JP5522427B2 - A method for converting long-lived fission products to short-lived nuclides. - Google Patents

Description

  The present invention relates to a method for efficiently converting long-lived fission products generated at nuclear power plants and nuclear research facilities into short-lived nuclides.

  Among the wastes generated from nuclear facilities, radioactive fission products with a long half-life represented by radioactive iodine-129 (hereinafter referred to as I-129) and radioactive technetium-99 (hereinafter referred to as Tc-99) are disposed of. Therefore, there is a method of fixing by vitrification and disposing of the geological formation (see, for example, Non-Patent Document 1).

  However, since the half-lives of I-129 and Tc-99 are 15.7 million years and 210,000 years, respectively, the solidified body is released to the outside by long-term diffusion even if it is fixed by vitrification. Even after geological disposal, management and confirmation over a long period of time are required. For this reason, in order to improve the safety after disposal, it is necessary to deepen the disposal depth, which increases the economic burden of the facility.

  In this way, long-lived radioactive fission products are in a situation where final disposal is difficult, so wastes containing I-129 generated from nuclear facilities such as nuclear power plants and nuclear fuel reprocessing plants are chemically treated. After conversion to stable iodine by treatment, etc., an annihilation target is prepared, positioned in the irradiation section provided facing the reactor core, irradiated with neutrons, and the nuclides themselves are transmuted by neutron capture reaction. A method for reducing radioactivity has been considered (Patent Document 1). In addition, the iodine containing I-129 is separated by chemical and physical methods and recovered as a simple iodine, then sealed in a container such as quartz glass or zirconium alloy and irradiated with thermal neutrons to reduce I-129 radioactivity. A method has also been proposed (Patent Document 2).

  However, Patent Documents 1 and 2 only describe a pretreatment method for stabilizing unstable iodine, and a problem of how to treat a large amount of long-lived radioactive fission products. Is not solved at all.

  A fuel assembly loaded in a fast reactor composed of a cladding tube containing a fission product and a compound containing hydrogen atoms has also been proposed (Patent Document 3). Patent Document 3 describes that a Tc-99 rod is loaded on the outermost cladding tube of the fuel assembly, and a Tc-99 rod or zirconium hydride is loaded on the central cladding tube.

  However, in the conventionally proposed methods using fast reactors and light water reactors, the amount of long-lived radioactive fission products disappeared by transmutation is slightly larger than the amount of long-lived radioactive fission products newly generated by fission. It is difficult to dramatically increase the effective transmutation amount, that is, the annihilation amount of long-lived radioactive fission products. On the other hand, in the transmutation using a fusion device, although there is no long-lived radioactive fission product newly generated by fission, a large amount of long-lived radioactive fission product cannot be loaded due to the structure of the device. In addition, although the transmutation efficiency can be increased with an accelerator, it is difficult to increase the absolute amount of transmutation. Furthermore, use of a test and research reactor results in high processing costs because no energy is produced. As described above, with the methods proposed so far, a large amount of long-lived radioactive fission products that can withstand practical use could not be extinguished by transmutation.

JP 10-48390 A JP-A-5-40198 JP-A-8-194082

Nuclear Industry, Vol. 39, No. 6 (1993), pp. 6-20

  It is an object of the present invention to provide a method for efficiently converting as much long-lived fission products as possible into short-lived nuclides by greatly exceeding the amount of annihilation due to nuclear transmutation compared with the amount of new products produced by fission. is there.

  According to the present invention, an irradiation target obtained by pelletizing one or more long-lived fission products selected from Tc-99, I-129, Se-79, and Cs-135 together with graphite is used as an inner reflector region. A method for converting a long-lived fission product into a short-lived nuclide characterized by loading the inner reflector region of a graphite moderation reactor having a fuel region and an outer reflector region and performing thermal neutron irradiation Is provided.

  Long-lived fission products (LLFP) that can be treated in the present invention are Tc-99, I-129, Se-79, and Cs-135, and Tc-99 and I-129 are particularly about 20 barn. It is preferable because it has a nuclear reaction cross section with a relatively large thermal neutron.

  An irradiation target may also be loaded into the outer reflector region of the graphite moderator reactor.

  The irradiation target is a coated particle type pellet formed by dispersing particles formed by coating microspheres of long-lived fission products with SiC or low density carbon in graphite and / or dispersing long-lived fission products in graphite. It is preferable that a plurality of the dispersion-mixed pellets thus prepared are loaded into a graphite vessel of the same type as the fuel rod for a graphite moderator nuclear reactor. Further, the irradiation target is more preferably an aggregate of a plurality of graphite containers of the same type as the high temperature gas reactor fuel rod loaded with a plurality of the pellets. In this way, a plurality of irradiation targets are loaded into a graphite vessel of the same type as the graphite rod fuel rod, and a plurality of rod-like graphite vessels are assembled into an aggregate, thereby reducing the graphite deceleration reactor. The inner reflector region and the outer reflector region can be easily loaded, and the transmutation efficiency by one thermal neutron irradiation can be remarkably improved.

  The content ratio of the coated particle containing the long-lived fission product in the coated particle type pellet is 10 to 30% by volume, and the content ratio of the long-lived fission product in the dispersed mixed type pellet is 3 to 24% by volume. Is preferred. By setting the content ratio within the above range, long-lived fission products can be uniformly dispersed without being biased during baking. By including a small amount of long-lived fission product in the graphite in this way, it is possible to prevent a decrease in the reactivity of the reactor and a decrease in the neutron flux in the vicinity of the irradiation target due to the self-blocking effect. There is no loss of effectiveness. In terms of the core load, the GTHTR300 (a high temperature gas reactor having a thermal output of 600 MX and an electrical output of 300 MW) having an inner reflector region with a diameter of about 3 m and a height of about 8 m and a diameter of about 7.5 m and a height of about 10 m. When the core is loaded, it is 696 to 2087 kg in the case of coated particle type pellets, and 1292 to 10399 kg in the case of dispersed mixed type pellets, and the reactor reactivity decrease is a maximum 2.6% Δk / k from FIG. Become.

  It is preferable that the graphite volume occupation reduction ratio with respect to the total volume after loading the irradiation target in the inner reflector region of the graphite moderation reactor is 5% or less. By setting the reduction rate of the graphite volume of the graphite decelerating reactor to 5% or less, it is possible to reduce the influence on the operation characteristics of the reactor such as the change in reactivity and the heat capacity.

  According to the present invention, the loading position, form, and amount of the irradiation target assembly are set so as to minimize the decrease in the reactivity of the reactor, and the adjustment is made so that changes in the core characteristics are negligible and can be ignored. Thus, long-lived fission products can be efficiently converted into short-lived nuclides.

  The inner reflector region of the graphite moderation reactor has many thermal neutron components, and the irradiation target assembly is less subject to irradiation damage due to fast neutrons, so that the deterioration over time is small and the useful life of the irradiation target assembly is extended. Further, it is not necessary to slow down fast neutrons leaking from the core as in the case of using a fast reactor, and the irradiation target assembly can be loaded into the entire core central region, and the total volume available for transmutation is large. In addition, by reducing the concentration of long-lived fission products as coated particle type pellets and dispersion type pellets using graphite as a matrix, the reactivity of the reactor is reduced and the neutron flux in the vicinity of the aggregate is prevented from decreasing due to the self-shielding effect. be able to. On the other hand, by loading a plurality of rod-shaped containers loaded with a plurality of irradiation targets into the reactor as an aggregate, a long life fission product can be loaded in large quantities, and the total amount that can be reduced by transmutation can be increased. . Therefore, according to the present invention, it is possible to greatly exceed the amount of annihilation by transmutation compared to the amount of new production by fission, and to convert as much long-lived fission products as possible into short-lived nuclides as efficiently as possible. , More efficient waste disposal / disposal.

FIG. 1 is a horizontal cross-sectional view and a vertical cross-sectional view of a core of a graphite moderation reactor showing a state before loading an irradiation target containing a long-lived fission product. FIG. 2 is a horizontal cross-sectional view and a vertical cross-sectional view of the core of a graphite moderation reactor showing a state in which an irradiation target containing a long-lived fission product is loaded. FIG. 3 is a conceptual diagram of an irradiation target including a long-lived fission product, an assembly composed of the irradiation target, and a core loaded with the assembly. FIG. 4 is a graph showing the relationship between the loading amount of the long-lived fission product loaded in the graphite moderation reactor and the change in reactivity of the graphite moderation reactor. FIG. 5 is a diagram showing the amount of long-lived fission products (Tc-99 and I-129) loaded into the graphite moderation reactor, which are transmuted during the reactor operation, according to the method of the present invention.

  The present invention will be specifically described with reference to the drawings, but the present invention is not limited thereto.

  As shown in FIG. 3, in the case of a graphite moderation nuclear reactor having an annular fuel arrangement, particularly a high temperature gas reactor, a large graphite region having a diameter of about 3 m and a height of about 8 m is provided in the center of the core for the purpose of improving thermal characteristics.

  FIG. 1 (a) shows a horizontal section of the core of the HTGR and FIG. 1 (b) shows the same vertical section. It is possible to fill a material other than graphite to a few percent of the graphite in this region, that is, to the extent that the thermal performance is not degraded. FIG. 2 (a) shows a horizontal section of the core when an irradiation target containing a long-lived fission product (hereinafter abbreviated as “LLFP”) and graphite is loaded, and FIG. 2 (b) shows the vertical section.

  The irradiation target is preferably configured as an LLFP aggregate shown in FIG. The LLFP assembly has a structure called a pin-in-block type, and about 50 LLFP rods for transmutation are loaded on a hexagonal column graphite block. The LLFP assembly is formed by loading a plurality of graphite rod-like containers of the same type as the fuel rod. A graphite rod-shaped container is loaded with a plurality of irradiation targets including pellet-like graphite and LLFP. Hereinafter, a graphite rod-shaped container loaded with a pellet-shaped irradiation target is referred to as an LLFP rod. As the irradiation target pellet, a coated particle type pellet and a dispersion type pellet can be used. The coated particle type pellet was produced by dispersing particles in which LLFP fine particles were triple-coated with SiC or low-density carbon in graphite and baking it. Dispersion type pellets were produced by dispersing LLFP fine particles in graphite and baking them.

  FIG. 3 shows the structure of an LLFP rod composed of aggregate blocks and coated particle type (CP type) or LLFP dispersion type (MP type) pellets. As an example, LLFP loaded into the core of a GTHTR300 (thermal gas power 600MX, electric power 300 MW high temperature gas reactor) having a diameter of about 7.5 m × height about 10 m with an inner reflector area of about 3 m in height × about 8 m. 344 assemblies, 57 LLFP rods loaded per block of one LLFP assembly, filling rate of CF type pellets into LLFP rods of 30% by volume, CP type pellets containing 16 LLFPs .15 volume%, LLFP Tc-99 (density 9.2 g / cc) 60 volume% and I-129 (YI3 density 4.4 g / cc, I density ratio 80%, I-129 isotope ratio 73%) If set to 40 volume% mixing, the total core loadings of Tc-99 and I-129 would be about 2,000 kg and 400 kg, respectively. In this case, the LLFP concentration with respect to graphite is 0.7%, and the graphite volume reduction in the central part of the core before and after the assembly loading is only 0.7%, and the influence on the thermal characteristics is negligible. On the other hand, if the MP type pellet is set to 344 aggregates, 57 block loads, and the volume mixing ratio of LLFP formed into graphite is 24%, the total core loading weight of Tc-99 and I-129 is about 10 respectively. 2,000 kg and 2,000 kg. In this case, the graphite volume reduction is 3.5%. It can be seen from FIG. 4 that the reactivity reduction of the reactor due to the above setting is 1.9% Δk / k and 2.6% Δk / k, respectively. In general, in the case of a high-temperature gas reactor, if the reactivity change of the core is up to −3% Δk / k, there is almost no influence, and the influence on the operation characteristics of the reactor is small. Note that the reactivity decrease of the reactor shown in FIG. 4 was obtained from the calculation of the Monte Carlo code of the three-dimensional system model of the graphite moderation type reactor.

  FIG. 5 shows a decrease characteristic of the LLFP amount accompanying the reactor operation. Tc-99 has an operating period of 14.5 years for reducing the initial load by half, and the half-life of natural decay can be greatly shortened to 210,000 years. The number of years that I-129 is halved is 7.5 years, and the half-life of 1570 years can be shortened as well. In the above-mentioned CP type pellet setting, 3.3 million units of Tc-99 and 4.7 units of I-129 are generated in a 1 million kilowatt class light water reactor per year, and 300,000 kilowatt class high temperature gas reactor for one year. It can be processed by driving. Furthermore, if MP type pellets with increased loading are adopted, more than 10 LLFP per year can be eliminated.

  The embodiment of the present invention is not limited to the above example, and various loading methods are possible as long as the reactor characteristics are not significantly impaired.

  The method of the present invention makes it possible to convert a large amount of long-lived fission products generated from nuclear power plants and the like into short-lived nuclides, thereby reducing the size of facilities for storage and geological disposal of fission products to be processed. Therefore, facility operation and construction costs can be reduced.

Claims (6)

An irradiation target obtained by pelletizing one or more long-lived fission products selected from Tc-99, I-129, Se-79 and Cs-135 together with graphite, and a graphite inner reflector region in the center of the core; Loading a fuel region around the graphite inner reflector region and an inner reflector region of a graphite-moderated reactor having a graphite outer reflector region around the fuel region, and performing thermal neutron irradiation wherein, the method for converting the long-lived fission products into short-lived nuclides.   The method of claim 1, further loading the irradiation target into a graphite outer reflector region of the graphite moderator reactor. The irradiation target, the long life SiC microspheres of fission products or low density carbon coated with comprising particles formed by dispersing in a graphite coated particle types pellets and / or the long-lived fission products through the graphite The method according to claim 1 or 2, wherein a plurality of dispersion-mixed type pellets dispersed in a graphite vessel are loaded into a graphite vessel of the same type as the graphite rod fuel rod.   The method according to claim 3, wherein the irradiation target is an aggregate formed by collecting a plurality of graphite containers of the same type as a high temperature gas reactor fuel rod loaded with a plurality of the pellets. The content of the coated particles comprising the long-lived fission products in coated particle types pellets is 10 to 30 volume%, the content of the long-lived fission products in the dispersive mixing type pellets with 3 to 24 volume% The method according to any one of claims 1 to 4, wherein:   6. The method according to claim 1, wherein a graphite volume occupancy reduction rate with respect to a total volume after loading the irradiation target in the graphite inner reflector region of the graphite moderation nuclear reactor is 5% or less.

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