JP2024049151A - High-level radioactive material processing system and high-level radioactive material processing method - Google Patents

Abstract

To enable solidifying a minor actinoid in a more stable state.SOLUTION: A high-level radioactive substance processing system includes: a uranium supply unit that supplies uranium to a liquid extracting a radioactive substance including at least one of a minor actinoid and lanthanoid; a solidification processing unit that, with the uranium supplied, heats the liquid extracting the minor actinoid and lanthanoid, and evaporates a part of the liquid, thereafter, further heats the liquid to prepare a solidified solidification body; and a stabilization processing unit that sinters the prepared solidification body, eliminates carbon, hydrogen, oxygen, and a part or all of a nitrogen component, and creates a fluorite structure of uranium dioxide taking the minor actinoid and lanthanoid inside. In the solidification processing unit, when a molar quantity of the solidification body is 1, a molar quantity X of the radioactive substance is 0≤X≤7.SELECTED DRAWING: Figure 4

Description

The present disclosure relates to a high-level radioactive material processing system and a high-level radioactive material processing method.

As a method for treating high-level radioactive waste, there is a method for extracting minor actinides, which are radioactive substances, from high-level radioactive materials (e.g., Patent Document 1 and Patent Document 2). In addition, there is a solidification method as a treatment method for storing spent fuel in a recyclable form (e.g., Patent Document 3).

JP 2018-63198 A JP 2019-15533 A Japanese Patent No. 6037168

By extracting minor actinides from high-level radioactive materials, the extracted minor actinides can be reused as fuel for fast breeder reactors and other reactors. In addition, removing the minor actinides can reduce the waste treatment burden. Here, in order to reuse minor actinides as fuel, nuclear transmutation technology for minor actinides is required. Furthermore, in situations where recycling technology has not yet been put to practical use, the separated minor actinides must be stored stably for long periods of time.

The present disclosure has been made in consideration of the above, and aims to provide a high-level radioactive material processing system and a high-level radioactive material processing method that can convert minor actinides into a solid in a more stable state.

In order to solve the above-mentioned problems and achieve the object, the present disclosure provides a high-level radioactive material treatment device, which includes a uranium supply unit that supplies uranium to a liquid from which radioactive material containing at least one of minor actinides and lanthanides has been extracted, a solidification processing unit that heats the liquid from which uranium has been supplied and from which the minor actinides and lanthanides have been extracted to evaporate some of the liquid, and then further heats the liquid to produce a solidified body, and a stabilization processing unit that sinters the solidified body produced, removes some or all of the carbon, hydrogen, oxygen, and nitrogen components, and produces a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanides therein, and the stabilization processing unit provides a solidified body in which the molar amount X of the radioactive material, when the molar amount of uranium is 1, is 0<X≦7.

In order to solve the above-mentioned problems and achieve the object, the present disclosure provides a high-level radioactive material processing method, which includes the steps of: supplying uranium to a liquid from which radioactive material containing at least one of minor actinides and lanthanides has been extracted, at a ratio such that the molar amount X of the radioactive material is 0<X≦7, where the molar amount of uranium is 1; heating the liquid from which the uranium has been supplied and from which the minor actinides and lanthanides have been extracted to evaporate some of the liquid, and then further heating the liquid to produce a solidified body; and performing, in a single device, a stabilization process to sinter the solidified body, remove some or all of the carbon, hydrogen, oxygen, and nitrogen components, and produce a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanides.

This disclosure makes it possible to convert minor actinides into more stable solids.

FIG. 1 is a schematic diagram showing a schematic configuration of a high-level radioactive material treatment apparatus according to this embodiment. FIG. 2 is a schematic diagram showing a schematic configuration of a high-level radioactive material treatment apparatus according to another embodiment. FIG. 3 is an explanatory diagram showing a ternary phase diagram at 1523 K (1250° C.). FIG. 4 is an explanatory diagram showing a ternary phase diagram at 1273 K (1000° C.). FIG. 5 is an explanatory diagram showing a ternary phase diagram at 1073 K (800° C.). FIG. 6 is an explanatory diagram showing a ternary phase diagram at 923K (650° C.). FIG. 7 is an explanatory diagram showing a ternary phase diagram at 773K (500° C.). FIG. 8 is an explanatory diagram showing a ternary phase diagram at 623 K (350° C.).

Below, an embodiment of the high-level radioactive material treatment device according to the present disclosure will be described in detail with reference to the drawings. The high-level radioactive material treatment device according to the present disclosure extracts MA (minor actinides) from high-level radioactive material and vitrifies and stabilizes the waste from which MA has been extracted. In addition, when lanthanides are contained in the high-level radioactive material, the high-level radioactive material treatment device extracts the lanthanides as well as the minor actinides. In this disclosure, "MA (minor actinides)" refers to transuranium elements belonging to the actinides, excluding Pu. "Actinides" is a general term for elements with atomic numbers from 89 to 103. Minor actinides include, for example, Np (neptunium), Am (americium), and Cm (curium). "Ln (lanthanoid)" is a general term for elements with atomic numbers from 57 to 71.

Figure 1 is a schematic diagram showing the outline of the high-level radioactive material treatment device of this embodiment. The high-level radioactive material treatment device (treatment device) 10 shown in Figure 1 includes an extraction device 12, a solidification device 14, a stabilization device 16, and a storage device 18. The treatment device of this embodiment will be described as using high-level radioactive waste (hereinafter also referred to as "HALW") as the high-level radioactive material. The waste liquid is, for example, a liquid obtained after U (uranium) and Pu (plutonium) are recovered from the solution of spent nuclear fuel discharged from a light water reactor in the reprocessing of spent nuclear fuel. The high-level radioactive waste contained in the waste liquid includes fission products (hereinafter also referred to as "FP"), MA, and lanthanides. Specifically, the high-level radioactive waste (hereinafter also referred to as "HALW") is waste liquid generated by reprocessing using the PUREX method. In the Purex process, a nitric acid solution containing U and Pu 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. Meanwhile, FP, MA, and Ln remain in the nitric acid solution (waste liquid). The nitric acid solution containing FP, MA, and Ln becomes the waste liquid to be treated.

The extraction device 12 extracts MA components from the waste liquid. The extraction device 12 includes a waste liquid supply unit 22, an extractant supply unit 24, a diluent supply unit 26, and an MA extract liquid production unit 28. The waste liquid supply unit 22 stores HALW, which is liquid high-level radioactive waste, and supplies it to the MA extract liquid production unit 28. In this embodiment, the extractant supply unit 24 and the diluent supply unit 26 are provided, but the extractant supply unit 24 and the diluent supply unit 26 may be combined into a single device to supply a liquid phase organic solvent in which the extractant is dissolved.

The extractant supply unit 24 supplies the extractant to the MA extractant generating unit 28. The extractant captures MA and Ln. The extractant is a liquid that migrates to the dilution liquid and mixes uniformly. For example, a complexing agent that forms a complex with MA or Ln can be used as the extractant. The complexing agent is preferably inexpensive and selectively forms a complex with MA. Examples of the extractant 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), and N,N'-dibutyl-N,N'-dimethyltetradecylmalonamide (DMDBTDMA). It is preferable to use a DGA-based material (complexing agent) as the extractant. Specific examples of complexing agents include N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA), tetra(2-ethylhexyl)diglycolamide (T2EHDGA), etc. Extractants may be used alone or in combination of two or more.

The dilution liquid supply unit 26 supplies the dilution liquid to the MA extraction liquid generation unit 28. The dilution liquid is an organic phase material (organic solvent) that is insoluble in the liquid components of the waste liquid. 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. The dilution liquid is an organic phase material that is insoluble in the liquid components of the waste liquid. The dilution liquid has a property of eluting the extractant from the liquid of the waste liquid. The dilution liquid preferably has a boiling point of 30°C or more and 100°C or less. In addition, it is preferable that the dilution liquid does not have a flash point or has a flash point equal to or higher than the decomposition temperature of the extractant. By using a liquid that satisfies the above range as the dilution liquid, the processing load in the solidification processing unit can be reduced. In addition, it is preferable that the dilution liquid is a material that is less deteriorated by heat or radiation even when reused, and it is preferable that the dilution liquid is a solvent having a hydrocarbon structure that is less deteriorated by heat or radiation even when reused. Here, the phrase "there is little deterioration due to heat or radiation even when reused" means that even if the material is exposed to the energy of heat or radiation and becomes a low molecular weight structure, it has the property of being difficult to form a complex with metal ions. As the diluent, for example, a hydrofluorocarbon (HFC), which is a fluorine-based material, is preferably used. The organic solvent may be used alone or in combination of two or more types. In addition, the diluent is preferably made into a reusable liquid after being separated from the MA in the solidification device 14.

The MA extract generating unit 28 is supplied with waste liquid, an extractant, and a diluent. When the MA and Ln in the waste liquid come into contact with the extractant by the solvent extraction method, the MA and Ln migrate to the extractant side. The extractant that has captured MA and Ln is encapsulated in the diluent. After the extraction process, the MA extract generating unit 28 separates the diluent from the waste liquid to generate an MA extract in which the extractant that has captured MA and Ln is encapsulated in the diluent. The MA extract generating unit 28 may be a continuous type in which each material is continuously supplied and an MA extract is generated, or a batch type in which each material is intermittently supplied and an MA extract is generated. The MA extract generating unit 28 may selectively capture MA with the extractant, or may simultaneously capture and process MA and Ln.

The solidification device 14 mixes uranium with the MA extract, removes the liquid components of the mixed liquid, and produces a solidified body. The solidification device 14 includes a solidification processing unit 40 and a uranium supply unit 44. The uranium supply unit 44 supplies uranium to the MA extract in the solidification processing unit 40. The uranium supply unit 44 supplies a uranium-containing substance that is uniformly soluble in the diluent (organic solvent) that constitutes the MA extract, such as an organic U complex, specifically a uranium complex of dipivaloylmethane. The uranium in the organic U complex is contained in a valence state of tetravalent or hexavalent.

The solidification treatment section 40 includes a distillation section 45, a heat treatment section 46, a dilution liquid recovery section 47, an oxidation gas supply section 48, and an exhaust gas recovery section 49. The distillation section 45 heats the MA extract mixed with uranium to, for example, 50°C to 100°C to evaporate the dilution liquid. The heat treatment section 46 is an apparatus integrated with the distillation section 45, and further heats the object from which the dilution liquid has been evaporated, for example to 300°C, to thermally decompose the residue and convert it to oxide. The dilution liquid recovery section 47 is connected to the distillation section 45 and recovers the dilution liquid evaporated in the dilution section 45. The dilution liquid recovery section 47 supplies the recovered dilution liquid to the dilution liquid supply section 26 for reuse. The oxidation gas supply section 48 supplies an oxidizing gas, for example, oxygen or air, to the heat treatment section 46. The exhaust gas supply section 49 recovers the exhaust gas discharged from the heat treatment section 46.

The solidification processing unit 40 distills the MA extract mixed with uranium in the distillation unit 45 to remove the dilution liquid components. By removing the dilution liquid, the solidification processing unit 40 leaves a solidified residue containing the extractant, MA, Ln, and uranium as a complex. The solidification processing unit 40 also recovers the gas discharged during distillation in the distillation unit 45 in the dilution liquid recovery unit 47. The solidification processing unit 40 heats the object obtained by distilling the dilution liquid in the heat treatment unit 46 while supplying an oxidizing gas from the oxidizing gas supply unit 48, thereby removing the components of the extractant, specifically, C, H, O, and N, and generating a solidified body. The solidified body becomes an oxide. The solidification processing unit 40 can use known evaporation methods such as a batch type (batch type) and a continuous type (plate tower or packed tower) for distillation and heat treatment. It is preferable that the distillation unit 45 sets the processing temperature to a boiling point of the dilution liquid of the MA extract liquid at a temperature between -10°C and +20°C, and below the flash point, based on the boiling point of the dilution liquid. The heat treatment unit 46 is at a temperature at which thermal decomposition and oxide conversion processes occur, for example, 300°C or higher.

The stabilization device 16 performs a stabilization process to remove some or all of the carbon, hydrogen, oxygen, and nitrogen components from the solidified body. The stabilization device 16 has a stabilization processing section 50.

The stabilization treatment section 50 includes a reduction section 52, a reducing gas supply section 54, and an exhaust gas recovery section 56. The reduction section 52 heats the solidified body in a reducing atmosphere, calcines it, and sinters it. The reduction section 52 is an apparatus integrated with the distillation section 45 and the heat treatment section 46. The reducing gas supply section 54 supplies a reducing gas, for example, hydrogen, to the reduction section 52. The exhaust gas recovery section 56 recovers the exhaust gas discharged from the reduction section 52. The stabilization treatment section 50 also generates _UO2 form while suppressing the generation of _UO3 form by adjusting the temperature and performing hydrogen reduction during treatment. The stabilization treatment section 50 can also generate _UO2 form while suppressing the generation of _UO3 form by supplying moisture during treatment and performing hydration reduction.

The stabilization treatment unit 50 supplies reducing gas from the reducing gas supply unit 54 to the reduction unit 52 while heating, calcining, and sintering the solidified body in the reduction unit 52 to remove some or all of the carbon, hydrogen, oxygen, and nitrogen components from the treatment target and solidify it into a fluorite structure. Specifically, the stabilization treatment unit 50 sinters the solidified body to perform denitration and pyrolysis treatment, remove impurity components (organic components, nitric acid components, moisture) in the solidified body, i.e., CHON components, and then sinter it. The sintering temperature is, for example, 600°C to 1250°C, preferably 600°C to 800°C.

The stabilization treatment unit 50 removes impurity components from the solidified body containing uranium, and then sinters the solidified body in a reducing atmosphere, thereby obtaining uranium oxide having a fluorite structure in which MA and Ln are dissolved. As an example, by distillation, a solidified matter of [M3 ⁺ .( _NO3- ⁾ _3.m (extractant)] + ([ ^U4+ .( _NO3- ⁾ _4.n (ligand)] or [M3 ⁺ .( _NO3- ⁾ _3.m (extractant)] + ([ _UO22 ⁺ .( _NO3- ⁾ _2.n (ligand)] is generated. Here, M is MA or Ln. By heat treating this solidified matter and removing the CHON component, it becomes M _a O _b + U ₃ O ₈ + ( _αCO2 + _βH2O + γNO _x ), and ( _αCO2 + _βH2O + γNO _x ) is removed. Furthermore, by sintering M _a O _b + U ₃ O ₈ , an oxide with a fluorite structure, UM _RM O _y , is formed. RM is M(Ln or MA)/U, which is the equivalent ratio of Ln or MA to U. The stabilization device 16 removes some or all of the carbon, hydrogen, oxygen, and nitrogen components, which are components that may gasify during storage, and turns the uranium dioxide into a fluorite structure containing MA and Ln, thereby producing a solidified material with high thermal and chemical stability.

The storage device 18 stores the stabilized treated material. The storage device 18 includes a storage section 60. The storage section 60 stores the stabilized MA-containing material in a solid state. The storage section 60 is, for example, a metal container or cask. The storage section 60 need only be capable of removing decay heat from the stabilized MA-containing material. The storage section 60 may also be provided underground, and the stabilized treated material may be buried underground.

The MA-containing material stored in the storage section 60 of the processing system 10 can be used as fuel for a nuclear power generation system. When nuclear transmuting MA as fuel, the stored solidified body with a fluorite structure is melted to produce a solution containing U, MA, and Ln (solidified body melting). Next, the obtained solution is subjected to a purification process for U and MA (MA purification). In MA purification, for example, U and MA in the obtained solution are separated from Ln, and highly exothermic MA is separated from MA as necessary. Next, the obtained U and MA (Np, Am, etc.) are mixed with U and Pu to obtain a mixed oxide, thereby producing fuel (MA fuel production). The obtained fuel is burned in a fast breeder reactor or the like (MA combustion).

In the processing device 10 of this embodiment, an extractant and a diluent are added to the waste liquid in the extraction device 12, and then mixed to perform a separation process to separate MA from the waste liquid together with Ln, thereby producing an MA extract. Next, in the solidification device 14, the processing device 10 distills the diluent from the MA extract and supplies U to produce a solidified body containing U, MA, and Ln. Next, in the stabilization device 16, the processing device 10 forms a fluorite structure by removing organic components (CHON) from the solidified body containing MA. The processing device 10 stores the processed product with the fluorite structure in the storage device 18.

It is preferable that the processing device 10 performs the heat treatment of the solidification stabilization device 13 in a single device. This allows the device configuration to be simplified. Also, since the processing device 10 can perform multiple processes in a single device, the number of times radioactive materials need to be transported can be reduced, making the processing safer.

In addition, by using a diluent with a boiling point of 30°C or higher and 100°C or lower, or with no flash point or with a flash point higher than the decomposition temperature of the extractant, the diluent can be easily recovered by solidification treatment and the required energy can be reduced.

The processing device 10 separates, solidifies, and stabilizes MA along with Ln. This allows MA to be efficiently extracted from liquid high-level radioactive waste, reducing the waste disposal burden and solidifying it. Furthermore, by extracting MA and Ln and turning them into a stabilized substance, the MA can be stored stably even when it is stored for reuse as fuel. Furthermore, when it is to be reused, it is possible to make it reusable by simply dissolving the solidified body.

In addition, by processing MA and Ln without separating them, the process of separating MA and Ln is unnecessary, and the processing load can be reduced.

In addition, when producing an MA extract from highly radioactive materials containing nitric acid-based substances, the processing device 10 preferably uses an extractant that does not form a third phase. Specifically, it is preferable to use a DGA-based material. One example is T2EHDGA. In addition, it is preferable for the extractant to be a material that can be removed during processing in the stabilization device 16, specifically a material composed only of carbon, hydrogen, nitrogen, and oxygen. In addition, it is preferable for the extractant to not contain any corrosive components.

In addition, the processing device 10 supplies uranium to the liquid containing MA and Ln in the uranium supply unit 44, generates a solidified body, and then sinters it in the stabilization device 16, thereby forming a fluorite structure of uranium dioxide in which MA and Ln are fixed. This allows the solid to have a long-term stable crystal structure after stabilization processing, and MA can be stored stably. Specifically, trivalent MA and Ln can be dissolved in the fluorite structure, and a stable fluorite structure can be obtained compared to a mixture of MA and Ln or an oxide structure of only one of them. Furthermore, by forming a crystal structure using uranium, when a solid containing MA is reprocessed and used, it is possible to obtain a structure that is free of unnecessary components such as Si, Na, Ca, etc. This allows the substance to be stabilized in a state that is easy to use when reprocessing.

The uranium supply unit 44 preferably supplies uranium in the form of an organic substance complex. This makes it easier to dissolve the uranium in the diluent. Here, it is preferable to use a ligand that is coordinated to the organic U complex and that is composed only of carbon, hydrogen, nitrogen, and oxygen. In addition, it is preferable that the ligand that is coordinated to the organic U complex is easy to stabilize (thermally decompose) and does not contain corrosive components, etc. This allows it to be suitably removed during the stabilization process, and the MA-containing substance can be made into a stabilized substance, which can be stored stably.

In the above embodiment, the extraction device 12 produces an MA extract by extraction with an organic solvent and performs a solidification process, but this is not limited to the above. The processing device may produce a stripped liquid in which MA is extracted into a liquid phase, solidify it, and stabilize it.

Figure 2 is a schematic diagram showing the general configuration of a high-level radioactive material treatment device of another embodiment. The high-level radioactive material treatment device (treatment device) 10A shown in Figure 2 includes an extraction device 12A, a solidification device 14A, a stabilization device 16A, and a storage device 18. The treatment device 10A of this embodiment will be described as a case in which high-level radioactive waste (hereinafter also referred to as "HALW") is used as the high-level radioactive material, similar to the treatment device 10.

The extraction device 12A extracts MA components from the waste liquid. The extraction device 12A includes a waste liquid supply unit 22, an extractant supply unit 24, a diluent supply unit 26, an MA extract liquid production unit 28, an MA stripping liquid production unit 30, a stripping agent supply unit 32, and a diluent supply unit 34. The waste liquid supply unit 22 stores HALW, which is a liquid high-level radioactive waste, and supplies it to the MA extract liquid production unit 28.

The extractant supply unit 24 supplies the extractant to the MA extractant generating unit 28. The extractant captures MA and Ln. The extractant is a liquid that transfers to the dilution liquid. For example, a complexing agent that forms a complex with MA or Ln can be used as the extractant. The complexing agent is preferably less expensive than the complexing agent that selectively forms a complex with MA. Examples of the extractant 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), and N,N'-dibutyl-N,N'-dimethyltetradecylmalonamide (DMDBTDMA). It is preferable to use a DGA-based material (complexing agent) as the extractant. Specific examples of complexing agents include N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA), tetra(2-ethylhexyl)diglycolamide (T2EHDGA), etc. Extractants may be used alone or in combination of two or more.

The diluent supply unit 26 supplies the diluent to the MA extract production unit 28. The diluent is an organic phase material (organic solvent) that is insoluble in the liquid components of the waste liquid. 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 an organic solvent is n-dodecane. One type of organic solvent may be used alone, or two or more types may be used in combination. In addition, it is preferable that the diluent is separated from MA in the MA stripping production unit 30 and then made into a reusable liquid.

The MA extract generating unit 28 is supplied with waste liquid, an extractant, and a diluent. When the MA and Ln in the waste liquid come into contact with the extractant by the solvent extraction method, the MA and Ln migrate to the extractant side. The extractant also migrates to the diluent side. The MA extract generating unit 28 separates the diluent from the waste liquid after the extraction process to generate an MA extract, which is an extractant that has captured MA and Ln. The MA extract generating unit 28 may be a continuous type in which each material is continuously supplied and an MA extract is generated, or a batch type in which each material is intermittently supplied and an MA extract is generated. In this embodiment, the extractant supply unit 24 and the diluent supply unit 26 are provided, but the extractant supply unit 24 and the diluent supply unit 26 may be combined into one device to supply an organic solvent in a liquid phase in which the extractant is dissolved.

The MA back-extraction liquid generation unit 30 is supplied with the MA extract from the MA extract generation unit 28 and with the back-extraction agent from the back-extraction agent supply unit 32.

The back extraction agent supply unit 32 supplies a substance that transfers MA and Ln from the MA extract to the liquid phase as a back extraction agent. The back extraction agent is, for example, nitric acid. The back extraction agent is diluted with a diluent of the liquid phase, for example, water. The hydrothermal treatment promoter supply unit 34 supplies the hydrothermal treatment promoter to the MA back extraction liquid production unit 30. The hydrothermal treatment promoter is a neutral to basic liquid and is a reducing agent composed of CNOH, for example, an aldehyde compound, an amine compound, etc. Here, the preparation of the mixture of the MA extract, the back extraction agent, and the hydrothermal treatment promoter is not limited to this embodiment. The hydrothermal treatment promoter supply unit 34 may supply the back extraction agent to the MA extract and supply the hydrothermal treatment promoter to the MA back extraction liquid.

The MA stripping liquid generator 30 brings the MA extract into contact with a diluent containing a stripping agent, and transfers the MA and Ln contained in the organic phase MA extract to the diluent containing a stripping agent. The extraction device 12A may reuse the organic solvent (extractant and diluent) in the MA extract after processing. The MA stripping liquid generator 30 controls the acid concentration of the liquid and the amount of stripping agent to set the component ratio during processing in the solidification process 14A to a predetermined ratio.

The solidification device 14A supplies uranium to the MA stripping liquid, removes the liquid components of the mixed liquid, and produces a solidified body. The solidification device 14A includes a solidification processing unit 40A and a uranium supply unit 44A. The uranium supply unit 44A supplies uranium to the MA stripping liquid of the solidification processing unit 40A. The uranium supply unit 44A supplies a uranium-containing material that is homogeneously soluble in the liquid phase solvent that constitutes the MA stripping liquid. Uranium dissolves in the liquid phase in a valence state of tetravalent or hexavalent.

The solidification treatment section 40A includes a hydrothermal treatment section 45A, a heat treatment section 46A, an exhaust gas recovery section 47A, an oxidizing gas supply section 48A, and an exhaust gas recovery section 49A. The hydrothermal treatment section 45A heats the MA stripping liquid mixed with uranium to, for example, 80°C to 150°C, causing hydrothermal treatment and oxide conversion treatment, and evaporating the liquid components. The liquid components also include a hydrothermal treatment promoter. The heat treatment section 46A is an apparatus integrated with the hydrothermal treatment section 45A, and further heats the object from which the liquid has been evaporated, for example to 300°C, to thermally decompose the residue. The exhaust gas recovery section 47A is connected to the hydrothermal treatment section 45A and recovers the exhaust gas evaporated by the hydrothermal treatment section 45A. The oxidizing gas supply section 48A supplies an oxidizing gas, for example oxygen or air, to the heat treatment section 46A. The exhaust gas supply section 49A recovers the exhaust gas discharged from the heat treatment section 46A.

The solidification treatment section 40A performs hydrothermal synthesis of the MA stripping solution mixed with uranium in the hydrothermal treatment section 45A to form a solidified material containing MA and uranium. The liquid component is removed by solid-liquid separation of the solidified body and the dilution liquid component. In addition, the hydrothermal reaction can be promoted by adding a hydrothermal treatment promoter. The solidification treatment section 40A heats the object from which the liquid has been removed and which has been oxidized while supplying an oxidizing gas from the oxidizing gas supply section 48A in the heat treatment section 46A, thereby removing the components of the extractant, specifically, C, N, O, H, and generating a solidified body. The solidified body contains U, MA, and Ln. Specifically, the solidified body contains _UO2 and an M oxide precursor. The M oxide precursor is MOOH, M(OH) ₃ , etc. M is at least one of MA and Ln. The hydrothermal treatment treatment section 45A is exemplified by a treatment temperature of 80°C or more and 150°C or less, and a treatment time of 0.5h or more and 20h or less. The heat treatment section 46 is heated to a temperature equal to or higher than the temperature at which pyrolysis occurs, for example, 300° C. or higher.

The stabilization device 16A performs a stabilization process to remove some or all of the carbon, hydrogen, oxygen, and nitrogen components from the solidified body and generate a solidified body with a fluorite structure. The stabilization device 16A has a stabilization processing unit 50A.

The stabilization treatment section 50A includes a reduction section 52A, a reduction gas supply section 54A, and an exhaust gas recovery section 56A. The reduction section 52A heats the solidified body in a reducing atmosphere and calcines and sinters it. The reduction section 52A is an integrated device with the distillation section 45A and the heat treatment section 46A. The reduction gas supply section 54A supplies a reducing gas, such as hydrogen, to the reduction section 52A. The exhaust gas recovery section 56A recovers the exhaust gas discharged from the reduction section 52.

The stabilization treatment unit 50A supplies reducing gas from the reducing gas supply unit 54 to the reduction unit 52A while heating, calcining, and sintering the solidified body in the reduction unit 52A, solidifying the object to be treated into a fluorite structure. Specifically, the stabilization treatment unit 50A sinters the solidified body to perform denitration and pyrolysis treatment, remove impurity components (organic components, nitric acid components, moisture) in the solidified body, i.e., CHON components, and then sinters it. The sintering temperature is, for example, 600°C or higher and 800°C or lower.

The stabilization processing section 50A removes impurity components from the solidified body containing uranium, and then sinters it to obtain uranium oxide with a fluorite structure containing MA and Ln in solid solution. The stabilization device 16A removes some or all of the carbon, hydrogen, oxygen, and nitrogen components that may gasify during storage, and creates a fluorite structure of uranium dioxide containing MA and Ln, resulting in a solidified body with high thermal and chemical stability.

The storage device 18 stores the stabilized treated material. The storage device 18 includes a storage section 60. The storage section 60 stores the stabilized MA-containing material in a solid state. The storage section 60 is, for example, a cask. The storage section 60 may also be provided underground, and the stabilized treated material may be buried underground.

In the processing device 10A of this embodiment, extraction and back-extraction are performed in the extraction device 12A, a solution containing MA in the liquid phase is processed, uranium is supplied, and a fluorite structure of uranium dioxide with MA and Ln fixed therein is formed in the stabilization device 16A.

It is preferable that the processing device 10A performs the heat treatment of the solidification stabilization device 13 in a single device. This allows the device configuration to be simplified. In addition, since the processing device 10A can perform multiple processes in a single device, the transportation of radioactive materials can be reduced, making processing safer.

In addition, even when performing back extraction, by forming a fluorite structure, it is possible to obtain a long-term stable crystal structure as a solid after stabilization processing, and MA can be stored stably. Specifically, trivalent MA and Ln can be dissolved in the fluorite structure, making the stable fluorite structure even more stable. In addition, by forming a crystal structure using uranium, it is possible to obtain a structure that is free of unnecessary components such as Si, Na, Ca, etc., when the solid containing MA is reprocessed and used. This makes it possible to obtain a substance that is stabilized in a state that is easy to use when reprocessing. Furthermore, the processing device 10A can obtain various effects similar to those of the processing device 10.

The uranium supply unit 44 of the processing device 10, 10A preferably supplies uranium in an amount equal to or greater than the total molar amount of minor actinides and lanthanides contained in the liquid extracted by the extraction device. This allows for a more stable structure than the fluorite structure of the solid material during storage.

As described above, the processing apparatus 10, 10A allows MA to have a long-term stable crystal structure, preferably a fluorite structure, thereby enabling MA to be stored stably for a long period of time. The MA-containing substance produced by the processing apparatus 10, 10A can be made more stable by setting the conditions during stabilization processing and the conditions during storage to specified conditions. Specifically, a more stable state can be achieved by setting the molar ratio of MA to U and O (oxygen) to a specified balance. The molar ratio of MA to U and O (oxygen) can be set to a specified range during stabilization processing by adjusting the amount of U added to MA and the atmosphere during processing.

Next, the preferred conditions for the processing method in the processing apparatus 10 will be described with reference to Figures 3 to 8. Figure 3 is an explanatory diagram showing a ternary phase diagram at 1523 K (1250°C). Figure 4 is an explanatory diagram showing a ternary phase diagram at 1273 K (1000°C). Figure 5 is an explanatory diagram showing a ternary phase diagram at 1073 K (800°C). Figure 6 is an explanatory diagram showing a ternary phase diagram at 923 K (650°C). Figure 7 is an explanatory diagram showing a ternary phase diagram at 773 K (500°C). Figure 8 is an explanatory diagram showing a ternary phase diagram at 623 K (350°C).

Figures 3 to 8 show the crystal structures formed at one atmosphere (1 atm) and at various temperatures with the molar ratios of the three elements oxygen (O), neodymium (Nd), and uranium (U). Niobium is a simulated MA that simulates a minor actinide. In other words, the crystal structures shown in Figures 3 to 8 can be produced by processing with the same amount of MA as Nd. In Figures 3 to 8, the atmospheric conditions were adjusted by controlling the supply of inert gas containing a trace amount of hydrogen.

3, when the temperature is 1523 K, i.e., 1250° C., a crystal including a fluorite structure can be produced by using the molar ratio of region 104A. Specifically, a substance in which the fluorite structure is mixed with U ₃ O ₈ , rhombohedrons, etc. is generated in region 104A. Also, when the temperature is 1523 K, i.e., 1250° C., a stabilized crystal structure can be produced by using region 104A. Specifically, a substance in which the fluorite structure is mixed with elemental uranium and rhombohedrons is generated in region 104A.

As shown in FIG. 3, when the temperature is 1523 K, that is, 1250° C., it is preferable to set the conditions in region 104A. As an example, it is preferable that the molar amount X of the radioactive material, when the molar amount of uranium is 1, is 0<X≦7. Also, it is preferable that the molar amount Y of oxygen (O), when the molar amount of uranium is 1, is 1.9≦Y≦12.

As shown in FIG. 4, when the temperature is 1273 K, i.e., 1000° C., a material having only the fluorite structure can be produced by using the molar ratio of region 100B. All of the crystals contained in region 100B have the fluorite structure. Also, as shown in FIG. 4, when the temperature is 1273 K, i.e., 1000° C., a material having a stabilized fluorite structure can be produced by using region 104B.

As shown in FIG. 4, when the temperature is 1273 K, i.e., 1000° C., it is preferable to use the conditions of region 104B, and more preferably the conditions of region 100B. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive material is preferably 0<X≦7, and more preferably 0.08≦X≦0.5. Also, when the molar amount of uranium is 1, the molar amount Y of oxygen is preferably 1.9≦Y≦12, and more preferably 2.2≦Y≦3.2.

As shown in FIG. 5, when the temperature is 1073 K, i.e., 800° C., a material having only a fluorite structure can be produced by using the molar ratio in region 100C, and a material having a stabilized fluorite structure can be produced by using the molar ratio in region 104C.

As shown in FIG. 5, when the temperature is 1073K, i.e., 800°C, it is preferable to use the conditions of region 104C, and more preferably the conditions of region 100C. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive material is preferably 0<X≦7, and more preferably 0<X≦0.6. Also, when the molar amount of uranium is 1, the molar amount Y of oxygen is preferably 1.9≦Y≦12, and more preferably 2.0≦Y≦3.2.

As shown in FIG. 6, when the temperature is 923 K, i.e., 650° C., a material having only a fluorite structure can be produced by using the molar ratio in region 100D, and a material having a stabilized fluorite structure can be produced by using the molar ratio in region 104D.

As shown in FIG. 6, when the temperature is 923 K, i.e., 650° C., it is preferable to use the conditions in region 104D, and more preferably the conditions in region 100C. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive material is preferably 0<X≦6.3, and more preferably 0.05≦X≦0.5. Also, when the molar amount of uranium is 1, the molar amount Y of oxygen is preferably 1.9≦Y≦12, and more preferably 2.2≦Y≦3.2.

As shown in FIG. 7, when the temperature is 773 K, i.e., 500° C., a material having only a fluorite structure can be produced by using the molar ratio in region 100E, and a material having a stabilized fluorite structure can be produced by using the molar ratio in region 104E.

As shown in FIG. 7, when the temperature is 773 K, i.e., 500° C., it is preferable to use the conditions of region 104E, and more preferably the conditions of region 100E. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive material is preferably 0<X≦6.3, and more preferably 0.1≦X≦0.5. Also, when the molar amount of uranium is 1, the molar amount Y of oxygen is preferably 1.9≦Y≦12, and more preferably 2.3≦Y≦3.2.

As shown in FIG. 8, when the temperature is 623 K, i.e., 350° C., a material having only a fluorite structure can be produced by using the molar ratio in region 100F, and a material having a stabilized fluorite structure can be produced by using the molar ratio in region 104F.

As shown in FIG. 8, when the temperature is 623 K, i.e., 350° C., it is preferable to use the conditions of region 104F, and more preferably the conditions of region 100F. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive material is preferably 0<X≦6.3, and more preferably 0.2≦X≦0.4. Also, when the molar amount of uranium is 1, the molar amount Y of oxygen is preferably 1.9≦Y≦12, and more preferably 2.3≦Y≦3.

From the above, in the processing apparatus, when the molar amount of radioactive material (MA) in the solidified body is set to 1 when the molar amount of uranium in the solidified body is set to 1, it is preferable that X is 0<X≦7, and more preferably 0<X≦6.3. By setting the molar ratio of uranium to MA in the above range, it is possible to produce a material containing a stabilized fluorite structure in the region 104A to the region 104F, and to stabilize the crystal. In addition, the processing apparatus 10, 10A can suppress the state of the crystal from changing during long-term storage by producing a crystal that satisfies the conditions of the region 104F. In addition, in the stabilization processing apparatus, it is preferable that the molar amount X of the radioactive material (MA) in the solidified body when the molar amount of uranium in the solidified body is set to 1 is 0<X≦7, and the molar amount Y of oxygen when the molar amount of uranium is set to 1 is 1.9≦Y≦12. By setting the molar ratio of uranium, MA, and oxygen within the above range, it is possible to produce a material containing a stabilized fluorite structure in regions 104A to 104F, and to stabilize the crystal.

In addition, it is more preferable that the stabilization treatment section has a molar amount of radioactive material (MA) of 0<X≦6.3 when the molar amount of uranium in the solidified body is 1, and the stabilization treatment temperature is 600°C or more and 1000°C or less. This allows stable crystals to be formed with a high probability during stabilization treatment. It is more preferable that the stabilization treatment section has a temperature of 0.08≦X≦0.5 when the stabilization treatment temperature is 600°C or more and 1000°C or less. In addition, by setting the ratio of the ternary system range to satisfy the condition under which only the fluorite structure is formed under the condition of a state of 500°C, it is possible to obtain crystals that can be stored for a long time even when the stabilization treatment temperature is 600°C or more and 1000°C or less. In addition, by making it possible to make a material with only the fluorite structure, it is possible to store it stably even if the storage temperature rises to 500°C.

In the stabilization processing unit, the molar amount X of the radioactive material (MA) when the molar amount of uranium in the solidified body is 1 is preferably 0<X≦7, more preferably 0<X≦6.3, and even more preferably 0.1≦X≦0.5. More specifically, the stabilization processing unit is preferably set to a condition that satisfies region 104E of the ternary phase diagram at 773K (500°C) shown in FIG. 7, and more preferably to a condition that satisfies region 100E. This makes it possible to suppress changes in the crystal structure even when the temperature of the manufactured material rises to 500°C during long-term storage, and allows for stable long-term storage. In the stabilization processing unit, the molar amount X of the radioactive material (MA) when the molar amount of uranium in the solidified body is 1 is preferably 0.1≦X≦0.5, and the molar amount Y of oxygen when the molar amount of uranium is 1 is preferably 2.3≦Y≦3.2. This makes it possible to suitably create only the fluorite structure.

Although the present invention has been described above with reference to the embodiments, the present disclosure is not limited to the above embodiments. Each configuration and their combinations in the above embodiments are merely examples, and additions, omissions, substitutions, and other modifications of configurations are possible without departing from the spirit of the present invention.

The present disclosure discloses the following inventions, but is not limited to the following.
(1) a uranium supply unit that supplies uranium to a liquid from which radioactive materials containing at least one of minor actinides and lanthanides have been extracted;
a solidification processing unit which is supplied with uranium, heats the liquid from which the minor actinides and the lanthanides have been extracted to evaporate a portion of the liquid, and then further heats the liquid to produce a solidified body;
a stabilization treatment section for sintering the produced solidified body to remove a part or all of the carbon, hydrogen, oxygen, and nitrogen components, thereby producing a fluorite structure of uranium dioxide having the minor actinides and the lanthanides incorporated therein;
The stabilization treatment unit is a high-level radioactive material treatment system in which a molar amount X of the radioactive material in the solidified body is 0<X≦7 when a molar amount of uranium is 1.

(2) The stabilization treatment unit is a high-level radioactive material treatment system as described in (1), in which the molar amount Y of oxygen when the molar amount of uranium is 1 is 1.9≦Y≦12.

(3) A high-level radioactive material processing system as described in (1) or (2) in which the stabilization processing unit has a molar amount X of the radioactive material in the solidified body when the molar amount of uranium in the solidified body is 1, which is 0<X≦6.3, and the stabilization processing temperature is 600°C or higher and 1000°C or lower.

(4) A high-level radioactive material processing system according to any one of (1) to (3), in which the stabilization processing unit is a system in which the molar amount of the radioactive material in the solidified body, when the molar amount of uranium in the solidified body is set to 1, is X, and 0.1≦X≦0.5.

(5) A high-level radioactive material processing system as described in (4) in which the stabilization processing unit has a molar amount Y of oxygen when the molar amount of uranium is 1, and the molar amount Y is 2.3≦Y≦3.2.

(6) A high-level radioactive material processing system according to any one of (1) to (5), wherein the stabilization processing unit is provided with a reducing gas supply unit that supplies reducing gas during stabilization processing, and the amount of oxygen is controlled by the amount of reducing gas supplied.

(7) A high-level radioactive material processing system according to any one of (1) to (6), comprising an extraction device that uses an organic solvent liquid to extract the minor actinides and the lanthanides from a liquid containing high-level radioactive materials, extracts the organic solvent containing the minor actinides and the lanthanides, and supplies a diluting liquid to produce a diluted liquid, and the uranium supply unit adds a uranium complex that dissolves in the organic solvent to the liquid.

(8) A high-level radioactive material treatment system as described in (7), in which the diluting liquid has a boiling point of 30°C or higher and 100°C or lower.

(9) A high-level radioactive material treatment system as described in (7) or (8), in which the dilution liquid has no flash point or a flash point equal to or higher than the decomposition temperature of the extractant.

(10) An extraction apparatus for extracting the minor actinides and the lanthanides from a liquid containing high-level radioactive materials by using an organic solvent liquid, mixing the extracted organic solvent with a diluent containing a liquid-phase stripping agent, and generating a liquid in which the minor actinides and the lanthanides have been transferred to the diluent containing the liquid-phase stripping agent,
The high-level radioactive material treatment system according to any one of (1) to (6), wherein the uranium supply unit adds uranium ions that dissolve in a liquid phase to the liquid.

(11) A high-level radioactive material processing system as described in (10), in which the solidification device produces a solidified body by heat treatment.

(12) A step of supplying uranium to a liquid from which a radioactive material containing at least one of a minor actinide and a lanthanide has been extracted, at a molar amount X of the radioactive material when the molar amount of uranium is taken as 1, such that 0<X≦7;
a step of supplying uranium, heating the liquid from which the minor actinides and the lanthanides have been extracted to evaporate a portion of the liquid, and then further heating the liquid to produce a solidified body;
and performing, in a single apparatus, a stabilization treatment of sintering the produced solidified body, removing some or all of the carbon, hydrogen, oxygen, and nitrogen components, and producing a fluorite structure of uranium dioxide having the minor actinides and the lanthanides incorporated therein.

10, 10A Processing system (high-level radioactive material processing system)
12, 12A Extraction device 14, 14A Solidification device 16, 16A Stabilization device 18 Storage device 22 Waste liquid supply section 24 Extractant supply section 26 Dilution liquid supply section 28 MA extract liquid production section 30 MA stripping liquid production section 34 Hydrothermal treatment accelerator supply section 40, 40A Solidification treatment section 44, 44A Uranium supply section 45, 45A Distillation section 46, 46A Heat treatment section 47, 47A Dilution liquid recovery section 48, 48A Oxidizing gas supply section 49, 49A Exhaust gas recovery section 50, 50A Stabilization treatment section 52, 52A Reduction section 54, 54A Reduction gas supply section 56, 56A Exhaust gas recovery section 60 Storage section

Claims (12)

a uranium supply unit that supplies uranium to the liquid from which the radioactive material containing at least one of a minor actinide and a lanthanide has been extracted;
a solidification processing unit which is supplied with uranium, heats the liquid from which the minor actinides and the lanthanides have been extracted to evaporate a portion of the liquid, and then further heats the liquid to produce a solidified body;
a stabilization treatment section for sintering the produced solidified body to remove a part or all of the carbon, hydrogen, oxygen, and nitrogen components, thereby producing a fluorite structure of uranium dioxide having the minor actinides and the lanthanides incorporated therein;
The stabilization treatment unit is a high-level radioactive material treatment system in which a molar amount X of the radioactive material in the solidified body is 0<X≦7 when a molar amount of uranium is 1. The high-level radioactive material processing system according to claim 1, wherein the stabilization processing unit has a molar amount Y of oxygen in a range of 1.9≦Y≦12 when the molar amount of uranium is 1. The high-level radioactive material processing system according to claim 1, wherein the stabilization processing unit is configured such that the molar amount of the radioactive material in the solidified body, when the molar amount of uranium in the solidified body is 1, is X, and the stabilization processing temperature is 600°C or higher and 1000°C or lower. The high-level radioactive material processing system according to claim 1, wherein the stabilization processing unit is configured such that the molar amount X of the radioactive material in the solidified body is 0.1≦X≦0.5 when the molar amount of uranium in the solidified body is 1. The high-level radioactive material processing system according to claim 4, wherein the stabilization processing unit has a molar amount Y of oxygen when the molar amount of uranium is 1, and the molar amount Y is 2.3≦Y≦3.2. The high-level radioactive material processing system according to claim 1, wherein the stabilization processing unit includes a reducing gas supply unit that supplies reducing gas during stabilization processing, and the amount of oxygen is controlled by the amount of reducing gas supplied. an extraction device that uses an organic solvent liquid to extract the minor actinides and the lanthanides from a liquid containing a high-level radioactive material, extracts the organic solvent containing the minor actinides and the lanthanides, and supplies a dilution liquid to produce a diluted liquid;
The high-level radioactive material processing system according to claim 1 , wherein the uranium supply unit adds a uranium complex that dissolves in the organic solvent to the liquid. The high-level radioactive material processing system according to claim 7, wherein the diluting liquid has a boiling point of 30°C or higher and 100°C or lower. The high-level radioactive material processing system according to claim 7, wherein the dilution liquid has no flash point or a flash point equal to or higher than the decomposition temperature of the extractant. an extraction device for extracting the minor actinides and the lanthanides from a liquid containing high-level radioactive materials using an organic solvent liquid, and then mixing the extracted organic solvent with a diluent containing a liquid-phase stripping agent to generate a liquid in which the minor actinides and the lanthanides have been transferred to the diluent containing the liquid-phase stripping agent;
The high-level radioactive material processing system according to claim 1 , wherein the uranium supply unit adds uranium ions that dissolve in a liquid phase to the liquid. The high-level radioactive material processing system according to claim 10, wherein the solidification device produces a solidified body by heat treatment. supplying uranium to a liquid from which a radioactive material containing at least one of a minor actinide and a lanthanide has been extracted at a molar amount X of the radioactive material, where X is 0≦X≦7, where X is a molar amount of uranium taken as 1;
a step of supplying uranium, heating the liquid from which the minor actinides and the lanthanides have been extracted to evaporate a portion of the liquid, and then further heating the liquid to produce a solidified body;
and performing, in a single apparatus, a stabilization treatment of sintering the produced solidified body, removing some or all of the carbon, hydrogen, oxygen, and nitrogen components, and producing a fluorite structure of uranium dioxide having the minor actinides and the lanthanides incorporated therein.

Images