Method for Reducing Disposal Load of High-Level Radioactive Waste

Abstract

To provide a method for reducing the load of processing a high-level radioactive waste with an excellent effect of suppressing a waste liquid amount.SOLUTION: The method for reducing the load of processing a high-level radioactive waste includes the steps of: collecting minor actinoid (MA) and lanthanoid (Ln) from a high-level radioactive waste which is a waste liquid of which U and Pu are separated from a solution of a used nuclear fuel so that the decontamination factor of the Ln is at least 1 and less than 100, and obtaining a radioactive composition including MA and Ln; manufacturing a fuel of a fast reactor by using the radioactive composition; and loading the fuel into the fast reactor and fuel-converting the minor actinoid.SELECTED DRAWING: Figure 1

Description

Technical Field

[0001] The present invention relates to a method for reducing the disposal load of high-level radioactive waste and a fuel for a fast reactor.

Background Art

[0002] In the reprocessing of spent nuclear fuel generated from a light water reactor or the like, U (uranium) and Pu (plutonium) are recovered from the solution of the spent nuclear fuel. After recovering U and Pu, the high-level radioactive waste (Highly Active Liquid Waste) (hereinafter also referred to as HALW) remaining contains fission products (hereinafter also referred to as FP), as well as minor actinides (hereinafter also referred to as MA) such as Np (neptunium), Am (americium), and Cm (curium). HALW is planned to be vitrified into a glass solid through a concentration process and disposed of in a geological formation.

[0003] Some of the MA have nuclides with a long half-life on the order of 10,000 years, which is a factor increasing the load in the disposal of HALW. Therefore, nuclear conversion technologies for recovering MA from HALW and converting the recovered MA into different elements using a fast reactor or an accelerator (ADS) have been studied. The FP in HALW contains lanthanoids (hereinafter also referred to as Ln) such as La (lanthanum), Ce (cerium), and Eu (europium). Since Ln has similar chemical properties to MA, Ln is likely to be entrained when recovering MA from HALW. Therefore, after the MA recovery process, an MA purification process for separating MA and Ln is performed. As typical methods for the MA recovery process and the MA purification process, a solvent extraction method and an extraction chromatography method using a solvent (extractant solution) for extracting MA are known.

[0004] Fig. 8 shows an example of the process from the recovery of MA to nuclear conversion. In this example, first, MA is separated from HALW (MA separation step). In the MA separation step, MA is extracted from HALW by solvent extraction and then back-extracted from the extract. Since Ln is entrained with MA in the MA separation step, after the MA separation step, the back-extract is concentrated to separate MA and Ln (MA purification step). In the MA purification step, MA is extracted from the concentrated solution by solvent extraction and then back-extracted from the extract into the aqueous phase to obtain an MA-containing solution from which Ln has been separated. Then, the obtained MA-containing solution is concentrated to produce MA fuel, which is then subjected to nuclear conversion (burning) in a fast reactor. The waste liquid (containing FP) from the MA separation step is concentrated, vitrified, stored, and then disposed of in a geological formation. The waste liquid (containing Ln) from the MA purification step is concentrated, vitrified, stored, and then disposed of in a geological formation.

[0005] In the prior art, in order to produce MA fuel that does not contain Ln, an extremely high degree of purification is required in MA purification. Therefore, in MA purification, the extraction-back-extraction operation is usually repeated until the desired degree of purification is achieved. The increase in the number of steps increases the amount of waste liquid and also increases the equipment cost. In response to such problems, Patent Document 1 proposes a method of solidifying a liquid medium containing MA, Ln, and a liquid medium obtained through an MA recovery step and an MA purification step from HALW, and storing the solidified body.

Prior Art Documents

Patent Documents

[0006]

Patent Document 1

Summary of the Invention

Problems to be Solved by the Invention

[0007] However, the method of Patent Document 1 still does not have a sufficient effect of suppressing the amount of waste liquid. The present disclosure has been made to solve the above problems, and an object thereof is to provide a method for reducing the disposal load of HALW and a fuel for a fast reactor, which are excellent in the effect of suppressing the amount of waste liquid.

Means for Solving the Problem

[0008] In order to solve the above problems, the method for reducing the disposal load of HALW according to the present disclosure separates MA and Ln from HALW, which is a waste liquid obtained by separating U and Pu from a solution of used nuclear fuel, so that the decontamination factor of the Ln is 1 or more and less than 100, to obtain a radioactive composition containing the MA and the Ln; a step of manufacturing fuel for a fast reactor using the radioactive composition; and a step of loading the fuel into the fast reactor and performing nuclear conversion of the MA. Further, the fuel for a fast reactor according to the present disclosure contains MA and Ln.

Advantages of the Invention

[0009] The method for reducing the disposal load of HALW and the fuel for a fast reactor according to the present disclosure are excellent in the effect of suppressing the amount of waste liquid.

Brief Description of the Drawings

[0010]

Figure 1

Figure 2

Figure 3

Figure 4

Figure 5

Figure 6

Figure 7

Figure 8

Modes for Carrying Out the Invention

[0011] In this specification, MA (minor actinide) is an element other than Pu among the transuranic elements belonging to actinides. Actinides are a general term for elements with atomic numbers from 89 to 103. Ln (lanthanoid) is a general term for elements with atomic numbers from 57 to 71. The decontamination factor of Ln in the radioactive composition (hereinafter also referred to as DF Ln .) is the value obtained by dividing the amount (mass) of Ln in HALW by the amount (mass) of Ln after extraction and purification, and is represented by the following formula. As the amount of Ln, the radioactivity concentration may be used instead of the mass. DF Ln = Amount of Ln in HALW / Amount of Ln after extraction and purification

[0012] <First Embodiment> As shown in FIG. 1, the method for reducing the disposal load of HALW according to the first embodiment of the present disclosure is performing a process (MA separation) of separating MA together with Ln 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 and Ln in step S1-1; performing a purification process of MA (MA purification) on the first radioactive composition obtained in step S1-1 so that the decontamination factor of Ln is greater than 1 and less than 100, to obtain a second radioactive composition containing MA and Ln in step S1-2; manufacturing a fuel for a fast reactor using the second radioactive composition obtained in step S1-2 in step S1-3; loading the fuel obtained in step S1-3 into a fast reactor and performing nuclear conversion of MA in step S1-4. The waste liquid (containing FP) from which MA and Ln have been separated in step S1-1 is, for example, vitrified, stored, and then disposed of in a geological formation. The waste liquid containing Ln separated in step S1-2 is, for example, vitrified, stored, and then disposed of in a geological formation.

[0013] (HALW) HALW is the waste liquid obtained by separating U and Pu from the solution of used nuclear fuel. HALW contains FP, MA, Ln, etc. and does not contain U and Pu. HALW typically contains at least Np, Am, and Cm as MA and at least La and Ce as Ln. Depending on the type of used nuclear fuel (pressurized water reactor (PWR) fuel, boiling water reactor (BWR) fuel, MOX fuel, burnup, cooling period, etc.), the amounts and ratios of MA and Ln contained in the used nuclear fuel and HALW are different.

[0014] HALW is generated, for example, in the reprocessing of used nuclear fuel. In the reprocessing of used nuclear fuel, the used nuclear fuel taken out from the reactor is made into a powder product containing U and Pu through, for example, shearing, dissolution, clarification, metering and adjustment, extraction and separation, purification, and denitration. In extraction and separation, U and Pu are separated from the solution of used nuclear fuel. Extraction and separation are performed, for example, by the PUREX (Plutonium Uranium Redox EXtraction) method. In the PUREX method, a nitric acid solution of used nuclear fuel, tributyl phosphate (TBP), and an organic solvent such as dodecane are brought into contact and mixed. 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.

[0015] The used nuclear fuel may be the one that has been interim stored during reprocessing. When the fuel obtained by step S1-3 is blanket fuel, it is preferable that the used nuclear fuel is the used nuclear fuel that has been interim stored during reprocessing. The MA of the used fuel immediately taken out from the reactor contains high-heat-generating nuclides (Cm 243 、Cm 244 etc.). If blanket fuel is manufactured using such MA, there is a risk that the heat generation amount will become too high. Therefore, conventionally, high-heat-generating nuclides have been separated during MA purification. In the spent nuclear fuel that has been interim stored during reprocessing, since the decay heat of the high-heat-generating nuclides has been reduced by the interim storage, after the separation of MA from HALW, it can be used for blanket fuel manufacturing without performing the separation process of the high-heat-generating nuclides. In reprocessing, the period of interim storage of spent nuclear fuel is, for example, about 50 to 60 years. Note that when separating high-heat-generating nuclides in step S1-2 or when performing a step of storing a radioactive composition before fuel manufacturing as in the third embodiment described later, the spent nuclear fuel does not necessarily have to be interim stored. When the fuel obtained by step S1-3 is core fuel, there is no problem even if high-heat-generating nuclides are included.

[0016] (Step S1-1: MA separation) The separation process can be carried out by a known method. Examples of the separation process include a process including a process of extracting MA together with Ln from HALW by a solvent extraction method (hereinafter also referred to as MA extraction process), or a process including a process of adsorbing MA and Ln of HALW onto an adsorbent and eluting MA and Ln adsorbed on the adsorbent (hereinafter also referred to as MA adsorption-elution process). When the separation process includes an extraction process, after the extraction process, a process of back-extracting MA and Ln from the extract obtained by the extraction process (hereinafter also referred to as MA back-extraction process) may further be included.

[0017] In the MA extraction process, for example, HALW is brought into contact with an organic solvent solution containing an extractant (extractant solution). When HALW is brought into contact with the extractant solution, MA and Ln migrate to the extractant solution side.

[0018] Examples of the extractant include complexing agents that form complexes with MA and Ln. Such complexing agents are preferred because they are less expensive than complexing agents that selectively form complexes with MA. Specific examples of the complexing agent include n-octyl(phenyl)-N,N'-diisobutylcarbamoylmethylphosphine oxide-tributyl phosphate mixture (CMPO-TBP mixture), diisodecyl phosphate, 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), N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA), and the like. The extractant may be used alone or in combination of two or more.

[0019] The organic solvent can be appropriately selected according to the extractant used. It is desirable that the organic solvent be reusable, inexpensive, and resistant to radiation degradation. Specific examples of the organic solvent include, for example, n-dodecane. The organic solvent may be used alone or in combination of two or more.

[0020] The extract obtained by the MA extraction treatment may be directly used as the first radioactive composition in step S1-2, or further MA back-extraction treatment may be performed. In terms of further reducing the number of steps and the amount of waste liquid, it is preferable to use the extract obtained by the MA extraction treatment as the first radioactive composition in step S1-2. In the MA extraction treatment, DF can be adjusted by the type and concentration of the extractant, nitric acid concentration, volume ratio of the organic phase to the aqueous phase, and the like. Ln can be adjusted.

[0021] In the MA back-extraction treatment, for example, the extract is brought into contact with an aqueous solution containing nitric acid (stripping agent). As a result, MA and Ln in the extract migrate to the aqueous solution side. On the other hand, the organic solvent in the extract does not migrate to the aqueous solution side and remains on the extract side, so it can be reused. The back-extract obtained by the MA back-extraction treatment is typically directly used as the first radioactive composition in step S1-2.

[0022] The adsorbent used in the MA adsorption-elution treatment only needs to be able to adsorb MA and Ln. As a method for adsorbing MA and Ln in HALW onto the adsorbent, a method of bringing HALW into contact with the adsorbent can be mentioned. Examples of the method of bringing HALW into contact with the adsorbent include column type, batch type, etc. MA and Ln adsorbed on the adsorbent can be eluted by bringing the adsorbent into contact with the eluent. The eluent obtained by the MA adsorption-elution treatment, that is, the eluent after contact with the adsorbent, contains MA and Ln. Typically, this eluent is directly used as the first radioactive composition in step S1-2. In the MA adsorption-elution treatment, DF Ln can be adjusted by the type of adsorbent, temperature, pressure, solid-liquid ratio, etc.

[0023] In the first radioactive composition obtained in step S1-1, DF Ln is not particularly limited and may be 1 or may exceed 1. Regarding DF Ln in the first radioactive composition, from the viewpoint of suppressing the amount of waste liquid, it is preferably 1 or more and less than 13.

[0024] (Step S1-2: MA purification) In MA purification, MA in the first radioactive composition obtained in step S1-1 and a part of Ln are separated. If necessary, highly exothermic MA is separated from MA. MA purification can be carried out by a known method. Examples of MA purification include extraction chromatography using extractants such as COMPO and TODGA.

[0025] The second radioactive composition obtained in step S1-2 contains MA and Ln. Also, typically, since the amount of Ln is reduced compared to the first radioactive composition, DF Ln in the second radioactive composition is Ln greater than DF in the first radioactive composition. DF in the second radioactive composition Ln is less than 100, preferably less than 20, and may be less than 1, from the viewpoint of suppressing the amount of waste liquid.

[0026] (Step S1-3: Fuel production) Fuel can be produced by a known method according to the type of fuel. Examples of the fuel include core fuel and blanket fuel. The core fuel is typically MOX fuel and contains a mixed oxide of U and Pu. The blanket fuel typically contains depleted U. When the fuel is core fuel, for example, U, Pu, and the second radioactive composition are mixed to obtain a mixed oxide, and if necessary, it is formed into an arbitrary shape such as a pellet. When the fuel is blanket fuel, for example, depleted U and the second radioactive composition are mixed and, if necessary, formed into an arbitrary shape such as a pellet.

[0027] Since the second radioactive composition contains MA and Ln, the fuel obtained in step S1-3 also contains MA and Ln. The mass ratio of MA to Ln in the fuel is the same as the mass ratio of MA to Ln in the second radioactive composition.

[0028] The content of Ln in the fuel obtained in step S1-3 is set according to the type of fuel. For example, when the fuel is core fuel, the content of Ln is preferably 1 to 13% by mass based on the total amount of core fuel to be loaded. When the fuel is blanket fuel, the content of Ln is preferably 31 to 33% by mass based on the total amount of blanket fuel to be loaded. If the content of Ln is below the above upper limit, the nuclear conversion of MA can be carried out favorably.

[0029] When the fuel obtained in step S1-3 is core fuel, examples of the composition of the core fuel include the following compositions. The ratio of each component is the ratio to the total weight of the core fuel to be loaded. The total of U oxide, Pu oxide, MA oxide, and Ln oxide does not exceed 100% by mass. U oxide: 64 to 65% by mass, Plutonium oxide: 29 - 30 mass%, MA oxide: 3.2 - 4.5 mass%, Ln oxide: 0.5 - 3.4 mass%. Here, the core fuel of the fast reactor includes "inner core fuel" and "outer core fuel". The "total amount of core fuel to be loaded" is the total amount of all these core fuels. The core fuel may further contain other components as necessary.

[0030] When the fuel obtained by step S1 - 3 is blanket fuel, examples of the composition of the blanket fuel include the following. The ratio of each component is the ratio with respect to the total amount of the blanket fuel to be loaded. Deteriorated U: 65 mass%, MA: 3.0 mass%, Ln: 32 mass%. Here, the blanket fuel of the fast reactor includes "axial blanket fuel" (for example, the upper blanket fuel 45 and the lower blanket fuel 46 described later) and "radial blanket fuel". The "total amount of blanket fuel to be loaded" is the total amount of all these blanket fuels. The blanket fuel may further contain other components as necessary.

[0031] (Step S1 - 4: MA transmutation) By loading the fuel obtained in step S1 - 3 into the fast reactor and burning it, the transmutation of MA can be carried out. The fast reactor may be a known fast reactor.

[0032] With reference to FIG. 2, an example of the configuration of the fast reactor will be described. As shown in the figure, the fast reactor 100 in 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.

[0033] The reactor core 1 is a heat source containing fissile material. The detailed configuration of the reactor core 1 will be described later. The reactor vessel 2 is a vessel that houses the reactor core 1. The reactor vessel 2 has a cylindrical shape with a bottom surface. The reactor core 1 is fixed at the lower part inside the reactor vessel 2 via the in-vessel structure 12. The upper opening of the reactor vessel 2 is covered by a fixed plug 8. The fixed plug 8 is supported by the structure of the reactor building (reactor vessel pedestal 6).

[0034] The guard vessel 3 covers the reactor vessel 2 from the outside. That is, the reactor vessel 2 and the guard vessel 3 form a double-wall structure. Thereby, even when the coolant leaks from the reactor vessel 2, the coolant is retained by the guard vessel 3, and leakage to the outside is suppressed.

[0035] The coolant inlet pipe 4 guides the coolant (primary coolant) introduced from the outside into the reactor vessel 2. Examples of the coolant include liquid metal sodium. The end of the coolant inlet pipe 4 is located below the reactor core 1 inside the reactor vessel 2. Thereby, the inside of the reactor vessel 2 is filled with the coolant. The coolant outlet pipe 5 discharges the coolant inside the reactor vessel 2 to the outside. The end of the coolant outlet pipe 5 is located above the reactor core 1 inside the reactor vessel 2.

[0036] The upper core structure 7 includes a control rod drive mechanism 9, a rotating plug 10, and a rotating plug drive device 11. The control rod drive mechanism 9 is a device for inserting and withdrawing control rods for controlling the progress of the nuclear fission reaction inside the reactor core 1, which will be described later. The control rod drive mechanism 9 moves the control rod forward and backward in the vertical direction. The rotating plug 10 is a device for positioning equipment for exchanging the nuclear fuel (core fuel assembly 30, which will be described later) inside the reactor core 1. The rotating plug 10 is driven by the rotating plug drive device 11.

[0037] Next, the configuration of the core 1 will be described with reference to FIG. 3. As shown in the figure, the core 1 is an aggregate of members each having a hexagonal cross-sectional shape, and is arranged without gaps so as to form a hexagonal shape as a whole. The core 1 has a neutron shield 21, a radial blanket section 22, control rods 23, a neutron source 24, an outer core section 25, and an inner core section 26. However, the neutron source 24 may not be arranged.

[0038] The neutron shield 21 is arranged on the outermost peripheral side of the core 1. A plurality of neutron shields 21 are arranged so as to form a hexagonal ring. The radial blanket section 22 is provided inside the neutron shield 21. A plurality of radial blanket sections 22 are arranged so as to form a hexagonal ring. The radial blanket section 22 is formed by arranging a plurality of blanket fuel assemblies. The outer core section 25 is provided inside the radial blanket section 22. An inner core section 26 is provided further inside the outer core section 25. A plurality of control rods 23 can be inserted into a part of the region inside the inner core section 26. The outer core section 25 and the inner core section 26 are formed by arranging a plurality of core fuel assemblies 30 described later.

[0039] The outer core section 25 and the inner core section 26 generate heat by causing nuclear fission of nuclear fissile materials using neutrons generated from themselves or neutrons generated from the neutron source 24 as a trigger. The insertion amount of the control rods 23 is adjusted to control the progress of this nuclear fission reaction. In the radial blanket section 22, the reaction proceeds in a state where the fast nuclear fission reaction is reduced compared to the outer core section 25 and the inner core section 26. Also, in the radial blanket section 22, the amount of plutonium generated by the nuclear fission reaction is large compared to the outer core section 25 and the inner core section 26. The neutron shield 21 is provided to shield neutrons and suppress leakage to the outside.

[0040] Next, referring to FIG. 4, the configuration of the core fuel assembly 30 will be described. 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.

[0041] The trumpet tube 31 has a cylindrical shape centered on an axis Ac extending in the vertical direction. Also, the trumpet tube 31 has a hexagonal cross-sectional shape when viewed from the direction of the axis Ac. The lower opening of the trumpet tube 31 is closed by the entrance nozzle 32. Inside the entrance nozzle 32, a flow path (not shown) for guiding the coolant into the trumpet tube 31 is formed. The entrance nozzle 32 is formed with an opening for communicating the flow path 32F with the outside. A handling head 33 is attached to the upper opening of the trumpet tube 31. The handling head 33 is a portion that is gripped by a device when transporting the core fuel assembly 30.

[0042] Inside the trumpet tube 31, 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. The coolant guided from the opening of the entrance nozzle 32 flows through the space around and above the core fuel element 40.

[0043] Next, referring to FIG. 5, the configuration of the core fuel element 40 will be described. 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, core fuel 44, upper blanket fuel 45, and lower blanket fuel 46 housed inside the cladding tube 41, and upper end plugs 48 and lower end plugs 49 provided at both ends of the cladding tube 41.

[0044] The core fuel 44, the upper blanket fuel 45, and the lower blanket fuel 46 are each formed into a cylindrical pellet shape. A plurality of lower blanket fuels 46, a plurality of core fuels 44, and a plurality of upper blanket fuels 45 are filled in this order so as to be stacked from below the cladding tube 41. The uppermost upper blanket fuel 45 is pressed downward by the plenum spring 43.

[0045] The blanket fuel assembly constituting the radial blanket portion 22 is the same as the core fuel assembly except that it includes a blanket fuel element instead of the core fuel element 40. The blanket fuel element is the same as the core fuel element 40 except that it includes a blanket fuel instead of the core fuel 44, the upper blanket fuel 45, and the lower blanket fuel 46.

[0046] When the fuel obtained in step S1-3 is core fuel, this fuel may be used for the core fuel 44 in the outer core portion 25, may be used for the core fuel 44 in the inner core portion 26, or may be used for both of them, but it is preferably used for the core fuel 44 in the inner core portion 26. Generally, the core fuel 44 in the outer core portion 25 and the core fuel 44 in the inner core portion 26 have different compositions. For example, the core fuel 44 in the outer core portion 25 tends to have a higher Pu enrichment than the core fuel 44 in the inner core portion 26. Conventionally, when manufacturing core fuels 44 with different compositions in a common line, it is necessary to perform a cleaning operation to remove the fuel remaining in the common line so that the fuel with a high Pu enrichment for the outer core portion 25 does not mix with the fuel with a low Pu enrichment for the inner core portion 26, and the burden of the cleaning operation was large. On the other hand, when using the fuel obtained in step S1-3 for the core fuel 44 in the inner core portion 26, it is possible to adjust the Pu enrichments of the outer core portion 25 and the inner core portion 26 by adjusting the amount of Ln contained in the fuel. Thereby, the fuel before containing Ln can use the fuel with the same Pu enrichment, and the cleaning operation of the common line during fuel production becomes easy.

[0047] When the fuel obtained in step S1-3 is blanket fuel, this fuel may be loaded as upper blanket fuel 45 or lower blanket fuel 46 into the core fuel element 40, or may be loaded into the blanket fuel element. From the viewpoint of MA conversion, it is preferable to load it into the blanket fuel element.

[0048] (Function and effect) In the method with the above configuration, the DF of the second radioactive composition used for fuel production Ln is greater than 1 and less than 100, so the number of steps for MA purification when obtaining the second radioactive composition can be reduced, for example, to about 3 times or less. Therefore, the amount of waste liquid and equipment cost due to MA purification can be suppressed.

[0049] <Second Embodiment> As shown in FIG. 6, the method for reducing the disposal load of HALW according to the second embodiment of the present disclosure is a process (MA separation) of separating MA together with Ln from HALW, which is a waste liquid obtained by separating U and Pu from the solution of spent nuclear fuel, so that the decontamination factor of Ln is 1 or more and less than 100, to obtain a radioactive composition containing MA and Ln in step S2-1, step S2-2 of manufacturing the fuel of the fast reactor using the radioactive composition obtained in step S2-1, step S2-3 of loading the fuel obtained in step S2-2 into the fast reactor and performing MA conversion. The waste liquid (containing FP) from which MA and Ln were separated in step S2-1 is, for example, vitrified, stored, and then disposed of in the ground.

[0050] (Step S2-1: MA separation) Step S2-1 is the same as step S1-1, except that the separation process is performed so that the DF in the radioactive composition obtained by the separation process Ln is less than 100.

[0051] The radioactive composition obtained in step S2-1 contains MA and Ln. The DF in the radioactive compositionLn is less than 100, more preferably less than 20, and may even be less than 1, from the viewpoint of suppressing the amount of waste liquid.

[0052] (Step S2-2: Fuel production) Step S2-2 is the same as step S1-3.

[0053] (Step S2-3: MA conversion) Step S2-3 is the same as step S1-4.

[0054] (Operational effect) In the method with the above configuration, the DF of the radioactive composition used for fuel production Ln is more than 1 and less than 100, and since the DF Ln is set to the target value during MA separation, MA purification can be omitted. Therefore, the amount of waste liquid and equipment cost due to MA purification can be suppressed.

[0055] <Third Embodiment> As shown in FIG. 7, the method for reducing the disposal load of HALW according to the third embodiment of the present disclosure is a step S3-1 of obtaining a radioactive composition containing MA and Ln by performing a process (MA separation) of separating MA together with Ln from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, so that the decontamination factor of Ln is 1 or more and less than 100; a step S3-2 of storing the radioactive composition obtained in step S3-1; a step S3-3 of manufacturing a fuel for a fast reactor using the radioactive composition stored in step S3-2; a step S3-4 of loading the fuel obtained in step S3-3 into a fast reactor and performing nuclear conversion of MA. The waste liquid (containing FP) from which MA and Ln have been separated in step S3-1 is, for example, vitrified, stored, and then disposed of in a geological formation.

[0056] (Step S3-1: MA separation) Step S3-1 is the same as step S2-1.

[0057] (Step S3-2: Storage) In step S3-2, the radioactive composition is temporarily stored. The radioactive composition may be stored in a liquid state or may be solidified and stored in the form of a solidified body. In terms of a small storage load, it is preferable to store it in the form of a solidified body.

[0058] Examples of the solidification treatment include decomposition treatment, hydrothermal treatment, and vitrification treatment. When performing vitrification treatment, it is preferable to perform a concentration treatment for concentrating the radioactive composition before the vitrification treatment.

[0059] When the radioactive composition is an extract obtained by an extraction treatment, the decomposition treatment is preferable as the solidification treatment. By performing the decomposition treatment on the extract, the organic solvent of the extract is removed and MA is oxidized. Thereby, a solidified body containing MA oxide is obtained. Examples of the decomposition treatment include distillation, thermal decomposition, and incineration. Distillation can be carried out using a known evaporation method such as a batch method or a continuous method (tray column or packed column).

[0060] When the radioactive composition is a stripping solution obtained by a stripping treatment or an eluate obtained by an adsorption-elution treatment, the hydrothermal treatment or a treatment of concentrating the liquid material and vitrifying the obtained concentrated solution (concentration-vitrification treatment) is preferable as the solidification treatment. By subjecting the radioactive composition to hydrothermal treatment, MA contained in the radioactive composition is oxidized and a solidified body containing MA oxide precipitates. The solidified body is separated from the liquid medium (water) by solid-liquid separation. Thereby, a solidified body containing MA oxide is obtained. Concentration of the radioactive composition and vitrification of the concentrated solution can be each carried out by a conventional method.

[0061] If necessary, before storing the obtained solidified body, a stabilization treatment for removing part or all of the carbon, hydrogen, oxygen, and nitrogen components may be performed on the solidified body. The solidified product may contain organic substances such as an extractant and radiolytic products of the extractant. If the solidified product contains organic substances, gas may be generated during storage, which may cause problems such as impairing the confinement function of radioactive substances. By performing a stabilization treatment to remove part or all of the carbon, hydrogen, oxygen, and nitrogen components from the solidified product, the occurrence of such problems can be suppressed. Examples of the stabilization treatment include calcination and sintering.

[0062] As a method for storing the solidified product, a method known as a dry storage method for radioactive waste can be used. For example, a method of storing the solidified product in a plurality of canisters, storing these plurality of canisters in a cask, and storing them in a storage facility can be mentioned.

[0063] The storage period of the radioactive composition can be set as appropriate. Highly heat-generating MAs such as Cm are not suitable for core fuel fabrication. Therefore, when the radioactive composition contains highly heat-generating MAs, in step S3-2, it is preferable to store the radioactive composition or its solidified product until the highly heat-generating MAs contained in the radioactive composition or its solidified product are sufficiently decayed. Thereby, the radioactive composition or its solidified product after storage can be used for core fuel fabrication without performing a separation process for highly heat-generating MAs. Note that Cm contained in HALW is mainly Cm with a relatively short half-life of about 18 years. 244 Cm.

[0064] (Step S3-3: Fuel Fabrication) Step S3-3 is the same as step S2-2. However, in step S3-2, when the radioactive composition is made into a solidified product, the solidified product is dissolved and fuel is fabricated using the solution. To dissolve the solidified product, for example, if it is a solidified product containing MA oxide, an aqueous nitric acid solution may be added. When the solidified product is a vitrified solidified product, after dissolving the solidified product with an acid solution or the like, the glass component is separated from the obtained solution.

[0065] (Step S3-4: MA Transmutation) Step S3-4 is the same as step S2-3.

[0066] (Function and effect) In the method with the above configuration, the DF of the radioactive composition used for fuel production Ln is greater than 1 and less than 100, and also, when separating MA, DF Ln is set to the target value, so MA purification can be omitted. Therefore, the amount of waste liquid and equipment cost due to MA purification can be suppressed. Also, since the radioactive composition is stored before fuel production, it can be used for core fuel production without performing the separation process of highly heat-generating MA.

[0067] As described above, the embodiments of the present disclosure have been explained. However, each configuration and their combinations in the above embodiments are examples, and additions, omissions, substitutions, and other changes of the configuration are possible without departing from the gist of the present disclosure.

[0068] <Supplementary Note> The method for reducing the disposal load of HALW described in each embodiment is understood as follows, for example. (1) The method for reducing the disposal load of HALW according to the first to third embodiments above includes steps of recovering MA and Ln from HALW so that DF Ln is 1 or more and less than 100 to obtain a radioactive composition containing MA and Ln (steps S1-1 and S1-2, S2-1, S3-1), manufacturing fuel for a fast reactor using the radioactive composition (steps S1-3, S2-2, S3-3), loading the fuel into the fast reactor, and performing nuclear conversion of MA (steps S1-4, S2-3, S3-4).

[0069] Conventionally, when performing nuclear conversion of MA in a fast reactor, a high purification degree with DF Ln = 100 or more has been required. MA purification for such a high purification degree results in an increase in the amount of waste liquid and equipment cost. According to the study by the present inventors, DF LnIt has been found that even fuels produced using radioactive compositions with less than 100 can perform nuclear conversion of MA without problems. Therefore, in the method according to the above embodiment, DF Ln was set to less than 100. DF Ln If it is less than 100, MA purification can be omitted, or even when MA purification is performed, the number of steps can be reduced. Therefore, the amount of waste liquid and equipment cost due to MA purification can be suppressed.

[0070] (2) In the method according to the first to third embodiments, it is preferable that the fuel is a core fuel. In this case, the decontamination factor of Ln is preferably 1 or more and less than 20. According to the above configuration, the recovered MA accompanied by Ln in the core fuel can be nuclear-converted.

[0071] (3) In the method according to the first to third embodiments, it is also preferable that the fuel is a blanket fuel. According to the above configuration, the recovered MA accompanied by Ln in the blanket fuel can be nuclear-converted.

[0072] (4) In the method according to the first to third embodiments, when the fuel is a blanket fuel, it is preferable to have a step (S3-2) of storing the radioactive composition before the step (S3-3) of manufacturing the fuel, as in the method according to the third embodiment. The MA in spent fuel contains high-heat-generating nuclides (Cm 243 , Cm 244 , etc.). When manufacturing blanket fuel using MA containing such nuclides, there is a risk that the heat generation amount will become too high. Therefore, conventionally, high-heat-generating nuclides have been separated during MA purification. According to the above configuration, since the decay heat of high-heat-generating nuclides can be reduced, after separating MA from HALW, it can be used for MA fuel manufacturing without performing separation processing of high-heat-generating nuclides.

[0073] (5) In the method according to the first to third embodiments, when the fuel is blanket fuel, it is also preferable that the spent nuclear fuel is the spent nuclear fuel that has been interim stored during reprocessing. The MA of spent fuel contains high heat-generating nuclides (Cm 243 , Cm 244 , etc.). When manufacturing blanket fuel using such MA, there is a risk that the calorific value will become too high. Therefore, conventionally, high heat-generating nuclides have been separated during MA purification. According to the above configuration, since the decay heat of high heat-generating nuclides is reduced by interim storage, after separating MA from HALW, it can be used for manufacturing blanket fuel without performing the separation process of high heat-generating nuclides.

[0074] The fuel of the fast reactor described in each embodiment is understood as follows, for example. (6) The fuel of the fast reactor according to the first to second embodiments contains MA and Ln.

[0075] Conventionally, when performing nuclear conversion of MA in a fast reactor, a high degree of purification has been required so that Ln does not mix into the MA fuel. In the fuel according to the above embodiment, since it contains Ln, the decontamination factor of Ln in the process of manufacturing the fuel can be made less than 100, and the amount of waste liquid and equipment cost due to MA purification can be suppressed.

[0076] (7) The fuel according to the first to second embodiments is preferably core fuel. In this case, the content of Ln is preferably 1 to 13 mass% with respect to the total amount of core fuel to be loaded. According to the above configuration, the recovered MA can be nuclear converted with Ln accompanying the core fuel.

[0077] (8) The fuel according to the first to second embodiments is preferably blanket fuel. In this case, the content of Ln is preferably 31 to 33 mass% with respect to the total amount of blanket fuel to be loaded. According to the above configuration, the recovered MA can be nuclear converted with Ln accompanying the blanket fuel.

Description of Symbols

[0078] 100 Fast reactor 1 Core 2 Reactor vessel 3 Guard vessel 4 Coolant inlet pipe 5 Coolant outlet pipe 6 Reactor vessel pedestal 7 Upper core structure 8 Fixed plug 9 Control rod drive mechanism 10 Rotating plug 11 Rotating plug drive device 12 In-vessel structure 21 Neutron shield 22 Radial blanket section 23 Control rod 24 Neutron source 25 Outer core section 26 Inner core section 30 Core fuel assembly 31 Trumpet tube 32 Entrance nozzle 33 Handling head 40 Core fuel element 41 Cladding tube 43 Plenum spring 44 Core fuel 45 Upper blanket fuel 46 Lower blanket fuel 48 Upper end plug 49 Lower end plug Axis of Ac

Claims

1. A step of recovering minor actinides and lanthanoids from high-level radioactive waste, which is waste liquid obtained by separating U and Pu from a spent nuclear fuel solution, and obtaining a radioactive composition containing the minor actinides and the lanthanoids; A step of manufacturing a fuel for a fast reactor using the radioactive composition; A step of loading the fuel into a fast reactor and performing nuclear conversion of the minor actinides, and having: The step of obtaining the radioactive composition includes a separation process of separating the minor actinides from the high-level radioactive waste together with the lanthanoids, and a purification process of separating a part of the minor actinides and the lanthanoids from the first radioactive composition obtained by the separation process to obtain a radioactive composition in which the amount of the lanthanoids is reduced compared to the first radioactive composition; The separation process is an extraction process of extracting the minor actinides from the high-level radioactive waste together with the lanthanoids by a solvent extraction method, or a process of back-extracting the minor actinides and the lanthanoids from the extraction liquid obtained by the extraction process, or a process of adsorbing the minor actinides and the lanthanoids of the high-level radioactive waste onto an adsorbent and eluting the minor actinides and the lanthanoids adsorbed on the adsorbent. The separation process is performed a number of times such that the decontamination factor of the lanthanoids, represented by the amount of lanthanoids in the high-level radioactive waste / the amount of lanthanoids in the radioactive composition, is 1 or more and less than 13; The purification process is a process by an extraction chromatography method, and the purification process is performed a number of times such that the decontamination factor of the lanthanoids is less than 20. A method for reducing the disposal load of high-level radioactive waste.

2. A step of recovering minor actinides and lanthanoids from high-level radioactive waste, which is waste liquid obtained by separating U and Pu from a spent nuclear fuel solution, and obtaining a radioactive composition containing the minor actinides and the lanthanoids; A step of manufacturing a fuel for a fast reactor using the radioactive composition; loading the fuel into a fast reactor and performing nuclear conversion of the minor actinides, The step of obtaining the radioactive composition consists only of a separation process of separating the minor actinides together with the lanthanoids from the high-level radioactive waste, The separation process is a process of extracting the minor actinides together with the lanthanoids from the high-level radioactive waste by a solvent extraction method, or a process of back-extracting the minor actinides and the lanthanoids from the extraction liquid obtained by the extraction process, or a process of adsorbing the minor actinides and the lanthanoids of the high-level radioactive waste onto an adsorbent and eluting the minor actinides and the lanthanoids adsorbed on the adsorbent. The separation process is performed a number of times such that the decontamination factor of the lanthanoids represented by the amount of lanthanoids in the high-level radioactive waste / the amount of lanthanoids in the radioactive composition is 1 or more and less than 20. A method for reducing the disposal load of high-level radioactive waste.

3. The reduction method according to claim 1 or 2, wherein the fuel is a core fuel.

4. The reduction method according to claim 1 or 2, wherein the fuel is a blanket fuel.

5. The reduction method according to claim 4, having a step of storing the radioactive composition before the step of manufacturing the fuel.

6. The reduction method according to claim 4 or 5, wherein the spent nuclear fuel is spent nuclear fuel that has been interim stored in reprocessing.

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