Nuclear transmutation method for minor actinoid

Abstract

To provide a nuclear transmutation method capable of improving a nuclear transmutation amount of minor actinoids (MA) using a fast reactor.SOLUTION: A nuclear transmutation method for MA includes: a step of performing processing of separating MA from high-level radioactive liquid waste being waste liquid obtained by separating U and Pu from a liquid solution of spent nuclear fuel and obtaining a first radioactive composition containing MA; a step of adding U to the first radioactive composition to obtain a second radioactive composition; a step of obtaining an oxide solid solution of MA and U from the second radioactive composition; a storage step of allowing high heat-generating MA contained in the oxide solid solution to decay; a step of producing fuel containing core fuel, radial blanket fuel, upper axial blanket fuel, and lower axial blanket fuel by using the oxide solid solution after storage; and a step of loading the produced fuel into the core of a fast reactor and performing nuclear transmutation of MA contained in the fuel.SELECTED DRAWING: Figure 1

Description

[Technical field]

[0001] The present invention relates to a method for transmuting minor actinides. [Background technology]

[0002] In the reprocessing of spent nuclear fuel generated from light water reactors, U (uranium) and Pu (plutonium) are recovered from the solution of spent nuclear fuel. Highly Active Liquid Waste (HALW) that remains after the recovery of U and Pu contains fission products (FP) as well as minor actinides (MA) such as Np (neptunium), Am (americium), and Cm (curium). HALW is planned to be vitrified through an enrichment process and disposed of in a geological repository.

[0003] MA includes nuclides with long half-lives on the order of 10,000 years, which increases the load of HALW disposal. Therefore, nuclear transmutation technology is being considered to separate MA from HALW and convert the separated MA into different elements using fast reactors or accelerators (ADS). Patent documents 1 and 2 disclose a core in which a target, which is a fuel for MA nuclear transmutation, is accommodated between a central region that accommodates the core fuel and a blanket region that accommodates the blanket fuel. Patent documents 1 and 2 also disclose that the target contains hydride fuel in order to increase the annihilation rate of MA. Patent Document 3 discloses a core in which fuel containing transuranium elements (TRUs) such as MA is arranged non-homogeneously in the central region, and also proposes mixing MA elements into an inner blanket region arranged in the central region of the core in order to improve the core characteristics.

[0004] On the other hand, MA recovered from HALW generally contains highly exothermic nuclides ( 244 Cm, etc. Therefore, Patent Document 4 discloses that MA recovered from HALW is solidified and stored for a long period of time to attenuate highly exothermic nuclides. [Prior art documents] [Patent documents]

[0005] [Patent Document 1] JP 2020-134296 A [Patent Document 2] JP 2020-153758 A [Patent Document 3] Japanese Patent Application Publication No. 5-180971 [Patent Document 4] JP 2020-125989 A Summary of the Invention [Problem to be solved by the invention]

[0006] However, with conventional nuclear transmutation technologies, the annual amount of MA transmutation is at most 0.006-0.02 t-MA / (GW·year). Assuming the reprocessing amount at the Rokkasho Reprocessing Plant (800 tHM / year), the annual amount of MA generated is approximately 1.0 t-MA / year, meaning that the transmutation of MA using fast reactors cannot keep up with the amount of MA generated. The present disclosure has been made to solve the above-mentioned problems, and aims to provide a method for transmuting MA that can improve the yield of MA transmutation in a fast reactor. [Means for solving the problem]

[0007] In order to solve the above problems, the method for transmuting MA according to the present disclosure includes the steps of: performing a process for separating MA from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, to obtain a first radioactive composition containing MA; adding U to the first radioactive composition to obtain a second radioactive composition; obtaining an oxide solid solution of MA and U from the second radioactive composition; storing the oxide solid solution to attenuate the highly exothermic MA contained in the oxide solid solution; using the oxide solid solution after storage, producing fuel including core fuel to be placed in a central region of a fast reactor core, radial blanket fuel to be placed in a radial blanket region radially outside the central region, upper axial blanket fuel to be placed in an upper axial blanket region axially above the central region, and lower axial blanket fuel to be placed in a lower axial blanket region axially below the central region; and loading the produced fuel into the core of a fast reactor and transmuting the MA contained in the fuel. Effect of the Invention

[0008] The method for transmuting MA according to the present disclosure can improve the transmutation yield of MA in a fast reactor. [Brief description of the drawings]

[0009] [Figure 1] FIG. 2 is a flow diagram illustrating a nuclear transmutation method for MA according to the first embodiment. [Diagram 2] FIG. 1 is a vertical cross-sectional view showing the configuration of a fast reactor. [Diagram 3] FIG. 2 is a radial cross-sectional view showing the configuration of a core in the first embodiment. [Figure 4] FIG. 4 is a schematic diagram of the cross section of the core taken along line III-III of FIG. [Diagram 5] FIG. 2 is a perspective view showing the configuration of a core fuel assembly. [Figure 6] FIG. 2 is a perspective view showing the configuration of a core fuel element. [Figure 7] FIG. 1 is an axial cross-sectional view showing an example of an arrangement of blanket fuel containing MA and blanket fuel not containing MA. [Figure 8]FIG. 1 is an axial cross-sectional view showing an example of an arrangement of blanket fuel containing MA and blanket fuel not containing MA. [Figure 9] FIG. 11 is a radial cross-sectional view showing an example of a configuration of a core in a second embodiment. [Figure 10] FIG. 10 is a schematic diagram of the XX cross section of the core in FIG. [Figure 11] FIG. 11 is a radial cross-sectional view showing another example of the configuration of the core in the second embodiment. [Figure 12] FIG. 12 is a schematic diagram of the cross section of the core taken along line XII-XII of FIG. 11. DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

[0010] In this specification, MA (minor actinides) refers to the transuranium elements belonging to the actinides excluding Pu, and mainly refers to Np, Am, and Cm. In this specification, blanket fuel is a general term for radial blanket fuel, upper axial blanket fuel, lower axial blanket fuel and inner blanket fuel. The symbol "~" indicating a range of numerical values ​​means that the numerical values ​​before and after it are included as the lower limit and upper limit. For ease of explanation, the scale of each part in the drawings may differ from the actual scale.

[0011] First Embodiment As shown in FIG. 1, the method for transmuting MA according to the first embodiment of the present disclosure includes: Step S1-1: performing a process (MA separation process) for separating MA from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, to obtain a first radioactive composition containing MA; Step S1-2 of adding U to the first radioactive composition to obtain a second radioactive composition; Step S1-3 of obtaining a solid solution of oxides of MA and U from the second radioactive composition; Step S1-4: storing the oxide solid solution to allow the highly exothermic MA contained in the oxide solid solution to decay; Step S1-5 of using the stored oxide solid solution to produce fuel including core fuel to be disposed in a central region of a fast reactor core, radial blanket fuel to be disposed in a radial blanket region radially outside the central region, upper axial blanket fuel to be disposed in an upper axial blanket region axially above the central region, and lower axial blanket fuel to be disposed in a lower axial blanket region axially below the central region; and step S1-6 of loading the produced fuel into the core of a fast reactor and transmuting the MA contained in the fuel. The waste liquid (containing FP) from which MA has been separated in step S1-1 is, for example, vitrified, stored, and then disposed of in a geological repository.

[0012] (HALW) HALW is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel. HALW contains FP, MA, lanthanides (hereinafter also referred to as Ln), etc., but does not contain U or Pu. HALW typically contains at least Np, Am, and Cm as MA. Ln is a general term for elements with atomic numbers from 57 to 71. HALW typically contains at least La and Ce as Ln. The amounts and ratios of MA and Ln contained in spent nuclear fuel and HALW vary depending on the type of spent nuclear fuel (pressurized water reactor (PWR) fuel, boiling water reactor (BWR) fuel, MOX fuel, burnup, cooling period, etc.). HALW is generated, for example, by the reprocessing of spent nuclear fuel. In the reprocessing of spent nuclear fuel, the spent nuclear fuel removed from the reactor is processed, for example, by shearing, dissolving, clarifying, weighing and adjusting, extraction and separation, refining, and denitrification to produce a powder product containing U and Pu. In the extraction and separation, U and Pu are separated from the spent nuclear fuel solution. Extraction and separation is performed, for example, by the PUREX (Plutonium Uranium Redox Extraction) method. In the PUREX method, a nitric acid solution of spent nuclear fuel is contacted and mixed with tributyl phosphate (TBP) and an organic solvent such as dodecane. As a result, U and Pu in the nitric acid solution form complexes with TBP and move to the organic solvent side. On the other hand, FP, MA, and Ln remain in the nitric acid solution (waste liquid) side.

[0013] (Step S1-1: MA separation) The MA separation treatment can be carried out by a known method. During the MA separation process, Ln may be entrained with MA. Since Ln has similar chemical properties to MA, Ln is likely to be entrained when MA is separated from HALW. After the MA separation process, an MA purification process for separating MA and Ln may be performed.

[0014] Examples of MA separation processes include a process involving extracting MA from HALW by solvent extraction (hereinafter also referred to as MA extraction process), or a process involving adsorbing MA from HALW onto an adsorbent and eluting the MA adsorbed onto the adsorbent (hereinafter also referred to as MA adsorption-elution process). When the MA separation treatment includes an extraction treatment, the treatment may further include, after the extraction treatment, a treatment of stripping MA from the extract obtained by the extraction treatment (hereinafter also referred to as MA stripping treatment).

[0015] In the MA extraction process, for example, the HALW is brought into contact with an organic solvent solution containing an extractant (extractant solution), and the MA migrates to the extractant solution.

[0016] The extractant may be, for example, a complexing agent that forms a complex with MA. The complexing agent may be one that forms a complex with Ln. Specific examples of the complexing agent include n-octyl(phenyl)-N,N'-diisobutylcarbamoylmethylphosphine oxide-tributyl phosphate mixture (CMPO-TBP mixture), diisodecyl phosphoric acid, 6,6'-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-1,2,4-benzotriazin-3-yl)-2,2'-bipyridine (BTBP), N,N'-dibutyl-N,N'-dimethyltetradecylmalonamide (DMDBTDMA), and N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA). The extractant may be used alone or in combination of two or more.

[0017] The organic solvent can be appropriately selected depending on the extractant used. It is desirable that the organic solvent is reusable, inexpensive, and resistant to radiation degradation. A specific example of the organic solvent is n-dodecane. The organic solvent may be used alone or in combination of two or more kinds.

[0018] The extract obtained by the MA extraction treatment may be directly subjected to step S1-2 as the first radioactive composition, or may be further subjected to MA stripping treatment.

[0019] In the MA stripping process, for example, the extract is contacted with an aqueous solution containing nitric acid (stripping agent). This causes MA and Ln in the extract to migrate to the aqueous solution. Meanwhile, the organic solvent in the extract does not migrate to the aqueous solution but remains in the extract, making it reusable. The stripping solution obtained in the MA stripping treatment is typically directly supplied to step S1-2 as the first radioactive composition.

[0020] The adsorbent used in the MA adsorption-elution treatment may be any adsorbent capable of adsorbing MA. The method for adsorbing MA in HALW onto an adsorbent includes contacting the HALW with an adsorbent, such as a column method or a batch method. The MA adsorbed on the adsorbent can be eluted by contacting the adsorbent with an eluent. The eluate obtained by the MA adsorption-elution process, i.e., the eluate after contact with the adsorbent, contains MA. The eluate may contain Ln. Typically, the eluate obtained by the MA adsorption-elution process is directly subjected to step S1-2 as the first radioactive composition.

[0021] (Step S1-2: U addition) In step S1-2, U is added to the first radioactive composition in order to obtain an oxide solid solution of MA and U in step S1-3 described below. This results in a second radioactive composition containing MA and U. U is preferably added so that the amount of MA relative to the amount of U is 10 mass % or less.

[0022] (Step S1-3: Solidification process) In step S1-3, a treatment for obtaining a solid solution of oxides of MA and U from the second radioactive composition (hereinafter, also referred to as a solidification treatment) is carried out. Since the second radioactive composition contains MA and U, it is possible to obtain an oxide solid solution having a stable fluorite crystal structure. It is also known that UO2 fuel has a fluorite crystal structure.

[0023] The solidification treatment may be any treatment capable of obtaining an oxide solid solution, and examples of such treatment include reduction / sintering treatment and hydrothermal treatment.

[0024] When the second radioactive composition contains an organic solvent (for example, when the first radioactive composition obtained in step S1-1 is an extract obtained by an extraction process), the solidification process is preferably a reduction / sintering process. When the reduction / sintering process is performed on the second radioactive composition containing an organic solvent, the extractant contained in the organic solvent is removed and MA and U are oxidized. As a result, an oxide solid solution of MA and U is obtained. Examples of the reduction / sintering treatment include distillation, pyrolysis, denitration, and incineration. Distillation can be carried out using a known evaporation method such as a batch method or a continuous method (plate tower or packed tower).

[0025] When the second radioactive composition contains water (for example, when the first radioactive composition obtained in step S1-1 is a stripping solution obtained by stripping treatment or an elution solution obtained by adsorption-elution treatment), the solidification treatment is preferably hydrothermal treatment. When the second radioactive composition containing water is subjected to hydrothermal treatment, MA and U contained in the second radioactive composition are oxidized, and an oxide solid solution of MA and U is precipitated in the water. The precipitated oxide solid solution is separated from the liquid medium (water) by solid-liquid separation.

[0026] If necessary, before storing the obtained oxide solid solution, the oxide solid solution may be subjected to a stabilization treatment for removing a part or all of the carbon, hydrogen, oxygen and nitrogen components. The oxide solid solution may contain organic matter such as the extractant or the radiolysis product of the extractant. If the oxide solid solution contains organic matter, gas may be generated during storage, impairing the containment function of the radioactive material and causing problems. The occurrence of such problems can be suppressed by performing a stabilization treatment on the oxide solid solution to remove some or all of the carbon, hydrogen, oxygen, and nitrogen components. The stabilization treatment may be, for example, calcination or sintering.

[0027] (Step S1-4: Storage) By storing the oxide solid solution, the highly exothermic MA contained in the oxide solid solution is attenuated. The fuel assemblies loaded into the reactor core are subject to limits on the amount of heat generated per assembly (for example, 3 kW or less per assembly). Immediately after being removed from the reactor, the MA in spent fuel contains highly heat-generating nuclides. When using such MA, due to the above-mentioned heat generation limit, it is necessary to either reduce the MA content or to carry out a process to separate the highly heat-generating nuclides (MA refining) prior to fuel production. By attenuating the highly heat-generating MA through storage, it is possible to increase the MA content in the fuel without carrying out a process to separate the highly heat-generating MA. Highly exothermic MAs include, for example, Cm-242 (half-life of approximately 163 days) and Cm-244 (half-life of approximately 18 years).

[0028] The oxide solid solution can be stored using any method known as a dry storage method for radioactive waste, such as storing the oxide solid solution in multiple canisters, housing these multiple canisters in a cask, and storing them in a storage facility. The period for storing the oxide solid solution can be appropriately set depending on the nuclide to be attenuated.

[0029] (Step S1-5: Fuel production) In step S1-5, at least the core fuel, radial blanket fuel, upper axial blanket fuel, and lower axial blanket fuel are produced as fuel. Inner blanket fuel to be placed in the central region together with the core fuel may also be produced. Both of these fuels produced in step S1-5 contain MA and U due to their oxide solid solutions.

[0030] <Core fuel> The core fuel is typically MOX fuel, which contains a mixed oxide of U and Pu. The composition of the core fuel may be, for example, as follows: U content: 65~75% by mass, Pu content: 20~30% by mass, MA content: 1~3% by mass Ln content: 0~0.3% by mass. Here, the total of U, Pu, MA and Ln does not exceed 100 mass%. Ln is likely to be entrained in MA during production, and is assumed to be entrained in an amount up to about 0.1 times that of MA. The content of each element is the ratio to the total mass of the core fuel placed in the central region of the core. The core fuel may further contain other components as necessary. Typically, there are two types of core fuel: one located in the inner region of the central region (hereafter also referred to as the inner core fuel) and the other located in the outer region (hereafter also referred to as the outer core fuel). The outer core fuel usually has a higher content of fissile material (e.g. Pu) than the inner core fuel.

[0031] As a method for producing core fuel using oxide solid solutions, for example, there is a method in which the oxide solid solutions are redissolved in nitric acid or the like to obtain a solution containing MA and U, and then fuel for fast reactors is produced from the solution by denitration, conversion, and sintering, and then formed into any shape such as pellets as necessary. As a method for dissolving oxide solid solutions, for example, there is a method in which an aqueous solution of nitric acid is added to the oxide solid solutions. During the production of core fuel, there is no need to separate MA and U contained in the oxide solid solution.

[0032] <Blanket fuel> Blanket fuel is fuel that contains a lower content of fissile material (e.g. Pu) than the core fuel. Blanket fuel typically contains U. When the U in the blanket fuel absorbs neutrons emitted from the core fuel, the U-238 in the U is converted to Pu-239, which becomes new fuel.

[0033] The composition of the radial blanket fuel may be, for example, as follows: U content: 70~90% by mass, MA content: 10~30% by mass, Ln content: 0~3% by mass. Here, the total of U, MA, and Ln does not exceed 100 mass%. Ln is easily entrained during the production of MA, and entrainment of Ln with respect to MA is expected to be up to about 0.1 times. The ratio of each component is the ratio to the total mass of the radial blanket fuel placed in the radial blanket region. The composition of each of the multiple radial blanket fuels in the radial blanket region may be the same or different. The radial blanket fuel may further contain other components as required.

[0034] The radial blanket fuel preferably does not contain moderators such as H, Be, or C. In the past, adding a moderator to the target has been considered, but when the MA content is around 30 mass%, adding a moderator does not significantly change the transmutation efficiency (annihilation rate) of MA. In addition, if the radial blanket fuel contains a moderator, it has an effect such as an increase in the amount of hydrogen gas generated during reprocessing of spent blanket fuel after MA transmutation. By not including a moderator in the radial blanket fuel, the effect on the reprocessing side after MA transmutation can be mitigated.

[0035] The composition of the upper axial blanket fuel may be, for example, as follows: U content: 70~90% by mass, MA content: 10~30% by mass, Ln content: 0~3% by mass. Here, the sum of U and MA does not exceed 100 mass%. Ln is easily entrained during the production of MA, and entrainment of up to about 0.1 times that of MA is expected. The proportion of each component is the proportion to the total mass of the upper axial blanket fuel placed in the upper axial blanket region. The composition of each of the multiple upper axial blanket fuels in the upper axial blanket region may be the same or different. The upper axial blanket fuel may further comprise other components as required. The upper axial blanket fuel, like the radial blanket fuel, preferably does not contain moderators such as H, Be, or C.

[0036] The composition of the lower axial blanket fuel may be, for example, as follows: U content: 70~90% by mass, MA content: 10~30% by mass, Ln content: 0~3% by mass. Here, the sum of U and MA does not exceed 100 mass%. Ln is easily entrained during the production of MA, and entrainment of up to about 0.1 times that of MA is expected. The proportion of each component is the proportion to the total mass of the lower axial blanket fuel placed in the lower axial blanket region. The composition of each of the multiple lower axial blanket fuels in the lower axial blanket region may be the same or different. The lower axial blanket fuel may further comprise other components as required. The lower axial blanket fuel, like the radial blanket fuel, preferably does not contain moderators such as H, Be, or C.

[0037] For example, the oxide solid solution is dissolved again in nitric acid to obtain a solution containing MA and U, and then fuel for fast reactors is produced by denitration, conversion, and sintering, and then molded into any shape such as pellets as necessary. For example, the oxide solid solution is dissolved by adding an aqueous solution of nitric acid to the oxide solid solution. In the production of blanket fuel, it is not necessary to separate MA and U contained in the oxide solid solution again.

[0038] (Step S1-6: MA transmutation) In step S1-5, for example, core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14, and neutron shield 15 are arranged in a predetermined region of the fast reactor core to form core 1 (homogeneous core), and the fast reactor is operated for a given period of time. This causes nuclear transmutation of MA contained in each fuel.

[0039] <Fast reactor core> An example of the configuration of a fast reactor will be described with reference to Fig. 2. As shown in the figure, a fast reactor 100 of this example includes a core 1, a reactor vessel 2, a guard vessel 3, a coolant inlet pipe 4, a coolant outlet pipe 5, an upper core mechanism 7, and a fixed plug 8.

[0040] The reactor core 1 is a heat generating source containing fissile material. The detailed configuration of the reactor core 1 will be described later. The reactor vessel 2 is a vessel that contains the reactor core 1. The reactor vessel 2 is cylindrical with a bottom surface. The reactor core 1 is fixed to the lower part of the reactor vessel 2 via reactor internal structures 9. The opening at the top of the reactor vessel 2 is covered by a fixing plug 8. The fixing plug 8 is supported by a structure of the reactor building (reactor vessel pedestal 6).

[0041] The guard vessel 3 covers the reactor vessel 2 from the outside. In other words, the reactor vessel 2 and the guard vessel 3 form a double-wall structure. As a result, even if coolant leaks from the reactor vessel 2, the coolant is retained by the guard vessel 3, and leakage to the outside is suppressed.

[0042] The coolant inlet pipe 4 introduces the coolant (primary coolant) introduced from the outside into the reactor vessel 2. An example of the coolant is liquid metallic sodium. An end of the coolant inlet pipe 4 is located below the core 1 in the reactor vessel 2. As a result, the reactor vessel 2 is filled with the coolant. The coolant outlet piping 5 discharges the coolant in the reactor vessel 2 to the outside. An end of the coolant outlet piping 5 is located above the core 1 in the reactor vessel 2.

[0043] The upper core mechanism 7 has a control rod drive mechanism 71, a rotating plug 72, and a rotating plug drive device 73. The control rod drive mechanism 71 is a device for inserting and withdrawing a control rod for controlling the progress of a nuclear fission reaction in the core 1, which will be described later. The control rod drive mechanism 71 moves the control rod forward and backward in the vertical direction. The rotating plug 72 is a device for positioning equipment for replacing nuclear fuel (for example, the core fuel assemblies 30, which will be described later) in the core 1. The rotating plug 72 is driven by the rotating plug drive device 73.

[0044] Next, the configuration of the core 1 will be described with reference to FIGS. As shown in FIGS. 3 and 4, the core 1 includes core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14, and a neutron shield 15.

[0045] The core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14, and neutron shield 15 are each contained in a hexagonal cross-sectional casing, treated as an assembly, and disposed in a container provided in each region. Each of the hexagons corresponding to the core fuel 11, radial blanket fuel 12, and neutron shield 15 shown in Fig. 3 indicates one assembly. In the following description, disposing in a container provided in a region is abbreviated to "disposing in a region."

[0046] The core fuel 11 is arranged in a central region of the core 1 . The central region is located in the center of the reactor core 1. The central region has a hexagonal prism shape. The central region is composed of an inner region and an outer region located radially outward of the inner region. The inner region is shaped like a hexagonal column coaxial with the central region. A plurality of control rods 16 can be inserted into a portion of the inner region. Nuclear fission of the core fuel 11 is controlled by controlling the amount of control rods 16 inserted between the core fuel 11. The outer region surrounds the inner region without any gaps. The outer region is shaped like a hexagonal ring in a plan view. The core 1 according to this embodiment is a homogeneous core in which only the core fuel 11 and the control rods 16 are arranged in the central region.

[0047] The upper axial blanket fuel 13 is arranged in an upper axial blanket region on the axially upper side of the central region (upper side in FIG. 4). An upper axial blanket region covers the axially upper side of the central region.

[0048] The lower axial blanket fuel 14 is disposed in a lower axial blanket region axially below the central region (lower side in FIG. 4). A lower axial blanket region covers the axially underside of the central region.

[0049] The radial blanket fuel 12 is disposed in a radial blanket region radially outward of the central region. The radial blanket region completely surrounds the radial outside of the central region, the radial outside of the upper axial blanket region, and the radial outside of the lower axial blanket region. The radial blanket region has a hexagonal ring shape in a plan view. In addition, the neutron source may be disposed at a position adjacent to the central region of the radial blanket region.

[0050] The neutron shielding body 15 is arranged in each shielding body region radially outside the radial blanket region, axially above the upper axial blanket region, and axially below the lower axial blanket region. The radially outer shielding region (hereinafter also referred to as the radial shielding region) surrounds the radially outer side of the radial blanket region over the entire circumference. The radial shielding region has a hexagonal ring shape in a plan view. The axially upper shield region of the upper shaft blanket region (hereinafter also referred to as the upper shaft shield region) covers the axially upper side of the upper shaft blanket region and the axially upper side of the radial blanket region. The shield region axially below the lower shaft blanket region (hereinafter also referred to as the lower shaft shield region) covers the axially below the lower shaft blanket region and the axially below the radial blanket region.

[0051] During operation of the reactor core 1, the core fuel 11 generates heat by fissioning the fission material using neutrons generated by the core fuel 11 itself or from a neutron source located at any location as a trigger. The insertion amount of the control rods 16 is adjusted to control the progress of this fission reaction. In blanket fuels such as the radial blanket fuel 12, the upper axial blanket fuel 13, and the lower axial blanket fuel 14, the reaction proceeds with a reduced amount of fast fission reactions compared to the core fuel 11. Also, in the blanket fuels, the amount of fission material produced by the fission reactions is greater than that of the core fuel 11. The neutron shield 15 blocks neutrons and prevents the neutrons from leaking outside the reactor core 1 .

[0052] Next, the configuration of the assembly will be described in detail with reference to FIG. 5 is an example of an assembly (hereinafter, also referred to as a core fuel assembly) arranged in the central region in a plan view. As shown in the figure, the core fuel assembly 30 has a trumpet tube 31, an entrance nozzle 32, a handling head 33, and a plurality of core fuel elements 40.

[0053] The trumpet tube 31 has a cylindrical shape centered on an axis Ac extending in the vertical direction. The trumpet tube 31 has a hexagonal cross-sectional shape when viewed from the direction of the axis Ac. The opening at the bottom of the trumpet tube 31 is closed by the entrance nozzle 32. A flow path (not shown) is formed inside the entrance nozzle 32 for guiding the coolant into the inside of the trumpet tube 31. The entrance nozzle 32 has an opening that connects this flow path with the outside. A handling head 33 is attached to the opening at the top of the trumpet tube 31. The handling head 33 is a part that is held by a device when the core fuel assembly 30 is transported. Inside the trumpet tube 31 and directly above the entrance nozzle 32, a plurality of core fuel elements 40 are arranged at intervals in a direction perpendicular to the axis Ac. A coolant introduced from the opening of the entrance nozzle 32 flows in the space around and above the core fuel elements 40.

[0054] The configuration of the core fuel element 40 will be described with reference to Fig. 7. As shown in the figure, the core fuel element 40 has a cylindrical cladding tube 41 extending in the vertical direction, a plenum spring 43 housed inside the cladding tube 41, core fuel 11, upper axial blanket fuel 13, and lower axial blanket fuel 14, and an upper end plug 48 and a lower end plug 49 provided at both ends of the cladding tube 41.

[0055] The core fuel 11, the upper axial blanket fuel 13, and the lower axial blanket fuel 14 are each formed into a cylindrical pellet shape. A plurality of lower axial blanket fuels 14, a plurality of core fuels 11, and a plurality of upper axial blanket fuels 13 are filled in this order, stacked from the bottom of the cladding tube 41. The uppermost upper axial blanket fuel 13 is pressed downward by a plenum spring 43. The plurality of core fuels 11 are located in the central region when the core fuel assemblies 30 are disposed in the central region. The plurality of upper axial blanket fuels 13 are located in the upper axial blanket region when the core fuel assemblies 30 are disposed in the central region. The plurality of lower axial blanket fuels 14 are located in the lower axial blanket region when the core fuel assemblies 30 are disposed in the central region.

[0056] The assembly arranged in the radial blanket region in a planar view (hereinafter also referred to as the radial blanket fuel assembly) is similar to the core fuel assembly 30 except that radial blanket fuel 12 is contained instead of the core fuel 11, upper axial blanket fuel 13, and lower axial blanket fuel 14.

[0057] The core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13 and lower axial blanket fuel 14 were each produced in step S1-5 and contain MA. The MA content of the core fuel is generally set to a maximum of about 5 mass%, and the MA content of the blanket fuel is generally set to a maximum of about 30 mass%. This is because as the MA content increases, the neutron emission rate or heat generation density increases, increasing the need for measures such as neutron shielding or cooling, and making manufacturing more difficult. By incorporating MA into the core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, and lower axial blanket fuel 14, the amount of MA loaded can be increased without increasing the MA content. Some of the multiple core fuels 11 arranged in the central region may not contain MA, as may the radial blanket fuels 12, the upper axial blanket fuels 13, and the lower axial blanket fuels 14.

[0058] In another preferred embodiment, as shown in Figure 7, the radial blanket fuel 12 arranged in a portion L2 adjacent to the upper axial blanket fuel 13 or the lower axial blanket fuel 14, i.e., in the portion in which the upper axial blanket fuel 13 or the lower axial blanket fuel 14 is present in the radial direction, does not contain MA. The radial blanket fuel 12 arranged in the portion L2 tends to have a low MA transmutation rate because there are other blanket fuels (upper axial blanket fuel 13, lower axial blanket fuel 14, radial blanket fuel 12 adjacent to the central region) between them and the core fuel 11. By distributing MA unevenly in the portions other than the portion L2, the MA transmutation efficiency can be improved.

[0059] In another preferred embodiment, as shown in FIG. 8, among the radial blanket fuels 12, the radial blanket fuels 12 arranged in a portion L3 adjacent to the upper axial blanket fuels 13 or the lower axial blanket fuels 14 and more than 20 cm from the core fuels 11 do not contain MA. Even in a portion adjacent to the upper axial blanket fuel 13 or the lower axial blanket fuel 14, the MA can be transmuted if it is located within 20 cm (particularly within 10 cm) of the core fuel 11. In the embodiment shown in Fig. 8, the amount of transmutation of MA can be increased more than in the embodiment shown in Fig. 7.

[0060] The neutron shield 15 may be any known neutron shield, including, for example, an iron or hydrogen material.

[0061] After step S1-6, step S1-7 may be performed in which the operation of the fast reactor is stopped and the fuel after MA transmutation (core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, and lower axial blanket fuel 14) is removed from the core 1. At this time, the neutron shield 15 may be removed together with the fuel.

[0062] After step S1-7, step S1-8 may be performed in which the fuel (spent fuel) after the MA transmutation is reprocessed. In blanket fuel reprocessing, MA and Pu can be separated by known methods such as solvent extraction.

[0063] (Action and effect) In the method configured as described above, U is added to the first radioactive composition obtained by MA separation, and the composition is stored as an oxide solid solution. This oxide solid solution is then used to produce fuels containing MA (core fuel, radial blanket fuel, upper axial blanket fuel 13, and lower axial blanket fuel 14), which are then loaded into the core of a fast reactor to perform MA transmutation. This makes it possible to increase the amount of MA loaded into the core of the fast reactor and improve the amount of MA transmutation compared to conventional methods.

[0064] Second Embodiment The method for transmuting MA according to the second embodiment of the present disclosure includes the steps of: Step S2-1: performing a process (MA separation process) for separating MA from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, to obtain a first radioactive composition containing MA; Step S2-2 of adding U to the first radioactive composition to obtain a second radioactive composition; Step S2-3 of obtaining a solid solution of oxides of MA and U from the second radioactive composition; Step S2-4: storing the oxide solid solution to allow the highly exothermic MA contained in the oxide solid solution to decay; Step S2-5 of using the stored oxide solid solution to produce fuel including core fuel to be disposed in a central region of a fast reactor core, inner blanket fuel to be disposed in the central region of the fast reactor core, radial blanket fuel to be disposed in a radial blanket region radially outside the central region, upper axial blanket fuel to be disposed in an upper axial blanket region axially above the central region, and lower axial blanket fuel to be disposed in a lower axial blanket region axially below the central region; and step S2-6 of loading the produced fuel into the core of a fast reactor and transmuting the MA contained in the fuel. The waste liquid (containing FP) from which MA has been separated in step S2-1 is, for example, vitrified, stored, and then disposed of in a geological repository.

[0065] (Step S2-1: MA separation) Step S2-1 is similar to step S1-1.

[0066] (Step S2-2: U addition) Step S2-2 is similar to step S1-2.

[0067] (Step S2-3: Solidification process) Step S2-3 is similar to step S1-3.

[0068] (Step S2-4: Storage) Step S2-4 is similar to step S1-4.

[0069] (Step S2-5: Fuel production) Step S2-5 is similar to step S1-5, except that, as fuel, in addition to the core fuel, radial blanket fuel, upper axial blanket fuel, and lower axial blanket fuel, inner blanket fuel is produced.

[0070] The composition of the inner blanket fuel may be, for example, as follows: U content: 70~90% by mass, MA content: 10~30% by mass, Ln content: 0~3% by mass. Here, the total of U, MA, and Ln does not exceed 100 mass%. Ln is likely to be entrained during the production of MA, and entrainment of Ln with respect to MA is expected to be up to about 0.1 times. The ratio of each component is the ratio to the total mass of the inner blanket fuel 17 placed in the central region. The composition of each of the multiple inner blanket fuels 17 in the central region may be the same or different. The inner blanket fuel 17 may further include other components as desired. Like the radial blanket fuel 12, the inner blanket fuel 17 preferably does not contain moderators such as H, Be, C, and the like. The manufacturing process for the inner blanket fuel is similar to that for other blanket fuels.

[0071] (Step S2-6: MA transmutation) Step S2-6 is similar to step S1-6, except that inner blanket fuel is loaded into the core in addition to the core fuel, radial blanket fuel, upper axial blanket fuel, and lower axial blanket fuel. When inner blanket fuel is loaded, the core becomes a heterogeneous core.

[0072] An example of a core in the second embodiment will be described with reference to FIGS. As shown in FIGS. 9 and 10, a core 1A of this example is similar to the core 1 of the first embodiment, except that it further includes inner blanket fuel 17.

[0073] The inner blanket fuel 17 is disposed in the central region of the core 1A. The inner blanket fuel 17 is disposed between a plurality of core fuels 11 in the axial direction of the inner region of the central region. The inner region has a structure in which a layer of core fuel 11, a layer of inner blanket fuel 17, and a layer of core fuel 11 are stacked in this order from the axial lower side. The core 1A in this example is an axially heterogeneous core in which the core fuel 11 and the inner blanket fuel 17 are arranged in layers in the axial direction of the central region. In this embodiment, a portion of the core fuel 11 contained in the core fuel assemblies 30 is replaced with the inner blanket fuel 17 .

[0074] Another example of the core in the second embodiment will be described with reference to FIGS. As shown in FIGS. 11 and 12, the core 1B of this example is similar to the core 1A, except that the arrangement of the inner blanket fuel 17 is different and that inner core fuel 11A is arranged as the core fuel 11 in the outer region of the central region. Incidentally, the outer core fuel 11B may be arranged in the outer region.

[0075] The inner blanket fuel 17 is disposed in the central region of the core 1B. The inner blanket fuel 17 is disposed among the multiple core fuels 11 in the radial direction of the central region. More specifically, the inner blanket fuel 17 is disposed in the center of the inner region in a plan view, and is disposed in a ring shape by connecting the multiple control rods 16 surrounding it, and is also dispersed and disposed at multiple locations in the outer region. The inner blanket fuel 17 is disposed over the entire axial area of ​​the central region. The core 1B in this example is a radially heterogeneous core in which the core fuel 11 and the inner blanket fuel 17 are mixed in the radial direction of the central region. In this embodiment, a part of the core fuel assemblies 30 is replaced with an assembly (inner blanket fuel assembly) containing inner blanket fuel 17 instead of the core fuel 11.

[0076] After step S2-6, step S2-7 may be performed in which the operation of the fast reactor is stopped and the fuel after MA transmutation (core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14, and inner blanket fuel 17) is removed from the core 1. At this time, the neutron shield 15 may be removed together with the fuel.

[0077] After step S2-7, step S2-8 may be performed in which the fuel after MA transmutation (spent fuel) is reprocessed. In blanket fuel reprocessing, MA and Pu can be separated by known methods such as solvent extraction.

[0078] (Action and effect) In the core 1A having the above configuration, the blanket fuel includes inner blanket fuel 17 in addition to the radial blanket fuel, upper axial blanket fuel 13, and lower axial blanket fuel 14, and the inner blanket fuel also contains MA, so that the amount of MA that can be loaded into the core of the fast reactor can be increased and the nuclear transmutation amount of MA can be improved even more than in the first embodiment.

[0079] In particular, in the radially heterogeneous core as shown in Figures 11 and 12, the core fuel 11 and the inner blanket fuel 17 are mixed in the radial direction of the central region, so that the inner core fuel 11A with a relatively low content of fissile material can be arranged also in the outer region of the central region. This makes it possible to improve the MA transmutation rate by utilizing high-energy neutrons from the core fuel.

[0080] <Other embodiments> The above describes an embodiment of the present disclosure, but each configuration and combination thereof in the above embodiment is merely an example, and addition, omission, substitution, and other modifications of the configuration are possible without departing from the gist of the present disclosure.

[0081] When performing MA transmutation, the core configuration may be changed depending on the supply and demand balance of fissile materials such as Pu. For example, if the demand for Pu increases, the supply of Pu can be increased by loading blanket fuel made of U that does not contain MA.

[0082] An example of changing the core configuration to perform MA conversion is shown below. For example, first, core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14 and neutron shield 15 are arranged in a predetermined region of the fast reactor core to form core 1 (homogeneous core), and step S3-1 is performed in which the fast reactor is operated for an arbitrary period of time to perform MA nuclear transmutation. After step S3-1, the operation of the fast reactor is stopped, and step S3-2 is performed in which the core fuel 11, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14 and neutron shield 15 are removed from the core 1. After step S3-2, core fuel 11, inner blanket fuel 17, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14 and neutron shield 15 are arranged in a predetermined region of the fast reactor core to form core 1A (axially heterogeneous core), and step S3-3 is performed in which the fast reactor is operated for a desired period of time. After step S3-3, the operation of the fast reactor may be stopped, and step S3-4 may be performed in which the core fuel 11, the inner blanket fuel 17, the radial blanket fuel 12, the upper axial blanket fuel 13, the lower axial blanket fuel 14, and the neutron shield 15 are removed from the core 1A. After step S3-4, step S3-1 or step S3-3 may be performed again, or step S3-5 may be performed in which core fuel 11, inner blanket fuel 17, radial blanket fuel 12, upper axial blanket fuel 13, lower axial blanket fuel 14 and neutron shield 15 are arranged in a predetermined region of the fast reactor core to form core 1B (radially heterogeneous core) and the fast reactor is operated for a desired period of time. Also, step S3-6 may be performed to reprocess the fuel (spent fuel) removed in step S3-2 or the like.

[0083] <Additional Notes> The method for transmuting MA described in each embodiment can be understood, for example, as follows. (1) The method for transmuting MA according to the first and second embodiments is as follows: A step of separating MA from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, to obtain a first radioactive composition containing MA (S1-1, S2-1); Adding U to the first radioactive composition to obtain a second radioactive composition (S1-2, S2-2); Obtaining a solid solution of oxides of MA and U from the second radioactive composition (S1-3, S2-3); Steps (S1-4, S2-4) of storing the oxide solid solution and attenuating the highly exothermic minor actinides contained in the oxide solid solution; Using the stored oxide solid solution, a step (S1-5, S2-5) is performed to produce fuel including core fuel to be disposed in a central region of a fast reactor core, radial blanket fuel to be disposed in a radial blanket region radially outside the central region, upper axial blanket fuel to be disposed in an upper axial blanket region axially above the central region, and lower axial blanket fuel to be disposed in a lower axial blanket region axially below the central region; loading the produced fuel into the core of a fast reactor and transmuting the MA contained in the fuel (S1-6, S2-6); has.

[0084] In the method described above, U is added to the first radioactive composition obtained by MA separation, and stored as an oxide solid solution. This oxide solid solution is then used to produce fuels (core fuel, radial blanket fuel, upper axial blanket fuel, and lower axial blanket fuel) containing MA, which are then loaded into the core of a fast reactor for MA transmutation. This allows the amount of MA loaded into the core of the fast reactor to be increased, and the amount of MA transmutation to be improved compared to conventional methods. For example, the annual amount of MA transmutation can be increased to about 0.1 to 0.15 t-MA / (GW year).

[0085] (2) In the method for transmuting an MA according to the second embodiment, the fuel further includes an inner blanket fuel disposed in the central region. According to the above configuration, the amount of MA loaded in the core of the fast reactor can be further increased, and the amount of MA transmutation can be further improved. In addition, by disposing the inner blanket fuel in the central region, the contact area between the core fuel and the blanket fuel can be increased, and the MA transmutation efficiency can be further improved.

[0086] (3) The MA transmutation method according to the above first and second embodiments can further include a step (S1-8, S2-8) of reprocessing the fuel after the transmutation. According to the above configuration, the amount of nuclear transmutation of MA can be increased. [Explanation of symbols]

[0087] 100 fast reactor 1. Core 2 Reactor vessel 3 Guard vessel 4 Coolant inlet pipe 5 Coolant outlet piping 6. Reactor vessel pedestal 7. Upper core mechanism 8 Fixing plug 9 Furnace internals 11 Core fuel 11A Inner core fuel 11B Outer core fuel 12-diameter blanket fuel 13 Upper shaft blanket fuel 14 Lower Axis Blanket Fuel 15 Neutron Shielding 16 Control Rods 17 Internal Blanket Fuel 30 Core fuel assembly 31 Trumpet 32 Entrance nozzle 33 Handling Head 40 Core fuel elements 41 Cladding tube 43 Plenum Spring 48 Upper end plug 49 Lower end plug 71 Control Rod Drive Mechanism 72 Rotating plug 73 Rotating plug drive AC axis

Claims

1. A step of separating minor actinides from high-level radioactive liquid waste, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, to obtain a first radioactive composition containing minor actinides; doping said first radioactive composition with U to obtain a second radioactive composition; obtaining a solid solution of oxides of minor actinides and U from the second radioactive composition; a storage step for attenuating highly exothermic minor actinides contained in the oxide solid solution; Using the stored oxide solid solution, a process is performed to produce fuel including core fuel disposed in a central region of a fast reactor core, radial blanket fuel disposed in a radial blanket region radially outside the central region, upper axial blanket fuel disposed in an upper axial blanket region axially above the central region, and lower axial blanket fuel disposed in a lower axial blanket region axially below the central region; loading the produced fuel into a fast reactor and transmuting minor actinides contained in the fuel; The method for transmuting minor actinides comprises:

2. 2. The method of transmuting minor actinides as recited in claim 1, wherein said fuel further comprises an inner blanket fuel disposed in said central region.

3. 3. The method for transmutation of minor actinides according to claim 1, further comprising a step of reprocessing the fuel after the transmutation.

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