JP2015161568A - Onsite spend nuclear fuel treatment method - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a spent nuclear fuel treatment method small in size, capable of being installed even in a nuclear power plant, and high in transuranium element reduction efficiency.SOLUTION: An onsite spent nuclear fuel treatment method comprises: a shear process 20 of shearing fuel rods of spent nuclear fuel assemblies 10; a dissolution process 21 of chemically dissolving the sheared spent nuclear fuel and separating an insoluble residue 14 mainly containing platinum groups from a solution 15 containing nuclear fission products, uranium, and transuranium elements; a uranium extraction process 22 of separating only uranium from the solution 15 containing the nuclear fission products, uranium, and transuranium elements, and separating a uranium solution 16 mainly containing uranium from a mixture 17 of the nuclear fission products and the transuranium elements; and a neutron irradiation process 23 of irradiating the mixture of the nuclear fission products and the transuranium elements with neutrons and reducing a transuranium element content in the mixture 17 to 1/100 or less.

Description

  The present invention relates to an on-site type spent nuclear fuel processing method that can be installed in a nuclear power plant.

  In the processing of spent nuclear fuel generated by the operation of commercial light water reactors, uranium 235 and plutonium 239 that can be reused by reprocessing are taken out, and the remaining high-level radioactive waste is geologically disposed as a stable vitrified material. High-level radioactive waste contains transuranium elements with a very long half-life, and it takes a period of several hundred thousand years or more to reach the same level of radiotoxicity as uranium ore. It can be said that it is impossible to prove and guarantee such long-term storage in advance.

  When protons are accelerated to 1 GeV by a high energy accelerator and irradiated to a heavy metal target, a large amount of neutrons are generated by the spallation reaction. In Non-Patent Document 1, an accelerator-driven subcritical reactor is configured using this spallation reaction, and the above storage period is increased from several hundred thousand years by burning the transuranium element in the accelerator-driven subcritical furnace. Means have been proposed for shortening to thousands of years.

Yoshitaka Kimura, “High Energy Accelerator”, Kyoritsu Publishing Co., Ltd., Experimental Physical Science Series, Vol. 7, p. 161-167

  However, as described above, the spent nuclear fuel produced from one nuclear power plant by the accelerator-driven subcritical reactor using the spallation reaction, that is, the spent nuclear fuel for 1 to 5 million kW class commercial light water reactors. Is required, a proton beam accelerator having an acceleration energy of 1 GeV and an acceleration current of 15 mA is required. With the current accelerator technology, only a superconducting linear accelerator can realize such a high-power, high-energy proton beam accelerator having a beam output of about 15 MW. Moreover, in the case of a superconducting linear accelerator, the length of the accelerator exceeds 500 m, and it is difficult to install it in the nuclear power plant premises. Also, in the superconducting accelerator, the power consumption of the cooling device is large, and the high frequency power source for driving each acceleration cavity requires a large output current capacity, which reduces the power efficiency of the high frequency power device. The accelerator efficiency is 1% or less.

  Therefore, in order to achieve a beam output of 15 MW, it is necessary to supply power of 1.5 GW or more as accelerator driving power. If all of this accelerator drive power is supplied from a light water reactor in a nuclear power plant, supplying this accelerator drive power will produce new transuranium elements in the reactor, and further considering transmission and substation efficiency, As a whole, there was a problem that it was impossible to eliminate the burning of the transuranium element efficiently.

  The present invention has been made in view of such circumstances, and a main object of the present invention is to provide an on-site type spent nuclear fuel processing method that can be installed in a nuclear power plant premises and has high transuranium element combustion efficiency. is there.

  In order to achieve the above object, an on-site spent nuclear fuel processing method according to the present invention includes a shearing process of shearing fuel rods of a spent nuclear fuel assembly, and chemically dissolving the sheared spent nuclear fuel, From a dissolution process that separates into an insoluble residue mainly composed of platinum group, a fission product, a solution containing uranium and a transuranium element, and a solution containing the fission product, uranium and a transuranium element, uranium Uranium extraction process that separates only the uranium solution mainly composed of uranium, a mixture of fission product and transuranium element, and neutron irradiation to the mixture composed of fission product and transuranium element And a neutron irradiation process for reducing the amount of transuranium element in the mixture.

  It is desirable to reduce the total weight of all kinds of transuranium elements contained in the mixture to 1% by weight or less by the neutron irradiation process.

In the on-site type spent nuclear fuel processing method, in the neutron irradiation process, the fission product has a neutron flux density of 1 × 10 ¹⁴ n / cm ² · second to 1 × 10 ¹⁶ n / cm ² · second. It is desirable to irradiate a mixture of transuranium elements with neutrons.

In the neutron irradiation process, a molten salt nuclear reactor neutron source is applied as a neutron source, and a mixture of uranium tetrafluoride and a fluoride salt is applied to the liquid nuclear fuel loaded in the molten salt nuclear reactor. As for uranium, it is preferable that the isotope ratio of 99% or more is uranium tetrafluoride by uranium 233, and the uranium tetrafluoride by other uranium isotopes is 1% or less.
In addition, when a lithium compound is added to the fluoride salt for the purpose of, for example, reducing the melting point of the molten salt, it is desirable that 100% of the lithium salt is a lithium compound of lithium 7.

  Lithium 7 is obtained by irradiating boron with neutrons.

  According to the on-site spent nuclear fuel processing method according to the present invention, only uranium is separated from the light water reactor spent nuclear fuel, and the transuranium elements and fission products present in the remaining residue from which uranium has been separated are converted into molten salt raw materials. Since the uranium element is burned by irradiating neutrons from the reactor neutron source, no extra uranium element is generated in the process, and a spent nuclear fuel processing method having a high uranium element combustion efficiency can be obtained. Moreover, since a high energy accelerator is not used, the apparatus can be reduced in size, and on-site type spent nuclear fuel processing that can be installed in a nuclear power plant premises becomes possible.

It is process drawing which shows one Embodiment of the on-site type | mold spent nuclear fuel processing method which concerns on this invention. It is a perspective view which shows the general external shape of the molten salt reactor neutron source used for the neutron irradiation process of FIG. It is a longitudinal cross-sectional view which shows the core structure and reflector structure of the molten salt reactor neutron source of FIG. It is an AA line expanded transverse sectional view showing the core structure of the molten salt reactor neutron source of FIG. It is a perspective view which shows one Embodiment of the to-be-irradiated object storage container of the on-site type | mold spent nuclear fuel processing method which concerns on this invention. It is a graph which shows the example of the position of the to-be-irradiated object storage container at the time of neutron irradiation, and the neutron flux distribution of the to-be-irradiated object storage container part of the on-site type spent nuclear fuel processing method which concerns on this invention. It is a flowchart which shows the process which manufactures the lithium 7 used for the on-site type | mold spent nuclear fuel processing method which concerns on this invention. It is sectional drawing which shows one Embodiment of the molten salt reactor neutron source used for the neutron irradiation process of FIG. As a comparative example of this invention, it is a graph which shows the neutron flux distribution of the irradiated object storage container part at the time of loading the irradiated object storage container at the time of neutron irradiation with the transuranium element which does not contain plutonium.

  The on-site type spent nuclear fuel processing method according to the present invention will be described below with reference to the drawings and tables.

  FIG. 1 is a process diagram for explaining an embodiment of an on-site type spent nuclear fuel processing method according to the present invention. FIG. 1 shows a case where uranium nuclear fuel that has been subjected to 33 GW · d / t combustion in a pressurized water reactor and passed through a cooling period of 3 years is applied to the on-site spent nuclear fuel processing method according to the present invention. The process configuration procedure is shown in a flow diagram. As shown in FIG. 1, the on-site spent nuclear fuel processing method according to the present invention has a configuration in which four processes of a shearing process 20, a melting process 21, a uranium extraction process 22, and a neutron irradiation process 23 are arranged in cascade. have.

  In the shearing process 20, after the end piece 11 is removed from the spent nuclear fuel assembly 10, the fuel rods in the spent fuel assembly 10 are cut into a size of several centimeters in length, and the so-called hull is covered with a cladding tube. Spent nuclear fuel pellets 13 are produced in the finished state. In the case of pressurized water reactor uranium nuclear fuel, about 530 kg of spent nuclear fuel is stored in the spent nuclear fuel assembly 10, and after the cooling period of 3 years, the components shown in Table 1 are included therein. .

In the dissolution process 21, first, the spent nuclear fuel pellets 13 generated in the shearing process 20 are treated with, for example, heated nitric acid. In this dissolution process, hull does not dissolve in nitric acid, and can be recovered as the residual cladding tube 12. Further, uranium, transuranium elements and fission products of Table 1 are dissolved in the solution portion, and the platinum group remains in the insoluble residue portion. The insoluble residue portion contains a very small amount of fission products, and the platinum group aggregate 14 is subjected to glass solidification treatment and discarded as low-level radioactive waste.
On the other hand, with respect to the solution portion 15, only uranium is extracted using, for example, the Purex method using TBP (tributyl phosphate) as an extraction solvent, and separated as a uranium aggregate 16. The uranium aggregate 16 is discarded as deteriorated uranium. Overseas, uranium waste is subject to radioactive waste regulations during transportation, but once it arrives at the disposal site, this regulation is lifted and disposed of in the same manner as other industrial waste.

  Transuranic elements and fission products remain in the solution after uranium extraction. These residues are taken out as a powdery mixture 17 by evaporating and thermally decomposing the solution after uranium extraction, and sent to the neutron irradiation process 23. Mixture 17 contains 14.6 kg of fission product and 5.62 kg of transuranium element. The neutron irradiation process 23 has a function of burning (fission) and extinguishing the transuranium element by irradiating the mixture 17 with neutrons.

  Next, the operation of the embodiment of the on-site type spent nuclear fuel processing method according to the present invention will be described focusing on the neutron irradiation process 23. In the neutron irradiation process 23, the mixture 17 is irradiated with neutrons using the molten salt reactor neutron source 30 whose conceptual diagram is shown in FIG. In FIG. 2, the code | symbol 31 has shown the neutron reflecting material which consists of graphite or beryllium oxide etc., for example. The neutron reflector 31 has a cylindrical outer shape with a diameter of 3 m and a height of 4 m, and a core of a molten salt nuclear reactor is disposed at the center thereof. Reference numeral 32 denotes an irradiation hole for inserting the irradiated object of neutrons into the core of the molten salt nuclear reactor, specifically, an inner hole of a bottomed pipe.

  FIG. 3 is a cross-sectional view of the molten salt reactor neutron source 30 of FIG. The core has a diameter of 1.2 m and a height of 3 m, and is built in a nickel alloy core vessel 33. In the reactor core 33, a plurality of neutron reflectors 34A, 34B, 34C and 34D and a molten salt nuclear fuel 35, which are concentrically spaced from each other, are stored. Here, the neutron reflectors 34 </ b> A, 34 </ b> B, 34 </ b> C, and 34 </ b> D are made of the same material as the neutron reflector 31. A horizontal sectional view (transverse sectional view) of the core of FIG. 3 is shown in FIG. The irradiation hole 32 has a diameter of 15 cm and is disposed so as to penetrate the center of the core.

  FIG. 5 is an external view of an irradiated object container 40 that houses the neutron irradiated object. The irradiated object storage container 40 is made of a nickel alloy having a thickness of about 2 mm, and has a diameter of 14.5 cm and a height of 1.5 m. When one spent nuclear fuel assembly is processed, the fission product / transuranic element mixture 17 output from the uranium extraction process 22 has a weight of 20.2 kg. The volume of the irradiated object storage container 40 is about 25 liters, and has a volume that can store all of about 20 kg of the mixture 17.

FIG. 6 is a graph showing the neutron irradiation position of the irradiated container 40 containing the mixture 17 and the radial neutron flux distribution in the irradiation effective length portion of the molten salt reactor neutron source in a two-group approximation. The vertical axis of the graph represents the neutron flux density in units of n / cm ² · sec, and the horizontal axis represents the radial distance in units of cm. Fast represents the neutron flux density of neutrons having an energy of 1 MeV or higher, and Thermal represents the neutron flux density of thermal neutrons. As can be seen from the graph, when the mixture 17 is placed in the irradiation object storage container 40 and inserted into the irradiation hole 32, the mixture 17 is 1.1 × 10 ¹⁵ n / cm ² · sec with fast neutrons and thermal neutrons. It will receive 8.2 × 10 ¹⁴ n / cm ² · second neutron irradiation.

  When the mixture 17 is irradiated with neutrons, the transuranium element therein is fissioned and converted into fission products. Table 2 shows the composition of the transuranium element contained in the mixture 17.

It is assumed that the amount of transuranium element shown in Table 2 is uniformly present in the irradiated container 40 having a volume of 25 liters. At this time, when 1 MeV neutrons are simultaneously irradiated with a neutron flux density of 1.1 × 10 ¹⁵ n / cm ² · sec and thermal neutrons with a neutron density of 8.2 × 10 ¹⁴ n / cm ² · sec, Each transuranium element changes to the values shown in Table 3 by the fission and transmutation reactions by. In Table 3, “-” indicates 1 mg or less. In calculating Table 3, the influence of neutron absorption by fission products generated by fission of transuranium elements is not considered.

That is, in FIG. 1, when the irradiation time in the neutron irradiation process 23 is 6 months,
Since 99.3% of the transuranium element in the mixture 17 is burned, the fission product and transuranium element mixture 18 output from the neutron irradiation process 23 has the composition shown in Table 4. Therefore, when the mixture 18 is vitrified and stored, the radioactive toxicity of the vitrified body is reduced to the level of natural uranium ore in a storage period of about 700 to 800 years. By reducing the total weight of the transuranium elements contained in the mixture 18 to 1% by weight or less in this way, the spent nuclear fuel processed product can be reduced to the radioactive level of natural uranium within 1000 years. .

In the embodiment of the on-site type spent nuclear fuel processing method according to the present invention, the molten salt reactor neutron source 30 was used in the neutron irradiation process 23 of FIG. The molten Shiokaku fuel 35 for loading the molten salt reactor neutron source 30, is considered the application of molten salts of various combinations, in the present embodiment ⁷ _LiF-BeF 2 -ThF ₄ - by applying the ²³³ UF ₄ Yes. In a normal molten salt nuclear fuel, the molar ratio of thorium tetrafluoride to uranium tetrafluoride is about 98: 2, but in this embodiment, in order to achieve the high neutron flux of FIG. The molar ratio with uranium fluoride is about 90:10. By further increasing the molar ratio of uranium tetrafluoride, the size of the neutron flux can be further increased from the value shown in FIG.

  Moreover, it is desirable to apply uranium tetrafluoride produced with uranium having an isotope ratio of uranium 233 of 99% or more as the uranium tetrafluoride added to the molten salt nuclear fuel 35. For example, when uranium tetrafluoride, which is a mixture of many non-fissionable isotopes with a large mass number, such as uranium 234 and uranium 236, is used as the molten salt nuclear fuel 35, the high neutron flux causes the above six months. During the irradiation period, a large amount of transuranium element is generated in the molten salt nuclear fuel 35. Therefore, in order to efficiently reduce transuranium elements in the entire system including the neutron source, uranium tetrafluoride applied to the molten salt nuclear fuel 35 has a uranium 233 isotope ratio of 99% or more. desirable.

  In the embodiment of the on-site spent nuclear fuel processing method according to the present invention, only uranium is extracted by the uranium extraction process of FIG. 1, and plutonium is burned by the neutron irradiation process 23 together with other superuranium elements. . Since plutonium has an isotope with a large fission cross section, even if a large amount of transuranium element containing plutonium is inserted into the irradiation hole 32, the neutron flux density of the irradiation hole 32 does not greatly decrease. Therefore, neptunium, americium, curium, or the like can be efficiently burned by neutron irradiation together with plutonium.

  As a comparative example of the present invention, the neutron flux of the molten salt reactor neutron source 30 when plutonium does not exist, that is, when only the neptunium, americium, and curium are loaded into the irradiated object storage container 40 and inserted into the irradiation hole 32. The distribution is shown in FIG. Neptunium, americium, and curium all have a small fission cross section and a large absorption cross section, so that the neutron flux density is attenuated at the center of the irradiation hole 32 as shown in FIG. Therefore, if it is attempted to surely burn the transuranium element in the center, it is necessary to lengthen the irradiation time by the amount of attenuation of the neutron flux, and the transuranium element reduction efficiency decreases.

  In addition, plutonium is a powerful nuclear weapon material, and in the present invention, it is burned together with other transuranium elements without being separated as a single element, thereby providing a spent nuclear fuel processing method with excellent proliferation resistance. An effect is also obtained.

  As the lithium fluoride to be added to the molten salt nuclear fuel 35, it is desirable to use lithium fluoride produced with lithium having an isotope ratio of lithium 7 of 100%. A general molten salt nuclear reactor can be operated continuously for about 30 years. However, the molten salt nuclear reactor neutron source 30 continues to operate under a high neutron flux as shown in FIG. A refresh of structural materials and nuclear fuel is required. On the other hand, if the molten salt nuclear fuel 35 contains a small amount of lithium 6 even if it is a trace amount, a large amount of tritium is generated by the lithium 6 absorbing neutrons. Even if lithium fluoride based on highly concentrated lithium 7 (lithium having 99.995% is lithium 7 and 0.005% is lithium 6 isotope ratio) is used, when the molten salt reactor neutron source 30 is continuously operated, 240 Curie of tritium is generated every day, and the amount of tritium is 180,000 Curie in the operation period of 2 years.

  Natural lithium has an isotope ratio of 92.5% lithium 7 and 7.5% lithium 6. Highly concentrated lithium 7 is obtained by increasing the isotope ratio of lithium 7 by, for example, a lithium ion permselective membrane method. However, the concentration rate of 99.995% is a technical limit, and the price is very high. Expensive. Therefore, in this embodiment, lithium 7 having an isotope ratio of 100% is manufactured by the lithium 7 manufacturing method 200 shown in FIG.

In FIG. 7, the method for producing lithium 7 includes a neutron irradiation process 71 and a lithium extraction process 72. Reference numeral 70 denotes a boron aggregate, that is, α-boron, β-boron, boron crystal, etc., and has the same size and volume as the irradiated container 40 shown in FIG. When the boron assembly 70 is irradiated with neutrons in the neutron irradiation process 71, lithium 7 is generated in the boron assembly 70 by a ¹⁰ B (n, α) ⁷ Li reaction. The boron assembly 72 in which lithium 7 after neutron irradiation is generated is chemically treated by a lithium extraction process 73 (for example, liquid phase separation treatment with hydroxide), so that the lithium 7 assembly 75 and boron waste 74 is separated. 100% purity lithium 7 is produced from the extracted lithium 7 aggregate 75.

In the neutron irradiation process 71, neutron irradiation is performed using the molten salt reactor neutron source 80 shown in FIG. The molten salt nuclear reactor neutron source 80 has exactly the same dimensions and structure as the molten salt nuclear reactor neutron source 30 shown in FIG. 3, but the composition of the molten salt nuclear fuel 81 is different from that of the molten salt nuclear fuel 35. Melting Shiokaku fuel 35 100% pure ⁷ _LiF-BeF 2 by lithium 7 of -ThF ₄ - ²³³ UF ₄ has been applied, but ⁷ by high concentration of lithium 7 in molten Shiokaku fuel _{75 LiF-BeF 2 -ThF 4 -} 233 UF 4 Apply. Therefore, also in the molten salt reactor neutron source 80, the neutron flux density is substantially equal to the value shown in FIG.

When the boron aggregate 70 is composed of natural α-boron, the density of the natural α-boron is 2.54 g / cm ³ , so the weight of the boron aggregate 70 is 63.5 kg. Since the boron 10 isotope ratio is 19.9%, there are approximately 1270 moles of boron 10 nuclei in the boron assembly 70. The thermal neutron absorption cross section of the boron 10 nucleus is 6000 bahn, and when the irradiation is performed for about 1 hour in the neutron irradiation process 71, about half of the boron 10 nucleus is transmuted to the lithium 7 nucleus. Therefore, by extracting lithium from the boron assembly 72 after irradiation, 650 mol of lithium 7, that is, 4.5 kg of lithium 7 is obtained with a purity of 100%.

In molten salt reactor neutron source 30 shown in FIG. 3, the molten Shiokaku ⁷ _LiF-BeF 2 to about 300kg as fuel 35 -ThF ₄ - ²³³ UF ₄ for loading nuclear fuel, it requires lithium 7 to about 60kg per group And That is, in FIG. 7, lithium 7 production by one hour irradiation is repeated about 14 times, so that lithium 7 having a purity of 100% for one molten salt reactor neutron source can be secured. In the present embodiment, the means for producing lithium 7 using a molten salt reactor neutron source has been described. However, even if an accelerator neutral source or an acceleration-driven subcritical neutron source is applied as the neutron source, the equivalent is achieved. An effect can be obtained.

DESCRIPTION OF SYMBOLS 10 Spent fuel assembly 20 Shearing process 21 Melting process 22 Uranium extraction process 23 Neutron irradiation process 30 Molten salt reactor accelerator neutron source 31 Neutron reflector 32 Irradiation hole 35 Molten salt nuclear fuel 40 Irradiation object storage container

Claims (6)

A shearing process for shearing the fuel rods of the spent nuclear fuel assembly;
A dissolution process in which the spent nuclear fuel that has been sheared is chemically dissolved and separated into an insoluble residue mainly composed of a platinum group, and a solution containing fission products, uranium, and transuranium elements;
A uranium extraction process in which only uranium is separated from the solution containing the fission product, uranium, and transuranium element, and separated into a uranium solution mainly composed of uranium, and a mixture of the fission product and transuranium element,
Neutron irradiation process for irradiating the mixture of fission products and transuranium elements with neutrons to reduce the amount of transuranium elements in the mixture;
An onsite-type spent nuclear fuel processing method.   2. The on-site spent nuclear fuel processing method according to claim 1, wherein in the neutron irradiation process, the total weight of the transuranium element contained in the mixture is reduced to 1% by weight or less. The neutron irradiation process irradiates neutrons at a neutron flux density of 1 × 10 ¹⁴ n / cm ² · second to 1 × 10 ¹⁶ n / cm ² · second, according to claim 1 or 2. On-site spent nuclear fuel processing method.   In the neutron irradiation process, a molten salt reactor neutron source is applied as a neutron source, and the liquid nuclear fuel loaded in the molten salt reactor includes uranium tetrafluoride and fluorine containing uranium having a uranium 233 isotope ratio of 99% or more. The on-site type spent nuclear fuel processing method according to any one of claims 1 to 3, wherein a mixture with a chemical salt is applied.   The on-site spent nuclear fuel processing method according to claim 4, wherein a lithium compound containing lithium having a lithium 7 isotope ratio of 100% is added to the fluoride salt. 6. The on-site spent nuclear fuel processing method according to claim 5, wherein the lithium 7 of the lithium compound is lithium 7 obtained by irradiating boron with neutrons.