JP2023119380A - Method for reducing load of processing high-level radioactive waste and fuel of fast reactor - Google Patents

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

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 spent nuclear fuel solution. Highly Active Liquid Waste (hereinafter also referred to as HALW) remaining after recovery of U and Pu includes fission products (hereinafter also referred to as FP) and Np ( neptunium), Am (americium), Cm (curium), and other minor actinides (hereinafter also referred to as MA). HALW is planned to be vitrified through a concentration process and disposed of in a geological layer.

MA has nuclides with long half-lives on the order of 10,000 years, and is a factor in increasing the load in disposal of HALW. Therefore, a nuclear transmutation technology is being studied that recovers MA from HALW and converts the recovered MA into different elements using a fast reactor or an accelerator (ADS).
FPs in HALW include 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 tends to accompany the recovery of MA from HALW. Therefore, after the MA recovery step, an MA purification step for separating MA and Ln is performed. Solvent extraction and extraction chromatography using a solvent (extractant solution) for extracting MA are known as representative methods of the MA recovery step and MA purification step.

FIG. 8 shows an example of the process from recovery of MA to nuclear transmutation. In this example, first, MA is separated from HALW (MA separation step). In the MA separation step, MA is solvent extracted from HALW and back extracted from the extract. Since Ln accompanies MA in the MA separation step, after the MA separation step, the reverse extract is concentrated to separate MA and Ln (MA purification step). In the MA purification step, MA is solvent-extracted from the concentrate, and MA is back-extracted from the extract into the aqueous phase to obtain a MA-containing solution from which Ln has been separated. After that, the obtained MA-containing solution is concentrated to produce MA fuel, which is then transmuted (combusted) in a fast reactor. The effluent (containing FP) from the MA separation process is concentrated, vitrified, stored, and then geologically disposed. The waste liquid (containing Ln) from the MA purification process is concentrated, vitrified, stored, and then geologically disposed.

In the prior art, an extremely high degree of purification is required in MA refining in order to produce MA fuel that does not contain Ln. Therefore, in MA purification, extraction-back extraction operations are usually repeated until the desired degree of purification is achieved. As the number of processes increases, the amount of waste liquid increases and the equipment cost also increases.
To address this problem, Patent Document 1 proposes a method of solidifying a liquid medium containing MA, Ln, and a liquid medium obtained from HALW through the MA recovery process and MA purification process, and storing the solidified product. It is

JP 2020-125989 A

However, the method of Patent Document 1 is still insufficient in the effect of suppressing the amount of waste liquid.
The present disclosure has been made to solve the above problems, and aims to provide a HALW disposal load reduction method and a fast reactor fuel that are excellent in the effect of suppressing the amount of waste liquid.

In order to solve the above problems, the HALW disposal load reduction method according to the present disclosure removes MA and Ln from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, and recovering to be less than 100 to obtain a radioactive composition containing the MA and the Ln; using the radioactive composition to produce fuel for a fast reactor; and loading the fuel into the fast reactor. and transmuting the MA.
Also, the fast reactor fuel according to the present disclosure includes MA and Ln.

The HALW disposal load reduction method and fast reactor fuel of the present disclosure are excellent in the effect of suppressing the amount of waste liquid.

FIG. 2 is a flow diagram illustrating a method for reducing the HALW disposal load according to the first embodiment of the present disclosure; 1 is a longitudinal sectional view showing the configuration of a fast reactor; FIG. FIG. 2 is a plan view showing the structure of a core; FIG. 2 is a plan view showing the structure of a core fuel assembly; FIG. 4 is a plan view showing the configuration of a core fuel element; FIG. 10 is a flow diagram illustrating a method for reducing the HALW disposal load according to the second embodiment of the present disclosure; FIG. 10 is a flow diagram illustrating a method for reducing the HALW disposal load according to the second embodiment of the present disclosure; FIG. 10 is a flow chart explaining a process according to a comparative example;

In this specification, MA (minor actinide) is an element other than Pu among transuranic elements belonging to actinide.
Actinide is a generic 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 for Ln in a 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 expressed by the following formula. be done. As the Ln amount, radioactivity concentration may be used instead of mass.
DF _Ln = Ln amount in HALW/Ln amount after extraction and purification

<First Embodiment>
As shown in FIG. 1, the HALW disposal load reduction method according to the first embodiment of the present disclosure includes:
Step S1-1 of obtaining a first radioactive composition containing MA and Ln by performing a treatment (MA separation) for separating MA and Ln from HALW, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel. and,
The first radioactive composition obtained in step S1-1 is subjected to MA purification treatment (MA purification) so that the decontamination coefficient of Ln is greater than 1 and less than 100, so that the first radioactive composition containing MA and Ln Step S1-2 of obtaining the radioactive composition of 2;
a step S1-3 of producing fast reactor fuel using the second radioactive composition obtained in step S1-2;
and a step S1-4 of loading the fuel obtained in step S1-3 into the fast reactor and transmuting the MA.
The waste liquid (containing FP) from which MA and Ln have been separated in step S1-1 is, for example, vitrified, stored, and then geologically disposed.
The waste liquid containing Ln separated in step S1-2 is, for example, vitrified, stored, and then geologically disposed.

(HALW)
HALW is the effluent from which U and Pu are separated from the solution of spent nuclear fuel. HALW includes FP, MA, Ln, etc. and excludes U and Pu. HALW typically contains at least Np, Am and Cm as MA, and at least La and Ce as Ln.
The amount and ratio of MA and Ln contained in spent nuclear fuel and HALW, depending on the type of spent nuclear fuel (pressurized water reactor (PWR) fuel, boiling water reactor (BWR) fuel, MOX fuel, burnup, cooling period, etc.) is different.

HALW is produced, for example, by reprocessing spent nuclear fuel. In the reprocessing of spent nuclear fuel, the spent nuclear fuel removed from the reactor is subjected to, for example, shearing, dissolution, clarification, weighing/adjustment, extraction/separation, refining, and denitrification, and is turned into a powder product containing U and Pu. be. In the extraction/separation, U and Pu are separated from the spent nuclear fuel solution. Extraction and separation are performed, for example, by a PUREX (Plutonium Uranium Redox EXtraction) method. In the Purex method, a nitric acid solution of spent nuclear fuel, tributyl phosphate (TBP), and an organic solvent such as dodecane are contacted and mixed. As a result, U and Pu in the nitric acid solution form a complex with TBP and move to the organic solvent side. On the other hand, FP, MA, and Ln remain on the nitric acid solution (waste liquid) side.

The spent nuclear fuel may have been intermediately stored during reprocessing.
If the fuel obtained by step S1-3 is blanket fuel, the spent nuclear fuel is preferably spent nuclear fuel interimly stored in reprocessing.
The MA of the spent fuel immediately after being removed from the nuclear reactor contains highly exothermic nuclides (Cm ²⁴³ , Cm ^244, etc.). If such MA is used to produce blanket fuel, the calorific value may become too high, so conventionally, highly exothermic nuclides were separated during MA refining.
In the spent nuclear fuel that has been interimly stored during reprocessing, the decay heat of highly exothermic nuclides has been reduced by interim storage. , can be used for blanket fuel production.
In reprocessing, the period of interim storage of spent nuclear fuel is, for example, about 50 to 60 years.
It should be noted that when separating the highly exothermic nuclides in step S1-2, or when performing a step of storing the radioactive composition before fuel production as in the third embodiment described later, the spent nuclear fuel It does not have to be intermediately stored.
If the fuel obtained in step S1-3 is core fuel, there is no problem even if it contains highly exothermic nuclides.

(Step S1-1: MA separation)
Separation treatment can be carried out by a known method.
The separation treatment includes, for example, a treatment of extracting MA from HALW together with Ln by a solvent extraction method (hereinafter also referred to as MA extraction treatment), or a treatment in which MA and Ln of HALW are adsorbed on an adsorbent, and A treatment including a treatment for eluting MA and Ln adsorbed to (hereinafter also referred to as MA adsorption-elution treatment).
When the separation process includes an extraction process, it may further include 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) after the extraction process.

In the MA extraction treatment, 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.

Extractants include, for example, complexing agents that form complexes with MA and Ln. Such a complexing agent is preferred because it is less expensive than a complexing agent that selectively forms a complex with MA. Specific examples of complexing agents 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. One extractant may be used alone, or two or more extractants may be used in combination.

The organic solvent can be appropriately selected according to the extractant to be used. Desirably, the organic solvent should be reusable, inexpensive, and resistant to radiation degradation. A specific example of the organic solvent is n-dodecane. An organic solvent may be used individually by 1 type, and may be used in combination of 2 or more types.

The liquid extract obtained by the MA extraction process may be directly used as the first radioactive composition in step S1-2, and may be further subjected to the MA reverse extraction process. From the viewpoint 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 process as the first radioactive composition for step S1-2.
In the MA extraction process, the DF _Ln can be adjusted by the type and concentration of the extractant, the nitric acid concentration, the volume ratio of the organic phase to the aqueous phase, and the like.

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

The adsorbent used for the MA adsorption-elution treatment should be capable of adsorbing MA and Ln.
As a method for adsorbing MA and Ln of HALW to an adsorbent, there is a method of contacting HALW with an adsorbent. Examples of methods for contacting HALW with an adsorbent include a column method and a batch method.
MA and Ln adsorbed on the adsorbent can be eluted by bringing the adsorbent into contact with an eluent.
The eluate obtained from the MA adsorption-elution process, ie after contact with the adsorbent, contains MA and Ln. Typically, this eluent is directly subjected to step S1-2 as the first radioactive composition.
In the MA adsorption-elution process, DF _Ln can be adjusted by the type of adsorbent, temperature, pressure, solid-liquid ratio, and the like.

In the first radioactive composition obtained in step S1-1, DF _Ln is not particularly limited and may be 1 or greater than 1.
DF _Ln in the first radioactive composition is preferably 1 or more and less than 13 from the viewpoint of suppressing the amount of waste liquid.

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

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

(Step S1-3: fuel production)
Fuel production can be carried out by known methods depending on the type of fuel.
Examples of fuel include core fuel and blanket fuel. The core fuel is typically MOX fuel and contains mixed oxides of U and Pu. 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, which is optionally formed into an arbitrary shape such as pellets. When the fuel is a blanket fuel, for example, the depleted U and the second radioactive composition are mixed and optionally formed into any shape such as pellets.

Since the second radioactive composition contains MA and Ln, the fuel obtained by step S1-3 also contains MA and Ln. The mass ratio of MA to Ln in the fuel is similar to the mass ratio of MA to Ln in the second radioactive composition.

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 with respect to the total amount of loaded core fuel. When the fuel is blanket fuel, the content of Ln is preferably 31 to 33% by mass with respect to the total amount of loaded blanket fuel. If the content of Ln is equal to or less than the above upper limit, the nuclear transmutation of MA can be carried out satisfactorily.

When the fuel obtained in step S1-3 is the core fuel, the composition of the core fuel includes, for example, the following composition. The ratio of each component is the ratio to the total weight of the loaded core fuel. The sum of U oxide, Pu oxide, MA oxide and Ln oxide does not exceed 100% by mass.
U oxide: 64 to 65% by mass,
Pu oxide: 29 to 30% by mass,
MA oxide: 3.2 to 4.5% by mass,
Ln oxide: 0.5 to 3.4% by mass.
Here, the core fuel of the fast reactor includes "inner core fuel" and "outer core fuel". "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 needed.

When the fuel obtained in step S1-3 is blanket fuel, the composition of the blanket fuel includes, for example, the following composition. The ratio of each component is the ratio to the total amount of blanket fuel to be loaded.
Degradation U: 65% by mass,
MA: 3.0% by mass,
Ln: 32% by mass.
Here, the blanket fuel of the fast reactor includes "axial blanket fuel" (for example, upper blanket fuel 45 and lower blanket fuel 46, which will be described later) and "radial blanket fuel". "Total amount of blanket fuel loaded" is the total amount of all these blanket fuels.
The blanket fuel may further contain other ingredients as needed.

(Step S1-4: MA transmutation)
The nuclear transmutation of MA can be performed by loading the fuel obtained in step S1-3 into the fast reactor and burning it.
The fast reactor may be a known fast reactor.

An example of the configuration of a fast reactor will be described with reference to FIG. As shown in the figure, the 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, a fixed plug 8;

Core 1 is a heat source containing fissile material. A detailed configuration of the 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 tubular shape with a bottom surface. The core 1 is fixed to the lower part inside the reactor vessel 2 via the core internals 12 . An upper opening of the reactor vessel 2 is covered with a fixed plug 8 . The fixed plug 8 is supported by the reactor building structure (reactor vessel pedestal 6).

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. As a result, even if the coolant leaks from the reactor vessel 2, the coolant is held by the guard vessel 3 and is prevented from leaking to the outside.

The coolant inlet pipe 4 guides the coolant (primary coolant) introduced from the outside into the reactor vessel 2 . Examples of coolants include liquid metal sodium. The end of the coolant inlet pipe 4 is located below the core 1 inside the reactor vessel 2 . As a result, 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 core 1 inside the reactor vessel 2 .

The core upper mechanism 7 has 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 in the core 1, which will be described later. The control rod drive mechanism 9 moves the control rods back and forth in the vertical direction. The rotating plug 10 is a device for positioning equipment for exchanging nuclear fuel (a core fuel assembly 30 to be described later) in the core 1 . The rotating plug 10 is driven by a rotating plug driving device 11 .

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

The neutron shields 21 are arranged on the outermost side of the core 1 . A plurality of neutron shields 21 are arranged in a hexagonal ring.
The radial blanket portion 22 is provided inside the neutron shield 21 . A plurality of radial blanket portions 22 are arranged to form a hexagonal ring. The radial blanket portion 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 inside the outer core section 25 . A plurality of control rods 23 can be inserted into a part of 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.

The outer core section 25 and the inner core section 26 generate heat by nuclear fission of fissionable materials triggered by neutrons generated by themselves or neutrons generated by the neutron source 24 . The amount of insertion of the control rod 23 is adjusted to control the progress of this nuclear fission reaction.
In the radial blanket portion 22 , reactions proceed with reduced fast fission reactions as compared to the outer core portion 25 and the inner core portion 26 . In addition, the amount of plutonium produced by the nuclear fission reaction is greater in the radial blanket portion 22 than in the outer core portion 25 and the inner core portion 26 .
The neutron shield 21 is provided to shield neutrons and prevent leakage to the outside.

Next, the configuration of the core fuel assembly 30 will be described with reference to FIG. 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 .

The trumpet tube 31 has a tubular shape centered on an axis line Ac extending in the vertical direction. Moreover, the trumpet tube 31 has a hexagonal cross-sectional shape when viewed from the direction of the axis Ac. A lower opening of the trumpet tube 31 is closed by an entrance nozzle 32 . Inside the entrance nozzle 32 , a flow path (not shown) is formed for guiding the coolant to the inside of the trumpet tube 31 . The entrance nozzle 32 is formed with an opening that communicates 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 the part that is gripped by the 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 guided from the opening of the entrance nozzle 32 flows in the space around and above the core fuel element 40 .

Next, the configuration of the core fuel element 40 will be described with reference to FIG. As shown in the figure, the core fuel element 40 includes a vertically extending cylindrical cladding tube 41, a plenum spring 43 housed inside the cladding tube 41, a core fuel 44, an upper blanket fuel 45 and a lower portion. It has a blanket fuel 46 and upper and lower end plugs 48 and 49 provided at both ends of the cladding tube 41 .

The core fuel 44, the upper blanket fuel 45, and the lower blanket fuel 46 are each formed in the shape of cylindrical pellets. A plurality of lower blanket fuels 46 , a plurality of core fuels 44 , and a plurality of upper blanket fuels 45 are stacked in this order from below the cladding tube 41 . The uppermost upper blanket fuel 45 is urged downward by a plenum spring 43 .

The blanket fuel assemblies that make up the radial blanket portion 22 are similar to the core fuel assemblies except that they include blanket fuel elements instead of the core fuel elements 40 .
The blanket fuel element is similar to core fuel element 40 except that instead of core fuel 44, upper blanket fuel 45 and lower blanket fuel 46, blanket fuel is provided.

If the fuel obtained by step S1-3 is core fuel, this fuel may be used for core fuel 44 in outer core section 25, may be used for core fuel 44 in inner core section 26, or both. However, it is preferably used for the core fuel 44 in the inner core section 26 .
In general, the core fuel 44 in the outer core section 25 and the core fuel 44 in the inner core section 26 differ in composition. For example, the core fuel 44 in the outer core section 25 tends to be more enriched with Pu than the core fuel 44 in the inner core section 26 .
Conventionally, when the core fuel 44 with different compositions is manufactured in a shared line in this way, the fuel with a high Pu enrichment for the outer core section 25 does not mix with the fuel with a low Pu enrichment for the inner core section 26. As a result, there was a need for cleaning work to remove the fuel remaining in the shared line, and the burden of cleaning work was heavy.
On the other hand, when the fuel obtained in step S1-3 is used as the core fuel 44 of the inner core section 26, the Pu enrichment of the outer core section 25 and the inner core section 26 is adjusted by adjusting the amount of Ln contained in the fuel. It is possible to adjust the degree of quenching. As a result, the same Pu-enriched fuel can be used as the fuel before containing Ln, which facilitates cleaning of the shared line during fuel production.

If the fuel obtained by step S1-3 is blanket fuel, this fuel may be loaded into the core fuel element 40 as upper blanket fuel 45 or lower blanket fuel 46, or may be loaded into the blanket fuel element. In terms of MA transmutation, it is preferred to load blanket fuel elements.

(Effect)
In the method having the above configuration, the DF _Ln of the second radioactive composition used for fuel production is more than 1 and less than 100, so the number of MA purification steps when obtaining the second radioactive composition is reduced, for example, three times. It can be reduced to the following extent. Therefore, it is possible to suppress the amount of waste liquid and equipment cost due to MA refining.

<Second embodiment>
As shown in FIG. 6, the HALW disposal load reduction method according to the second embodiment of the present disclosure includes:
By performing a treatment (MA separation) to separate MA and 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 coefficient of Ln is 1 or more and less than 100. Step S2-1 of obtaining a radioactive composition containing Ln;
Step S2-2 of producing fast reactor fuel using the radioactive composition obtained in step S2-1;
and a step S2-3 of loading the fuel obtained in step S2-2 into the fast reactor and transmuting the MA.
The waste liquid (containing FP) from which MA and Ln have been separated in step S2-1 is, for example, vitrified, stored, and then geologically disposed.

(Step S2-1: MA separation)
Step S2-1 is the same as step S1-1 except that the separation treatment is performed so that the DF _Ln of the radioactive composition obtained by the separation treatment is less than 100.

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

(Step S2-2: fuel production)
Step S2-2 is similar to step S1-3.

(Step S2-3: MA transmutation)
Step S2-3 is similar to step S1-4.

(Effect)
In the method having the above configuration, the DF _Ln of the radioactive composition used for fuel production is more than 1 and less than 100, and the DF _Ln is set to the target value during MA separation, so MA refining can be omitted. Therefore, it is possible to suppress the amount of waste liquid and equipment cost due to MA refining.

<Third embodiment>
As shown in FIG. 7, the HALW disposal load reduction method according to the third embodiment of the present disclosure includes:
By performing a treatment (MA separation) to separate MA and 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 coefficient of Ln is 1 or more and less than 100. Step S3-1 of obtaining a radioactive composition containing Ln;
Step S3-2 of storing the radioactive composition obtained by step S3-1;
Step S3-3 of producing fast reactor fuel using the radioactive composition stored in step S3-2;
and a step S3-4 of loading the fuel obtained in step S3-3 into the fast reactor and transmuting the 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 geologically disposed.

(Step S3-1: MA Separation)
Step S3-1 is similar to step S2-1.

(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 a solid form. Storage in the solidified form is preferable in terms of a small storage load.

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

When the radioactive composition is an extract obtained by an extraction treatment, the solidification treatment is preferably a decomposition treatment. By subjecting the extract to decomposition treatment, the organic solvent in the extract is removed and MA is oxidized. As a result, a solidified body containing MA oxide is obtained.
Decomposition treatments include, for example, distillation, thermal decomposition, and incineration. Distillation can be carried out using a known evaporation method such as a batch method or a continuous method (a plate tower or a packed tower).

When the radioactive composition is a back-extraction solution obtained by back-extraction treatment or an eluent obtained by adsorption-elution treatment, the solidification treatment may be hydrothermal treatment, or concentration of the liquid and concentration obtained A treatment for vitrifying the liquid (concentration-vitrification treatment) is preferred.
By hydrothermally treating the radioactive composition, the MA contained in the radioactive composition is oxidized and a solidified body containing MA oxide is precipitated. The solidified material is separated from the liquid medium (water) by solid-liquid separation. As a result, a solidified body containing MA oxide is obtained.
Concentration of the radioactive composition and vitrification of the concentrate can be carried out by conventional methods.

If necessary, before storing the obtained solidified body, the solidified body may be subjected to a stabilization treatment for removing part or all of the carbon, hydrogen, oxygen and nitrogen components.
The solidified material may contain organic substances such as extractants and radiation decomposition products of the extractants. If the solidified material contains organic matter, gas is generated during storage, which may impair the function of confinement of radioactive substances. Such problems can be suppressed by subjecting the solidified body to a stabilization treatment that removes some or all of the carbon, hydrogen, oxygen, and nitrogen components.
Examples of stabilization treatments include calcination and sintering.

As a method for storing the solidified material, a method known as a dry storage method for radioactive waste can be used. For example, the solidified material is stored in a plurality of canisters, these canisters are stored in casks, and stored in a storage facility. method.

The period during which the radioactive composition is stored can be set as appropriate.
Since highly exothermic MA such as Cm is not suitable for core fuel production, in the case where the radioactive composition contains highly exothermic MA, in step S3-2, the radioactive composition or its solidified body is mixed with the radioactive composition or It is preferable to store until the highly exothermic MA contained in the solidified body is sufficiently attenuated. As a result, the stored radioactive composition or its solidified body can be used for core fuel production without subjecting the highly exothermic MA to a separation process. Note that Cm contained in HALW is mainly ²⁴⁴ Cm, which has a relatively short half-life of about 18 years.

(Step S3-3: fuel production)
Step S3-3 is similar to step S2-2.
However, when the radioactive composition is solidified in step S3-2, the solidified body is dissolved and the solution is used to produce the fuel.
To dissolve the solidified body, for example, in the case of the solidified body containing MA oxide, an aqueous nitric acid solution may be added. When the solidified body is a vitrified body, the solidified body is dissolved in an acid solution or the like, and then the glass component is separated from the resulting solution.

(Step S3-4: MA transmutation)
Step S3-4 is similar to step S2-3.

(Effect)
In the method having the above configuration, the DF _Ln of the radioactive composition used for fuel production is more than 1 and less than 100, and the DF _Ln is set to the target value during MA separation, so MA refining can be omitted. Therefore, it is possible to suppress the amount of waste liquid and equipment cost due to MA refining.
In addition, since the radioactive composition is stored before fuel production, it can be used for core fuel production without separation of highly exothermic MA.

As described above, the embodiments of the present disclosure have been described, but each configuration and combination thereof in the above embodiments are examples, and additions, omissions, replacements, and other Change is possible.

<Appendix>
The HALW disposal load reduction method described in each embodiment can be grasped, for example, as follows.
(1) The method for reducing the disposal load of HALW according to the first to third embodiments described above recovers MA and Ln from HALW so that DF _Ln is 1 or more and less than 100, Steps of obtaining a composition (S1-1 and S1-2, S2-1, S3-1), and steps of producing fast reactor fuel using the radioactive composition (S1-3, S2-2, S3) -3), and a step of loading the fuel into the fast reactor and transmuting the MA (S1-4, S2-3, S3-4).

Conventionally, when transmuting MA in a fast reactor, a high refinement of DF _Ln =100 or more was required. MA refining to achieve such a high degree of purification results in an increase in the amount of waste liquid and facility costs.
According to the studies of the present inventors, it has been found that MA nuclear transmutation can be performed without problems even with fuel produced using a radioactive composition having a DF _Ln of less than 100. Therefore, DF _Ln is set to less than 100 in the method according to the above embodiment. If DF _Ln is less than 100, it is possible not to perform MA refining, or to reduce the number of steps even if MA refining is performed. Therefore, it is possible to suppress the amount of waste liquid and equipment cost due to MA refining.

(2) In the methods according to the first to third embodiments, the fuel is preferably core fuel. In this case, the decontamination coefficient of Ln is preferably 1 or more and less than 20.
According to the above configuration, the recovered MA can be transmuted while the core fuel is accompanied by Ln.

(3) In the methods according to the first to third embodiments, the fuel is preferably blanket fuel.
According to the above configuration, the recovered MA can be transmuted while the blanket fuel is accompanied by Ln.

(4) In the methods according to the first to third embodiments, when the fuel is blanket fuel, the step of producing the fuel (S3-3 ), it is preferable to have a step (S3-2) of storing the radioactive composition.
The spent fuel MA contains highly exothermic nuclides (Cm ²⁴³ , Cm ^244, etc.). If MA containing such nuclides is used to produce blanket fuel, the calorific value may become too high, so conventionally, nuclides with high calorific value have been separated during MA refining.
According to the above configuration, the decay heat of the highly exothermic nuclides can be reduced, so that after the MA separation from the HALW, the MA fuel can be produced without performing the separation treatment of the highly exothermic nuclides.

(5) In the methods according to the first to third embodiments, when the fuel is blanket fuel, it is also preferable that the spent nuclear fuel is spent nuclear fuel that has been interimly stored during reprocessing.
The spent fuel MA contains highly exothermic nuclides (Cm ²⁴³ , Cm ^244, etc.). If such MA is used to produce blanket fuel, the calorific value may become too high, so conventionally, highly exothermic nuclides were separated during MA refining.
According to the above configuration, since the decay heat of the highly exothermic nuclides is reduced by the intermediate storage, after the MA separation from the HALW, it can be used for blanket fuel production without performing the separation treatment of the highly exothermic nuclides. can be done.

The fast reactor fuel described in each embodiment is grasped, for example, as follows.
(6) Fast reactor fuels according to the first and second embodiments contain MA and Ln.

Conventionally, when transmuting MA in a fast reactor, a high degree of refinement was required so as not to mix Ln into the MA fuel.
Since the fuel according to the above embodiment contains Ln, the decontamination coefficient 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 refining can be suppressed.

(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% by mass with respect to the total amount of loaded core fuel.
According to the above configuration, the recovered MA can be transmuted while the core fuel is accompanied by Ln.

(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% by mass with respect to the total amount of loaded blanket fuel.
According to the above configuration, the recovered MA can be transmuted while the blanket fuel is accompanied by Ln.

100 Fast reactor 1 Core 2 Reactor vessel 3 Guard vessel 4 Coolant inlet pipe 5 Coolant outlet pipe 6 Reactor vessel pedestal 7 Core upper mechanism 8 Fixed plug 9 Control rod drive mechanism 10 Rotating plug 11 Rotating plug drive device 12 Inside the reactor Structure 21 Neutron Shield 22 Radial Blanket 23 Control Rod 24 Neutron Source 25 Outer Core 26 Inner Core 30 Core Fuel Assembly 31 Bugle 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 plug 49 Lower plug Ac Axis

Claims (11)

A minor actinide and a lanthanide are recovered from high-level radioactive waste, which is a waste liquid obtained by separating U and Pu from a solution of spent nuclear fuel, so that the decontamination coefficient of the lanthanide is 1 or more and less than 100, and the minor actinide and the obtaining a radioactive composition comprising a lanthanide;
using the radioactive composition to produce fuel for a fast reactor;
and loading the fuel into a fast reactor to transmute the minor actinide. 2. The method of claim 1, wherein the fuel is core fuel. 3. The reduction method according to claim 2, wherein the lanthanoid has a decontamination factor of 1 or more and less than 20. 2. The method of claim 1, wherein the fuel is blanket fuel. 5. The method of claim 4, comprising storing the radioactive composition prior to manufacturing the fuel. 6. A reduction method according to claim 4 or 5, wherein said spent nuclear fuel is spent nuclear fuel intermediately stored in reprocessing. Fast reactor fuel containing minor actinides and lanthanides. 8. The fuel of claim 7, which is core fuel. 9. The fuel according to claim 8, wherein the lanthanide content is 1 to 13% by mass with respect to the total amount of loaded core fuel. 8. The fuel of claim 7, which is blanket fuel. 11. The fuel according to claim 10, wherein the lanthanoid content is 31-33% by mass with respect to the total amount of blanket fuel loaded.

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