging a body, comprising:a camera head positioned in a location and configured to detect radiation emitted from regions in the body so as to produce sliced images of radiation intensity emitted from a three dimensional structure of the regions, the camera head comprising:a two-dimensional array of D CdZnTe (CZT) detector elements, wherein D is an integer greater than 1, the array of D detector elements being defined by two orthogonal repetition vectors, the elements being mounted in the camera head and being respectively coupled to D electrodes, each electrode being configured to output signals indicative of intensities of radiation that are incident on a given detector element; andD adjustable collimators disposed respectively, in a collimator array defined by two orthogonal repetition vectors, in registration between the D detector elements and the body so as to define respective regions of the body from which the radiation emitted is incident on the detector elements, each of the adjustable collimators having N dimensional configurations defining different respective volumes of each of the regions, wherein N is an integer greater than 1, wherein each of the collimators is in registration with each of the detector elements so that the emitted radiation from the body traverses the collimators to impinge on the detector elements and wherein for each collimator the respective volumes of each of the regions have a respective common axis of symmetry which is normal to the two orthogonal repetition vectors, wherein each collimator comprises a respective different cavity which is configured to receive liquid opaque to the radiation; anda processor which is configured:to receive DN signals from the D electrodes corresponding to the respective volumes of each collimator while the camera head is in the location and while the respective common axis of symmetry of each collimator is fixed and while the adjustable collimators are in the N different dimensional configurations,to process the signals in order to form a three dimensional image of the body comprising a number of volume elements from which the radiation is emitted, andto produce sliced images of the three dimensional image while the two-dimensional array of CZT detector elements and the respective common axis of symmetry of each collimator are fixed with respect to the body,wherein the number of volume elements is a function of D and N. 2. The apparatus according to claim 1, wherein each collimator comprises a first collimator channel aligned with a second collimator channel and separated therefrom by an adjustable gap. 3. The apparatus according to claim 2, wherein the first collimator channel is aligned with one of the detector elements and is separated therefrom by a variable gap. 4. The apparatus according to claim 3, wherein the processor is coupled to adjust at least one of the variable gap and the adjustable gap. 5. The apparatus according to claim 2, wherein the first and second collimator channels comprise different cross-sectional areas. 6. The apparatus according to claim 1, wherein the liquid comprises mercury. 7. The apparatus according to claim 1, wherein each respective different cavity alters a length of each collimator on receipt of the liquid. 8. The apparatus according to claim 1, wherein each respective cavity comprises a different first cylinder and a different second cylinder having a different cross- section from the respective different first cylinder, and wherein each collimator on receipt of the liquid changes from the respective different first cylinder to the respective different second cylinder. 9. The apparatus according to claim 1, wherein the emitted radiation comprises gamma rays. 10. The apparatus according to claim 1, wherein the processor is configured to generate a representation of radioisotopes in the body in response to an intensity of the radiation. 11. The apparatus according to claim 1, wherein the processor is coupled to compute the number of the volume elements iteratively, so as to determine a largest number of the volume elements. 12. The apparatus according to claim 1, wherein the different respective volumes comprise respective first volumes and respective second volumes, and wherein the respective first volumes include the respective second volumes. 13. The apparatus according to claim 12, wherein the respective first volumes comprise respective first conic volumes, and wherein the respective second volumes comprise respective second conic volumes concentric with the respective first conic volumes. 14. The apparatus according to claim 1, wherein each respective different cavity comprises a plurality of isolated sub-compartments, each sub-compartment being independently filled with the liquid to form the N dimensional configuration, and so that each respective different cavity is operative in the N dimensional configurations in non-horizontal orientations. 15. The apparatus according to claim 1, wherein the volume elements are uniquely identifiable by an ordered triple (i, j, k) wherein i, j, k are positive integers having respective values 1, . . . , I; 1, . . . , J; and 1, . . . , K; and wherein a number IJK of volume elements is a function of DN proportionality coefficients βi,j,kd,n for each of the volume elements, wherein d, n are positive integers respectively identifying a particular detector element and a particular configuration of the adjustable collimators, d, n having respective values 1, . . . , D; 1, . . . N, each proportionality coefficient being a ratio between the emitted radiation from a given volume element of the body and the radiation received by a given detector element for a respective collimator having a given dimensional configuration. 16. The apparatus according to claim 15, wherein βi,j,kd,n=αi,j,kd,n·θi,j,kd,n·Vi,j,kd,n wherein:αi,j,kd,n is an attenuation factor between volume element (i, j, k) and detector d in collimator dimensional configuration n, θi,j,kd,n is a solid angle in which volume element (i, j, k) is viewed from detector d in collimator dimensional configuration n, and Vi,j,kd,n is a volume fraction of volume element (i, j, k) enclosed by the body as viewed from detector d in collimator dimensional configuration n. 17. The apparatus according to claim 16, wherein the processor is configured:to calculate values of βi,j,kd,n, αi,j,kd,n, θi,j,kd,n, and Vi,j,kd,n in response to geometrical relations between positions of the body, the D detector elements and the N dimensional configurations of the collimators, and wherein processing the signals comprises evaluating DN simultaneous linear equations: S d , n = ∑ i , j , k α i , j , k d , n · C i , j , k · θ i , j , k d , n · V i , j , k d , n wherein Sd,n is a total radiation intensity received by detector d in collimator configuration n, andwherein Ci,j,k is an average radioisotope concentration in volume element (i, j, k). 18. A method for imaging a body, comprising:configuring a camera head to detect radiation emitted from regions in the body so as to produce sliced images of radiation intensity emitted from a three dimensional structure of the regions, the camera head comprising:a two-dimensional array of D CdZnTe (CZT) detector elements, wherein D is an integer greater than 1, the array of D detector elements being defined by two orthogonal repetition vectors and being mounted in the camera head, the elements being respectively coupled to D electrodes, each electrode being configured to output signals indicative of intensities of radiation that are incident on a given detector element, andD adjustable collimators respectively disposed, in a collimator array defined by the two orthogonal repetition vectors, in registration between the D detector elements and the body so as to define respective regions of the body from which the radiation emitted is incident on the detector elements, each of the adjustable collimators having N dimensional configurations defining different respective volumes of each of the regions, wherein N is an integer greater than 1, and wherein each of the collimators is in registration with each of the detector elements so that the emitted radiation from the body traverses the collimators to impinge on the detector elements, wherein for each collimator the respective volumes of each of the regions have a respective common axis of symmetry which is normal to the two orthogonal repetition vectors, and wherein each collimator comprises a respective different cavity which is configured to receive a liquid opaque to the radiation;positioning the camera in a location;receiving DN signals from the electrodes corresponding to the respective volumes of each collimator while the camera head is in the location and while the respective common axis of symmetry of each collimator is fixed and while the adjustable collimators are in the N different dimensional configuration;processing the signals in order to form a three dimensional image of the body comprising a number of volume elements from which the radiation is emitted, wherein the number is a function of D and N; andproducing sliced images of the three dimensional image while the two-dimensional array of CZT detector elements and the respective common axis of symmetry of each collimator are fixed with respect to the body. 19. The method according to claim 18, wherein each collimator comprises a first collimator channel aligned with a second collimator channel and separated therefrom by an adjustable gap. 20. The method according to claim 19, wherein the first collimator channel is aligned with one of the detector elements and is separated therefrom by a variable gap. 21. The method according to claim 20, and comprising adjusting at least one of the variable gap and the adjustable gap. 22. The method according to claim 19, wherein the first and second collimator channels comprise different cross-sectional areas. 23. The method according to claim 18, wherein the liquid comprises mercury. 24. The method according to claim 18, wherein each respective different cavity alters a length of each collimator on receipt of the liquid. 25. The method according to claim 18, wherein each respective different cavity comprises a respective different first cylinder and a respective different second cylinder having a different cross-section from the respective different first cylinder, and wherein on receipt of the liquid, each collimator changes from the respective different first cylinder to the respective different second cylinder. 26. The method according to claim 18, wherein the emitted radiation comprises gamma rays. 27. The method according to claim 18, and comprising generating a representation of radioisotopes in the body in response to an intensity of the radiation. 28. The method according to claim 18, wherein processing the signals comprises computing the number iteratively, so as to determine a largest number of the volume elements. 29. The method according to claim 18, wherein each respective different cavity comprises a plurality of isolated sub-compartments, each sub-compartment being independently filled with the liquid to form the N dimensional configurations, so that each respective different cavity is operative in the N dimensional configurations in non-horizontal orientations. 30. The method according to claim 18, wherein the different respective volumes comprise respective first volumes and respective second volumes, and wherein the respective first volumes include the respective second volumes. 31. The method according to claim 30, wherein the respective first volumes comprise respective first conic volumes, and wherein the respective second volumes comprise respective second conic volumes concentric with the respective first conic volumes. 32. The method according to claim 18, wherein the volume elements are uniquely identifiable by an ordered triple (i, j, k) wherein i, j, k are positive integers having respective values 1, . . . , I; 1, . . . , J; and 1, . . . , K; and wherein a number IJK of volume elements is a function of DN proportionality coefficients βi,j,kd,n for each of the volume elements, wherein d, n are positive integers respectively identifying a particular detector element and a particular configuration of the adjustable collimators, d, n having respective values 1, . . . , D; 1, . . . N, each proportionality coefficient being a ratio between the emitted radiation from a given volume element of the body and the radiation received by a given detector element for a respective collimator having a given dimensional configuration. 33. The method according to claim 32, wherein βi,j,kd,n=αi,j,kd,n·θi,j,kd,n·Vi,j,kd,n wherein:αi,j,kd,n is an attenuation factor between volume element (i, j, k) and detector d in collimator dimensional configuration n,θi,j,kd,n is a solid angle in which volume element (i, j, k) is viewed from detector d in collimator dimensional configuration n, andVi,j,kd,n is a volume fraction of volume element (i, j, k) enclosed by the body as viewed from detector d in collimator dimensional configuration n. 34. The method according to claim 33, and comprising:calculating values of βi,j,kd,n, αi,j,kd,n, θi,j,kd,n, and Vi,j,kd,n in response to geometrical relations between positions of the body, the D detector elements and the N dimensional configurations of the collimators,and wherein processing the signals comprises evaluating DN simultaneous linear equations: S d , n = ∑ i , j , k α i , j , k d , n · C i , j , k · θ i , j , k d , n · V i , j , k d , n wherein Sd,n is a total radiation intensity received by detector d in collimator configuration n, andwherein Ci,j,k is an average radioisotope concentration in volume element (i, j, k). |
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046831150 | description | Referring now to the drawing and first, particularly, to FIG. 1 thereof, there is shown a square, grid-shaped spacer, formed of a nickel-iron alloy in a nuclear reactor fuel assembly according to the invention, the spacer being mode up of two flat, planar outer straps 2 and 3 which are disposed at right angles to one another. On the inside of these outer straps 2 and 3, inner straps 21 to 23 are arranged which are parallel to the outer strap 2, and inner straps 31 to 33 which are parallel to the outer strap 3 and which mutually intersect at right angles, forming square grid meshes. In these meshes, respectively, a control rod guide tube or a nuclear fuel-containing fuel rod of the nuclear reactor fuel assembly is arranged, the inner and outer straps being disposed perpendicularly thereto. In the interest of greater clarity, only a single fuel rod 4 is shown in the outermost square corner grid mesh 5 formed by the outer straps 2 and 3 of the spacer, the lateral surfaces of the inner and outer straps being parallel to the longitudinal direction of the fuel rod 4 i.e. the inner and outer straps 2 and 3 being arranged on edge. Inside the square grid meshes, the inner straps 21 to 23 and 31 to 33 are formed with firm bumps 311, 321 and 331 as well as 211, 221 and 231 and are provided with springs corresponding to inwardly-directed springs 6 of the outer strap 2 and inwardly-directed springs 7 of the outer strap 3. The bumps and springs provide the grid-shaped spacer with a positive locking support for the control rod guide tubes extending through individual grid meshes, and a positive locking support in the spacer for fuel rods extended through other grid meshes. At the edges extending in the spacer perpendicularly to the mutually parallel control-rod guide tubes and fuel rods, the outer straps 2 and 3 have slightly inwardly inclined rejection tabs 8 and 9. The outer straps 2 and 3 are connected to one another by means of an intermediate strip 10 which is located between these two outer straps 2 and 3 at the outer edge of the spacer, that outer edge being parallel to the mutually parallel control rod guide tubes and fuel rods, the intermediate strip 10 being likewise parallel to these control rod guide tubes and fuel rods. This intermediate strip 10 is formed by a fishplate on the outer strap 2 which overlaps with a corresponding fishplate on the outer strap 3 and is welded at an inner side thereof to the fishplate on the outer strap 3. The intermediate strip 10 is flat and planar and is inclined relative to the two outer straps 2 and 3, respectively, at an angle of 45.degree. i.e. this intermediate strip 10 is perpendicular to the diagonal of the corner grid mesh 5 and, thereby, to the spacer diagonal between the two outer straps 2 and 3. On the outside of this intermediate strip 10, in the middle between the ends thereof, at the edges of the outer straps 2 and 3 perpendicular to the control rod guide tubes and the fuel rods in the spacer, a rejection or refusal rise 11 is provided which is formed by being embossed or stamped out of the fishplate on the outer strap 2 in the respective directions of the diagonal of the corner grid mesh 5 and the spacer diagonal between the two outer straps 2 and 3. This rejection rise 11 is inclined transversely to the two outer straps 2 and 3 and towards the two ends of the intermediate strip 10 i.e. in longitudinal direction of the control rod guide tubes and the fuel rods contained in the spacer, forming respective chamfered surfaces 12. In addition, the two outer straps 2 and 3 are inclined at the corners, starting from the edges thereof perpendicularly to the control rod guide tubes and fuel rods contained in the spacer and towards the intermediate strip 10 located between these outer straps 2 and 3, forming thereby chamfers 13 and 14 of like inclination. As is shown in FIG. 2, the rejection rises 11 alternatingly engage the respective fuel rods 4 in the corner grid mesh 5 if there is relative movement in longitudinal direction of two parallel nuclear reactor fuel assemblies arranged diagonally adjacent one another in the checkerboard pattern with spacers constructed in accordance with FIG. 1, and thereby force the two intermediate strips 10 apart to such an extent that the two intermediate strips 10 of the two diagonally adjacent spacers, and consequently the nuclear reactor fuel assemblies, cannot become hooked at the ends thereof when there is relative movement of the fuel rods 4 in the longitudinal direction. In FIG. 3, like parts of the grid-shaped square spacer are identified by the same reference characters as in FIG. 1, but the bumps and springs in the individual grid meshes are not shown. The spacer according to FIG. 3 differs from that of FIG. 1 in that the intermediate strip 10 is formed flat and planar between the two outer straps 2 and 3 which are welded together therewith, and in that a spring strip 15 is suspended from the ends of this intermediate strip 10 which are located at the edges of the outer straps 2 and 3 which are perpendicular to the control rod guide tubes and fuel rods in the spacer. This spring strip 15 is formed of spring steel and covers the intermediate strip 10. Furthermore, this spring strip 15 has, in the middle, between the ends thereof, a rejection rise 11 which, like the rejection rise 11 of the spacer according to FIG. 1 projects outwardly, respectively, in the diagonal of the corner grid mesh 5 and in the spacer diagonal between the two outer straps 2 and 3, and is inclined transversely to the two outer straps 2 and 3, forming chamfered surfaces 12 in longitudinal direction of the fuel rod 4. This spring strip 15 can be retrofitted or applied afterwards to the previously manufactured grid-shaped spacer and is therefore well suited for retrofitting previously completed and already available nuclear reactor fuel assemblies. In the square spacer according to FIG. 4, like parts are also provided with the same reference characters as in the spacer according to FIG. 1. The spacer according to FIG. 4 differs from that of FIG. 1 in that the rejection rise 11 is formed by a tab 16 which is cut out in the middle of the fishplate forming the intermediate strip 10 on the outer strap 2. This tab 16 has, at both lateral edges thereof, a respective pointed cutout 31, 32. Respective points of these two cutouts 31 and 32 lie on a bending or folding edge 33 which is parallel to the rods in the spacer i.e. especially also to the control rod guide tube 4 in the corner grid mesh 5. At this folding edge 33, the tab 16 is bent or folded perpendicularly. The bent-away end of the tab 16 is soldered or welded to the outer side of the other outer strap 3 in a pre-stamped or pre-embossed depression 34 formed in the outer strip 3. The one part of the tab 16 which is formed on the one outer strap 2 and is limited by the bending or folding edge 33 lies in the plane of this outer strap 2, while the other part of the tab 16 virtually lies in the plane of the other outer strap 3 starting from the bending or folding edge 33. In the spacer according to FIG. 4, embossing of the rejection rise 11 is therefore avoided, which is particularly advantageous if this spacer is formed of a zirconium alloy. The chamfers or inclinations of the rejection rise 11 of FIG. 4 are formed by the edges of the tab 16 in the cutouts 31 and 32. Advantageously, a rectilinear inclination 13 and 14 is continued, in the spacer according to FIG. 4 at the corners of the two outer straps 2 and 3, into the intermediate strips 10, forming a pointed cutout 19 at the two ends of this intermediate strip 10, so that thereat, a further improvement in the sliding-off of a spacer on a diagonally adjacent nuclear reactor fuel assembly in the checkerboard pattern is effected. The foregoing is a description corresponding, in substance, to German application P 3330 850.0, dated Aug. 26, 1983, International priority of which is being claimed for the instant application, and which is hereby made part of this application. Any material discrepancies between the foregoing specification and the specification of the aforementioned corresponding German application are to be resolved in favor of the latter . |
053612922 | summary | BACKGROUND OF THE INVENTION This invention relates to condensers that collect radiation and deliver it to a ring field. More particularly, this condenser collects radiation, here soft x-rays, from a small, incoherent source and couples it to the ring field of a camera designed for projection lithography. Projection lithography is a powerful and essential tool for microelectronic processing. As feature sizes are driven smaller and smaller, optical systems are approaching their limits caused by the wavelengths of the optical radiation. Soft x-rays are now at the forefront of research in the efforts to achieve the desired feature sizes. This radiation has its own problems however. The complicated and precise optical lens systems used in conventional projection lithography do not work well for a variety of reasons. Chief among them is the fact that most x-ray reflectors have efficiencies of only about 60%. This alone dictates very simple beam guiding optics with very few surfaces. One approach has been to develop cameras that use only a few surfaces which can image with acuity only along a narrow arc or ring field 48. Such cameras then use the ring field to scan a reflective mask 46 and translate that onto the wafer 16 for processing. Although cameras can be designed to do this, there is as yet no available condenser system which can efficiently couple the soft x-rays from a source to the ring field required by this type of camera. Further, full field imaging, as opposed to ring field, requires severely aspheric mirrors. Such mirrors cannot be manufactured to the necessary tolerances with present technology for use at these wavelengths. SUMMARY OF THE INVENTION A series of aspheric mirrors on one side of a small, incoherent source of radiation produces a series of beams. If the mirrors were continuously joined into a parent mirror, they would image the quasi point source into a ring image at some radial distance, here some number of feet. Since only a relatively small arc (about 60 degrees) of the ring image is needed by the camera, the most efficient solution is to have all of the beams be so manipulated that they all fall onto this same arc needed by the camera. Also, all of the beams must be aimed through the camera's virtual entrance pupil. These requirements are met in two steps. First, the beams are individually rotated and translated, as necessary, using mirrors so that they overlap at the ring field and pass through the real entrance pupil without interfering with each other. The second step is to image this real entrance pupil into the camera's virtual entrance pupil using a powered imaging mirror. This places the corrected, combined images of the mirrors into the proper position for use by the camera. This system may be configured in a variety of ways, two of which are explained below. Although this condenser was initially designed for use with soft x-rays, it is also suitable for use with any other lithographic system that uses a small, bright source of radiation with a ring field camera with modifications that are well within the ordinary skill those working in this technical area. |
abstract | A hazardous material repository includes a drillhole formed from a terranean surface into a subterranean zone that includes a geologic formation, where the drillhole includes a vertical portion and a non-vertical portion coupled to the vertical portion by a transition portion, the non-vertical portion includes a storage volume for hazardous waste; a casing installed between the geologic formation and the drillhole, the casing including one or more metallic tubular sections; at least one canister positioned in the storage volume of the non-vertical portion of the drillhole, the at least one canister sized to enclose a portion of hazardous material and including an outer housing formed from a non-corrosive metallic material; and a backfill material inserted into the non-vertical portion of the drillhole to fill at least a portion of the storage volume between the at least one canister and the casing. |
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048250880 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT With reference now to FIG. 1, wherein like numerals designate like components throughout all the several figures, the invention is a lightweight cask assembly 1 having structural walls formed from a titanium alloy that is particularly useful in carrying radioactive materials of different activities aboard a vehicle such as a tractor-trailer. In use, the cask assembly is typically mounted within a novel biaxial restraint cradle 3, which in turn is secured onto the trailer of a tractor-trailer (not shown). Generally, the cask assembly itself has a cylindrical body 5 which is circumscribed on either end by toroidal impact limiters 7a and 7b. Each of these impact limiters 7a, 7b is a donut-shaped shell of yieldable aluminum which is approximately one-half of an inch thick. Each of the toroidal impact limiters 7a, 7b is mounted around its respective end of the cylindrical body 5 by means of a support ring assembly 8a, 8b which in turn is secured to the cylindrical body 5 by a plurality of bolts 9. Support ring assembly 8b is formed from a titanium alloy and is part of the end plate assembly 15 discussed hereinafter. Disposed between the impact limiters 7a, 7b are a pair of opposing trunnions 11a, 11b and 11c, 11d. The two pairs of trunnions are disposed 180 degrees apart around the cylindrical body 5 of the cask assembly 1, and are receivable within two pairs of turnbuckle assemblies 12a, 12b, and 12c, 12d (of which only 12a and 12b are visible) that form part of the cradle 3. The cylindrical body 5 is capped by a closure 13 at one end, and an end plate assembly 15 formed from a titanium alloy (shown in FIG. 3) at the other end. As is best seen in FIGS. 3 and 5, the cylindrical body 5 of the cask assembly 1 is generally formed by an outer container 18 which is surrounded by a thermal protection shell 20 on its exterior, and which contains in its interior one of two different shield inserts 22 or 23, depending upon the activity and type of radiation emitted by the material to be transported. While only two specific types of shield inserts 22 and 23 are specifically disclosed herein, it should be noted that the inserts 22 and 23 are merely exemplary, and that the improved cask assembly may in fact be used with any number of different types of shield inserts formed of different shielding materials having different wall thicknesses for handling radioactive material within a broad range of activity and radiation type. With reference now to FIGS. 2A, 2B, and 2C, the thermal protection shell 20 which circumscribes the outer container 18 of the cask assembly 1 is formed from a pair of semi-cylindrical shell sections 24a, 24b which are rigidly interconnectable into thermal contact with one another. Each of the shell sections 24a, 24b includes a pair of cut-outs 26 for admitting the trunnions 11a, 11b, 11c, and 11d. Each of the shell sections 24a, 24b is formed from a metal having a thermal coefficient of expansion which is greater than that of the metal that forms the walls of the outer container 18, and which is at least as heat-conductive as the metal which forms the walls 54 of the outer container 18. When the outer wall of the outer container 18 is formed from titanium, the shell sections 24a, 24b are preferably formed from aluminum or magnesium or an alloy of either or both of these metals. The coefficient of thermal expansion of these metals is approximately twice that of the thermal coefficient of expansion of titanium. Moreover, the high coefficient of thermal conductivity of each such metal insures that the thermal protection shell 20 will not significantly obstruct the conduction of decay heat conducted through the walls of the outer container 18 which is generated by the radioactive material held within the cask assembly 1. When the diameter of the outer container 18 is between forty and sixty inches, a wall thickness of approximately one-half of an inch is preferred for both of the shell sections 24a, 24b. Such a wall thickness renders the thermal protection shell 20, as a whole, thin enough to be conveniently retrofitted over many existing transportation casks without significantly adding to the weight thereof, yet is thick enough to maintain the structural integrity needed to expand away from the outer walls of the outer container when exposed to a source of intense thermal radiation, such as a fire. Finally, the preferred thickness of one-half on an inch provides enough mass to give the entire thermal protection shell 20 a significant latent heat of fusion, which will provide still more thermal protection through ablation should the cask 1 be exposed to intense heat. A plurality of top and bottom connecting assemblies 28, 29 are used to rigidly interconnect the two semi-cylindrical shell sections 24a, 24b. Since each of the connecting assemblies 28, 29 are identical in structure, a description will be made only of the top connecting assembly 28 circled in FIG. 2A. This connecting assembly 28 is formed from a pair of opposing semicircular lugs 30a and 30b which are integrally formed along the edges of the shell sections 24a and 24b respectively. These lugs 30a, 30b include mutually alignable bore holes 31a and 31b for receiving a connecting bolt 32. The threaded end 33 of the bolt 32 is engaged to a tension nut 34 as shown in FIG. 2B. The distance between the two lugs 30a, 30b (and hence the distance between the edges of the shell sections 24a, 24b) is largely determined by the extent of which the end 33 of the bolt 32 is threaded through the tension nut 34. A lock washer 35 is disposed between the tension nut 34 and the lug 30a to prevent the nut 34 from becoming inadvertently loosened. A pair of lock nuts 36a, 36b are threadedly engaged near the center portion of the connecting bolt 32 between the two lugs 30a and 30b. These lock nuts provide two functions. First, when properly adjusted, they prevent the tension nut 34 from applying excess tensile forces between the two shell sections 24a and 24b which might interfere with their expansion away from the outer container 18 in the event the cask assembly is exposed to a fire or other source of intense heat. Second, the nuts 36a, 36b eliminate all slack or play between the lugs 30a, 30b, thus insuring that the connecting assembly 28 rigidly interconnects the two shield sections 30a, 30b. Again, lock washers 37a, 37b are disposed between the lock nuts 36a and 36b and their respective lugs 30a and 30b to prevent any inadvertent loosening from occurring. An overlap 40 is provided between the edges of the two shell sections 24a and 24b to establish ample thermal contact and hence thermal conductivity between these shell sections. The overlap 40 is formed from an outer flange 42 and recess 44 provided along the edge of shell section 24a which interfits with a complementary outer flange 46 and recess 48 provided along the opposing edge of shield section 24b. The actual length of the overlap 40 will vary depending upon the distance between the two lugs 30a and 30b as adjusted by the bolt 32, tension nut 34, and lock nuts 36a and 36b. In operation, the two sections 24a, 24b of the thermal protection shell 20 are installed over the cask assembly 1 by aligning the various cutouts 26a, 26b, 26c, and 26d with the corresponding trunnions of 11a, 11b, 11c, and 11d which project from the cylindrical body 5, and placing the sections 24a, 24b together so that the lugs 30a and 30b of each of the connecting assemblies 28, 29 are in alignment with one another and the flanges and recesses 42, 44, and 48, 46 of each overlaps 40 are interfitted. Next, the bolt 32, tension nut 35, lock nuts 36a, 36b, and lock washers 35, 37a, and 37b are installed in their proper positions with respect to the lugs 30a, 30b of each of the connecting assemblies 28, 29. The tension nut 34 is then screwed over the threaded end 33 of connecting bolt 32 until the interior surface of each of the shell sections 24a and 24b is pulled into intimate thermal contact with the outside wall 54 of the outer container 18. In the preferred method of installing the thermal protection shield, the tension nut 34 of each of the connecting assemblies 28, 29 is initially torqued to a selected maximum on the threaded shaft of the bolt 32 until the nut 34 imparts a significant tensile force between the two lugs 30a and 30b. This tensile force tends to squeeze the two shell sections 24a and 24b together around the outer wall 54 of the outer container 18 in a clamp-like fashion, which in turn removes any significant gaps between the outer surface of the wall 54 and the inner surface of the shell sections 24a and 24b by bending these sections into conformity with one another. In the next step, each of the nuts 34 is relaxed enough to prevent these tensile clamping forces from interfering with the expansion of the thermal protection shell 20 in the event of a fire, yet not so much as to cause the surfaces of the shell 20 and the outer container from becoming disengaged with one another. Thereafter, the lock nuts 36a and 36b are tightened against the faces of their respective lugs 30a and 30b to remove all slack in each connecting assembly 28, 29. The end result is a rigid interconnection between opposing edges of the shield sections 24a and 24b, wherein each of the opposing lugs 30a and 30b is tightly sandwiched between the tension nut 34 and lock nut 36a, or the head of the bolt 38 and lock nut 36b, respectively. If the outer container has no trunnions 11a, 11b, 11c, 11d, or other structural members which would prevent the surfaces of the shell 20 and outer container 18 from coming into intimate thermal contact, the shell 20 may assume the form of a tubular sleeve which may be, in effect, heat shrunk into contact over the container 18. This alternative method of installation comprises the steps removing the impact limiters 7a, 7b, of heating the shell to a temperature sufficient to radially expand it, sliding it over the wall 54 of the outer container 18, allowing it to cool and contract into intimate thermal contact with the wall 54, and reinstalling the impact limiters 7a, 7b. FIG. 2C illustrates the typical gap condition between the inner surface of the thermal protection shell 20 and the outer surface of the outer container 18. Under ambient conditions, these two opposing surfaces are either in direct contact with one another, or separated by only a tiny gap 50 which may be as much as one mil. Such a one mil separation at various points around the cask assembly 1 does not significantly interfere with the conduction of heat between the wall 54 of outer cask 18, and the thermal protection shell 20. However, when the cask assembly 1 is exposed to a source of intense thermal radiation such as a fire, the substantially higher thermal coefficient of expansion of the aluminum or magnesium forming the shell 20 will cause it to expand radially away from the outer surface of the outer container 18, leaving an air gap 53 (shown in phantom) between the two surfaces. Moreover, since the thermal protection shield 20 is formed from a metal having good heat conductive properties, this differential thermal expansion is substantially uniform throughout the entire circumference of the shield 20, which means that the resulting insulatory air gap 53 is likewise substantially uniform. When this gap exceeds approximately two and one-half mils, the primary mode of heat transfer switches from conductive and convective to radiative. Thus the three mil gap provides a substantial thermal resistor between the fire or other source of intense infrared radiation in the outer container 18 of the cask 1. With reference now to FIGS. 3, 4A, 4B, and 5, the side walls of the outer container 18 of the improved cask 1 are a laminate formed from the previously mentioned outer wall 54, an inner wall 56, and a center layer 58 of shielding material. In the preferred embodiment, both the outer wall 54 and inner wall 56 is formed from a high strength alloy of titantium, such as Ti-3-Al-2.5-V, or Ti-6-Al-4-V. Such a titanium alloy is approximately three to four times stronger than most stainless steel on a pound-per-pound basis. Moreover, because titanium is about half the density of most stainless steels, this titanium alloy is about 75% to 100% stronger than most stainless steel on a volume-per-volume basis. The end result is that both the outer wall 54 and the inner wall 56 may be made substantially thinner with a material only about one-half as dense as stainless or low alloy steel. Hence the savings in weight are manifest. While other high strength alloys of titanium may be used, Ti-3-Al-2.5-V is preferred for its easy weldability. Disposed between the outer wall 54 and the inner wall 56 is a layer of Boro-Silicone, which is a shielding material formed from particles of boron suspended in a matrix of silicone. This material advantageously absorbs neutrons from neutron-emitting radioactive materials (such as transuranic elements), and further is a relatively good conductor of heat. It is a rubbery material easily cast, and may be melted and poured between the inner and outer walls 54, 56 of the outer container 18 during it manufacture. Boro-Silicone is available from Reactor Experiments, Inc., and is a registered trademark of this corporation. The bottom of the outer container 18 is formed by an end plate assembly 15 that includes an outer plate 60, an inner plate 62, a layer of center shielding material 64, the previously mentioned support ring assembly 8b and a lower reinforcing ring 65. In the preferred embodiment, the outer plate 60 is again formed from a titanium alloy such as Ti-3-Al-2.5-V approximately one-eighth inch thick. The inner plate 62, like the inner wall 56, is again formed from Ti-3-Al-2.5-V approximately one inch thick. The center shielding material 64 is again preferably Boro-Silicone for all the reasons mentioned in connection with the center shielding material 58 of the side walls of the container 18. The titanium alloy inner plate 62 is joined around the bottom edge of the inner wall 56 360 degrees via weld joint 66. The top of the outer container 18 includes a reinforcing ring 68 again made of Ti-3-Al-2.5-V. This ring 68 is preferably about two inches thick throughout its length, and is integrally connected to the inner wall 56 of the container 18 by a 360 degree weld joint 69. The upper edge of the ring 68 is either threaded or stepped to accommodate one of the two types of improved closures 115b or 117b, as will be explained in detail hereinafter. With specific reference now to FIGS. 3 and 5, the cask assembly 1 is formed from the outer container 18 and shell 20 in combination with one of two different shield inserts 22 (illustrated in FIG. 3) or 23 (illustrated in FIG. 5). Each of the shield inserts 22, 23 is formed from an outer cylindrical wall 72 which is preferably one-half inch thick and a cylindrical inner wall 74 which is approximately one-eighth of Ti-3-Al-2.5-V. Each of the shield inserts 22 and 23 includes a layer of shielding material 76 between their respective outer and inner walls 72, 74. However, in shield insert 22, this shielding material is formed from a plurality of ring-like sections 78a, 78b, and 78c of either depleted uranium or tungsten. These materials have excellent gamma shielding properties, and are particularly well adapted to contain and shield radioactive material emitting high intensity gamma radiation. Of course, a single tubular layer of depleted uranium or tungsten could be used in lieu of the three stacked ring-like sections 78a, 78b, and 78c. However, the use of the stacked ring-like sections is preferred due to the difficulty of fabricating and machining these metals. To effectively avoid radiation streaming at the junctions between the three sections, overlapping tongue and groove joints 79 (see FIG. 4A) are provided at each junction. By contrast, in shield insert 23, a layer of poured lead 80 is used as the shielding material 76. While lead is not as effective a gamma shield as depleted uranium, it is a better material to use in connection with high-neutron emitting materials, such as the transuranic elements. Such high neutron emitters can induce secondary neutron emission when depleted uranium is used as a shielding material. While such a secondary neutron emission is not a problem with tungsten, this metal is far more difficult and expensive to fabricate than lead, and is only marginally better as a gamma-absorber. Therefore, lead is a preferred shielding material when high-neutron emitting materials are to be transported. In both of the shield inserts 22, 23, the bottom edges of the inner and outer walls 72, 74 are welded around a bottom plate 82, while the upper edges of these walls are both welded around an insert reinforcing ring 89. Both bottom plate 82 and ring 89 are formed from Ti-3-Al-2.5V. It should be noted that in the preferred embodiment, the use of a high strength titanium alloy such as Ti-3-Al-2.5V allows the inner wall 74 of each of the shield inserts 22 and 23 to be much thinner than if this wall were made of steel and yet still comply with the U.S. government impact stress criteria. Such thin inner wall minimizes the distance between the shielding material 76 and the radioactive waste disposed inside the insert 22, 23, which in turn minimizes the weight of the shielding material 76 required to meet U.S. government surface radiation requirements. The radius of the interior of the shield inserts 22 and 23 will be custom dimensioned with a particular type of waste to be transported so that the inner wall 74 of the insert comes as close as possible into contact with the radioactive material contained therein. The applicants have noted that fulfillment of the foregoing criteria provides the most effective shielding configuration per weight of shielding material. Additionally, the thickness and type of shielding material 76 will be adjusted in accordance with the activity of the material contained within the shield insert 22, 23 so that the surface radiation of the cask assembly 1 never exceeds 200 mr. The fulfillment of all these criteria maximizes the capacity of the cask assembly 1 to carry radioactive materials while simultaneously minimizing the weight of the cask. The use of titanium alloy in the outer and inner walls 54, 56, 72, 74 of both the outer container 18 and shield inserts 22, 23 has the further advantage of enhancing the overall thermal conductivity of the cask assembly. Despite the fact that the heat conductivity of titanium is only about half as great as the heat conductivity of conventional structural materials such as 304 stainless steel, the fact that the walls 54, 56, 72 and 74 may be made so much thinner as a result of the higher strength of titanium more than offsets the difference in thermal conductivity. The end result is that the use of titanium not only results in a lighter-weight cask, but a safer cask capable of more effectively dissipating the heat of decay of the radioactive materials contained therein, hence insuring that this heat will not create unwanted pressures with the cask assembly 1. FIGS. 4A and 4B illustrate the vent, purge, and drain assembly 90 of the outer container 18. This assembly 90 includes a threaded drain pipe 92 for receiving a drain plug 94. The inner end 96 of the drain plug 94 is conically shaped and seatable in sealing engagement with a complementary valve seat 97 located at the inner end of the pipe 92. Wrench flats 98 integrally formed at the outer end of the drain plug 94 allow the plug 94 to be easily grasped and rotated into or out of sealing engagement with the valve seat 94. A vent pipe 100 is obliquely disposed in fluid communication with the end of the drain pipe 92. A threaded vent plug 102 is engageable into and out of the vent pipe 100. A screw head 103 is provided at the outer end of the vent plug 102 to facilitate the removal and insertion of the threaded plug 102 into the threaded interior of the vent pipe 100. A drain tube 104 is fluidly connected at is upper end to the bottom of the valve seat 97 by way of a fitting 106. In the preferred embodiment, the drain tube 104 is formed from stainless steel, and is housed in a side groove 108 provided along the inner surface of the wall 56 of the outer container 18. As is most easily seen in FIG. 4B, the lower open end 109 of the drain tube 104 is disposed in a bottom groove 110 which extends through the shallowly conical floor 112 of the outer container 18. In operation, the vent, purge, and drain assembly may be used to vent the interior of the outer container 18 by removing the vent plug 102 from the vent pipe 100, screwing an appropriate fitting (not shown) into the threaded vent pipe 100 in order to channel gases to a mass spectrometer, and simply screwing the conical end 96 of the drain plug 94 out of sealing engagement with the valve seat 97. If drainage is desired, the drain plug 94 is again removed. A suction pump is connected to the drain pipe 92 in order to pull out, via drain tube 104, any liquids which may have collected in the bottom groove 110 of the conical floor 112 of the outer container 18. Gas purging is preferably accomplished after draining by removing the vent plug 102, and connecting a source of inert gas to the drain pipe 92. The partial vacuum within the container 18 that was created by the suction pump encourages inert gas to flow down through the drain tube 104. Although not specifically shown, the interior of the drain plug 98 may be provided with one or more rupture discs to provide for emergency pressure relief in the event that the cask assembly 1 is exposed to a source of intense thermal radiation, such as a fire, over a protracted period of time. The closure 13 used in connection with the cask 1 may be either screw-type double-lidded closures 115a, 115b (illustrated in FIG. 3), or breech-lock double-lidded closures 117a, 117b (illustrated in FIG. 5). With reference now to FIG. 3, each of the screw-type closures 115a, 115b includes an outer lid 120a, 120b, and an inner lid 122a, 122b. The inner lid 122a, 122b in turn includes an outer edge 124a, 124b which is seatable over the ledge 126a, 126b provided around the opening 128a, 128b of the shield insert 22 or the outer container 18 respectively. A gasket 130a, 130b circumscribes the outer edge 124a, 124b of each of the inner lids 122a, 122b of the two closures 115a, 115b. In the preferred embodiment, these gaskets 130a, 130b are formed of Viton because of its excellent sealing characteristics and relatively high temperature limit (392 degrees F.) compared to other elastomers. The gasket 130a, 130b of each of the inner lids 122a and 122b is preferably received and held within an annular recess (not shown) that circumscribes the outer edge 124a, 124b of each of the inner lids 122a, 122b and the ledges 126a, 126b. To facilitate the insertion of shield insert 22 into the container 18, it is important to note that the opening 128b of the container 18 is at least as wide as the interior of the container 18 at all points. Each of the outer lids 120a, 120b of the screw-type closures 115a, 115b includes a threaded outer edge 134a, 134b which is engageable within a threaded inner edge 136a, 136b that circumscribes the openings 128a, 128b of the shield insert 22 and the outer container 18 respectively. Swivel hooks 137a, 137b (indicated in phantom) may be detachably mounted to the centers of the outer lids 120a, 120b to facilitate the closure operation. Finally, both the outer lids 120a, 120b of the screw-type closures 115a, 115b includes a plurality of sealing bolts 138a-h, 139a-h, threadedly engaged in bores extending all the way through the outer lids 120a, 120b for a purpose which will become apparent shortly. To seal the cask assembly 1, inner lid 122a is lowered over ledge 126a of the shield insert 22 so that the gasket 130 is disposed between the outer edge 124a of the inner ledge 122a and ledge 126a. The detachably mountable swivel hook 137 is mounted onto the center of the outer lid 120a. The outer lid 120a is then hoisted over the threaded inner edge 136a of the shield insert 22. The threaded outer edge 136a of the shield insert is then screwed into the threaded inner edge 136a to the maximum extent possible. The axial length of the screw threads 134a and 136a are dimensioned so that, after the outer lid 120a is screwed into the opening 128a to the maximum extent possible, a gap will exist between the inner surface of the outer lid 120a and the outer surface of the inner lid 122a. Once this has been accomplished, the securing bolts 138a-h are each screwed completely through their respective bores in the outer lid 120a so that they come into engagement with the inner lid 122a, thereby pressing the gasket 130a and into sealing engagement between the ledge 126a and the outer edge 124a of the lid 122a. The particulars of this last step will become more apparent with the description of the operation of the breech-lock double-lidded closures 117a, 117b described hereinafter. To complete the closure of the cask assembly 1, the outer screw-type closure 115b is mounted over the opening 128b of the outer container 18 in precisely the same fashion as described with respect to the opening 128a of the shield insert 22. With reference now to FIGS. 5, 6A, and 6B, the breech-lock double-lidded closure 117a, 117b also includes a pair of outer lids 140a, 140 which overlie a pair of inner lids 142a, 142b respectively. Each of the inner lids 142a, 142b likewise includes an outer edge 144a, 144b which seats over a ledge 146a, 146b that circumscribes the opening 148a, 148b of the shielding insert 23 and outer container 18, respectively. Each of the outer edges 144a, 144b is circumscribed by a gasket 150a, 150b for effecting a seal between the edges 144a, 144b and their respective ledges 146a, 146b. Like opening 128b, opening 148b is at least as wide as the interior of the outer container 18. Thus far, the structure of the breech-lock double-lidded closures 117a, 117b has been essentially identical with the previously described structure of the screw-type double-lidded closures 115a, 115b. However, in lieu of the previously described screw threads 134a, 134b, the outer edges 154a, 154b of each of the outer lids 140a, 140b are circumscribed by a plurality of uniformly spaced arcuate notches 156a, 156b which define a plurality of arcuate flanges 158a, 158b. Similarly, the inner edges 160a, 160b which circumscribe each of the openings 148a, 148b of the shield insert 23 and outer container 18, respectively, include notches 162a, 162b which circumscribe the inner edges 160a, 160b of the shield insert 23 and the outer container 18. As may best be seen in FIG. 6A and 6C, such dimensioning allows the flanges 164a, 164b of each of the outer lids 140a, 140b, to be inserted through the notches 162a, 162b of each of the openings 148a. 148b and rotated a few degrees to a securely locked position wherein the arcuate flanges 158a, 158b of the outer lids 140a, 140b are overlapped and captured by the arcuate flanges 164a, 164b that circumscribe the inner edges 160a, 160b. It should be further noted that the axial length L1 (illustrated in FIG. 6B) of the interlocking flanges 158a, 158b and 164a, 164b is sufficiently short to leave a small gap L2 between the inner surface of the outer lids 140a, 140b and the outer surface of the inner lids 142a, 142b. The provision of such a small distance L2 between the outer and inner lids allows the outer lids 140a, 140b to be rotated a few degrees into interlocking relationship with their respective notched inner edges 160a, 160b without transmitting any rotary motion to the inner lids 142a, 142b which could cause the inner lid gaskets 150a, 150b to scrape or wipe across their respective ledges 146a, 146b. Connected around the outer edges of the outer lids 140a, 140b are three suspension pin assemblies 166a, 166b, and 166c and 167a, 167b, 167c are uniformly spaced 120 degrees apart on the edges of their respective outer lids 140a, 140b. As the structure of each suspension pin assembly is the same, only a suspension pin assembly 166a will be described. With reference now to FIG. 6C, suspension pin assembly 166a includes a suspension pin 168 which is slideably movable along an annular groove 170 provided around the circumference of each of the inner ledges 142a, 142b. A simple straight-leg bracket 172 connects the suspension pin 168 to the bottom edge of its respective outer lid. In operation, the suspension pin assemblies 166a, 166b, 166c, and 167a, 167b, 167c, serve two functions. First the three suspension pin assemblies attached around the edges of the two outer lids 140a and 140b mechanically connect and thus unitize the inner and outer lids of each of the breech-lock closures 117a, 117b so that both the inner and the outer lids of each of the closures 177a and 117b may be conveniently lifted and lowered over its respective opening 148a, 148b in a single convenient operation. Secondly, the pin-and-groove interconnection between the inner and the outer lids of each of the two breech-lock type closures 117a and 117b allows the outer lids 140a and 140b to be rotated the extent necessary to secure them to the notched outer edges 160a, 160b of their respective containers without imparting any significant amount of torque to their respective inner lids 142a, 142b. This advantageous mechanical action in turn prevents the gaskets 150a and 150b from being wiped or otherwise scraped across their respective ledges 146a, 146b. In the preferred embodiment, the width of the groove 170 is deliberately made to be substantially larger than the width of the pin 168 so that the pin 168 may avoid any contact with the groove 170 when the outer lids 140a, 140b are rotated into interlocking relationship with their respective containers 23 and 18. With reference again to FIG. 6A and 6C, each of the outer lids 140a, 140b includes eight sealing bolts 174a-h, 174.1a-h equidistantly disposed around its circumference. Each of these sealing bolts 174a-h, 174.1a-h is receivable within a bore 175 best seen in FIG. 6C. Each of these bores 175 includes a bottom-threaded portion 176 which is engageable with the threads 176.1 of its respective bolt 174a-h, 174.1a-h as well as a centrally disposed, non-threaded housing portion 177. At its upper portion the bore 175 includes an annular retaining shoulder 178 which closely circumscribes the shank 179 of its respective bolt 174a-h, 174.1a-h. The retaining shoulder 178 insures that none of the sealing bolts 174a-h, 174.1a-h will inadvertently fall out of its respective bore 175 in the outer lid 140a, 140b. In operation, each of the sealing bolts 174a-h is screwed upwardly into its respective bore 175 until its distal end 179.1 is recessed within the threaded portion 176 of the bore 175. After the outer lid 140a or 140b has been secured into the notched inner edge 160a or 160b of its respective container 23 or 18, the sealing bolts 174a-h are screwed down into the position illustrated in FIG. 6C until their distal ends 179.1 forcefully apply a downward-direction force around the outer edges 144a, 144b of their respective inner lids 142a, 142b. Such a force presses the gaskets 150a and 150b into sealing engagement against their respective ledges 146a, 146b. It should be noted that the same bolt and bore configuration is heretofore described is utilized in the screw-type double-lidded closures 115a, 115b. To insure that the outer lids 140a and 140b will not become inadvertently rotated out of locking engagement with their respective vessels 23 or 18, a locking bracket 180 is provided in the position illustrated in FIG. 6A and 6B in each of the outer lids 140a, 140b after they are rotated shut. Each locking bracket 180 includes a lock leg 182 which is slid through mutually registering notches 156a, 156b, and 162a, 162b after the outer lids 140a and 140b have been rotated into locking engagement with the inner edges 160a, 160b of either the shielding insert 23 or the outer container 18. In the case of outer lid 140b, the mounting leg 184 is secured by means of locking nuts 186a, 186b. In the case of outer lid 140a, the mounting leg 184 is captured in place by inner lid 142b which abuts against it. Although not specifically shown in any of the drawings, each of the outer lids 120a, 120 b of the screw-type double-lidded closures 115a, 115b is similarly secured. However, instead of a locking bracket 180, a locking screw (not shown) is screwed down through the outer edges of each of the outer lids 120a, 120b and into a recess precut in each of the inner lids 122a, 122b. |
056299630 | summary | The present invention relates to a tank for the storage tank for radioactive fissile material solutions. Said radioactive fissile material solutions have to be warehoused or stored in tanks or reservoirs whose geometry and/or constituents make it possible to avoid the risks of criticality. Said tanks are referred to as subcritical tanks. The tanks known hitherto contain a volume of neutron absorbing material in contact with which or within which there has been created at least one internal space or channel for containing the radioactive solution. The tanks used are especially parallelepipedal tanks containing a channel of rectangular cross-section for radioactive solution, or cylindrical tanks containing an annular channel for radioactive solution. To prevent the formation of critical masses, said parallelepipedal channel must have a rectangular cross-section of small width; said annular channel is likewise of limited width. Such configurations greatly limit the storage capacities and are therefore unsatisfactory. It is not possible, however, to imagine increasing the length or diameter of said tanks without restriction, since this would present problems of space requirement, rigidity and manufacture. Patent application FR-A-2 212 820 has proposed an improvement to cylindrical tanks with annular channels. A novel arrangement is now proposed for storage tanks for radioactive fissile material solutions. Said novel arrangement makes it possible in particular to reduce the space requirement for a given stored volume. The tanks according to the invention comprise at least one cell for containing the radioactive solution and solid neutron absorbing material for avoiding the risks of criticality; the opposite surfaces between said radioactive solution and said neutron absorbing material are increased within said cell, said radioactive solution being separated from said neutron absorbing material by metal walls. According to a characteristic feature, an array of substantially vertical tubes containing solid neutron absorbing material is arranged within said cell, said tubes having metal walls and being located in compartments delimited by other metal walls and distributed throughout said radioactive solution. Said tubes can be of any shape. In particular, their cross-section can be circular, square, oval (ellipsoidal) etc. The same comment applies to the shape of the cell or cells and that of the tank. Said solid neutron absorbing material is the one conventionally used in this technical field. It can consist especially of boron polyethylene plaster (BPP), boron concrete, boron bitumens or permali, etc. Said radioactive solution and said solid neutron absorbing material are separated by metal walls, generally two metal walls, one of them being the wall of the tubes of the array. The constituent material of said walls is generally stainless steel. However, other materials, such as zirconium or titanium, can be used, especially for walls in contact with the radioactive solution. The presence of these walls within the tanks according to the invention is particularly advantageous. Irrespective of the variant, said walls perform numerous functions. In addition to their function of retaining material (solid neutron absorbing material or radioactive solution), they contribute especially to enhancing the rigidity of the assembly, governing its geometry and protecting said assembly from corrosion, fire etc. They also play an advantageous role in its production. The value of said walls will be understood more clearly from the following description of the invention. It will be noted incidentally here that the present invention is not limited to these variants. According to the invention, the tubes contain the neutron absorbing material and are located in compartments delimited by metal walls and distributed throughout the radioactive solution. In this embodiment, the metal walls which separate the radioactive solution from the neutron absorbing material therefore consist on the one hand of the wall of the tubes containing said neutron absorbing material, and on the other hand of the wall of the compartments within which said tubes are positioned. These two types of walls can be made of stainless steel. It is also possible to provide stainless steel tubes for the neutron absorbing material and zirconium or titanium walls for the compartments inside which said tubes of neutron absorbing material are positioned. In a variant of the invention, the tubes of neutron absorbing material advantageously take the form of stainless steel can filled with neutron absorbing material. Said cans are located in the compartments distributed throughout the radioactive solution. Said compartments have been arranged within the cell before it is filled. It has been possible, in particular, to provide the bottom plate of said cell with welding lips in which said compartments have been positioned and then welded. These are actually tubes dimensioned so as to accommodate the tubes of neutron absorbing material. Said tubes of neutron absorbing material are advantageously placed--inside said compartments--on the bottom of the cell or on positioning studs which are themselves located on the bottom of the cell. Said tubes are thus easy to manipulate from the top, making it possible in particular for the neutron absorbing material to be completely inspected or even replaced. Defined between the two types of walls--the walls of the tubes of neutron absorbing material and the walls of the compartments inside which said tubes are located--there is a volume which can advantageously be used for the circulation of a cooling fluid. Said circulation of a cooling fluid can be effected according to different variants. It is recommended that said volumes communicate with orifices created at the base of the tank. It will be possible to make provision for the forced injection of a cooling fluid through said orifices, the fluid being discharged through the top. The same orifices can be used to effect intrinsic cooling by natural convection of ambient air. This is done by raising the tank and opening said orifices to the ambient air. This embodiment of the invention, put into effect with said intrinsic cooling, is particularly preferred. Irrespective of the embodiment of the invention, the cell which contains the array of tubes is advantageously fitted with a cover. Said cover contributes to the stabilization and rigidity of the assembly. Irrespective of the arrangement of the array or arrays of tubes within the storage tanks according to the invention (it being possible for said tanks to contain several cells), the following can also be included: means for homogenizing the radioactive solution; PA1 means for cooling said radioactive solution. As an example of homogenizing means, there may be mentioned a system for circulating said solution or a system for bubbling gas into the intermediate spaces. Such means are well known to those skilled in the art. For example, said homogenizing means may include an immersed air-lift system for circulating the solution. The system pumps the solution at a low level and discharges solution at a higher level. As an example of cooling means, there may be mentioned immersed coils or submerged cooling pins. According to the invention, said forced cooling means can coexist with the above-described means for ensuring intrinsic cooling by natural convection. Such a cooling means may consist of an immersed coil including a number of turns. A cooling liquid flows through the coil. A further possibility is to position a cell with an array of tubes, according to the invention, in the central space of a traditional annular tank so as to optimize the utilization of the volumes. Such a tank of the invention makes it possible to store radioactive solution on the one hand in the annular channel and on the other hand in the intermediate space between the walls of the compartments within which the tubes containing the neutron absorbing material are located. The number of tubes and their distribution within a cell of tanks according to the invention are quite obviously governed by the criticality calculation as a function of the chosen embodiment. The tanks of the invention make it possible to store a given quantity of radioactive solution on a greatly reduced floor space. Said tanks, which are more compact, also possess a better stability. Their rigidity and mechanical stability give them a good resistance to earthquakes. The reduction in space requirement makes the equipment easier to transport and to install in nuclear buildings, especially during renovation operations. In addition, a given radioactive solution can be stored using a tank of the invention with intermediate spaces whose dimensions correspond to approximately twice the width of a traditional tank channel. This enables measuring instruments (sensors, probes etc.) or homogenizing means larger than those used in channel tanks to be introduced into the stored solutions. |
abstract | A nuclear reactor module includes a reactor vessel and a reactor housing mounted inside the reactor vessel, wherein the reactor housing comprises a shroud and a riser located above the shroud. The nuclear reactor module further includes a heat exchanger proximately located about the riser, and a reactor core located in the shroud. A steam generator by-pass system is configured to provide an auxiliary flow path of primary coolant to the reactor core to augment a primary flow path of the primary coolant out of the riser and into the shroud, wherein the auxiliary flow path of primary coolant exits the reactor housing without passing by the heat exchanger. |
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059057700 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a storage framework for nuclear reactor fuel elements, having a plurality of structural wells of essentially rectangular cross section fastened on a baseplate. In order to provide for the intermediate storage of spent fuel elements from a nuclear power plant, in particular a pressurized water or boiling water reactor, the fuel elements are stored within the nuclear power plant, in particular within fuel element storage basins. It is desirable for technical and economic reasons to utilize the available storage space as efficiently as possible. In order to increase the storage capacity, storage frameworks with incorporated nuclear poison, for example with boron-containing material, are employed. The storage framework, while having sufficient mechanical stability and neutron-absorbing properties, should require as little space as possible for itself, in order to allow a high loading density for spent fuel elements. Published European Patent Application 0 537 615 A1, corresponding to U.S. Pat. No. 5,311,563, specifies a storage framework for nuclear reactor fuel elements, in which a plurality of wells of essentially rectangular cross section are fastened on a baseplate, the wells standing vertically on the baseplate and in each case being disposed diagonally opposite one another in a checkered manner. Some of the wells located diagonally opposite one another are in each case connected to one another, along their mutually adjoining longitudinal edges with an offset, through the use of at least two connecting elements that bridge a gap formed by the offset. The pair of longitudinal edges is assigned at least one first connecting element of high rigidity, in each case in a first direction parallel to the baseplate, and at least one second connecting element of high rigidity, in each case in a second direction parallel to the baseplate. Due to the high rigidity of the connecting elements, internal transverse forces acting on the storage framework can be absorbed, without the need for an additional supporting grid in the upper region of the storage framework. It is thus also possible to load the interspaces of the storage framework which are not provided with wells, so that intermediate positions or intermediate locations for fuel elements are formed. In order to provide neutron absorption, the walls of the wells of the storage framework are composed of austenitic boron steel with a boron content of up to 2%, and the connecting elements are composed of a soft austenitic steel, the carbon content of which is lower than 0.1%. As a result, when extreme external forces act on the storage framework, the boron steel wells experience virtually no deformation, since the external forces are absorbed, where appropriate, by virtue of plastic deformations of the connecting elements. The use of an austenitic boron steel as a load-bearing structure for the wells makes those wells particularly complicated to manufacture, in order to ensure the required mechanical stability of the boron-treated steel. Further embodiments of a storage framework for spent fuel elements, in which the respective storage framework has a neutron-absorbing material, are disclosed, for example, in U.S. Pat. No. 4,088,897, U.S. Pat. No. 4,630,738, U.S. Pat. No. 4,695,424 and U.S. Pat. No. 4,119,859. A fact common to the storage frameworks known from those publications is that the neutron-absorbing material is integrated firmly into the load-bearing structure of the storage framework. In the storage framework disclosed in U.S. Pat. No. 4,088,897, boron-containing material is disposed between an inner and an outer wall of a well for receiving a spent fuel element. In the storage framework known from U.S. Pat. No. 4,630,738, a plurality of square neutron-absorbing wells parallel to one another on a baseplate are connected firmly to the baseplate, with the neutron-absorbing material, in the form of plates made from sintered boron-treated aluminum, being fastened firmly to the respective sides of the rectangular well. The boron-treated plates are disposed between an inner and an outer well. U.S. Pat. No. 4,695,424 discloses a storage framework, in which the neutron-absorbing material, for example a boron carbide, in the form of plates is non-displaceably fastened, particularly welded, to the outside of a well for receiving spent fuel elements. U.S. Pat. No. 4,119,859 describes a storage framework for fuel elements of a nuclear power plant, with wells for receiving a fuel element in each case. There, the wells have a sandwich structure with an inner wall and an outer wall, between which a neutron-absorbing material, for example boron carbide, is intercalated. The known measures for introducing neutron-absorbing material into a fuel element storage framework involve integrating the neutron-absorbing material firmly into the load-bearing structure of the fuel element storage basin, thus necessitating a considerable construction outlay in production terms and, where appropriate, requiring a monitoring of the neutron-absorbing material. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a storage framework for nuclear reactor fuel elements, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which ensures easier assembly, a high stability of an entire interconnected well system, a high loading density as well as uncoupling of neutron-absorbing material from load-bearing structures of the interconnected well system. With the foregoing and other objects in view there is provided, in accordance with the invention, a storage framework for nuclear reactor fuel elements, comprising a baseplate; a plurality of structural wells of substantially rectangular cross section fastened on and standing upright on the baseplate, the structural wells disposed approximately diagonally opposite one another in a checkered manner, and each two of the structural wells disposed diagonally opposite one another spanning an interspace; a neutron-absorbing structure with boron-treated steel disposed in at least one of an interspace and a structural well for displacement relative to the structural wells, and the neutron-absorbing structure having at least a partial region with spacers, such as strips, plates or edge beads, toward a respective one of the structural wells. Due to the above-described separation of the load-bearing function in the structural wells from the function of neutron absorption, the structural wells intended for receiving the fuel elements can be manufactured simply from known non-boron-treated austenitic steel or other steels permitted for this purpose, in particular in mass production. The neutron-absorbing structure has no load-bearing function of any kind, so that the use of boron-treated steel presents no problem at all both from a production standpoint and for reasons of mechanical stability. By virtue of an appropriate shaping of the neutron-absorbing structure, the interspace may likewise be used for receiving spent fuel elements. The use of boron-treated steel additionally has the advantage that there is no need to monitor the neutron absorption capacity. The boron-treated steel has, for example, a boron fraction of up to 2%. It is likewise possible to place the structure in the structural wells, in which case a configuration both in the structural wells and in the interspaces is particularly effective, since two layers of, for example, plates made from boron-treated steel are thereby laid one behind the other, with the result that it is also possible to store fuel elements having relatively high radioactive radiation capacity. Since the neutron-absorbing structure is constructed so as to be displaceable relative to the structural wells, an exchange of the structure can be carried out, for example if it becomes necessary to adapt the shape of the structure to the fuel elements to be received. Furthermore, this ensures that mechanical forces are transmitted to the structure at most to only a slight degree. Moreover, a displaceable structure is suitable for the retrofitting of existing fuel element storage frameworks. In accordance with another feature of the invention, the neutron-absorbing structure has plates made of boron-treated steel which, particularly for especially effective neutron absorption, are disposed essentially parallel to the side walls of the structural wells. In accordance with a further feature of the invention, the plates are connected to an essentially rectangular absorber well which, as a non-load-bearing element, merely has to have sufficient self-supporting stability. Such an absorber well may also be inserted into a formed interspace after an interconnected storage framework system has been produced from the structural wells. In accordance with an added feature of the invention, the absorber well is formed in each case from four plates which are connected releasably to one another, in particular intermeshed. As a result, the absorber well formed from the plates connected releasably to one another is held in its position by the structural wells surrounding it, thereby forming a preferably rectangular cell which is loosely installed and is not subjected to load. This cell formed by the absorber well is, in turn, suitable for receiving a spent fuel element. In accordance with an additional feature of the invention, in order to support and position the absorber well, at least some of the plates adjoin a respective structural well in at least a part region. For this purpose, the plates have suitable spacers, such as strips or distance plates. The plates may also have edge beads that are bent outward, that is to say bent toward the side walls of the structural wells, particularly in order to improve the utilization of the corners formed by the structural wells. In accordance with yet another feature of the invention, a structural well has, at an end located opposite the baseplate, a guide strip which, in particular, runs parallel to the baseplate and points into the interspace and through the use of which an unintentional removal of the neutron-absorbing structure from the interspace, for example when a fuel element disposed therein is extracted, is avoided. In accordance with yet a further feature of the invention, in order to increase the mechanical stability of the fuel element storage framework which is constructed from the structural wells, structural wells located diagonally opposite one another are in each case connected at the corresponding longitudinal edges through at least one connecting element, the connecting element having high rigidity and running in a first direction parallel to the baseplate. The connection, in particular a welded connection, of the structural wells to the connecting element, forms a mechanically stable interconnected system of load-bearing structural wells for receiving spent fuel elements. In accordance with a concomitant feature of the invention, in order to provide a further increase in mechanical stability, a further connecting element of high rigidity, which runs along a second direction parallel to the baseplate, is provided at the longitudinal edges. There is an angle which is preferably between 70.degree. and 90.degree. between the first direction and the second direction. The connecting elements accordingly run largely perpendicularly to one another and intersect within the gap present between the longitudinal edges of the structural wells located diagonally opposite one another. The structural wells are thereby fixed within the plane spanned by the baseplate and forces occurring between them are absorbed by the connecting elements. 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 storage framework for nuclear reactor fuel elements, 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. |
039986918 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a method of producing the radioisotope of iodine, I-131, from a tellurium oxide intermediate irradiated by a neutron flux in a nuclear reactor. 2. Description of the Prior Art The radioisotope of iodine, I-131 (hereinafter referred to as I-131), has so far been produced from metallic tellurium or tellurium compounds irradiated by a neutron flux. This is, when tellurium is exposed to a neutron flux from a nuclear reactor or other neutron source for a suitable period of time, the isotope of tellurium, Te-130, which is contained in tellurium in the proportion of 34.49%, is converted to radioactive tellurium, Te-131, and this radioactive isotope is converted to I-131 with spontaneous emission of .beta.-rays. The technical difficulty in the application of the above method resides in the complicated separation process in which pure I-131 in the useful form is to be separated from the tellurium starting material. Either a wet method or dry method has so far been used for the separation process. With the wet method metallic tellurium powder was, at first, used as a starting material. This material is irradiated with a neutron beam in a nuclear reactor, and then dissolved in a mixture comprising conc. sulfuric acid and chromic acid anhydride. A very vigorous reaction takes place upon the dissolution, and the tellurium is converted to telluric acid and the iodine (I-131) to iodic acid. By adding oxalic acid to the above solution, the I-131 is reduced to elemental iodine and is separated by means of distillation. The I-131 thus obtained is heavily contaminated by the reagents which are used in a large amount in the above process and, therefore, a further purification process is required. The above problem was partly overcome by using telluric acid, which is soluble in water or mineral acids, as a starting material. R. Constant, Journal of Inorganic and Nuclear Chemistry, Vol. 7, pp. 133-139, (1958). By using telluric acid, the separation process is made fairly simple. However, the stability of telluric acid is low at high temperature. This brings about a large disadvantage in irradiation by a nuclear reactor; namely, there is danger that the material will decompose during irradiation and the container for irradiation may burst owing to the increase of internal pressure resulting from generation of gases accompanying the decomposition. On the other hand, with the dry method, which was reported by K. Taugbol and J. B. Dahl in JENER REPORT No. 52, I-131 is separated from the irradiated powder of tellurium dioxide by means of dry distillation at a temperature ranging from 680.degree. - 700.degree. C in an air or oxygen stream. This process does not require complicated dissolution and separation processes which require a large amount of reagents, and the apparatus and the operation are simplified. In addition, with the dry method only I-131 is distilled and the radioactivity of I-131 is not diluted with water, whereas with the wet method water is also distilled with I-131. However, this method has an unavoidable defect in that the final product, distilled I-131, is contaminated by the tellurium volatilized together with I-131. The distillation temperature should be high in order to separate I-131 from said powder efficiently; but at the same time, this increases the volatilization of tellurium, and thereby the purity of the distilled I-131 is lowered. The present inventors previously found that the above problems could be solved by using tellurium trioxide as a starting material, and completed a novel and excellent method of producing I-131. [See Japanese application No. 23537/1970 and U.S. application Ser. No. 125,337, filed Mar. 17, 1971 now U.S. Pat. No. 3,772,146]. In this method the properties of tellurium trioxide are utilized which comprise stability at temperatures up to about 400.degree. C (i.e. tellurium trioxide does not substantially change chemically), stability under irradiation by the nuclear reactors usually used for research and the thermal property that tellurium trioxide readily releases I-131 included in the crystal lattice accompanying the decomposition that begins at about 400.degree. C. The temperature characteristics of tellurium trioxide, when used as a starting material, result in large advantages over telluric acid which decomposes at about 110.degree. C. Further, the advantages of this method are that I-131 can be recovered with high recovery ratio in a short time at a much lower temperature than is required for the process using tellurium dioxide as a starting material (actually, it is suitable to heat irradiated tellurium trioxide at 450.degree. C in order to promote the decomposition and the accompanying release of I-131 from said material), and moreover, the low distillation temperature eliminates worry about the volatilization of tellurium, which was a defect of conventional methods. However, it is feared that the temperature of such material may be raised higher than about 400.degree. C when it is irradiated with a high density neutron flux in the nuclear reactor used for research in order to produce a large amount of I-131. Accordingly, there are some practical problems in using a large nuclear reactor for the irradiation of tellurium trioxide. SUMMARY OF THE INVENTION The object of this invention is to use, as a starting material, a tellurium oxide (hereinafter referred to as tellurium oxide intermediate) characteristics of which are intermediate between those of tellurium trioxide and of tellurium dioxide. That is to say, the tellurium oxide intermediate is more stable at high temperatures than tellurium trioxide and releases I-131 atoms without the volatilization of tellurium at lower temperatures than tellurium dioxide does. Accordingly, the object of this invention is, in other words, to propose a novel method of producing radioactive iodine keeping the advantages of the dry method and removing the defects which reside in prior methods using tellurium dioxide or tellurium trioxide as starting materials, i.e. volatilization of tellurium at the high temperatures which accompany the distillation of I-131 from tellurium dioxide and low stability for a high temperature of tellurium trioxide as mentioned above, by using a tellurium oxide intermediate as a starting material. In the present invention, the neutron flux density and irradiation time are not especially restricted. One skilled in the art can easily determine a suitable irradiation period according to the neutron flux density to be employed so as to convert Te-130 to Te-131, and, hence, to I-131 of a required quantity. In case it is feared that the temperature will raise as high as the decomposition temperature of the tellurium oxide intermediate by employing a large flux density (e.g. 3 .times. 10.sup.14 n/cm.sup.2.sup.. sec or more), it may be necessary to cool the material in the course of irradiation. The period of time required for the decomposition of the tellurium oxide intermediates depends on the temperature, the amount of the material and the surface area of the material to be treated. The temperature ranges from the decomposition temperature (about 560.degree. C) to about 650.degree. C, preferably from about 600.degree. C. to about 650.degree. C. |
summary | ||
claims | 1. A radiotherapy apparatus comprising:a source of radiation configured to emit a beam of radiation; anda collimator structure configured to limit a lateral extent of the beam, the collimator structure comprising:a primary collimator positioned adjacent to the radiation source, the primary collimator comprising a primary aperture configured to shape the beam;a first collimator comprising a plurality of adjacent elongate leaves, the leaves being extendable into the beam in a first direction transverse to the beam; anda block collimator positioned away from the radiation source, the block collimator comprising a spine region extending in a second direction and an aperture, located on the spine region, configured to permit the beam to pass through, the block collimator being extendable into the beam in the second direction transverse to the beam and transverse to the first direction. 2. The radiotherapy apparatus according to claim 1, wherein the aperture is cone-shaped. 3. The radiotherapy apparatus according to claim 1, wherein the aperture is a through-hole. 4. The radiotherapy apparatus according to claim 1, wherein the aperture is filled with a radiotransparent material. 5. The radiotherapy apparatus according to claim 4, wherein the aperture has a frusto-conical shape configured to limit the beam to a desired diameter. 6. The radiotherapy apparatus according to claim 1, wherein the block collimator includes a plurality of apertures including the aperture and at least one additional aperture. 7. The radiotherapy apparatus according to claim 6, wherein the plurality of apertures have different dimensions. 8. The radiotherapy apparatus according to claim 6, further comprising:a control apparatus configured to, in a first mode:extend the block collimator such that the aperture is placed in the path of the beam; andextend leaves of the first collimator in order to cover the at least one additional aperture. 9. The radiotherapy apparatus according to claim 1, further comprising:a control apparatus configured to, in a first mode, extend leaves of the first collimator in order to cover the aperture. 10. The radiotherapy apparatus according to claim 9, wherein the control apparatus is further configured to, in a second mode:extend the block collimator such that the aperture is placed in the path of the beam. 11. The radiotherapy apparatus according to claim 1, wherein the block collimator further comprises:a first side region and a second side region flanking the spine region, the first and second side regions having a second thickness that is less than the first thickness. 12. The radiotherapy apparatus according to claim 11, wherein the block collimator further comprises:a front edge portion having the first thickness. 13. The radiotherapy apparatus according to claim 1, wherein the block collimator further comprises:a frontal portion extending transversely to the spine region in either direction to define a T shape. 14. The radiotherapy apparatus according to claim 13, wherein the frontal portion extends in the first direction. 15. The radiotherapy apparatus according to claim 13, wherein the spine region is of a first thickness and the block collimator further comprises:a first side region and a second side region flanking the spine region, the first and second side regions having a second thickness that is less than the first thickness. 16. A collimator structure for a radiotherapy apparatus, the collimator structure comprising:a primary collimator comprising a primary aperture configured to shape a beam of radiation emitted from a source of radiation, the primary collimator being positioned adjacent to the radiation source;a first collimator comprising a plurality of adjacent elongate leaves, the leaves being extendable into the beam in a first direction transverse to the beam; anda block collimator positioned away from the radiation source, the block collimator comprising a spine region extending in a second direction and an aperture, located on the spine region, configured to permit the beam to pass through, the block collimator being extendable into the beam in the second direction transverse to the beam and transverse to the first direction. 17. The collimator structure according to claim 16, wherein the block collimator comprises a plurality of apertures having different dimensions. 18. The collimator structure according to claim 16, wherein the aperture has a frusto-conical shape configured to limit the beam to a desired diameter. 19. The collimator structure according to claim 16, wherein the spine region has a first thickness and the block collimator further comprises:a first side region and a second side region flanking the spine region, the first and second side regions having a second thickness that is less than the first thickness. 20. The collimator structure according to claim 19, wherein the block collimator further comprises:a front edge portion having the first thickness. |
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051125676 | description | DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows the vessel 1 of a pressurized water nuclear reactor, arranged inside a vessel well forming a part of the concrete structure 2 of the nuclear reactor building. The vessel 1 contains the nuclear reactor core 3, consisting of juxtaposed fuel assemblies. FIG. 1 shows the guide tube 4 of a core assembly 3 forming the end part, in a vertical direction, of an instrumentation tube 5 enabling neutron flux measurements to be carried out along the height of the reactor core 3. The U-shaped tube 5 comprises a guide tube 6 passing through the bottom of the vessel in a leakproof manner and fastened to the bottom in the extension of the guide tube 4 and of a junction channel between the bottom of the vessel and the lower part of the guide tube 4 passing through a part of the lower internal equipment of the reactor. Following the vertical section 6a fastened to the bottom of the vessel, the guide tube 6 comprises a horizontal section 6b and then a vertical section 6c whose upper end is fastened to a part 2a of the nuclear reactor structure. The sections 6a, 6b and 6c of the guide tube 6 are joined together by curved parts of the guide tube. The upper end part of the section 6c of the guide tube 6 is connected to a passage device 7 permitting the leakproof passage into the guide tube 6 of a conduit 8 intended to ensure the guiding and the positioning of at least one measurement probe inside the guide tube 4. The thimble-shaped conduit 8 comprises a closed end intended to occupy a working position inside the guide tube 4 and an open end situated opposite, at the end of the part of the conduit 8 which is situated outside the guide tube 6, above the leakproof passage device 7. In its upper end part situated outside the guide tube 6, the conduit 8 comprises a manual closure valve 9 intended to shut off the end of the guide conduit in a leakproof manner in the case where the conduit 8 has a sealing defect which may be detected while the reactor is in operation. The guide conduit 8 or guide conduit makes it possible to perform the guiding and the positioning of one or more measurement probes fastened to a flexible cable which is introduced into the guide conduit 8 through its open upper end. The tube 5 and the guide conduit 8 forming its internal part intended to receive the measurement probes are in the shape of a U, one of whose vertical legs is inside the reactor vessel and whose other vertical leg emerges in its upper part into a measuring room 10. The insertion and withdrawal of the guide conduits 8 from the guide tubes, the insertion and withdrawal of the measurement probes from the guide conduits, and the recovery and treatment of the measurement signals originating from the probes can all be carried out from the measuring room 10. The leakproof passage device 7 is positioned above the height of the reactor vessel 1. When reactor refuelling operations, and possibly maintenance and repair work, are carried out, the reactor vessel is open as shown in FIG. 1, and is full of water. The water level in the guide tubes 6 communicating with the inside of the vessel is then situated below the leakproof passage device 7 of the guide conduits 8. When the reactor is in operation, the vessel is closed by a closure head and filled with primary water at a pressure of the order of 155 bars. The guide tubes 6, in each of which a guide conduit 8 is engaged, fill with primary water under pressure up to the height of the leakproof passageway 7, where the closure of the upper end of the guide tube 6 is ensured by the guide conduit 8 interacting with the seals of the device 7. The outer surface of the wall of the guide conduits 8 is subjected to the pressure of the primary fluid. In the case where one of the guide conduits 8 has a sealing defect, primary water enters inside the guide conduit and spreads inside the room 10 through the open end of the guide conduit emerging into this room. It then becomes necessary to close the manual valve 9 very rapidly, and this, as indicated above, involves some hazard to the operator. FIG. 2 shows the upper part of an instrumentation tube such as the tube 5 of FIG. 1, in the region of the measuring room 10. The corresponding elements in FIGS. 1 and 2 bear the same references. The guide tube 6 in which the guide conduit 8 is engaged is fastened, in its upper part, to a metal structure 12, itself firmly fastened to the part 2a of the concrete structure of the reactor. Leakproof passage devices 7 for the guide conduits 8 are fastened to the upper surface of the structure 12 by means of supports 13. The upper end part of the guide conduit 8, vertical in direction and situated outside the guide tube 6 and the sealing device 7, carries at its upper end a manual closure valve 9 allowing the end of the guide conduit 8 to be shut off in a completely leakproof manner in the case where the wall of this guide conduit exhibits a sealing defect. A shutoff device according to the invention, which will be described in greater detail with reference to FIGS. 3 to 5, is inserted in the vertical upper end part of the guide conduit 8, between a lower section 8a of this upper part and an upper section 8b, to the end of which the manual closure valve 9 is fastened. In FIGS. 3, 4 and 5 it can be seen that the shutoff device 14 according to the invention is made up in the shape of a housing comprising a thick wall made up of two complementary parts 15 and 16 assembled together by bolts 17 and clamping nuts 17'. The bolts 17 pass through the two parts 15 and 16 of the housing throughout their thickness and ensure effective clamping of these two parts, between which a seal 18 is inserted, engaged in an annular groove machined into the part 15 of the housing, inside the region of the component 15 through which the bolts 17 pass. The heads of the bolts 17 are housed in recesses machines into the part 16 forming the over of the housing, and the nuts 17' are locked against rotation in recesses machined into the second component 15 forming the body of the housing to which the cover 16 is fastened. As can be seen in FIGS. 3 and 4, the body 15 of the housing comprises, along its axis 19, two connection nozzles 20 and 21. The nozzle 20 comprises an internal bore 20a tapped over a part of its height and forming a housing of cylindrical shape. The nozzle 21 comprises an external threading 21a. The parts 20a and 21a of the connection nozzles 20 and 21 are intended to receive couplings firmly attached to the corresponding ends of the sections 8a and 8b of the guide conduit, which are attached by screwing to the housing 14. The sealing of the coupling is ensured by seals or by a weld. The housing body 15 comprises a cavity 24 which is closed in a leakproof manner by the cover 16 when the housing is assembled by virtue of the bolts 17 and of the seal 18. The smooth part of the bore of the connection nozzle 20 forms the housing of the head of a bolt 25 fastened in a tapped bore 26 machined into the body 15 of the housing 14 in the extension of the bore 20a. A seal 27 is inserted between a shoulder of the bolt head 25 and the end shoulder of the bore of the connection nozzle 20. The body 15 of the housing 14 is pierced in the extension of the bore of the nozzle 20 and of the tapped hole 26, so as to form an opening emerging into the chamber 24, inside which is arranged a valve seat 30 made of expanded graphite. The bolt 25 and the seat 30 are pierced axially and throughout their length to form a perfectly calibrated channel 28. At its upper end, the channel 28 emerges into a prismatic opening 29 making it possible to perform the clamping of the bolt 25 engaged in the bore of the nozzle 20, from the entry part of this bore. The clamping is performed so as to keep the seat 30 in its housing, the limitation of the clamping being obtained by squashing the seal 27, onto which the shoulder of the bolt head is applied through the intermediary of a washer. A resilient washer is also inserted between the end of the bolt 25 and the upper surface of the seat 30, onto which the pressure exerted by the bolt is applied. The nozzle 21 is pierced axially to form a perfectly calibrated channel 31 whose diameter is equal to the diameter of the channel 28. The channel 31 emerges into the cavity 24 in the axial extension of the channel 28 passing through the seat 30 and the securing bolt 25. As can be seen in FIGS. 4 and 5, the housing 14 comprises a shutter 32 mounted for oscillation around a horizontal axis 33, inside the cavity 24, between the body 15 and the cover 16 of the housing. The movable shutter 32 comprises a balance arm 34 which consists of a cast component obtained by precision molding, a spherical closure member 35 and a counter-weight 36 made of sintered material consisting of nickel-bonded tungsten carbide. Passing through the central part of the balance arm 34 is an opening which has a cross-section of hexagonal shape, in which is engaged a shaft 37 having a central part of prismatic shape with a hexagonal cross-section and two smooth frustoconical end parts forming pivots. The pivots of the shaft 37 are engaged in frustoconical cavities of two bearings 38a and 38b, made of hard material and arranged in a cavity provided in the body 15 of the housing and in a cavity of the cover 16 respectively. The shaft 37 is held by resilient clamping between the bearings 38a and 38b, by virtue of resilient washers 39a and 39b inserted between a shoulder of the corresponding bearing and a shoulder of the cavity in which the bearing is engaged. When the cover 16 is clamped onto the body 15 of the housing by means of the bolts 17, the shaft 37 is clamped resiliently between the bearings 38a and 38b. The balance arm 34, which is locked onto the prismatic shaft 37 and the whole of the shutter 32 firmly attached to the balance arm 34 are thus mounted movably in rotation inside the cavity 24 of the housing 14. The prismatic shaft 37 and the bearings 38a and 38b consist of hard material components which can be obtained by powder metallurgy. The shutter 35 consists of a ball made of hard material and pierced by spark erosion so as to arrange a cavity into which is introduced a spindle or pin 40 of small diameter, brazed at its other end to a part of the balance arm 34, which is situated at some distance from the axis of rotation 33 of this balance arm. The cavity machined into the ball 35 permits a floating mounting of the ball onto the pin 40, so that when the shutter is in its closing position, as shown in FIG. 4, the ball 35 is engaged in a hemispherical cavity machined into the seat 30 at the end of the channel 28. The floating mounting of the ball makes it possible to improve the contact between its outer surface and the hemispherically shaped seat bearing to ensure leakproof closure of the channel 28 passing through the seat 30 and the bolt 25. The counterweight comprises an internal spot facing in which there is engaged an end part of the balance arm 34 remote from the part to which the pin 40 is fastened, in relation to the axis of rotation 33 of the shutter. The counterweight 36 is fastened by a screw 41 to the balance arm 34. The position of the counterweight 36 enables the ball 35 to be returned to a closing position inside the hemispherical cavity of the seat 30, as shown in FIG. 4. The body 15 of the housing 14 forming a wall surrounding the cavity 24 is traversed throughout its thickness by a tapped hole 42 in which there is fastened by screwing the connecting end of a manometer 43 enabling the pressure inside the cavity 24 to be monitored continuously. The manometer 43 consists of a pressure sensor which activates an electrical device such as an all-or-nothing contact whose signal is conveyed by a cable to an alarm device placed in the measuring room 10. If the pressure inside the chamber 24 exceeds a certain predetermined value, an alarm signal is transmitted, so as to alert the operator responsible for supervising the measuring room 10. As can be seen in FIG. 2, the shutter 14 is inserted between two sections 8a and 8b of a guide conduit of the pressurized water nuclear reactor, intended to ensure the support and the guiding of a measurement probe fastened to the end of a flexible cable. The diameter of the measurement probe and of the flexible cable is very slightly smaller than the internal diameter of the channels 28 and 31 passing through the connection nozzles of the housing 14. The probe fastened to the end of its measurement cable can be introduced into the guide conduit so that it enters the upper section 8b of the guide conduit, and then into the connection nozzle 20 of the housing 14 and into the channel 28. On leaving the channel 28, the probe exerts a downward thrust on the ball 35 for closing the seat 30 in a sealed manner, and this makes the shutter 32 tilt into its position 32,, shown in broken lines in FIG. 4 and in continuous lines in FIG. 6. The measurement probe and its connection cable 45 are introduced into the channel 31 passing through the nozzle 21, and then into the lower section 8a of the measurement conduit engaged inside the sealed passage device 7. The probe and its connection cable are guided by the guide conduit until the probe has reached, in its working position, the inside of the reactor core 3. As can be seen in FIG. 6, when the probe is in place in the measurement conduit, the cable 45 is engaged inside the channels 31 and 28 with a very small clearance. If the guide conduit in which the measurement probe is engaged exhibits a sealing defect, the pressurized water of the nuclear reactor entering the guide conduit 8 is subjected to a very large pressure drop in the region of the very narrow annular space remaining between the outer surface of the cable 45 and the surface of the channel 31. Water at a pressure which is much lower than the primary pressure begins to fill the cavity 24 of the housing 14. When the pressure in the cavity 24 exceeds the predetermined surveillance threshold, the manometer 43 sends a signal to the measuring room, which alerts the operator to the presence of a leakage through the wall of the guide conduit 8. The operator can then close the valve 9 ensuring the sealing of the guide conduit before contaminated liquid spreads to the room 10. When the measurement probe and its cable 45 are not in position inside the guide conduit, the shutter 32 is in its position shown in continuous lines in FIG. 4. The ball 35 is applied against the seat 30 by the counterweight 36 of the shutter 32. If the thimble has a leakage, pressurized water from the nuclear reactor enters this guide conduit and comes to the point of filling the cavity 24. The pressurized water exerts an upward vertical force on the ball 35, which is thus applied against the hemispherical cavity of the seat 30 with a pressure which is proportionally higher, the higher the pressure in the cavity 24. The shutter 32 thus ensures leakproof closure of the upper part of the guide conduit, and this prevents a leakage of pressurized water into the room 10. An alarm signal is transmitted to the measuring room by the pressure sensor 43 as soon as the pressure in the chamber 24 exceeds the predetermined surveillance threshold. The operator is thus alerted and closes the valve 9 of the guide conduit, in order to increase the safety of the leakproof closure of this guide conduit exhibiting a leakage. In all cases, if a leakage through the wall of a guide conduit arises, the operator can perform the leakproof closing of this guide conduit before contaminated liquid spreads into the measuring room. In the case where the measurement probe is not introduced into the thimble, the shutter according to the invention ensures leakproof closure of the guide conduit in an automatic and passive manner when a leakage appears through the wall of the guide conduit. The invention is not limited to the embodiment which has been described. The wall of the housing 14 forming the shutoff device can be made in a manner other than that described, starting with different components of a body and of a cover of suitable shape which are obtained by precision molding. The method of construction of the housing in two parts added onto one another and fastened by bolts with the insertion of a seal makes it possible to facilitate the mounting of the shutoff device and in particular of the movable shutter situated in its central cavity. It is also obvious that the passage channels for the measurement probe and for its connection cable can be made in a way which differs from that described. Similarly, the seat onto which the closing device of the movable shutter is applied can be made and fastened inside the housing in a way which differs from that described. Similarly, the movable shutter may differ in shape from that described. The counterweight may be made as a single component with the body or balance arm of the shutter and the shaft of rotation of this shutter may be made in a form which differs from that described. The closing member of the shutter may be other than a ball, and may be fastened to the body or balance arm of the shutter in any suitable manner. Lastly, the shutoff device according to the invention may be employed not only in the case of a guide conduit introduced into a pressurized water nuclear reactor and receiving a measurement probe fastened to the end of a flexible cable, but also in the case of any measurement conduit of a nuclear reactor or of any other industrial plant employing a fluid under pressure and requiring remote measurements. Still more generally, the shutoff device according to the invention may be employed in any vertical conduit used to support and guide an element of elongate shape, and into which a fluid under pressure is liable to enter accidentally. |
summary | ||
description | Embodiments of the invention relate generally to computed tomography (CT) imaging and, more particularly, to a composite material pre-patient x-ray collimator for use as part of a CT imaging system and a method of manufacturing thereof. Typically, in computed tomography (CT) imaging systems, an x-ray source emits a fan-shaped beam toward a subject or object, such as a patient or a piece of luggage. Hereinafter, the terms “subject” and “object” shall include anything capable of being imaged. The beam, after being attenuated by the subject, impinges upon an array of radiation detectors. The intensity of the attenuated beam radiation received at the detector array is typically dependent upon the attenuation of the x-ray beam by the subject. Each detector element of the detector array produces a separate electrical signal indicative of the attenuated beam received by each detector element. The electrical signals are transmitted to a data processing system for analysis, which ultimately produces an image. In operation, the x-ray source and the detector array are rotated about the gantry within an imaging plane and around the subject. The x-ray source is typically in the form of an x-ray tube that emits x-rays at a focal point, with the x-rays being emitted along diverging linear paths in an x-ray beam. A pre-patient collimator is employed for shaping a cross-section of the x-ray beam and for directing the shaped beam through the patient and toward the detector array. The detector array typically includes a collimator for collimating x-ray beams, a scintillator for converting x-rays to light energy adjacent the collimator, and photodiodes for receiving the light energy from the adjacent scintillator and producing electrical signals therefrom. In CT imaging systems, the pre-patient collimator used for shaping the x-ray beam has historically been constructed by machining a monolithic piece of tungsten. Forming the pre-patient collimator from tungsten was appropriate because of the material's radiation blocking ability and structural properties. However, it is recognized that tungsten is an expensive material and difficult to machine. Additionally, in newer CT imaging systems that implement larger patient coverage, faster rotation speed, and larger bore sizes, pre-patient collimators formed from tungsten become even less desirable. That is, in such newer CT imaging systems, the centripetal acceleration (i.e., G-load) increases dramatically on the pre-patient collimator due to the increasing radius from the center of rotation, faster rotation speed of the components in the gantry, and larger pre-patient collimator size needed to block the beam in large-coverage systems. The weight and forces imposed on the pre-patient collimator are of concern as it affects dynamic balance of the CT imaging system, as well as agility of motion for the collimator. Lead has also been recognized as a possible material from which to construct a pre-patient collimator, as lead also exhibits ideal radiation blocking capabilities associated with its material density. Unfortunately, similar to the use of tungsten pre-patient collimators, the high density of lead means that a pre-patient collimator constructed of lead is affected by the G-load increase in newer CT imaging systems. Additionally, lead is recognized as being too soft to be useful as a monolithic material and is not compliant under the Restriction of Hazardous Substances Directive (RoHS). Therefore, it would be desirable to design a pre-patient collimator that combines the blocking power of a high-density material with the structural support of a lower density substrate material, therefore cutting back on weight and cost of the collimator, while preserving robustness, radiation blocking ability, and RoHS compliance. The invention is a directed method and apparatus for providing a composite material pre-patient x-ray collimator for use as part of a CT imaging system. According to one aspect of the invention, a pre-patient collimator for shaping an x-ray beam in a computed tomography (CT) system includes a base comprised of a first material, the first material having a first material density, and an insert mechanically coupled to the base and being comprised of a second material, the second material comprising a moldable material having a second material density greater than the first material density and that is sufficient to block high frequency electromagnetic energy. The base comprises a plurality of structural features by which the insert is molded to the base, with the moldable material of the insert forming a connection with the plurality of structural features to mechanically couple the base and the insert. According to another aspect of the invention, a method of manufacturing a pre-patient collimator for use in a computed tomography (CT) system includes the steps of forming a base from a first material, the base being formed so as to have a plurality of geometrical features thereon and molding a second material onto the base to form an insert, the second material comprising a material having a material density greater than that of the first material and that is sufficient to block high frequency electromagnetic energy. The second material is injection molded onto the base such that the second material forms a mechanical bond with the plurality of geometrical features to secure the insert to the base. According to yet another aspect of the invention, a computed tomography (CT) system includes a rotatable gantry having an opening to receive an object to be scanned, a high frequency electromagnetic energy projection source configured to project a high frequency electromagnetic energy beam toward the object, and a collimator positioned between the high frequency electromagnetic energy projection source and the object configured to shape the high frequency electromagnetic energy beam, the collimator comprising a pair of blades. The CT system also includes a detector array configured to detect high frequency electromagnetic energy passing through the object and generate a detector output responsive thereto, a data acquisition system (DAS) connected to the detector array and configured to receive the detector output, and an image reconstructor connected to the DAS and configured to reconstruct an image of the object from the detector output received by the DAS. Regarding the collimator, each blade of the collimator further includes a metallic base formed of a first material and comprising a plurality of geometrical features thereon formed therein and an insert mechanically coupled to the base that is formed of a radiation blocking material having a material density greater than a material density of the first material, with the insert being mechanically coupled to the metallic base by way of the plurality of geometrical features, such that the blade is free of adhesives and fasteners for coupling the metallic base and the insert. Various other features and advantages will be made apparent from the following detailed description and the drawings. The operating environment of the invention is described with respect to a sixty-four-slice computed tomography (CT) system. However, it will be appreciated by those skilled in the art that the invention is equally applicable for use with other multi-slice configurations. Moreover, the invention will be described with respect to the detection and conversion of x-rays. However, one skilled in the art will further appreciate that the invention is equally applicable for the detection and conversion of other high frequency electromagnetic energy. The invention will be described with respect to a “third generation” CT scanner, but is equally applicable with other CT systems. Referring to FIG. 1, a computed tomography (CT) imaging system 10 is shown as including a gantry 12 representative of a “third generation” CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays toward a detector assembly or collimator 18 on the opposite side of the gantry 12. Referring now to FIG. 2, detector assembly 18 is formed by a plurality of detectors 20 and data acquisition systems (DAS) 32. The plurality of detectors 20 sense the projected x-rays 16 that pass through a medical patient 22, and DAS 32 converts the data to digital signals for subsequent processing. Each detector 20 produces an analog electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuated beam as it passes through the patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to an x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has some form of operator interface, such as a keyboard, mouse, voice activated controller, or any other suitable input apparatus. An associated display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28 and gantry motor controller 30. In addition, computer 36 operates a table motor controller 44 which controls a motorized table 46 to position patient 22 and gantry 12. Particularly, table 46 moves patients 22 through a gantry opening 48 of FIG. 1 in whole or in part. As shown in FIG. 2, CT system 10 also includes a pre-patient collimator 50 mounted on gantry 12 and positioned in proximity to x-ray source 14. Collimator 50 is constructed to shape the cross-section of x-ray beam 16 into a shape that matches the shape of detector array 18, such as a rectangular shape, for example. The collimator 50 thus ensures that a patient being scanned is not subjected to an unnecessary dose of x-rays. Referring now to FIG. 3, an exploded view of the collimator 50 is provided according to one embodiment of the invention. As shown in FIG. 3, collimator 50 is formed of a pair of blades 51 that are adjustable relative to one another so as to vary the size of an aperture 53 formed therebetween for allowing the x-ray beam 16 (FIG. 2) to pass there through. Varying of the size/shape of aperture 53 thus determines the cross-section of x-ray beam 16 and provides for the control and modification of the beam to a desired size/shape, such as to match the shape of detector array 18. As shown in FIG. 3, each blade 51 is further formed of a base 52 and an insert 54 attached to the base 52. According to embodiments of the invention, each blade 51 of collimator 50 is formed as a composite component, in that the base 52 and the insert 54 are constructed of different materials. More specifically, the base 52 and the insert 54 are formed of materials having different densities. Each blade 51 is formed from a lower density structural material and a high-density radiation blocking material to achieve a resulting composite part that serves to selectively block radiation while also minimizing a weight and cost of the overall collimator 50. The base(s) 52 of each blade 51 of collimator 50 is constructed to provide structural support to the collimator 50 and allow for securing to gantry 12 of the CT system 10 (FIG. 1), while also being designed to lower the overall weight and cost of the collimator 50. As such, the base 52 is composed of a lower density structural material that is selected based on its ability to preserve the robustness of the collimator 50 without adding undue weight and cost to the collimator 50. As used herein, the term “lower density structural material” refers to a material having a density that is not sufficient to block high frequency electromagnetic energy (e.g., x-ray radiation) from passing there through. According to embodiments of the invention, the base 52 may therefore be formed of aluminum, steel, or another similarly acceptable material, that can be machined to have desired structural characteristics, as set forth in detail below. The insert 54 of collimator 50 is constructed to provide radiation blocking within the collimator 50. As such, the insert 54 is composed of a high-density radiation blocking material. As used herein, the term “high-density radiation blocking material” refers to a material having a density that is sufficient to block high frequency electromagnetic energy (e.g., x-ray radiation) from passing there through. According to embodiments of the invention, the insert 54 may therefore be formed, in part, of tungsten or another similarly acceptable material, that serves to block and shape the beam of x-rays 16 emitted from x-ray source 14 (FIG. 1), for example. According to an exemplary embodiment, the insert 54 is formed of a moldable high-density radiation blocking material, such as tungsten impregnated plastic, so that the insert 54 can be secured to base 52 by way of mechanical bonding, as set forth in detail below. According to embodiments of the invention, the composite material blades 51 of collimator 50 are constructed such that base 52 is mechanically bonded to insert 54 without the use of adhesives or mechanical fasteners (e.g., bolts, screws, etc.). According to an exemplary embodiment, insert 54 is formed of a moldable material (e.g., tungsten impregnated plastic) that is molded onto base 52, such as by way of mechanical over-molding or injection molding, to form an inseparable blade 51 in collimator 50. To facilitate the mechanical bonding of the base 52 and the insert 54, the base 52 is constructed to include a plurality of geometrical or structural features thereon that “lock” the insert 54 to the base 52 during a molding of the insert 54 thereto. Referring now to FIG. 4, a detailed view of base 52 is shown according to an exemplary embodiment of the invention. As shown in FIG. 4, base 52 includes a plurality of geometrical/structural features 56, 58, 60 thereon that provide for a mating of insert 54 thereto when insert 54 is applied/formed via a mechanical over-molding or injection molding process. The base 52 includes a series of undercuts 56 for receiving the insert 54, with the undercuts 56 being formed at opposing ends and sides of an insert placement area 62, for example. The base 52 also includes a series of holes 58 spaced apart in the insert placement area 62, with the holes 58 having counter bores 60 on an exit surface 64 of the base 52 opposite from where insert 54 is placed. According to an exemplary embodiment, undercuts 56 and holes 58 with counter bores 60 are formed in base 52 by way of a machining operation, such as according to standard machining procedures of an aluminum/steel material. According to another embodiment, a notch 65 is cut into the back of the base to serve as a gate for the introducing high-density radiation blocking material of the insert in the molding process. Referring now to FIG. 5, a detailed view of insert 54 is shown according to an exemplary embodiment of the invention. As shown in FIG. 5, insert 54 includes a main face 66 that is generally formed in insert placement area 62 of base 52. According to an exemplary embodiment, face 66 is formed as a curved face that accommodates an asymmetric x-ray beam, optimizes placement from x-ray source 14 (FIG. 1) based on a radius of curvature of the face, and provides full x-ray beam blocking. Alternatively, it is recognized that face 66 could also be formed as a straight (i.e., non-curved) face, according to another embodiment of the invention. Included on insert 54 are lips or protrusions 68 formed at opposing ends of face 66, with the lips/protrusions 68 being formed to mate with undercuts 56 formed on base 52 (FIG. 4). Also included on insert 54 are a series of anchors 70 that extend out from a back surface 72 of face 66 and down through holes 58 formed in base 52. At an end of the anchors 70 distal from face 66, circular flanges 74 are formed that mate with the counter bore 60 of holes 58, to lock the anchor 70 to base 52. As indicated above, insert 54 is formed by way of a mechanical over-molding or injection molding process, such that the lips/protrusions 68 of face 66 and the anchors 70 form a locking mechanical bond with the geometrical features 56, 58, 60 of the base 52, i.e., the undercuts 56 and holes 58 with counter bores 60 formed on/in the base 52. According to an embodiment of the invention, each blade 51 of the collimator 50 can thus be manufactured by first forming an base 52 from a piece of aluminum or steel, for example, with the aluminum/steel being machined to form an base having a plurality of geometrical features formed thereon. As set forth above, the geometrical features may be in the form of a series of undercuts 56 and holes 58 having counter bores 60 formed therein. Upon machining of the base 52, the insert 54 is molded to the base by way of an over-molding or injection molding process. In molding the insert 54 to the base 52, a number of protrusions 68 are formed on the insert that mate with the undercuts 56 of the base. Additionally, a number of anchors 70 are formed on the insert 54 that mate with the holes 58 and counter bores 60 of the base 52. The molding of the insert 54 to the base 52 forms a mechanical bond there between that secures the insert to the base, without the need for any adhesives and/or fasteners. Referring now to FIG. 6, a package/baggage inspection system 100 is shown according to an embodiment of the invention, with the package/baggage inspection system 100 incorporating a pre-patient collimator 50 such as shown in FIG. 3. As shown in FIG. 6, package/baggage inspection system 100 includes a rotatable gantry 102 having an opening 104 therein through which packages or pieces of baggage may pass. The rotatable gantry 102 houses a high frequency electromagnetic energy source 106 as well as a detector assembly 108 having scintillator arrays comprised of scintillator cells similar to that shown in FIG. 6 or 7. A conveyor system 110 is also provided and includes a conveyor belt 112 supported by structure 114 to automatically and continuously pass packages or baggage pieces 116 through opening 104 to be scanned. Objects 116 are fed through opening 104 by conveyor belt 112, imaging data is then acquired, and the conveyor belt 112 removes the packages 116 from opening 104 in a controlled and continuous manner. As a result, postal inspectors, baggage handlers, and other security personnel may non-invasively inspect the contents of packages 116 for explosives, knives, guns, contraband, etc. Beneficially, the composite material pre-patient collimator 50 can be optimized for radiation blocking ability, rigidity, and weight. The reduced weight of the collimator reduces demand on motors and bearings of the CT system since it is lighter. Additionally, the structure of the composite material pre-patient collimator blade, and the mechanical bonding/locking provided thereby, eliminates the need for any separate fasteners or adhesives to be used, thereby also eliminating any leakage points that might be created by the use of such fasteners. Additionally, the molding of the blocking material insert to the base allows for flexibility in geometry and application, and reduces waste and cost by providing a near net shape of the molded insert, such that the resulting insert is more environmentally friendly than using adhesives, for example. Still further, the structure of the composite material pre-patient collimator provides for machining of the assembly that is easier and cheaper than machining a pure tungsten collimator assembly. Therefore, according to one embodiment of the invention, a pre-patient collimator for shaping an x-ray beam in a computed tomography (CT) system includes a base comprised of a first material, the first material having a first material density, and an insert mechanically coupled to the base and being comprised of a second material, the second material comprising a moldable material having a second material density greater than the first material density and that is sufficient to block high frequency electromagnetic energy. The base comprises a plurality of structural features by which the insert is molded to the base, with the moldable material of the insert forming a connection with the plurality of structural features to mechanically couple the base and the insert. According to another embodiment of the invention, a method of manufacturing a pre-patient collimator for use in a computed tomography (CT) system includes the steps of forming a base from a first material, the base being formed so as to have a plurality of geometrical features thereon and molding a second material onto the base to form an insert, the second material comprising a material having a material density greater than that of the first material and that is sufficient to block high frequency electromagnetic energy. The second material is injection molded onto the base such that the second material forms a mechanical bond with the plurality of geometrical features to secure the insert to the base. According to yet another embodiment of the invention, a computed tomography (CT) system includes a rotatable gantry having an opening to receive an object to be scanned, a high frequency electromagnetic energy projection source configured to project a high frequency electromagnetic energy beam toward the object, and a collimator positioned between the high frequency electromagnetic energy projection source and the object configured to shape the high frequency electromagnetic energy beam, the collimator comprising a pair of blades. The CT system also includes a detector array configured to detect high frequency electromagnetic energy passing through the object and generate a detector output responsive thereto, a data acquisition system (DAS) connected to the detector array and configured to receive the detector output, and an image reconstructor connected to the DAS and configured to reconstruct an image of the object from the detector output received by the DAS. Regarding the collimator, each blade of the collimator further includes a metallic base formed of a first material and comprising a plurality of geometrical features thereon formed therein and an insert mechanically coupled to the base that is formed of a radiation blocking material having a material density greater than a material density of the first material, with the insert being mechanically coupled to the metallic base by way of the plurality of geometrical features, such that the blade is free of adhesives and fasteners for coupling the metallic base and the insert. This written description uses examples to disclose the invention, including the best mode, and also to enable any person skilled in the art to practice the invention, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the invention is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims. |
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abstract | The present invention provides a charged particle beam apparatus used to measure micro-dimensions (CD value) of a semiconductor apparatus or the like which captures images for measurement. For the present invention, a sample for calibration, on which a plurality of polyhedral structural objects with known angles on surfaces produced by the crystal anisotropic etching technology are arranged in a viewing field, is used. A beam landing angle at each position within a viewing field is calculated based on geometric deformation on an image of each polyhedral structural object. Beam control parameters for equalizing the beam landing angle at each position within the viewing field are pre-registered. The registered beam control parameters are applied according to the position of the pattern to be measured within the viewing field when performing dimensional measurement. Accordingly, the present invention provides methods for reducing the variation in the CD value caused by the variation in the electron beam landing angle with respect to the sample with an equal beam landing angle and methods for reducing the instrumental error caused by the difference in the electron beam landing angle between apparatuses. |
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044341334 | summary | BACKGROUND OF THE INVENTION Coal can be efficiently converted into hydrocarbons of a more useful gaseous or liquid form by coal gasification or liquefaction techniques, utilizing energy from a high-temperature, gas-cooled nuclear reactor for the endothermic and/or electrolytic processing required, as taught by Jones, in U.S. Pat. No. 4,158,673. While the United States, the Soviet Union, and China still contain major deposits of coal, this mineral is considered precious in most other parts of the world, where deposits are either lacking or have been largely used up. Thus, while the earth's supply of precious fossil fuels is being steadily depleted to provide electricity and petrochemicals, a virtually unlimited worldwide supply of other carbon bearing minerals remains untapped as an energy source. Salotti, in U.S. Pat. No. 3,558,724, taught that inorganic crystalline carbonates could provide gaseous products containing up to 4% methane, if the carbonates were first heated in an oxygen-free atmosphere at from about 400.degree. C. to 700.degree. C., and then contacted with excess hydrogen gas at from about 200 psi. to 10,000 psi. This process, however, uses large quantities of valuable hydrogen gas, which is becoming increasingly important itself as an energy source. In addition, this process provides a poor yield of methane, leaves carbon residue and maintains explosive reaction conditions. Tamers, in U.S. Pat. No. 4,009,219, taught the production of benzene from inorganic carbonates such as limestone. Tamers reacted limestone (CaCO.sub.3) with lithium metal in a vacuum at 500.degree. C., raised the temperature of reaction to 1,000.degree. C. for 30 minutes to reverse secondary reactions that produce carbon and metal oxide, and then hydrolyzed the resulting product lithium carbide (Li.sub.2 C.sub.2) with water. This produced acetylene, with calcium oxide and lithium hydroxide by-products. The lithium hydroxide was converted to lithium metal by fused salt electrolysis, and recycled back to the limestone reaction. The acetylene was then purified and dried. Benzene was produced from this treated acetylene, using dried potassium chromate activated silica-alumina catalyst at 120.degree. C. to 200.degree. C. Benzene, however, is now known to be toxic and a carcinogenic agent. What is needed is a method to produce high carbon chain hydrocarbons without using valuable fuels such as coal or hydrogen or producing toxic substances such as benzene. SUMMARY OF THE INVENTION It has been discovered that the above-described need can be met by a process comprising the steps of: (1) reacting inorganic, crystalline or non-crystalline, carbon containing mineral material, such as CaCO.sub.3, with a stoichiometric excess of molten lithium metal, in a lithium reactor means, in an inert atmosphere, at a temperature of between about 300.degree. C. and about 1,200.degree. C., to provide lithium salt compounds such as Li.sub.2 C.sub.2 and Li.sub.2 O, and formation of CaO; (2) hydrolyzing the Li.sub.2 C.sub.2 to provide C.sub.2 H.sub.2 (acetylene gas). A lithium hydroxide slurry formed from the Li.sub.2 O can be recovered from the hydrolysis step and converted to lithium metal by a variety of means, such as fused salt electrolysis, for recycle to the lithium reactor means; (3) reacting the C.sub.2 H.sub.2 with steam at between about 250.degree. C. and about 475.degree. C. in the presence of catalysts such as ZnO, to provide CH.sub.3 COCH.sub.3 (acetone). The C.sub.2 H.sub.2 does not have to be purified for this reaction to occur; (4) pyrolyzing the CH.sub.3 COCH.sub.3 at between about 600.degree. C. and about 800.degree. C., to provide ketene gas, which is then cooled to -60.degree. C. by a suitable cooling means, to provide a ketene product (CH.sub.2 .dbd.C.dbd.O) in liquid form, and separable methane gas. The ketene can then be stored as a liquid at 25.degree. C. under a pressure of about 40 psi, if desired. In reduction step (1) and hydrolysis step (2), a total of approximately 976 kJ. of energy is gained for each mole of CaCO.sub.3 used. Some of this energy can be used to keep the lithium reactor at temperature. The CaO still present in step (2) can be separated from the lithium hydroxide slurry for use in other industries, such as an alkali for water treatment, etc. By-product hydrogen from step (3) and methane from step (4) can be separated and used as fuels. The electrolysis step to provide lithium metal, and the pyrolysis step to provide ketene require a large expenditure of energy, which could be supplied by a pressurized water nuclear reactor, or a high-temperature, gas-cooled nuclear reactor, without major modifications in design or structure of the reactor. Energy from the nuclear reactor could also be used to help keep the lithium reactor at temperature, and to provide steam and heat to produce acetone from acetylene. Thus, uranium would be the chief fuel consumed in the process of converting the inorganic carbon into an organic carbon, which can be used to produce liquid hydrocarbon fuels such as diesel oil or gasoline. The ketene can be photochemically decomposed or thermally decomposed to produce extremely reactive, organic methylene. A particular type of insertion chain reaction sequence could then be started between methylene and alkanes, such as methane, which could be easily supplied as a by-product from the pyrolysis reaction. This could ultimately result in a mixture of ethane, propane, butane, pentane, hexane, heptane, and possibly higher carbon chain hydrocarbons such as octane. The pyrolysis of acetone, the decomposition of ketene and the methylene insertion chain reaction to form ethane and higher hydrocarbons, can be accomplished as separate steps, or they can be combined in various fashions. Thus, by the method of this invention, limestone type minerals can be reacted to form gasoline-type molecules via a ketene reaction sequence, utilizing energy from an already existing nuclear power source. |
abstract | Provided is a fuel assembly 1 for a PWR nuclear reactor capable of stably creating transverse flows of coolant for pressing and securing control rods so as to restrain vibration of the control rods in order to restrain the outer surfaces of the control rods and the inner surfaces of control rod guide tubes from being worn, wherein an upper nozzle arranged above the fuel assembly comprises an adapter plate 6 constituting the lower structure of the upper nozzle, a side wall extended along the periphery of the adapter plate, an overhang projected into a space above the adapter plate from the upper part of the side wall, and apertures for attaching control guide tubes and passage holes, which are formed in the passage surface of the adapter plate, and wherein those 15A, 15B, 15C of the passage holes which are located at positions where the coolant impinges upon the overhang are generally arranged, line-symmetric with respect to diagonal lines of the passage surface serving as symmetric axes Q, and ligaments 21, 23 around the passage holes located inside and outside of those 11a of the attaching apertures which are located on the outer peripheral side are set to be larger than ligaments 22 around the passage holes which are located on opposite transverse sides of those of the attaching apertures. |
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abstract | A graphical interface is provided to manage interfaces with hardware and software devices. The graphical communication interface can be operated on an electronic device to simplify management of one or more interfaces, while providing opportunities for enhanced capabilities and control of the interfaces. The graphical communication interface can create objects that are associated with hardware or software devices. The objects are representative of the device and are depicted in the graphical interface. The object is configured to be interactive with the device and enable communication between the graphical interface and the hardware device. The graphical interface can include both software objects and hardware objects and the objects can include user-defined protocols to communicate with the device, allowing communications with a wide variety of devices. Analysis objects may also be created for interaction with the hardware objects or software objects. |
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description | The present application claims priority from Japanese application JP 2005-140588 filed on May 13, 2005, the content of which is hereby incorporated by reference into this application. The present invention relates to a charged particle beam device, in particular to a scanning electron microscope, an electron beam semiconductor inspector, an electron beam semiconductor dimension measurement device, a focused ion beam device and the like, each of which has an aberration corrector. A scanning electron microscope (SEM) has a higher resolution than an optical microscope in the observation of the surface of an object, and hence it is widely used not only as a device for research but also as an industrial device for the dimension measurement of semiconductor wafer patterns which have increasingly been miniaturized in recent years, the observation of foreign matters on a surface and the like. In the case of the dimension measurement of a semiconductor, a high resolution of several nm at a low acceleration voltage of 1 kV or lower has increasingly been required. The resolution of a SEM depends on how to focus an electron beam into a smaller spot on the surface of a specimen, and hence it is dominated by a diffraction aberration, the chromatic aberration and the spherical aberration of an electron lens as well as the size of the electron source reduced and focused with the lens. The resolution has heretofore been improved by devising an electron optics system, in particular by increasing the reduction ratio of an electron source, optimizing the shape of the object lens by the combination of acceleration and deceleration electric fields, and thus decreasing the aberration. However, it has already been proved by Scherzer that it is impossible to make the spherical and chromatic aberrations zero with an object lens rotationally symmetrical to the optical axis, and the improvement of a resolution by such conventional measures has been restricted from the aspects of a shape and dimension, machining accuracy, material quality, breakdown voltage and others. In view of the above situation, a method for canceling the aberration of an object lens with a chromatic and spherical aberration corrector made by combining quadrupoles and octupoles is proposed (refer to H. Rose, Optik 33 (1971), pp. 1 to 24), and a SEM having an aberration corrector has been put into practical application by Zach and others in 1995 (refer to J. Zach and M. Haider, Nuclear Instruments and Methods in Physics Research A363 (1995), pp. 316 to 325). In the event of the actual use of an aberration corrector, the adjustment of the strength of each pole, the alignment of the poles, and the alignment of the whole system including an object lens and the aberration corrector are very important. The above document by Zach and others discloses the method of judging and adjusting the controlled variables of a multipole in consideration of the amounts, directions and symmetry of the blurring of SEM images. Further, the document of S. Uno, K. Honda, N. Nakamura, M. Matsuya, J. Zach Proc. of 8APEM (2004), pp. 46 to 47 and Published Japanese Translation of PCT No. 521801/2003 disclose the method of estimating the magnitude of various kinds of geometric aberrations by deconvolution through the Fourier transformation of plural SEM images and feeding it back to the control of a multipole. Furthermore, Published Japanese Translation of PCT No. 505899/2005 discloses the method of, in the event of astigmatic correction, modulating the beam energy of charged particles, thus obtaining scanning images, and adjusting the alignment of columns from the deviation of the images and the change of the definition. In addition, JP-A No. 355822/2004 discloses the method of applying beam scanning three times to an identical line on a specimen while changing the energy thereof, thus forming three images, and adjusting a chromatic aberration corrector from the deviation among the images and the change of the definition. When an aberration corrector is used, how to actually adjust a multipole is important and the operation of a multipole is complicated and requires sufficient experiences. Though the aforementioned Published Japanese Translation of PCT Nos. 521801/2003 and 505899/2005 explain the operation principle of the adjustment method of an aberration corrector, they disclose neither a means of simply adjusting an aberration corrector nor a device having a configuration necessary for the means. For example, the aforementioned Published Japanese Translation of PCT No. 521801/2003 does not disclose at all the chromatic aberration correction which is necessary as the preliminary step of spherical aberration correction. Further, the aforementioned Published Japanese Translation of PCT No. 505899/2005 only describes that “astigmatism can be corrected with an astigmatic corrector” and does not disclose at all the internal configuration of the astigmatic corrector and the concrete adjustment method thereof. Furthermore, in the case of the means disclosed in JP-A No. 355822/2004, it is necessary to adjust plural images while comparing them on a display. In view of the above situation, the object of the present invention is to provide a charged particle beam device provided with an aberration corrector which has a simpler device configuration and better operability than conventional one and is capable of correcting a chromatic aberration and/or a spherical aberration. In a charged particle beam device having an aberration correction means, a multi-hole stop is used at the time of adjusting the aberration correction means and scanning charged particle beam images (for example, SEM images or SIM images) are formed with plurally divided beams. A specimen is scanned with plural beams and hence overlapping images are obtained. Then it is possible to correct the chromatic aberration and the spherical aberration with good operability by judging the correction state of the aberrations from the directions and symmetry of the overlapping and feeding back the judgment result to the adjustment of the aberration corrector so as to eliminate the overlapping. Further, it is possible to secure a deep focal depth by using an orbicular zone aperture the center of which is shielded in the state of the aberration correction. The optical axis of a charged particle beam supplied from an identical charged particle source is divided into plurality, and an aberration corrector is adjusted so that the plurally divided optical axes are superimposed. Since the degree of the adjustment of the aberration corrector is visually observed, the operability of the adjustment of the aberration corrector improves outstandingly. Further, since the process of the adjustment under the visual observation can easily be replaced with image processing, the measure is suitable for the automation of the adjustment of the aberration corrector. An embodiment wherein the present invention is applied to a scanning electron microscope is explained hereunder. A means basically identical to the present embodiment can also be applied to another electron beam application device or a device of the beam of other charged particles such as protons and ions though the configuration of the lens and the aberration corrector is to be changed in accordance with the kinds thereof. FIG. 1 shows a general configuration of a scanning electron microscope as an embodiment according to the present invention. The scanning electron microscope of the present embodiment is roughly composed of: a SEM column 101 to irradiate or scan a specimen with an electron beam; a specimen chamber 102 to contain a specimen stage; a control unit 103 to control components of the SEM column 101 and the specimen chamber 102; and others. Further, to the control unit 103, connected are: a data storage 76 to store prescribed information; a monitor 77 to display obtained images; and an operator console 78 to function as a man-machine interface between the device and device users. The operator console comprises information input means such as a keyboard, a mouse and the like, for example. Firstly, the components in the SEM column 101 are explained. A Schottky electron source 1 is an electron source which is made by diffusing oxygen, zirconium and others in a monocrystal of tungsten and makes use of the Schottky effect thereof. Then, in the vicinity of the Schottky electron source 1, a suppressor electrode 2 and an extraction electrode 3 are placed. Schottky electrons are emitted by heating the Schottky electron source 1 and applying a voltage of about +2 kV between the Schottky electron source 1 and the extraction electrode 3. A negative voltage is applied to the suppressor electrode 2 and thus the discharge of electrons from places other than the tip of the Schottky electron source 1 is inhibited. The electrons passed through the aperture of the extraction electrode 3 are accelerated and converged with an electrostatic lens comprising a first anode 4 and a second anode 5 and then enter the components of the latter stages along an optical axis 60. The electrons are converged with a first condenser lens 6, the beam diameter is regulated with a movable stop 31, and the electrons enter an aberration corrector 10 through a second condenser lens 7 and a deflector 8. Here, a multi-hole stop may be placed at the latter stage of the aberration corrector 10. The deflector 8 is adjusted so that the axis of the condenser lens 7 coincides with the axis of the aberration corrector 10. In the present embodiment, explanations are given on the basis of a quadrupole-octupole type chromatic and spherical aberration corrector 10. A quadrupole and an octupole are formed at the respective stages of the aberration corrector 10 and twelve electrodes (magnetic poles may also be used as those) are used for that purpose. Then, it is also possible to form a dipole, a sextupole or a twelve-pole other than the quadrupole and octupole and those poles are used in order to electrically correct the field distortion caused by the assembly error of the electrodes or magnetic poles and the unevenness of the magnetic pole material. The electron beam to which the chromatic and spherical aberrations, which are to be compensated with an object lens 17, are rendered by the aberration corrector 10 is converged on a specimen 18 with the object lens 17, and the surface of the specimen is scanned with the converged spot by using a scanning deflector 15. The reference numeral 38 represents an objective aligner. A specimen stage 80 having a specimen mounting table on which the specimen 18 is mounted is contained in the specimen chamber 102. Secondary electrons generated by electron beam irradiation pass through the object lens 17, hit a reflector 72, and generate electrons. The generated electrons are detected by a secondary electron detector 73 but it is also possible to adjust the position of the reflector 72 hit by the secondary electrons with an E×B deflector 71. The detected secondary electron signals are introduced into a control computer 30 as luminance signals synchronizing with the scanning. The control computer 30 applies appropriate processing to the introduced luminance signal information and the processed information is displayed as SEM images on the monitor 77. Only one detector is shown in the figure but it is also possible to dispose plural detectors so that images can be obtained by selecting the energy and angle distributions of reflected electrons and secondary electrons. The reflector 72 is not necessarily required if either the secondary electrons are directly accumulated to the secondary electron detector 73 with the E×B deflector 71 or a secondary electron detector of a coaxial disk shape having an aperture in the center thereof is disposed on the optical axis 60. The control unit 103 is composed of: an electron gun power source 20; a control voltage source 21; an acceleration voltage source 22; a first condenser lens power source 23; a second condenser lens power source 24; a deflection coil power source 25; an aberration corrector power source 26; a scanning coil power source 27; an object lens power source 28; a retarding power source 29; a movable stop micromotion mechanism 32; an anastigmatic coil power source 35; an objective aligner power source 37; a secondary electron detector power source 74; an E×B deflector power source 75; a specimen stage control mechanism 81; and others. Then, they are connected to relevant components in the SEM column with signal transmission lines, electric wiring and the like. FIG. 2 shows an example of the configuration of the movable stop 31. The movable stop shown in FIG. 2 has, in addition to a round aperture 310 which is usually used for regulating the amount of the beam electric current, apertures 311 and 312 for chromatic aberration correction, and apertures 317 and 318 for spherical aberration correction. Because of space limitation, the apertures 311 and 312 are shown on the upper side and the apertures 317 and 318 are shown on the lower side. However, in reality, those apertures are formed on the same single sheet. Here, each of the apertures 311, 312, 317 and 318 has plural apertures respectively and thus, in order to distinguish the former “aperture” from the latter “aperture,” the apertures 311, 312, 317 and 318 are referred to as the aperture groups 311, 312, 317 and 318. The aperture group 311 is a five-hole aperture having four small holes located around the center hole at angles of 90 degrees, namely disposed at fourfold symmetrical positions. Then, the aperture group 312 is a five-hole aperture formed by rotating the small holes of the aperture group 311 at an angle of 45 degrees. Here, if the center hole is not formed, the probe current is insufficient and that makes the adjustment difficult. This is because, in the state where the aberration is corrected, the beam passing through the center hole and the beams passing through the small holes of the periphery are converged into single spot with the object lens. Further, the reason why the peripheral small holes are smaller than the center hole is to clarify the direction of correction (the adjustment is carried out in the direction where the peripheral images coincide with the center image). By increasing the strength of the image formed by the beam passing through the center hole and reducing the strength of the images formed by the peripheral holes in the adjustment of aberration correction, the strength of the image formed by the beam passing through the center hole increases and the strength of the images formed by the peripheral holes is reduced in the obtained secondary electron images. Thereby, the visibility improves at the time of the aberration correction and it becomes easy to carry out the image processing when it is automated. The aperture groups 317 and 318 are the aperture groups used for spherical aberration correction and the distances between the center hole and the peripheral small holes are larger than those of the aperture groups 311 and 312 used for chromatic aberration correction. It is possible to make the degree of separation of the images easily visible by changing the amplitude of high voltage wobbler in the case of the adjustment of the chromatic aberration correction, but it is necessary to carry out adjustment by changing the distances between the holes in the case of the adjustment of the spherical aberration. Though it is basically desirable to prepare aperture groups having various distances between holes, a stop having large distances between holes is prepared. Meanwhile, the stop may be used for both the chromatic aberration adjustment and the spherical aberration adjustment. However, in this case, image separation caused by the spherical aberration still remains at the time of the chromatic aberration correction and hence the endpoint of the adjustment of the chromatic aberration correction is hardly identified. With a stop having small distances between holes, the image separation caused by the spherical aberration can be ignored and thus the endpoint of the adjustment of the chromatic aberration correction is easily identified. Next, the adjustment procedures of the aberration corrector 10 are explained with the flowchart shown in FIG. 3. Here, the control of aberration correction explained below can also be automated but, unless otherwise noted, it is assumed that the adjustment is manually operated by a device user. Note that, for better understanding, the steps which can be operated by the control computer 30 when automated are shown by the oval frames in the flowchart shown in FIG. 3. The procedures of the adjustment of the aberration correction in the present embodiment roughly comprise the step of chromatic aberration correction and the step of spherical aberration correction. At the time of the chromatic aberration correction, the adjustment is sometimes carried out by changing the paraxial trajectory of electrons. In that case, the conditions of the spherical aberration correction are changed and hence it is necessary to carry out the chromatic aberration correction in advance. Firstly, the step of the chromatic aberration correction is explained. (1) Firstly, a series of ordinary axis matching of a SEM is carried out in the state of the deactivation of the aberration corrector. Next, (2) the quadrupole at each stage of the aberration corrector 10 is excited in sequence and the axis of the aberration corrector 10 is made to roughly coincide with the optical axis with the deflector 8 and the scanning deflector 15 so that the beam passes through nearly the center of the quadrupole. In the adjustment of the chromatic aberration correction, the strength of the quadrupole at each stage of the aberration corrector is set at an initial value determined beforehand by calculation or the like so as to create the state wherein images can be observed, and thereafter (3) the five-hole aperture 312 is inserted into the optical axis, and (4), in the state, acceleration voltage is cyclically changed. In the present embodiment, the control of cyclically applying variable voltage to acceleration voltage is referred to as “high voltage wobble.” The image magnification of a SEM in this case is lowered to the extent that the influence of the spherical aberration is not observed. In the case of the high voltage wobble of the present embodiment, AC voltage is applied between the second anode 5 and an earth potential with a high voltage wobbler power source 33. Thereby, the acceleration voltage of the electron beam is modulated at a prescribed cycle and high voltage wobble is realized. FIGS. 4A to 4C schematically show how a SEM image displayed on a monitor is varied by the high voltage wobble in the state of inserting the five-hole aperture 312 into the optical axis. The electron beam having passed through the first condenser lens shown in FIG. 1 passes through the five-hole stop, thereby the optical path thereof is separated into five paths, and the separated electron beams are introduced into the object lens. The focal plane moves vertically along the optical axis due to the variation of the acceleration voltage and hence the specimen surface is scanned with five spots between the upper limit and the lower limit of the varying acceleration voltage. As a result, the SEM image is observed as a quintuple image formed by superimposing the center image and the peripheral images located around the center image at angles of 90 degrees as shown in FIGS. 4A to 4C. As a specimen used for the adjustment, a specimen having a symmetrical shape is desirable, in particular a spherical shape is suitable. Here, in the present embodiment, a latex ball is used as the specimen. FIG. 4A shows the state of not applying variable voltage. Among the electron beams of the five optical paths having passed through the object lens, the electron beam having passed through the center hole is controlled with the object lens so as to form a focus on the viewing screen (on the surface of the specimen 18). Meanwhile, the electron beams of the four peripheral optical paths other than the center optical path form a crossover en route and hit the specimen surface. This is because the focal plane of the electron beams passing through the four peripheral optical paths is different from the focal plane of the electron beam passing through the center optical path. In the case of the present embodiment, the focal plane is formed at the crossover position located on the front side of the specimen, namely on the side of the electron source. As a result, in the SEM image observed in the state shown in FIG. 4A, a sharp (namely, well-focused) ring-shaped image formed by the electron beam having passed through the center hole and four ring-shaped images of blurring profiles (namely somewhat defocused) formed around the periphery thereof are observed. FIG. 4B is a schematic view of the case where the amplitude of the applied variable voltage is negative. In the figure, the focal plane of the electron beams having passed through the five holes, namely the crossover position, moves toward the object lens. As a result, in the SEM image observed in the state shown in FIG. 4B, the images corresponding to the electron beams having passed through the five holes are observed in the state where the blurring of the focus is larger than that observed in the case shown in FIG. 4A. Further, with regard to the images formed by the electron beams having passed through the four peripheral small holes, the crossover position of the beams becomes farther from the viewing screen and hence the distances between the center position of the ring-shaped image in the center and the center positions of the four peripheral ring-shaped images increase. FIG. 4C is a schematic view of the case where the amplitude of the applied variable voltage is positive. In the figure, the focal plane of the electron beams having passed through the five holes moves toward the specimen surface. As a result, in the observed SEM image, the blurring of the focus in all the five ring-shaped images becomes larger than that observed in the case shown in FIG. 4A. Further, in the case shown in FIG. 4C, the focal plane moves toward the stage side farther from the specimen and apparently the crossover position is also formed inside the specimen. The crossover position of the four peripheral electron beams also moves toward the stage side farther from the specimen surface and hence, in the observed SEM image, the distances between the center position of the ring-shaped image in the center and the center positions of the four peripheral ring-shaped images decrease. The width of the deviation of the five ring-shaped images explained above increases in proportion to the amplitude of the high voltage wobble. Further, the aforementioned adjustment procedures of the aberration corrector is based on the premise of manual operation and a device user adjusts the aberration corrector so that the five rings may be superimposed while visually confirming the SEM image with a monitor. Hence, the amplitude of the applied voltage of the high voltage wobbler is set at such a degree of amplitude as to be able to easily observe the deviation of the quintuple image. When the electron beams enter in the state of deviating from the optical axis of the object lens, the whole overlapping image moves not vertically (when the object lens is a magnetic lens) but transversely (to the right, left, top or bottom, or diagonally) and hence it is possible to carry out optical axis matching with a high degree of accuracy by adjusting the position of the stop 31 so that the whole overlapping image may move vertically around the center of the visual field. (5) When the chromatic aberration correction is advanced while the ratio of the strengths of the electric field and magnetic field of the quadrupole of the second stage is varied, the SEM image which has been separated and superimposed comes to be superimposed on the center image in a certain direction (regarded as x direction). The process is shown in FIG. 5. The adjustment of the strengths of the electric field and magnetic field is carried out by changing the setting parameters of the aberration corrector power source 26 with the operator console 78 while the device user visually confirms the monitor 77. In this case, by setting the phase of the quadrupole of the aberration corrector beforehand so that the direction of the image deviation in which the SEM image changes may coincide with the direction of the small holes among the five holes (for example, the x direction 313), the operation of the adjustment can be facilitated. (6) Next, the ratio of the strengths of the electric field and magnetic field of the quadrupole of the third stage is adjusted so that the SEM image which has been separated in the direction orthogonal to the previous direction may be superimposed. The operation is also carried out while the setting parameters of the aberration corrector power source 26 are changed with the operator console 78. (7) The above processes are repeated while the amplitude of the high voltage wobble is reduced and the processes terminate when the deviated images come close to the center image evenly in the four directions. FIGS. 6A to 6C show the change of the SEM image caused by the high voltage wobble at the time of the completion of the chromatic aberration correction. The primary chromatic aberration is corrected and the secondary chromatic aberration remains. Hence, the deviation of the electron orbit and the change of the SEM image corresponding to the positive and negative changes of the acceleration voltage show similar tendency in both FIGS. 6B and 6C. Meanwhile, in the above explanations, the adjustment method of poles by manual operation has been explained. However, automatic adjustment by computer control can also be adopted practically. In order to realize automatic adjustment, the secondary electron images obtained in the adjustment processes of the quadrupole are taken in the control computer 30 in each of the processes (5), (6) and (7) described above and the degree of the overlapping of the images is quantified. As the means, for example, used is an image processing method such as a phase limit correlation method wherein the image of a specimen before the insertion of a multi-hole stop and the images of the specimen obtained in the processes (5), (6) and (7) are subjected to Fourier transformation and phase limit processing, and the matching of them is obtained. A reference table is stored in the data storage 76. Control information obtained as the result of image processing, and the voltage values to be applied to the poles in relation to the control information are compared and stored in the reference table. The control computer 30 refers to the table, and decides and transmits the voltage to be fed back to the aberration correction power source 26. The table may be stored in the computer 30 itself. In order to automate all the adjustment steps, it is necessary to automate all the steps in the flowchart shown in FIG. 3. However, as long as the adjustment of the poles can be automated, the other steps can easily be automated. Next, the steps of the spherical aberration correction are explained. (8) The high voltage wobble is deactivated and the magnification is raised until the quintuple SEM image may be observed at in-focus by the influence of the spherical aberration. Otherwise, the movable stop 31 is driven and a five-hole stop having wide spaces between small holes, namely the aperture group 317, is selected so that the quintuple SEM image may easily be observed at in-focus. When the voltage is raised by operating the octupole of the second stage, the spherical aberration is corrected in a direction (tentatively regarded as x direction) and the images come close to the center image. The voltage of the octupole of the second stage is adjusted so that the images coincide with the center image and likewise the voltage of the octupole of the third stage is adjusted so that the images coincide in the orthogonal y direction. Successively, a five-hole stop having small holes the angles of which deviate from the currently-used five-hole stop by 45 degrees, namely the aperture group 318, is inserted and the voltage of the octupoles of the first and fourth stages is adjusted on the basis of the movement of the images in the direction rotationally deviated from the previous direction by 45 degrees. By so doing, the correction of the spherical aberration (aperture aberration) is completed. In the above adjustment, in the case of the manual adjustment, the device user carries out the above adjustment, in the same way as the chromatic aberration adjustment, by changing the setting parameters of the aberration corrector power source 26 with the operator console 78 while visually observing the monitor 77. Then, in the case of the adjustment by automatic control, in the same way as explained earlier, an appropriate image processing algorithm is used, the degree of the overlapping of the SEM image is quantified, and the poles are controlled so that all the five images are superimposed. Here, though a chromatic and spherical aberration corrector of a quadrupole-octupole type is adopted in the present embodiment, the present invention is not limited to the present embodiment and applicable also to another type of aberration correction means, the process of only the chromatic aberration correction, and the process of only the spherical aberration correction. Then, in the above-explained adjustment method of an aberration corrector, in principle, it is possible to adjust the aberration corrector even when the high voltage wobbler is not applied. The high voltage wobbler is the control which is adopted in order to facilitate the visual confirmation at the time of manual operation and it is also possible to adjust poles so that five ring-shaped images coincide while observing only a SEM image corresponding to FIG. 4A. However, when the high voltage wobbler is used in combination, a SEM image is displayed in the manner of intensifying the degree of aberration and hence there is an effect of facilitating the adjustment. Further, the adjustment is basically possible even when a three-hole stop or a four-hole stop is used as shown in FIG. 7. However, from the viewpoint of visual sense, the use of a stop having small holes disposed fourfold rotationally symmetrically as shown in FIG. 2 makes the adjustment easier. As stated above, by the means explained in the present embodiment, it is possible to realize an aberration corrector which facilitates adjustment and remarkably improve the operability of a charged particle beam application device having the aberration corrector. The present embodiment is explained on the basis of a case where the present invention is applied to a critical-dimension-measurement SEM (scanning electron microscope). As a specimen for dimension measurement, a semiconductor wafer or a semiconductor chip on which a circuit pattern is formed, or a specimen produced by cutting out a part of the wafer may be adopted. FIG. 8 shows the configuration of the hardware of a critical-dimension-measurement SEM in the present embodiment. The general configuration of the critical-dimension-measurement SEM is the same as that of the charged particle beam application device explained in the first embodiment in the fact that the critical-dimension-measurement SEM is composed of: a SEM column 101; a specimen chamber 102 to contain a specimen stage; a control unit 103; and others. However, the critical-dimension-measurement SEM has a specimen preparation chamber (load chamber) 40 to introduce a specimen to be subjected to dimension measurement into the device. The specimen preparation chamber 40 and the specimen chamber 102 in the device main body are separated from each other with a gate valve 42 and, when a specimen is introduced into the main body side, the gate valve opens and the specimen is transferred into the specimen chamber in the device main body by a specimen transfer mechanism 41. With regard to the control unit 103 too, though the general configuration is the same as that of the device used in the first embodiment, in the case of the critical-dimension-measurement SEM, the control computer 30 has the functions of: driving the specimen stage in accordance with the recipe which is input beforehand; obtaining an image at a prescribed position on a wafer; applying image analysis to the detected secondary electron image signals; and thus carrying out the prescribed dimension measurement of a line pattern. Here, the electron optics components contained in the SEM column 101 are the same as those in the first embodiment and hence the explanations thereof are omitted. The operation of the electron optics system is essentially the same as that of the charged particle beam device in the first embodiment and the adjustment operation of the aberration corrector and the operation of alignment are carried out through the procedures explained in the first embodiment. However, in the case of the critical-dimension-measurement SEM, automatic operation is dominant and hence the operation of the movable stop 31 is automated. In each of the processes shown in FIG. 3, the control computer 30: judges the orientation of the overlapping of images and the symmetry at angles of 90 degrees; commands the movable stop micromotion mechanism 32 to drive the air-pressuring mechanism, the pulse motor and others; and thus carries out the selection from among plural stops and the position micromotion. It is possible to detect a notch and a marking of a stop and control the position of the stop. That is, in the flowchart shown in FIG. 3, automated are the steps of: the insertion of a stop for chromatic aberration correction; the position adjustment of a stop for chromatic aberration correction; the selection and insertion of a stop for spherical aberration correction; and the switching from a multi-hole stop to a single-hole stop. Other adjustment steps, namely the primary setting of the voltage and electric current of a quadrupole power source, the setting of the amplitude of the applied voltage of the high voltage wobble and others, can be automated by setting a reference table or the like. In the present embodiment, it is preferable to use such a nine-hole stop as shown by the reference numeral 322 in FIG. 9. This is because, at the time of chromatic and spherical aberration correction, it is possible to carry out the correction in the directions at angles of 45 degrees in relation to the correction in the x and y directions with only one multi-hole stop. In other words, by using a nine-hole stop, it is possible to save the time spent for the change of a stop and shorten the time required for device adjustment. Since the critical-dimension-measurement SEM is used as an in-line measuring device, it is important for the device to increase the throughput as much as possible. Therefore, by shortening the time for the device adjustment, it is possible to increase the device operation time per day and improve the cost advantage for a device user. When a critical-dimension-measurement SEM is equipped with an aberration corrector, the chromatic and spherical aberrations of the electron lens system (the object lens in particular) are cancelled and hence the resolution of the obtained SEM image improves. Thereby it becomes possible to use a beam having a large aperture angle, which has heretofore been restricted because of the spherical and chromatic aberrations, for probe formation. However, the focal depth decreases inversely as shown by the following expression.DOF˜dp/2α (1),where, DOF represents a focal depth, dp a probe diameter, and α a beam aperture angle (semi-angle). In a semiconductor circuit which is a typical measuring object specimen of a critical-dimension-measurement SEM, the circuit pattern formed on a substrate is multi-layered and further the pattern itself has some degrees of jogs. Hence, a SEM image used for the dimension measurement of a pattern has to be focused at least on the layer to be subjected to the dimension measurement. Therefore, it is also necessary to cope with such a problem that the focal depth decreases when the chromatic and spherical aberrations are corrected. As a prior art aimed at the improvement of resolution and the increase of a focal depth in the field of electron beam application devices, JP-A 124205/2002 discloses a technology of using an orbicular zone light stop, thus cutting an electron beam in the vicinity of an optical axis, and preventing the focal depth from deteriorating. The reason why the focal depth is prevented from deteriorating by cutting the electron beam in the vicinity of an optical axis is briefly explained below. When the electron beams having passed through an orbicular zone light stop are converged in the vicinity of a focus with an object lens, any of the electron beams enters the vicinity of the focus at almost the same angle to the optical axis. If the wave nature of electrons is taken into consideration, by the interference of an electron wave group forming an identical angle with the optical axis, the probability of the existence of electrons on a plane group parallel with the optical axis increases. When the probability is integrated around the optical axis, the field where the probability of the existence of electrons increases extends backward and forward around the focal point in the center along the optical axis. In other words, the focal depth increases. In contrast, when the beam in the vicinity of the center is not cut, the probability of the existence of electrons on the plane group obliquely intersecting with the optical axis increases in the case of the interference between the center beam and the beams from the periphery of the stop. When the probability is integrated around the optical axis, the phenomenon wherein they are compensated with each other and thus the probability of the existence of electrons increases occurs only in the vicinity of the focus position and the focal depth becomes smaller than the case of the orbicular zone light. However, the document does not disclose the means for correcting the chromatic and spherical aberrations. The focus position of the center beam is different from that of the peripheral beams due to the spherical aberration and hence, when the stop is changed from an ordinary round single-hole stop to an orbicular zone aperture stop, focus deviation occurs and the focus position needs to be moved. Then, even though the focus position is moved, the chromatic aberration is not corrected, thus the spots are in the state of separation, and therefore the obtained image is an overlapped image similar to the one shown in FIG. 4A. Therefore, the application of the above prior art is limited. In view of the above situation, in the critical-dimension-measurement SEM of the present embodiment, the device is designed so that the device may have the function of being operable in both the long focal depth mode and the aberration correction mode and a device user may select either of the modes. FIG. 10 shows a flowchart of the operation of the device in the operation flow after spherical aberration correction. When the step of the end of the spherical aberration correction is terminated in the flowchart shown in FIG. 7, an icon for selecting either the long focal depth mode or the aberration correction mode is exhibited on the monitor 77 shown in FIG. 8 and a device user selects either of them. When a user selects the long focal depth mode, by the instruction of the control computer 30, the movable stop micromotion mechanism 32 moves the eight-hole aperture 323 shown in FIG. 11 (pseudo orbicular zone aperture since a shield cannot be placed on the optical axis without a support in the case of a charged particle beam) onto the optical path. Thereby, the focal depth increases by the multiple, which is defined by the expression below, of that in the case of a round hole stop.1/{1−(a′/a)2} (2),where, a represents the outer radius of a pseudo orbicular zone aperture, and a′ represents the inner radius of the pseudo orbicular zone aperture. In this case, since it is after the correction of the chromatic and spherical aberrations, as long as the center position of the stop is restored, the deviation of a focus and the shift of an image do not occur. By so doing, it is possible to lessen the decrease of the focal depth caused when an aberration corrector is used and also improve the operability and quality of the SEM image at focusing. In contrast, in the case of the aberration correction mode, a high resolution image can be obtained by aberration correction but the beam aperture angle increases and thus the focal depth decreases. In the case of the long focal depth mode, in the automated dimension-measurement image processing, as long as the focus is well converged in the vicinity of the position where the dimension measurement is carried out, the focus is also well converged at the position where the dimension measurement is carried out due to the long focal depth and the dimension-measurement image processing does not stop and rather advances. Since it is not necessary to manually adjust the focus again, the ratio of automated dimension measurement improves, and also the throughput improves. Meanwhile, when the close observation of the roughness and shape of a pattern is desired, a high resolution image can be obtained by adjusting the focus to the position to be observed with the aberration correction mode. Therefore, in the critical-dimension-measurement SEM, by allocating the long focal depth mode and the aberration correction mode to the automated dimension measurement and the image observation respectively, the advantage of obtaining high degrees of performances (high throughput and high image quality) of the device can be obtained. Examples of other orbicular zone apertures are shown as the reference numerals 324 and 325 in FIG. 11. In those cases, the branch portions are few and the opening ratio is large in comparison with the orbicular zone aperture of the reference numeral 323 and hence the probe electric current is reduced only by about {1−(a′/a)2} times that of a single round aperture having the radius a and it is possible to carry out dimension measurement of a high resolution long focal depth which less deteriorates the brightness. The present invention is applicable to a scanning electron microscope, a semiconductor inspector, a scanning transmission electron microscope, a focused ion beam device and the like. |
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061334987 | summary | TECHNICAL FIELD This invention relates to the use of chemically bonded phosphate ceramic (CBPC) waste forms for immobilizing large volumes of low-level, radioactive and/or hazardous waste, and, in particular, to an improved process and CBPC product. BACKGROUND OF INVENTION Low-level mixed wastes contain hazardous chemical and low-level radioactive materials. Of particular concern are low-level mixed waste streams that contain heavy metals, such as lead, cadmium, copper, zinc, nickel, and iron, among others, and waste streams from nuclear materials processing applications that contain technetium-99, chromium, and antimony. The U.S. Environmental Protection Agency classifies waste as hazardous, under the Resource Conservation and Recovery Act (RCRA), if excessive amounts of heavy metals leach from the waste during the Toxicity Characteristic Leaching Procedure (TCLP). Land disposal of leachable heavy metal waste is very expensive and strictly regulated, and therefore cost-effective, safe, leach resistant methods for encapsulating heavy metal waste is of current environmental importance. Stabilization of low-level mixed waste requires that the contaminants, including soluble heavy metals ions, are effectively immobilized. No single solidification and stabilization technology is known to successfully treat and dispose of low-level mixed waste, due to the physical and chemical diversity of the waste streams. Conventional high-temperature waste treatment methods (e.g., incineration, vitrification) are largely unsuitable for the treatment of low-level mixed waste streams, because their reliance on high temperature risks the release of volatile contaminants and they generate undesirable secondary waste streams. A low-temperature approach is to stabilize hazardous waste by using inorganic (e.g., pozzolanic) binders, such as cement, lime, kiln dust, and/or fly ash. Disadvantages of this approach include a high sensitivity to the presence of impurities, high porosity, and low waste loading volume. Organic binders (e.g., thermosetting polymers) are used even less frequently, because of cost and greater complexity of application. Organic binders are not compatible with water-based wastes, unless the waste is first pre-treated and converted to an emulsion or solid, and organic binders are subject to deterioration from environmental factors, including biological action and exposure to ultraviolet light. Recently, an alternative low-temperature approach has been developed at Argonne National Laboratory for stabilizing and solidifying low-level mixed waste by incorporating or loading the waste into a phosphate ceramic waste form. This technique immobilizes the waste by solidification, such that the waste is physically micro-encapsulated within the dense matrix of the phosphate ceramic waste form, and stabilizes the waste by converting the waste into their insoluble phosphate forms. Ceramic encapsulation systems are particularly attractive given that the bonds formed in these systems are ether ionic or covalent, and hence stronger than the hydration bonds in cement systems. Also, the ceramic formulation process is exothermic and economical. Phosphates are particularly good candidates for stabilization of radioactive and hazardous waste, because phosphates of radio nuclides and hazardous metals are essentially insoluble in groundwater. A salient feature of the low-temperature ceramic phosphate formulation process is an acid-base reaction. For example, magnesium phosphate ceramic waste forms have been produced by reacting magnesium oxide (MgO) with phosphoric acid to form a phosphate of magnesium oxide, Newberyite, as represented in Equation (1), below. EQU MgO+H.sub.3 PO.sub.4 +2H.sub.2 O.fwdarw.MgHPO.sub.4.3H.sub.2 O(1) The acid-base reaction results in the reaction of the waste components with the acid or acid-phosphates, leading to chemical stabilization of the waste. In addition, encapsulation of the waste in the phosphate ceramic results in physical containment of the waste components. The reaction represented by Equation (1) above occurs rapidly and generates heat, and upon evaporation of the water, a porous ceramic product results. U.S. Pat. No. 5,645,518 issued to Wagh, et al., incorporated herein by reference, describes in detail the process steps for setting liquid or solid waste in CBPC products using acid-base reactions. Accordingly, the process involves mixing ground solid waste, including salt waste spiked with heavy metals, with a starter powder of oxide and hydroxide powders of various elements; slowly adding the waste-powder mixture to an acid solution of phosphoric acid or soluble acid phosphates; thoroughly mixing the waste-powder-acid mixture for about a half hour to an hour at ambient temperatures (less than 100.degree. C.), such that the components of the mixture chemically react to form stable phases and a reacted viscous slurry or paste results; and allowing the slurry or paste to set for a few hours into the final CBPC product. Liquid waste is similarly stabilized by mixing the liquid waste with the acid solution (preferably 50:50), and then reacting the waste-acid mixture with the starting powders. The maximum temperature for the process is about 80.degree. C. The CBPC products attain full strength in about three weeks, and exhibit a complex structure, including a major crystalline phase, e.g., Newberyite (MgHPO.sub.4.3H.sub.2 O), and an insoluble, stable phase. The waste components are generally homogeneously distributed within the phosphate ceramic matrix. Unfortunately, however, the porous product (Newberyite) is unsuitable for waste treatment on a large scale. U.S. Pat. No. 5,830,815 issued to Wagh, et al., incorporated herein by reference, describes improving the CBPC fabrication process by incorporating two temperature control processes for both reducing heat generation during the encapsulation (reaction) steps and moderating pH conditions (some wastes are unstable at a low pH). The first temperature control process involves pre-treating the phosphoric acid with a carbonate, bicarbonate or hydroxide of a monovalent metal (e.g., K, Na, Li, Rb) prior to mixing with an oxide or hydroxide powder so as to buffer the acid. In particular, potassium containing alkali compounds (e.g., K.sub.2 O, KHCO.sub.3, KOH) result in a more crystalline waste form, and the higher the concentration of potassium in the potassium containing compound, the more crystalline the final product, resulting in a higher compression strength, lower porosity, and greater resistance to weathering, compressive forces, and leaching. The second temperature control process involves bypassing the use of the acid and mixing the oxide powder with dihydrogen phosphates of potassium, sodium, lithium, or other monovalent alkali metal, to form a ceramic at a higher pH. Neutralizing the phosphoric acid solution in equation (1) by adding potassium hydroxide (KOH), as represented in the chemical equation (2) below, reduces the reaction rate and heat generation, and results in the formation of a superior magnesium potassium phosphate (MKP) mineral product, MgKPO.sub.4.6H.sub.2 O (magnesium potassium phosphate hexahydrate), as represented in chemical equation (3) below. EQU H.sub.3 PO.sub.4 +KOH.fwdarw.KH.sub.2 PO.sub.4.H.sub.2 O (2) EQU MgO+KH.sub.3 PO.sub.4 +5H.sub.2 O.fwdarw.MgKPO.sub.4.6H.sub.2 O(3) The chemically bonded ceramic phosphate (CBPC) waste form (e.g, MgKPO.sub.4.6H.sub.2 O) is a dense, hard material with excellent durability and a high resistance to fire, chemicals, humidity, and weather. The low-temperature (e.g., room-temperature) process encapsulates chloride and nitrate salts, along with hazardous metals, in magnesium potassium phosphate (MKP) ceramics, with salt waste loadings of up to between approximately 70 weight percent and approximately 80 weight percent. This durable MKP ceramic product has been extensively developed and used in U.S. Department of Energy waste treatment projects. Phosphates in general are able to bind with hazardous metals in insoluble complexes over a relatively wide pH range and most metal hydroxides have a higher solubility than their corresponding phosphate forms. In addition to the magnesium and magnesium-potassium phosphate waste products discussed above, known waste encapsulating phosphate systems include, but not limited to, phosphates of magnesium-ammonium, magnesium-sodium, aluminum, calcium, iron, zinc, and zirconium (zirconium is preferred for cesium encapsulation). A non-exclusive summary of known phosphate systems and processing details is provided in Table I below, selected according to the ready availability of materials and low cost. It is also known to add other materials to either the waste or ceramic binder ingredients, such as fly ash. TABLE I ______________________________________ Phosphate Systems and Processing Details Curing System Starting Materials Solution Time ______________________________________ MKP Ground MgO, ground K Water 1 hour dihydrophosphate crystals Mg phosphate Calcined MgO Phosphoric acid- >8 days water (50/50) Mg-NH.sub.4 Crushed dibasic NH.sub.4 Water 21 days phosphate phosphate crystals mixed with calcined MgO Mg-Na phosphate Crushed dibasic Na Water 21 days phosphate crystals mixed with calcined MgO Al phosphate Al(OH).sub.3 powder Phosphoric acid Reacted (.apprxeq. 60.degree. C.) powder, pressed Zr phosphate Zr(OH).sub.4 Phosphoric acid 21 days ______________________________________ Iron oxides including either iron oxide (FeO) itself or magnetite (Fe.sub.3 O.sub.4) have also been used in the formation of phosphate ceramic products, however, these materials are uncommon and expensive. Haematite (Fe.sub.2 O.sub.3) is a very unreactive powder and efforts to form a chemically bonded phosphate ceramic (CBPC) product using haematite have been unsuccessful. When mixed with phosphoric acid, and even highly concentrated phosphoric acid, the haematite either does not react or reacts at such a slow rate that the reaction is impractical for the development of CBPC products. The slow rate of reaction is due to the insolubility of haematite, which is in its highest oxidation state. Appropriate oxide powders include, but are not limited to, oxides or hydroxides of aluminum, calcium, iron, magnesium, titanium, and zirconium, and combinations thereof. The oxide powders may be pre-treated (e.g., heated, calcined, washed) for better reactions with the acids. While no pressure is typically applied to the reacted paste, about 1,000 to 2,000 pounds per square inch may be applied when zirconium-based powders are used. The acid component may be dilute or concentrated phosphoric acid or acid phosphate solutions, such as dibasic or tribasic sodium, potassium, or aluminum phosphates, and the paste-setting reactions are controllable either by the addition of boric acid to reduce the reaction rate, or by adding powder to the acid while concomitantly controlling the temperature. Examples of appropriate phosphates include phosphates of aluminum, beryllium, calcium, iron, lanthanum, magnesium, magnesium-sodium, magnesium-potassium, yttrium, zinc, and zirconium, and combinations thereof. Salt waste may be reacted with phosphoric acid to consume any carbon dioxide (CO.sub.2) present, prior to mixing the salt waste with the oxide powders or binding powders, as the evolution of CO.sub.2 results in very porous final ceramic products. Unfortunately, the acid-base reactions involved in the phosphate ceramic systems described above occur very rapidly, resulting in the generation of considerable exothermic heat that prevents the formation of homogeneous large scale phosphate ceramic monoliths. The rapid setting of the CBPC products also hinders the proper conversion of hazardous or radioactive contaminants into stabilized phosphate forms. As a result, the CBPC products formed by methods known in the art have very poor density and strength. Encapsulation of waste containing heavy metals in known CBPC systems is also of limited practical use. Although heavy metals in the form of soluble nitrates (e.g., Cr(NO.sub.3).sub.3.9H.sub.2 O, Ni(NO.sub.3).sub.2.6H.sub.2 O, Pb(NO.sub.3).sub.2, and Cd(NO.sub.3).sub.2.4H.sub.2 O) are reportedly converted to insoluble phosphates by the CBPC forming chemical reactions, there is a critical need to improve their leach resistance and to provide greater stabilization for the metal anions of technetium-99, chromium, and antimony. Efforts to encapsulate heavy metal waste in phosphate ceramic products are further hampered by low maximum waste loading capacities, because of interference of the metal anions with ceramic-setting reactions, leaching of soluble metal anions from the resulting highly porous ceramic product (especially in aqueous environments), and rapid structural degradation of the ceramic product caused by the high leach rates. Also, environmental stresses degrade the integrity of known CBPC waste forms over time. For example, exposure to repeated cycles of wetting, drying and/or freezing, or acidic or other conditions conducive to leaching may affect the long term effectiveness of waste encapsulated CBPC waste forms. A need exists for improved phosphate ceramic systems and improved methods for disposing of wastes containing metal anions in phosphate ceramic products. The present invention is a surprisingly effective process step that significantly improves known phosphate ceramic formulations, enables the production of iron-based phosphate ceramic systems, and critically increases the stabilization of wastes containing heavy metals within CBPC composites. The invention involves adding oxidants or reductants to the ceramic phosphate formulations to retard or accelerate the acid-base reactions and thereby control the exothermic temperature of the reactions. In addition, the use of reducing agents significantly improves the stabilization of the metal anions within the phosphate ceramic composition, and thus the leach resistance of the encapsulated metals, by changing the valence of the metal to a lower oxidation state, such that the metal is more stable in the presence of the phosphate ions and/or the metal is more reactive with the phosphate ions. Therefore, in view of the above, a basic object of the present invention is to control the reactions rates and heat generation in phosphate ceramic processes to allow homogeneous large scale phosphate ceramic monoliths. Another object of the present invention is to significantly improve the density and strength of phosphate ceramic products formulated from methods known in the art. Another object of the present invention is to form chemically bonded phosphate ceramic products from inexpensive iron-based materials, such as haematite. Yet another object of the present invention is to provide an improved method for stabilizing waste containing metal anions in a phosphate ceramic composite having increased loading capacity and improved leach resistance. A further object of the invention is to provide an improved, safe, low temperature, economical method for stabilizing large volumes of waste containing metal anions in a durable, long term storage phosphate ceramic product. Additional objects, advantages, and novel features of the invention are set forth in the description below and/or will become apparent to those skilled in the art upon examination of the description below and/or by practice of the invention. The objects, advantages, and novel features of the invention may be realized and attained by means of instrumentation and combinations particularly pointed out in the appended claims. BRIEF SUMMARY OF THE INVENTION Briefly, this invention is a surprisingly effective method for significantly improving phosphate ceramic formulations and enabling the production of iron-based phosphate ceramic systems. The invention involves the addition of an oxidizing or reducing step to known phosphate ceramic formulations during the acid-base reactions between the oxide powders and phosphoric acid or acid phosphate solutions. The additives allow control of the rate of the acid-base reactions and concomitant heat generation. As a result, phosphate ceramic systems incorporating iron-based materials are practical, including the formation of iron phosphate ceramic products from haematite, a readily available, inexpensive material. The CBPC products may be crystalline ceramics and/or glass (non-crystalline). In an alternate embodiment, the addition of reducing agents to the ceramic phosphate system significantly improves the stabilization of heavy metal waste encapsulated within chemically bonded phosphate ceramic (CBPC) waste forms. Addition of the reducing agent, preferably a stannous salt, changes the valence of the metal to a lower oxidation state, such that the metal is more stable in the presence of the phosphate ions and/or the metal is more reactive with the phosphate ions. Importantly, the reduced metal ions are more stable and/or more reactive with the phosphate ions, resulting in the formation of insoluble metal species within the final phosphate ceramic matrix. |
abstract | A process is disclosed for the treatment of a solvent which has been used in nuclear fuel reprocessing or uranium ore purification, the solvent comprising an organophosphate ester and a hydrocarbon diluent. The process includes distilling the solvent under reduced pressure to remove substantially all the diluent and a major proportion of the organic ester, converting organophosphate to inorganic phosphate and encapsulating the residual material. |
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053902211 | claims | 1. In a boiling water reactor fuel bundle, a debris catcher construction for placement within the flow volume defined by an inlet plenum to the lower tie plate assembly to the upper fuel rod supporting grid comprising: first means including spatially axially separated strainer structures across said plenum for imparting a flow direction (momentum) change to passing water coolant, said first means not forming a continuum across said plenum such that at least one substantially unobstructed flow path is preserved through said first means; and second means defining a trapping structure for trapping debris separated from said water coolant by said flow direction change imparted by said first means. said means defining spatially separated structures across said plenum includes a deflecting cone. said means defining said spatially separated structures is located within a lower tie plate of a fuel bundle between an inlet orifice and said rod supporting grid. a plurality of upstanding side-by-side fuel rods; a lower tie plate including a rod supporting grid for supporting said plurality of upstanding side-by-side fuel rods; means for maintaining said fuel rods in upstanding side-by-side relation; a channel surrounding said plurality of upstanding fuel rods from the vicinity of said lower rod supporting grid along the length of said fuel rods to form a discrete flow path through said fuel bundle; an inlet plenum to the lower tie plate assembly to the upper fuel rod supporting grid; first means defining spatially separated strainer structures across said inlet plenum for imparting flow direction change to passing water coolant, said means defining said spatially separated strainer structure not forming a continuum across said plenum such that at least one substantially unobstructed flow path is preserved through said first means; and second means defining a trapping structure for the trapping separation of the debris separated from said water coolant by said flow direction change imparted by said first means. said inlet plenum includes a ring for trapping debris fastened to a portion of said plenum. said inlet plenum is located in said lower tie plate between an inlet orifice and said rod supporting grid. 2. The invention of claim 1 and wherein: 3. The invention of claim 1 and wherein: 4. In a boiling water reactor fuel bundle comprising in combination: 5. The invention of claim 4 and wherein said spatially separated strainer structures are placed in overlapping relationship across said plenum with axial spatial separation therebetween. 6. The invention of claim 5 and wherein said means defining trapping structures includes solid portions and perforate portions. 7. The invention of claim 4 and wherein: 8. The invention of claim 7 and wherein: |
description | The present invention relates generally to a system and method of disposing of nuclear waste. More particularly, the invention relates to disposing of nuclear waste temporarily or permanently in underground rock formations using multilateral boreholes. Numerous methods for disposing of nuclear waste are provided in the art. For example, an existing disposal method for nuclear waste is to bury the waste in shallow vaults also known as deep vertical wells. This method places the waste in vertical silos drilled into a mountain by a tunnel boring machine. The storage chambers are to be drilled approximately 1,000 feet into the mountain and can cost billions of dollars. Another method proposed for disposing of nuclear waste is burial of the waste in suitable canisters in mud in the bottom of the ocean. This method is dangerous as the canisters may rupture and pollute the ocean, killing life found in the surrounding area. A further proposal for disposing of nuclear waste is to place the waste into specially designed modules and launch the modules into space using the space shuttle. The modules will then be propelled into the sun for final incineration. This system would cost many billions of dollars and thus is not very practical. Another method proposed is to bury the nuclear waste in suitable canisters and placing such canisters within salt caverns below the surface of earth. This method is not suitable as the salt caverns are located at quite shallow depths and in case there is a leakage, the water table may get contaminated. It has also been proposed to bury the waste in near surface trenches or wells as used in landfills. This approach is not viable due to the great danger associated with disposing of the waste so close to the surface where leakage of the waste may do great harm to all life in the surrounding area. It has further been proposed to bury the waste in deep vertical wells which will be sealed with cement or mud. Burying the waste in the polar ice caps whereby the great masses of ice could enclose and isolate the radioactive material has also been proposed. U.S. Pat. Nos. 5,850,614 and 6,238,138 teach an application of horizontal wellbores to serve as repositories of nuclear waste in deep underground reservoirs. U.S. Pat. No. 5,863,283 teaches waste storage application in which nuclear waste filled liners are hung in the wellbores. The above described methods are all illustrative of prior art methods of nuclear waste disposal. While these methods may be suitable for the particular purpose to which they address, they would not be as suitable for the purposes of the present invention as heretofore described. The present invention is concerned with disposing of nuclear waste and, more specifically, to a method of disposing of nuclear waste in underground rock formations using multilateral horizontal boreholes. A primary object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations. Another object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which will provide prolonged safety from the nuclear waste and added protection to human health and the environment. An additional object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which will provide protection in case of rupturing or leaking of the canister in which the waste is stored. Another object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which will provide safe storage of the waste for at least 10,000 years. A further object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which is impervious to surface effects such as flooding, glaciation or seismic interference. A still further object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which will bury the waste in horizontally extending boreholes positioned well below the earth's surface. An even further object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations which will drill a primary vertical wellbore and secondary horizontal laterals extending therefrom. A yet further object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations wherein the secondary laterals will include an inner lining made from layers of steel and optionally a lead lining. A still further object of the present invention is to provide a method of disposing of nuclear waste in underground rock formations wherein front and end plugs will be placed within the secondary laterals for retaining canisters filled with waste. An additional object of the present invention is to provide a method of retrieving back the disposed nuclear waste stored in the horizontal laterals in underground rock formations. A method of disposing nuclear waste in underground rock formations is disclosed by the present invention. The method includes the steps of selecting an area of land having a rock formation positioned therebelow. The rock formation must be of a depth able to prevent radioactive material placed therein from reaching the surface and must be at least a predetermined distance from active water sources and drilling a vertical wellbore from the surface of the selected area which extends into the underground rock formation. A primary horizontal lateral is drilled from the vertical wellbore whereby the surface of the horizontal lateral is defined by the underground rock formation. A steel casing is placed within the horizontal lateral and cemented in place by circulating cement in the annular space between the steel casing and the wall of the wellbore. Nuclear waste to be stored within the lateral is placed in a canister and the encapsulated nuclear waste is positioned within the primary horizontal lateral. The primary horizontal lateral is then filled with cementitious material to seal the encapsulated nuclear waste therein. Additional primary horizontal laterals can be drilled from the vertical wellbore and secondary and tertiary horizontal laterals can be drilled from the primary horizontal lateral. Additional layers of lead, cement and steel may be used to cover the laterals and shield the rock formation from any radiation leakage. Furthermore, front and end plugs may be positioned at either end of the laterals, retaining the canisters therein and providing added protection from leakage of any solid, liquid or gaseous material. The foregoing and other objects, advantages and characterizing features will become apparent from the following description of certain illustrative embodiments of the invention. The novel features which are considered characteristic for the invention are set forth in the appended claims. The invention itself, however, both as to its construction and its method of operation, together with additional objects and advantages thereof, will be best understood from the following description of the specific embodiments when read and understood in connection with the accompanying drawings. Attention is called to the fact, however, that the drawings are illustrative only, and that changes may be made in the specific construction illustrated and described within the scope of the appended claims. Description of ElementsReference numeralsDrilling rig10Earth's surface12vertical wellbore14surface layers16cap rock layer18primary lateral20angle between primary laterals22secondary laterals24tertiary laterals26horizontal plane28protective zone29first layer of cement30first steel casing32second layer of cement34lead liner36canister38centralizers40far end of lateral42front end of lateral44windows46cement filler48front plug49end plug50layer of lead52second steel casing54liner support56sandwiched layer of lead58first layer of steel60second layer of steel62third layer of steel64connector73retrieving tool74drill pipe76sealing material75plugging device77 The present invention can be more fully understood by reading the following detailed description of some of the embodiments, with reference made to the accompanying drawings. FIG. 1 shows a preferred embodiment of the equipment used and the results obtained when practicing the method of the present invention. A drilling rig 10 is positioned on an isolated surface 12 of the earth and is used to create a vertical wellbore 14 which will extend vertically into the earth's surface. The vertical distance is contemplated to be in the range of 5,000 feet to 25,000 feet. The vertical wellbore 14 extends through a plurality of layers of the earth's surface 16 and into a layer of rock 18 herein called the repository. The repository layer of rock 18 is a specially selected rock formation deep enough below the earth's surface to prevent radiation which may leak from reaching the surface. The selected rock formations have existed for billions of years as is evidenced by the chronological fossil history found in the rock strata. Branching off and extending horizontally from the vertical wellbore 14 at a depth below the earth's surface occupied by the layer of cap rock 18 are primary laterals 20. The primary laterals 20 may be at different depths or at the same depth and extending at an angle 22 from one another. Any number of primary laterals 20 may be drilled from the vertical wellbore, two primary laterals are shown in FIG. 1 for purposes of example only. In one embodiment of the invention there are extending from the primary laterals 20 and along the same horizontal plane 28 are secondary laterals 24 and extending from the secondary laterals 24 and also along the same horizontal plane 28 are tertiary laterals 26. The primary, secondary and tertiary laterals 20, 24, and 26 respectively of a single branch extending from the vertical wellbore 14 all extend in the same horizontal plane 28 while each branch may extend in different horizontal planes as shown in FIG. 1. The formation of cap rock 18 should enclose the primary, secondary and tertiary laterals 20, 24 and 26 on all surfaces to thereby define the dimensions of the laterals and ensure isolation for an indefinite period. FIG. 2 shows a drilling rig 10 which is well known in art and is similar to those used in oil drilling and exploration to reach oil deposits located deep beneath the earth's surface. While a preferred structure for the drilling rig 10 is shown and described herein, those of ordinary skill in the art who have read this description will appreciate that there are numerous other structures for the drilling rig 10 and, therefore, should be construed as including all such structures as long as they achieve the desired result of creating a primary wellbore extending a predetermined distance below a surface of the earth, and therefore, that all such alternative mechanisms are to be considered as equivalent to the one described herein. FIG. 3 illustrates a single branch extending from the vertical wellbore 14. Extending vertically through the repository rock 18 is the vertical wellbore 14. A primary lateral 20 branches out horizontally from the vertical wellbore 14 along the horizontal plane 28 and a plurality of secondary laterals 24 extend from the primary lateral 20 in the horizontal plane 28. A plurality of tertiary laterals 26 extends from the secondary laterals 24 and in the horizontal plane 28. Any number of secondary laterals 24 can extend from each primary lateral 20 and any number of tertiary laterals 26 can extend from each secondary lateral 24. The number of secondary and tertiary laterals 24, 26 are for purposes of description only and not meant to be limiting. The only requirement on the positioning of the secondary and tertiary laterals 24 and 26 is that they cannot overlap one another. Overlapping of the laterals causes communication there-between and will act to reduce the effectiveness of the structure. The lateral wellbores 20, 24, 26 in the embodiments are contemplated to be in the range of 500 feet to 40,000 feet. A distance easily implemented with current drilling technologies. The lateral wellbores will range in size from 4 inch diameter to 20 inch diameter. FIG. 4 illustrates a preferred construction of the tertiary lateral 26 within the circle labeled 4 of FIG. 3 in greater detail, the construction of the primary and secondary laterals 20, 24 respectively, are identical thereto. The tertiary lateral 26 is comprised of a protective zone 29 therein to prevent the leakage of the nuclear waste from the tertiary lateral 26. As illustrated in FIG. 4 the protective zone 29 comprises a first layer of cement 30 within the lateral 26 which forms the first outer layer. A second outer layer is of a steel casing 32 and is sealed within the first outer casing of cement 30. The first layer of cement is thus formed by circulating cement between the first steel casing 32 and the walls of tertiary lateral 26. Further, a second layer of cement 34 and a layer of lead 36 is provided within the steel casing of the tertiary lateral 26. Nuclear waste is placed and secured within a radioactive capsule or canister 38. The radioactive canister 38 is well known in the art and presently used for securing nuclear waste. Any known method for securing nuclear waste in a container or capsule for placement in a lateral as produced by the present method may be used. The nuclear waste includes the potentially hazardous to the environment waste including nuclear, chemical, warfare waste, biomedical waste. In the present embodiment, the nuclear waste canisters 38 are mechanically modified with a connector 73 as shown in FIG. 4. The canister connector 73 has a disengagable mechanism that allows the canisters to be taken hold of and lifted or pulled back to the surface. This process is well known in the oil and gas industry and is customarily referred to as “fishing” and it is a well-defined discipline in oil-field well service work. The connector 73 provides a means for attaching the canisters 38 to a retrieving tool 74 and helps in retrieving the canisters 38 when needed. A disengagable connector 73 is defined as an adapter than can be engaged or connected and subsequently disengaged or disconnected. The canister 38 is positioned in the laterals 20, 24, 26 and may be held in a fixed position within the laterals 20, 24, 26 by a plurality of centralizers 40. The sequence of layers coating the laterals 20, 24, 26 act to protect the rock formation 18 in which the laterals 20, 24, 26 extend from leakage of any nuclear waste. By storing the heavy nuclear waste in horizontal laterals 20,24,26 the weight and pressure exerted in the wellbores by the dense waste is very small compared to the weight and pressures developed in a long vertical well where thousands of feet of waste is stored vertically. The total weight component is based on a few inches, the diameter of the lateral, whereas in the prior art using vertical nuclear storage, the pressures are based on several thousand feet of dense material. FIG. 5 illustrates a partial view of a nuclear waste storage network including a wellbore 14 and primary and secondary laterals 20, 24, respectively, extending therefrom. In order to produce a primary lateral 20, a window 46 must be cut into the vertical wellbore 14 at the point from which the primary lateral 20 is to extend. The window cutting process is well known in the drilling industry. The primary lateral 20 is then drilled through the window 46 and extending horizontally into the rock formation 18. The same is true for producing the secondary and tertiary laterals 24, 26. A window 46 must be cut into the lateral at the point from which the dependent lateral will extend. The dependent lateral will then be drilled through the window 46 and into the rock formation 18 in the identical plane in which the primary lateral lies. FIG. 6 depicts the cross sectional view of the three secondary laterals along the line 6-6 of the FIG. 1. The figure shows a typical location in the lateral upstream of the location of the stored canisters 38 where cement material 48 fills the wellbore and is surrounded by the steel casing 32 and the first layer of cement 30. In order to provide additional protection from leaking nuclear waste, a front plug 49 and an end plug 50 may be positioned within the lateral as is illustrated in FIG. 7. The front plug 49 is positioned adjacent the window 46 at the point at which the lateral 20, 24, 26 branches and the end plug 50 is positioned at an end 42 of the lateral 20, 24, 26 opposite the front plug 49. The end plug 50 is inserted into the lateral 20, 24, 26 prior to placement of the canisters 38 and the entry into and exit from the lateral of any liquid, solid or gaseous material thereby providing additional safety from leakage of nuclear waste into the host rock formation 18. These plugs 49, 50 are known and preferably similar to oil field “packers” used to cover the vertical wellbores and prevent oil from exiting the well. However, these plugs 49, 50 may be in any other form which achieves the necessary purpose of providing additional protection from leakage of nuclear waste from the lateral 20, 24, 26. Other embodiments for the protective zone of the laterals are also possible. One such embodiment is illustrated in FIG. 8 wherein a first steel casing 32 is then deployed within the lateral. A first layer of cement 30 is circulated around the steel casing 32 of the laterals 20, 24 or 26. Further a second steel casing 54 is deployed within the first steel casing 32 and a second layer of cement 34 is circulated between the second steel casing 54 and the first steel casing 32. Further, a layer of lead 52 is placed within the second steel casing 54, said layer of lead 52 acting as a liner is separated from the second steel casing 54 by a plurality of liner supports 56 placed between the second steel casing 54 and the layer of lead 52. This lead liner acts as a radiation shield. The liner supports 56 extend only to the entry point of the lateral 20, 24, 26 i.e. the position at which the window 46 is cut. FIG. 9 illustrates another embodiment of the protective zone wherein the second steel casing 54 is replaced with a three tired structure. The three tiered structure includes a layer of lead 58 sandwiched between layers of steel 60, 62. As in the embodiment illustrated in FIG. 9, the lead layer 58 only extends to the entry point of the lateral 20, 24, 26. A third layer of steel 64 extends between the sandwiching layers of steel 60, 62 from the entry point of the lateral 20, 24, 26 to the top of the vertical wellbore 14. These additional layers 58, 60, 62 and 64 also provide added protection from radiation which may leak from the canisters, preventing the radiation from leaving the lateral 20, 24, 26 and entering the host rock formation 18. FIG. 10 shows an embodiment wherein the canisters 38 are positioned within the lateral and a plugging device 77 is installed. This is generally called a bridge plug in the industry and this type of plug is utilized as the primary mechanism in plugging the millions of depleted oil and gas wells around the world. These plugs can vary in properties, and in their emplacement methods but in general they form a permanent mechanical seal in the wellbore such that material below the plug does not migrate above the plug. Therefore, the plugging device 77 is installed in the lateral containing canisters in order to keep the canisters at its own place and prevent movement of the canisters. Further, a plurality of plugging device 77 can be used for multiple safety barriers in sequence. The lateral 20, 24, 26 above the plugging device 77 is now filled with a sealing material 75. There are many other types of sealing material available in market for protection of the nuclear waste canisters 38. However, in the present embodiment the sealing material 75 is a special blend of cement designed to provide leak protection, safety and durability. A detachable or disengagable connector 73 is connected to the canisters 38. For placing the canister 38 in the lateral 20, 24, 26, the canister 38 is attached to the drill pipe 76 and inserted in to the lateral 20, 24, 26 from the surface of earth. Once deployed within the lateral, the drill pipe 76 is detached from the connector 73 of the canister 38 and is pulled back to the surface, thus depositing the canister 38 with the connector 73 within the lateral 20, 24, 26. The drill pipe 76 may then be used to place additional canisters 38 within the laterals until either the laterals are filled or all the canisters are stored. Further the plugging device 77 and the sealing material 75 is placed within the lateral 20, 24, 26 to avoid movement of nuclear waste from the lateral in case of an inadvertent leakage of the nuclear waste from the canister 38. FIG. 11 illustrates the components necessary for retrieving the canister 38 back to the surface from the laterals 20, 24, 26. As shown in the FIG. the retrieving tool 74 comprises a special drilling bit for drilling the sealing material 75 and a fishing tool for connecting to the connector 73 of the canister 38. For retrieving the canister 38 back to the surface the retrieving tool 74 is inserted in the lateral 20, 24, 26 by the drill pipe 76. The sealing material 75 is drilled out as shown in the FIG. with a special drill bit supported by the pipe 76 which allows the sealing material to be removed without damaging the canister 38. Once the sealing material is removed, the plugging device 77 is also removed and the fishing tool of the retrieving tool 74 is attached to the connector 73 of the canister 38. The fishing tool fishes the canister 38 back to the surface of the earth by drill pipe 76. The engagement of the fishing tool of the retrieving tool with the connector of the canister 38 is illustrated in FIG. 12. In operation, an isolated area is selected for placement of the wellbore 14 and laterals 20, 24 and 26. The area must include a rock formation 18 therebelow and at a depth great enough to prevent any nuclear waste which may leak from reaching the surface. The rock formation 18 must also be a predetermined safe distance from any underground active water sources. Upon selection of an appropriate area, a drilling rig 10 such as is used to drill oil wells is used to create a vertical wellbore 14 which extends into the selected rock formation 18. A window 46 is then cut into the vertical wellbore 14 at a depth occupied by the rock formation 18 and at each position from which a primary lateral 20 is desired to extend. A primary lateral 20 is then drilled into the rock formation 18 extending from each window 46 to form each primary lateral 20. The primary laterals 20 may be at differing depths below the surface from one another as long as they extend more or less horizontally, i.e. perpendicular to the vertical wellbore 14, and have dimensions, i.e. sides, defined by the rock formations 18. Windows 46 are then cut into each primary lateral 20 at each position from which a secondary lateral 24 is desired to extend. The secondary laterals 24 are each then drilled to extend from their respective window 46 and each extend horizontally through the rock formation 18 in the same plane as the primary lateral 20 from which they depend. Windows 46 are then cut into each secondary lateral 20 at each position from which a tertiary lateral 24 is desired to extend. The tertiary laterals 24 are each then drilled to extend from their respective window 46 and each extend horizontally through the rock formation 18 in the same plane as the primary and secondary laterals 20, 24 from which they depend. A first steel casing 32 is installed in each lateral 20, 24, 26 and is cemented in place by circulating the cement to form the cement layer 30 which forms the first outer layer. A second outer layer is of a steel casing 54 and is sealed within the first outer casing of cement 30. The first layer of cement is thus formed by circulating cement between the first steel casing 32 and the walls of tertiary lateral 20, 24, 26. Further, a second layer of cement 34 and a layer of lead 36 is provided within the steel casing of the tertiary lateral 20, 24, 26. In order to provide added protection from radiation, which may leak within the laterals 20, 24, 26, the second inner layer of lead may be replaced by alternate constructions. In an alternate embodiment of the present invention, a first steel casing 32 is installed in the laterals 20, 24, 26. Then a first layer of cement 30 is circulated in the laterals 20, 24 or 26. Further a second steel casing 54 is placed within the first steel casing 32 and a second layer of cement 34 is formed by circulating cement between the second steel casing 54 and the first steel casing 32. Further, a layer of lead 52 is placed within the second steel casing 54, said layer of lead 52 is separated from the second steel casing 54 by a plurality of liner supports 56 placed between the second steel casing 54 and the layer of lead 52. This liner material acts as a radiation shield. The liner support 56 extends only to the entry point of the lateral 20, 24 or 26, i.e. the position at which the window 46 is cut. A second alternate construction for the second inner layer 36 is also formed of a three-tiered structure. In this construction, a first layer of steel 60 is positioned within the first inner layer of cement 34. A layer of lead 58 is then positioned within the first inner layer of steel 60 and a second layer of steel 62 is positioned within the layer of lead 58 acting to sandwich the layer of lead 58 between the first and second layers of steel 60, 62. In an alternate construction, the layer of lead 58 only extends to the entry point of the lateral. The first and second layers of steel 60, 62 are positioned to cover the entire surface of the lateral in which they are placed and extend through each lateral from which it depends and the vertical wellbore 14. A third layer of steel 64 is positioned between the first and second layers of steel 60, 62 and extends between the sandwiching layers of steel 60, 62 from the entry point of the lateral to the top of the vertical wellbore 14. Portions of the third steel layer 64 may be replaced by a layer of lead 58 within the depending laterals which will house canisters 38 containing nuclear waste. These additional layers 58, 60, 62 and 64 provide added protection from radiation, which may leak from the canisters, preventing the radiation from leaving the lateral and entering the host rock formation 18. An end plug may then be inserted into each of the lateral 20, 24 or 26 in which it is desired to store canisters 38 containing nuclear waste. The laterals are now prepared for storing the canisters containing nuclear waste. A plurality of centralizers 40 may be connected to the canisters 38 to hold the canisters 38 stationary within the lateral in which they are stored. The canister 38 is modified with a detachable connector 73. The connector 73 is attached to the drill pipe 76 and is then directed through the vertical wellbore 14 and through the network of laterals until it reaches its final destination for storage. The connector 73 is then separated from the drill pipe 76 and is removed from the network through the laterals and the vertical wellbore 14 and up to the surface 12 of the selected area by hoisting up the tubular string 72. The tubular string 72 is then used to position another canister 38 within the network of laterals 20, 24 or 26. This process is repeated until the network is full or all the canisters 38 are positioned within the network. The plugging device 77 may then be positioned at the entry point of the lateral 20, 24 or 26, i.e. at the point at which the windows 46 are cut, to seal each lateral and prevent any solid, liquid or gaseous material from escaping from the sealed lateral. The network is further filled with sealing material 75 to seal the canisters 38 in place within their respective lateral 20, 24 or 26 and also act to prevent any nuclear waste which may leak from reaching either the rock formation 18 housing the laterals or the surface of the selected area. Alternatively, a front plug 49 may be placed within the lateral 20, 24 or 26 and an end plug 50 be placed at the terminating end of the lateral 20, 24 or 26 for sealing the canisters 38 within the lateral 20, 24 or 26 to prevent leakage of the nuclear waste. From the above description, it is evident that the present invention provides a method of disposing of nuclear waste in underground rock formations and provides prolonged safety from the nuclear waste and added protection to human health and the environment. This method also provides protection in case of rupturing or leaking of the canister in which the waste is stored and safe storage of the waste for at least 10,000 years. It also provides storage of nuclear waste, which is impervious to surface effects such as flooding, glaciation or seismic interference. The laterals in which the waste is stored include an inner lining made from layers of cement, steel and lead and possibly also include front and end plugs to provide the above benefits. It will be understood that each of the elements described above, or two or more together, may also find a useful application in other types of applications differing from the type described above. While the invention has been illustrated and described as shown in the drawings, it is not intended to be limited to the details shown, since it will be understood that various omissions, modifications, substitutions and changes in the forms and details of the formulation illustrated and in its operation can be made by those skilled in the art without departing in any way from the spirit of the present invention. Without further analysis, the foregoing will so fully reveal the gist of the present invention that others can, by applying current knowledge, readily adapt it for various applications without omitting features that, from the standpoint of prior art, fairly constitute essential characteristics of this invention. |
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claims | 1. A method of making a silicon-based semiconductor material comprising:introducing a master alloy comprising Si and Si31 radioisotope into a base Si material to form a transition material comprising atoms of the Si31 radioisotope within a Si crystal lattice structure; andaging the transition material to convert at least a portion of the Si31 radioisotope atoms to P31 atoms retained in the Si crystal lattice structure to thereby form a silicon-based material doped with P31. 2. The method of claim 1, wherein the amount of P31 in the silicon-based semiconductor material is above an equilibrium solubility limit of phosphorous in silicon as determined by a standard Si—P phase diagram. 3. The method of claim 2, wherein the equilibrium solubility limit is a room temperature solubility limit. 4. The method of claim 1, wherein the amount of P31 retained in the Si crystal lattice structure is from 0.0001 to 40 atomic percent of the Si crystal lattice structure. 5. The method of claim 1, wherein the amount of P31 retained in the Si crystal lattice structure is at least 0.001 atomic percent of the Si crystal lattice structure. 6. The method of claim 1, wherein the amount of P31 retained in the Si crystal lattice structure is from 0.01 to 30 atomic percent of the Si crystal lattice structure. 7. The method of claim 1, wherein the master alloy is introduced into molten base Si material. 8. The method of claim 7, wherein the master alloy is introduced at a temperature above room temperature into the molten base Si material. 9. The method of claim 1, wherein the master alloy comprises from 0.005 to 50 atomic percent Si31 radioisotope. 10. The method of claim 1, wherein the master alloy comprises from 0.05 to 40 atomic percent Si31 radioisotope. 11. The method of claim 1, wherein the Si crystal lattice structure is face-centered-cubic, and the P31 atoms randomly occupy sites within the face-centered-cubic structure. 12. The method of claim 1, wherein the silicon-based semiconductor material has a conductivity of from 105 to 1017 per ohms·m. |
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claims | 1. A method for storing nuclear fuel in a storage container including a concrete body and a fuel receiver embedded in the concrete body, comprising the steps of introducing the nuclear fuel into the fuel receiver, providing formwork for the concrete body and mounting the fuel receiver within the formwork, placing the formwork in an immersed position in a pool containing a body of water, placing concrete in the immersed formwork to form the concrete body after said mounting of the fuel receiver within the framework step and after said introducing of the nuclear fuel in the fuel receiver step, and removing the formwork with the concrete body formed therein from the pool. 2. A method according to claim 1 in which the nuclear fuel is introduced into the fuel receiver after the formwork has been placed in the immersed position in the pool. claim 1 3. A method according to claim 2 , in which the fuel is transferred to the fuel receiver from an underwater position in an adjacent pool or pool section. claim 2 4. A method according to claim 2 in which subsequent to introduction of the nuclear fuel into the fuel receiver and sealing of the fuel receiver the formwork is transferred while in an immersed position to an adjacent pool or pool section in which placement of the concrete in the formwork is effected. claim 2 5. A method according to claim 3 , in which the nuclear fuel is placed in the adjacent pool or pool section while accommodated in a shipping container and in which the shipping container is placed in an immersed position in that pool or pool section. claim 3 6. A method according to claim 1 in which the fuel is introduced into the fuel receiver before the fuel receiver is introduced into the formwork and in which the fuel receiver with the fuel introduced into it is placed in the formwork after the formwork has been placed in the immersed position in the pool. claim 1 7. A method according to claim 6 in which the fuel receiver with the fuel introduced into it is transferred while in an immersed position from a pool or pool section containing a body of water to the immersed formwork. claim 6 8. A method according to claim 1 in which the fuel receiver is jointlessly embedded in the concrete. claim 1 9. A method according to claim 1 in which the concrete body is cast in the shape of a substantially straight upstanding cylinder. claim 1 10. A method according to claim 9 in which the concrete body is formed with a central, axially through passage and the fuel receiver is provided in the concrete body as a number of individually sealable receiver sections distributed about the central passage and in which the nuclear fuel is distributed to the receiver sections during its introduction into the fuel receiver. claim 9 11. A method according to claim 1 in which the formwork is assembled as permanent formwork from at least the following components: a lower end cover, an upper end cover and a cylindrical outer wall which is joined with the end covers, and in which a reinforcement is mounted in the form work and anchored in the end covers. claim 1 12. A method according to claim 11 in which the reinforcement is provided as two groups of reinforcing members which extend helically along two imaginary cylindrical surfaces inwardly of and close to the inner side of the outer formwork wall, the reinforcing members of each group being uniformly spaced apart circumferentially and are of the same hand, whereas the hand of the reinforcing members of the other group is opposite to the hand of the reinforcement members of the first group. claim 11 13. A method according to claim 1 in which the concrete is placed in the formwork through at least one vertical placing tube the mouth of which is positioned adjacent the lowest part of the formwork cavity when the placing commences and raised as the placing proceeds, such that it is constantly slightly below the surface of the placed concrete. claim 1 14. A method as claimed in claim 12 in which the reinforcing members are stressed at least to some degree after the concrete has hardened partially but not completely. claim 12 15. A method according to claim 1 in which the formwork is assembled and provided with the fuel receiver outside the pool in which the nuclear fuel is introduced into the fuel receiver. claim 1 |
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051577003 | claims | 1. An exposure apparatus usable with synchrotron radiation generating means for injecting electrons into a ring to produce synchrotron radiation, said apparatus comprising: exposure means for exposing a wafer, through a mask, to the synchrotron radiation introduced from the synchrotron radiation generating means through a window; intensity detecting means for detecting an intensity distribution of the synchrotron radiation; and control means for controlling said exposure means on the basis of the intensity distribution provided by said detecting means immediately after injection of the electrons into the ring. detecting means for detecting an intensity of X-rays after injection of electrons to generate the synchrotron radiation; and means for calculating an X-ray intensity during an exposure operation on the basis of a result of detection by said detecting means and on an attenuation curve of the injected electrons. a shutter for controlling exposure; a sensor for detecting an intensity of exposure radiation at a portion of an exposure area; a controller for controlling said shutter; and a signal processor for calculating a cumulative amount of exposure on the basis of the intensity of the exposure radiation detected by said sensor at the exposure area portion upon a preliminary actuation of said shutter by said controller, and for supplying an output to said controller, wherein said controller effects control of said shutter in accordance with the output supplied from said signal processor, to control an actual exposure operation. a shutter for controlling exposure; a sensor for detecting an intensity of exposure radiation at a position of an exposure area; a shutter driver for opening and closing said shutter; and calculating means for receiving outputs of said sensor when said shutter is opened and when said shutter is closed, in synchronism with opening and closing of said shutter by said shutter driver, and for processing the outputs to calculate an intensity of the exposure radiation, wherein said calculating means calculates the intensity of the exposure radiation by storing and processing information from said semiconductor sensor immediately before and after actuation of said shutter by said shutter driver. first detecting means for detecting an intensity of X-ray rays, corresponding to an exposure area; and second detecting means, disposed outside of the exposure area, for detecting attenuation of the synchrotron radiation. a shutter for controlling an exposure period; a semiconductor sensor for detecting an intensity of exposure radiation; a wafer stage; a wafer stage driver for driving said wafer stage; a shutter driver for opening and closing said shutter; and calculating means for receiving an output of said semiconductor sensor in synchronism with opening and closing of said shutter and for processing the output to calculate an intensity of the exposure radiation, wherein said wafer stage, upon a detection of exposure radiation intensity by said sensor, retracts the wafer from exposure radiation, so that the wafer is not exposed to the exposure radiation during a detection. a shutter for controlling an exposure period; a semiconductor sensor for detecting an intensity of exposure radiation; a wafer stage; a wafer stage driver for driving said wafer stage; a shutter driver for opening and closing said shutter; calculating means for receiving an output of said semiconductor sensor in synchronism with opening and closing of said shutter by said shutter driver, and for processing the output to calculate an intensity of the exposure radiation; and an auxiliary shutter for preventing the wafer from being exposed to the radiation during detection of the exposure radiation intensity by said sensor. a stage for supporting a member to be exposed; optical control means for controlling radiation projected from the exposure radiation source to the member; optical detecting means for detecting one of an intensity of the radiation of the radiation source and the intensity of the radiation on a surface of the member to be exposed; driving profile determining means, coupled with said optical control means and said optical detecting means, for determining a driving profile of said optical control means based on the intensity on the surface as a function of position of the surface and based on the intensity of the radiation source, to provide a uniform amount of exposure on the surface; and driving profile compensating means for one of expanding and contracting a time axis of the driving profile in accordance with a change in the intensity of the radiation source. a stage for supporting a member to be exposed; exposure control means, disposed in a radiation path between the radiation source and said stage, including a movable member for selectively limiting the radiation from said radiation source to the member; determining means for determining a driving speed for providing a constant amount of exposure of the member to radiation, despite any attenuation, with time, of an intensity of the radiation to the member; and driving means for driving said movable member of said exposure control means in accordance with the driving speed determined by said determining means. a stage for supporting a member to be exposed; exposure control means, disposed in a radiation path between the radiation source and said stage, including a movable member for selectively limiting the radiation from said radiation source to the member; determining means for determining a driving speed for providing a constant amount of exposure of the member to radiation, despite any attenuation, with time, of an intensity of the radiation to the member; and driving means for driving said stage in accordance with the driving speed determined by said determining means. injecting electrons into a ring to generate a synchrotron radiation beam; detecting an intensity distribution of the synchrotron radiation beam immediately after injection of the electrons in said injecting step; controlling exposure of a substrate to the synchrotron radiation beam on the basis of the detected intensity distribution. injecting electrons into a ring to generate a synchrotron radiation beam; detecting an intensity distribution of x-rays in the synchrotron radiation beam after injection of the electrons in said injection step; determining an intensity of the x-rays during a predetermined period on the basis of the detection in said detecting step and an attenuation curve of the injected electrons; and controlling exposure of a substrate to the synchrotron radiation beam on the basis of the determination. opening a shutter for preliminary exposure of an exposure region; detecting exposure radiation through the shutter by a sensor located at a position in the exposure region when the shutter is preliminarily opened in said opening step; calculating an exposure amount of the position on the basis of the detection, and providing an information signal relating to the exposure amount; exposing a substrate to the exposure radiation, while the shutter is controlled on the basis of the information signal, to expose the substrate to the radiation for semiconductor manufacturing. providing a sensor for detecting exposure radiation; operating a shutter for controlling exposure with the exposure radiation; detecting, immediately before and after the operation of the shutter an output of the sensor when the shutter is closed and an output of the sensor when the shutter is opened; determining a value related to an intensity of the exposure radiation on the basis of the outputs detected in said detecting step. providing an exposure control mechanism for controlling exposure of a substrate to exposure radiation, and operating the exposure control mechanism on the basis of a driving profile; performing one of expanding and contracting a time base of the driving profile in accordance with a change in an intensity property of the exposure radiation determined on the basis of detecting the exposure radiation to determine a corrected driving profile; and operating the exposure control mechanism in accordance with the corrected profile to expose the substrate to the exposure radiation. providing an exposure control mechanism for controlling exposure of a substrate to exposure radiation, and driving a movable member at a predetermined speed by the exposure control mechanism; determining the driving speed of the movable member in accordance with attenuation of intensity of the exposure radiation with time; and moving the movable member at the predetermined driving speed, while exposing the substrate. providing an exposure control mechanism for controlling exposure of a substrate to exposure radiation, and driving a stage carrying the substrate at a predetermined speed by the exposure control mechanism; determining the driving speed of the stage in accordance with attenuation of intensity of the exposure radiation with time; and moving the stage at the predetermined driving speed, while exposing the substrate. obtaining information, related to intensity of x-rays contained in a synchrotron radiation beam supplied to an exposure region, by a first detector disposed in the exposure region; retracting the first detector outside the exposure region; detecting attenuation of the synchrotron radiation beam with a second sensor disposed outside the exposure region; and controlling exposure of the exposure region on the basis of outputs of the first and second sensors. 2. An apparatus according to claim 1, wherein said control means comprises means for controlling said exposure means on the basis of the intensity distribution of the synchrotron radiation immediately after one of mounting and exchanging the window. 3. An apparatus according to claim 1, wherein said intensity detecting means comprises a wafer carrying stage for carrying the wafer and further comprising an X-ray detector mounted thereon. 4. An apparatus according to claim 1, wherein said exposure means comprises a shutter for controlling the amount of radiation introduced, and wherein said control means controls the shutter in accordance with the intensity distribution detected by said intensity detecting means. 5. An X-ray exposure apparatus using synchrotron radiation, said apparatus comprising: 6. An exposure apparatus, comprising: 7. An apparatus according to claim 6, further comprising means for moving said sensor in the exposure area, and wherein said signal processor calculates an amount of exposure from a plurality of exposure area portions. 8. An apparatus according to claim 6, wherein said controller comprises means for correcting control of said shutter on the basis of the cumulative amount of exposure fed back thereto from said signal processor. 9. An exposure apparatus, comprising: 10. An X-ray exposure apparatus for transferring by synchrotron radiation a pattern of a mask onto a member coated with a sensitive resist, said apparatus comprising: 11. An exposure apparatus comprising: 12. An exposure apparatus comprising: 13. An exposure apparatus usable with an exposure radiation source, said apparatus comprising: 14. An apparatus according to claim 13, wherein the radiation source is an X-ray source, and said optical control means comprises a movable aperture. 15. An apparatus according to claim 13, wherein the radiation source is a synchrotron radiation source, and said optical control means comprises an actuator for swinging a mirror. 16. An exposure apparatus usable with a radiation source, said apparatus comprising: 17. An apparatus according to claim 16, wherein said exposure control means comprises a movable aperture stop. 18. An apparatus according to claim 16, wherein said exposure control means comprises a movable mirror. 19. An exposure apparatus usable with a radiation source, said apparatus comprising: 20. An exposure method for manufacturing semiconductor devices, comprising: 21. A method according to 20, further comprising providing a window, which is substantially transparent to the synchrotron radiation beam, between the ring and the substrate, and effecting said detecting step with respect to the beam passed through the window. 22. A method according to claim 21, further comprising effecting detecting step immediately after the window is one of mounted and replaced. 23. A method according to claim 21, further comprising effecting said detecting step using a detector mounted on a stage for carrying the substrate. 24. A method according to claim 23, wherein the detector includes a x-ray detector, and further comprising detecting the x-rays contained in the synchrotron radiation beam with the x-ray detector. 25. A method according to claim 21, further comprising using a shutter in said exposure controlling step. 26. An exposure method for manufacturing semiconductor devices, comprising: 27. A method according to claim 26, further comprising providing a window, which is substantially transparent to the synchrotron radiation beam, between the ring and the substrate, and effecting said detecting step to the beam passed through the window. 28. A method according to claim 27, further comprising effecting said detecting step using a detector mounted on a stage for carrying the substrate. 29. A method according to claim 28, further comprising using a shutter in said controlling step. 30. An exposure method for manufacturing semiconductor devices, comprising: 31. A method according to claim 30, further comprising repeating the preliminary exposure for different positions of the sensor, and providing the information signal on the basis of exposure amounts calculated for the different positions. 32. A method according to claim 30, further comprising providing a window which is substantially transparent to the radiation and effecting said detecting step to radiation passed through the window. 33. A method according to claim 30, wherein the radiation includes x-rays in synchrotron radiation. 34. An exposure method for manufacturing semiconductor devices, comprising: 35. A method according to claim 34, further comprising providing a window which is substantially transparent to the radiation and effecting said detecting step to radiation passed through the window. 36. A method according to claim 35, wherein the radiation includes x-rays in synchrotron radiation. 37. An exposure method for manufacturing semiconductor devices, comprising: 38. A method according to claim 37, further comprising providing a window which is substantially transparent to the radiation and effecting said detecting step to radiation passed through the window. 39. A method according to claim 38, wherein the radiation includes x-rays in synchrotron radiation. 40. An exposure method for manufacturing semiconductor devices, comprising: 41. A method according to claim 40, further comprising providing a window which is substantially transparent to the radiation, and detecting radiation passed through the window. 42. A method according to claim 41, wherein the radiation includes x-rays in synchrotron radiation. 43. An exposure method for manufacturing semiconductor devices, comprising: 44. A method according to claim 43, further comprising providing a window which is substantially transparent to the radiation and detecting radiation passed through the window. 45. A method according to claim 44, wherein the radiation includes x-rays in synchrotron radiation. 46. An exposure method for manufacturing semiconductor devices, comprising: 47. A method according to claim 46, further comprising providing a window which is substantially transparent to the radiation and effecting said detecting step to radiation passed through the window. |
abstract | One form of the invention is directed to an apparatus that comprises step-down circuitry to better match impedance between an input and an output that includes a number of stages each electrically coupled to another and each including a charge storage device. The circuitry further includes a number of switching devices operable in a first electrical connectivity state to connect the charge storage device of each of the stages in series to receive electrical charge from the input and in a second electrical connectivity state opposite the first state to connect the charge storage device of each of the stages in parallel to discharge electricity through the output. This circuitry can be used in connection with a radioisotopic conversion cell. |
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claims | 1. A method for irradiating living tissue using a balloon applicator, comprising:placing the balloon applicator in a cavity of the living tissue, inflating the balloon applicator, and verifying the balloon placement by x-ray imaging taken exterior to the patient, the balloon applicator having a balloon wall adjacent to tissue to be irradiated, and the balloon wall being doped with x-ray contrast material, so that the imaging passes x-ray radiation tangentially through edges of the balloon and detects an outline of the balloon as a thin circle and the verifying of balloon placement is performed by detecting the thin circle outline of the balloon,positioning a switchable x-ray source at a desired position in the balloon applicator, andmoving the source through a series of positions within the balloon applicator, to administer radiation to tissue adjacent to the balloon. 2. The method of claim 1, wherein the step of moving the switchable x-ray source comprises stepping the position of the x-ray source through a series of distinct dwell positions, and turning the x-ray source on at each dwell position and off between dwell positions. 3. The method of claim 1, further including modulating x-ray dose depth at different positions within the balloon applicator by adjusting voltage in the switchable x-ray source, and adjusting intensity of x-rays during the treatment in accordance with a dose prescription, as the source is moved through the cavity. 4. The method of claim 1, including a plurality of iterations of the steps of placing the x-ray source into the balloon applicator within the cavity and removing the x-ray source, for a series of dose fractions over a treatment period which may extend over several days. 5. The method of claim 1, wherein the balloon is generally spherical in shape. 6. The method of claim 1, wherein the balloon is generally ellipsoidal in shape. 7. The method of claim 1, wherein the balloon is generally hotdog shaped. 8. The method of claim 1, wherein the cavity in the tissue is irregularly shaped, and wherein the balloon is shaped to match the cavity. 9. The method of claim 1, including administering x-ray dose from the switchable x-ray source at about 5-50 Gy per hour. 10. The method of claim 1, wherein the switchable x-ray source is operated at greater than about 40 kVp. 11. The method of claim 1, wherein the switchable x-ray source is operated at about 40 kVp to 80 kVp. 12. The method of claim 1, wherein, during treatment using the switchable x-ray source, the x-ray source is modulated in intensity of radiation by modulating current in the x-ray source to accurately achieve a prescribed isodose profile. 13. The method of claim 12, further including modulating x-ray dose depth at different positions within the balloon applicator by adjusting voltage in the switchable x-ray source. 14. A method for irradiating living tissue using a balloon applicator, comprising:placing the balloon applicator in a cavity of the living tissue, the applicator having a flexible shaft, and inflating the balloon applicator,positioning a switchable x-ray source in the flexible shaft and at a desired position in the balloon applicator, using a flexible, bendable cable with the x-ray source positioned at the end of the cable, the x-ray source being a miniature x-ray tube no greater than about 3.2 mm in diameter,switching on the x-ray source, andmoving the source through a series of positions within the balloon applicator, to administer radiation to tissue adjacent to the balloon. 15. The method of claim 14, wherein the step of moving the switchable x-ray source comprises stepping the position of the x-ray source through a series of distinct dwell positions, and turning the x-ray source on at each dwell position and off between dwell positions. 16. The method of claim 14, wherein the step of moving the x-ray source comprises moving the source continuously while modulating radiation from the x-ray source by at least one of: (a) switching the x-ray source on and off, (b) varying current in the x-ray source to modulate dose intensity, and (c) varying voltage in the x-ray source to modulate x-ray dose depth at different positions. 17. The method of claim 14, further including modulating x-ray dose depth at different positions within the balloon applicator by adjusting voltage in the switchable x-ray source, and adjusting intensity of x-rays during the treatment in accordance with a dose prescription, as the source is moved through the cavity. 18. The method of claim 14, wherein the step of moving the switchable x-ray source comprises stepping the position of the x-ray source through a series of distinct dwell positions, and modulating radiation from the x-ray source at different positions by at least one of: (a) switching the x-ray source on and off, (b) varying current in the x-ray source to modulate dose intensity, and (c) varying voltage in the x-ray source to modulate x-ray dose depth. 19. The method of claim 18, including a plurality of iterations of the steps of placing the x-ray source into the balloon applicator within the cavity and removing the x-ray source, for a series of dose fractions over a treatment period which may extend over several days. 20. The method of claim 14, including a plurality of iteration of the steps of placing the x-ray source into the balloon applicator within the cavity and removing the x-ray source, for a series of dose fractions over a treatment period which may extend over several days. 21. The method of claim 14, wherein the balloon is generally spherical in shape. 22. The method of claim 14, wherein the balloon is generally ellipsoidal in shape. 23. The method of claim 14, wherein the balloon is generally hotdog shaped. 24. The method of claim 14, including administering x-ray dose from the switchable x-ray source at about 5-50 Gy per hour. 25. The method of claim 14, wherein the switchable x-ray source is operated at greater than about 40 kVp. 26. The method of claim 14, wherein the switchable x-ray source is operated at about 40 kVp to 80 kVp. 27. The method of claim 14, further including draining liquids from the cavity of the living tissue while the inflated balloon is within the cavity, using drain lumens formed in the applicator and extending to exterior of the tissue. 28. The method of claim 27, wherein the balloon wall has an exterior surface with texture to define drain channels to provide a path for flow of liquids toward the exterior of the cavity. 29. The method of claim 28, wherein the balloon applicator has a generally central flexible shaft having drain holes in a distal end of the shaft, distal of the balloon, connected to the drain lumens which pass through the shaft, and further including drain holes in the exterior of the shaft proximal of the balloon for collecting liquids traveling over the surface of the balloon. |
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abstract | A system and method are provided for implanting ions into a workpiece in a plurality of operating ranges. A desired dosage of ions is provided, and a spot ion beam is formed from an ion source and mass analyzed by a mass analyzer. Ions are implanted into the workpiece in one of a first mode and a second mode based on the desired dosage of ions, where in the first mode, the ion beam is scanned by a beam scanning system positioned downstream of the mass analyzer and parallelized by a parallelizer positioned downstream of the beam scanning system. In the first mode, the workpiece is scanned through the scanned ion beam in at least one dimension by a workpiece scanning system. In the second mode, the ion beam is passed through the beam scanning system and parallelizer un-scanned, and the workpiece is two-dimensionally scanned through the spot ion beam. |
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description | The present invention relates to charged particle beam apparatuses and pattern measuring methods and more particularly, to a charged particle beam apparatus for aligning charged particles given off from a sample to detect them and a pattern measuring method therefor. As a semiconductor pattern becomes corpuscular, a delicate difference in configuration has an influence upon operational characteristics in a device and accordingly, needs for management of configuration have been raised. Therefore, a scanning electron microscope (SEM) used for inspection and measurement of semiconductors has been required of high sensitivity and high accuracy than in the past more and more. Especially, highly efficient detection of signals given off from a bottom or the like of such a pattern of large aspect ratio (depth/width) as a deep hole or trench has been demanded. In this case, how to detect efficiently a signal among discharged signals of less amounts is important and thus, making full use of discrimination of angle and direction of detection signals can be one of solving measures. JP-A-09-507331 discloses a SEM having two-stages of detectors adapted to discriminatively detect, for the sake of forming a high contrast mage of a hole bottom on the basis of electrons discharged from the bottom of such a deep hole as a contact hole, electrons passing through a trajectory close to the optical axis of an electron beam by having a small relative angle to the optical axis, that is, high angle electrons and electrons passing through a trajectory relatively distant from the optical axis of the electron beam by having a larger relative angle to the optical axis in contrast to the high angle electrons, that is, low angle electrons. Further, disclosed in JP-A-2006-228999 and JP-A-2006-332038 is a SEM having an aligner for secondary electrons adapted to control trajectories of secondary electrons with the aim of unifying signals detected by means of a plurality of detectors. By selectively detecting electrons in the specified directions on the basis of angle discrimination as explained in the JP-A-09-507331, JP-A-2006-228999 and JP-A-2006-332038, an image emphasizing information indicative of the hole bottom or the like can be formed. Of the electrons discharged from the hole bottom, however, electrons discharged from a portion near the side wall have a high possibility of impinging upon the side wall and the efficiency of detection of them is lowered. Especially when outputs of the right and left detectors are attempted to be uniform as explained in JP-A-2006-228999 and JP-A-2006-332038, values of brightness at the left and right edges are unified in the case of an objective of a hole pattern and there is a possibility that a signal of one edge cannot be detected sufficiently. In particular, with a critical dimension-scanning electron microscope (CD-SEM) used, highly accurate measurement is difficult to realize unless both the two edges representing measurement references have each a high S/N ratio. A charged particle beam apparatus and a pattern measurement method aiming at revealing information indicative of the edge of bottom of such a high-aspect structure as a deep hole and deep trench will be described hereinafter. In addition, a charged particle beam apparatus and a pattern measurement method which aim at discriminating a pattern which is difficult for judgment through the use of a top/down image. To comply with the above object and cope with the above problem, according to one embodiment of the present invention, a charged particle beam apparatus having an opening formation member formed with an opening for passage of a charged particle beam emitted from a charged particle source, and either a detector for detecting charged particles discharged from a sample and having passed through the passage opening or a detector for detecting charged particles resulting from bombardment upon another member of the charged particles having passed through the passage opening, comprises an aligner for aligning the charged particles discharged from the sample and a control unit for controlling the aligner, wherein the control unit controls the aligner to cause it to shift trajectories of the charged particles discharged from the sample and length measurement or critical dimensioning is executed on the basis of detection signals before and after alignment by the aligner. To comply with the above object and cope with the above problem, according to another embodiment of the invention, a charged particle beam apparatus having an opening formation member formed with an opening for passage of a charged particle beam emitted from a charged particle source, and either a detector for detecting charged particles discharged from a sample and having passed though the passage opening or a detector for detecting charged particles resulting from bombardment upon another member of the charged particles having passed through the passage opening, comprises an aligner for aligning the charged particles discharged from the sample and a control unit for controlling the aligner, wherein the control unit forms a signal waveform on the basis of an output of the detector, aligns the charged particles discharged form the sample such that a brightness indicated by a peak of the signal waveform satisfies a predetermined condition and executes length measurement of a pattern dimension on the sample by using a signal waveform obtained on the basis of the alignment. To comply with the above object and cope with the above problem, according to still another embodiment of the invention, a charged particle beam apparatus having an opening formation member formed with an opening for passage of a charged particle beam emitted from a charged particle source and either a detector for detecting charged particles having discharged from a sample and passed though the passage opening or a detector for detecting charged particles resulting from bombardment upon another member of the charged particles having passed through the passage opening, comprises an aligner for aligning the charged particles discharged from the sample and a control unit for controlling the aligner, wherein the control unit judges a line and/or a space formed on the sample on the basis of a signal waveform detected after alignment by means of the aligner. Advantageously, with the above construction, information indicative of the edge of bottom of a structure of high aspect such as a deep hole or deep trench can be revealed. Further, identification of a pattern whose judgment is difficult through a top/down image can be made possible. In order to selectively detect charged particles given off in a specified direction, a detector or a secondary electron conversion electrode is considered to be arranged at a specified angle or in a specified direction. Employed as a principal method for angle/direction discrimination is a method using a plurality of detectors. But, the arrangement and number of the detectors limits the angle and direction effective for signal discrimination, making it difficult to carry out discrimination at arbitrary angles and/or in arbitrary directions. To solve these difficulties, the use of an annular detector comprised of a plurality of detection elements is considerable. With the annular detector used, however, when the number of division of the elements is increased to provide flexibility in discrimination performance, the detection signal per element decreases. Further, errors in manufacture process of the apparatus will induce the possibility of degrading the accuracy of discrimination. The reduction in accuracy due to the manufacture error can be suppressed to some extent by using a secondary electron aligner but a sufficient signal quantity cannot sometimes be ensured. A charged particle beam apparatus and a pattern dimension measuring method which can sufficiently maintain a signal based on charged particles discharged in a specified direction while excluding signals in other directions will be described in the following by making reference to the accompanying drawings. Especially, in the present embodiment, a charged particle beam apparatus and a pattern dimension measuring method will be described which can permit efficient angle discrimination excluding a signal quantity reduction per a detector and detection element and a shift of field of view as well by using a plurality of detectors or detection elements at the time that angles/directions of secondary electrons are discriminatively detected. Of signals discharging from the sample, necessary information only is selectively detected, unnecessary signals are discarded as noises and charged particles discharged from the sample are aligned in an arbitrary direction by means of the aligner. By doing so, charged particles arriving at the detector are detected selectively in accordance with discharge angles. Particularly, by suppressing unnecessary information in signals discharged from a recess whose configuration is difficult to observe, useful information can be made conspicuous to permit execution of configuration inspection and the like. More specifically, a restrictive member for selective passage of part of charged particles discharged from the sample is arranged and the aligner is so controlled as to perform switchover between a first alignment state for passing charged particles in a first discharge direction and a second alignment state for passing charge particles in a second discharge direction. Through this control, in an aligned status, information in a specified direction is emphasized but information in a direction other than the specified direction is limited. Namely, by changing the aligned status and by detecting signals before and after the change, a signal in each of the plural directions can be improved in S/N. Especially, in the CD-SEM for measuring dimensions among a plurality edges at different positions, its measurement accuracy can be improved. According to the present embodiment, information around the bottom side wall of deep hole/deep trench (for example, edge information) can be detected with high accuracy and in the semiconductor manufacture process, for example, more accurate and effective process management can be ensured. Referring now to FIG. 1, the conceptual structure of a scanning electron microscope will be described. By applying an extraction voltage 12 between an electric field emission cathode 11 and an extraction electrode 13, a primary electron beam 1 is extracted. The primary electron beam 1 undergoes a convergence action by means of a condenser lens 14 and a scanning deflection by means of an upper scan deflector 21 and a lower scan deflector 22. Deflection intensities the upper and lower deflectors 21 and 22 have are so adjusted as to cause the primary electron beam 1 to scan on a sample 23 two-dimensionally with respect to a fulcrum at the lens center of objective lens 17. Similarly, the primary electron beam is subject to deflection actions by upper and lower image shift deflectors 25 and 26 which are adapted to change the scanning position. The primary electron beam 1 subject to the deflection is further accelerated by means of an accelerating cylinder 18 arranged in a passage of the objective lens 17. The primary electron beam 1 undergoing further acceleration at the lower stage is focused by the lens action of objective lens 17, finally bombarding on the sample 23 held by a holder 24. Under irradiation of the primary electrons, secondary electrons are discharged from the surface of sample 23. The secondary electrons can be sorted into secondary electrons 2 (a) at high angles in directions parallel to the optical axis and secondary electrons 2 (b) at low angles representing low angle components in directions inclined towards the sample surface. The secondary electrons propagate along the optical axis inversely to the primary electrons and reach a secondary electron limit plate 31. The high-angle secondary electrons 2 (a) pass through a hole of the secondary electron limit plate 31 and bombard on a reflector 27 (conversion electrode), that is a member different from a detector so as to be converted into tertiary electrons which in turn are detected by mean of an upper detector 28 (a). A secondary electron aligner 32 does not align the electron beam but aligns the secondary electrons 2(a) selectively towards the reflector 27 in order that the secondary electron 2(a) can be prevented from passing though the electron beam pass opening and from directing towards the electric field emission cathode 11. The secondary electrons 2 (b) at low angles impinge upon the secondary electron limit plate 31 and converted into tertiary electrons which in turn are detected by a lower detector 28 (b). Detected signals are processed by using an arithmetic unit 40. The signals detected by the individual detectors are converted into digital images. With a view to promoting an image S/N, the thus obtained images may be added and then imaged. A control unit 41 is connected to arithmetic unit 40, objective lens control power supply 42, retarding voltage power supply 43, accelerating voltage power supply 44, memory unit 45 and secondary electron aligner control power supply 46 so as to control operation of these components. For the purpose of angle discrimination, the secondary electrons 2(a) at high angles must pass through the hole of secondary electron limit plate 31 but the trajectory of the secondary electrons will sometimes depart from the optical axis under the influence of the use of image shift and the passage through objective lens 17. Conversely, the hole of secondary electron limit plate 31 will sometimes be disposed at a site distant from the optical axis under the influence of accuracies of assemblage and optical axis adjustment or will sometimes be disposed distantly from the optical axis intentionally with the aim of selecting the hole diameter. Typically, the secondary electron aligner is used in order for the secondary electrons 2 (a) at high angles to be caused to constantly pass through the hole of secondary electron limit plate 31. By using lower and upper secondary electron aligners 33(b) and 33(a), the alignment of secondary electron trajectory is controlled. To avoid the influence upon the trajectory of primary electrons, a Wien filter comprised of electrodes and magnetic field coils is used as the secondary electron aligner. Turning to FIG. 2, the concept of a method for controlling the secondary electron aligners 33 (a) and 33 (b) will be explained. Through the use of image shift deflectors 25 and 26, the primary electron beam is irradiated on a spot Ls distant from the optical axis on the sample surface. Secondary electrons 2 (a) at high angles discharged therefrom propagate inversely along the optical axis. When passing through the objective lens 17 and image shift deflectors 25 and 26, the secondary electrons undergo action of deflection and enter the lower secondary electron aligner 33 (b) with an axis eccentricity LL and at an angle θSE. Subsequently through the use of the lower secondary electron aligner 33 (b), the secondary electrons are so aligned at an angle θL as to be allowed to pass through the center of upper secondary electron aligner 33 (a). Next, by using the upper secondary electron aligner 33 (a), the thus aligned secondary electrons are so aligned at an angle θU as to be parallel to the optical axis, thus being permitted to pass through the center of secondary electron limit plate 31 in parallel with the optical axis. The foregoing explanation has been given by making reference to the longitudinally sectional view in FIG. 2. When taking actual rotation due to the magnetic field in the objective lens 17, however, the secondary electrons need to be controlled also in azimuthal directions. At that time, rotational angles due to the magnetic field are determined in advance in connection with the individual components of azimuths in the individual secondary electron aligners and control operation is conducted through matrix operation. The rotational angles due to the magnetic field can be determined by using a secondary electron arrival position detecting method or electron trajectory simulation. Next, a method for controlling alignment toward off-optical axis in the secondary electron limit plate 31 will be described with reference to FIG. 3. By controlling secondary electrons 2 (a) during non-use of image shift in such a manner that θL, and θU are equal to each other and in opposite directions, only position of the trajectory can be shifted while the trajectory being kept to be parallel to the optical axis. In the present embodiment, this is utilized so that signals at desired angles of the high angle secondary electrons 2(a) may be passed through the hole of secondary electron limit plate 31 by controlling the position of arrival of the high angle secondary electrons 2 (a) at the secondary electron limit plate 31 through the use of the secondary electron aligner. In addition, the present apparatus is also provided with the function to form a line profile on the basis of detected secondary electrons or reflected electrons. The line profile is formed on the basis of information indicative of an electron detection amount or brightness of a sample image which is obtained by scanning the primary electron beam linearly or two dimensionally and the thus obtained line profile is used for dimension measurement of a pattern formed on a semiconductor wafer, for instance. The control computer has been described in connection with FIG. 3 as being integral with the scanning electron microscope or so correspondingly but this is by no means limited and procedures as will be described below may be conducted with a control processor arranged separately from the scanning electron microscope. In that case, it is necessary to provide transmission media for transmitting a detection signal detected by the secondary signal detector 13 to the control processor and for transmitting signals from the processor to the lens and aligner of the scanning electron microscope and to provide an input/output terminal for inputs/output of signals transmitted by way of the aforementioned transmission media. In embodying the present invention as described below, a method and an apparatus therefor will be described in which only charged particles discharged in specified directions are detected by aligning secondary electrons discharged from a scanning spot by means of the secondary electron aligner without changing the incidence on the spot of a charged particle beam vertical to a substrate normally used, by inducing a profile of the detected signal and by detecting a corpuscular configuration of a pattern on the basis of the profile. While the present invention can be applicable to various kinds of charged particle beam apparatuses (SEM, FIB and so on), an example using a SEM as a typical apparatus will be described. In this embodiment, corpuscular configurations in deep trench/hole will be taken up. A deep trench pattern is diagrammatically illustrated at (a) in FIG. 4. When a deep trench pattern is observed with the normal SEM, a signal generating from the bottom of a hole is shielded by the side wall and hardly detected, with the result that its signal amount is decreased as compared to that from the upper portion of hole, producing an image and profile as shown at (b) in FIG. 4. Under this condition, however, the shape of trench bottom cannot be confirmed and accordingly, the whole of signal amounts from the trench bottom is increased typically by an expedient of raising the pre-charge or intensifying the electric field. With the signal amounts increased, the trench bottom configuration can be confirmed correspondingly easily but this is insufficient in many cases. Accordingly, by using the present method, selective detection of signals is carried out. Firstly, secondary electrons are curved toward a direction a in FIG. 3 by means of the SE aligner so that an image and a signal profile as shown at (c) in FIG. 4 may be obtained. Subsequently, when the secondary electrons are curved toward a direction b in FIG. 3, an image and a signal profile as shown at (d) in FIG. 4 can be obtained. As a result, at the bottoms of right and left side walls of the deep trench pattern shown at (a) in FIG. 4, the presence of a corpuscular configuration 102 of bottom failing to be detected at (b) in FIG. 4 can be detected. Since the SE aligner can be so operated as to be set in arbitrary direction and at arbitrary angle, even for a hole configuration having an edge through 360 degrees as shown at (a) in FIG. 4, bottom observation in all directions of edge can be permitted. Further, the SE aligner can be controlled by the application of only voltage and current so as to operate rapidly and therefore, a drift of the field view can be prevented during right/left switchover, thus ensuring that the SE aligner can be applied without decreasing the throughput even in the automated process. Turning now to FIG. 6, a flowchart is illustrated showing procedures for execution of dimension measurement on a sample on the basis of control of a position where secondary electrons reach by the action of the secondary electron aligner. Firstly, measurement conditions of a pattern are set from an input unit connected to the arithmetic unit 40 in step 601. Illustrated in FIG. 9 is an example of a measurement system in which the input unit 902 for setting the measurement conditions is coupled to the measurement apparatus including a scanning electron microscope 901 by way of a network. In FIG. 10, an example of screen of a GUI (Graphical User Interface) for measurement condition setting to be displayed on a display of input unit 902. In the GUI screen, conditions for measurement using an electron beam (for example, optical conditions of an electron beam) can be set. On the basis of the optical conditions set therein, an optical condition setter 905 sets conditions for control by the control unit 41 and setting information thereof is stored in the memory unit 45. On the basis of these conditions, conditions for controlling the sample stage and the deflectors adapted to adjust the positions of field of view. On the basis of inputs of coordinate information (location) and size of field of view (FOV size), a design data extractor 906 of an operation processor 903 built in the arithmetic operation unit 40 reads design data of the coordinates out of a design data medium 904 and displays it as figure data in a display area 1001. Along with optical conditions of the electron beam, length measurement boxes 1002 and 1003 indicative of measurement references (measurement start point and end point) of a measurement objective pattern are also set in the display area 1001. On the basis of detection of a predetermined brightness portion or a peak top, a pattern measurer 908 sets a measurement start point and a measurement end point in the length measurement boxes 1002 and 1003 and measures a dimension between the two points. Next, the optical condition setter 905 sets conditions for secondary electron aligners in accordance with the set measurement box position and on the basis of setting conditions stored in a memory 45 (step 602). As the conditions for secondary electron aligners, secondary electron alignment conditions are selected which can sufficiently assure brightness of edges contained in the measurement box. When secondary electron alignment conditions are stored in advance according to measurement conditions, two alignment conditions necessary for measurement are read out of the memory unit 45. In addition, in the event that proper alignment conditions are unknown, the secondary electron trajectory may be aligned in the measurement direction in a unit of predetermined alignment intensity and a signal waveform may be selected which occurs when a peak indicative of the edge becomes a predetermined value or more. Illustrated in FIG. 8 is a graph illustrating how the brightness shifts in response to alignment intensities of the plural secondary electron aligners in respect of a right or left edge of a hole. For the brightness at the edge portion being low (for example, a status at (b) in FIG. 4) and the brightness at the edge portion being high (for example, a statuses at (c) and (d) in FIG. 4), pieces of information can be obtained for both the statuses and therefore, measurement may be conducted by using signal waveforms obtained under such conditions for alignment that the brightness is the highest or the brightness is higher than a predetermined value. Alternatively, an approximation function passing through the peak position may be prepared and a peak position may be identified on the basis of the approximation function. Since, in the case of a hole-pattern, conditions for proper alignment are different for the right and left edges, peak extraction based on selection of two proper alignment conditions selected from a plurality of choices is carried out. Further, when a space between lower-layer and upper-layer patterns is to be measured, clarification of one edge position of the lower pattern is sufficient and therefore, only one alignment condition may be set. Next, under one or plural conditions for alignment of secondary electrons, the beam scanning based on the set optical conditions is carried out to form signal waveforms or images (step 603). In this step, beam scanning is conducted in accordance with setting of frame number (number of frames) exemplified in FIG. 10. When the secondary election alignment conditions are known in advance, images or signal waveforms are formed in accordance with the number of edges representing measurement reference but when the secondary electron alignment conditions are unknown, secondary electron alignment conditions of predetermined alignment intensity are set and images or signal waveforms are formed in accordance with the thus set individual conditions. After the processing, edge positions are detected and dimension measurements among the plural edges are executed (steps 604 and 605). For example, in the dimension measurement, a waveform 701 emphasized in a left edge and a waveform 702 emphasized in a right edge are selected and superimposed on each other by means of a waveform synthesizer 902 and a dimension between desired edges (x1 in the case of FIG. 7) is measured by means of the pattern measurer 908. Theoretically, the secondary electron aligner does not align the electron beam emitted from the cathode and so, fields of view used for acquisition of the waveforms 701 and 702 are quite the same. Accordingly, by superimposing the obtained waveforms each other in such a manner that their end positions are coincident, for example, a relative position between the two edges can indicate a distance between edges of an actual pattern and highly accurate dimension measurement can be accomplished. The foregoing embodiment has be explained by way of example of pattern dimension measurement conducted by using the secondary electron aligners built in the scanning electron microscope exemplified in FIG. 1 but the dimension measurement may be carried out by using a scanning electron microscope as exemplified in FIG. 11, for instance. The scanning electron microscope in FIG. 11 comprises a detector including, instead of the reflection plates 27 and 31 in FIG. 1, micro-channel plates 1101 and 1102 adapted to directly detect electrons the present example, the micro-channel plate 1102 substitutes for the opening formation member having an electron beam passage opening. Signals detected by the micro-channel plates 1101 and 1102 are amplified by amplifiers 1103 and 1104, respectively, and transmitted to a signal processing unit not shown. The scanning electron microscopes exemplified in FIGS. 1 and 11 comprise detectors for detecting electrons at low angles (namely, lower detector 28 (b) and micro-channel plate 1102). In the foregoing embodiment, the alignment is effected such that the trajectory of secondary electrons are positioned at a proper spot in relation to the member for limiting the passage of electrons at low angles and an example has been explained in which proper alignment conditions are determined on the basis of a change in upper-side detection signal but the alignment condition may be found out on the basis of the output of lower detector. In the example of FIG. 8, on the basis of the output of the upper side detector, an alignment intensity is determined at which the brightness of a specified edge is the highest but in contract to the maximization of the output at the upper detector, minimization of the output of the lower detector may be considered. Accordingly, on the basis of the output of the lower detector, conditions for alignment may be determined Embodiment 2 is directed to LS judgment An example of line and space image in the ordinary SEM is illustrated at (a) in FIG. 5. When the line and space are equal in size and a difference in clearness is mall therebetween, the profile becomes as shown at (b) in FIG. 5 and the line and space are difficult in discriminating from each other. According to the present invention, however, a space portion can be shaded as shown at (c) in FIG. 5 by aligning secondary electrons in a direction a in FIG. 3, making it easy to judge a concave/convex configuration. A component of angle of elevation of discharging secondary electrons is different at the line portion from that at the space portion. Since secondary electrons emitting from the space portion will shield secondary electrons the adjoining line pattern generates, the range of elevation angles at which discharging is possible is limited by the width of space and the height of line pattern and consequently, almost secondary electrons are of high angle components. Accordingly, these secondary electrons are more sensitive to the detection signal control using the SE aligner and secondary electron limit plate than those from the line portion, so that an image having a shade is formed at the space portion as shown at (c) in FIG. 5. At that time, the line can be discriminated from the space from a difference in profile between the line portion and the space portion. For the above reasons, the space portion is caused to have a asymmetrical brightness after the alignment. Consequently, as shown at (d) in FIG. 5, the gradient of profile at the space portion becomes larger than that at the line portion. By making full use of this, the concave/convex configuration can be judged to determine that the image at (a) in FIG. 5 corresponds to a configuration at (e) in FIG. 5. More specifically, since the signal after alignment becomes different depending on whether the peak generates from an interval “line to space” or “space to line”, the concave/convex configuration is judged by extracting the feature of the peak. For example, as shown at (d) in FIG. 5, it can be understood that when the alignment direction is as shown exemplarily at “a” in FIG. 3, the space exists in the alignment direction of high peak and the line exists oppositely to the alignment direction. It can be seen that the space exists in a direction of a peak having its skirt spreading and the line exists in a direction opposite to the alignment direction. It also can be seen that the line exists in an alignment direction of low peak and the space exists in a direction opposite to the alignment direction. Then, it can be seen that in an alignment direction of a peak having its skirt portion becoming narrow, the line exists in the alignment direction of peak and the space exists in a direction opposite to the alignment direction. Further, in the case of the alignment direction being “b” in FIG. 3, results inverse to those in direction “a” in FIG. 3 can be obtained. Preferably, by determining changes in peak as described above on the basis of comparison, for example, between signal waveforms before and after the alignment, between a signal waveform without alignment and a signal waveform with alignment or between signal waveforms with alignment (comparison between a plurality of signal waveforms obtained through alignment in inverse directions, for instance), line or space may be judged. In addition, by comparing adjoining peaks contained in a signal waveform after alignment, the aforementioned features may be extracted. On the basis of the judgment reference as above, the arithmetic operation unit 903 makes a decision as to whether a region surrounded by two peaks is a line or a space. Advantageously, the secondary electron aligner does not deflect the primary electron beam and therefore, a waveform required for measurement and a relative position of a waveform for concave/convex configuration judgment do not change, thereby ensuring a highly accurate judgment without misunderstanding the judgment results. By executing the secondary electron alignment continuously and making the aforementioned concave/convex judgment without performing movement of the stage or shift of field of view by the deflector before or after beam scanning for measurement, the aforementioned effects can be obtained. To add, after differentiating a signal waveform to make features of waveform clear, the concave/convex configuration judgment may be carried out. Further, by using the signal waveforms after alignment, both the judgment and measurement of the concave/convex configuration may be executed. In this case, a pattern can be measured by using a signal waveform in which a peak of an edge portion is revealed. According to the present embodiment, even for a pattern having continuous similar configurations and being hardly judged by only its feature of waveform such as the line and space in a SEM image, the height information can be utilized and more accurate concave/convex configuration judgment can therefore be assured. Further, since the incident electrons are not slanted and besides, the stage is not inclined, amendment for a drift of field of view is not necessary, having a small influence upon the throughput and present invention is effective even in the automated production process. In addition, the alignment of secondary electrons can be conducted in all directions to cause both a longitudinal pattern and a transverse pattern to be treated and besides, not only even a linear configuration such as a simple line and space but also a two-dimensional pattern having longitudinal and transverse edges can be judged. |
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description | This application is a continuation of U.S. patent application Ser. No. 09/730,338, now U.S. Pat. No. 6,789,046, entitled “PERFORMANCE LOGGING SOLUTION,” filed on Dec. 5, 2000, and issued on Sep. 7, 2004, the entirety of which is incorporated herein by reference. The present invention relates generally to computer systems, and more particularly to a system and method for gathering and aggregating performance metrics of a plurality of computers cooperating as an entity wherein the entity may be interfaced collectively as a whole and/or individually. Additionally, the system and method may be employed to gather and aggregate performance metrics of a plurality of entities cooperating as a higher entity where a parent entity may be interfaced directly or as part of an even higher collection of parent entities. The gathering of performance metrics is hierarchical with no predefined limits. With the advent of Internet applications, computing system requirements and demands have increased dramatically. Many businesses, for example, have made important investments relating to Internet technology to support growing electronic businesses such as E-Commerce. Since companies are relying on an ever increasing amount of network commerce to support their businesses, computing systems generally have become more complex in order to substantially ensure that servers providing network services never fail. Consequently, system reliability is an important aspect to the modern business model. A first approach for providing powerful and reliable services may be associated with a large multiprocessor system (e.g., mainframe) for managing a server, for example. Since more than one processor may be involved within a large system, services may continue even if one of the plurality of processors fail. Unfortunately, these large systems may be extraordinarily expensive and may be available to only the largest of corporations. A second approach for providing services may involve employing a plurality of lesser expensive systems (e.g., off the shelf PC) individually configured as an array to support the desired service. Although these systems may provide a more economical hardware solution, system management and administration of individual servers is generally more complex and time consuming. Currently, management of a plurality of servers is a time intensive and problematic endeavor. For example, managing server content (e.g., software, configuration, data files, components, etc.) requires administrators to explicitly distribute (e.g., manually and/or through custom script files) new or updated content and/or configurations (e.g., web server configuration, network settings, etc.) across the servers. If a server's content becomes corrupted, an administrator often has no automatic means of monitoring or correcting the problem. Furthermore, configuration, load-balance adjusting/load balance tool selection, and monitoring generally must be achieved via separate applications. Thus, management of the entity (e.g., plurality of computers acting collectively) as a whole generally requires individual configuration of loosely coupled servers whereby errors and time expended are increased. Presently, there is not a straightforward and efficient system and/or process for providing system wide performance metric data of the collection of servers. Additionally, there is no system and/or process for providing system wide performance metric data of a collection of arrays of servers. Some applications may exist that provide performance metrics of an individual server, however, these applications generally do not provide performance metrics across the logical collection of loosely coupled servers. For example, many times it is important to view information from the collection of servers to determine relevant system-wide performance. Thus, getting a quick response view of pertinent performance metrics associated with the plurality of servers may be problematic, however, since each server generally must be searched independently. Downloading all performance metric information from each individual server would overwhelm the network and be extremely cumbersome to an administrator to review all of the performance metric information to find problems or determine a state of the array. Furthermore, the complexity would be substantially increased for a collection of arrays. The present invention relates to a system and method of monitoring, gathering and aggregating performance metrics for a plurality of entities configured as a single entity. For example, the entities may include a plurality of members (e.g., computers, servers, clusters) collectively cooperating as a whole. In accordance with the present invention, a system interface is provided wherein a consistent and unified result set of performance information of a plurality of the entities as a whole may be obtained from any of the members associated with the entity. The system and method provides for configuration settings to be provided on a single computer or member wherein the configuration setting information (e.g., performance information to be logged) is propagated or replicated to each member of the entity. The configuration setting information is then employed by each member for determining which performance metric types (e.g., counters) to log. The members are notified of any changes to the configuration settings and a performance monitoring system dynamically adjusts the performance metric type logging accordingly. In one aspect of the invention, the performance metric types are logged to a data store based on a predefined time period and resolution for each member. The data is then dynamically aggregated to data of larger time periods and larger time resolutions. This is accomplished by performing mathematical operation on the data values of the data points for the predefined time period and time resolution to provide data points of higher time periods and time resolutions for each performance metric being logged. A performance gathering and aggregation system is provided that receives requests from a source or requestor to receive performance metric data of a single member or of the entity as a whole. The data gathering and aggregation system provides a request to a query component, which queries the members for the data values for the particular time period and resolution stored in the data store and passes the results to the data gathering and aggregation system. The performance gathering and aggregation system aggregates and formats the results for transmitting to the requestor. The query component includes error handling for handling members that are non-responsive or send invalid results. If performance metrics information has been requested for the entity as a whole, the performance gathering and aggregation system matches up data point values with respect to time for each member that provides valid results and provides aggregated data values for each time point over a specified time period and time resolution to the requestor. The data is aggregated by performing mathematical operations on each time data point for a particular metric type for each entity that provides valid performance data. The following description and the annexed drawings set forth in detail certain illustrative aspects of the invention. These aspects are indicative, however, of but a few of the various ways in which the principles of the invention may be employed and the present invention is intended to include all such aspects and their equivalents. Other advantages and novel features of the invention will become apparent from the following detailed description of the invention when considered in conjunction with the drawings. The present invention is now described with reference to the drawings, wherein like reference numerals are used to refer to like elements throughout. The present invention is described with reference to a system and method for monitoring, gathering and aggregating performance data of a plurality of members forming an entity. The performance data to be logged is defined by a configuration setting at any member of the entity and this information replicated to other members of the entity. Each member monitors its performance data and logs this data locally to a data store based on a predefined time period and time resolution. The performance data of the predefined time period and time resolution is then aggregated to a plurality of data sets of larger time periods and time resolutions. An interface can then request performance data from the members via a performance gathering and aggregation system. The performance gathering and aggregation system requests or queries the plurality of member for performance data for a performance metric type via a query component. The performance gathering and aggregation system receives performance data from the members based on a requested time period and resolution and a performance metric type. The request can be either for a single member or from all members of the entity. The performance gathering and aggregation system will then aggregate and format performance data for the particular performance metric type based on the requested time period and resolution. If the request is for receiving the performance metric type for the entity, the performance gathering and aggregation system will aggregate the performance data values of similar time points to provide a unified performance result set for the entity over the particular time period and resolution. The unified result set is then returned to the requestor. In accordance with the present invention, a performance system is provided that greatly facilitates management and administration of an entity. The performance system substantially automates performance information retrieval by enabling an application to retrieve the performance metric data of the entity from any of a plurality of systems operatively coupled to the entity. A consistent experience is therefore provided wherein the performance metric data of the entity may be retrieved as if the entity were a singular machine—thereby providing a substantial improvement over conventional systems that may require an administrator to individually retrieve performance metric data from each machine comprising the entity. Thus, the present invention saves time and administration costs associated with conventional systems. Moreover, system troubleshooting is improved since entity members may be considered upon as a collective whole (e.g., retrieving system wide performance) and/or individual members may be identified and operated upon. Referring initially to FIG. 1, a system 10 illustrates a particular aspect of the present invention related to a performance system for monitoring, gathering and aggregating performance metrics of a plurality of systems cooperating as an entity. A plurality of systems (e.g., computers, servers, machines) for example, computer systems 1 through N (N being an integer) 22a through 22d may be operatively coupled to a network 14 thereby forming an entity 12. Other sources that may not be part of the entity 12, may also be coupled to the network 14 for retrieving, gathering and aggregating performance metric data from the entity 12 or for gathering and aggregating performance raw metric data from the entity 12 by employing its own gathering and aggregation system. For example, an external consumer of data 26 can connect to one of the computer systems 22 through the network 14 to retrieve raw or aggregated performance metric data or connect to one of the interfaces 16a through 16d to retrieve raw or aggregated performance metric data. Additionally, an external user interface 27 can connect to one of the computer systems 22 through the network 14 to retrieve raw or aggregated performance metric data or connect to one of the interfaces 16a through 16d to retrieve raw or aggregated metric data. Furthermore, a parent entity 28, parallel entities 29 and/or a child entity 30 can connect to any member of the entity for retrieving and passing performance metric data between entities for gathering and/or aggregating. In order to request and provide specific gathered and aggregated operation performance information of the entity 12, a plurality of interfaces (e.g., computer monitor) 16a through 16d may provide output, and an input device (e.g., mouse, keyboard) 24a through 24d may provide input requests to the operation gathering and aggregation system 18a through 18d. As depicted by the system 10, the interface 16 enables an application or process to retrieve, display or monitor the entity 12 from each member 22a–22d and/or from non-members such as any of the components 26–30. The interface 16 provides a consistent interface for an application or process to measure the operational performance metrics of the entity 12 as if it was a singular machine. Consequently, the user does not have to administer (e.g., gain access to each machine) and configure (e.g., download new content/software) each machine individually. Thus, time is saved and errors are mitigated. It is noted that the interface 16 generally does not have to run on each computer in the system 10. As will be described in more detail below, full entity operation monitoring may be achieved by interfacing to a single member, for example. The interface 16 may be served with information provided from each member 22a through 22d employing any of the performance gathering and aggregation systems 18a through 18d. This may be achieved by enabling each member to distribute information to the entity 12. Therefore, the interface 16 may provide aggregated performance information of the entity as a whole through the performance gathering and aggregation system 18—in contrast to conventional systems wherein performance information of a member may be received and/or displayed only at the individual member employing a performance monitoring system 20a–20d. For example, computer systems 22a–22d processor performance may be retrieved and/or displayed as an aggregation of the output of each member of the entity 12. Any of the interfaces 16a through 16d may be provided with a similar consistent result set. It is noted that the members 22a through 22d may also be entities. For example, some members could also be a collection of members represented by an entity. Thus, the entity 12 may include members that are entities in their own right. Alternatively, the interface 16 is provided with individual operational performance metrics from any of the performance gathering and aggregation systems 18a through 18d by requesting this information from that particular member. Furthermore, entity configurations may be modified from any of the interfaces 16 by enabling the user to provide input to the interface and thereby distribute resultant modifications throughout the entity 12. This may be achieved for example, by providing the input to a single member wherein the single member may then distribute the modified configuration throughout the entity 12. It is to be appreciated that other distribution systems may be provided. For example, rather than have entity operation information centrally distributed and aggregated at the single member, individual members 22a–22d may share a master file (e.g., XML) describing the configuration information of each member. FIG. 2 illustrates entity configuration with respect to configuration of the types of performance metrics that are to be monitored by members of the entity, so that there is uniformity throughout the entity of the types of metrics to be monitored. This uniformity allows for performance information to be monitored and retrieved for not only individual members, but also for the entity as a whole. An entity 45 is provided having a first member 50 and a plurality of additional members 60 coupled to the first member 50. The first member 50 includes global performance configurations settings 54 and the members 60 include member specific configurations settings 62. Setting of the performance metric types to be logged in the global performance configuration settings 54 on the first member 50 provides for propagation of these settings to the configuration settings 62 of the members 60. This is accomplished by employing a replication engine 52. This provides for logging of the same performance metrics for each member 60 of the entity 45. Any change to the global performance configuration settings 54 causes the member configuration settings to be dynamically updated. It is to be appreciated that the global performance configuration settings may be set at any of the plurality of members of the entity and the setting propagated to the other members. FIG. 3 illustrates a block schematic view of the components employed to provide both a singular member result set of performance metric information and an aggregates entity result set of performance metric information utilizing the performance system of the present invention. Each member 60 and an aggregator member 50 can include a performance monitor system 66 coupled to a performance configuration source 62 and a performance data source 68. The performance monitor system 66 utilizes the configuration setting information in the performance configuration source 62 to determine the different performance metrics to log. The performance monitor system 66 periodically retrieves the performance data values of the different performance metrics from the performance data source 68. The performance monitor system 66 then periodically logs the performance data values in the data store 64 related to that particular member. The performance metric data values are repeatedly logged based on a predefined time period and time resolution, until the configurations settings are changed. It is to be appreciated that component(s) may reside between the performance monitor system 66 and the data store 64 for setting up communication links, accessing data and/or transforming data. The performance monitor system 66 also dynamically aggregates the performance metric data values based on the predefined time period and time resolution to data sets of higher time periods and higher time resolution (e.g., 10 seconds, 1 minute, 15 minutes, 1 hour, 1 day) (e.g., aggregates across time, hereinafter referred to as “time aggregation”). The aggregation of data reduces the amount of storage required in addition to increasing the speed of the queries and the user interface. Various mathematical methodologies may be employed to perform time aggregation. For example, for an aggregation from ten seconds to one minute, the performance data values would include six points. The data values of these six points could be aggregated to a minute by taking the average, the minimum, the maximum, the last, the weighted average or some other value of the data values of these six points for supplying the one minute data value. An interface 56 can provide a request to a performance aggregation system 58 for performance data for a particular performance metric over a particular time period and time resolution based on a single member result set or based on an aggregated result set of the performance data values over the entire entity. If the interface 56 is a user interface, it may be constrained by the number of data points that can be plotted, therefore, interpolation of the data to smaller sets can be done at query time or by database aggregation. The performance gathering and aggregation system 58 requests this information from the data stores 64 through a query component 49. The query component 49 includes error handling. For example, if a member is not available results are returned from the other members and aggregated appropriately, while an error is returned for the unavailable member, which is not utilized to provide the aggregated results. For example, the query component 49 can pass back a value of −1 when no data is available for a respective member. The performance gathering and aggregation system 58 takes into account data not available. Furthermore, the interface 56 can be operable to properly convey the error data. The interface 56 may also be operable to provide selectability for which performance metrics to be returned to the interface 56. It is to be appreciated that component(s) may reside between the performance gathering and aggregation system 58 and the query component 49 for setting up communication links, accessing data and/or transforming data. The performance metric data is provided to the performance gathering and aggregation system 58 for the particular metric requested. The performance gathering and aggregation system 58 can transform the data points to fit within a particular time period and resolution. If the performance metric data is to be returned for the entire entity, the performance gathering and aggregation system collapses or aggregates the performance metric data values for time points within the specified time period and time resolution for each member submitting data values (e.g., aggregates across members, hereinafter referred to as “entity aggregation”). Various mathematical methodologies may be employed to perform entity aggregation. For example, for an aggregation of 4 members, the performance data values for each time data point would include four points. The data values of these four points could be aggregated by taking the average, the minimum, the maximum, the last, the weighted average or some other value of the data values of these four points for supplying a single aggregated data value for the entity for that particular time point. The above methodology can be repeated for all time data points within the specified time period and time resolution. It is to be appreciated that not all members will return data values or have data values for a particular point in time. In this situation, the gathering and aggregation system 58 disregards the data value and determines an appropriate aggregated data value on data values considered valid that were returned by the members and ignores the invalid data values in the aggregation. FIG. 4a illustrates a block schematic diagram of the operation of the performance monitor system 66. The performance monitor system 66 includes a configuration store 70, a configuration consumer component 76 and a metric consumer component 80. Configuration settings are transmitted to the configuration store 70 from the member itself or from another member of the entity. Any changes in the configuration settings to the configuration store 70 causes an instance operation event 74 to notify the configuration consumer component 76 that there has been a change in the configuration settings. The configuration consumer component 76 determines any changes, additions or deletion to the metrics being logged. The configuration consumer component 76 creates and updates an active metrics table (e.g., list of metrics currently being logged). If changes to global configuration class indicate that logging is turned on, the configuration consumer component 76 creates an event timer 78 that periodically informs the counter consumer component 80 to update the data store 64 with the current metric values. The counter consumer component 80 queries a metric source 72 periodically and logs the metric data to the data store 64 based on a predefined time period and time resolution defined by the timer event 78. The metric source 72 could be any of a variety of metric data source types based on an operating system environment. For example, in the Microsoft® Windows® Operating System environment, the metric data source type could be Windows Management Instrumentation (WMI), which is a support mechanism for management of systems in an enterprise. WMI allows developers to use a simple, consistent mechanism to query for information on computers across an enterprise (e.g., hardware settings, performance information, driver configuration, BIOS information, application settings, event log information). WMI allows for both hardware and software to be modeled. It is to be appreciated that other computer management systems that provide performance metric information may be employed to carry out the present invention. The metric consumer component 80 could employ any of a variety of services to query performance information from the metric source component 72. Additionally, the metric source could be a variety of metric source types based on an operating system environment. For example, in Microsoft® Windows® Operating System environment, the metric consumer component 80 could employ performance data helper (PDH) and the metric source 72 could be the performance library (PERFLIB), which is a dynamically linked library residing on the Microsoft® Windows® Operating System. The metric consumer component 80 then stores the metric information in the data store 64. If changes to global configuration class indicate that logging is turned off, the configuration consumer component 76 deletes the event timer 78 and logging is ceased. FIG. 4b illustrates a block schematic diagram of the operation of the performance monitor system 66 in cooperation with the performance gathering and aggregation system 58. As previously described, the performance monitor system 66 logs performance metric data periodically based on the configuration settings and the event timer setting. The performance metric data is stored in separate predefined time period and time resolution data sets for each metric being logged. The performance data stored for each metric is based on a time period of the timer event 78. The performance monitor system 66 includes a member time aggregation component 82, which dynamically collapses or aggregates data to additional data sets based on larger time periods and larger time resolutions from a first data set based on the predefined time period and time resolution. For example, FIG. 5 illustrates a number of stored data sets residing in the data store 64. The performance monitor system 66 logs a metric list 90, a member list 110 and ten second performance data stored for each metric being logged. The ten second performance metric data is stored for metric #1 92A, metric #2 94A, metric #3 96A up to metric #N 100A. The ten second performance metric data includes performance metric data logged every ten seconds defined by the event timer or the like. The time aggregation component 82 then dynamically collapses or aggregates performance metric data to data sets of larger time periods and resolutions employing the ten second tables. FIG. 5 illustrates that the ten second data being aggregated up to one minute performance metric data sets 92B, 94B, 96B up to 100B, which is then aggregated to additional performance metric data sets, all the way up to one day performance metric data sets 92N, 94N, 96N up to 100N. Referring again to FIG. 4b, the performance data gathering and aggregation system 58 includes an entity aggregation component 84. The performance data gathering and aggregation system 58 will receive a request from a source (e.g., the interface component 16 to receive performance information on a metric over a certain time period and time resolution for either a particular member or for the entity as a whole. The performance gathering and aggregation system 58 will then access or query a particular performance metric data set relating to the time period and time resolution to be received for that performance metric over a single member or over all members. If the request is for performance metric data for the entity, the entity aggregation component 84 will aggregate the metric data over the members to find a single metric value for each time data point. The aggregated values will then be transformed to appropriate data points for the particular time resolution requested. The aggregated and transformed values will then be transmitted back to the source requesting the data. Referring now to FIGS. 6a and 6b, a particular example is illustrated where the interface component 16 is a user interface adapted to display performance metric data graphically over a particular time period and time resolution. A user interface 120 is provided for viewing performance metric data for either a single member or for an entity as a whole. The user interface 120 includes a scope pane 125 and a results pane 132. FIG. 6a depicts an entity wide view 128 and FIG. 6b depicts a member view 140, respectively. Referring to FIG. 6a, the entity wide view 128 may be selected via a mouse for example by selecting display object 122 (e.g., entity node MyCluster). From the scope pane 125, a user may then navigate to a plurality of pages (e.g., displayed in the results pane 132 and/or via other menus) that provide performance and status views of the entity as a whole. Referring briefly to FIG. 6b, the members view 140 may be provided to enable a user to view pages associated with a particular member by selecting display objects in the scope pane 125 associated with that particular member (e.g., by selecting display object 124 or 126). As described above in relation to FIG. 6a, an entity node display object 122 may be provided to display and enable selection of an entity. It is to be appreciated that a plurality of entities having associated members may be defined. In order to facilitate management and navigation, each member (e.g., member node display objects 124 and 126) may be presented in the entity node view. For example, demobrick-01 and demobrick-02 represent member nodes 124 and 126, and appear under entity node 122-MyCluster. The entity node view 122 may be independent of each member's actual topology and additionally may allow for the inclusion of members that are not part of the same subnet, domain, and/or physically near. The user interface 120 may provide performance views to enable a user to display to a chart control (e.g., performance counters). The performance metric data may be aggregated for the entity and/or related to a specific member. If a user selects an entity wide view as described above, a performance metric display 128 may be provided as depicted in the results pane 132. As illustrated in the scope pane 125, an entity node 122 may be highlighted indicating to the user that metric information is provided as an aggregated set from members 124 and 126. As illustrated in the display output 128 performance metric information for the entity may be aggregated and displayed. The aggregated information may be provided from a plurality of sources such as from metrics associated with performance aspects of members serving the entity. For example, a second display output window 134 may provide information regarding particular metrics such as processor utilization, memory available, and server requests per second. Inputs 136 and 138 (e.g., Add/Remove) may be provided to add and remove metrics from the display 128 respectively. For example, if input Add 136 were selected, a list (not shown) may be provided to enable the user to select a performance metric for display output. Similarly, counters may be removed by selecting (e.g., mouse highlighting) a metric within the display 134 and then selecting the Remove input 138. A selection input/output 130 (e.g., rectangle with selection arrow) may be provided to enable the user to see and/or select a suitable time period for monitoring the aggregated data described above. As the time period is modified, the resolution of the display output 128 may thereby be altered accordingly. FIG. 6b illustrates a view similar to FIG. 6a, however, the display output 140 is directed from a particular member. As shown in the scope pane 125, demobrick-01 124 may be highlighted to indicate that data is provided from a member. The change in selection of any member or addition of a metric to a current screen causes the user interface to send a request to the data gathering and aggregation system 58. The data gathering and aggregation system 58 then performs the functions as previously discussed and returns the performance metric data values to the user interface 120 for display. FIG. 7a illustrates one particular methodology for providing similar configuration data settings for metrics to be logged at each member of an entity. In step 150, the global configuration settings for the counters to be logged are set at a first member 50. The first member 50 then propagates these global configuration settings to the configuration settings 62 of each member 60 (including the first member 50) by employing the replication engine 52 in step 160. FIG. 7b illustrates one particular methodology for each member in responding to the methodology for providing similar configuration data settings for metrics to be logged at each member of an entity of FIG. 7a. In step 200, the member 60 receives the initial configuration settings and begins logging metric data based on the initial configuration in step 210. In step 220, the performance monitor system 66 monitors if the configuration settings have changed. If the configuration settings have not changed (NO), the performance monitor system 66 continues logging metric data according to the initial configuration in step 210. If the configuration settings have changed (YES), the performance monitor system 66 begins logging metric data according to the new configuration settings in step 230. The performance monitor system 66 then returns to step 220 to determine if any new changes have occurred to the configurations settings. FIG. 8a illustrates one particular methodology for logging and member time aggregation of performance metric data. In step 300, the configuration consumer component 76 receives a metric logging configuration change. In step 310, the configuration consumer component 76 updates the metrics being logged. In step 320, the performance monitor system 66 begins logging the performance data for selected metrics for data sets of a first time period and time resolution. In step 330, the data of a first time period and resolution is aggregated to data sets of higher time periods and resolutions (e.g., 1 minute, 15 minute, 1 hour, 1 day). FIG. 8b illustrates one particular methodology for requesting and receiving performance metric data. In step 350, the performance data gathering and aggregation system 58 receives a request from interface component 56 for performance metric data over a specific time period and time resolution. The request includes an entity specific request and time period or resolution request for the metric to be viewed. In step 360, the data and gathering aggregation system 58 builds and passes the request to the respective members. The results are returned for the requested performance metric data of a data set of a specific time period and resolution from the data stores 64 of the responding members 60, in step 370. In step 380, the data gathering and aggregation system 58 aggregates the performance metric information data values for each data time point and formats the results for the appropriate time resolution. In step 390, the data gathering and aggregation system 58, returns the results to the interface 56. In order to provide a context for the various aspects of the invention, FIG. 9 and the following discussion are intended to provide a brief, general description of a suitable computing environment in which the various aspects of the present invention may be implemented. While the invention has been described above in the general context of computer-executable instructions of a computer program that runs on a computer and/or computers, those skilled in the art will recognize that the invention also may be implemented in combination with other program modules. Generally, program modules include routines, programs, components, data structures, etc. that perform particular tasks and/or implement particular abstract data types. Moreover, those skilled in the art will appreciate that the inventive methods may be practiced with other computer system configurations, including single-processor or multiprocessor computer systems, minicomputers, mainframe computers, as well as personal computers, hand-held computing devices, microprocessor-based or programmable consumer electronics, and the like. The illustrated aspects of the invention may also be practiced in distributed computing environments where tasks are performed by remote processing devices that are linked through a communications network. However, some, if not all aspects of the invention can be practiced on stand-alone computers. In a distributed computing environment, program modules may be located in both local and remote memory storage devices. With reference to FIG. 9, an exemplary system for implementing the various aspects of the invention includes a conventional computer 420, including a processing unit 421, a system memory 422, and a system bus 423 that couples various system components including the system memory to the processing unit 421. The processing unit may be any of various commercially available processors, including but not limited to Intel x86, Pentium and compatible microprocessors from Intel and others, including Cyrix, AMD and Nexgen; Alpha from Digital; MIPS from MIPS Technology, NEC, IDT, Siemens, and others; and the PowerPC from IBM and Motorola. Dual microprocessors and other multi-processor architectures also may be employed as the processing unit 421. The system bus may be any of several types of bus structure including a memory bus or memory controller, a peripheral bus, and a local bus using any of a variety of conventional bus architectures such as PCI, VESA, Microchannel, ISA and EISA, to name a few. The system memory includes read only memory (ROM) 424 and random access memory (RAM) 425. A basic input/output system (BIOS), containing the basic routines that help to transfer information between elements within the server computer 420, such as during start-up, is stored in ROM 424. The computer 420 further includes a hard disk drive 427, a magnetic disk drive 428, e.g., to read from or write to a removable disk 429, and an optical disk drive 430, e.g., for reading a CD-ROM disk 431 or to read from or write to other optical media. The hard disk drive 427, magnetic disk drive 428, and optical disk drive 430 are connected to the system bus 423 by a hard disk drive interface 432, a magnetic disk drive interface 433, and an optical drive interface 434, respectively. The drives and their associated computer-readable media provide nonvolatile storage of data, data structures, computer-executable instructions, etc. for the server computer 420. Although the description of computer-readable media above refers to a hard disk, a removable magnetic disk and a CD, it should be appreciated by those skilled in the art that other types of media which are readable by a computer, such as magnetic cassettes, flash memory cards, digital video disks, Bernoulli cartridges, and the like, may also be used in the exemplary operating environment, and further that any such media may contain computer-executable instructions for performing the methods of the present invention. A number of program modules may be stored in the drives and RAM 425, including an operating system 435, one or more application programs 436, other program modules 437, and program data 438. The operating system 435 in the illustrated computer may be a Microsoft operating system (e.g., Windows NT operating system). It is to be appreciated that other operating systems may be employed such as UNIX for example. A user may enter commands and information into the server computer 420 through a keyboard 440 and a pointing device, such as a mouse 442. Other input devices (not shown) may include a microphone, a joystick, a game pad, a satellite dish, a scanner, or the like. These and other input devices are often connected to the processing unit 421 through a serial port interface 446 that is coupled to the system bus, but may be connected by other interfaces, such as a parallel port, a game port or a universal serial bus (USB). A monitor 447 or other type of display device is also connected to the system bus 423 via an interface, such as a video adapter 448. In addition to the monitor, computers typically include other peripheral output devices (not shown), such as speakers and printers. The computer 420 may operate in a networked environment using logical connections to one or more remote computers, such as a remote client computer 449. The remote computer 449 may be a workstation, a server computer, a router, a peer device or other common network node, and typically includes many or all of the elements described relative to the server computer 420, although only a memory storage device 450 is illustrated in FIG. 9. The logical connections depicted in FIG. 9 include a local area network (LAN) 451 and a wide area network (WAN) 452. Such networking environments are commonplace in offices, enterprise-wide computer networks, intranets and the Internet. When employed in a LAN networking environment, the server computer 420 may be connected to the local network 451 through a network interface or adapter 453. When utilized in a WAN networking environment, the server computer 420 generally may include a modem 454, and/or is connected to a communications server on the LAN, and/or has other means for establishing communications over the wide area network 452, such as the Internet. The modem 454, which may be internal or external, may be connected to the system bus 423 via the serial port interface 446. In a networked environment, program modules depicted relative to the computer 420, or portions thereof, may be stored in the remote memory storage device. It will be appreciated that the network connections shown are exemplary and other means of establishing a communications link between the computers may be used. In accordance with the practices of persons skilled in the art of computer programming, the present invention has been described with reference to acts and symbolic representations of operations that are performed by a computer, such as the computer 420, unless otherwise indicated. Such acts and operations are sometimes referred to as being computer-executed. It will be appreciated that the acts and symbolically represented operations include the manipulation by the processing unit 421 of electrical signals representing data bits which causes a resulting transformation or reduction of the electrical signal representation, and the maintenance of data bits at memory locations in the memory system (including the system memory 422, hard drive 427, floppy disks 429, and CD-ROM 431) to thereby reconfigure or otherwise alter the computer system's operation, as well as other processing of signals. The memory locations wherein such data bits are maintained are physical locations that have particular electrical, magnetic, or optical properties corresponding to the data bits. What has been described above are preferred aspects of the present invention. It is, of course, not possible to describe every conceivable combination of components or methodologies for purposes of describing the present invention, but one of ordinary skill in the art will recognize that many further combinations and permutations of the present invention are possible. Accordingly, the present invention is intended to embrace all such alterations, modifications and variations that fall within the spirit and scope of the appended claims. |
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description | The present application is related to controlling reactivity in a nuclear fission reactor. Illustrative embodiments provide a reactivity control assembly for a nuclear fission reactor, a reactivity control system for a nuclear fission reactor having a fast neutron spectrum, a nuclear fission traveling wave reactor having a fast neutron spectrum, a method of controlling reactivity in a nuclear fission reactor having a fast neutron spectrum, methods of operating a nuclear fission traveling wave reactor having a fast neutron spectrum, a system for controlling reactivity in a nuclear fission reactor having a fast neutron spectrum, a method of determining an application of a controllably movable rod, a system for determining an application of a controllably movable rod, and a computer program product for determining an application of a controllably movable rod. The foregoing summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. The present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Illustrative Reactivity Control Assembly Referring now to FIG. 1A and given by way of overview, an illustrative reactivity control assembly 10 for a nuclear fission reactor (not shown) is shown. A reactivity control rod 12 includes neutron absorbing material 14 configured to absorb neutrons (not shown). At least a portion of the neutron absorbing material 14 includes fertile nuclear fission fuel material 16. At least one sensor 18 is physically associated with the reactivity control rod 12. The sensor 18 is configured to sense status of at least one reactivity parameter associated with the reactivity control rod 12. Illustrative details will be set forth below by way of non-limiting examples. It will be appreciated that the reactivity control rod 12 may be any type of suitable reactivity control rod. In some embodiments the reactivity control rod 12 may be a stand-alone reactivity control rod. That is, in such an arrangement the reactivity control rod 12 is not grouped into an assembly with other rods, such as nuclear fission fuel rods and/or other reactivity control rods. In some other embodiments, the reactivity control rod 12 may be part of an assembly that includes nuclear fission fuel rods and/or other reactivity control rods. It will also be appreciated that the reactivity control rod 12 may have any physical shape as desired for a particular application. Given by way of non-limiting examples, in various embodiments the reactivity control rod 12 may have a cross-sectional shape that is square, rectangular, circular, ovoid, or any shape as desired. In various embodiments the reactivity control rod 12 may be embodied as a blade, and may have any cross-sectional shape as desired, such as a rectangle, an “X”, a “+”, or any other shape. The reactivity control rod 12 may have any shape that is suited for the nuclear fission reactor in which the reactivity control rod 12 is to be used. No limitation regarding shape of the reactivity control rod 12 is implied, and none should be inferred. In some embodiments the neutron absorbing material 14 may be configured to absorb fast spectrum neutrons. For example, the neutron absorbing material 14 may have an absorption cross-section that permits absorption of fast spectrum neutrons—that is, neutrons having an energy level on the order of at least around 0.11 MeV. Given by way of non-limiting example, the neutron absorbing material 14 may have an absorption cross-section on the order of around 10 barns or less. In some embodiments the fertile nuclear fission fuel material 16 may serve as the component of the neutron absorbing material 14 that absorbs the fast neutrons. In some other embodiments, other component(s) of the neutron absorbing material 14 may also serve as additional component(s) of the neutron absorbing material 14 (in addition to the fertile nuclear fission fuel material 16) that absorbs the fast neutrons. Illustrative details regarding fertile nuclear fission fuel material 16 and other components of the neutron absorbing material 14 will be set forth below. In some applications, it may be desirable to maintain the neutron spectrum of a nuclear fission reactor within the fast neutron spectrum. Given by way of non-limiting examples, the reactivity control assembly 10 may be used to help control reactivity in a fast nuclear fission reactor, such as without limitation a traveling wave reactor or a fast breeder reactor, like a liquid metal fast breeder reactor or a gas-cooled fast breeder reactor, or the like. To that end, in some other embodiments the neutron absorbing material 14 may be configured to reduce moderation of neutrons. For example, the neutron absorbing material 14 may have a suitably large atomic mass that can help reduce the amount of slowing down of fast spectrum neutrons. As such, a reduction may be made in softening of the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. It will be appreciated that, given by way of non-limiting examples, elements of the actinide series, such as without limitation uranium and thorium, present a sufficiently large atomic mass to help reduce moderation of neutrons. In some embodiments the fast spectrum neutrons may be part of a nuclear fission traveling wave. A nuclear fission traveling wave may also be referred to as a nuclear fission deflagration wave. Non-limiting examples of initiation and propagation of a nuclear fission traveling wave is described in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; and U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of which are hereby incorporated by reference. The fertile nuclear fission fuel material 16, that is included in the neutron absorbing material 14, can include any type of fertile nuclear fission fuel material as desired for a particular application. For example, in some embodiments the fertile nuclear fission fuel material 16 may include uranium, such as 238U. It will be appreciated that the absorption cross-spectrum for fast neutrons of 238U is on the order of around 0.2 barns. In some other embodiments the fertile nuclear fission fuel material 16 may include thorium, such as 232Th. It will be appreciated that the absorption cross-spectrum for fast neutrons of 232Th is on the order of around 0.2 barns. The fertile nuclear fission fuel material 16 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. In some applications it may be desirable to soften the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. For example, in such applications the reactivity control assembly 10 may be used to help control reactivity in a thermal nuclear fission reactor, such as without limitation a pressurized water reactor. As another example, in some other applications the reactivity control assembly 10 may be used to help control reactivity in a fast nuclear fission reactor in which it is desired to soften the neutron spectrum to reduce irradiation damage. To that end and referring now to FIGS. 1B-1G, in some embodiments the reactivity control rod 12 may also include neutron moderating material 20 in addition to the fertile nuclear fission fuel material 16. The neutron moderating material 20 may include any suitable neutron moderating material as desired for a particular application. Given by way of non-limiting example, the neutron moderating material 20 may include any one or more of hydrogen, deuterium, helium, lithium, boron, carbon, graphite, sodium, lead, and the like. When the neutron moderating material 20 is provided, the neutron moderating material 20 may be distributed within the reactivity control rod 12 in any manner as desired for a particular application. For example and as shown in FIGS. 1B-1F by way of illustration and not of limitation, in some embodiments the neutron moderating material 20 may be substantially heterogeneously distributed within the reactivity control rod 12. Given by way of non-limiting examples, the neutron moderating material 20 may be heterogeneously distributed in disks 21 (FIGS. 1B and 1C). The disks 21 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 12 (as shown in FIG. 1B) or substantially transverse to the axial orientation of the reactivity control rod 12 (as shown in FIG. 1C). Given by way of further non-limiting examples, the neutron moderating material 20 may be heterogeneously distributed toward ends of the reactivity control rod 12 (as shown in FIG. 1D) or toward a middle of the reactivity control rod 12 (as shown in FIG. 1E). Given by way of a further non-limiting example, the neutron moderating material 20 may be provided as a rod follower 23 (as shown in FIG. 1F). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 1G, the neutron moderating material 20 may be substantially homogeneously distributed within the reactivity control rod 12. Referring now to FIGS. 1H-1M, in some embodiments the neutron absorbing material 14 may also include neutron absorbing poison 22 in addition to the fertile nuclear fission fuel material 16. The neutron absorbing poison 22 may include any suitable neutron absorbing poison as desired. For example and given by way of non-limiting examples, the neutron absorbing poison 22 may include any one or more of silver, indium, cadmium, gadolinium, hafnium, lithium, 3He, fission products, protactinium, neptunium, boron, and the like. The neutron absorbing poison 22 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. When the neutron absorbing poison 22 is provided, the neutron absorbing poison 22 may be distributed within the reactivity control rod 12 in any manner as desired for a particular application. For example and as shown in FIGS. 1H-1L by way of illustration and not of limitation, in some embodiments the neutron absorbing poison 22 may be substantially heterogeneously distributed within the reactivity control rod 12. Given by way of non-limiting examples, the neutron absorbing poison 22 may be heterogeneously distributed in disks 25 (FIGS. 1H and 1I). The disks 25 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 12 (as shown in FIG. 1H) or substantially transverse to the axial orientation of the reactivity control rod 12 (as shown in FIG. 1I). Given by way of further non-limiting examples, the neutron absorbing poison 22 may be heterogeneously distributed toward ends of the reactivity control rod 12 (as shown in FIG. 1J) or toward a middle of the reactivity control rod 12 (as shown in FIG. 1K). Given by way of a further non-limiting example, the neutron absorbing poison 22 may be provided as a rod follower 27 (as shown in FIG. 1L). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 1M, the neutron absorbing poison 22 may be substantially homogeneously distributed within the reactivity control rod 12. In some embodiments and referring now to FIGS. 1H-1P, an effect on reactivity achievable by the fertile nuclear fission fuel material 16 may equalized toward an effect on reactivity achievable by portions of the neutron absorbing poison 22. For example, such equalization may be desirable to mitigate localized flux peaking. It will be appreciated that such equalization may be effected regardless of whether the fertile nuclear fission fuel material 16 is distributed heterogeneously or homogeneously and regardless of whether the neutron absorbing poison 22 is distributed heterogeneously (FIGS. 1H-1L and FIGS. 1O-1P) or homogeneously (FIG. 1M). In some other embodiments and still referring to FIGS. 1H-1P, reactivity effect of the fertile nuclear fission fuel material 16 and reactivity effect of the neutron absorbing poison 22 may be locally tailored as desired for a particular application. For example, in some embodiments and as shown generally in FIG. 1N the reactivity control rod 12 has a region 24 and a region 26. It will be appreciated that the regions 24 and 26 may be located anywhere within the reactivity control rod 12 as desired. No limitation is implied, and is not to be inferred, by virtue of appearance in the drawings which are provided for illustration purposes only. A concentration 28 of the neutron absorbing poison 22 is disposed in the region 24 and a concentration 30 of the neutron absorbing poison 22 is disposed in the region 26. A concentration 32 of the fertile nuclear fission fuel material 16 is disposed in the region 24 and a concentration 34 of the fertile nuclear fission fuel material 16 is disposed in the region 26. It will be appreciated that concentration may be determined per volume basis, per area basis, or per length basis, as desired. It will be appreciated that reactivity effects of the concentrations 28 and 30 of the neutron absorbing poison 22 and reactivity effects of the concentrations 32 and 34 of the fertile nuclear fission fuel material 16 may be tailored as desired for a particular application. For example, in some embodiments and as shown in FIGS. 1H-1P a reactivity effect of the concentration 30 of the neutron absorbing poison 22 may be substantially equalized with a reactivity effect of the concentration 32 of the fertile nuclear fission fuel material 16. In some other embodiments and as also shown in FIGS. 1H-1P a reactivity effect of the concentration 28 of the neutron absorbing poison 22 may be substantially equalized with a reactivity effect of the concentration 34 of the fertile nuclear fission fuel material 16. In some other embodiments and as shown in FIGS. 1H-1P a reactivity effect of the concentration 30 of the neutron absorbing poison 22 may be different from a reactivity effect of the concentration 32 of the fertile nuclear fission fuel material 16. In other embodiments a reactivity effect of the concentration 28 of the neutron absorbing poison 22 may be different from a reactivity effect of the concentration 34 of the fertile nuclear fission fuel material 16. Other reactivity effects may be effected as desired. For example and as shown in FIGS. 1H-1P, in some embodiments a sum of reactivity effects of the concentration 28 of the neutron absorbing poison 22 and the concentration 32 of the fertile nuclear fission fuel material 16 may be substantially equalized toward a sum of reactivity effects of the concentration 30 of the neutron absorbing poison 22 and the concentration 34 of the fertile nuclear fission fuel material 16. In some other embodiments, reactivity effect is substantially constant between the region 24 and the region 26. If desired, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may vary. For example and as shown in FIGS. 1O and 1P, in some embodiments concentration the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a continuous gradient. Given by way of non-limiting example, as shown in FIG. 1O the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in wedges 36 and 38, respectively, that abut each other along their hypotenuse 40. Given by way of another non-limiting example, as shown in FIG. 1P the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in mated frustoconical sections 42 and 44, respectively. It will be appreciated that the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be provided in other suitable arrangements in which their concentrations change along a continuous gradient, and arrangements are not to be limited to those shown in FIGS. 1G and 1H by way of illustration and not of limitation. In some other embodiments, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a non-continuous gradient. For example, concentration of the fertile nuclear fission fuel material 16 and/or the neutron absorbing poison 22 may change along a non-continuous gradient as a result of heterogeneous distribution as shown in FIGS. 1H-1L. In such cases, concentration of the neutron absorbing poison 22 can vary along a non-contiguous gradient because the neutron absorbing poison 22 is provided in discrete locations (as opposed to homogeneous distribution). Also in such cases, concentration of the fertile nuclear fission fuel material 16 can vary along a non-contiguous gradient because the fertile nuclear fission fuel material 16 is provided in discrete locations that are separated from each other by the discrete locations of the neutron absorbing poison 22. In some embodiments the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be spatially fixed relative to each other. That is, in such arrangements the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 do not physically move in relation to each other. However, in some other embodiments the fertile nuclear fission fuel material 16 and the neutron absorbing poison 22 may be spatially movable relative to each other. Given by way of non-limiting example and referring briefly to FIGS. 1H-1L and 1O-1P, any one or more of the discrete locations of the neutron absorbing poison 22, such as without limitation the disks 25, may be slidably received in the reactivity control rod 12 and may be moved in and out of the reactivity control rod 12 as desired by a suitable mechanism, such as a control rod drive mechanism (not shown) or the like. The sensor 18 may be physically associated with the reactivity control rod 12 in any suitable physical association as desired. For example, referring now to FIGS. 1A-1P and also to FIG. 1Q, in some embodiments physical association may include the sensor 18 being located within an interior 46 of the reactivity control rod 12. For example, the sensor 18 may be located via any suitable attachment method on an interior surface 48 of a cladding wall 50 of the reactivity control rod 12. As a further example and referring now to FIGS. 1A-1P and also to FIG. 1R, in some other embodiments physical association may include the sensor 18 being located proximate an exterior 52 of the reactivity control rod 12. For example, the sensor 18 may be located via any suitable method on an exterior surface 54 of the cladding wall 50. Any one or more of various reactivity parameters associated with the reactivity control rod 12 may be sensed with the sensor 18. Given by way of non-limiting examples, the sensed reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. The sensor 18 may include any suitable sensor that is configured to sense the reactivity parameter that is desired to be sensed. Given by way of non-limiting example, in some embodiments the sensor 18 may include at least one fission detector, such as without limitation a micro-pocket fission detector. In some other embodiments the sensor 18 may include a neutron flux monitor, such as without limitation a fission chamber and/or an ion chamber. In some embodiments the sensor 18 may include a neutron fluence sensor, such as without limitation an integrating diamond sensor. In some embodiments the sensor 18 may include a fission product detector, such as without limitation a gas detector, a β detector, and/or a γ detector. In some embodiments, when provided, the fission product detector may configured to measure a ratio of isotope types in fission product gas. In some embodiments the sensor 18 may include a temperature sensor. In some other embodiments the sensor 18 may include a pressure sensor. In some embodiments the sensor 18 may include a power sensor, such as without limitation a power range nuclear instrument. In some embodiments, if desired the sensor 18 may be replaceable. In some applications it may be desirable to mitigate effects of internal pressure within the reactivity control rod 12 exerted by fission products, such as fission product gases. In such cases and referring now to FIG. 1S, in some embodiments the reactivity control rod 12 may define at least one chamber 56 configured to accumulate fission products. For example, when provided the chamber 56 may include a plenum 58. In some embodiments the plenum 58 may be located at least one mean free path λT for fission-inducing neutrons from the fertile nuclear fission fuel material 16. In some embodiments a backflow prevention device 60, such as a check valve like a ball check vale or the like, may be provided to help prevent re-entry into the reactivity control rod 12 of fission product gases that have outgassed from the reactivity control rod 12. Referring now to FIG. 1T, in some embodiments a calibration device 62 configured to calibrate the sensor 18 may be provided. It will be appreciated that, when provided, the calibration device 62 suitably is a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensor 18. Referring now to FIG. 1U, in some embodiments at least one communications device 64 may be operatively coupled to the sensor 18 as generally indicated at 66. The communications device 18 suitably is any acceptable device that can operatively couple the sensor 18 in signal communication with a suitable communications receiving device (not shown) as generally indicated at 68. Given by way of non-limiting examples, the communications device 64 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. Illustrative Reactivity Control System Referring now to FIG. 2A, an illustrative reactivity control system 210 is provided for a nuclear fission reactor (not shown) having a fast neutron spectrum. Given by way of overview, the reactivity control system 210 includes a reactivity control rod 212. The reactivity control rod 212 includes neutron absorbing material 214 configured to absorb fast spectrum neutrons. At least a portion of the neutron absorbing material 214 includes fertile nuclear fission fuel material 216. An actuator 217 is responsive to at least one reactivity parameter and is operationally coupled, as indicated generally at 219, to the reactivity control rod 212. Illustrative details will be set forth below by way of non-limiting examples. The actuator 217 may be responsive to any one or more of various reactivity parameters as desired for a particular application. In some embodiments, the reactivity parameter may include any one or more reactivity parameter of the nuclear fission reactor. In some other embodiments the reactivity parameter may include any one or more reactivity parameter of the reactivity control rod 212. Given by way of non-limiting examples, the reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and neutron spectrum. As mentioned above, the nuclear fission reactor (not shown) has a fast neutron spectrum. In some embodiments the nuclear fission reactor may include a traveling wave reactor, in which case the fast spectrum neutrons may be part of a nuclear fission traveling wave. In some other embodiments the nuclear fission reactor may include a fast breeder reactor, like a liquid metal fast breeder reactor or a gas-cooled fast breeder reactor, or the like. In some embodiments the neutron absorbing material 214 may be configured to reduce moderation of neutrons. For example, the neutron absorbing material 14 may have a suitably large atomic mass that can help reduce the amount of slowing down of fast spectrum neutrons. As such, a reduction may be made in softening of the neutron spectrum from the fast neutron spectrum toward neutron spectrums having neutron energy levels less than around 0.1 MeV, such as an epi-thermal neutron spectrum or a thermal neutron spectrum. It will be appreciated that, given by way of non-limiting examples, elements of the actinide series, such as without limitation uranium and thorium, present a sufficiently large atomic mass to help reduce moderation of neutrons. The fertile nuclear fission fuel material 216, that is included in the neutron absorbing material 214, can include any type of fertile nuclear fission fuel material as desired for a particular application. For example, in some embodiments the fertile nuclear fission fuel material 216 may include uranium, such as 238U. In some other embodiments the fertile nuclear fission fuel material 16 may include thorium, such as 232Th. The fertile nuclear fission fuel material 16 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. In some applications it may be desirable to soften the neutron spectrum within the fast neutron spectrum toward a softer neutron spectrum that is still within the fast neutron spectrum—that is, at least around 0.1 MeV. For example, in some applications it may be desired to soften the neutron spectrum to reduce irradiation damage. To that end and referring now to FIGS. 2B-2G, in some embodiments the reactivity control rod 212 may also include neutron moderating material 220 in addition to the fertile nuclear fission fuel material 216. The neutron moderating material 220 may include any suitable neutron moderating material as desired for a particular application. Given by way of non-limiting example, the neutron moderating material 220 may include any one or more of hydrogen, deuterium, helium, lithium, boron, carbon, graphite, sodium, lead, and the like. When the neutron moderating material 220 is provided, the neutron moderating material 220 may be distributed within the reactivity control rod 212 in any manner as desired for a particular application. For example and as shown in FIGS. 2B-2F by way of illustration and not of limitation, in some embodiments the neutron moderating material 220 may be substantially heterogeneously distributed within the reactivity control rod 212. Given by way of non-limiting examples, the neutron moderating material 220 may be heterogeneously distributed in disks 221 (FIGS. 2B and 2C). The disks 221 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 212 (as shown in FIG. 2B) or substantially transverse to the axial orientation of the reactivity control rod 212 (as shown in FIG. 2C). Given by way of further non-limiting examples, the neutron moderating material 220 may be heterogeneously distributed toward ends of the reactivity control rod 212 (as shown in FIG. 2D) or toward a middle of the reactivity control rod 212 (as shown in FIG. 2E). Given by way of a further non-limiting example, the neutron moderating material 220 may be provided as a rod follower 223 (as shown in FIG. 2F). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 2G, the neutron moderating material 220 may be substantially homogeneously distributed within the reactivity control rod 212. Referring now to FIGS. 2H-2M, in some embodiments the neutron absorbing material 214 may also include neutron absorbing poison 222 in addition to the fertile nuclear fission fuel material 216. The neutron absorbing poison 222 may include any suitable neutron absorbing poison as desired. For example and given by way of non-limiting examples, the neutron absorbing poison 222 may include any one or more of silver, indium, cadmium, gadolinium, hafnium, lithium, 3He, fission products, protactinium, neptunium, boron, and the like. The neutron absorbing poison 222 may be provided in any suitable form as desired, such as without limitation powdered form, discrete particle form like beads or pellets, or any other form as desired. When the neutron absorbing poison 222 is provided, the neutron absorbing poison 222 may be distributed within the reactivity control rod 212 in any manner as desired for a particular application. For example and as shown in FIGS. 2H-2L by way of illustration and not of limitation, in some embodiments the neutron absorbing poison 222 may be substantially heterogeneously distributed within the reactivity control rod 212. Given by way of non-limiting examples, the neutron absorbing poison 222 may be heterogeneously distributed in disks 225 (FIGS. 2H and 2I). The disks 225 may be oriented substantially coaxially with an axial orientation of the reactivity control rod 212 (as shown in FIG. 2H) or substantially transverse to the axial orientation of the reactivity control rod 212 (as shown in FIG. 2I). Given by way of further non-limiting examples, the neutron absorbing poison 222 may be heterogeneously distributed toward ends of the reactivity control rod 212 (as shown in FIG. 2J) or toward a middle of the reactivity control rod 212 (as shown in FIG. 2K). Given by way of a further non-limiting example, the neutron absorbing poison 222 may be provided as a rod follower 227 (as shown in FIG. 2L). It will be appreciated that any heterogeneous distribution may be used as desired. No particular heterogeneous distribution is intended to be implied by way of illustration and none should be inferred. In some other embodiments and as shown in FIG. 2M, the neutron absorbing poison 222 may be substantially homogeneously distributed within the reactivity control rod 212. In some embodiments and referring now to FIGS. 2H-2P, an effect on reactivity achievable by the fertile nuclear fission fuel material 216 may equalized toward an effect on reactivity achievable by portions of the neutron absorbing poison 222. For example, such equalization may be desirable to mitigate localized flux peaking. It will be appreciated that such equalization may be effected regardless of whether the fertile nuclear fission fuel material 216 is distributed heterogeneously or homogeneously and regardless of whether the neutron absorbing poison 222 is distributed heterogeneously (FIGS. 2H-2L and FIGS. 2O-2P) or homogeneously (FIG. 2M). In some other embodiments and still referring to FIGS. 2H-2P, reactivity effect of the fertile nuclear fission fuel material 216 and reactivity effect of the neutron absorbing poison 222 may be locally tailored as desired for a particular application. For example, in some embodiments and as shown generally in FIG. 2N the reactivity control rod 212 has a region 224 and a region 226. It will be appreciated that the regions 224 and 226 may be located anywhere within the reactivity control rod 212 as desired. No limitation is implied, and is not to be inferred, by virtue of appearance in the drawings which are provided for illustration purposes only. A concentration 228 of the neutron absorbing poison 222 is disposed in the region 224 and a concentration 230 of the neutron absorbing poison 222 is disposed in the region 226. A concentration 232 of the fertile nuclear fission fuel material 216 is disposed in the region 224 and a concentration 234 of the fertile nuclear fission fuel material 216 is disposed in the region 226. It will be appreciated that concentration may be determined per volume basis, per area basis, or per length basis, as desired. It will be appreciated that reactivity effects of the concentrations 228 and 230 of the neutron absorbing poison 222 and reactivity effects of the concentrations 232 and 234 of the fertile nuclear fission fuel material 216 may be tailored as desired for a particular application. For example, in some embodiments and as shown in FIGS. 2H-2P a reactivity effect of the concentration 230 of the neutron absorbing poison 222 may be substantially equalized with a reactivity effect of the concentration 232 of the fertile nuclear fission fuel material 216. In some other embodiments and as also shown in FIGS. 2H-2P a reactivity effect of the concentration 228 of the neutron absorbing poison 222 may be substantially equalized with a reactivity effect of the concentration 234 of the fertile nuclear fission fuel material 216. In some other embodiments and as shown in FIGS. 2H-2P a reactivity effect of the concentration 230 of the neutron absorbing poison 222 may be different from a reactivity effect of the concentration 232 of the fertile nuclear fission fuel material 216. In other embodiments a reactivity effect of the concentration 228 of the neutron absorbing poison 222 may be different from a reactivity effect of the concentration 234 of the fertile nuclear fission fuel material 216. Other reactivity effects may be affected as desired. For example and as shown in FIGS. 2H-2P, in some embodiments a sum of reactivity effects of the concentration 228 of the neutron absorbing poison 222 and the concentration 232 of the fertile nuclear fission fuel material 216 may be substantially equalized toward a sum of reactivity effects of the concentration 230 of the neutron absorbing poison 222 and the concentration 234 of the fertile nuclear fission fuel material 216. In some other embodiments, reactivity effect is substantially constant between the region 224 and the region 226. If desired, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may vary. For example and as shown in FIGS. 2O and 2P, in some embodiments concentration the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a continuous gradient. Given by way of non-limiting example, as shown in FIG. 2O the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in wedges 236 and 238, respectively, that abut each other along their hypotenuse 240. Given by way of another non-limiting example, as shown in FIG. 2P the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in mated frustoconical sections 242 and 244, respectively. It will be appreciated that the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be provided in other suitable arrangements in which their concentrations change along a continuous gradient, and arrangements are not to be limited to those shown in FIGS. 2G and 2H by way of illustration and not of limitation. In some other embodiments, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a non-continuous gradient. For example, concentration of the fertile nuclear fission fuel material 216 and/or the neutron absorbing poison 222 may change along a non-continuous gradient as a result of heterogeneous distribution as shown in FIGS. 2H-2L. In such cases, concentration of the neutron absorbing poison 222 can vary along a non-contiguous gradient because the neutron absorbing poison 222 is provided in discrete locations (as opposed to homogeneous distribution). Also in such cases, concentration of the fertile nuclear fission fuel material 216 can vary along a non-contiguous gradient because the fertile nuclear fission fuel material 216 is provided in discrete locations that are separated from each other by the discrete locations of the neutron absorbing poison 222. In some embodiments the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be spatially fixed relative to each other. That is, in such arrangements the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 do not physically move in relation to each other. However, in some other embodiments the fertile nuclear fission fuel material 216 and the neutron absorbing poison 222 may be spatially movable relative to each other. Given by way of non-limiting example and referring briefly to FIGS. 2H-2L and 2O-2P, any one or more of the discrete locations of the neutron absorbing poison 222, such as without limitation the disks 225, may be slidably received in the reactivity control rod 212 and may be moved in and out of the reactivity control rod 212 as desired by a suitable mechanism, such as a control rod drive mechanism or the like. Referring now to FIG. 2Q, in some embodiments the reactivity control rod 212 may define at least one chamber 256 configured to accumulate fission products. For example, when provided the chamber 256 may include a plenum 258. In some embodiments the plenum 258 may be located at least one mean free path λT for fission-inducing neutrons from the fertile nuclear fission fuel material 216. In some embodiments a backflow prevention device 260, such as a check valve like a ball check vale or the like, may be provided to help prevent re-entry into the reactivity control rod 212 of fission product gases that have outgassed from the reactivity control rod 212. As mentioned above, the actuator 217 is responsive to at least one reactivity parameter. In some embodiments, the reactivity control system 210 may include an apparatus configured to determine the reactivity parameter. Given by way of non-limiting examples and referring now to FIGS. 2R-2AL, the apparatus may include at least one sensor 218. As shown in FIGS. 2R-2AL, in some embodiments the sensor 218 may be physically associated with the reactivity control rod 210. Given by way of non-limiting examples, in FIGS. 2R-2AH the sensor 218 may be physically associated with embodiments of the reactivity control rod 210 that have been shown and explained with reference to FIGS. 2A-2Q. In such cases, details have already been set forth regarding embodiments of the reactivity control rod 210 with reference to FIGS. 2A-2Q and need not be repeated for an understanding. In such embodiments the sensor 218 may be physically associated with the reactivity control rod 212 in any suitable physical association as desired. For example and referring to FIG. 2AI, in some embodiments physical association may include the sensor 218 being located within an interior 246 of the reactivity control rod 212. For example, the sensor 218 may be located via any suitable attachment method on an interior surface 248 of a cladding wall 250 of the reactivity control rod 212. As a further example and referring now to FIG. 2AJ, in some other embodiments physical association may include the sensor 218 being located proximate an exterior 252 of the reactivity control rod 212. For example, the sensor 218 may be located via any suitable method on an exterior surface 254 of the cladding wall 250. It will be appreciated that the sensor 218 need not be physically associated with the reactivity control rod 212. To that end, in some embodiments, the sensor 218 is not physically associated with the reactivity control rod 212. For example, in some embodiments the sensor 218 may be located at a position that is separate from the reactivity control rod 212 but that permits the sensor 218 to sense the reactivity parameter desired to be sensed. Given by way of non-limiting example, the sensor 218 may be located at a position that is separate but no more than one mean free path λT for fission-inducing neutrons from the reactivity control rod 212. Any one or more of various reactivity parameters associated with the reactivity control rod 212 may be sensed with the sensor 218. Given by way of non-limiting examples, the sensed reactivity parameter may include any one or more of parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. The sensor 218 may include any suitable sensor that is configured to sense the reactivity parameter that is desired to be sensed. Given by way of non-limiting example, in some embodiments the sensor 218 may include at least one fission detector, such as without limitation a micro-pocket fission detector. In some other embodiments the sensor 218 may include a neutron flux monitor, such as without limitation a fission chamber and/or an ion chamber. In some embodiments the sensor 218 may include a neutron fluence sensor, such as without limitation an integrating diamond sensor. In some embodiments the sensor 218 may include a fission product detector, such as without limitation a gas detector, a β detector, and/or a γ detector. In some embodiments, when provided, the fission product detector may be configured to measure a ratio of isotope types in fission product gas. In some embodiments the sensor 18 may include a temperature sensor. In some other embodiments the sensor 218 may include a pressure sensor. In some embodiments the sensor 218 may include a power sensor, such as without limitation a power range nuclear instrument. In some embodiments, if desired the sensor 218 may be replaceable. In some other embodiments, the reactivity parameter may be determined without being sensed by a sensor. Given by way of non-limiting example, in some embodiments the apparatus may include electrical circuitry (not shown) configured to determine at least one reactivity parameter (which have been discussed above). The reactivity parameter may be determined in any suitable manner. Given by way of non-limiting example, the reactivity parameter may be retrieved from a look-up table using operating parameters, such as temperature, pressure, power level, time in core life (as measured in effective full power hours), and the like, as entering arguments. Given by way of another non-limiting example, the reactivity parameter may be modeled, such as by running suitable neutronics modeling software on a suitable computer. Given by way of illustration, suitable neutronics modeling software includes MCNP, CINDER, REBUS, and the like. In a further non-limiting example, the reactivity parameter may be determined by a reactor operator or any other person skilled in the art based on prior knowledge or experience. In a general sense, those skilled in the art will recognize that various aspects described herein (including the electrical circuitry configured to determine at least one reactivity parameter) can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof that can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Referring to FIG. 2AK, in some embodiments a calibration device 262 configured to calibrate the sensor 218 may be provided. It will be appreciated that, when provided, the calibration device 262 suitably is a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensor 218. Referring to FIG. 2AL, in some embodiments at least one communications device 264 may be operatively coupled to the sensor 218 as generally indicated at 266. The communications device 218 suitably is any acceptable device that can operatively couple the sensor 218 in signal communication with a suitable communications receiving device (not shown) as generally indicated at 268. Given by way of non-limiting examples, the communications device 264 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. Referring now to FIGS. 2A-2AL, the reactivity control rod 212 is operationally coupled, as indicated generally at 219, to the actuator 217 in any suitable manner as desired. For example, in some embodiments the reactivity control rod 212 may be electromagnetically coupled to the actuator 217. In some other embodiments the reactivity control rod 212 may be mechanically linked to the actuator 217. Referring to FIG. 2AM, in some embodiments the reactivity control system 210 may include an actuator controller 270 that is configured to generate a rod control signal 272. In such embodiments, the actuator 217 is configured to move the reactivity control rod 217 that is operationally coupled thereto (as generally indicated at 219) responsive to the rod control signal 272. The actuator controller 270 generates the rod control signal 272 and communicates the rod control signal 272 in signal communication to the actuator 217. Referring to FIG. 2AN, in some embodiments a communications device 274 is configured to communicate the rod control signal 272 from the actuator controller 270 to the actuator 217. The communications device 274 suitably is any acceptable device that can operatively couple the actuator controller 270 in signal communication with the actuator 217. Given by way of non-limiting examples, the communications device 274 may include an electrical cable, a fiber optic cable, a telemetry transmitter, a radiofrequency transmitter, an optical transmitter, or the like. The actuator controller 270 may generate the rod control signal 272 in any suitable manner as desired. For example and referring to FIG. 2AO, in some embodiments the actuator controller 270 may include an operator interface 276. Given by way of non-limiting example, in some embodiments the operator interface 276 may include a shim switch. Referring to FIG. 2AP, in some other embodiments the actuator controller 270 may include electrical circuitry 278 that is configured to automatically generate the rod control signal 272 based upon at least one reactivity parameter (which have been discussed above). Referring now to FIGS. 2A-2AP, the actuator 217 may be any suitable actuator as desired for a particular application. Given by way of non-limiting example, in some embodiments the actuator 217 may include a reactivity control rod drive mechanism. In some embodiments the actuator 217 may be configured to drive the reactivity control rod 212 bidirectionally. That is, when the reactivity control rod 212 is provided for use in a nuclear fission reactor, the reactivity control rod 212 may be driven into and/or out of a core of the nuclear fission reactor as desired. In some other embodiments, the actuator 217 may be further configured to stop driving the reactivity control rod 217 at least one intermediate position between a first stop position and a second stop position. Illustrative Nuclear Fission Traveling Wave Reactor Referring now to FIG. 3, in some embodiments an illustrative nuclear fission traveling wave reactor 300 having a fast neutron spectrum may include any illustrative embodiment of the reactivity control system 210 (FIGS. 2A-2AP). Given by way of non-limiting example, the nuclear fission traveling wave reactor 300 includes an illustrative nuclear fission reactor core 331. The nuclear fission reactor core 331 includes suitable nuclear fission fuel material 333 that is configured to propagate therein a nuclear fission traveling wave having a fast neutron spectrum. As described above, the reactivity control system 210 includes reactivity control rods 212. Each reactivity control rod 212 includes neutron absorbing material configured to absorb fast spectrum neutrons of the nuclear fission traveling wave. At least a portion of the neutron absorbing material includes fertile nuclear fission fuel material. The reactivity control system 210 also includes actuators 217. Each of the actuators 217 is responsive to at least one reactivity parameter and is operationally coupled to at least one of the reactivity control rods 212, as indicated generally at 219. In some embodiments, the reactivity parameter may include at least one reactivity parameter of the nuclear fission traveling wave reactor. However, in some other embodiments and as discussed above, the reactivity parameter may include at least one reactivity parameter of at least one of the reactivity control rods 212. In various embodiments the reactivity parameter may include one or more reactivity parameters such as neutron fluence, neutron flux, neutron fissions, fission products, radioactive decay events, temperature, pressure, power, isotopic concentration, burnup, and/or neutron spectrum. It will be appreciated that the reactivity control system 210 included in the nuclear fission traveling wave reactor 300 may be embodied in any manner desired as discussed above. For example, the reactivity control system and any of its components may be embodied, without limitation, as discussed above with reference to any one or more of FIGS. 2A-2AP. Because embodiments of the reactivity control system 210 have been discussed in detail above, for sake of brevity details need not be repeated for an understanding. Illustrative details of embodiments of the nuclear fission traveling wave reactor 300 will be set forth below. It will be appreciated that the nuclear fission traveling wave reactor 300 is a non-limiting example that is set forth below for purposes of illustration and not of limitation. The nuclear fission reactor core 333 is housed within an illustrative reactor core enclosure 335 which acts to maintain vertical coolant flow through the core. In some embodiments the reactor core enclosure 335 may also function as a radiation shield to protect in-pool components such as heat exchangers and the like from neutron bombardment. The reactivity control rods 212 longitudinally extend into the nuclear fission reactor core 331 for controlling the fission process occurring therein, as discussed above. The nuclear fission reactor core 331 is disposed within an illustrative reactor vessel 337. In some embodiments the reactor vessel 337 is filled to a suitable amount (such as about 90% or so) with a pool of coolant 339, such as liquid metal like sodium, potassium, lithium, lead, mixtures thereof, and the like, or liquid metal alloys such as lead-bismuth, to such an extent that the nuclear fission reactor core 331 is submerged in the pool of coolant. Suitably, in an illustrative embodiment contemplated herein, the coolant is a liquid sodium (Na) metal or sodium metal mixture, such as sodium-potassium (Na—K). In addition, in some embodiments a containment vessel 341 sealingly surrounds parts of the nuclear fission traveling wave reactor 300. In some embodiments a primary coolant pipe 343 is coupled to the nuclear fission reactor core 331 for allowing a suitable coolant to flow through the nuclear fission reactor core 331 along a coolant flow stream or flow path 345 in order to cool the nuclear fission reactor core 331. In various embodiments the primary coolant pipe 343 may be made from materials such as, without limitation, stainless steel or from non-ferrous alloys, zirconium-based alloys, or other suitable structural materials or composites. In some embodiments the heat-bearing coolant generated by the nuclear fission reactor core 331 flows along the coolant flow path 345 to an intermediate heat exchanger 347 that is also submerged in the pool of coolant 339. The intermediate heat exchanger 347 may be made from any suitable material, such as without limitation stainless steel, that is sufficiently resistant to heat and corrosive effects of the coolant, such as without limitation liquid sodium, in the pool of coolant 339. The coolant flowing along the coolant flow path 345 flows through the intermediate heat exchanger 347 and continues through the primary coolant pipe 343. It will be appreciated that the coolant leaving intermediate heat exchanger 347 has been cooled due to heat transfer occurring in the intermediate heat exchanger 347. In some embodiments a pump 349, which may be an electro-mechanical pump or an electromagnetic pump as desired, is coupled to the primary coolant pipe 343. In such embodiments the pump 349 is in fluid communication with the coolant carried by the primary coolant pipe 343. The pump 349 pumps the coolant through the primary coolant pipe 343, through the nuclear fission reactor core 331, along the coolant flow path 345, and into the intermediate heat exchanger 347. A secondary coolant pipe 351 is provided for removing heat from the intermediate heat exchanger 347. The secondary coolant pipe 351 includes a secondary hot leg pipe segment 353 and a secondary cold leg pipe segment 355. The secondary hot leg pipe segment 353 and the secondary cold leg pipe segment 355 are integrally connected to the intermediate heat exchanger 347. The secondary coolant pipe 351 contains a secondary coolant, that is a fluid such as any one of the coolant choices previously mentioned. The secondary hot leg pipe segment 353 extends from the intermediate heat exchanger 347 to a steam generator 357. In some embodiments, if desired, the steam generator 357 may include a superheater. After passing through the steam generator 357, the secondary coolant flowing through the secondary loop pipe 351 and exiting the steam generator 357 is at a lower temperature and enthalpy than before entering the steam generator 357 due to heat transfer occurring within the steam generator 357. After passing through the steam generator 357, the secondary coolant is pumped, such as by means of a pump 359, which may be an electro-mechanical pump or an electromagnetic pump or the like, along the secondary cold leg pipe segment 355, which extends into the intermediate heat exchanger 347 for providing the previously mentioned heat transfer. Disposed in the steam generator 357 is a body of water 361 having a predetermined temperature. The secondary coolant flowing through the secondary hot leg pipe segment 353 will transfer its heat by means of conduction and convection to the body of water 361, which is at a lower temperature than the secondary coolant flowing through the secondary hot leg pipe segment 353. As the secondary coolant flowing through the secondary hot leg pipe segment 353 transfers its heat to the body of water 361, a portion of the body of water 361 will vaporize to steam 363 according to the predetermined temperature within the steam generator 357. The steam 363 will then travel through a steam line 365. One end of the steam line 365 is in vapor communication with the steam 363 and another end of the steam line 365 is in liquid communication with the body of water 361. A rotatable turbine 367 is coupled to the steam line 365 such that the turbine 367 rotates as the steam 363 passes therethrough. An electrical generator 369 is coupled to the turbine 367 by a rotatable turbine shaft 371. The electrical generator 369 generates electricity as the turbine 367 rotates. A condenser 373 is coupled to the steam line 365 and receives the steam 363 passing through the turbine 367. The condenser 373 condenses the steam 363 to liquid water and passes any waste heat via a recirculation fluid path 375 and a condensate pump 377, such as an electro-mechanical pump, to a heat sink 379, such as a cooling tower, which is associated with the condenser 373. The feed water condensed by the condenser 373 is pumped along a feed water line 381 from the condenser 373 to the steam generator 357 by a feed water pump 383, which may be an electro-mechanical pump that is interposed between the condenser 373 and the steam generator 357. Embodiments of the nuclear fission reactor core 331 may include any suitable configuration as desired to accommodate the reactivity control system 210. In this regard, in some embodiments the nuclear fission reactor core 331 may be generally cylindrically shaped to obtain a generally circular transverse cross section. In some other embodiments the nuclear fission reactor core 331 may be generally hexagonally shaped to obtain a generally hexagonal transverse cross section. In other embodiments the nuclear fission reactor core 331 may be generally parallelepiped shaped to obtain a generally rectangular transverse cross section. Regardless of the configuration or shape selected for the nuclear fission reactor core 331, the nuclear fission reactor core 331 is operated as a traveling wave nuclear fission reactor core. For example, a nuclear fission igniter (not shown for clarity), which includes an isotopic enrichment of nuclear fissionable material, such as, without limitation, U-233, U-235 or Pu-239, is suitably located in the nuclear fission reactor core 331. Neutrons are released by the igniter. The neutrons that are released by the igniter are captured by fissile and/or fertile material within the nuclear fission fuel material 333 to initiate a nuclear fission chain reaction. The igniter may be removed once the fission chain reaction becomes self-sustaining, if desired. The igniter initiates a three-dimensional, traveling wave or “burn wave”. When the igniter generates neutrons to cause “ignition”, the burn wave travels outwardly from the igniter so as to form the traveling or propagating burn wave. Speed of the traveling burn wave may be constant or non-constant. Thus, the speed at which the burn wave propagates can be controlled. For example, longitudinal movement of the reactivity control rods 210 in a predetermined or programmed manner can drive down or lower neutronic reactivity of vented nuclear fission fuel modules 30. In this manner, neutronic reactivity of nuclear fuel that is presently being burned behind the burn wave or at the location of the burn wave is driven down or lowered relative to neutronic reactivity of “unburned” nuclear fuel ahead of the burn wave. Controlling reactivity in this manner maximizes the propagation rate of the burn wave subject to operating constraints for the nuclear fission reactor core 331, such as amount of permissible fission product production and/or neutron fluence limitations of reactor core structural materials. The basic principles of such a traveling wave nuclear fission reactor are disclosed in more detail in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006; and U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, and LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of which are hereby incorporated by reference. It will be appreciated that the embodiment of the nuclear fission traveling wave reactor 300 described above is set forth as a non-limiting example for purposes of illustration only and not of limitation. In some other embodiments, the nuclear fission traveling wave reactor 300 may be a gas-cooled fast nuclear fission traveling wave reactor that includes a suitable gas coolant, such as helium or the like. In such an embodiment, a gas-driven turbine-generator may be driven by the gas coolant. Illustrative Methods, Systems, and Computer Software Program Products Following are a series of flowcharts depicting implementations of processes. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an overall “big picture” viewpoint and thereafter the following flowcharts present alternate implementations and/or expansions of the “big picture” flowcharts as either sub-steps or additional steps building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an overall view and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Referring now to FIG. 4A, a method 400 is provided for controlling reactivity in a nuclear fission reactor having a fast neutron spectrum. The method 400 starts at a block 402. At a block 404 a desired reactivity parameter within a selected portion of the nuclear fission reactor having a fast neutron spectrum is determined. At a block 406 at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, is adjusted responsive to the desired reactivity parameter. The method 400 stops at a block 408. It will be appreciated that the method 400 may be performed with respect to any nuclear fission reactor having a fast neutron spectrum. In some embodiments, the method 400 may be performed with respect to a nuclear fission traveling wave reactor, in which case the fast spectrum neutrons may be part of a nuclear fission traveling wave. In some other embodiments, the method 400 may be performed with respect to any suitable fast breeder reactor, such as a liquid metal fast breeder reactor, a gas-cooled fast breeder reactor, or the like. Thus, no limitation to any particular type of nuclear fission reactor having a fast neutron spectrum is intended and should not be inferred. Illustrative details will be set forth below by way of non-limiting examples. In various embodiments the desired reactivity parameter may be determined with respect to any portion of a nuclear fission reactor as desired. For example and referring to FIG. 4B, in some embodiments determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the fertile nuclear fission fuel material at a block 410. In some other embodiments and referring to FIG. 4C, determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the at least one reactivity control rod at a block 412. In some other embodiments and referring to FIG. 4D, determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum at the block 404 may include determining at least one desired reactivity parameter of the nuclear fission reactor at a block 414. In some embodiments the reactivity control rod may be adjusted responsive to a difference between the desired reactivity parameter and a determination of the reactivity parameter. For example and referring to FIGS. 4A and 4E, in some embodiments at a block 416 at least one determined reactivity parameter may be determined. Referring additionally to FIG. 4F, in some embodiments at a block 418 a difference between the desired reactivity parameter and the at least one determined reactivity parameter may be determined. Referring additionally to FIG. 4G, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 406 may include adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter at a block 420. The determined reactivity parameter may be determined in any suitable manner as desired. For example and referring now to FIGS. 4E and 4H, in some embodiments determining at least one determined reactivity parameter at the block 416 may include predicting at least one reactivity parameter at a block 422. Referring to FIGS. 4E and 4I, in some embodiments determining at least one determined reactivity parameter at the block 416 may include modeling at least one reactivity parameter at a block 424. Referring to FIGS. 4E and 4J, in some embodiments determining at least one determined reactivity parameter at the block 416 may include selecting at least one predetermined reactivity parameter at a block 426. Referring to FIGS. 4E and 4K, in some other embodiments determining at least one determined reactivity parameter at the block 416 may include sensing at least one reactivity parameter at a block 428. It will be appreciated that any desired reactivity parameter may be sensed at the block 428 in any suitable manner. For example and referring to FIGS. 4K and 4L, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing a time history of at least one reactivity parameter at a block 430. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. Referring to FIGS. 4K and 4M, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing at least one radioactive decay event at a block 432. Referring to FIGS. 4K and 4N, in some embodiments sensing at least one reactivity parameter at the block 428 may include detecting fission at a block 434. Referring to FIGS. 4K and 4O, in some embodiments sensing at least one reactivity parameter at the block 428 may include monitoring neutron flux at a block 436. Referring to FIGS. 4K and 4P, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing neutron fluence at a block 438. Referring to FIGS. 4K and 4Q, in some embodiments sensing at least one reactivity parameter at the block 428 may include detecting fission products at a block 440. Referring to FIGS. 4K and 4R, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing temperature at a block 442. Referring to FIGS. 4K and 4S, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing pressure at a block 444. Referring to FIGS. 4K and 4T, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing power level at a block 446. Referring now to FIGS. 4A and 4U, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 406 may include moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at a block 448. In various embodiments the directions may include axial directions in the nuclear fission reactor, radial directions in the nuclear fission reactor, and/or lateral directions in the nuclear fission reactor. Referring now to FIGS. 4A and 4V, in some embodiments sensing at least one reactivity parameter at the block 428 may include sensing a difference in reactivity in association with a change in position of the reactivity control rod at a block 450. Referring to FIGS. 4A and 4W, in some embodiments a sensor that is configured to sense at least one reactivity parameter may be calibrated at a block 452. Referring now to FIG. 5A, a method 500 is provided for operating a nuclear fission traveling wave reactor having a fast neutron spectrum. The method 500 starts at a block 502. At a block 503 a nuclear fission traveling wave having a fast neutron spectrum is propagated in a nuclear fission traveling wave reactor core. At a block 504 a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor is determined. At a block 506 at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, is adjusted responsive to the desired reactivity parameter. The method 500 stops at a block 508. Illustrative details will be set forth below by way of non-limiting examples. Referring now to FIGS. 5A and 5B, in some embodiments a nuclear fission traveling wave having a fast neutron spectrum may be initiated in the nuclear fission traveling wave reactor core at a block 509. In various embodiments the desired reactivity parameter may be determined with respect to any portion of the nuclear fission traveling wave reactor as desired. For example and referring to FIG. 5C, in some embodiments determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the fertile nuclear fission fuel material at a block 510. In some other embodiments and referring to FIG. 5D, determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the at least one reactivity control rod at a block 512. In some other embodiments and referring to FIG. 5E, determining a desired reactivity parameter within a selected portion of the nuclear fission traveling wave reactor at the block 504 may include determining at least one desired reactivity parameter of the nuclear fission traveling wave reactor at a block 514. In some embodiments the reactivity control rod may be adjusted responsive to a difference between the desired reactivity parameter and a determination of the reactivity parameter. For example and referring to FIGS. 5A and 5F, in some embodiments at a block 516 at least one determined reactivity parameter may be determined. Referring additionally to FIG. 5G, in some embodiments at a block 518 a difference between the desired reactivity parameter and the at least one determined reactivity parameter may be determined. Referring additionally to FIG. 5H, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 506 may include adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter at a block 520. The determined reactivity parameter may be determined in any suitable manner as desired. For example and referring now to FIGS. 5F and 5I, in some embodiments determining at least one determined reactivity parameter at the block 516 may include predicting at least one reactivity parameter at a block 522. Referring to FIGS. 5F and 5J, in some embodiments determining at least one determined reactivity parameter at the block 516 may include modeling at least one reactivity parameter at a block 524. Referring to FIGS. 5F and 5K, in some embodiments determining at least one determined reactivity parameter at the block 516 may include selecting at least one predetermined reactivity parameter at a block 526. Referring to FIGS. 5F and 5L, in some other embodiments determining at least one determined reactivity parameter at the block 516 may include sensing at least one reactivity parameter at a block 528. It will be appreciated that any desired reactivity parameter may be sensed at the block 528 in any suitable manner. For example and referring to FIGS. 5L and 5M, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing a time history of at least one reactivity parameter at a block 530. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. Referring to FIGS. 5L and 5N, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing at least one radioactive decay event at a block 532. Referring to FIGS. 5L and 5O, in some embodiments sensing at least one reactivity parameter at the block 528 may include detecting fission at a block 534. Referring to FIGS. 5L and 5P, in some embodiments sensing at least one reactivity parameter at the block 528 may include monitoring neutron flux at a block 536. Referring to FIGS. 5L and 5Q, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing neutron fluence at a block 538. Referring to FIGS. 5L and 5R, in some embodiments sensing at least one reactivity parameter at the block 528 may include detecting fission products at a block 540. Referring to FIGS. 5L and 5S, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing temperature at a block 542. Referring to FIGS. 5L and 5T, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing pressure at a block 544. Referring to FIGS. 5L and 5U, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing power level at a block 546. Referring now to FIGS. 5A and 5V, in some embodiments adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at the block 506 may include moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter at a block 548. In various embodiments the directions may include axial directions in the nuclear fission traveling wave reactor, radial directions in the nuclear fission traveling wave reactor, and/or lateral directions in the nuclear fission traveling wave reactor. Referring now to FIGS. 5A and 5W, in some embodiments sensing at least one reactivity parameter at the block 528 may include sensing a difference in reactivity in association with a change in position of the reactivity control rod at a block 550. Referring to FIGS. 5A and 5X, in some embodiments a sensor that is configured to sense at least one reactivity parameter may be calibrated at a block 552. Referring now to FIG. 6A, an illustrative system 610 is provided for controlling reactivity in a nuclear fission reactor (not shown) having a fast neutron spectrum. The system 610 includes means 612 for determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum. The system 610 also includes means 614 for adjusting at least one reactivity control rod (not shown) having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter. In various embodiments the determining means 612 may include suitable electrical circuitry. As discussed above, various aspects described herein (including the means 612 for determining a desired reactivity parameter within a selected portion of a nuclear fission reactor having a fast neutron spectrum) can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof that can be viewed as being composed of various types of “electrical circuitry.” Consequently, it is emphasized that, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. In various embodiments the adjusting means 614 may include any suitable electro-mechanical system, such as without limitation an actuator. Given by way of illustration and not limitation, a non-limiting example of an actuator includes a control rod drive mechanism. However, it will be appreciated that, in a general sense, the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, and electro-magnetically actuated devices, or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment), and any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, as well as other systems such as motorized transport systems, factory automation systems, security systems, and communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In some embodiments the fast spectrum neutrons may be part of a nuclear fission traveling wave. In such cases, the nuclear fission reactor may include a nuclear fission traveling wave reactor. However, it will be appreciated that in other embodiments the fast spectrum neutrons need not be part of a nuclear fission traveling wave. Thus, in some embodiments, the nuclear fission reactor may include any suitable nuclear fission reactor having a fast neutron spectrum. Referring to FIG. 6B, in some embodiments the means 612 for determining a desired reactivity parameter may include means 616 for determining at least one desired reactivity parameter of the fertile nuclear fission fuel material. In some other embodiments and referring to FIG. 6C, the means 612 for determining a desired reactivity parameter may include means 618 for determining at least one desired reactivity parameter of the at least one reactivity control rod. In some other embodiments and referring to FIG. 6D, the means 612 for determining a desired reactivity parameter may include means 620 for determining at least one desired reactivity parameter of the nuclear fission reactor. The means 616, 618, and 620 may include suitable electrical circuitry, as described above. Referring now to FIG. 6E, in some embodiments the system 610 may also include means 622 for determining at least one determined reactivity parameter. In some embodiments the means 622 may include suitable electrical circuitry, as described above. Some other embodiments of the means 622 will be discussed below. Referring now to FIG. 6F, in some embodiments the system 610 may also include means 624 for determining a difference between the desired reactivity parameter and the at least one determined reactivity parameter. The means 624 may include suitable electrical circuitry, as described above, such as without limitation a comparator. Referring additionally to FIG. 6G, in some embodiments the adjusting means 614 may include means 626 for adjusting at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the difference between the desired reactivity parameter and the at least one determined reactivity parameter. The means 626 may include any suitable electro-mechanical system as described above, such as without limitation a control rod drive mechanism. In various embodiments the determining means 622 may determine a determined reactivity parameter in any manner as desired for a particular application. For example and referring to FIG. 6H, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 628 for predicting at least one reactivity parameter. The means 628 may include suitable electrical circuitry, as described above. Given by way of non-limiting example, the predetermined reactivity parameter may be retrieved from a look-up table using operating parameters, such as temperature, pressure, power level, time in core life (as measured in effective full power hours), and the like, as entering arguments. Referring to FIG. 6I, in some other embodiments the means 622 for determining at least one determined reactivity parameter may include means 630 for modeling at least one reactivity parameter. The means 630 may include suitable electrical circuitry, as described above, such as without limitation a suitable computer. The means 630 may also include suitable neutronics modeling software that runs on the electrical circuitry. Given by way of illustration, suitable neutronics modeling software includes MCNP, CINDER, REBUS, and the like. Referring to FIG. 6J, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 632 for selecting at least one predetermined reactivity parameter. The means 632 may include suitable electrical circuitry, as described above. Referring to FIG. 6K, in some embodiments the means 622 for determining at least one determined reactivity parameter may include means 634 for sensing at least one reactivity parameter. In various embodiments, the sensing means 634 may include any one or more of various sensors and detectors as desired for a particular purpose, as will be discussed below. Referring to FIG. 6L, in some embodiments the sensing means 634 may include means 636 for sensing a time history of at least one reactivity parameter. Sensing a time history may be performed as desired, such as by sensing and recording or storing the sensed reactivity parameter more than one time. Given by way of non-limiting examples, a time history of at least one reactivity parameter may include, without limitation, a rate of the reactivity parameter, accumulation of the reactivity parameter, total fissions, or the like. In various embodiments the means 636 may include suitable storage, such as computer memory media or computer memory storage or the like, configured to store values of the reactivity parameter over time. Referring to FIG. 6M, in some other embodiments the sensing means 634 may include 638 means for sensing at least one radioactive decay event. Given by way of non-limiting examples, the means 638 may include any one or more of suitable sensors or detectors for sensing α, β, and/or γ radiation as desired. Referring back to FIG. 6K, in various embodiments the sensing means 634 may include any suitable sensor as desired for a particular application. Given by way of illustrative examples and without limitation, in various embodiments the sensing means 634 may include any one or more sensor, such as at least one fission detector, a neutron flux monitor, a neutron fluence sensor, a fission product detector, a temperature sensor, a pressure sensor, and/or a power level sensor. Referring to FIG. 6N, in some embodiments the adjusting means 614 may include means 640 for moving, in at least one of two directions, at least one reactivity control rod having fast spectrum neutron absorbing material, at least a portion of the fast spectrum neutron absorbing material including fertile nuclear fission fuel material, responsive to the desired reactivity parameter. In some embodiments, the means 640 may include an actuator such as a control rod drive mechanism and/or a rod handling system. In various embodiments, the directions may include any one or more of axial directions in the nuclear fission reactor, radial directions in the nuclear fission reactor, and/or lateral directions in the nuclear fission reactor. Referring to FIG. 6O, in some embodiments the sensing means 634 may include means 642 for sensing a difference in reactivity in association with a change in position of the reactivity control rod. In various embodiments, the means 642 may include electrical circuitry, as described above. In some embodiments the electrical circuitry may implement a suitable comparator. Referring to FIG. 6P, in some embodiments the system 610 may also include means 644 for calibrating the sensing means 634. In various embodiments the calibration means 644 suitably includes a source having known characteristics or attributes of the reactivity parameter, discussed above, that is sensed by the sensing means 634. Referring now to FIG. 7A, a method 700 is provided for determining an application of a controllably movable rod. The method 700 starts at a block 702. At a block 704 at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor is determined, the controllably movable rod including fertile nuclear fission fuel material. At a block 706 an application of the controllably movable rod, chosen from a reactivity control rod and a nuclear fission fuel rod, is determined. The method 700 stops at a block 708. In various embodiments, the application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to the at least one determined reactivity parameter in the controllably movable rod. Non limiting examples given by way of illustration and not of limitation will be described below. Referring to FIG. 7B, in some embodiments at a block 710 the determined reactivity parameter and a target reactivity parameter may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined reactivity parameter and the target reactivity parameter. Referring back to FIG. 7A, in some embodiments the at least one reactivity parameter may include a neutron absorption coefficient. Referring to FIG. 7C, in some embodiments at a block 712 the determined neutron absorption coefficient a target neutron absorption coefficient may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined neutron absorption coefficient and the target neutron absorption coefficient. For example, a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron absorption coefficient is at least the target neutron absorption coefficient. As another example, a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron absorption coefficient is less than the target neutron absorption coefficient. Referring back to FIG. 7A, in some other embodiments the at least one reactivity parameter may include a neutron production coefficient. Referring to FIG. 7D, in some embodiments at a block 714 the determined neutron production coefficient and a target neutron production coefficient may be compared. In some embodiments, an application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to comparison of the determined neutron production coefficient and the target neutron production coefficient. For example, a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron production coefficient is at least the target neutron production coefficient. As another example, a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron production coefficient is less than the target neutron production coefficient. Referring back to FIG. 7A, the at least one reactivity parameter may be determined in any manner as desired for a particular application. Given by way of non-limiting examples, determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor may be based on neutron exposure history of the controllably movable rod, a property of fertile nuclear fission fuel material of the controllably movable rod, a property of fissile nuclear fission fuel material of the controllably movable rod, a property of neutron absorbing poison of the controllably movable rod, and/or a property of fission products of the controllably movable rod. Referring to FIG. 7E, in some embodiments determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 704 may include monitoring at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 716. Referring to FIG. 7F, in some other embodiments determining at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 704 may include predicting at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 718. Referring to FIG. 7G, in some embodiments predicting at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at the block 718 may include calculating at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor at a block 720. Referring now to FIG. 8A, a system 810 is provided for determining an application of a controllably movable rod. An apparatus 812 is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor, the controllably movable rod including fertile nuclear fission fuel material. Electrical circuitry 814 is configured to determine an application of the controllably movable rod chosen from a reactivity control rod and a nuclear fission fuel rod. In various embodiments, the application of the controllably movable rod (chosen from a reactivity control rod and a nuclear fission fuel rod) may be determined responsive to the at least one determined reactivity parameter in the controllably movable rod. Non limiting examples given by way of illustration and not of limitation will be described below. Referring to FIG. 8B, a comparator 816 may be configured to compare the determined reactivity parameter and a target reactivity parameter. In such a case, the electrical circuitry 814 may be responsive to the comparator 816. Still referring to FIG. 8B, in some embodiments the at least one reactivity parameter may include a neutron absorption coefficient. In such cases, the comparator 816 may be further configured to compare the determined neutron absorption coefficient with a target neutron absorption coefficient. The electrical circuitry 814 may be responsive to comparison of the determined neutron absorption coefficient and the target neutron absorption coefficient by the comparator 816. In some embodiments a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron absorption coefficient is at least the target neutron absorption coefficient. In some other embodiments a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron absorption coefficient is less than the target neutron absorption coefficient. Still referring to FIG. 8B, in some other embodiments the at least one reactivity parameter may include a neutron production coefficient. In such cases, the comparator 816 may be further configured to compare the determined neutron production coefficient with a target neutron production coefficient. The electrical circuitry 814 may be responsive to comparison of the determined neutron production coefficient and the target neutron production coefficient by the comparator 816. In some embodiments a chosen application of the controllably movable rod may include a nuclear fission fuel rod when the determined neutron production coefficient is at least the target neutron production coefficient. In some other embodiments a chosen application of the controllably movable rod may include a reactivity control rod when the determined neutron production coefficient is less than the target neutron production coefficient. Referring back to FIG. 8A, in various embodiments the apparatus 812 may be configured as desired to determine the reactivity parameter. For example and referring to FIG. 8C, in some embodiments the apparatus 812 may include electrical circuitry 818 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on neutron exposure history of the controllably movable rod. Referring to FIG. 8D, in some other embodiments the apparatus 812 may include electrical circuitry 820 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fertile nuclear fission fuel material of the controllably movable rod. Referring to FIG. 8E, in some embodiments the apparatus 812 may include electrical circuitry 822 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fissile nuclear fission fuel material of the controllably movable rod. Referring to FIG. 8F, in some embodiments the apparatus 812 may include electrical circuitry 824 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of neutron absorbing poison of the controllably movable rod. Referring to FIG. 8G, in some embodiments the apparatus 812 may include electrical circuitry 826 that is configured to determine at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor based on a property of fission products of the controllably movable rod. Referring to FIG. 8H, in some embodiments the apparatus 812 may include at least one monitoring device 828 that is configured to monitor at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. Given by way of non-limiting examples, the monitoring device 828 may include any one or more of the sensors and/or detectors described above. Referring to FIG. 8I, in some embodiments the apparatus 812 may include electrical circuitry 830 that is configured to predict at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. For example, in some embodiments the electrical circuitry 830 may be further configured to calculate at least one reactivity parameter of a controllably movable rod in a nuclear fission reactor. Referring to FIG. 9A, an illustrative method 900 is provided for operating a nuclear fission traveling wave reactor. The method 900 starts at a block 902. At a block 904 a reactivity control apparatus having a first worth is inserted into a first location of a reactor core of a nuclear fission traveling wave reactor. At a block 906, worth of the reactivity control apparatus is modified. At a block 908 the reactivity control apparatus is moved from the first location to a second location of the reactor core of the nuclear fission traveling wave reactor such that the reactivity control apparatus has a second worth that is different from the first worth. The method 900 stops at a block 910. Referring to FIG. 9B, in some embodiments modifying worth of the reactivity control apparatus at the block 906 may include absorbing neutrons by the reactivity control apparatus at a block 912. Referring to FIG. 9C, in some embodiments, absorbing neutrons by the reactivity control apparatus at the block 912 may include absorbing neutrons by fertile nuclear fission fuel material of the reactivity control apparatus at a block 914. In some of the cases, the second worth may be greater than the first worth. Referring to FIG. 9D, in some other embodiments modifying worth of the reactivity control apparatus at the block 906 may include modifying absorptive effect of self-shielded burnable poison of the reactivity control rod at a block 916. Referring to FIG. 9E, in some embodiments modifying absorptive effect of self-shielded burnable poison of the reactivity control rod at the block 916 may include modifying self-shielding effect of the self-shielded burnable poison at a block 918. Referring to FIG. 9F, in some embodiments modifying self-shielding effect of the self-shielded burnable poison at the block 918 may include modifying exposure of the self-shielded burnable poison to a neutron flux at a block 920. Referring to FIG. 9G, in some embodiments modifying exposure of the self-shielded burnable poison to a neutron flux at the block 920 may include modifying neutron energy at a block 922. In some embodiments the second worth may be less than the first worth. In some other embodiments the second worth may be greater than the first worth. In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, and electro-magnetically actuated devices, or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment), and any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, as well as other systems such as motorized transport systems, factory automation systems, security systems, and communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of random access memory), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, or optical-electrical equipment). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware and software implementations of aspects of systems; the use of hardware or software is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link, etc.). It will be appreciated that each block of block diagrams and flowcharts, and combinations of blocks in block diagrams and flowcharts, can be implemented by computer program instructions. These computer program instructions may be loaded onto a computer or other programmable apparatus to produce a machine, such that the instructions which execute on the computer or other programmable apparatus create computer-readable media software program code configured to implement the functions specified in the block diagram or flowchart block(s). These computer program instructions may also be stored in a computer-readable memory that can direct a computer or other programmable apparatus to function in a particular manner, such that the instructions stored in the computer-readable memory produce an article of manufacture including computer-readable media software program code instructions which implement the function specified in the block diagram or flowchart block(s). The computer-readable media software program code instructions may also be loaded onto a computer or other programmable apparatus to cause a series of operational steps to be performed on the computer or other programmable apparatus to produce a computer implemented process such that the instructions which execute on the computer or other programmable apparatus provide steps for implementing the functions specified in the block diagram or flowchart block(s). Accordingly, blocks of the block diagrams or flowcharts support combinations of means for performing the specified functions, combinations of steps for performing the specified functions, and computer-readable media software program code for performing the specified functions. It will also be understood that each block of the block diagrams or flowcharts, and combinations of blocks in the block diagrams or flowcharts, can be implemented by special purpose hardware-based computer systems which perform the specified functions or steps, or combinations of special purpose hardware and computer instructions. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures can be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled”, to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable”, to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components and/or wirelessly interactable and/or wirelessly interacting components and/or logically interacting and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to.” Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, etc. unless context requires otherwise. In some instances, one or more components may be referred to herein as “configured to.” Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. Furthermore, it is to be understood that the invention is defined by the appended claims. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to inventions containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that virtually any disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms. For example, the phrase “A or B” will be understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. With respect to context, even terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Those skilled in the art will appreciate that the foregoing specific illustrative processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. One skilled in the art will recognize that the herein described components (e.g., process blocks), devices, and objects and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are within the skill of those in the art. Consequently, as used herein, the specific examples set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific example herein is also intended to be representative of its class, and the non-inclusion of such specific components (e.g., process blocks), devices, and objects herein should not be taken as indicating that limitation is desired. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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description | This application claims the benefit of U.S. Provisional Application No. 62/330,726, filed May 2, 2016, which application is hereby incorporated by reference. The utilization of molten fuels in a nuclear reactor to produce power provides significant advantages as compared to solid fuels. For instance, molten fuel reactors generally provide higher power densities compared to solid fuel reactors, while at the same time having reduced fuel costs due to the relatively high cost of solid fuel fabrication. Molten fluoride fuel salts suitable for use in nuclear reactors have been developed using uranium tetrafluoride (UF4) mixed with other fluoride salts such as UF6, and UF3. Molten fluoride salt reactors have been operated at average temperatures between 600° C. and 860° C. Binary, ternary, and quaternary chloride fuel salts of uranium, as well as other fissionable elements, have been described in co-assigned U.S. patent application Ser. No. 14/981,512, titled MOLTEN NUCLEAR FUEL SALTS AND RELATED SYSTEMS AND METHODS, which application is hereby incorporated herein by reference. In addition to chloride fuel salts containing one or more of PuCl3, UCl4, UCl3F, UCl3, UCl2F2, and UClF3, the application further discloses fuel salts with modified amounts of 37Cl, bromide fuel salts such as UBr3 or UBr4, thorium chloride (e.g., ThCl4) fuel salts, and methods and systems for using the fuel salts in a molten fuel reactor. Average operating temperatures of chloride salt reactors are anticipated between 300° C. and 600° C., but could be even higher, e.g., >1000° C. This disclosure describes various configurations and components of a molten fuel nuclear reactor. For the purposes of this application, embodiments of a molten fuel reactor that use a chloride fuel, such as a mixture of one or more fuel salts such as PuCl3, UCl3, and/or UCl4 and one or more non-fissile salts such as NaCl and/or MgCl2, will be described. However, it will be understood that any type of fuel salt, now known or later developed, may be used and that the technologies described herein may be equally applicable regardless of the type of fuel used. For example, a fuel salt may include one or more non-fissile salts such as, but not limited to, NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, AmCl3, LaCl3, CeCl3, PrCl3 and/or NdCl3. Note that the minimum and maximum operational temperatures of fuel within a reactor may vary depending on the fuel salt used in order to maintain the salt within the liquid phase throughout the reactor. Minimum temperatures may be as low as 300-350° C. and maximum temperatures may be as high as 1400° C. or higher. Similarly, except were explicitly discussed otherwise, heat exchangers will be generally presented in this disclosure in terms of simple, single pass, shell-and-tube heat exchangers having a set of tubes and with tube sheets at either end within a shell. However, it will be understood that, in general, any design of heat exchanger may be used, although some designs may be more suitable than others. For example, in addition to shell and tube heat exchangers, plate, plate and shell, printed circuit, and plate fin heat exchangers may be suitable. FIG. 1 illustrates, in a block diagram form, some of the basic components of a molten fuel reactor. In general, a molten fuel reactor 100 includes a reactor core 104 containing a fissionable fuel salt 106 that is liquid at the operating temperature. Fissionable fuel salts include salts of any nuclide capable of undergoing fission when exposed to low-energy thermal neutrons or high-energy neutrons. Furthermore, for the purposes of this disclosure, fissionable material includes any fissile material, any fertile material or combination of fissile and fertile materials. The fuel salt 106 may or may not completely fill the core 104, and the embodiment shown is illustrated with an optional headspace 102 above the level of the fuel salt 106 in the core 104. The size of the reactor core 104 may be selected based on the characteristics and type of the particular fuel salt 106 being used in order to achieve and maintain the fuel in an ongoing state of criticality, during which the heat generated by the ongoing production of neutrons in the fuel causes the temperature of the molten fuel to rise when it is in the reactor core. Criticality refers to a state in which loss rate of neutrons is equal to or less than the production rate of neutrons in the reactor core. The performance of the reactor 100 is improved by providing one or more reflectors 108A, 108B, 108C around the core 104 to reflect neutrons back into the core. Reflectors may be made of any neutron reflecting material, now known or later developed, such as graphite, beryllium, steel, or tungsten carbide. The molten fuel salt 106 is circulated between the reactor core 104 and one or more primary heat exchangers 110 located outside of the core 104. The circulation may be driven using one or more pumps 112. The primary heat exchangers 110 transfer heat from the molten fuel salt 106 to a primary coolant 114 that is circulated through a primary coolant loop 115. In an embodiment the primary coolant may be another salt, such as NaCl—MgCl2, or lead. For example, in an embodiment, the primary coolant is 42MgCl2+58NaCl salt. Other coolants are also possible including Na, NaK, supercritical CO2 and lead bismuth eutectic. In an embodiment, a reflector 108 is between each primary heat exchanger 110 and the reactor core 104 as shown in FIG. 1. For example, in an embodiment a cylindrical reactor core 104, having a diameter of 2 meters (m) and a height of 3 m, is oriented vertically so that the flat ends of the cylinder are on the top and bottom respectively. The entire reactor core 104 is completely encased in reflectors 108 between which are provided channels for the flow of fuel salt 106 into and out of the reactor core 104. Although FIG. 1 illustrates one heat exchanger 110, depending on the embodiment any number of heat exchangers 110 may be used, the heat exchangers 110 being spaced around the exterior of the core 104. For example, embodiments having two, four, six, eight, ten, twelve and sixteen primary heat exchangers are contemplated. As discussed above, any design of heat exchanger may be used but, generally, the heat exchangers 110 will be discussed in terms of a shell and tube heat exchanger. In shell and tube heat exchanger embodiments, the fuel salt may flow through the tubes which are contained within a shell filled with the primary coolant. The fuel salt enters the tubes via one or more tube sheets in the shell to prevent the fuel salt from mixing with the primary coolant. This is referred to as either a tube-side fuel or a shell-side coolant configuration. Alternatively, the fuel salt may flow through the shell and the primary coolant may flow through the tubes, which is referred to either as a tube-side coolant or shell-side fuel configuration. Salt contacting surfaces of heat exchanger components may be clad to protect against corrosion. Other protection options include protective coatings, loose fitting liners or press-fit liners. In an embodiment, cladding on the internal surface of the tubes is molybdenum that is co-extruded with the base heat exchanger tube material. For other fuel salt contacting surfaces (exterior surfaces of the tube sheets and exterior surface of the shell), the cladding material is molybdenum alloy. Nickel and nickel alloys are other possible cladding materials. Molybdenum-rhenium alloys may be used where welding is required. Components in contact with primary cooling salt may be clad with Alloy 200 or any other compatible metals, such as materials meeting the American Society of Mechanical Engineers' pressure vessel code. The tube primary material may be 316 stainless steel or any other compatible metals. For example, in an embodiment alloy 617 is the shell and tube sheet material. In a tube-side fuel embodiment the fuel salt flows through the tubes of the heat exchanger 110 and exits into the fuel salt outlet channel. The primary coolant within the shell of the heat exchanger 110 removes heat from the fuel salt traveling through the tubes and heated coolant is then passed to the power generation system 120. As shown in FIG. 1, heated primary coolant 114 from the primary heat exchangers 110 is passed to a power generation system 120 for the generation of some form of power, e.g., thermal, electrical, or mechanical. The reactor core 104, primary heat exchangers 110, pumps 112, molten fuel circulation piping (including other ancillary components that are not shown such as check valves, shutoff valves, flanges, drain tanks, etc.) and any other components through which the molten fuel circulates during operation can be referred to as the fuel loop 116. Likewise, the primary coolant loop 115 includes those components through which primary coolant circulates, including the primary heat exchangers 110, primary coolant circulation piping (including other ancillary components that are not shown such as coolant pumps 113, check valves, shutoff valves, flanges, drain tanks, etc.). The molten fuel reactor 100 further includes at least one containment vessel 118 that contains the fuel loop 116 to prevent a release of molten fuel salt 106 in case there is a leak from one of the fuel loop components. Note that not all of the primary coolant loop 115 is within the containment vessel 118. In an embodiment fuel salt flow is driven by a pump 112 so that the fuel salt circulates through the fuel loop 116. In the embodiment shown, there is one pump 112 for each primary heat exchanger 110. Fewer or more pumps may be used. For example, in alternative embodiments multiple, smaller pumps may be used for each heat exchanger 110. In an embodiment, a pump 112 may include an impeller at some location within the fuel loop 116. In an embodiment, the channel or component of the fuel loop containing the impeller also serves as the pump casing or body, so that rotation of the impeller drives the flow of fuel salt around the fuel loop. Impellers may be of any design such as open, semi-open or closed and the impeller blades may be any configuration such as backward-curved, forward-curved or radial. One or more diffuser vanes may also be provided at or near an impeller location to assist in directing the flow driven by the rotation of the impeller. The impeller may be attached to a rotating shaft that connects the impeller to a motor which may be located outside of the fuel loop and containment vessel. An example of this embodiment can be found in FIGS. 6A-6C, discussed below. Other pump configurations are also possible. Broadly speaking, this disclosure describes multiple alterations and component configurations that improve the performance of the reactor 100 described with reference to FIG. 1. Auxiliary Cooling System (ACS) In an embodiment, an ACS may be provided for auxiliary cooling of the primary coolant. The ACS may share some components of the primary coolant loop and be designed to take over cooling during certain events or under certain circumstances. In an embodiment, the ACS may be composed of one or more independent ACS loops (i.e., independent from the other ACS loops) in which each ACS loop shares some of its flow path with a primary coolant loop. The major components that together constitute the ACS are: pipes such as 316 stainless steel pipes with nickel cladding; containment isolation valves; auxiliary heat exchangers; air ducts; support structures; and instrumentation and diagnostics. FIGS. 2A and 2B illustrate an embodiment of a layout of the primary cooling loop and the ACS adapted for use in a molten fuel nuclear reactor. In the embodiment shown, a molten fuel reactor 202 is shown connected to four primary coolant circuits. The auxiliary cooling system is integrated into the four independent parallel primary coolant circuits. Each circuit includes a primary cooling leg 210 and an ACS leg 212 and a shared reactor coolant inlet and outlet portion 204 that pipes the primary coolant into and out of the reactor 202. Because it is shared, the reactor inlet and outlet portion 204 is considered part of both the primary coolant loop and the ACS loop. The primary cooling loop has the function of transporting thermal energy from the primary heat exchangers 206 inside the reactor 202, which may be within the reactor's containment vessel, to the power generation system (not shown) during normal operation. The primary cooling loop is made up of the reactor inlet and outlet portion 204 and the primary cooling leg 210. In the embodiment shown, the primary cooling legs 210 include the heat exchangers 206 and coolant management equipment needed to maintain the normal cooling operation for the reactor 202. In the embodiment shown, each primary cooling leg 210 includes, among other things, two heat exchangers 206, a coolant pump 220, a power recovery system in the form of a steam generator 214, a drain tank 216, and a coolant makeup tank 218. A steam reheater 226 may also be provided. The power recovery system converts the thermal energy from the heated primary coolant into electrical, thermal or mechanical power. Many types of power recovery systems are known and any system, now known or later developed, may be used. In the specific embodiment shown, the steam generator 214 transfers energy from the heated primary coolant to a water stream to generate steam from which electrical, thermal or mechanical power is generated, for example, by using the steam to run a steam turbine cycle to generate electricity. The drain tank 216 is provided at a low spot in the circuit piping to allow the salt to be drained from the circuit to allow for servicing, repair, or replacement of components in the circuit. The makeup tank 218 serves as an expansion chamber and contains extra coolant to maintain the desired level and pressure of the coolant in the circuit throughout the operational temperature range even though the volume of the primary coolant may fluctuate over that temperature range. The ACS loop has the function of transporting thermal energy from the primary heat exchangers to atmospheric air in some non-normal operating scenarios via entirely passive physical processes. In the embodiment shown, the ACS loop is made up of the reactor inlet and outlet portion 204 and the ACS leg 212. The ACS leg 212 includes an auxiliary heat exchanger (AHX) 222 and an air duct 224. The air duct 224 permits the flow of cool ambient air to the AHX 222 and vents heated air to the atmosphere in order to remove heat from the primary coolant flowing in the ACS loop. The AHX 222 is placed to generate buoyancy-driven natural circulation of coolant salt when the ACS is active. As mentioned above, the AHX 222 may be any suitable air-cooled heat exchanger design including, but not limited to, a fin, a fin fan heat exchanger, a plate and shell, or a shell and tube heat exchanger. In an embodiment, the ACS 200 may be bypassed during normal reactor operation so that as much of the heat generated by the reactor 202 as possible is available for generating power. In an alternative embodiment, the ACS 200 may be in continuous use regardless of the reactor's operating condition. In yet another embodiment, the circuits may be designed in such a way that the ACS 200 has only a reduced flow of primary coolant or otherwise causes a reduced amount of heat transfer to the air during normal operation, but a larger flow and/or heat transfer during non-normal operation. For example, in an embodiment during a loss of power, the air duct 224 may automatically open or may open more fully to increase the air flow through the AHX 222. As mentioned above, in an embodiment the ACS 200 may be designed to be entirely passive in operation. That is, the cooling provided by the ACS 200 occurs in the absence of externally provided power or control. In such a design, during a loss of forced flow event the primary coolant flow may be driven by the heat generated by the reactor 220, such as the decay heat generated by the fuel salt in the reactor when the reactor is subcritical. In addition, in an embodiment the cooling circuits 202 may be designed so that in the event of a loss of power flow is directed from the primary loop to the ACS loop. For example, valves between the reactor inlet and outlet portion 204 and the primary cooling leg 210 may be automatically closed in the event of a loss of power forcing the coolant flow through the ACS loop. In the embodiment shown, check valves 228 are provided that prevent flow from the ACS loop from circulating into the primary cooling leg 210. Note that even though the ACS 200 may be able to operate completely passively in an emergency, in a non-emergency the ACS 200 may also be controllable to augment or take over cooling duties from the primary coolant loop 206 when desired. Such may occur at times when the power generation is not needed or desired, but the operator still has need to operate the reactor 202, such as during startup, shutdown, periods of low power demand, or testing. As such, the ACS 200 can provide both flexibility in operation while serving as the emergency heat removal system. In an embodiment, the primary coolant is a molten salt, such as 42MgCl2+58NaCl, and all salt-facing surfaces are made of or coated with material suitable for use as the coolant salt. For example, the pipes may be 316 stainless steel with a nickel cladding on their interior surfaces and have a trace heating system used during startup to melt frozen salt. In an embodiment, check valves may be provided to control flow of coolant into the reactor 202 when the ACS 200 activates to avoid overcooling the fuel. Isolation valves 230 may be provided to serve as part of the containment boundary for non-normal operating scenarios. The embodiment illustrated in FIGS. 2A and 2B makes use of four independent ACS loops. The ACS loops may be sized for redundancy so that if any one of them fails, the system still provides sufficient cooling to keep all structures, systems, and components within design limits. Direct Reactor Auxiliary Cooling System (DRACS) As discussed above, in the event of a power failure, natural circulation of the fuel salt through the primary heat exchanger can prevent buildup of too much thermal energy in the fuel salt. However, a direct reactor auxiliary cooling system (DRACS) may also be provided. In an embodiment, during a power failure the DRACS may be responsible for removing a sufficient amount of thermal energy from the reactor to prevent damage to any of the components. A DRACS may include one or more dedicated secondary heat exchangers that transfer thermal energy from the fuel salt to the primary coolant and, subsequently, the atmosphere via the ACS. Examples of such DRACS heat exchanger designs can be found in U.S. Provisional Patent Application Ser. No. 62/422,474, filed Nov. 15, 2016, titled THERMAL MANAGEMENT OF MOLTEN FUEL NUCLEAR REACTORS, which application is hereby incorporated herein by reference. In an embodiment, the DRACS system has a dedicated DRACS heat exchanger in the reactor pool in addition to any primary heat exchangers. The DRACS may also include a dedicated DRACS coolant loop completely independent from the primary coolant loop. In an embodiment, the DRACS may be sized to be capable of removing the expected decay heat from the reactor in the event that the primary coolant loop has become completely inoperative. Embodiments of the reactors described herein may provide one or more DRACS heat exchangers in addition to the primary heat exchangers. The DRACS heat exchangers may be located at a level higher than the thermal center of the reactor core and the primary heat exchanger, in order to take greater advantage of the natural circulation which is more important during a loss of forced flow event. For example, in an embodiment a DRACS heat exchanger is provided above the primary heat exchanger in the flow of fuel salt. Separate DRACS heat exchangers may or may not be used during normal operation to provide cooling. In an alternative embodiment, the upper reflector may incorporate a DRACS heat exchanger. In this embodiment, the DRACS heat exchanger may be contained within the upper reflector. This may use a coolant that also serves as a reflector or neutron absorber, as described in greater detail in U.S. patent application Ser. No. 15/282,814, filed Sep. 30, 2016, titled NEUTRON REFLECTOR ASSEMBLY FOR DYNAMIC SPECTRUM SHIFTING. In an embodiment, the coolant may be solid at operating temperatures but, upon the top of the reactor core reaching some higher temperature, the coolant may melt at which point the DRACS may begin operation. For example, lead and alloys of lead such as lead-bismuth alloy (e.g., lead-bismuth eutectic 44.5Pb-55.5Bi) and lead-copper alloy (e.g., molybdockalkos) may be used. Molten Fuel Pump Configurations FIGS. 3A-3C illustrate an embodiment of a molten fuel reactor design that has a pump for each primary heat exchanger to drive fuel salt flow. In the embodiment shown, eight primary heat exchangers 310 are spaced around a central reactor core 304. FIG. 3A is a plan view looking down from the top of the reactor 300. FIG. 3B is a cross-sectional view through the center of the reactor 300 and two opposing heat exchangers 310. FIG. 3C illustrates a perspective view of an eight-exchanger configuration of a molten fuel reactor 300 partially cutaway to show different internal components including the impellers, shafts, and motors of the pumps. The reactor core 304 is defined on top by a vessel head 319, which may be a reflector or incorporate a reflector, and on bottom by a neutron reflector 308B. Laterally, the reactor core 304 is defined by the shells of the eight heat exchangers 310. In operation, the heated fuel salt from the reactor core 304 is pumped through the heat exchangers where it is cooled and the cooled fuel salt returned to the core 304. In the embodiment shown, the reactor core 304 and heat exchangers 310 are within a containment vessel 318. The primary containment vessel 318 is defined by a liner or set of liners that create an open-topped vessel. The cooled primary coolant enters and exits the vessel 318 from the top, which allows the containment vessel to be unitary and have no penetrations. The primary coolant loop is integrated into the reactor 300 so that the entering primary coolant first cools at least a portion of the containment vessel 318. After being routed next to an interior surface of the containment vessel 318 for some distance in a primary coolant inlet channel 330, in the embodiment shown the coolant is then routed into the bottom of the primary heat exchanger 310. The coolant exits the top of the primary heat exchanger 310 and is then routed out of the containment vessel 318 and to a power generation system (not shown). In the embodiment shown fuel salt is driven through the fuel loop by eight separate impellers 312A located above the heat exchangers 310 in the upper channels. In the location of the impellers 312A, the sides of the channels serve as the casings or pump bodies shaped to complement the impellers 312A in order to obtain efficient flow. In the embodiment shown, the impellers 312A are between the upper tube sheet 332 of the heat exchangers and the horizontal portion of the channels from the top of the reactor core 304. Each impeller 312A is connected by a rotating shaft 312B to a motor 312C located above the reactor 300. This removes the electronic components of the pump from the region of high neutron flux and high temperatures. One or more access ports may be provided in the vessel head 319 so that the impeller 312A may be removed and serviced or replaced. The impellers 312A and shafts 312B may be made of any material suitable for the high temperature and neutron flux fuel salt environment that will exist in the fuel loop at the fuel inlet of the primary heat exchanger 310. For example, the fuel-facing components may be formed from one or more molybdenum alloys, one or more zirconium alloys (e.g., ZIRCALOY™), one or more niobium alloys, one or more nickel alloys (e.g., HASTELLOY™ N) or high temperature ferritic, martensitic, or stainless steel and the like. The impellers 312A and shafts 312B may be clad on the fuel salt-facing surfaces to protect against corrosion. Other protection options include protective coatings. In an embodiment, cladding may be molybdenum that is co-extruded with the base impeller or shaft material. Alternative cladding material includes molybdenum alloys, nickel and nickel alloys, and molybdenum-rhenium alloys. FIG. 4 illustrates an alternative embodiment of a molten fuel reactor design, similar to that of FIGS. 3A-3C but provided with an inner reflector 408C. In the embodiment shown, the inner reflector 408C is provided separating the reactor core 404 from the heat exchangers 410. This reduces the neutron flux through the components of the heat exchangers as well as through the impellers 412A and shafts 412B. Otherwise, the reactor 400 is similar in operation and configuration to the reactor of FIGS. 3A-3C. FIGS. 5, 6, and 7 illustrate embodiments of an alternative pump configuration in which the impeller is bottom mounted. In a bottom-mounted impeller configuration, the impeller is located in the fuel loop in the cooled fuel salt outlet channel below the primary heat exchanger, in which the outlet channel acts as the casing or body for the impeller. In this configuration, the impeller is in a lower temperature environment than in a top-mounted configuration as shown in the FIGS. 3A-3C and 4, above. Depending on the embodiment, the reduced wear on the impeller and portion of the shaft immediately adjacent the impeller may justify the additional complexity in overall design. FIG. 5 illustrates a reactor 500 having a reactor core 504 defined by an upper reflector 508A, a lower reflector 508B and an inner reflector 508C. In the embodiment shown, the lower reflector 508B extends laterally and up the sides of the containment vessel 518 for added protection. The primary heat exchanger 510 configured to have shell-side coolant flow (illustrated by dotted lines 514), the coolant entering through a coolant inlet channel 530 and heated coolant exiting from coolant outlet channel 536. In the embodiment shown, fuel flows (illustrated by dashed lines 506) from the reactor core 504, through an upper channel above the inner reflector 508C, and into the heat exchanger 510 through the inlet tube sheet 532. After passing through the tube set, the now-cooled fuel exits the lower tube sheet 531 and flows back into the reactor core 504 via a lower channel under the inner reflector 508C. In FIG. 5, the fuel-flow impeller 512A is located below the fuel salt outlet of the primary heat exchanger 510 configured to have shell-side coolant flow. The impeller 512A is attached to a shaft 512B connected to a top-mounted motor 512C above the vessel head (not shown) and the upper reflector 508A. In this embodiment, the shaft 512B passes through the heat exchanger 510. This may increase the complexity of the heat exchanger 510. In an embodiment, the impeller 512A and shaft 512B are integrated into the heat exchanger 510 whereby servicing is achieved by removing the heat exchanger/impeller and shaft assembly as a unit. In an alternative embodiment (not shown), the shaft 512B may not penetrate heat exchanger, but rather be located so that it is adjacent to but outside of the heat exchanger 510. FIG. 6 illustrates a reactor 600 similar to that in FIG. 5. In the embodiment shown, the reactor 600 has a reactor core 604 defined by an upper reflector 608A, a lower reflector 608B and an inner reflector 608C. Again, the lower reflector 608B extends laterally and up the sides of the containment vessel 618 for added protection. The primary heat exchanger 610 configured to have shell-side coolant flow (illustrated by dotted lines 614), the coolant entering through a coolant inlet channel 630 and heated coolant exiting from coolant outlet channel 636. In the embodiment shown, fuel flows (illustrated by dashed lines 606) from the reactor core 604, through an upper channel above the inner reflector 608C, and into the heat exchanger 610 through the inlet tube sheet 632. After passing through the tube set, the now-cooled fuel exits the lower tube sheet 631 and flows back into the reactor core 604 via a lower channel under the inner reflector 608C. In FIG. 6, the impeller 612A is still located below the fuel salt outlet channel below the primary heat exchanger 610 and attached to a shaft 612B. However, in FIG. 6, the shaft 612B extends downward and is coupled to a bottom-mounted motor 612C located outside of the containment vessel 618 by an electro-magnetic coupler 650. In this embodiment, the shaft 612B does not penetrate the containment vessel 618. This may increase the complexity of the containment vessel's construction but maintains the containment vessel 618 as a unitary vessel. FIG. 7 illustrates a reactor 700 similar to that in FIG. 5. In the embodiment shown, the reactor 700 has a reactor core 704 defined by an upper reflector 708A, a lower reflector 708B and an inner reflector 708C. Again, the lower reflector 708B extends laterally and up the sides of the containment vessel 718 for added protection. The primary heat exchanger 710 configured to have shell-side coolant flow (illustrated by dotted lines 714), the coolant entering through a coolant inlet channel 730 and heated coolant exiting from coolant outlet channel 736. In the embodiment shown, fuel flows (illustrated by dashed lines 706) from the reactor core 704, through an upper channel above the inner reflector 708C, and into the heat exchanger 710 through the inlet tube sheet 732. After passing through the tube set, the now-cooled fuel exits the lower tube sheet 731 and flows back into the reactor core 704 via a lower channel under the inner reflector 708C. In FIG. 7, the impellers 712A are located in the reactor core 704. In this embodiment, each cooled fuel salt channel is provided with an impeller 712A located near the bottom of the reactor core 704. As with FIG. 6, the impellers 712A have shafts 712B that extend downward and are coupled to bottom-mounted motors 712C located outside of the containment vessel 718 by electro-magnetic couplers 750. In this embodiment, the shafts 712B do not penetrate the containment vessel 718. In an alternative embodiment, instead of a separate and independent pump for each primary heat exchanger, fewer or more pumps may be provided. For example, in an alternative embodiment of the reactor 700 of FIG. 7, a single impeller 718A may be provided at the bottom of the reactor core 704 that draws flow from the cooled fuel outlet of two or more of the heat exchangers 710. FIG. 8 illustrates yet another embodiment of a pump configuration in which a single impeller 812A is located within the reactor core 804. In the embodiment shown, a single impeller 840 is rotated about the central axis of the reactor 800. In the embodiment shown, the impeller 840 includes a number of blades 842, a hub 844 from which the blades 842 extend laterally, and a shaft 812B coupled to the hub 844. Upon rotation of the shaft 812B, the hub and blades also rotate and drive the circulation of the fuel salt within the reactor 800 as shown by arrows 806. In the embodiment shown, the blades 842 extend from the hub 844 to a point adjacent to the side of the reactor core, in this case defined the internal reflectors 808C. In the embodiment shown, the side of the reactor core is provided with a complementary casing surface for the ends of the blades 842 in order to more efficiently drive the flow of the molten fuel through the reactor core 804. One or more diffusers 848 may be provided in each of the upper channels to make the flow of salt more uniform as it circulates into the heat exchanger 810. The diffusers may be as simple as a flow directing baffle somewhere within the fuel loop or may be a more complicated set of baffles, orifice plates or other static elements. In an alternative embodiment (not shown), the impeller is bottom mounted as discussed with reference to FIGS. 6 and 7 and the shaft 846 extends down from the hub, rather than upward, and is rotated by an electromagnetic coupling below the reactor containment vessel 818. FIG. 9 illustrates yet another pump configuration in which the impeller is intermediately located between two sections of a primary heat exchanger. In the embodiment shown, a reactor 900 is provided having an upper reflector 908A, a lower reflector 908B and one or more heat exchangers 910 enclosing a reactor core 904, all contained within a containment vessel 918. An inner reflector (not shown) may or may not be provided, depending on the embodiment, between the reactor core 904 and the heat exchanger 910, between the heat exchanger 910 and the coolant inlet channel 930 or both. In the embodiment shown, fuel salt is circulated through the shell of the heat exchanger 910, as illustrated by dashed line 906, and coolant is passed through the tubes of the tube set, as illustrated by dotted line 914. The heat exchanger 910 is divided into two sections 910A and 910B by an intermediate wall 928. The tube set is continuous throughout and extends from a lower tube sheet 931, which is the coolant inlet, to the upper tube sheet 932, which is the coolant outlet. In the embodiment shown, heated fuel salt flows past the tube set in the upper section 910A of the heat exchanger 910, which is open to the reactor core 904. At least a portion of the opposite side of the shell is also open allowing the fuel salt to flow into a pump channel 912D containing an impeller 912A. The impeller 912A is connected via a shaft 912B to a motor (not shown) as described above). Rotation of the impeller 912A drives the fuel salt into the lower section 910B heat exchanger shell, through the tube set, and out the bottom of the shell through another opening into the reactor core 904. One or more baffles 929 may also be provided to route the flow of fuel salt through the tubeset. FIG. 9 illustrates a region 934 within the shell of the heat exchanger 910 that is above the level of fuel salt in the reactor core 904. This region may either be solid except for the penetrating tubes, for example filled with a reflector material, or may be a headspace filled with inert gas. In the embodiment shown in FIG. 9, the impeller 912A is within the heat exchanger, that is, within the shell of the heat exchanger 910. It is located away from the area of high neutron flux and also not exposed to the highest temperatures of the reactor 900. The impeller 912A and shaft 912B may be integrated into the heat exchanger 910 so that all are removed as an assembly for servicing or replacement. In an alternative embodiment using a different heat exchanger design, the heat exchanger may be similarly adapted to include an impeller or impellers within the heat exchanger. For example, in a plate and frame heat exchanger an impeller may be located within a corner port of the inter plate transfer path or, alternatively, a plate within the stack of heat exchanging plates could be provided with an impeller. It will be clear that the systems and methods described herein are well adapted to attain the ends and advantages mentioned as well as those inherent therein. Those skilled in the art will recognize that the methods and systems within this specification may be implemented in many manners and as such are not to be limited by the foregoing exemplified embodiments and examples. In this regard, any number of the features of the different embodiments described herein may be combined into one single embodiment and alternate embodiments having fewer than or more than all of the features herein described are possible. While various embodiments have been described for purposes of this disclosure, various changes and modifications may be made which are well within the scope contemplated by the present disclosure. For example, electromagnetic couplers could be used with top-mounted motors to reduce the number of penetrations of the vessel head, in which case the shafts need not penetrate the vessel head, for instance, as shown in FIGS. 3A-C. Numerous other changes may be made which will readily suggest themselves to those skilled in the art and which are encompassed in the spirit of the disclosure. |
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abstract | A detector signal-processing circuit comprises the following: a current/voltage conversion part that converts the current value of a neutron detector to a voltage value; a variable gain amplification part that performs amplification by a first-step variable gain using a D/A converter; a current level response-use resistance circuit that selects the measurement range in accordance with the voltage value; temperature measurement units for measuring the temperature of the resistance circuit for current level response; a temperature compensation part for commanding gain compensation by the D/A converter on the basis of the measured temperature; and a selective adjustment control part for selective control of the measurement range and adjustment of the variable gain of the variable gain amplification part. Due to this configuration, neutron flux can be measured with high precision while maintaining a constant output precision, before and after switching of the measurement range. |
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052710548 | description | DETAILED DESCRIPTION OF THE INVENTION The numeral 10 generally designates a fuel assembly grid corner region which incorporates the principals of the invention. Evident are fragments of two perimeter strip pieces, 12 and 14 joined by a perimeter strip seam weld 16 which are formed to overlap at the extremities of flat plane side sections thereof, 18 and 18', respectively. Another flat plane section 20 of the strip 12 is connected to the flat side section 18 by means of a transverse continuous corner bend 22 of varying radii, with its outer longitudinal end portions 24 having a greater radius that its inner portion 26. At the top and bottom margins of the flat side sections 18 and 20, are upper and lower anti-hang-up tabs 28 and 28', respectively. In flat plane section 20 of perimeter strip piece 12 with its perimeter in said flat plane is a stress-relieving slot 30 which extends an equal slot perimeter length on either side of the longitudinal center line of the flat section 20. The length of slot 30 is approximately 1/3 to 1/2 of the width of the flat section 20. The width of slot 30 is preferably less than twice the thickness of the strip material which is typically a stainless steel such as "Inconel 625" or zirconium alloy such as "Zircaloy 4". As a comparison of FIG. 1 and its' flat sections 18 and 20, with FIGS. 2 and 3, and their flat sections 20 and 32, will show, it is possible to have transverse slots 30 in the flat sections on each side of the bend or bend line 22. This depends on whether the flat section 18, 20 or 32 is interrupted by fuel rod support arches 34 or by springs 36. The flat side sections 20 and 32 with arches 34 adjacent bend line 22 having much less cut-out periphery for stress relief that the flat side 18 which has a large spring 36 cut-out adjacent bend line 22 which will inherently stress relieve the corner-bend 22. It will thus be seen that a perimeter strip grid cornerpiece of increased flatness having two flat-side sections on either side of transverse bend line 22 is provided whether or not the fuel rod support structure on one or both sides of said bend line is a small cut-out for arches 34. This stress-relief is accomplished without a significant weakening of the strip in the portion or area near bend 22 containing the slot 30. It therefore maintains its required rigidity. |
summary | ||
description | 1. Field of the Invention This invention relates to a technique of estimating the time of failure of a product including plural components and the degree of degradation of each component, and reflecting these to a maintenance plan. 2. Description of the Related Art Conventional maintenance planning depends on the experience and intuition of individual servicemen. Therefore, the risk of damage to a user caused by the unavailability of a product and the maintenance cost cannot be balanced with each other. That is, in order to reduce the risk of product failure, a replacement operation takes place more than necessary so as to replace a component in which failure is very likely to occur, before its service life end, and the maintenance cost increases. Conversely, to lower the maintenance cost by using up the component to the end of its service life, a visit is made after failure occurs. Therefore, it is a trade-off with a longer downtime and increased damage to the user caused by the unavailability of the product. In another type of industry, there is an example of maintenance planning based on the cost and risk (see, JP-A-2004-152017). However, this is limited to a judgment on whether to conduct maintenance or not, based on the risk, and it is difficult to estimate the cost if the number of components to be replaced (consumable parts) increases. That is, when judging whether to replace consumable parts or not, a very high calculation cost is required for calculating which combination of consumable parts should be replaced as the best maintenance plan. Also, it cannot be seen how long the time for next visit can be prolonged, and the cost cannot be calculated simply. This invention is made in order to solve the foregoing problems, and it is an object of this invention to provide a technique that enables reduction in the cost of maintenance services and reduction in the downtime of a product. To solve the foregoing problem, a maintenance system according to this invention adapted for making a maintenance plan for consumable parts of an apparatus that is a maintenance target, includes: an interval information acquiring unit configured to acquire information related to a combination of a visit interval that prescribes a time interval at which a visit should be made for maintenance operation for each consumable part, and a replacement interval that prescribes a time interval at which each consumable part should be replaced and that is associated with the visit interval; a counter value acquiring unit configured to acquire a counter value that indicates actual use of consumable parts in the apparatus that is a maintenance target; and a maintenance plan calculating unit configured to calculate timing at which a next visit should be made for the apparatus that is a maintenance target, and a consumable part that should be replaced at the timing, on the basis of the information acquired by the interval information acquiring unit and the counter value acquired by the counter value acquiring unit. To solve the foregoing problem, a maintenance system according to this invention adapted for making a maintenance plan for consumable parts of an apparatus that is a maintenance target, includes: interval information acquiring means for acquiring information related to a combination of a visit interval that prescribes a time interval at which a visit should be made for maintenance operation for each consumable part, and a replacement interval that prescribes a time interval at which each consumable parts should be replaced and that is associated with the visit interval; counter value acquiring means for acquiring a counter value that indicates actual use of consumable parts in the apparatus that is a maintenance target; and maintenance plan calculating means for calculating timing at which a next visit should be made for the apparatus that is a maintenance target, and a consumable part that should be replaced at the timing, on the basis of the information acquired by the interval information acquiring means and the counter value acquired by the counter value acquiring means. To solve the foregoing problem, a maintenance method according to this invention adapted for making a maintenance plan for consumable parts of an apparatus that is a maintenance target, includes: an interval information acquiring step of acquiring information related to a combination of a visit interval that prescribes a time interval at which a visit should be made for maintenance operation for each consumable part, and a replacement interval that prescribes a time interval at which each consumable parts should be replaced and that is associated with the visit interval; a counter value acquiring step of acquiring a counter value that indicates actual use of consumable parts in the apparatus that is a maintenance target; and a maintenance plan calculating step of calculating timing at which a next visit should be made for the apparatus that is a maintenance target, and a consumable part that should be replaced at the timing, on the basis of the information acquired in the interval information acquiring step and the counter value acquired in the counter value acquiring step. Hereinafter, an embodiment of this invention will be described with reference to the drawings. FIG. 1 is a view of a system configuration showing the outline of a maintenance system according to an embodiment of this invention. FIG. 1 shows an example in which a maintenance plan for an image processing apparatus or multifunction peripheral (MFP) 201 installed at a user's location is made by the maintenance system according to this embodiment, and in which a serviceman 202 carries out maintenance services according to the maintenance plan. In the conventional maintenance services, the serviceman 202 visits an MFP that is a maintenance target at every timing for PM (preventive maintenance) set for each MFP, and carries out replacement of consumable parts, cleaning, and confirmation of operations of the MFP. Other than PM, when failure accidentally occurs, the serviceman receives a service call from the user and makes a visit to repair the MFP. As for the PM operation, since there are plural consumable parts in one MFP, if all the consumable parts are note degraded at the time of PM and the consumable parts that have not reached the end of their lives are replaced, it causes loss. Also, when accidental failure occurs, if only the consumable part that has failure is replaced at that time, the replacement time deviates from the PM cycle. Thus, while basically following a preset PM cycle, the serviceman 202 adjusts the replacement time for each consumable part individually on the basis of experience and thus tries to reduce the loss. However, if reduction in the cost is attempted by indiscriminately delaying the replacement time and extending the time of using the consumable parts, the risk of failure of the consumable parts increases adversely and therefore the unavailability of the MFP causes damage to the user. Moreover, if the replacement time is changed individually for each consumable part, it may be considered that the number of visits increases, which conversely increases the maintenance cost. Thus, the maintenance system 1 according to this embodiment is configured to calculate failure rate distribution for each consumable part on the basis of the past maintenance history data and to calculate the time for next visit and a list of consumable parts to be replaced at the time, on the basis of the calculated failure rate distribution. The maintenance system 1 in this case is adapted for making a maintenance plan for consumable parts of an apparatus that is a maintenance target, and includes a failure rate distribution calculating unit 204, a storage unit 205, a maintenance planning unit (visit interval calculating unit, replacement interval calculating unit, combination calculating unit, interval information acquiring unit, maintenance plan calculating unit) 206, a replacement difficulty judging unit 207, a notifying unit 208, a counter value acquiring unit 209, a CPU 801, and a memory 802. The failure rate distribution calculating unit 204 calculates failure rate distribution for each consumable part on the basis of maintenance history data as history information related to the maintenance operation that has been carried out to the apparatus that is the maintenance target. Here, the history information related to the maintenance operation that has been carried out to the apparatus that is the maintenance target is inputted, for example, by the serviceman 202 who has carried out the maintenance operation, and thereby stored in the storage unit 205 at a service center 203. The maintenance planning unit (visit interval calculating unit) 206 randomly calculates, for each consumable part, a “visit interval” prescribing a time interval at which a visit should be made for maintenance operation for each consumable part, on the basis of the failure rate distribution of each consumable part calculated by the failure rate distribution calculating unit 204. The maintenance planning unit (replacement interval calculating unit) 206 also randomly calculates, for each consumable part, a “replacement interval” prescribing a time interval at which replacement of each consumable part should be carried out, on the basis of the failure rate distribution of each consumable part. The “consumable parts” here may include, for example, a photoconductor drum, charger wire, fixing roller, transfer belt and the like. In this embodiment, however, the “consumable parts” include a cartridge in which plural consumable parts having different functions from each other are integrally formed as a unit. Now, the maintenance planning unit (visit interval calculating unit and replacement interval calculating unit) 206 calculates a value close to an interval with which it is predicted that the failure probability is equal to or higher than a predetermined probability, on the basis of the failure probability distribution of each consumable part. The visit interval calculated by the maintenance planning unit (visit interval calculating unit) 206 for each consumable part is set to be longer than the replacement interval calculated by the maintenance planning unit (replacement interval calculating unit) 206. The maintenance planning unit (combination calculating unit) 206 performs search processing using the Monte Carlo method or genetic algorithm on the basis of the visit interval calculated by the maintenance planning unit (visit interval calculating unit) 206 and the replacement interval calculated by the maintenance planning unit (replacement interval calculating unit) 206, and thereby calculates a combination of a visit interval and a replacement interval that minimizes a predetermined cost, of combinations of time intervals at which a visit should be made for maintenance operation and consumable parts that should be replaced during the visit. The “predetermined cost” described here is the sum of the labor costs required for the maintenance operation by the serviceman, the material costs of the consumable parts, and the amount of loss caused by the unavailability of the apparatus that is the maintenance target to the user. The maintenance planning unit (interval information acquiring unit) 206 acquires information related to a combination of the “visit interval” and the “replacement interval” associated with the visit interval, which is the information calculated by the maintenance planning unit (combination calculating unit) 206. The counter value acquiring unit 209 acquires a counter value indicating actual use of the consumable parts of the apparatus that is the maintenance target. The counter value in this case refers to an actual use value that is effective for grasping the degree of degradation of each consumable part mounted in the image processing apparatus 201, such as the number of sheets processed by the image processing apparatus 201, for example, the number of scanned pages of a manuscript, the number of printed sheets or the like. The counter value acquired by the counter value acquiring unit 209 is sent to the service center 203 and stores into the storage unit 205. The maintenance planning unit (maintenance plan calculating unit) 206 calculates timing at which the next visit should be made for the apparatus that is the maintenance target and a list of consumable parts that should be replaced at the timing, on the basis of the information acquired by the maintenance planning unit (interval information acquiring unit) 206 and the counter value acquired by the counter value acquiring unit 209. The replacement difficulty judging unit 207 judges whether the consumable parts that should be replaced, calculated by the maintenance planning unit (maintenance plan calculating unit) 206, are components that can only be replaced by the serviceman carrying out the maintenance operation for the apparatus (serviceman replacement units or SRU) or not. The notifying unit 208 includes, for example, a liquid crystal display or the like. If it is judged by the replacement difficulty judging unit 207 that the consumable parts are components that can only be replaced by the serviceman, the notifying unit 208 issues a notification that a visit should be made for the apparatus that is the maintenance target in order to replace the components, in the form of screen display. The notification by the notifying unit 208 is not necessarily limited to the screen display and it may be issued, for example, in the form of audio notification, print processing and the like. The CPU 801 is responsible for performing various types of processing in the maintenance system 1 and also responsible for realizing various functions by executing programs stored in the memory 802. The memory 802 includes, for example, a ROM, RAM or the like, and is responsible for storing various types of information and programs used in the maintenance system 1. FIG. 2 is a view showing the relation between a data format and each data table used in the maintenance system 1 according to this embodiment. As shown in FIG. 2, the maintenance system 1 uses nine data tables, that is, a “user” table in which a constant for each user is set, a “support center” table in which a constant for a support center is set, a “machine type” table in which a constant for a machine type is set, a “machine” table in which a constant for each machine and a variable calculated from the status of use are set, a “consumable part” table in which a constant for a consumable part and a failure rate variable calculated from market data are set, a “user/machine correspondence” table indicating the correspondence of a machine owned by a user, a “maintenance history” table in which maintenance operation history by a serviceman is recorded, a “consumable part status” table in which the status of each consumable part is set, and a “counter history” table in which counter history of each machine is recorded. The arrows in FIG. 2 indicate that the attribute of the start of the arrow is set into the attribute of the end of the arrow. The table name of the reference source is arranged before “.” of the attribute of the end of the arrow, and the attribute name of the reference source is shown after “.” For example, “machine. ID” in the “maintenance history” table indicates “ID” of the “machine” table. The serviceman 202 follows the format of the maintenance history table on the basis of the operation record gathered as a maintenance operation report, and updates the maintenance history table in the storage unit 205, for example, by using an operation input unit, not shown, provided at the service center 203. The MFP 201 is connected with the service center 203 so that they can communicate with each other via the Internet, public telephone line or the like. When a regular communication time (for example, 10 o'clock every day) set in the MFP 201 comes, the MFP 201 carries out communication with the service center 203. In this communication, the MFP 201 sends the ID number of the MFP 201, the current date and time, and the current total counter value to the service center 203. The service center 203 reflects the information received from the MFP 201 to the counter history table (FIG. 2) stored in the storage unit 205. After that, the MFP 201 confirms the status of communication and additional information, and ends the communication with the service center 203. FIG. 3 is a view showing an exemplary maintenance history table 301 with its contents updated as described above. The “total counter value” is a counter value indicating how many sheets are outputted as of A4 size, where 1 represents copying/printing in A4 and 2 represents copying/printing in A3. Here, the “counter” in the “maintenance history” table inputted by the serviceman 202 takes a similar value. The failure rate distribution calculating unit 204 performs fitting to Weibull distribution (where m represents shape parameter and η represents scale parameter), which is broadly used for failure distribution analysis,F(t)=1−e{−(t/η)^m}on the basis of the maintenance history table 301, thus estimating failure rate distribution of each consumable part. In the following description, a “photoconductive drum”, which is a consumable part of machine type A (FIG. 3), will be used. In order to find failure rate distribution of the photoconductive drum, the failure rate distribution calculating unit 204 extracts data related to the photoconductive drum from the maintenance history table 301 read from the storage unit 205 and calculates the failure interval. That is, all the tuples having “machine type.name” of “machine type A” and “consumable part.abbreviation” of “photoconductive drum”, and all the tuples having “machine type.name” of “machine type A” and “consumable part.abbreviation” of “PM all-replacement” are extracted, and the failure interval is calculated from the difference from the counter value in the previous replacement. The “end of life” column at the right end in the maintenance history table 301 is added for this description. An entry having “x” in the end of life column is data acquired when failure occurs before the photoconductive drum reaches PM, and an entry having “◯” in the end of life column is data acquired when replacement is done because PM is reached without failure. The data thus acquired when replacement is made before failure (entry having “◯”) is referred to as “abort data”. As an analysis method for such data, a cumulative hazard method is known. By the cumulative hazard method, the shape parameter m and the scale parameter η of Weibull distribution are estimated on the basis of the failure interval of each extracted tuple, and a failure distribution-related variable in the “consumable part” table is updated. That is, tuples in which “machine type.name” and “abbreviation” in the consumable part table coincide with “machine type A” and “photoconductive drum” are extracted. “Failure distribution parameter 1” is substituted into the shape parameter m and “failure distribution parameter 2” is substituted into the scale parameter η. In the “failure distribution classification” section, a constant (0) corresponding to Weibull distribution is set. This calculation is used for each consumable part and the consumable part table in the storage unit 205 is sequentially updated. FIG. 4 is a view showing an exemplary consumable part table 302 that is updated as described above. The failure rate distribution calculating unit 204 also calculates proceeding distribution of the number of copied sheets per day for each apparatus that is a maintenance target, on the basis of the counter history table (FIG. 2). That is, tuples having the same “machine.ID” are extracted from the counter history table, and an average value and distribution of counter proceeding are calculated on the basis of the difference in the counter acquisition date (number of days) and the quantity of change in the counter (proceeding). Thus, “counter proceeding average” and “counter proceeding distribution” in the machine table (FIG. 2) are updated. Similarly, the up-to-date status of use and the replacement date of each consumable part are updated for each machine. FIG. 5 is a view showing an exemplary machine table 303 with its contents updated. As for the status of use, from all the tuples extracted by “machine.ID” from the “counter history” table, the “counter acquisition date” and the “total counter” value at the time in the up-to-date “counter history” table are set into “counter acquisition date” and “total counter” in the machine table as the up-to-date counter data. The machine table 303 shown in FIG. 5 shows an example of updated data. For the replacement date for each consumable part, tuples having “consumable part.abbreviation” of the consumable part in question or tuples having “PM all-replacement” and “setup” are extracted from all the tuples extracted by “machine.ID” from the “maintenance history” table, and the up-to-date “visit date” is set into “counter acquisition date” in the “consumable part status” table (FIG. 2). In the “counter” section in the “consumable part status” table, 0 is set. FIG. 6 is a view showing an example of data of a consumable part status table 304 with its contents updated as described above. Next, the maintenance planning unit 206 will be described. The maintenance planning unit 206 is capable of executing a “strategy planning mode” and a “visit date presentation mode”. The “strategy planning mode” is executed when a given quantity of maintenance history data is additionally registered to the storage unit 205 or at periodic timing such as once a month. The “visit date presentation mode” is executed every day. First, the “strategy planning mode” will be described. The maintenance planning unit 206 calculates and sets a “visit interval” and a “replacement interval” for each consumable part of each machine. The serviceman 202 is to carry out maintenance operations based on these “visit interval” and “replacement interval”. That is, if at least one of a consumable part of a specific machine that is a maintenance target (here, the image processing apparatus 201) has reached the “visit interval”, the serviceman 202 makes a visit to the apparatus. Then, the serviceman 202 replaces all the consumable parts that have reached the “replacement interval” at the time of this visit. Next, the method for calculating a “visit interval” and a “replacement interval” will be described in detail. The maintenance planning unit 206 conducts a maintenance operation simulation of a period set as a “simulation period” in the “support center” table (see FIG. 2) and calculates a “visit interval” and a “replacement interval” that minimize the required cost. As the calculation method, a heuristic method, for example, the Monte Carlo method or genetic algorithm, is used, and a maintenance operation simulation is repeated with randomly set “visit interval” and “replacement interval”, thus calculating the cost. Of these, the visit interval and the replacement interval that minimize the cost is employed. A larger preset value of the simulation period is better, but the calculation time becomes longer accordingly. Therefore, it is desired that a period considered to be sufficient relatively to the average failure time of the machine is set. The cost in this case refers to the sum of the labor costs required for repair by the serviceman, the material costs of replaced consumable parts, and the loss (downtime loss) caused by the unavailability of the machine to the user due to unexpected machine failure. Specifically, an exemplary simulation using the Monte Carlo method will be described. FIG. 7 shows samples of setting maintenance planning strategies (combinations of visit intervals and replacement intervals for each consumable part) with respect to a machine having “machine.ID” of “100213”. Two thousand patterns of maintenance planning strategy samples are randomly created. In the creation of the samples, basically, the samples are randomly generated. However, in order to prevent creation of unwanted samples, it is desired that samples are generated closely to visit intervals and replacement intervals that are expected from experience based on the failure probability distribution or the like of each consumable part. The maintenance operation simulation is conducted with these 2000 patterns of maintenance planning strategies, and the sample that minimizes the cost is found. Next, the maintenance operation simulation with the maintenance planning strategy sample 1 (see FIG. 7) will be described. FIG. 8 is a flowchart showing the flow of the maintenance operation simulation by the maintenance system 1. The maintenance planning unit 206 collates “machine type.name” in the “machine” table with “machine type.name” in the “consumable part” table with respect to “machine.ID” of 100213 and extracts all the matching tuples in the “consumable part” table. Random numbers are generated on the basis of the failure probability represented by “failure distribution classification”, “failure distribution parameter 1” and “failure distribution parameter 2” in the “consumable part” table, and the next failure time of each consumable part is calculated (S901). The shortest one of the calculated next failure times is set as a next failure occurrence time candidate (S902). Meanwhile, “ID” in the “machine” table is collated with “machine.ID” in the “consumable part status” table, and all the matching tuples in the “consumable part status” table are extracted, thus calculating the next visit schedule of the serviceman. That is, with respect to each extracted consumable part status, the “visit interval” set in the sample 1 of FIG. 7 is referred to (S903), and the shortest time is set as a next visit time candidate (S904). Next, the calculated next failure occurrence time candidate is compared with the next visit time candidate, thus deciding an event (S905). If the next failure time candidate is shorter than the next visit time candidate (Y in S905), the next failure time candidate is adopted as an elapsed time for a failure occurrence event (S906). The consumable part to be replaced is decided and the required cost is calculated. With respect to all the consumable parts except for the consumable part having occurrence of failure, the preset “replacement interval” in the “consumable part status” table is referred to, and the consumable part having a replacement interval shorter than the next failure replacement time candidate is decided as the consumable part to be replaced. The cost is the sum of the following costs (S907).Labor costs=(“user.traveling time”+Σ“consumable part.replacement time” of consumable part to be replaced)×serviceman unit priceMaterial costs=Σ“consumable part.unit price” of consumable part to be replacedDowntime loss=“user.traveling time”דmachine.downtime loss unit price” Here, “user.traveling time” represents the traveling time from the support center to the user's location. If the next failure time candidate is equal to or longer than the next visit time candidate (N in S905), the next visit time candidate is adopted as an elapsed time for a pre-maintenance event (S908). The consumable part to be replaced is decided and the required cost is calculated. With respect to all the consumable parts except for the consumable part for which pre-maintenance is to be performed (the visit interval is reached), the preset “replacement interval” in the “consumable part status” table is referred to, and the consumable part having a replacement interval shorter than the next visit time candidate is decided as the consumable part to be replaced. The cost is the sum of the following costs (S909).Labor costs=(“user.traveling time”+Σ“consumable part.replacement time” of consumable part to be replaced)×serviceman unit priceMaterial costs=Σ“consumable part.unit price” of consumable part to be replacedDowntime loss=0 The idea of downtime is that the serviceman's operation time itself causes zero downtime because it is applied when the user is not using the machine in accordance with an agreement with the user. Here, the time until the serviceman comes in the case of unexpected failure is considered to be downtime. When generating an event, a new next failure time is calculated for the replaced consumable part, and for the consumable part that has not been replaced, the elapsed time is subtracted from each of the calculated next failure time and the visit interval, thus updating the next failure time and the visit interval (S910). Then, similarly, the decision of a next failure time candidate and a next visit time candidate (S911), the decision of an event, the decision of a consumable part to be replaced, and the calculation of the cost are repeated until the elapsed time is reached during the simulation period (N in S912). This simulation is considered to one set, and the simulation is conducted with respect to all the samples shown in FIG. 7. The sample that minimizes the calculated cost is employed and set in “consumable part status.visit interval” and “consumable part status.replacement interval” as the optimum strategy. FIG. 9 is a view showing the result of the simulation with respect to each sample. In this example, the 112th sample has the minimum cost per count and is thus employed as the optimum strategy. Next, the “visit date presentation mode” will be described. FIG. 10 is a flowchart showing the flow of processing in the visit date presentation mode in the maintenance system 1. FIG. 11 is a view showing an exemplary “counter history” table. In the “visit date presentation mode”, a case where the serviceman 202 constantly confirms the next visit date (visit timing for the apparatus) is considered. A desired “visit interval” and “replacement interval” are set in advance by the above-described “strategy planning mode”, and the serviceman 202 inputs “machine.ID” of the machine which the serviceman takes charge of, by the operation input unit, not shown, at the service center, thereby confirming the next visit date. The maintenance planning unit 206 collates “ID” in the “machine” table (see FIG. 2) with “machine.ID” in the “consumable part status” table (see FIG. 2), extracts all the matching tuples in the “consumable part status” table (see FIG. 2), and refers to “counter acquisition date”, “counter”, “visit interval” and “replacement interval”. Also, it refers to “counter proceeding average” from the “machine” table. The maintenance planning unit 206 finds out the next scheduled visit date for each consumable part by the following calculation (S701).Scheduled visit date=counter acquisition date+(visit interval−counter)/counter processing average Of the scheduled visit dates for the respective consumable parts, the nearest one is decided as the visit date (S702). Next, with respect to the consumable parts having the other schedule visit dates than the nearest one, the following calculation is performed to find out the scheduled replacement date (S703).Scheduled replacement date=counter acquisition date+(replacement interval−counter)counter proceeding average The consumable part having a scheduled replacement date that is nearer than the visit date is decided as the consumable part to be replaced (S704) and is presented together with the visit date by the notifying unit 208 (S705). FIG. 12 shows exemplary output results. Other than the above-described processing, “counter proceeding distribution” in the “machine” table (see FIG. 2) can be utilized to estimate the visit date by period. Next, a second embodiment of this invention will be described. This embodiment is a modification of the above-described first embodiment and the basic system configuration is the same. Hereinafter, the same parts as those described already in the first embodiment are denoted by the same numerals and will not be described further in detail. In this embodiment, the MFP 201 as an apparatus that is a maintenance target has a cartridge in which a photoconductive unit, a charger, a cleaner, a developing unit and the like are integrally formed as a unit. The cartridge can be attachable to/removable from the main body. In such a cartridge with various components integrated therein, replacement is necessary if one of the components constituting the cartridge is broken. Therefore, in the “strategy planning mode”, the “visit interval” and “replacement interval” set in the “consumable part status” table (see FIG. 2) are set to the same value with respect to all the components constituting the cartridge, and then set to minimize the calculated cost. The cartridge can be easily attached and removed, the user can replace it (equivalent to a so-called customer replacement unit or CRU). In view of the cost and efficiency of the maintenance services, it is preferable that the replacement of such consumable parts that can be easily replaced is carried out on the user side, if possible. FIG. 13 is a flowchart showing the flow of processing in the “visit date presentation mode” in the maintenance system according to this embodiment. The processing of S601 to S604 in the flowchart shown in FIG. 13 is similar to the processing of S701 to S704 shown in FIG. 10 in the first embodiment. Therefore, the processing of S605 and the subsequent steps will be described. When the consumable part to be replaced on the visit date for the apparatus that is the maintenance target is decided by the maintenance planning unit 206 (S604), the replacement difficult judging unit 207 judges whether or not a component of the cartridge is included in the list of consumable parts to be replaced (S605). If a component of the cartridge is not included in the list of consumable parts to be replaced on the visit date (No in S605), the decided visit date and the list of consumable parts to be replaced on the visit date are presented by the notifying unit 208 (S608). On the other hand, if a component of the cartridge is included in the list of consumable parts to be replaced on the visit date (Yes in S605), and if the decided visit date is before a preset date (Yes in S606), the visit date is set as “cartridge replacement date”. The replacement difficulty judging unit 207 registers the above-described “cartridge replacement date” to the storage unit 205. The MFP 201 downloads the information of the “cartridge replacement date” stored in the storage unit 205 as additional information at the time of regular communication, and displays a message of cartridge replacement on a control panel, not shown, provided in the MFP 201. Thus, for consumable parts that can be replaced on the user side, the replacement operation is carried out on the user side without having the serviceman visit, and for consumable parts that are difficult to replace on the user side, the user can have the serviceman visit. Thus, improvement in the operation efficiency in the maintenance operation can be realized. FIG. 14 is a flowchart for explaining a schematic flow of processing (maintenance method) in the maintenance system according to this embodiment. The maintenance planning unit (visit interval calculating unit) 206 randomly calculates, for each consumable part, a “visit interval” prescribing a time interval at which a visit should be made for maintenance operation for each consumable part, on the basis of the failure rate distribution of each consumable part (visit interval calculating step) (S101). The maintenance planning unit (replacement interval calculating unit) 206 randomly calculates, for each consumable part, a “replacement interval” prescribing a time interval at which replacement should be made for each consumable part, on the basis of the failure rate distribution of each consumable part (replacement interval calculating step) (S102). In the visit interval calculating step and the replacement interval calculating step, a value close to an interval with which it is predicted that the failure probability is equal to or higher than a predetermined probability, on the basis of the failure probability distribution of each consumable part. The visit interval calculated for each consumable part in the visit interval calculating step is set to be longer than the replacement interval calculated by the replacement interval calculating step. Next, the maintenance planning unit (combination calculating unit) 206 performs search processing using the Monte Carlo method or genetic algorithm on the basis of the “visit interval” calculated in the visit interval calculating step and the “replacement interval” calculated in the replacement interval calculating step, thereby calculating a combination of a visit interval and a replacement interval that minimizes a predetermined cost, of combinations of time intervals at which a visit should be made for maintenance operation and consumable parts that should be replaced during the visit (combination calculating step) (S103). The “predetermined cost” described here is the sum of the labor costs required for the maintenance operation by the serviceman, the material costs of the consumable parts, and the amount of loss caused by the unavailability of the apparatus that is the maintenance target to the user. Then, the maintenance planning unit (interval information acquiring unit) 206 acquires information related to a combination of the “visit interval” and the “replacement interval” associated with the visit interval, which is the information calculated in the combination calculating step (interval information acquiring step) (S104). The counter value acquiring unit 209 acquires a counter value indicating actual use of the consumable parts of the apparatus that is the maintenance target (counter value acquiring step) (S105). The maintenance planning unit (maintenance plan calculating unit) 206 calculates “timing at which the next visit should be made for the apparatus that is the maintenance target” and a “list of consumable parts that should be replaced at the timing”, on the basis of the information acquired in the interval information acquiring step and the counter value acquired in the counter value acquiring step (maintenance plan calculating step) (S106). The replacement difficulty judging unit 207 judges whether the consumable parts that should be replaced, calculated in the maintenance plan calculating step, are components that can only be replaced by the serviceman carrying out the maintenance operation for the apparatus (SRU) or not (replacement difficult judging step) (S107). If it is judged in the replacement difficulty judging step that the consumable parts are components that can only be replaced by the serviceman, the notifying unit 208 issues a notification that a visit should be made for the apparatus that is the maintenance target in order to replace the components (notification step) (S108). In this description, the example where the visit interval calculating step is executed prior to the replacement interval calculating step is described. However, the order is not limited to this as long as both of these processing steps are completed before the execution of the combination calculating step. In this description, the example where the visit interval calculating step to the sense information acquiring step are executed prior to the counter value acquiring step is described. However, it suffices that the processing of the counter value acquiring step and the interval information acquiring step is completed before the execution of the maintenance plan calculating step. Each step in the above-described processing in the maintenance system 1 is realized by causing the CPU 801 to execute a maintenance planning program stored in the memory 802. In this embodiment, the case where the function to carry out the invention has been recorded in advance within the apparatus is described. However, other than this, the similar function may be downloaded to the apparatus from a network, or the similar function stored in a recording medium may be installed into the apparatus. As a recording medium, any form of recording medium that can store a program and that is readable by the apparatus, such as a CD-ROM, may be used. Also, the function acquired in advance by installing or downloading may be realized in cooperation with the OS (operating system) or the like in the apparatus. As described above, according to this embodiment, by setting the two judgment references of “visit interval” and “replacement interval” for individual consumable parts, the serviceman can grasp “when to visit” and “which consumable parts should be replaced”. Also, since a desired visit interval and replacement interval (strategy) is calculated in advance and the next visit date is usually calculated on the basis of the decided strategy, the cost required for the calculation can be reduced. Moreover, since the status of use of the apparatus that is a maintenance target can be gathered in real time, the certainty of prediction of a visit date for the apparatus is significantly improved. This invention has been described in detail by using the specific modes. However, it is obvious to those skilled in the art that various changes and modifications can be made without departing from the spirit and scope of this invention. According to this invention, as described above in detail, a technique can be provided that enables reduction in the cost related to the maintenance services and that also enables reduction the downtime of the product. |
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055457967 | summary | FIELD OF THE INVENTION The present invention relates to an article made from waste materials. The invention also relates to an article for storage, isolation, or other management of "contaminated material", herein defined as radioactive, hazardous, or mixed wastes, and methods of making such an article, where the article itself is made in part from substantial amounts of radioactive, hazardous, or mixed waste materials, such as radioactive metal particulates, radioactive concrete particulates, hazardous waste solids, radioactive liquids, and the like. The term "mixed waste" is herein defined as a combination of radioactive and hazardous waste. The term "hazardous waste" is herein defined as set forth generally in 40 C.F.R. PART 261 "Identification And Listing Of Hazardous Waste", at the time of its use. BACKGROUND AND SUMMARY OF THE INVENTION Recycling of materials is often practiced to conserve raw materials. Contaminated materials are often treated to decontaminate their main constituents to allow reuse of these constituents. In general, isolated contaminants or contaminated articles not readily susceptible to decontamination are intended, in accordance with past practice, to be disposed of in ways that put the emphasis on destruction or isolation rather than on considerations of reuse. What therefore has happened and is continuing to happen is that increasing quantities of clean materials are turned into contaminated materials that then have to be disposed of in safe ways. A wide variety of articles and their uses are discussed herein, one type of which is a container to confine waste. Before the use of modern storage modules, in many instances nuclear waste and hazardous waste were stored in 55-gallon steel drums. In the case of very low level radioactive waste, such drums are still used, many times being overpacked in plastic containers such as polyethylene, polypropylene and the like. More sophisticated, all steel containment systems, having thick steel walls and an adjustable shielding core, are described in U.S. Pat. No. 4,451,739 (Christ et al.). There, exterior or interior lead or steel wire which serves to attenuate of gamma rays, can be wrapped in as many layers as appropriate to the contaminant waste material's radioactivity. More recently, outer waste container systems have been made with uncontaminated concrete reinforced with large, uncontaminated metal bars, called "rebar" construction, or a steel reinforcing mesh basket, to improve the strength of the container. However, use of large metal bars and steel mesh generally require thick walled containers. Examples of such containers are shown in U.S. Pat. Nos. 4,950,426 (Markowitz et al.) and 4,845,372 (Mallory et al.). In these containers, filler is used to seal the void space between the outer container and, for example, compacted steel-walled storage drums which alone, contain the radioactive material or hazardous material. in some instances, various types of fibers are used in place of bars or mesh to reinforce concrete in waste containment vessels. In U.S. Pat. No. 4,995,019 (Cataloyoud et al.), a tight-sealing, drum-covered containment vessel is taught, where cast iron or stainless steel fibers are distributed in a random manner in the concrete container. U.S. Pat. No. 4,167,491 (Gablin et al.) describes disadvantages associated with concrete, the most serious of which is the potential of some radioactive material leaching therefrom. There, water which contains radioactive nickel and cobalt-60 is passed through cation resin exchange beads to concentrate the contaminants, then the wet beads are mixed with an aqueous dispersion of a hydrophilic urea-formaldehyde plastic resin obtaining an acidic curing agent, to form a solid waste block for disposal in a steel or cast iron outer container. In U.S. Pat. No. 4,594,513 (Suzuki et al.), steel or carbon fibers, or metal gauze, are used to reinforce concrete containers, which containers are then impregnated with a polymerizable monomer, to provide a water impermeable lining. Other ingredients that can be added to concrete multipurpose contaminated waste containers having polymeric liners include amorphous metal fibers, fly ash, and silica fume, as taught in U.S. Pat. No. 5,225,114 (Anderson et al.). There, the container needs no exterior concrete overpack barrier and is also transportable and storable. In other instances, hazardous waste is high-density packed within a solidified radiation shielding by centrifugally casting waste material and polyorganic compounds or cementitious material, to form a monolith having high strength and structural integrity, such as taught in U.S. Pat. No. 5,075,045 (Manchak, Jr.). Radioactive wastes can also be classified, segregated, and cast with a shielding material which encapsulates it and prevents the escape of radiation, as taught in U.S. Pat. No. 4,897,221 (Manchak, Jr.) In Atomkernenergie-Kerntechnick, Bd. 41 (1982), No. 4, pp. 279 to 280, "Proposal for the Disposal of Contaminated Steel Parts from Shut Down Nuclear Power Plants", by W. M. Francioni, shielding materials and disposal of highly radioactive material are discussed. This article describes disposal of waste material by using it to form containers for other waste. It discloses, for example, that, theoretically, any suitable material, such as concrete, iron, or lead can be used for the container. It describes in detail a thick concrete transport container, with an additional, separate, inner shielding liner made of melted, cast, solidified, radioactive steel from reactor tubes, and the like. This article also discusses questions regarding Secondary wastes during melting the radioactive steel, and, whether the use of such metal shields would be economical since the metal used, at that time, would have had to be decontaminated to a maximum surface activity of 1 .mu.Ci/cm.sup.2 (37 Bq/g, or 1 nano Ci/g) for transport from the nuclear facility to a melting facility. In this same area of using radioactive components in waste transport containers, U.S. Pat. Nos. 4,767,572 (Sappok) and 4,882,092 (Sappok), issued in 1989, teach use of radioactive residues in the formation of radiation shielding structures. They state that one would normally expect that the last thing which could be tolerated in a shielding material is a substance which itself is radioactive. In one embodiment, 25 weight % radioactive steel from reactor tubes, and the like, are reacted with 75 weight % of uncontaminated cast iron, and the mixture is melted to provide an alloy filler material. However, here, the amount of radioactive waste is tripled by reactive dilution with uncontaminated material. Other embodiments include use of broken up radioactive concrete as a shielding structure alone or in combination with comminuted, radioactive metal alloys. In all instances, however, to minimize detrimental contribution of radiation to the environment, the radioactive material must have a cobalt-60 equivalent between 1 to 100 Bq/g (0.027 to 2.7 nano Ci/g) before being used to make radiation shielding structures, or transport or storage containers. They state that this lower level is two orders of magnitude higher than natural activity levels which they identify as 0.01 Bq/g. It is also known to melt radioactive metal, such as scrap from nuclear power plants and the like, and recast it to form blocks that are used for shielding, such as is described in SEG Brochure "Metal Processing" Feb. 1991 No. 1165-291. What is needed is an article which can effectively utilize high level radioactive waste as well as hazardous and mixed waste for a variety of useful purposes. What is also needed is a method of making a containment system where contaminated material, previously defined as radioactive waste, hazardous waste, or mixed waste can be used as part of the containment system itself, at high radioactivity levels and high hazardous waste levels, without expensive chemical decontamination steps, and without any substantial initial dilution by mixing or reaction with substantial amounts of uncontaminated materials, and where the radioactive, hazardous, or mixed elements in the waste can be fixed in the system, so that leaching is controlled and is not a problem. It is one of the objects of this invention to provide such an article, method, and containment system. The present invention resides, generally, in the concept that contaminated material, of a wide variety, is suitable for use in making a wider variety of articles than heretofore has been recognized. The present invention also resides, generally, in the concept that it is possible to reduce to a more marked extent than previously recognized, the ratio of clean material used, to contaminated material, in making a large number and variety of articles, including but in no way limited to waste containers. The contaminated materials to which this invention is concerned are primarily radioactive, hazardous, or mixed waste. Among the many uses for which this invention is applicable, the following are "articles", which serve to illustrate the scope of this invention: Containers of every conceivable dimension, shape, weight, and capacity for processing temporarily or permanently holding, isolating, disposing, or preserving radioactive or hazardous materials, wastes, waste residues, spent materials, or by-products therefrom; which include radioactive waste or hazardous waste. Shielding casks, pigs, bells, racks, grids, walls, panels, bricks, blocks, shot, sheet, wool, slabs, etc. that use contaminated lead, polyethylene or other plastics, water, depleted uranium, steels, and other metals, or other radioactive waste or hazardous waste. Building structures including structural steel or members (beams, columns, posts), panels, t-sections, hollow core slabs, cinder blocks, bollards, curb stops, floors, floors, footer, skin, or modules thereof; which use contaminated concrete, steels, lead or other metals, or plastics or other radioactive waste or hazardous waste. Linings, insulations, refractories, blankets, coatings that include contaminated materials including materials like contaminated steel fibers, shot, grit, dust and powders. Such linings etc. could be used on rail cars, truck bodies, caissons, open-top waste containers, impoundments, kilns, calciners, secondary combustion chambers, bulk storage facilities, silos, afterburners, quench towers, spray dryers, furnaces, ovens, and other similar thermal or chemical treatment equipment; which use radioactive waste or hazardous waste. Impact limiters on corners, sides, or surrounding a waste container, made from contaminated polymers, plastics, rubber, organic composites (wood or cellulosic fibers), steel or metal shapes (tubes, grids, cages, etc.); which use at least one of radioactive waste and hazardous waste. Rollers, breaks, spindles, shears, cutting and shaping tooling for contaminated materials; which use radioactive waste or hazardous waste. Components for use inside normally contaminated environments including tools, dams, reactor fuel grids, tubes and tube sheets, nozzles, ducts, etc. in nuclear reactors, hot cells, glove boxes, fuel reprocessing facilities, nuclear weapon manufacturing and disassembly facilities, devices holding or containing radioactive sources, etc; which use at least one of radioactive and hazardous waste. Seals, gaskets, o-rings and the like in applications exposed to radioactive or hazardous wastes and made from contaminated polymer, rubber, plastics, steels, metals, and the like; which use at least one of radioactive waste and hazardous waste. Dunnage, shoring, bracing, and the like made from contaminated materials to secure radioactive or hazardous materials or wastes; which use at least one of radioactive and hazardous waste. Water quality systems including settling tanks, ponds, clarifiers, flocculating tanks, sludge beds, grease and grit chambers and the like; which use at least one of radioactive and hazardous waste. Molds for steel ingots, plastic components, concrete shapes, RIM containers, rubber containers, etc; which use at least one of radioactive and hazardous waste. Road materials made with or consisting in part or in whole of contaminated materials such as rubber, concrete or stone, steel or other metals for subgrade, bitum course, precast slabs, level course bitum concrete filler and the like; which use at least one of radioactive and hazardous waste. Conveyers, sluices, tunnels, pipes, galleys, weirs, box culverts, bridges, bridge decks and the like used to convey radioactive or hazardous materials; which use at least one of radioactive and hazardous waste. Metal shapes of all kinds made for equipment supports, stands, piping, components, teeth, clogs, and other wearing components; which use contaminated steel or other metal. Process vessels, glove boxes, conveyors, skid plates, rails, wheels, platforms, grids and catwalks; which use contaminated steel or other metal. Steel wire for reinforcing, welded wire fabric, welded wire cages for holding filters in disposal containers, baghouse bags in place, bows for various tarp-like covers, reinforcing in rubber or polymer parts (conveyor belts, sheets, drop curtains); which use contaminated steel or other metal. Steel and other metal fibers of all shapes, twists, bends, lengths, thicknesses and aspect ratios to reinforce, add bulk, densify, stiffen, strengthen, toughen or otherwise modify various polymer, rubber, concrete, and refractory materials; where the steel or other fibers are contaminated. Filters and membranes, and other applications for containing or excluding radioactive or hazardous materials; which use sintered, contaminated steel or other metal. Boots, gloves, bellows, sleeves, and flexible joints to isolate radioactive or hazardous materials from an environment, or to be used in such an environment such as a hot cell, reactor cavity, glove box, or air lock; which use contaminated polymers or plastics. Targets for depleted uranium or other projectiles, articles being drop tested, items being crushed, obliterated, or made unrecognizable, such as for classified components, munitions, etc; which use contaminated steel, other metals, concrete or plastics. Tanks, tank liners, bearings, skid or slip sheets, insulators, impact limiters, bumpers, and balls; which use contaminated polymers or plastics. Spheres for limiting vaporization from chemical processing tanks; which use contaminated plastic. Sheeting and bags, used extensively to control contamination in nuclear power facilities; which use contaminated thermoplastics. Accordingly, in one of many embodiments, the invention relates to an article characterized in consisting essentially of: waste selected from the group consisting of a) radioactive waste, b) hazardous waste, and mixtures thereof. In one of many embodiments, the article can be a contaminated waste article, made solely from cast, cooled, melted, radioactive metal components, where the melted metal used in the article is preferably substantially free of slag residue and has a specific activity over 130 Bq/g, and where the article consists of unsupported cast metal. We have found, surprisingly, that contaminated metal tubing from, for example nuclear power plants, when melted and separated from slag residue can be used without dilution or alloying with uncontaminated metals, to provide slabs and stand-alone containers for contaminated waste. Even if the contaminated metal has a specific activity level substantially above 100 Bq/g, such as, above 130 Bq/g, it is not detrimental and the metal is still useful. The ability to use materials with high specific activities, such as above 130 Bq/g, minimizes the need for uncontaminated filler or the like in the container or other article. Therefore, other embodiments of the invention are an article, comprising a member, the member consisting essentially of a material selected from the group consisting of a) radioactive waste, b) hazardous waste, and mixtures thereof in a matrix of either concrete binder or plastic resin binder. Also, an article, comprising a member, the member consisting essentially of a material selected from the group consisting of a) radioactive thermoplastic b) hazardous thermoplastic and mixtures thereof; or an article, comprising a member, the member consisting essentially of metal of which more than 35 weight % of the metal is radioactive. Thus, the article could be a member section component or other part structure. We have discovered that hazardous waste and mixed waste can be used alone to make a wide variety of slab, brick, wall or other type articles. Another aspect of the invention resides in an article characterized as consisting essentially of: waste selected from the group consisting of a) radioactive waste, b) hazardous waste, and mixtures thereof, where, if radioactive waste is in metal form, such metal constitutes more than 35 weight % of the article. This article, during use, can and in most cases will be exposed to radioactive or hazardous waste and therefore become further contaminated. However the article, such as a container is not limited to being exposed to or containing waste. The article could be exposed to or contain various chemicals or other materials not considered contaminated or waste, or could be exposed to or contain "fresh" radioactive or hazardous materials. Use of more than 35 weight % metal will provide integrity for the structure and allow its use as a substantially self-supporting container or the like. Preferably when metal is used it will be substantially free of slag residue. A preferred high density containment system with a variety of particulate size gradings has also been discovered, and the invention also resides in a containment system for radioactive, hazardous or mixed waste, characterized by having a structure containing a series of different sized particles to provide high interior void volume filling, where at least one fine particulate selected from the group consisting of silica fume and flyash particles and mixtures thereof is close packed between coarse particulate selected from the group consisting of filler, cement and aggregate particles and mixtures thereof, and also containing additives distributed therethrough, selected from the group consisting of uniformly dispersed bars, fibers, generally spherical particles and amorphous particles, and mixtures thereof, such that the containment system has a density over about 90% of theoretical density. In the above embodiment, the particles and additives can be non-contaminated materials or contaminated materials. The containment system can be thin walled, that is, less than 5 cm thick. It can be a round, square, or other configured storage module, having a bottom, sidewalls and an associated attached lid. The containment system can have a closely attached plastic sheet about 0.2 cm to about 2.5 cm thick, covering at the inside of the system and/or the outside of the system. The contaminated material when used is distributed in the module walls and is not concentrated as a separate inner or outer layer or shield. The article/structure containment system can be a thick or thin wall type structure of various configurations, for example, primarily plastic containers, radioactive containment shielding; a variety of other rigid or flexible structures, including enclosures, dividers, barriers, burial and storage modules, vaults, trench walls and bunkers. We have also found very useful containment systems characterized by a structure containing a material selected from concrete or plastic resin, containing within its walls radioactive metal, in the form of discrete fibers constituting from 2 weight % to 55 weight % of the system having lengths from about 0.5 cm to about 20 cm where the system contains different sized particles to provide high interior void volume filling in the case of concrete, or resin impregnated porous metal mesh or discrete fibers in the case of plastic resin. The invention further resides in a method of making a contaminated waste article characterized by the steps: (A) providing radioactive metal material selected from the group consisting of nickel, chromium, iron, steel, and mixtures and alloys thereof; (B) inspecting said radioactive material to segregate it according to metal type, to provide a metal feed; (C) transporting the radioactive metal feed to a melting furnace operating at a temperature over 1400.degree. C., to melt the radioactive metal feed and form a top impure radioactive slag phase if the feed is impure, and generally a lower level radioactive molten metal phase; (D) casting the radioactive molten metal phase into a radioactive article substantially free of the slag phase; and (E) cooling the cast article to provide a solid radioactive article. These articles could be bricks, wall structures, slabs, containers or the like. They could be transportable and placed in direct or indirect contact with contaminated material. Also the slag phase could be cast into a radioactive article. The invention also resides in a method of mixing radioactive, hazardous, or mixed waste into a binder matrix to form a containment system for additional, highly concentrated radioactive or hazardous waste characterized by the steps of: (A) providing a contaminated material selected from at least one of: (i) radioactive material in small discrete form, (ii) hazardous waste material in small discrete form; and (iii) mixed waste in small discrete form; (B) mixing thoroughly: (i) about 100 parts by weight of a binder material and (ii) about 2 to about 570 parts by weight of the contaminated material to which no more than about 15 weight % of uncontaminated material has been mixed, to provide a homogeneous composition, in which the contaminated material is in discrete, non-agglomerated form throughout the binder; (C) forming the composition into a unitary, solid containment system which contains contaminated material, and binder acting as a matrix for the contaminated material; and (D) placing the containment system in direct or indirect contact with highly concentrated, radioactive, hazardous, or mixed waste. The invention very specifically, also resides in a method of making a structure for radioactive, hazardous or mixed waste, utilizing radioactive, hazardous or mixed waste as a component of the structure, characterized by the steps: (A) providing quantities of radioactive material selected from the group consisting of radioactive metal, radioactive concrete, radioactive sand, radioactive gravel, radioactive plastic, radioactive liquid, and mixtures thereof; (B) processing the radioactive material without dilution with any more than about 15 weight % of nonradioactive material, to provide at least one of (i) bars, (ii) fibers, (iii) generally spherical particles, (iv) amorphous particles, (v) sheet plastic, and (vi) stabilized liquids; (C) mixing (i) 100 parts by weight of a binder material and (ii) 0 to about 25 parts by weight of hazardous waste material selected from the group consisting of hazardous solids, hazardous liquids and mixtures thereof, to which is then added (iii) about 2 to about 570 parts by weight of the processed, radioactive material, to provide a homogeneous composition; and (D) forming the composition into a unitary, solid structure. The term "amorphous" as used herein means not having a standard geometrical shape or having an irregular shape. Most advantageously, the invention resides in a container characterized by having concrete and from 2 weight % to 55 weight % contaminated metal fibers having lengths from 0.5 cm to about 20 cm and a length:width aspect ratio of between 200:1 and 20:1 where the container contains different sized particles to provide high interior void volume filling and a high density, generally over about 90% of theoretical density. The term "high interior void volume filing" as used herein means most voids are filled, resulting in a low porosity low void structure. The initial contaminated material provided can include, radioactive stainless steel tubes used in cooling nuclear reactors, which have been cut into small pieces, or melt cast into small fibers; radioactive concrete chunks and dust resulting from demolition of or around nuclear reactor structures; "plastic" or "plastic resin" which is meant herein to also include rubber sheets or gloves used to deal with hazardous materials, and other materials described later in the specification; ion exchange resins, beads, powders or slurries used in purification processes; powdered hazardous soil; polychlorinated biphenyls; and the like. The contaminated materials are usually processed by one or more of cutting, grinding, shearing, heating, melting, melt-casting, pressing, and the like, to form small pieces or particulates less than about 50 mm diameter, or squares or fibers less than about 20 cm long. Preferably, only a small portion, of "uncontaminated material", herein defined as virgin, non-contaminated and non-radioactive material is mixed or reacted during processing, so that the volume of contaminated material is not substantially increased prior to mixing with binder. The binder material can be, for example metal, plastic, or a mixture of sand, aggregate and cement, with the possible addition of silica fume, flyash, and plasticizer. It is essential to thoroughly mix and disperse the contaminated material into its binder, so that the binder forms a matrix containing and firmly binding the discrete pieces or particles of contaminated material. In most cases where the binder is concrete, the cement used will be clean and non-contaminated so that good bonding is achieved, the same is true if a thermoset resin, such as an epoxy resin, is used as the binder. Also, contaminated thermoset resin cannot be remelted and would not provide good bonding. As distinguished from U.S. Pat. No. 5,402,455 (Angelo et al.), dealing primarily with fiber mesh reinforcement in a concrete container, this invention deals primarily with discrete particles of contaminated material in a wide variety of articles. In order to uniformly disperse radioactive fibers and particles, when they are used, and prevent clumping/agglomeration and thus concentration of radioactive material, the fibers are preferably processed to within narrow length:width aspect ratios and the particles are preferably processed to within narrow particle sizes and gradation. In the case of concrete binder material, they are preferably combined with chemical plasticizer when incorporated into the binder material. Preferably, a series of different sized contaminated material and binder materials are provided, to allow a close, high density packing and elimination of most void volume in the cured system. The term "radioactive" as used herein is defined to mean a level of activity, due to contamination or activation by, for example, radio-isotopes of cobalt, lead, cadmium, cesium, barium, or the like, ranging from 0.1 nano Ci/g to well over 10 nano Ci/g (3.7 Bq/g to well over 370 Bq/g). The contaminated materials used in substantially large volumes in this invention would otherwise have limited usefulness, and only be suitable for direct disposal in the absence of some dramatic technology that could remove the contamination. The containment system can contact or hold loose contaminated material directly, or contaminated material placed in standard or compressed steel drums, plastic containers, or the like, so that there is "indirect" contact of the containment system with the waste through contact with the steel drum or plastic container wall. The processes of this invention as previously described provide an article, such as a containment system, which has low permeability to water, excellent leach resistance, and, additionally in the case of concrete, high tensile and compressive strength; and which can contain from 2 to 570 parts of radioactive or hazardous material per 100 parts of matrix material, providing a major means for disposal of radioactive or hazardous waste. In the case of a container for contaminated waste, this invention allows an increased payload of contaminated material in the walls of the container of from 10 weight % to 100 weight %, based on contained contaminated material filling the container, usually in the form of compressed drums of contaminated material. Thus, a substantial number of waste containers can be eliminated in transport and burial operations, representing tremendous savings, and minimizing transport operations. |
claims | 1. A method for identifying sections of an existing schematic that are consistent with design practices, the method comprising the steps of:providing a template set, each template specifying a sub-set of components and relationships that are consistent with design practices; andusing a processor for examining the existing schematic to identify sections of the existing schematic that are inconsistent with the design practices specified in the template set;wherein the section that is inconsistent with the design practices is an inconsistent section, the method further including the step of, when a section of the existing schematic is inconsistent with the design practices specified in the template set, performing a function on the existing schematic; andwherein the function includes visually displaying the inconsistent section in a distinguishing manner. 2. The method of claim 1 wherein the existing schematic is an electrical schematic and wherein the step of providing a template set includes providing templates that specify both electrical icons corresponding to electrical components and relationships between the electrical icons. 3. The method of claim 1 wherein the design practices are best design practices. 4. A method for identifying sections of an existing schematic that are consistent with design practices, the method comprising the steps of:providing a template set, each template specifying a sub-set of components and relationships that are consistent with design practices; andusing a processor for examining the existing schematic to identify sections of the existing schematic that are inconsistent with the design practices specified in the template set;wherein the section that is inconsistent with the design practices is an inconsistent section, the method further including the step of, when a section of the existing schematic is inconsistent with the design practices specified in the template set, performing a function on the existing schematic; andwherein the function includes automatically identifying a template that indicates a possible replacement for the inconsistent section and automatically providing at least a section of the identified template. 5. An apparatus for identifying sections of an existing schematic that are consistent with design practices, the apparatus comprising:a database storing a template set, each template specifying a sub-set of components and relationships that are consistent with design practices; anda processor linked to the database and receiving the existing schematic, the processor programmed to examine the existing schematic to identify sections of the existing schematic that are inconsistent with the design practices specified in the template set;wherein the section that is inconsistent with the design practices is an inconsistent section, the processor further programmed to, when a section of the existing schematic is inconsistent with the design practices specified in the template set, perform a function on the existing schematic; andfurther including an interface, wherein the function performed by the processor when a section of the existing schematic is inconsistent with the design practices includes visually displaying the inconsistent section in a distinguishing manner via the interface. 6. The apparatus of claim 5 wherein the existing schematic is an electrical schematic and wherein the template set stored in the database includes templates that specify both electrical icons corresponding to electrical components and relationships between the electrical icons. 7. The apparatus of claim 5 wherein the design practices are best design practices. 8. An apparatus for identifying sections of an existing schematic that are consistent with design practices, the apparatus comprising:a database storing a template set, each template specifying a sub-set of components and relationships that are consistent with design practices; anda processor linked to the database and receiving the existing schematic, the processor programmed to examine the existing schematic to identify sections of the existing schematic that are inconsistent with the design practices specified in the template set;wherein the section that is inconsistent with the design practices is an inconsistent section, the processor further programmed to, when a section of the existing schematic is inconsistent with the design practices specified in the template set, perform a function on the existing schematic; andfurther including an interface, wherein the function performed by the processor when a section of the existing schematic is inconsistent with the design practices includes automatically identifying a template that indicates a possible replacement for the inconsistent section and automatically providing at least a section of the identified template via the interface. 9. A method for identifying sections of an existing schematic that are consistent with design practices, the method comprising the steps of:providing a template set, each template specifying a sub-set of components and relationships that are consistent with design practices; andusing a processor for examining the existing schematic to identify sections of the existing schematic that are inconsistent with the design practices specified in the template set;wherein a section that is inconsistent with the design practices is an inconsistent section, the method further including the step of, when a section of the existing schematic is inconsistent with the design practices specified in the template set, visually displaying the inconsistent section in a distinguishing manner;wherein, when a section of the existing schematic is inconsistent with the design practices specified in the template set, automatically identifying a template that indicates a possible replacement for the inconsistent section and automatically providing at least a section of the identified template. 10. The method of claim 9 wherein the existing schematic is an electrical schematic and wherein the step of providing a template set includes providing templates that specify both electrical icons corresponding to electrical components and relationships between the electrical icons. 11. The method of claim 9 wherein the existing schematic is an electrical schematic and wherein the step of providing a template set includes providing templates that specify both electrical icons corresponding to electrical components and relationships between the electrical icons. 12. The method of claim 9 wherein the design practices are best design practices. |
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056065868 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Embodiment 1 The first embodiment will be described. FIG. 3 shows an X-ray exposure apparatus using a convex mirror, and more particularly, a situation in which an SR-X-ray L0 which is an X-ray radiation ray along an optical path indicated by a solid line when an exposure operation is started, changes to SR-X-ray L1 along an optical path indicated by broken lines after the exposure is interrupted and orbit electrons are injected. The SR-X-rays L0 and L1 are emitted from a point of emission 1 of the charged particle accumulation ring and are enlarged in a y-axis direction perpendicular to an orbit plane of the charged particle accumulation ring by a convex mirror 2. Then, the X-rays are introduced into a pressure-reduced chamber not shown through an X-ray transmitting film, and are incident on a substrate 5 supported on a substrate stage 4. At the upper end of the substrate stage 4 shown in the Figure, an X-ray detector 6 is disposed, and a shutter 7 for controlling the exposure period is disposed between the X-ray transmitting film 3 and the substrate stage 4. Before the start of the exposure, the substrate stage 4 is moved downwardly in the Figure, and an X-ray intensity distribution V0 (y), in the y-axis direction, of the SR-X-ray L0 (the solid line in FIG. 4A) is measured by an X-ray detector 6. Since an output Vx (y, t) of the X-ray detector 6 changes in proportion to an orbit current I(t), it is normalized (I(t)) so that the X-ray intensity distribution V(y) is constant relative to time, as follows: EQU V(y)=Vx(y, t)/I(t). The orbit current I(t) is measured by an ammeter such as a DCCT (not shown) in the charged particle accumulation ring. Subsequently, the substrate 5 is exposed, and the resist is processed, and then, the distribution of the exposure amount is calculated on the basis of the remaining film ratio of the resist and a resist characteristics curve predetermined. On the basis of that, the exposure intensity distribution DO (y, t.sub.0) (the solid line in FIG. 4B) is determined. In addition, by an ammeter such as a DCCT or the like, an initial orbit current I(t.sub.0) of the charged particle accumulating ring 1 is measured. Then, the exposure period t.sub.exp at the time t during the exposure and at an exposure position y by the following equation (1), and on the basis of this, a moving speed curve of a shutter 7 is set, and the mask pattern is printed on a substrate 5 for a semiconductor device. ##EQU2## where, I(t) is an orbit current of the charged particle accumulating ring measured at the time t, and C.sub.0 is a constant on the basis of a set exposure amount. The orbit current I(t) of the charged particle accumulating ring at the time t may be measured by a known ammeter such as a DCCT or the like as with the initial orbit current I(t.sub.0). Alternatively, an X-ray sensor is disposed adjacent to the shutter 7, and it may be determined from a variation of the output thereof. When the orbit electrons of the charged particle accumulation ring reduce, re-injection is carried out. Then, the optical path of the SR-X-ray may change due to a change of the point of emission 1 of the charged particle accumulation ring and/or a radiation angle of the SR-X-ray. If this occurs, the exposure intensity distribution D1 (t.sub.1) of the changed SR-X-ray L1 is estimated in the following manner, and the moving speed curve of the movable shutter 7 is corrected on the basis thereof. First, the substrate stage 4 is moved down in the Figure, and the X-ray intensity distribution V1 (y) of the changed SR-X-ray L1 (indicated by a broken line in FIG. 4A) is measured by an X-ray detector 6. Then, a difference .DELTA.y between a peak position Y.sub.p0 of the X-ray intensity distribution V0 of the SR-X-ray L0 upon a start of the exposure and a peak position Y.sub.p1 of the X-ray intensity distribution V1 of the changed SR-X-ray L1 is determined. The exposure intensity distribution D1 (y, t.sub.1) of the changed SR-X-ray L1 (indicated by a broken line in FIG. 4A) is estimated as follows: ##EQU3## Thus, the change in the exposure intensity distribution due to the change of the optical path of the SR-X-ray is deemed as being derived from the same positional deviation which is the same as the positional deviation ay of the X-ray intensity distribution. Using the equations (1) and (2), the exposure period t.sub.1exp after the change is determined by the following equation (3), and the shutter 7 moving speed curve is corrected thereby: ##EQU4## In place of correcting the shutter 7 moving speed curve, the substrate stage 4 may be shifted by -.DELTA.y. An example of a calculation will be described. An X-ray exposure apparatus is taken in which an orbit radius of the charged particle accumulation ring is 0.6 m; and an accelerating voltage is 800 MV; a radius of the mirror is 50 m; a reflection angle is 15 mrad; a distance between the point of emission of the charged particle accumulation ring and the mirror is 7 m; a distance between the mirror and the X-ray transmitting film is 4 m; a distance between the X-ray transmitting film and the substrate is 0.8 mm; the resist material is PMMA; the mask material is SiN; an X-ray detector is GaAs; and the exposure area is .+-.15 mm. When a radiation angle of the SR-X-ray is changed by 0.01 mrad as a result of the orbit electron injection into the charged particle accumulation ring, the X-ray intensity distribution V1 (y, t.sub.1) is detected by the X-ray detector. Several plots of the measurements adjacent the peak are approximated by a quadratic equation curve, and the peak position is determined using the least square approximation. The comparison is made with the peak position of the X-ray intensity distribution V0 measured before the start of the exposure, and the positional deviation Ay is determined. Using equation (2), the exposure intensity distribution D1 (y, t.sub.1) is estimated by equation (2). The actual exposure intensity distribution is determined on the basis of the actual printing on the substrate, and the error therefrom is as shown in FIG. 5 and is not more than .+-.0.2%. The broken line shows an error of the exposure intensity distribution DO (y, t.sub.0) measured by the actual printing before the exposure start relative to the exposure intensity distribution measured by the actual printing after the orbit electron injection. The average is .+-.4%. As another method of calculating the positional deviation .DELTA.y of the peak position in the X-ray intensity distribution, an error .epsilon..sub.n is determined by the following equation, and it may be determined from .DELTA.y when the error is minimum. ##EQU5## The correction required when the exposure intensity distribution expands or reduces in the y direction together with the change of the peak position of the exposure intensity distribution may be accomplished in the following manner. When the X-ray intensity distributions measured before and after the variation can be approximated by the following equations (5) and (6), and the exposure intensity distribution before the change can be approximated by a quadratic equation (7): EQU V0 (y)=a.sub.0 (y-y.sub.p0).sup.2 +V.sub.p0 (5) EQU V1 (y)=a.sub.1 (y-y.sub.p1).sup.2 +V.sub.p1 (6) EQU D0 (y.sub.1 t.sub.1)=(b.multidot.D(y-y.sub.4).sup.2 +D.sub.p).multidot.f(t)(7) then, the exposure intensity distribution after the change is as follows: ##EQU6## where .DELTA.y=y.sub.p1 -y.sub.p0, and f(x) is a value dependent on the time D1 (y, t). When the X-ray intensity distributions before and after the change are Gaussian distributions defined by equations (9) and (10), and the exposure intensity distribution before the change is expressed by equation (11), the exposure intensity distribution is corrected by equation (12): ##EQU7## If the X-ray intensity distribution changes so great that it cannot be approximated by a particular equation or function, the following correction will be made. If the original X-ray intensity distribution V0 (y) having the peak at y.sub.p0 changes to an X-ray intensity distribution V1 (y) having the peak at y.sub.p1, the exposure intensity distribution D1 (y, t) is corrected to: ##EQU8## The X-ray detector is slightly damaged when it is irradiated with the X-rays. The degree of the damage is dependent on the total dose, and therefore, the amount of X-rays incident thereon is preferably reduced. For this purpose, a filter may be provided in front of the X-ray detector. However, the reflectance of the mirror relative to the SR-X-rays is dependent on the wavelength and the angle of incidence, and therefore, the position of the peak is deviated upwardly with an increase of the thickness of the filter even to the extent that it is out of the exposure region. FIG. 6 shows a relationship between the transmissivity and the peak position of the X-ray intensity distribution in the case of an aluminum filter, a Cr filter, and a Ti filter. Therefore, it is preferable that the thickness of the filter is selected in accordance with the material of the filter used so that the peak position remains in the exposure region. This embodiment is usable with a swingable mirror type X-ray exposure apparatus as shown in FIG. 7. In this case, sheet-like X-rays L are swung as indicated by an arrow by swinging a flat surface mirror 200, thus expanding the X-ray irradiation area. By changing the swinging speed of the mirror, the exposure period is controlled. In the beforehand measurement of the X-ray intensity distribution, the stage is fixed such that the X-ray detector 6 is at a proper position in the exposure region, and the mirror is swung. Then, the X-ray intensity is measured, and the stage is sequentially moved to determine the X-ray intensity distribution in the exposure area. On the basis of the X-ray intensity distribution thus determined, the X-ray intensity distribution is corrected through the method described above, and the mirror swinging speed is controlled to provide a proper exposure period. This embodiment is applicable to a stage scanning type exposure apparatus in which a mask and a wafer are fixed on a stage, and the stage is moved to expand the exposure area in effect. In this case, the stage moving speed may be corrected to provide the proper exposure period in accordance with the corrected exposure intensity distribution. According to this embodiment, when the SR-X-ray optical path changes due to the variation or vibration of the radiation angle due to the orbit electron injection and the displacement of the point of emission due to the temperature change or the like, the change of the exposure intensity distribution is estimated through a simple method without measuring the exposure intensity distribution, again, and therefore, the shutter moving speed curve or profile can be properly corrected. As a result, the exposure non-uniformity attributable to the change of the optical path of the SR-X-rays can be easily reduced. Embodiment 2 FIG. 8 shows an X-ray exposure apparatus according to a second embodiment of the present invention. The SR-X-rays 12 emitted from the SR generator 11 functioning as an X-ray source are incident on a cylindrical mirror (convex mirror) 13 of SiC having a radius of curvature R=56.7 m with an inclined incident angle of 15 mrad. The mirror 13 has a convex configuration because the function thereof is to expand the SR radiation 12 in the form of a sheet when emitted from the SR generator 11. Therefore, the curvature is so formed that the surface is away from the SR generator 11. The SR radiation 14 reflected by the mirror 13 is transmitted through a reticle in the form of a transmission type mask 17 having an X-ray transmission film on which a desired pattern is formed by an X-ray absorbing material, so that the desired pattern of the X-rays is incident on a substrate (wafer) 18 coated with resist material sensitive to the X-rays. Upstream of the mask 17, there is disposed a shutter 15 for controlling the exposure period over the entire exposure area. The shutter 15 is driven by a shutter driving unit 16 controlled by a shutter control unit 21. On the wafer stage 19, an X-ray detector 20 is disposed. At a light receiving portion of the X-ray detector 20, there is a pin-hole having a diameter of 0.7 mm. Although not shown in the Figure, a thin Be film having a thickness of 12 .mu.m is disposed downstream of the mirror 13 and upstream of the shutter 15. The upstream side of the thin film is under ultra-high vacuum, and the downstream side thereof is in a pressure reduced He environment. The description will be made as to the method of obtaining an exposure intensity distribution on the basis of the remaining film ratio of the resist material. When the exposure is carried out under constant conditions except for changing the exposure amount, the film remaining ratio is a function of the amount of exposure in the case of a negative resist. Conversely, the exposure amount is a function of the film remaining ratio. In view of this, at a predetermined position in the exposure area and at a constant accumulated current, several exposure operations are carried out with only the exposure period being changed, and thereafter, the resist is developed, and then the film remaining ratio is determined. By doing so, the functional relation between the film remaining ratio and the exposure period is determined. If it is assumed that the film remaining ratio 90% is the optical exposure, the exposure amount corresponding to the exposure period is the optimum amount of the exposure. Therefore, the entire exposure area is exposed to the radiation for a constant period of time at the accumulated current, and thereafter, the resist is developed, and the film remaining ratio is determined. By doing so, the exposure amount distribution is determined in the form of a ratio relative to the optimum exposure amount over the entire exposure area. If the exposure amount thus determined is divided by the exposure period, the exposure intensity distribution is determined. If a positive resist material is used, the same analysis is possible using the fact that the thickness of the remaining film is a function of the exposure amount. The exposure time period is not necessarily constant all over the exposure area, but the exposure period may be any if it is known. FIG. 9 shows an exposure intensity distribution (solid line) in a unit mW/cm.sup.2 and an output (broken line) of an X-ray detector 20 placed on a wafer stage 19. The exposure intensity is energy per unit volume and unit time. In this embodiment, the thickness of the resist is 1 .mu.m, and therefore, the unit mW/cm.sup.2 is used in place of mW/cm.sup.2 /.mu.m in the sense that the energy is absorbed by the 1 .mu.m-thick resist per unit volume. As will be understood from this Figure, the profile of the output of the X-ray detector 20 is significantly different from the exposure intensity distribution, and therefore, the output of the X-ray detector 20 cannot be deemed as the exposure intensity distribution. In the X-ray exposure, the relative positional deviation among the point of emission, at least one mirror and the mask and the wafer results in exposure non-uniformity. FIG. 10 shows an exposure intensity distribution determined by a film remaining ratio of the resist material when the mirror 13 changes its position by 10 .mu.m and 50 .mu.m in the y direction (the direction normal to the mirror surface, that is, inclined by 15 mrad relative to the normal line of the SR orbit plane). The solid line represents no position change: a dot line represents a 10 .mu.m deviation; and a broken line represents a 50 .mu.m position change. As will be understood, the maximum exposure intensity change from a 10 .mu.m positional change is approx. 0.4%, and the maximum exposure intensity change resulting from a 50 .mu.m positional change is approx. 2%. Therefore, without the means for measuring the exposure intensity distribution, the 0.4% and 2% exposure non-uniformities occur with the result of reduced yield. On the other hand, it is not practical to determine the exposure intensity distribution on the basis of the film remaining ratio of the resist material whenever the position change occurs, because there are many causes of the position change, and the amounts thereof are not constant. The exposure amount intensity distribution in the exposure area and the output of the X-ray detector 20 have substantially a one dimensional intensity distribution in a direction (y-direction) perpendicular to the SR orbit plane in this embodiment. When an exposure intensity distribution is DO (y), and an output of the X-ray detector 20 is O.sub.0 (y) under certain conditions, the coefficient of proportion A (y) is: EQU A (y)=D0 (y)/O.sub.0 (y) (14) FIG. 11 shows the coefficient of proportion A (y). The coefficient is obtained without the position change of the mirror 12. FIG. 12 shows the exposure intensity distribution and the output of the X-ray detector 20 when the mirror 13 changes its position by 10 .mu.m in the y direction, and FIG. 13 shows the exposure intensity distribution and the output of the X-ray detector 20 when the mirror 13 changes its position by 50 .mu.m in the y direction. An exposure intensity distribution and an output of the X-ray detector 20 under a condition different from that when the coefficient of proportion A (y) is determined by equation 14, are D1 (y), and O.sub.1 (y). Then the D1' (y) is defined as the output O.sub.1 (y) multiplied by the coefficient A (y). EQU D1'(y)=A (y).times.O.sub.1 (y) (15) The exposure intensity distribution D1' (y) is different from D1 (y), but it is determined on the basis of the coefficient A (y) predetermined under a condition and an output O.sub.1 (y) of the X-ray detector 20 during exposure, and therefore, it can be even more quickly and easily than when the exposure intensity distribution D1 (y) is determined on the basis of the remaining resist film ratio after the exposure operation. FIG. 14 shows an exposure intensity distribution D1 (y) and D1' (y) when the mirror 13 displaces by 10 .mu.m in the y direction. The difference between D1 (y) and D1' (y) is 0.04% over the entire exposure area. FIG. 15 shows exposure intensity distributions D1 (y) and D1' (y) when the mirror 13 displaces by 50 .mu.m in the y direction. The difference between D1 (y) and D1' (y) is 0.2% over the entire exposure area. Similarly, even if the mirror 13 changes its position by an unknown distance in the y direction, the exposure intensity distribution can be determined with such a high precision that the exposure non-uniformity is tolerable, by measuring the output O.sub.1 (y) by the X-ray detector 20 and multiplying it by the coefficient A (y), even if the displacement of the mirror 13 is unknown. FIG. 16 shows an exposure intensity distribution determined by the remaining resist film ratio when the mirror 13 rotates through 10 .mu.rad and 50 .mu.rad about an x-axis, a direction perpendicular to the emitting direction of SR radiation 2 in the SR orbit plane. The solid line represents no-change; the dot line represents a 10 .mu.rad case; and the broken line represents a 50 .mu.rad case. The exposure intensity distribution is approx. 0.3% at the maximum with a 10 .mu.rad rotation, and is approx. 1.7% at the maximum with a 50 .mu.rad rotation. Therefore, without the means for measuring the exposure intensity distribution, the exposure non-uniformities of 0.3% and 1.7% result. FIG. 17 shows an exposure intensity distribution and an output of the X-ray detector 20 when the mirror 13 rotates through 10 .mu.rad, and FIG. 18 shows an exposure intensity distribution and an output of the X-ray detector 20 when the mirror 13 rotates through 50 .mu.rad. The broken line represents the exposure intensity distribution D1' (y) determined by multiplying the output of the X-ray detector 20 when the mirror rotates through 10 .mu.rad by the coefficient A (y) shown in FIG. 11, and the solid line represents the exposure intensity distribution D1 (y) with the 10 .mu.rad rotation, in FIG. 19. In FIG. 20, a broken line represents the exposure intensity distribution D1' (y) determined by multiplying the output of the X-ray detector 20 when the mirror 13 rotates through 50 .mu.rad by the coefficient A (y), and the solid line represents the exposure intensity distribution D1 (y) when it is rotated through 50 .mu.rad. As will be understood from FIG. 19, the exposure intensity distributions D1 (y) and D1' (y) are in accord with each other with the errors of 0.04% and 0.2% at the maximum over the entire exposure area. Similarly, even if the mirror 13 is rotated through an unknown distance about the x-axis, the exposure intensity distribution can be determined with such an accuracy that the exposure non-uniformity is tolerable by measuring the output O.sub.1 (y) of the X-ray detector 20 and multiplying it by the coefficient A (y), even if the angular position of the mirror 13 changes. Thereafter, the proper exposure period is calculated at each point in the exposure area, and the speed of the shutter 15 in the exposure area is determined by a shutter control unit 21 so that the shutter 5 is opened for a proper exposure period to expose the resist properly. A method of determining the speed of the shutter 15 in the exposure area when the proper exposure period for each point in the exposure area is given, and driving the shutter 15, is disclosed in Japanese Laid-Open Patent Application No. 243519/1989. In this embodiment, the coefficient of proportion A (y) is determined under the condition that the mirror 13 does not make its position change. However, the accuracy of the exposure intensity distribution is hardly influenced even if the coefficient A (y) is determined under the condition that the mirror 13 changes in the y-direction by an unknown amount, or even if the coefficient A (y) is determined under the condition that it is rotated about the x-axis by an unknown amount. The present invention is applicable to such cases. In the case other than the case of the change in the y position or rotational position Ax of the mirror, the exposure intensity distribution can be determined with such an accuracy that the exposure non-uniformity is tolerable by measuring the output O.sub.1 (y) of the X-ray detector and multiplying it by the coefficient A (y). This applies to the position change of the point of emission, mask or wafer. When the mirror 13 rotates about the y-axis or the z-axis, the X-ray intensity distribution and the exposure intensity distribution are two dimensional distributions. At this time, the exposure intensity distribution can be determined by the following equation: EQU D1'(x, y)=A (y).times.O.sub.1 (x, y) (16) Embodiment 3 The shape of the electron beam and the angular component of the speed of the SR generator 11 are both a Gaussian distribution or substantially a Gaussian distribution. It is assumed that a position, in a direction perpendicular to the SR orbit plane, of the electrons in the electron beam of the SR generator 11 is y, and an angular component, in a direction perpendicular to the SR orbit plane, of the speed of the electrons is y' and the standard deviations of the variations are .sigma.y and .sigma.y'. The SR generator apparatus 11 is such a light source that the accumulation current attenuates with time in an exponential function fashion, but it has recently been found that the standard deviations .sigma.y and .sigma.y' change with the accumulation current. In FIG. 21, the solid line indicates the exposure intensity distribution when the accumulation current is 300 mA (.sigma.y=0.8 mm, .sigma.y'=0.3 mrad) and when it is 200 mA (.sigma.y=0.74 mm, .sigma.y'=0.26 mrad). The higher intensity represents the 300 mA case. The broken line indicates the exposure intensity distribution multiplied by 2/3 when the accumulation current is 300 mA. Since the standard deviations .sigma.y and .sigma.y' change with accumulation current, the exposure intensity distribution of the 200 mA case is different from the exposure intensity distribution of the 300 mA case multiplied by 2/3. For this reason, it is desirable that the exposure intensity distribution is determined at each accumulation current with such an accuracy that the exposure non-uniformity is tolerable. FIG. 22 shows a ratio A (y) of the exposure intensity distribution at 300 mA of the accumulation current and the output of the X-ray detector 20. In FIG. 23, a solid line shows an exposure intensity distribution D1 (y) when the accumulation current is 200 mA, and a broken line shows D1' (y) obtained by multiplying A (y) by the output O.sub.1 (y) of the X-ray detector when the accumulation current is 200 mA. They are different by 0.4% approximately adjacent to the center of the exposure. If the exposure intensity distribution is determined by multiplying the output of the X-ray detector 20 by A (y), it can be determined with 0.4% error. Embodiment 4 FIG. 24 illustrates an X-ray exposure apparatus according to a fourth embodiment of the present invention. The SR radiation 23 emitted from the SR generator 22 functioning as the X-ray source is incident on a swingable flat surface mirror 26 of S1 disposed at a position 3 m away from the point of emission with an angle of inclined incidence of 11-19 mrad. The SR radiation 23 in the form of a sheet is expanded. The SR radiation 27 reflected by the mirror is transmitted through a reticle in the form of a transmission type mask 30 having an X-ray transmission film on which a desired pattern is formed by an X-ray absorbing material, so that the desired pattern of the X-rays is incident on a substrate (wafer) 31 coated with resist material sensitive to the X-rays. In front of the mask, there is an opening (Be window 28) movable in synchronism with a mirror 26, and the thin film of Be having the thickness of 12 .mu.m functions as a vacuum isolator. The upstream side of the thin film is under the ultra-high vacuum, and the downstream side is in a pressure reduced He environment. In order to assure sufficient strength against the pressure difference, the Be window 28 has a width of 10 mm in the y-direction (the direction perpendicular to the SR orbit plane). It is vibrated in synchronism with the mirror 26 so as not to block the SR radiation 27 when the mirror 26 vibrates. When the SR radiation 27 is expanded to cover the exposure area of the wafer 31 by swinging the mirror, the exposure intensity is defined on the basis of the exposure amount when the SR radiation 27 in the form of a sheet swings at a constant speed on the wafer substrate 31. In other words, the exposure intensity is determined on the basis of the remaining resist film ratio or the line width accuracy after such an exposure operation. The inclination, relative to a horizontal plane, of the SR light 23 emitted from the point of emission of the SR generator 22 changes, when, for example, the temperature distribution changes in a space in which the SR generator 22 is placed. More particularly, it rotates about the X-axis in FIG. 24. FIG. 25 shows an exposure intensity distribution when the emitting direction of the SR radiation 23 changes by .DELTA..omega.x=0.05 mrad or 0.15 mrad, together with the exposure intensity distribution without these changes. The higher intensity represents the 0.05 mrad case. This is on the basis of the exposure amount provided by one swinging operation at the constant speed of 40 mm/sec of the sheet-like SR radiation 27 on the wafer substrate 31 by swinging motion of the mirror 26. By rotation of the SR radiation 23 emitting direction by 0.05 mrad and 0.15 mrad, the exposure intensity distribution changes by 0.4% and 2.7%. Therefore, when the change of the exposure intensity distribution is not detected, the exposure non-uniformities of 0.4% and 2.7% result. FIG. 26 shows the output of the X-ray detector 33 when the SR radiation 23 emitting direction changes by .DELTA..omega.x=0.05 mrad and 0.15 mrad, together with the output of the X-ray detector 33 without the change. The higher intensity represents the 0.05 mrad case. FIG. 27 shows a ratio A (y) of the exposure intensity distribution and the X-ray detector 33 when the emitting direction of the SR light 23 does not change. FIG. 28 shows the exposure intensity distribution and a value of D1' (y) which is the output of the X-ray detector 33 O.sub.1 (y) multiplied by a coefficient of proportion A (y). In the exposure region (-15 mm-15 mm), they are in accord with an accuracy of 0.2%. In FIG. 29, there is shown the exposure intensity distribution and a value of D1' (y) which is the output O.sub.1 (y) of the X-ray detector 33 multiplied by the coefficient A (y) when the emitting direction of the SR radiation 23 rotates by .DELTA..omega.x=0.15 mrad. They are in accord with each other with an accuracy of 0.5% in the exposure region. This embodiment is more efficient when the positional change is larger. The exposure amount control during the swinging motion of the mirror is effected on the basis of the following. When the emitting direction of the SR radiation 23 is not changed under the condition that 6 mJ/cm.sup.2 is required to expose the resist having a thickness of 1 .mu.m, for example, the exposure amount is 3.47 mW/cm.sup.2 at y=0 mm when the speed of the swinging motion is 40 mm/sec on the wafer substrate 31. Therefore, the mirror 26 is driven by a mirror driving unit 25 through a mirror control unit 24 so as to provide the speed of 23.1 (mm/sec)=40/(6/3.47), at y=0 mm. Generally, the swinging motion of the mirror 26 is controlled so as to provide the speed of 40/(6/E) (mm/sec) at y, when the exposure amount is E (mW/cm.sup.2) at y. As described in the foregoing, the exposure intensity distribution can be quickly determined with such an accuracy that the exposure non-uniformity is tolerable, without correct measurement of the amount of the profile change of the electron beam, which may result from a reduction of the accumulation current or the relative positional deviation among the point of emission, at least one mirror, mask and the wafer. Therefore, the exposure amount non-uniformity can be avoided. Embodiment 5 FIG. 30 illustrates an X-ray exposure apparatus according to a fifth embodiment of the present invention. The SR radiation emitted from the X-ray source in the form of an SR generator 61, is incident on a convex mirror 63 of SiC having a radius of curvature R=56.7 m and disposed 3 m away from the point of emission. The SR radiation reflected by the mirror 63 is transmitted through a transmission type mask 67 (reticle) having a pattern of X-ray absorbing material on an X-ray transparent film so that it is shaped into the pattern, and then, it is incident on the substrate (wafer) 68 on the wafer stage 69 on which the resist sensitive to the SR radiation is applied. Upstream of the mask 67, there is disposed a shutter 65 for controlling the exposure period over the entire exposure area. The shutter 65 is driven by a shutter driving unit 66 controlled by a shutter control unit 70. An unshown thin film of Be having a thickness of 12 .mu.m is disposed downstream of the mirror 63 and upstream of the shutter 65. The thin film makes isolation between the upstream ultra-high vacuum and the downstream reduced pressure He environment. The accumulation current of the SR generator 61 is measured by DCCT 62. Referring to FIG. 33, there is shown in a solid line, a calculated energy which is absorbed by a chemical sensitization resist containing halogen element and having a thickness of 1 .mu.m from the SR radiation 64 emitted from a typical SR generator having a dependency .sigma.y and .sigma.y' shown in FIGS. 31 and 32. Because the optimum exposure is provided when the energy absorbed by the resist has a predetermined level, the value is in accord with the exposure intensity predetermined. However, in the axial exposure, it changes depending on the non-uniformity of the reflectance of the mirror 63, the non-uniformity of the thickness of the Be thin film or the like. In view of this, it is desirably determined on the basis of measurement of the remaining resist film ratio or the like. In FIG. 33, the four lines represent the accumulation currents 300 mA, 250 mA, 200 mA and 150 mA cases in an order from the high exposure intensity side. For a reference, the broken lines represent the exposure intensities obtained on the assumption that it is proportional to the accumulation current on the basis of the intensity distribution at the accumulation current of 300 mA and further on the assumption that the .sigma.y and .sigma.y' do not change depending on the accumulation current. They are for 250 mA, 200 mA and 150 mA of the accumulation currents, respectively, in the order from the higher exposure intensity side. The optimum exposure amount of the chemical sensitization type resist material containing the halogen element used in this embodiment is 60 J/cm.sup.3 and the optimum exposure period therefor at each point in the exposure area is indicated by a solid line in FIG. 34. In the Figure, they are for 300 mA, 250 mA, 200 mA and 150 mA in the order named from the shorter exposure period. In this embodiment, the speed of the shutter 65 is controlled so that the shutter 65 is opened for the exposure period at each point. For reference, the broken lines are the exposure periods at the points determined on the assumption that the .sigma.y and .sigma.y' do not change depending on the accumulation current, and the exposure intensity distribution is proportional to the accumulation current on the basis of the exposure intensity distribution at the accumulation current 300 mA. They are for 250 mA and 150 mA of the accumulation current in the order from the shorter exposure period. Accordingly, despite the use of the SR generator exhibiting the accumulation current dependency of the .sigma.y and .sigma.y' if the exposure period is determined on the assumption that the exposure intensity distribution is proportional to the accumulation current, the resulting exposure period is longer than the proper exposure period by 7% at the maximum at the center of the exposure area at 150 mA, for example. As a result, the amount of the exposure is larger than the optimum exposure amount, and therefore, the exposure non-uniformity is produced. As a measure against the deviation of the exposure amount from the proper exposure amount, the exposure intensity distribution is measured at all of the accumulation current levels, and the exposure period at each point in the exposure area is set at all of the accumulation current levels. However, this is inefficient, and therefore, this embodiment uses the following method. The dependency of .sigma.y and .sigma.y' upon the accumulation current exhibits high reproducibility if the control parameter of the SR generator is constant, and the same .sigma.y and .sigma.y' result for the same accumulation current. As shown in FIGS. 31 and 32, .sigma.y and .sigma.y' exhibit substantially linear inclination relative to the accumulation current. The level and inclination is different for the individual SR generator. When the .sigma.y and .sigma.y' of the electron beam expressed by equation (A) exhibits the dependency upon such an accumulation current level, the SR radiation from the electron beam is incident on the substrate having the resist material sensitive to the SR radiation, through an optical system having at least one mirror and through the mask. As a result, the exposure intensity distribution on the substrate having the resist is dependent on the accumulation current. In this case, the exposure intensity distribution is generally different from a Gaussian distribution, since the SR radiation has been applied by way of the optical system comprising at least one mirror. In view of this, the exposure intensities for at least two accumulation currents are measured beforehand, and upon the exposure operation, the accumulation current is measured, and the change of the exposure intensity distribution attributable to the dependency of .sigma.y and .sigma.y' on the accumulation current is corrected. By doing so, the optimum exposure amount can be provided over the entire exposure area. It is assumed that the exposure intensity distributions at the accumulation currents I.sub.1 and I.sub.2 are P1 (x, y) and P2 (x, y). Then, the exposure intensity distribution for a given accumulation current level I, is ##EQU9## After the exposure intensity distribution is measured at each of three or more accumulation currents, the exposure intensity distribution at a given accumulation current can be determined by interpolation by using a quadrant or higher equation at each point. Referring to FIG. 35, broken lines indicate the exposure intensity distributions determined by equation (B) for the accumulation currents of 250 mA and 200 mA from the exposure intensity distributions at the accumulation currents of 300 mA and 150 mA. In FIG. 35, there is also shown the exposure intensity distributions of FIG. 33 at the accumulation currents of 300 mA, 250 mA, 200 mA and 150 mA. According to this embodiment, at a given accumulated current level not limited to 250 mA or 200 mA, the exposure intensity distribution can be determined with an error of 1% or lower. Since the tolerance of the exposure intensity is 2%, the error is within the tolerance. Thereafter, the proper exposure period is calculated at each point in the exposure area, and the speed of the shutter 65 in the exposure area is determined so that the shutter 65 is opened for a proper exposure period to expose the resist properly. A method of determining the speed of the shutter 65 in the exposure area when the proper exposure period for each point in the exposure area is given, and driving the shutter 65, is disclosed in Japanese Laid-Open Patent Application No. 243519/1989. In an alternative, the exposure intensity distribution is determined at at least two accumulation current levels, and the optimum exposure period at each point in the exposure area at the accumulation current levels, and thereafter, the proper exposure period at a given accumulation current level can be determined by interpolation on the basis of the proper exposure periods at at least two exposure current levels. Embodiment 6 FIG. 36 illustrates an X-ray exposure apparatus according to a sixth embodiment of the present invention. The SR radiation irradiated from the X-ray source in the form of an SR generator 71 is incident on a cylindrical convex spherical mirror 73 of SiC disposed at 3 m away from the point of emission with an inclined incident angle of 15 mrad approx. The mirror 73 has a convex surface in order to enlarge the SR light 76 emitted from the SR generator 71, and therefore, the curvature is away from the SR generator 78. The SR radiation 76 reflected by the mirror 13 is transmitted through a reticle in the form of a transmission type mask 79 having an X-ray transmission film on which a desired pattern is formed by an X-ray absorbing material, so that the desired pattern of the X-rays is incident on a substrate (wafer) 80 coated with resist material sensitive to the X-rays on the wafer stage 81. Upstream of the mask 79, there is disposed a shutter 77 for controlling the exposure period over the entire exposure area. The shutter 77 is driven by a shutter driving unit 78 controlled by a shutter control unit 83. In order to monitor the accumulation current, an X-ray detector 12 having a radiation receiving surface of sufficient size is disposed between the mirror 73 and the point of emission. An unshown thin film of Be having a thickness of 12 .mu.m is disposed downstream of the mirror 73 and upstream of the shutter 77 to function as an isolation between the upstream ultra vacuum environment and the downstream pressure reduced He environment. The radius of curvature R of the SiC mirror 73 is variable between 50 m and 57 m by a mirror shape changing unit 74 controlled by a mirror shape control unit 75. The shutter correction is carried out on the assumption that the .sigma.y and .sigma.y' of the electron beam do not change and that the exposure intensity distribution is proportional to the accumulation current as in the prior art, and the radius of curvature R is controlled so that the exposure non-uniformity is within the tolerance. FIG. 37 shows the exposure intensity distribution at the accumulation current level of 300 mA and the radius of curvature of 56.7 m. This is indicated by the highest intensity line among the four solid lines. If it is assumed that .sigma.y and .sigma.y' do not change and that the exposure intensity distribution is proportional to the accumulation current as in the prior art, the exposure intensity distributions are as indicated by broken lines in FIG. 37 at the accumulation currents 250 mA, 200 mA and 150 mA. On the other hand, in order that the exposure intensity distribution at the accumulation current level of 150 mA is in accord with the exposure intensity distribution provided on the assumption that the exposure intensity distribution is proportional to the accumulated current without the change of .sigma.y and .sigma.y' within the tolerance in the exposure area (20 mm-width), the radius of curvature R is 51.3 m. The exposure intensity distribution at this time is indicated by the lowest level solid line in FIG. 37. Between 150 mA and 300 mA of the accumulation current, the radius of curvature is determined by linear interpolation from 51.3 m-56.7 m. The resultant exposure intensity distribution at 200 mA and 250 mA of the accumulation current is indicated by solid lines in FIG. 37. They are in accord with the exposure intensity distribution provided on the assumption that .sigma.y and .sigma.y' do not change and that the exposure intensity distribution is proportional to the accumulation current, within the tolerance in the exposure region (20 mm-width). In this embodiment, the exposure intensity distribution is frequently measured at the accumulation current of 150 mA. A calibrated photodiode is usable as an X-ray detector 72. Generally, in an X-ray exposure apparatus using an SR generator, the X-ray intensity distribution measured by an X-ray detector 82 disposed on a wafer stage 81 is different from the exposure intensity distribution for the following reasons: 1. The spectrum of SR radiation is continuous; 2. The SR radiation is reflected by at least one mirror, and the reflectance is significantly dependent on the wavelength of the X-rays and the incident angle; 3. Transmissivities of the Be window and membrane passed by the SR radiation before reaching the resist are significantly dependent on the wavelength thereof; 4. The absorption of the X-rays by the resist is significantly dependent on the wavelength (the resist exposure is proportional to the energy absorbed thereby, and therefore, this means that the spectrum sensitivity of the resist is not uniform); and 5. The spectrum sensitivity of the X-ray detector 72 is significantly different from the spectrum sensitivity of the resist. Therefore, the amount of exposure is different even if the X-ray intensity is the same. However, the linearity of the X-ray detector 82 is high, and the reproducibility thereof is also high. Therefore, the interrelation therebetween is determined at each position in the exposure area, and on the basis of which an exposure intensity distribution can be calculated from the X-ray intensity at a given position. The exposure operation is carried out for a predetermined period at an accumulation current, and the exposure intensity distribution is determined from the film remaining ratio. The X-ray intensity distribution is determined from the output of the X-ray detector 82, while the wafer stage 81 is being moved in the y-direction (perpendicular to the SR orbit plane) with the accumulation current level which is substantially the same during the exposure. FIG. 38 shows a normalized exposure intensity distribution and an X-ray intensity distribution measured by the X-ray detector 82 disposed on the wafer stage 81 when the accumulation current is 150 mA, and the radius of curvature is 51.3 m. FIG. 39 shows an interrelation between the X-ray intensity and the exposure intensity at each point in the exposure region. Since the spectrum distribution, the spectrum sensitivity of the X-ray detector 82 and the spectrum sensitivity of the resist are different at each position of the exposure area, the inclination of the interrelation at each point is different. However, the linearity is so high that the exposure amount is determined from the X-ray intensity. The exposure intensity is determined by the following equation: EQU D (y)=A (y).times.O(y) (17) where D (y) is the exposure intensity, and O (y) is the X-ray intensity, and A (y) is the inclination. FIG. 40 shows the inclination A (y). If another X-ray detector is used with the result of a change of A (y), it is required to determine A (y), again. As for the method of determining the exposure intensity, the exposure operation is carried out while changing the exposure period, and thereafter, the development is carried out. Then, a ratio of the line width of the mask pattern and the line width of the resist pattern is expressed as a function of the exposure period. The optimum exposure period is determined as the exposure period resulting in 1 of the ratio. In this manner, the exposure intensity can be determined. The method of determining the accumulation current by the X-ray detector is based on the fact or assumption that the output of the X-ray detector has a sufficiently large ray receiving surface as compared with the expanded SR radiation in the direction perpendicular to the SR orbit plane. An output of an X-ray detector at an accumulation current is measured beforehand, and when the output of the X-ray detector is one half, for example, it is assumed that the accumulation current is also one half. If a photodiode is used as the X-ray detector, the linearity can be assured in the wide range, and therefore, it is preferable. Embodiment 7 FIG. 41 illustrates an X-ray exposure apparatus according to a seventh embodiment of the present invention. The SR radiation emitted from the X-ray source in the form of an SR generator 84 is incident on a swingable flat surface mirror 86 of SiC disposed 3 m away from the point of emission at an inclined incident angle of 11-19 mrad. By the swinging motion of the mirror 86, the sheet-like SR radiation 89 is expanded in effect. The mirror 86 is swung by a mirror swinging unit 87 at a swinging speed controlled by the mirror swing control unit 88. The SR radiation 89 reflected by the mirror 86 is transmitted through a reticle in the form of a transmission type mask 92 having an X-ray transmission film on which a desired pattern is formed by an X-ray absorbing material, so that the desired pattern of the X-rays are incident on a substrate (wafer) 93 coated with resist material sensitive to the X-rays, on the wafer stage 94. At the front of the mask, there is provided an opening (Be window 90) movable in synchronism with a mirror 86. A thin film of Be having a thickness of 12 .mu.m functions as a vacuum isolation between an upstream ultra high vacuum environment and a downstream pressure-reduced He environment. In order to assure the sufficient strength of the Be window 90 against the pressure difference, it has a thickness of 10 mm in the y-direction (perpendicular to the SR orbit plane). It is vibrated by a Be window driving unit 91 in synchronism with the vibration of the mirror 86 so as not to block the SR radiation 89. When the mirror swings, the exposure intensity is defined by the exposure amount when the sheet-like SR radiation 89 swings at a constant speed on the wafer substrate 93. In other words, the exposure intensity is determined from the remaining resist film ratio or the line width accuracy after such exposure. FIG. 42 shows, in a solid line, an exposure intensity distribution on the wafer substrate 93 when the mirror 86 is fixed such that the SR radiation 89 is incident on the flat mirror 86 at 12 mrad. The solid lines represent 300 mA, 250 mA, 200 mA and 150 mA of the accumulation current in the order named from the high intensity side. Through the width of the Be window 90, only the inside part of the SR radiation 89 reaches the wafer substrate 93 to expose the resist to the radiation. In FIG. 43, solid lines show the exposure amounts provided by one vibration of the sheet like SR radiation 89 on the wafer 86 at a constant speed of 40 mm/sec. The solid lines represent 300 mA, 250 mA, 200 mA and 150 mA of the accumulation current in the order named from the larger exposure amount side. They indicate the exposure intensities for the respective accumulation currents. The exposure amount control upon the mirror swinging is effected on the basis of the following. When 6 mJ/cm.sup.2 of the exposure amount is required to expose the resist having a thickness of 1 .mu.m, for example, the SR radiation 89 swings at the speed of 40 mm/sec on the wafer substrate 93 with the accumulation current of 300 mA, and the exposure amount if 3.42 mW/cm.sup.2 at y=0 mm. On the basis of this, the mirror 86 is swung to provide the speed of 22.8 (mm/sec) =40/(6/3.42) at y=0 mm. FIG. 44 shows the proper swinging speed of the sheet-like SR radiation 89 on the wafer substrate 93 at each accumulation current. They represent 300 mA, 250 mA, 200 mA and 150 mA cases of the accumulation currents in the order named from the high speed side. In FIG. 43, the broken lines are the values provided on the assumption that the exposure amount is proportional to the accumulation current at 300 mA of the accumulation current. If it is assumed that the exposure intensity is proportional to the accumulation current, the 6% exposure non-uniformity occurs when the accumulation current is 150 mA, for example. In FIG. 45, the solid line indicates the exposure intensity (the exposure amount at a certain swinging speed of the SR radiation 89), in the similar manner as in FIG. 43. The broken lines are the exposure intensities provided by measuring the exposure amount at 150 mA and 300 mA of the accumulation currents and by determining the exposure amount intensities at 200 mA and 250 mA of the accumulation current by equation (B). The error is approx. 1%. Therefore, if the speed of the flat mirror 86 is determined on the basis of the exposure intensity distribution determined by the interpolation, the resultant exposure amount is approx. 1%. As described in the foregoing, the exposure intensity distributions are estimated beforehand for at least two accumulation currents of the SR generator, and the dose is determined at a given position on the mask for the accumulation current during the exposure, in accordance with the estimated exposure intensity distribution and the accumulated current monitored during the exposure, by which the exposure amount non-uniformity attributable to the shape instability of the SR radiation resulting from the accumulation current can be removed. Embodiment 8 The description will be made as to an embodiment of a device manufacturing method using the exposure apparatus described in the foregoing. FIG. 46 is a flow chart of a manufacturing process of fine devices such as a semiconductor chip such as IC, LSI or the like, a liquid crystal panel, CCD, thin film magnetic head, micromachine or the like. In step 41 (circuit design), the circuit of the device is designed. At step 42 (mask manufacturing), a mask having a circuit pattern of the designed circuit is manufactured. On the other hand, at step 43 (wafer manufacturing), a wafer is manufactured using a proper material such as silicon or the like. At step 44, (wafer process), a pre-process is carried out in which the circuit is printed on the wafer through a lithographic process using the prepared mask and the wafer. At step 45 (assembling), a post-process is carried out in which a semiconductor chip is manufactured using the wafer produced at step 44. It includes an assembling step (dicing, bonding), packaging step (chip sealing) or the like. At step 46 (inspection), the operation of the device manufactured by step 45 is inspected, and durability thereof is inspected. The device is manufactured through such steps, and is delivered at step 47. FIG. 47 is a flow chart of wafer processing described above. At step 51 (oxidation), the surface of the wafer is oxidized, and at step 52 (CVD) an insulating film is formed on the surface of the wafer. At step 53 (electrode formation), an electrode is evaporated on the wafer. At step 54 (ion implantation), the ions are implanted into the wafer. At step 55 (resist processing), the wafer is coated with sensitive material. At step 56 (exposure), the circuit pattern is printed on the wafer by any one of the exposure apparatuses described hereinbefore. At step 57 (development), the exposed wafer is developed. At step 58 (etching), the part other than the resist image is removed. At step 59 (resist removal), the unnecessary resist material after the etching is removed. These steps are repeatedly carried out to an overlaid circuit pattern on the wafer. Through the manufacturing method of this embodiment, a high accuracy device which has been difficult to manufacture can be manufactured with high productivity. While the invention has been described with reference to the structures disclosed herein, it is not confined to the details set forth and this application is intended to cover such modifications or changes as may come within the purposes of the improvements or the scope of the following claims. |
050948032 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 represents the first embodiment of the present invention and, referring to FIG. 1, a steam generator 50 comprises a body shell 40, a shell supporting skirt 41 and other structural members inclusive of a heat transfer tube assembly 42 located in association with the shell 40 and the supporting skirt 41. An inlet port for liquid metal is composed of a piping system for the liquid metal, not shown, and distributing pipes 43 for the liquid metal inlet. Inside the body shell 40, as shown in FIG. 1, at the upper portion, is formed a liquid surface 44 for absorbing the change of the inner volume of the liquid metal. The shell 40 is provided with a dish portion 45 at the lower portion thereof and a nozzle 46 is formed to the dish portion 45. As shown in FIG. 1, an outlet rising pipe 47 for the liquid metal is arranged at the inner central portion of the steam generator 50. Further referring to FIG. 1, water inlet pipes 49 through which water is made to flow by the operation of a water feed pump, not shown, water outlet pipes 51 and water inlet headers 52 are arranged at a water inlet portion, whereas a steam outlet portion comprises outlet steam chambers 53, outlet steam distributing pipes 54, outlet steam headers 54a, and outlet steam pipes 56. An electromagnetic pump 60 is also arranged in the body shell 40, as shown in FIG. 1, at the upper inner portion of the outlet rising pipe 47 through an installing flange member 57. With respect to a power source, which is not shown in its entirety, for the electromagnetic pump 60, a three phase power source cable is connected to a cable for high temperature use through a terminal box, then through the insides of electric wire tube 58, and is then connected to an outer stator coil 59 and an inner stator coil 61 arranged in the liquid metal. The inner stator coil 61 is composed of a material of excellent strength and conductivity in a high temperature environment such an alumina dispersion-strength type copper alloy which is enclosed by a material having high electric insulation at a high temperature such as an inorganic ceramic series material, and each of the coil layers are of an annular form. The inner stator coil 61 is integrally composed of a plurality of assemblies each mounted in a close fit manner in recessed portions of a comb-shaped inner iron core 62 formed by laminated plates as shown in FIG. 3 or 4. The thus integrated structure of the inner stator coil exhibits an outer surface which is completely covered with a sealing member 64 made of such as austenite stainless to be isolated from the liquid metal. The inner iron core 62 is of a hollow cylindrical shape having an inner central through hole 62a, and the sealing member 64 is provided with corn shaped projections 64a extending outwardly at both vertical ends of the sealing member 64. Passages 65 are formed as axial through holes at the central portions of the projections 64a and these passages 65 communicate with the through hole 62a of the inner iron core 62 to thereby form the bypass passage 62a and 65 for cooling the same. The sealing member 64 is also disposed on the side of the bypass passage 62a and 65 so as to completely enclose the inner iron core 62 and the inner stator coil 61. A hollow cylindrical outer iron core 67 is disposed outside the inner iron core 62 in a concentric arrangement therewith with an annular space 66 therebetween. The outer iron core 67 is provided with an inner peripheral portion formed in comb shape and an annular outer stator coil 59 is mounted in a close fit manner in the comb shaped portion. The outer peripheral portion of the outer iron core 67 are embedded in the inner peripheral surface of an annular member 68, to which a plurality of cooling bypass passages 69 are formed in parallel with the outer peripheral surface of the outer iron core 67. The respective upper and lower ends of each cooling bypass passage 69 communicate with the upper and lower portions of the annular space 66, respectively. The outer iron core 67 and the assembled outer stator coil are covered with a sealing member 71. The inner iron core 62 is supported inside the outer iron core 67 by a support member 72 arranged between the inner iron core 62 and the outer iron core 67. The outer peripheral surface of the annular member 68 is supported by an electromagnetic pump supporting cylinder 73 which is arranged inside the body shell 40 of the steam generator 50. The steam generator of the structure according to the described embodiment operates as follows. When the electromagnetic pump 60, accommodated in the steam generator 50, is energized, the liquid metal, operable as a secondary coolant, is circulated and delivered to the intermediate heat, exchanger (see FIG. 8) to obtain heat and the temperature of the liquid metal becomes high. The liquid metal of high temperature is delivered to the steam generator and is fed into the body shell 40 through the liquid metal inlet distributing pipes 43. The liquid metal of high temperature then rises into the outlet rising pipe 47 opening at the lower portion of the steam generator 50 and is circulated, after applied with a discharge pressure by the electromagnetic pump 60, into the intermediate heat exchanger 4 through the piping 7 shown in FIG. 8. On the other hand, with the water and steam supplying side, the feed water fed by the operation of the water feed pump, not shown, is pumped into the water inlet header 51a through the water inlet pipe 49 and is then distributed into a plural number of flows through the water inlet distributing pipes 51 and finally into the water inlet chamber 52. The water distributed from the water inlet chamber 52 and fed into the interior of the heat transfer tube assembly 42 flows upwardly in the heat transfer tube assembly 42 and performs heat exchange operation between the water and the liquid metal. During this heat exchange operation, the water is highly heated to change into steam, which is then fed into the outlet steam chamber 53. The steam thus generated is distributed into a plurality of flows by the outlet steam distributing pipes 54 and is then combined at the outlet steam header 54a. The steam is finally delivered to the steam turbine, not shown, through the outlet steam distributing pipes 56. In the meantime, the electromagnetic pump 60 according to the present embodiment includes the inner and outer iron stator coils 61 and 59 which are tightly fitted to the comb shaped portions formed to the inner and outer iron cores 62 and 67. Accordingly, the magnetic fluxes caused by the stator coils 61 and 59 can be almost converged by the respective iron cores 62 and 67, and even in a case of an electromagnetic pump of large capacity, a sufficient pumping efficiency such as of 40 to 50% can be achieved. Moreover, excessive or abnormal temperature rise in the respective stator coils 61 and 59 and, hence, the inner and outer iron cores 62 and 67, can be definitely eliminated by the backward flow of the liquid metal from the discharge side of the pump to the suction side due to the location of the liquid metal bypass passages 62a, 65 and 69 through the heat removing function. This is to say that in detail, a part of the liquid metal of low temperature flowing upwardly in the main passage, i.e. the annular space 66, due to the operation of the electromagnetic pump 60, flows into the cooling bypass passages 65, 62a and 69 and, in these bypass passages, the liquid metal flows downwardly as shown by arrows in FIG. 3 because the upper portion of the annular space 66 is in a high pressure state and the lower portion thereof is in a low pressure state. Accordingly, the inner stator coil 61 and the inner iron core 62 are subjected to the heat removing function from the front and rear sides thereof by the liquid metal passing the annular space 66 and the bypass passages 65 and 62a, whereby the heat generated by the inner stator coil 61 is effectively absorbed and recovered to the liquid metal. In addition, the temperature rise of the outer iron core due to the heat generation of the outer stator coil 59 can be effectively suppressed by the heat removing function of the liquid metal passing through the bypass passage 69 formed outside the outer iron core 67. All the heat energy recovered to the liquid metal can be effectively utilized for a power generation system such as turbine generator. For example, in a case where the heat energy is utilized for a turbine generator, it is supposed that the energy conversion efficiency into electric energy is about 40%, almost all the energy loss (60 to 50%) of the electric energy which is not converted into fluid energy by the electromagnetic pump 60 is the energy loss of the respective stator coils 59 and 61 and the iron cores 62 and 67. Such energy loss can be almost completely recovered however, and accordingly, electric energy of about 24 to 30% of that applied to the electromagnetic pump 60 is again converted into electric energy. Thus, the apparent pumping efficiency of the electromagnetic pump becomes 64 to 70%, which is not substantially less than that due to the operation of a mechanical pump. FIGS. 5 and 6 represent the second embodiment according to the present invention in which a reaction pressure controlling cylinder 75 is disposed between the outlet rising pipe 47 and an inside shroud 48 of the heat transfer tube as shown in FIG. 1 with reference to the first embodiment of the present invention. Referring to FIGS. 5 and 6, an intermediate space of double cylindrical shape, formed between the outlet rising pipe 47, for the liquid metal and the inside shroud 48 of the heat transfer pipes, is formed as the reaction pressure controlling cylinder 75 is filled with liquid having a liquid surface communicating with the liquid metal of the low temperature side of a steam generator 80. A reaction pressure releasing pipe 76 is arranged as shown in FIG. 6 at the upper portion of the steam generator 80, and a rupture disk, not shown, is arranged for the reaction pressure releasing pipe 76. Except for the above structure of the steam generator, the second embodiment is provided with substantially the same structure as that of the first embodiment, and the operation or function of the steam generator 80 itself is substantially the same as that of the steam generator 50 of the first embodiment. The arrangement of the reaction pressure controlling cylinder 75 can effectively attain a remarkable damping effect by means of the liquid metal filling therein against a violent pressure propagation and a pressure rising due to accidents of the heat transfer tube assembly 42, for example. Particularly, the soundness and the reliability of the structure such as electromagnetic pump 60, located on the low temperature side of the steam generator 80, and the intermediate heat exchanger 4 (FIG. 8) and pipes, both arranged external to the steam generator 80, can be effectively improved. The arrangement of the reaction pressure releasing pipe 76 enables auxiliary equipment such as a liquid metal recovery tank required for the protection of the steam generator 80. FIG. 7 further represents the third embodiment according to the present invention in which stators have structures different from those of the first embodiment. Referring to FIG. 7, the electromagnetic pump of this embodiment comprises an inner iron core 62 and an inner stator coil 61 which is assembled in the outer periphery of the inner iron core 62. The inner stator coil 61 and the inner iron core 62 are covered with a sealing member 64. The central through hole 62a of the inner iron core 62 is formed as a bypass passage through which the liquid metal passes. An annular space 66, formed between the outer iron core 67 and the inner iron core 62, is also formed as a main passage for the liquid metal. According to the structure of the third embodiment, the central through hole 62a is formed as a cooling bypass passage, so that the heat generated from the inner stator coil 61 is effectively absorbed by the liquid metal flow. As described hereinbefore, according to the present invention, the main flow passage of the liquid metal is formed on the side on which the stator coil of the cylindrical iron core is assembled and the cooling bypass passage is formed to penetrate the central portion of the steam generator, so that the heat generated by the stator coil can be effectively absorbed and the excessive temperature rising of the stator coils and iron cores can be suppressed, whereby the heat generation efficiency of the steam generator can be remarkably improved. |
claims | 1. A method of controlling a substrate temperature during a plasma ion implantation process comprising:(a) supplying an implantation energy to perform a first portion of a plasma ion implantation process on a substrate having a magnetically susceptible layer formed thereon in a processing chamber for a first time period, wherein a temperature of the substrate is maintained below about 150 degrees Celsius;(b) cooling the temperature of the substrate after the first portion of the plasma ion implantation process has been completed by turning off bias power while maintaining the implantation energy on; and(c) performing a second portion of the plasma ion implantation process on the substrate, wherein the temperature of the substrate is maintained below 150 degrees Celsius. 2. The method of claim 1, further comprising:repeating step (b)-(c) in-situ a processing chamber. 3. The method of claim 1, further comprising:repeating (b)-(c) until a dopant concentration of the substrate has reached between about 1×1018 atoms/cm3 and about 1×1023 atoms/cm3. 4. The method of claim 1, wherein the first portion of the plasma ion implantation process has a duration of between about 5 seconds and about 40 seconds. 5. The method of claim 1, wherein performing the plasma ion implantation process on the substrate further comprises:implanting ions into a portion of the magnetically susceptible layer exposed by a patterned mask layer disposed on the substrate. 6. The method of claim 5, wherein the ions implanted into the magnetically susceptible layer are selected from a group consisting of boron, phorosphine, and arsine. 7. The method of claim 1, wherein the magnetically susceptible layer includes a first layer disposed on a second layer. 8. The method of claim 7, wherein the first layer is selected from a group consisting of iron, nickel, platinum, and combinations thereof; and the second layer is selected from a group consisting of cobalt, chromium, platinum, tantalum, iron, terbium, gadolinium, and combinations thereof. 9. The method of claim 1, wherein the implantation energy is supplied to a gas mixture in the processing chamber to ionize at least a portion of the gas mixture. 10. The method of claim 9, wherein the RF energy is supplied in a pulsed mode. 11. The method of claim 1, wherein the cooling process further comprises:supplying a cooling gas to the surface of the substrate. 12. The method of claim 11, wherein the cooling gas includes at least one of He, Ar, H2, N2 or N2O. 13. A method of controlling a substrate temperature during a plasma ion implantation process comprising:(a) supplying an implantation energy to perform a first portion of a plasma ion implantation process on a substrate having a magnetically susceptible layer formed thereon in a processing chamber for a first time period, wherein a temperature of the substrate is maintained below about 150 degrees Celsius;(b) turning off bias power while maintaining the implantation energy on during the first portion of the plasma ion implantation process to cool down the temperature between about 15 degrees Celsius and about 30 degrees Celsius; and(c) continuing performing a second portion of the plasma ion implantation process on the substrate, wherein the temperature of the substrate is maintained below 150 degrees Celsius. 14. The method of claim 13, further comprising:repeating step (b)-(c) in-situ a processing chamber. 15. The method of claim 13, further comprising:repeating (b)-(c) until a dopant concentration of the substrate has reached between about 1×1018 atoms/cm3 and about 1×1023 atoms/cm3. 16. The method of claim 13, wherein turning off the bias power further comprises:supplying a cooling gas to the substrate surface. 17. The method of claim 16, wherein the cooling gas includes at least one of He, Ar, H2, N2 or N2O. |
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abstract | A system and method of transferring a radioactive payload using a shield-gate apparatus, including a method of performing work within a cavity of a shielding container using a shield-gate apparatus and shielding block, in one embodiment, the invention is a system comprising: a shield-gate apparatus comprising a body, a passageway extending along an axis through the body, and one or more movable shielding gates; a shielding block positioned atop the body of the shield-gate apparatus to enclose a first opening of the passageway; and a retaining feature that prevents relative transverse movement between the shielding block and the shield-gate apparatus while allowing relative rotation between the shielding block and the shield-gate apparatus about a central axis of the shielding block. |
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summary | ||
055725597 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The .sup.16 O(n,p).sup.16 N reaction leads to activation of ordinarily benign pure water (H.sub.2 O) when it is bombarded with sufficiently energetic neutrons. The natural isotopic abundance of .sup.16 0 is 99.76%. The Q-value for this reaction is -9.637 MeV, and that corresponds to a relatively high neutron reaction threshold energy of 10.245 MeV. The reaction cross section is essentially negligible below 11 MeV but increases rapidly to around 80 millibarns near 12 MeV, apparently due to a cross section resonance near threshold. The cross section is around 40-50 millibarns in the range 14-15 MeV. The .sup.17 O(n,d+n'p).sup.16 N .sup.18 O(n,t).sup.16 N, .sup.16 O(n,.gamma.).sup.17 O(n,d+n'p).sup.16 N, .sup.17 O(n,2n).sup.16 O(n,p).sup.16 N and .sup.17 O(n,t).sup.15 N(n,.gamma.).sup.16 N reactions also contribute to .sup.16 N production when pure water is irradiated with 14-MeV neutrons. However, because of low isotopic abundances and small cross sections, these secondary contributions are extremely small. Relative to the .sup.16 O(n,p).sup.16 N reaction, the yield from the one-step secondary reactions is estimated to be less than one part in 10.sup.4. For the two-step secondary processes the relative yield is estimated to be less than one part in 10.sup.5, even when it is assumed that the cooling water has been exposed continuously for one year to fusion neutrons at assumed flux levels as high as 10.sup.15 neutrons/cm.sup.2 /second (roughly corresponding to a fusion power reactor operating at full power). In any event, it does not matter from the perspective of radiography which processes are involved in generating the .sup.16 N activity. The decay by beta (.beta..sup.-) emission of the product nucleus .sup.16 N with a 7.13 second half life to .sup.16 0 is a very energetic process. The transition to the ground state of .sup.16 0 involves beta particles with energies up to 10.419 MeV. There are also beta-decay transitions to excited levels of .sup.16 O followed by gamma-ray emission. The average energy of the composite beta spectrum is 2.693 MeV. Of interest in the present invention is the fact that 68.8% of all decays of .sup.16 N produce a 6.129-MeV gamma ray while 4.7% produce a 7.115-MeV gamma ray. The 6.129-MeV gamma rays thus outnumber those of 7.115-MeV by nearly 15-to-1. Furthermore, the transmission cross sections for these two energies differ by only a few percent across the Periodic Table. Therefore, water which is activated by sufficiently high energy neutrons becomes a source of nearly monoenergetic high-energy gamma rays which can be used for a variety of purposes. For completeness, it should be noted here that the neutron inelastic scattering reaction, .sup.16 O(n,n,).sup.16 O, also leads to the emission of these same gamma rays for neutron energies above the threshold for exciting the specific excited levels in .sup.16 0. The cross section for this process is several hundred millibarns for 14-MeV neutrons. However, the gamma-ray emission is prompt so neutron inelastic scattering from water does not contribute a source of delayed gamma radiation from water which has been transported away from the region in the D-T fusion reactor where the neutron irradiation occurs. The limiting conversion efficiency for 14-MeV neutrons to 6.129+7.115 MeV photons in an infinite water medium is approximately the ratio of the .sup.16 O(n,p).sup.16 N reaction cross section (40-50 millibarns) to the neutron total cross section for water (about 3 barns) multiplied by the photon-emission branching factor (about 0.74). This amounts to an efficiency of about 1% which is not large but nevertheless leads to significant gamma-ray production when water is exposed to 14-MeV neutron fields such as those produced by a D-T neutron generator or in a D-T fusion reactor. This is clearly evident from the recent calculations by Sato et al. in "Evaluation of Skyshine Dose Rate Due to Gamma-rays from Activated Cooling Water in Fusion Experimental Reactors," p. 946, Proceedings of the 8th International Conference on Radiation Shielding, American Nuclear Society, La Grange Park, Ill. (1994) for the ITER (International Thermonuclear Experimental Reactor) conceptual design as discussed below. Although the 14-MeV neutron fields produced by D-T neutron generators are much less intense than those anticipated for D-T fusion devices such as ITER, these accelerators are readily available in many laboratories. It has been possible to demonstrate using the present invention that sufficient numbers of .sup.16 N gamma rays can be produced with a D-T neutron generator to allow photon radiography to be carried out with moderate resolution. In any event, the present invention will operate with virtually any source of D-T fusion neutrons. The present invention was carried out at the Fusion Neutron Source (FNS) accelerator located at the Japan Atomic Energy Research Institute (JAERI) in Tokai, Japan. At this D-T neutron generator facility, deuterons can be accelerated up to 350-key energy, with beam currents up to 20 milliamperes. The deuterons impinge upon a titanium-tritide target to produce neutrons via the .sup.3 H(d,n).sup.4 He reaction. This arrangement leads to neutron production up to 3.times.10.sup.12 neutrons per second (into 4.pi. steradian). Since the reaction Q-value is 17.591 MeV, the energies of the emitted neutrons are in the range 13-15 MeV, depending upon the angle of emission relative to the incident deuterons. As D-T neutron generators go, this is a very powerful facility. Consequently, it was possible to carry out the present invention without considerations as to the optimization of the geometrical coupling between the neutron source and circulating water that was activated for radiography purposes. Referring to FIG. 1, there is shown a simplified schematic diagram of a radiography apparatus 10 using gamma rays emitted by water activated by fusion neutrons in accordance with the present invention. The radiography apparatus 10 includes a circulating loop of water 12 comprised of plastic tubing having an inner diameter of approximately 1 cm, a water pump 14 and a flow meter 16. The water within the circulating loop 12 flows in the direction of arrow 26 and through a shielding arrangement 28. The circulating loop of water 12 is arranged in a straight line along a path approximately 10 cm from a point neutron source 15 at its closest approach. Neutron source 15 includes a source of energetic deuterons 20 such as the aforementioned FNS accelerator for directing 350-keV deuterons represented by arrow 22 onto a titanium-tritide target 18. The deuterons 22 impinging upon the titanium-tritide target 18 produce neutrons represented by arrow 24 via the .sup.3 H(d,n).sup.4 He reaction. This reaction leads to neutron production up to 3.times.10.sup.12 neutrons per second (into 4.pi. steradian). The flow rates used in the circulating loop 12 could be varied by means of water pump 14 and were measured by means of flow meter 16. The intensity of the photon field could also be adjusted by changing the coupling of the circulating water loop to the neutron radiation field or, more simply by varying the speed of the water pump 14. In the disclosed embodiment, the water flow rate was such that any individual volume element of water spent no more than about 0.1 second in the high-fluence region near the titanium-tritide target 18. Because this time period is much shorter than the .sup.16 N half life, the activity generated in the water was always far short of saturation. The physical parameters available for optimization of the neutron irradiation configuration are dwell time in the neutron field, solid angle relative to the point neutron source, and average neutron energy. It is estimated that by coiling the water line and placing it closer to the target of the Fusion Neutron Source (FNS) accelerator, it would have been possible to achieve .sup.16 N concentrations in the flowing water of two orders of magnitude (10.sup.2) higher than were actually attained in the present embodiment. A maximum flow rate of about 10 liters per minute (corresponding to about 2 meters per second velocity in the tubing) could be achieved with the water pump 14 utilized in the disclosed embodiment. It was found that this particular flow rate provided nearly the highest possible delivered intensity of .sup.16 N activity at the position of the radiography apparatus (located approximately 25 meters from the accelerator target) for the particular geometry shown in FIG. 1 the .sup.16 N activity in the transported water decreased to about 30% of its value near the accelerator target due to radioactive decay during the required transit time of approximately 12 seconds between the titanium-tritide target and the radiographic portion of the apparatus. As indicated below, sufficient .sup.16 N activity was present at this position to perform the radiography measurements reported below. An estimate was made of the 6.129+7.115 MeV gamma ray emission rate from the water in the circulating loop 12. These calculations were based on physical data discussed above and details of the inventive radiography apparatus 10. The result obtained was approximately 1.times.10.sup.4 photons per second per milliliter of water (i.e., about 0.27 microCuries per milliliter). The actual volume of water viewed by the detector (described below) was about 7.3 milliliters. Referring to FIG. 2, as well as to FIG. 1, details of the photon detection arrangement used in the radiography apparatus 10 will now be described. In the photon detection portion of the radiography apparatus 10, the circulating loop of water 12 is completely surrounded by shielding 38 comprised of lead bricks to a thickness of at least 10 cm, except for a single collimator slot 30 which in the disclosed embodiments is 10 cm wide by 2.5 cm high through which the photons shown in simplified form as arrow 32 in the figures could emerge. A 20 cm gap between the shielded source of photons, i.e., the circulating loop of water 12, and a shielded scintillation detector 36 is provided for placement of an object 34 to be studied by radiography. The shielded scintillation detector 36 includes a 12.7 cm diameter.times.5.2 cm thick sodium iodide scintillator 52. The sodium iodide scintillator 52 is surrounded by lead shielding 42 at least 10 cm thick, except for a single slot 44 which is 13 cm wide by 2.5 cm high and is aligned with the collimator slot 30 in the shielding 38 of the circulating water loop 12. Table I shows that 10 cm of lead shielding limits the transmission of 6 MeV photons to less than 1%. TABLE I ______________________________________ Transmission (I/I.sub.0) Element x(cm) = 0.1 0.5 1.0 5.0 10.0 ______________________________________ Carbon (C) 0.9944 0.9725 0.9457 0.7563 0.5720 Aluminum 0.9929 0.9649 0.9309 0.6992 0.4889 (Al) Iron (Fe) 0.9763 0.8870 0.7868 0.3015 0.0909 Copper 0.9727 0.8706 0.7580 0.2502 0.0626 (Cu) Lead (Pb) 0.9518 0.7811 0.6102 0.0846 0.0072 ______________________________________ A rectangular slot geometry was selected because it provides a greater sensitivity than that available with a cylindrical or square collimator arrangement, without sacrificing resolution in the direction along which object 34 is scanned in the radiography apparatus 10. The rectangular collimator configuration shown in FIG. 2 permits photons to pass through object 34 at various angles. However, in the embodiment of the radiography apparatus shown in FIGS. 1 and 2, the range of angles due to this effect was relatively small, i.e., <14.degree. corresponding to a variation of less than 3% in path length through the object or target 34. The detector electronics include a photomultiplier tube 45 coupled to the sodium iodide scintillator 52 and disposed within lead shielding 42. The remaining portion of the electronics and data acquisition system 50 is coupled to the photomultiplier tube 45 by means of an electrical lead 48 extending through a narrow second slot 46 within lead shielding 42. The electronics and data acquisition system 50 is conventional in design and operation and includes a preamplifier, a high voltage power supply, an amplifier, a delay amplifier, a pulse selector, and a linear gate, which are not shown in the figure for simplicity. The latter three components allow pulses below an equivalent photon energy of 2.506 MeV to be rejected. Signals corresponding to higher energy gamma rays were acquired on line with a computer, although it would have been possible to alternatively record data using either a multichannel analyzer or a scaler. Object 34 was scanned in the direction of arrow 40 by the incident gamma rays 32 by displacing the object in the direction of the arrow. FIG. 3 is a graphic representation of a typical sodium iodide scintillation detector spectrum produced by 6.129+7.115 MeV gamma rays from radioactive water produced in accordance with the present invention, as seen by the shielded scintillation detector 36 through the above-described collimator system without an intervening object 34 present. Four test objects were prepared for use in demonstrating the feasibility of performing radiographic studies with the radiography apparatus of the present invention. Object A 54 as shown in FIGS. 4a and 4b is a featureless, 5 cm.times.15 cm.times.20 cm rectangular block of stainless steel (mostly iron). Object B 56 shown in FIGS. 5a and 5b is identical to Object A except for a 2 cm diameter hole drilled through the center along its axis. Object C 58 shown in the end and side views of FIGS. 6a and 6b consists of two 1 cm-thick copper plates 58a and 58b with a hidden rectangular lead block 58c which is 2.5 cm.times.20 cm situated between the two copper plates. Object D 60 shown in the end and side views of FIGS. 7a and 7b consists of two 5 cm.times.5 cm.times.20 cm stainless steel blocks and one pure lead block of the same dimensions stacked together. Each of objects "A" "B" "C" and "D" was scanned in the collimated photon beam, typically in steps of 0.5 cm, across a range of about 10 cm that fully encompassed the features of the object. Measurements were made periodically without an object in place (100% transmission). A gamma ray spectrum was recorded at each position. A fission chamber located near the accelerator target was used to measure the accumulated neutron output from the accelerator during each measurement interval. The intensity of .sup.16 N decay photons available for radiography is directly proportional to the neutron field intensity for a steady-state condition of water flow in the system. These recorded neutron fluence data were used to normalize each photon transmission measurement. The exposure times for each sample position were generally about 5 minutes. Therefore, it took about an hour to scan each individual object and thereby generate the desired radiograph which displayed its characteristic features. Additional measurements were performed at various times in carrying out the present invention to determine the extent and origin of the background. One such set of measurements was made for a 10 cm-thick lead brick blocking the collimator that defined the photon source. Spectral data was also acquired with the water turned off (so that no .sup.16 N activity was transported from the target area to the radiography apparatus) and with the FNS accelerator turned off to determine ambient and cosmic ray background. These measurements showed that the signal-to-noise ratio for the arrangement used in the present invention was about 20-to-1, and that a significant portion of the background came from ambient sources and cosmic ray interactions. It was also found that there was little change in the shape of the spectrum produced by the .sup.16 N gamma rays when various objects were placed between the gamma ray source and detector for radiography investigation. In other words, although the spectrum yield was reduced, the actual appearance of the spectrum was not noticeably distorted by passage of the gamma rays through the various materials considered. This result served to indicate that most of the detected gamma rays were either primary ones or those which inexperienced at most only small angle scattering interactions that did not significantly alter their energies. The events recorded in each spectrum produced by the sodium iodide scintillation detector 52 were summed from just above the lower level cutoff defined by the pulse selector and linear gate to just below the position where the amplifier saturated. These spectral sums constituted the raw transmission data. It was not necessary to calibrate the response of the detector any further. This approach to the analysis of these experimental data was possible because the shape of the spectrum was not noticeably altered by the passage of photons through the studied objects. The summed counts were corrected for recording dead time, and were further adjusted for neutron exposure of the water, to yield values of relative transmitted photon intensity. The relative integrated neutron fluence for each measurement time interval was deduced from the output of a fission chamber neutron monitor as discussed above. Periodic measurements of gamma ray spectra with no object present defined the equivalent incident photon intensity I.sub.o so that meaningful transmission ratios I/I.sub.o could be calculated. One dimensional radiographs for the various investigated objects were constructed from these ratios. Referring to FIGS. 8a-8d, there are shown graphic results of one-dimensional photon scans of the objects respectively shown in FIGS. 4a, 4b; 5a, 5b; 6a, 6b; and 7a, 7b, as measured and recorded by the present invention. The indicated uncertainties are based on the combined statistics for the summed counts from the sodium iodide scintillation detector spectra and for the neutron fluence monitor counts. The data points are connected with solid lines to provide eye guides. The dotted line segments indicate values of the transmissions which were calculated using the exponential law equation for the transmission of photons through matter, in combination with photon cross sections and pertinent material parameters. Qualitative agreement is observed in regions where the transmission is "flat" versus scan distance. However, precise agreement should not be expected because of uncertainties in density, thickness and composition of the materials involved, and the effects of small angle photon scattering. As indicated above, most of the data were acquired in increments of 0.5 cm along the scanning direction. Scanning was accomplished by moving the investigated object past the fixed collimator system in the direction of the scanning arrows shown in the aforementioned figures. It is clear from the data presented that the spatial resolution observed for these radiographs is consistent with the dimensions of the collimator arrangement. The graphic representations shown in FIGS. 8a-8d provide evidence of the individual features of the investigated objects shown in FIGS. 4a, 4b through 7a, 7b. For example, FIG. 8a shows object "A" as uniform with no distinguishing features as is evident from the featureless one-dimensional radiograph of this figure The hidden hole in object "B" shown in FIGS. 5a, 5b is apparent from the large peak in FIG. 8b. Similarly, the lead block hidden between the two copper plates in object "C" as shown in FIGS. 6a, 6b appears as the large trough in the graphic representation of FIG. 8c. Finally, the iron-lead-iron discontinuity characterizing object "D" shown in FIGS. 7a, 7b appears as the deep trough in the graphic representation of FIG. 8d. The collimator geometry of the radiography apparatus 10 of the present invention shown in FIGS. 1 and 2 could be modified to provide improved resolution if the coupling of the circulating loop of water 12 to the neutron radiation field from the neutron source 15 were optimized. For example, with a factor of two orders of magnitude (10.sup.2) enhancement in gamma ray source strength, which should be quite feasible at the FNS facility, it would be possible to reduce the collimator dimensions to 0.5 cm.times.0.5 cm and still achieve the same statistical precision in the transmission data for exposures of equivalent duration. Two dimensional scans would be feasible using such a rectangular collimator, but an array of several small detectors would be necessary to permit radiographs to be generated in a more reasonable time than would be required for a single large detector arrangement. These changes could be implemented by simple engineering design revisions and would not involve changes in the fundamental principles of the present invention. There is a difference of about seven orders of magnitude (10.sup.7) in the photon intensity observed from the radioactive water produced in the present invention and that which is likely to be encountered with the cooling water exiting from a D-T fusion reactor such as the International Thermonuclear Experimental Reactor (ITER). With such enhanced gamma ray source strengths at a D-T fusion reactor facility, it would be possible to achieve much better resolution and far shorter exposure times than appears to be possible with any existing D-T neutron generator. Resolution on the order of 1 mm and exposures no longer than a few seconds could be easily obtained, even allowing for some reduction in the gamma ray source strength due to the time required to transport water from a D-T fusion reactor to the remote location where radiography is performed. Since the volume of radioactive water available would be very large, it would also be possible employing a continuous, extended sheet of radioactive water and a two-dimensional array of collimators and detectors to obtain a complete radiographic image of a large, complex object in a matter of a few seconds. There has thus been shown a radiography apparatus for producing and directing essentially monoenergetic gamma rays onto an object for radiographic analysis. The substantial penetrating power of the monoenergetic gamma rays allows for accurate determination of the thickness of an object under investigation as well as its elemental composition, particularly for metals and high atomic number materials. The monoenergetic gamma rays are generated by exposing a circulating loop of water to energetic neutrons which may be produced by irradiating a tritium target with a deuteron beam such as obtained from a D-T fusion neutron generator. Oxygen in the pure water in the circulating loop is activated via the .sup.16 O(n,p).sup.16 N reaction using 14-MeV neutrons produced at the neutron source via the .sup.3 H(d,n).sup.4 He reaction. The object to be analyzed is located at a remote location to which the water circulating in the loop flows. The characteristic decay half life of 7.13 seconds is sufficient to permit gamma ray generation at a remote location, while not presenting a chemical or radioactivity hazard because the radioactivity falls to negligible levels after one-two minutes. While particular embodiments of the present invention have been shown and described, it will be obvious to those skilled in the art that changes and modifications may be made without departing from the invention in its broader aspects. Therefore, the aim in the appended claims is to cover all such changes and modifications as fall within the true spirit and scope of the invention. The matter set forth in the foregoing description and accompanying drawings is offered by way of illustration only and not as a limitation. The actual scope of the invention is intended to be defined in the following claims when viewed in their proper perspective based on the prior art. |
053596390 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Preferred embodiments of this invention will be described in detail hereinafter with reference to the drawings. First Embodiment FIG. 1 is a view showing an outline of a CT apparatus in a first embodiment of this invention. FIG. 2 is a view showing an interior structure and a control system of the first embodiment. An X-ray tube 1 and an X-ray collimator 2 constitute an X-ray emitting device 3 secured to an inner peripheral surface of a rotary frame mounted in an opening of a gantry not shown. An X-ray detector 4 is secured to the rotary frame opposite the X-ray emitting device 3 to detect transmitted X-rays and produce electric signals corresponding thereto. The rotary frame is rotatable around an examinee M lying on a top board 9 and inserted into the gantry opening. Consequently, the X-ray emitting device 3 and X-ray detector 4 revolve in an opposed relationship to each other to scan sectional planes of concern of the examinee M. The X-ray tube 1 acting as an X-ray emitting means is connected to a high voltage generator 30 which supplies thereto a high voltage necessary for X-ray generation. The X-ray detector 4 has a data acquisition system (DAS) 31 for receiving detection signals and converting these signals into digital signals, an image reconstructor 32 for reconstructing sectional images based on the digital signals received, and a CRT 33 for displaying the reconstructed sectional images. The X-ray collimator 2 is disposed adjacent the X-ray tube 1 to adjust a slice thickness and diverging angle of X-rays emitted from the X-ray tube 1. The X-ray collimator 2 is formed of an X-ray shielding material such as brass or lead to have a cylindrical shape. The X-ray collimator 2 defines a slit 5a extending in X direction which is a direction of divergence of the X-rays emitted. (Y direction perpendicular to X direction corresponds to the axial direction of examinee M.) The X-rays entering the slit 5a exit through a slot 6a having a circumferential dimension determining a slice thickness "t", and a dimension in X direction determining a diverging angle .alpha. of the X-rays. A slot 6b defining an X-ray incidence port has a larger circumferential dimension than the slot 6a, and the same dimension in X direction as the slot 6a, wherefore the slit 5a is wedge-shaped. Thus, X-rays emitted from the X-ray tube 1 enter the slit 5a through the slot 6b and converge in departing from the slot 6a to form an X-ray beam B having a desired shape. To provide varied slice thicknesses, the X-ray collimator 2 defines, besides the slit 5a, a plurality of slits 5b and 5c arranged circumferentially thereof and having different opening dimensions in the circumferential direction. The X-ray collimator 2 has an axis extending in X direction, with an output shaft of a reversible drive motor 7 attached to one axial end thereof, the other end being attached to the rotary frame in the gantry opening through a bearing not shown. Thus, the X-ray collimator 2 is rotatable about its axis with rotation of the drive motor 7. Not only is it possible to select one of the slits 5a-5c but the slit 5a, for example, may be minutely shifted to swing the X-ray beam B, as converged through the slot 6a, in Y direction. Even with a shift of the slit 5a, the slot 6b having a large circumferential dimension can direct X-rays emitted from the X-ray tube 1 into the slit 5a. The rotation of the X-ray collimator 2, i.e. reversible rotation of the drive motor 7, is controlled by a control unit 8. The X-ray collimator 2 corresponds to the X-ray emitting direction switching means of this invention. The control unit 8 controls movement of the top board 9 supporting the examinee M besides the rotation of the X-ray collimator 2. The top board 9 includes a rack 10 extending longitudinally thereof, and a pinion 11 meshed with the rack 10. An output shaft of a reversible drive motor 12 is attached to the pinion 11. The pinion 11 is rotatable with rotation of the drive motor 12 to slide the top board 9 longitudinally (in Y direction) through the rack 10. The reversible rotation of the drive motor 12 is controlled by the control unit 8. The top board 9 corresponds to the examinee moving means of this invention. The control unit 8 comprises, for example, a microcomputer including a CPU (central processing unit), a CPU memory, a ROM (read-only memory) and the like. A sequence (program) of collecting data for sectional images, which will be described later, is stored in the ROM, and the CPU carries out processing according to the program. The CPU memory is used for temporarily storing data being processed. The control unit 8 corresponds to the control means of this invention. The X-ray detector 4 is in the form of a curved band and attached to the rotary frame in the gantry opening as noted hereinbefore. The X-ray detector 4 has a dimension in X direction to detect all X-rays diverging in X direction through the slit 5a (5b or 5c) and transmitted through the examinee M. Its dimension in Y direction is one to detect all X-rays when the X-ray collimator 2 is minutely rotated to photograph a plurality of (n) sectional images. The "n" is the number of sectional planes to be scanned by X-rays whose emitting direction is shifted in Y direction by the control unit 8 as described hereinafter. The sequence of collecting data of sectional images with the CT apparatus having the above construction will be described next with reference to the flowchart shown in FIG. 3. In this sequence, as shown in FIG. 1, sectional planes #1 to #4 of the examinee M are scanned one after another. The sectional planes #1 to #4 are close to one another and each has a 2 mm slice thickness. The control unit 8 moves the top board 9 supporting the examinee M to adjust the first sectional plane #1 to a position to which the X-ray beam B travels vertically from the X-ray collimator 2 (the position at this time of slit 5a in the X-ray collimator 2 being called an initial position) (step S1). In this state, the X-ray emitting device 3 and X-ray detector 4, as opposed to each other, are revolved around the examinee M to scan the sectional plane #1 (step S2). When the sectional plane #1 has been scanned, the control unit 8 minutely rotates the X-ray collimator 2 to shift X-rays 2 mm in Y direction to irradiate the next sectional plane #2 (step S3). In this state, the X-ray emitting device 3 and X-ray detector 4, as opposed to each other, are revolved around the examinee M to scan the sectional plane #2 (step S4). Although the X-rays are not perpendicular to Y direction at this time, no problem arises in image reconstruction since the sectional plane is scanned with one revolution therearound so that the center of irradiation of the X-rays with respect to the sectional plane is substantially perpendicular to Y direction. When the sectional plane #2 has been scanned, the control unit 8 rotates the X-ray collimator 2 back to the initial position (i.e. resets the collimator 2), shifting the X-rays -2 mm in Y direction, to scan sectional planes #3 and #4. Simultaneously with the reset operation, the control unit 8 starts an operation to move the top board 9 to adjust the sectional plane #3 to the initial position (steps S5 and S6). The sectional planes #3 and #4 are scanned as at steps S2 through S4 in place of sectional planes #1 and #2. When data of all the sectional planes #1 through #4 have been collected, the top board 9 is driven to withdraw the examinee M from the gantry (step S7). Where other sectional planes #5, #6 and so on are to be photographed, steps S2-S6 are repeated to scan two sectional planes by minutely rotating the X-ray collimator 2, and the top board 9 (examinee M) is moved each time two sectional planes have been scanned. In this way, data of numerous sectional planes may be collected. It is also possible to scan sectional planes #1-#3 first by minutely rotating the X-ray collimator 2, and move the top board 9 each time three sectional planes have been scanned. In this case, however, the X-ray detector 4 must have a sufficient dimension in Y direction to allow scanning of sectional planes #1-#3. Where, for example, the CT apparatus is capable of scanning a 10 mm slice thickness, a site of concern is divided into sectional planes each having a 2 mm thickness to be scanned. The resulting data are processed to form a sectional image of 10 mm slice thickness free of a partial volume artifact. In this sequence, therefore, the X-ray collimator 2 may be minutely rotated five times to effect a 2 mm shift in Y direction each time. The time taken in collecting data in the above sequence will be described with reference to the time chart shown in FIG. 4. As seen from FIG. 4, assuming that step S1 is completed at a point of time "0", the scanning of sectional plane #1 (step S2) consumes about 1 second, then the rotation of X-ray collimator 2 (step S3) about 0.05 second, and the scanning of sectional plane #2 about 1 second. Next, the resetting rotation of X-ray collimator 2 (about 0.05 second) is started simultaneously with the movement of top board 9 (about 3 seconds) (step S6). The resetting rotation of X-ray collimator 2 is finished during the movement of top board 9. Thereafter, the scanning of sectional plane #3 (step S2) consumes about 1 second, then the rotation of X-ray collimator 2 (step S3) about 0.05 second, and the scanning of sectional plane #4 about 1 second. The time consumed up to completion of the scanning of sectional plane #4 is 1+0.05+1+3+1+0.05+1 which is about 7.1 seconds. The movement of top board 9 is regarded as consuming about 3 seconds since moving the sectional plane #3 to the position of sectional plane #1 (4 mm) in one stroke is considered to consume a longer time than a 2 mm movement as described later. Next, the time taken in collecting data with a conventional apparatus will be described with reference to the time chart shown in FIG. 5. As seen from FIG. 5, assuming that the sectional plane #1 is set to a scan position (corresponding to step S1 above) at a point of time "0", the scanning of sectional plane #1 consumes about 1 second, and then about 2 seconds are consumed in moving the top board 9 to set the sectional plane #2 to the scan position. Next, about 1 second is consumed in scanning the sectional plane #2, about 2 seconds in moving the top board 9 to set the sectional plane #3 to the scan position, about 1 second in scanning the sectional plane #3, about 2 seconds in moving the top board 9 to set the sectional plane #4 to the scan position, and finally about 1 second in scanning the sectional plane #4. Thus, the time consumed up to completion of the scanning of sectional plane #4 is 1+2+1+2+1+2+1 which is about 10 seconds. The movement of top board 9 is regarded as consuming about 2 seconds since this is a 2 mm movement. A comparison between FIG. 4 and FIG. 5 shows that the time needed to collect the data of sectional planes #1-#4 has been reduced by as much as about 2.9 (10-7.1) seconds. In collecting the data of sectional planes #1-#6, for example, this embodiment takes about 12.15 (7.1+3+1+0.05+1) seconds whereas the conventional apparatus takes about 16 (10+2+1+2+1) seconds. Thus, the larger the number of sectional planes scanned is, the more processing time is saved by omitting the movement of top board 9, which promotes the processing time reduction. Second Embodiment A second embodiment will be described next with reference to FIG. 6, which employs a different X-ray collimator to swing X-rays in Y direction. This X-ray collimator 20 is penetrated and slidably supported by parallel support shafts 26 extending in Y direction. These support shafts 26 are fixed to inner walls of a rotary frame mounted in a gantry opening. The X-ray collimator 20 is formed of an X-ray shielding material such as brass or lead to have a plate shape, and defines slits 21a-21c extending in a diverging direction of X-rays. Each of the slits 21a-21c has dimensions to provide a diverging angle .alpha. and slice thickness "t" of the X-rays. The slit 21a, for example, is wedge-shaped with a slot 22b defining an X-ray incidence port thereof and having a larger dimension in Y direction than a slot 22a (not seen in FIG. 6) defining an exit port. The X-ray collimator 20 is slidable in Y direction by a drive mechanism described hereunder. The X-ray collimator 20 includes a rack 23 extending in Y direction, and a pinion 24 meshed with the rack 23. An output shaft of a reversible drive motor 25 is attached to the pinion 24. The pinion 24 is rotatable with rotation of the drive motor 25 to reciprocate the rack 23 in Y direction. The rotation of the drive motor 25 is controlled by a control unit, not shown, similar to the control unit 8 in the first embodiment. The movement of a top board is also controlled by the control unit. The other details of this embodiment may be the same as in the first embodiment. Following the sequence described with reference to FIG. 3, data of a plurality of sectional planes may be collected with the reciprocating movement in Y direction of the X-ray collimator 20 replacing the rotation of the X-ray collimator 2 at steps S3 and S6. Third Embodiment A third embodiment of this invention will be described next. In each of the first and second embodiments, the X-ray collimator 2 or 20 is rotatable or reciprocable to shift the X-ray emitting direction. In the third embodiment, the X-ray emitting device 3 including the X-ray tube 1 and X-ray collimator 2 or 20 is arranged slidable axially of the examinee M (in Y direction). A specific construction will be described with reference to FIGS. 7 and 8. FIGS. 7 and 8 depict a cylindrical X-ray collimator 2 as described in the first embodiment, and this embodiment will be described hereinafter using this collimator 2 as an example. Like reference numerals are used to identify like parts in FIGS. 1 and 2 which are the same as in the first embodiment and will not be described again. The X-ray collimator 2 is rotatably mounted in a case 40, with the X-ray tube 1 fixed to the top of the case 40, so that the X-ray tube 1 and X-ray collimator 2 are movable together. The case 40 has an upper surface and a lower surface including X-ray penetrable windows not shown. X-rays emitted from the X-ray tube 1 travel through the X-ray penetrable window in the upper surface of the case 40 into the slit 5a of the X-ray collimator 2. An X-ray beam B exiting the slit 5a travels through the X-ray penetrable window in the lower surface of the case 40 to the examinee M. The case 40 is supported by four slide rails 41 fixed to a rotary frame 42 rotatably mounted in a gantry. The case 40 includes a motor 43 for reversibly rotating a screw shaft 44 meshed with the rotary frame 42. The screw shaft 44 is rotatable with rotation of the motor 43 to slide the case 40 containing the X-ray collimator 2, together with the X-ray tube 1, in Y direction along the slide rails 41. The rotation of the motor 43 is controlled, as is the motor 12 for moving the top board 9, by a control unit 45 similar to the control unit 8 in the first embodiment. An X-ray detector 4 is attached to the rotary frame 42 in an opposed relationship to the X-ray emitting device 3. For scanning sectional planes, the X-ray collimator 2 is first rotated to select a desired slit. At this time, the X-ray collimator 2 is rotated to a position to allow the X-ray beam B exiting the slit to travel perpendicular to Y direction. In this state, the control unit 45 rotates the motor 43 to move the X-ray emitting device 3 to an initial position, and at the same time moves the top board 9 to set a first sectional plane of the examinee M to the initial position. After the first sectional plane is scanned, the control unit 45 rotates the motor 43 to slide the X-ray emitting device 3 in Y direction for enabling scanning of a sectional plane adjacent the first sectional plane scanned. In this way, a desired number of sectional planes are scanned by sliding the X-ray emitting device 3 in Y direction. Subsequently, the control unit 45 effects a reset control to rotate the motor 43 backward to return the X-ray emitting device 3 to the initial position. At the same time, the top board 9 is moved to set a next sectional plane to be scanned, adjacent the scanned sectional planes, to the initial position. This sectional plane and adjacent sectional planes are successively scanned by moving the X-ray emitting device 3 in Y direction as above. After a desired number of sectional planes are scanned by sliding the X-ray emitting device 3 in Y direction, the control unit 45 effects the reset control and moves the top board 9 to set a next sectional plane to be scanned to the initial position. The above operations are thereafter repeated until completion of data collection. With the X-ray emitting device 3 arranged slidable in Y direction as described above, the time taken in sliding the X-ray emitting device 3 in Y direction is shorter than the time taken in moving the top board 9. Consequently, as in the first and second embodiments, data collection requires a reduced processing time. Moreover, in the third embodiment, the X-rays are emitted perpendicular to Y direction for all sectional planes scanned after sliding the X-ray emitting device 3 in Y direction. The X-ray collimator mounted in the case 40 is not limited to the cylindrical collimator, but may be the plate-shaped X-ray collimator 20 (see FIG. 6) described in the second embodiment. Fourth Embodiment A fourth embodiment of this invention will be described next with reference to FIGS. 9 and 10. The fourth embodiment is characterized in that the X-ray emitting device 3 and X-ray detector 4 are synchronously slidable in Y direction. Like reference numerals are used to identify like parts in FIGS. 1, 2, 7 and 8 which are the same as in the first and third embodiments and will not be described again. Specifically, the X-ray emitting device 3 and X-ray detector 4 are movable in Y direction together, with a case 40 having an X-ray tube 1 fixed thereto and an X-ray collimator 2 rotatably mounted therein, and the X-ray detector 4, being attached to a support ring 50 in an opposed relationship with each other. The support ring 50 includes four motors 51 for rotating screw shafts 52 meshed with a rotary frame 42. The motors 51 are synchronously rotatable to slide the X-ray emitting device 3 and X-ray detector 4 in Y direction by means of the support ring 50. The respective motors 51 are controlled to rotate synchronously by a control unit 53 similar to the control unit 8 in the first embodiment. This control unit 53 also controls a motor 12 for moving a top board 9. With this construction, as in the third embodiment, a desired number of sectional planes may be scanned by sliding the X-ray emitting device 3 and X-ray detector 4 in Y direction to reduce the processing time for data collection. Further, in the fourth embodiment, as in the third embodiment, the X-rays are emitted perpendicular to Y direction for all sectional planes scanned after sliding the X-ray emitting device 3 in Y direction. The X-ray detector 4 in the fourth embodiment may have a width in Y direction just enough to enable scanning of one sectional plane. The X-ray collimator mounted in the case 40 is not limited to the cylindrical collimator, but may be the plate-shaped X-ray collimator 20 (see FIG. 6) described in the second embodiment. The present invention may be embodied in other specific forms without departing from the spirit or essential attributes thereof and, accordingly, reference should be made to the appended claims, rather than to the foregoing specification, as indicating the scope of the invention. |
047215976 | summary | FIELD OF THE INVENTION This invention relates to a method and apparatus for compacting spent nuclear reactor fuel rods and more particularly for preparing such spent fuel rods for long-term water pool storage. STATEMENT OF PRIOR ART Nuclear reactor installations employ nuclear fuel materials in the form of fuel rods which are supported in fuel rod assemblies. The fuel rods are metal pipes which are filled with nuclear fuel material and are about 0.4-0.6 inch in diameter and from 8 to 15 feet in length. Groups of 64, 128, 220 or more such fuel rods are assembled in a fuel rod assembly which includes grids for alignment and support of the fuel rods, a lower end fitting, an upper end fitting, and guide tubes. The fuel rod assembly is introduced into a nuclear reactor as the fuel source. After the nuclear fuel in the fuel rod assembly is spent to a pre-established level, the entire fuel rod assembly is withdrawn from the nuclear reactor and is stored vertically in appropriate metal racks in a wet pool until the radioactive properties have dissipated sufficiently for transfer to other storage locations. Within the fuel rod assembly, the individual fuel rods are spaced-apart in a pre-established array, usually a rectangular array. The fuel rod assemblies are spaced-apart in the array and are maintained under water in the reactor for the purpose of moderating or slowing the neutrons. In the fuel rod assembly, the ratio of cross-sectional area of fuel rod to cross-sectional area of water is approximately 1:1. At the present time, spent nuclear fuel rod assemblies are withdrawn from the nuclear reactors and are stored vertically in appropriate storage racks under water in storage pools without any deliberate change in the fuel rod assembly. The fuel rod storage pools are filled with the spent fuel rod assemblies whose activity has dissipated as a result of extended storage in the pool. A number of suggestions have been made for removing long-term storage fuel rod assemblies from the pool and for withdrawing individual spent fuel rods from the fuel rod assembly and thereafter for assembling the individual spent fuel rods in new containers or canisters wherein the fuel rods are more closely aligned, i.e., more densely compacted, and for returning such newly filled canisters to appropriate storage racks within a water storage pool for long-term storage or until appropriate fuel recovery processing is economically feasible. None of these compacting processes have been carried out except with simulated fuel rod assemblies containing simulated fuel rods. Some of the anticipated difficulties with the proposed fuel rod compacting processes which have been suggested arise from the knowledge that the actual fuel rods are twisted and bent out of alignment as a result of their long-term exposure in nuclear reactors. In some cases, the distortion may be as much as 1.5 inches in an 8-foot long rod. Such permanent distortion of the fuel rods will interfere with the proposed alignment techniques. The casing of the fuel rods is usually embrittled due to irradiation in the nuclear reactor. A further problem is that the long, thin fuel rods are whippy and may be difficult to manipulate. A still further problem relates to the inherent safety of compacting spent fuel rods. There is a possibility that the fuel rods might become spaced-apart by a critical distance while removed from the fuel rod assembly and before compaction and confinement in a storage canister. Such possibility should be precluded. At the present time there is a need to compact spent fuel rods which are contained in wet storage pools in the fuel rod assemblies. STATEMENT OF THE INVENTION According to the present invention, a method and related apparatus are proposed for transferring spent fuel rods from a fuel rod assembly in an underwater pool directly into a fuel rod canister where the density of the fuel rods greatly exceeds the fuel rod density in the fuel rod assembly. As a result of the present invention, the spent fuel storage capacity in a particular water storage pool can be approximately doubled. Moreover, the fuel rod consolidation process of the present invention is carried out without altering the relative position of the fuel rods whereby after consolidation the identity of a fuel rod is known at each position in the fuel rod canister which facilitates accounting procedures. According to the invention, the top end of a fuel rod assembly is removed, by cutting or otherwise, and the exposed fuel rod tops are individually connected by welding to individual pulling members such as tubes or other tensioning devices. The pulling elements are presented in an array which corresponds to the array of the individual fuel rods. The pulling elements are drawn through a fuel rod directing chamber such as a transition funnel which has a relatively wide cross-section at its base corresponding to the array of the tops of the fuel rods in the fuel rod assembly. The transition funnel at its top end has a relatively narrow cross-section which corresponds to an array of fuel rods in a compact storage presentation. For each individual pulling element there is a separate guide within the transition funnel for directing the pulling element and the fuel rod which is welded thereto so as to pull the fuel rod from the fuel rod assembly through the transition funnel into a permanent storage container which is positioned above the top of the transition funnel. The pulling elements each includes welding means at their lower ends which can pass downwardly through the container and through the transition funnel. The upper ends of the pulling elements are connected to a tensioning device such as a reeling drum to permit movement of each pulling element and the associated fuel rod upwardly out of the fuel rod assembly into a selected one of the passageways through the transition funnel and thence into a pre-established position in a compacted array of fuel rods within the container. Welding means are provided at the bottom end of each pulling element for securing one fuel rod. A preferred pulling element is a plastic tube and a preferred welding means is an arc welder, e.g., an inert gas metal-arc type welder having an electrode extending to form an arc gap with the bottom wall of a fixture that is releasably connected to the plastic tube. After the fixture is welded to the fuel rod, the fuel rod is pulled into the canister. Thereafter, the pulling element is separated from the fuel rod by breaking the weld joint or by withdrawing the plastic tube from the fixture. The upper ends of the plastic tube forming the pulling elements are connected to a wench by a cable through a manifold used to supply inert gas to the plastic tubes for the welding process. Preferably the individual fuel rods are withdrawn upwardly concurrently from the common fuel rod assembly so that the upper ends of all of the fuel rods enter into the container at about the same level to facilitate stacking within the container. Preferably, within the container, the array of spent fuel rods is a rectangular array which provides maximum fuel rod density in the container. Preferably the fuel rod density in the container is approximately twice that of the fuel rod density in the fuel rod assembly. The transition funnel is so arranged that the guide tubes therein merge toward one another. As a consequence, the fuel rods, in passing from the fuel rod assembly into the fuel rod container, do not move apart so that critical distances between fuel rods cannot occur. Also, since the fuel rods are advanced under tension they can be straightened in the guide tubes without breakage due to irradiation embrittlement. By providing fuel rod containers of the same cross-sectional dimensions as the fuel rod assemblies, the containers can be stored in the same underwater fuel rod storage racks which have been employed for the fuel rod assemblies. When the present invention is practiced accordingly, the capacity of the fuel rod storage pools for spent nuclear fuel rods can be approximately doubled. The structural components of the empty fuel rod assembly are collected and stored for appropriate disposal. Accordingly, it is an object of this invention to provide a method for transferring spent fuel rods from a fuel rod assembly directly into a compact fuel rod container for compact storage of the spent fuel rods. It is a further object of this invention to carry out the described method without extracting the fuel rods above the surface of the water in the fuel rod storage pool in an area of the storage pool which is minimized by the fact that the fuel rods are moved unidirectionally from a fuel rod assembly and to a standard canister. |
summary | ||
claims | 1. A composition of matter comprising 195m Pt characterized by a specific activity of at least 30 mCi/mg Pt. 2. A composition of matter in accordance with claim 1 further characterized by a specific activity of at least 50 mCi/mg Pt. claim 1 3. A composition of matter in accordance with claim 2 further characterized by a specific activity of at least 70 mCi/mg Pt. claim 2 4. A composition of matter in accordance with claim 3 further characterized by a specific activity of at least 90 mCi/mg Pt. claim 3 5. High-specific-activity 195m Pt made by a method comprising the steps of: a. exposing 193 Ir to a flux of neutrons sufficient to convert a portion of said 193 Ir to 195m Pt to form an irradiated material; b. dissolving said irradiated material to form an intermediate solution comprising Ir and Pt; and c. separating said Pt from said Ir by cation exchange chromatography to produce a product comprising 195m Pt. 6. High-specific-activity 195m Pt in accordance with claim 5 wherein said dissolving step is carried out at a temperature of at least 210xc2x0 C. claim 5 7. High-specific-activity 195m Pt in accordance with claim 6 wherein said dissolving step is carried out at a temperature of at least 217xc2x0 C. claim 6 8. High-specific-activity 195m Pt in accordance with claim 5 wherein said intermediate solution further comprises aqua regia. claim 5 9. High-specific-activity 195m Pt in accordance with claim 5 wherein said separating step further comprises the steps of: claim 5 a. loading said intermediate solution onto a cation exchange column; b. eluting said Pt with a first eluent solution comprising HCl and thiourea. c. eluting said Pt with an essentially thiourea-free second eluent solution comprising HCl. 10. High-specific-activity 195m Pt in accordance with claim 5 wherein said 195m Pt product is characterized by a specific activity of at least 30 mCi/mg Pt. claim 5 11. High-specific-activity 195m Pt in accordance with claim 10 wherein said 195m Pt product is further characterized by a specific activity of at least 50 mCi/mg Pt. claim 10 12. High-specific-activity 195m Pt in accordance with claim 11 wherein said 195m Pt product is further characterized by a specific activity of at least 70 mCi/mg Pt. claim 11 13. High-specific-activity 195m Pt in accordance with claim 12 wherein said 195m Pt product is further characterized by a specific activity of at least 90 mCi/mg Pt. claim 12 |
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