High-level radioactive material processing system and high-level radioactive material processing method

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

[Technical Field]

[0001] The present disclosure relates to a high-level radioactive material processing system and a high-level radioactive material processing method. [Background technology]

[0002] One method for treating high-level radioactive waste is to extract minor actinides, which are radioactive materials, from high-level radioactive materials (e.g., Patent Document 1 and Patent Document 2).Another method for storing spent fuel in a recyclable form is to solidify it (e.g., Patent Document 3). [Prior art documents] [Patent documents]

[0003] [Patent Document 1] Japanese Patent Application Publication No. 2018-63198 [Patent Document 2] Japanese Patent Application Publication No. 2019-15533 [Patent Document 3] Patent No. 6037168 Summary of the Invention [Problem to be solved by the invention]

[0004] By extracting minor actinides from highly radioactive materials, the extracted minor actinides can be reused as fuel for fast breeder reactors and other reactors. Furthermore, removing the minor actinides can reduce the burden of waste disposal. However, 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 into practical use, the separated minor actinides must be stored stably for long periods of time.

[0005] The present disclosure has been made in view of the above, and an object thereof is to provide a high-level radioactive substance treatment system and a high-level radioactive substance treatment method capable of making minor actinides into a solid in a more stable state.

Means for Solving the Problems

[0006] In order to solve the above-described problems and achieve the object, the present disclosure provides a high-level radioactive substance treatment apparatus including: a uranium supply unit that supplies uranium to a liquid from which a radioactive substance including at least one of minor actinides and lanthanoids has been extracted; a solidification treatment unit that heats the liquid to which uranium has been supplied and from which the minor actinides and the lanthanoids have been extracted, evaporates a part of the liquid, and then further heats the liquid to produce a solidified body; and a stabilization treatment unit that sinters the produced solidified body, removes some or all of carbon, hydrogen, oxygen, and nitrogen components, and generates a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanoids therein, wherein in the stabilization treatment unit, when the molar amount of uranium is set to 1, the molar amount X of the radioactive substance satisfies 0 < X ≦ 7.

[0007] In order to solve the above-described problems and achieve the object, the present disclosure provides a high-level radioactive substance treatment method including: a step of supplying uranium to a liquid from which a radioactive substance including at least one of minor actinides and lanthanoids has been extracted at a ratio such that when the molar amount of uranium is set to 1, the molar amount X of the radioactive substance satisfies 0 < X ≦ 7; a step of heating the liquid to which uranium has been supplied and from which the minor actinides and the lanthanoids have been extracted, evaporating a part of the liquid, and then further heating the liquid to produce a solidified body; and a step of performing, in one apparatus, a stabilization treatment of sintering the produced solidified body, removing some or all of carbon, hydrogen, oxygen, and nitrogen components, and generating a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanoids therein.

Advantages of the Invention

[0008] According to the present disclosure, minor actinides can be made into a solid in a more stable state. [Brief explanation of the drawings]

[0009] [Figure 1] FIG. 1 is a schematic diagram showing the general configuration of a high-level radioactive material treatment apparatus according to this embodiment. [Figure 2] FIG. 2 is a schematic diagram showing a schematic configuration of a high-level radioactive material treatment apparatus according to another embodiment. [Figure 3] FIG. 3 is an explanatory diagram showing a ternary phase diagram at 1523 K (1250° C.). [Figure 4] FIG. 4 is an explanatory diagram showing a ternary phase diagram at 1273 K (1000° C.). [Figure 5] FIG. 5 is an explanatory diagram showing a ternary phase diagram at 1073 K (800° C.). [Figure 6] FIG. 6 is an explanatory diagram showing a ternary phase diagram at 923 K (650° C.). [Figure 7] FIG. 7 is an explanatory diagram showing a ternary phase diagram at 773 K (500° C.). [Figure 8] FIG. 8 is an explanatory diagram showing a ternary phase diagram at 623 K (350° C.). DETAILED DESCRIPTION OF THE INVENTION

[0010] An embodiment of a high-level radioactive material treatment device according to the present disclosure will be described in detail below with reference to the drawings. The high-level radioactive material treatment device according to the present disclosure extracts MAs (minor actinides) from high-level radioactive material and stabilizes the MA-extracted waste by vitrifying it. Furthermore, when lanthanides are contained in the high-level radioactive material, the high-level radioactive material treatment device extracts the lanthanides along with the minor actinides. In this disclosure, "MAs (minor actinides)" refers to transuranic 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 (lanthanoids)" is a general term for elements with atomic numbers from 57 to 71.

[0011] FIG. 1 is a schematic diagram showing the overall configuration of a high-level radioactive material treatment device according to this embodiment. The high-level radioactive material treatment device (treatment device) 10 shown in FIG. 1 includes an extraction device 12, a solidification device 14, a stabilization device 16, and a storage device 18. The treatment device according to this embodiment will be described assuming that high-level radioactive waste (hereinafter also referred to as "HALW") is used as the high-level radioactive material. The waste liquid is, for example, a liquid remaining after U (uranium) and Pu (plutonium) are recovered from a solution of spent nuclear fuel discharged from a light-water reactor during reprocessing. The high-level radioactive waste contained in the waste liquid includes fission products (hereinafter also referred to as "FPs"), MAs, and lanthanides. A specific example of high-level radioactive waste (hereinafter also referred to as "HALW") is waste liquid generated during reprocessing using the PUREX process. In the Purex process, a nitric acid solution containing U and Pu is brought into contact with tributyl phosphate (TBP) and an organic solvent such as dodecane. This causes U and Pu in the nitric acid solution to 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.

[0012] The extractor 12 extracts MA components from waste liquid. The extractor 12 includes a waste liquid supply unit 22, an extractant supply unit 24, a diluent supply unit 26, and an MA extract liquid generator 28. The waste liquid supply unit 22 stores HALW, which is liquid high-level radioactive waste, and supplies it to the MA extract liquid generator 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, which supplies a liquid organic solvent in which the extractant is dissolved.

[0013] The extractant supply unit 24 supplies the extractant to the MA extract generation unit 28. The extractant captures MA and Ln. The extractant is a liquid that migrates into the diluent 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 should preferably selectively form a complex with MA and be inexpensive. Examples of extractants include n-octyl(phenyl)-N,N'-diisobutylcarbamoylmethylphosphine oxide-tributyl phosphate mixture (CMPO-TBP mixture), diisodecyl phosphoric acid, 6,6'-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-1,2,4-benzotriazin-3-yl)-2,2'-bipyridine (BTBP), and N,N'-dibutyl-N,N'-dimethyltetradecylmalonamide (DMDBTDMA). A DGA-based material (complexing agent) is preferably used as the extractant. Specific examples of the complexing agent include N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA), tetra(2-ethylhexyl)diglycolamide (T2EHDGA), etc. The extracting agent may be used alone or in combination of two or more.

[0014] 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 selected appropriately depending on the extractant used. The organic solvent is preferably reusable, inexpensive, and resistant to radiation degradation. The diluent is an organic phase material that is insoluble in the liquid components of the waste liquid. The diluent has the ability to elute the extractant from the liquid waste. The diluent preferably has a boiling point of 30°C or higher and 100°C or lower. Furthermore, the diluent preferably has no flash point or a flash point equal to or higher than the decomposition temperature of the extractant. Using a liquid that satisfies the above range as the diluent can reduce the processing load in the solidification processing unit. Furthermore, the diluent is preferably a material that is less susceptible to degradation by heat or radiation even when reused, and preferably a solvent with a hydrocarbon structure that is less susceptible to degradation by heat or radiation even when reused. Here, being less susceptible to deterioration by heat or radiation even when reused means that even if it is exposed to heat or radiation energy and becomes a low-molecular-weight structure, it has the property of being less likely to form a complex with metal ions. As the diluent, for example, a fluorine-based material such as hydrofluorocarbon (HFC) is preferably used. One type of organic solvent may be used alone, or two or more types may be used in combination. Furthermore, it is preferable that the diluent be made into a reusable liquid after being separated from the MA in the solidification device 14.

[0015] The MA extract production unit 28 is supplied with waste liquid, an extractant, and a diluent. The MA extract production unit 28 uses solvent extraction to bring MA and Ln in the waste liquid into contact with the extractant, resulting in the transfer of MA and Ln to the extractant. The extractant that has captured MA and Ln is encapsulated in the diluent. After the extraction process, the MA extract production unit 28 separates the diluent from the waste liquid to produce an MA extract in which the extractant that has captured MA and Ln is encapsulated in the diluent. The MA extract production unit 28 may be a continuous system in which each material is continuously supplied and an MA extract is produced, or a batch system in which each material is intermittently supplied and an MA extract is produced. The MA extract production unit 28 may selectively capture MA with the extractant, or may simultaneously capture and process MA and Ln.

[0016] The solidification device 14 mixes uranium with the MA extract, removes the liquid components from the mixture, 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 tetravalent or hexavalent state.

[0017] The solidification treatment unit 40 includes a distillation unit 45, a thermal treatment unit 46, a diluent recovery unit 47, an oxidizing gas supply unit 48, and an exhaust gas recovery unit 49. The distillation unit 45 heats the MA extract mixed with uranium, for example, to 50 to 100°C, to evaporate the diluent. The thermal treatment unit 46, an apparatus integrated with the distillation unit 45, further heats the object from which the diluent has been evaporated, for example, to 300°C, to thermally decompose the residue and convert it to an oxide. The diluent recovery unit 47 is connected to the distillation unit 45 and recovers the diluent evaporated in the dilution unit 45. The diluent recovery unit 47 supplies the recovered diluent to the diluent supply unit 26 for reuse. The oxidizing gas supply unit 48 supplies an oxidizing gas, such as oxygen or air, to the thermal treatment unit 46. The exhaust gas supply unit 49 recovers the exhaust gas discharged from the thermal treatment unit 46.

[0018] The solidification unit 40 distills the MA extract mixed with uranium in the distillation unit 45 to remove the diluent components. By removing the diluent, the solidification unit 40 leaves a solidified residue containing the extractant, MA, Ln, and uranium as a complex. The solidification unit 40 also recovers the gas emitted during distillation in the distillation unit 45 in the diluent recovery unit 47. The solidification unit 40 heats the dilution product obtained by distilling the diluent in the heat treatment unit 46 while supplying an oxidizing gas from the oxidizing gas supply unit 48, thereby removing the extractant components, specifically, C, H, O, and N, and producing a solidified product. The solidified product becomes oxides. The solidification unit 40 can use known evaporation methods, such as batch-type and continuous-type (plate tower or packed tower) distillation and heat treatment. The distillation unit 45 preferably operates at a temperature between −10°C and +20°C, based on the boiling point of the diluted MA extract, and below the flash point. The temperature in the heat treatment section 46 is equal to or higher than the temperature at which thermal decomposition and oxide conversion occur, for example, 300° C. or higher.

[0019] The stabilization device 16 performs a stabilization treatment to remove some or all of the carbon, hydrogen, oxygen, and nitrogen components from the solidified body. The stabilization device 16 includes a stabilization treatment section 50.

[0020] The stabilization treatment unit 50 includes a reduction unit 52, a reducing gas supply unit 54, and an exhaust gas recovery unit 56. The reduction unit 52 heats the solidified body in a reducing atmosphere to calcinate and sinter it. The reduction unit 52 is an integrated device with the distillation unit 45 and the heat treatment unit 46. The reducing gas supply unit 54 supplies a reducing gas, such as hydrogen, to the reduction unit 52. The exhaust gas recovery unit 56 recovers the exhaust gas discharged from the reduction unit 52. The stabilization treatment unit 50 also produces UO2 form while suppressing the generation of UO3 form by adjusting the temperature and performing hydrogen reduction during treatment. The stabilization treatment unit 50 can also produce UO2 form while suppressing the generation of UO3 form by supplying water during treatment and performing hydration reduction.

[0021] The stabilization treatment unit 50 supplies a 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, nitrate components, and moisture) in the solidified body, i.e., CHON components, and then sinters it. The sintering temperature is, for example, 600°C or higher and 1250°C or lower, preferably 600°C or higher and 800°C or lower.

[0022] The stabilization treatment unit 50 removes impurities from the solidified body containing uranium, and then sinters it in a reducing atmosphere to obtain uranium oxide with a fluorite structure in which MA and Ln are dissolved. 3+ ·(NO3 - )3·m(extractant)]+([U 4+ ·(NO3 - )4·n(ligand)] or [M 3+ ·(NO3 - )3·m(extractant)]+([UO2 2+ ·(NO3 - )2·n(ligand)] is generated. Here, M is MA or Ln. By heat treating this solidified body and removing the CHON component, M a O b +U3O8+(αCO2+βH2O+γNO x ) and (αCO2 + βH2O + γNO x ) is removed. Furthermore, M a O b +U3O8 is sintered to form an oxide with a fluorite structure, UM RM O y is formed. RM is M(Ln or MA) / U, which is the equivalence ratio of Ln or MA to U. The stabilization device 16 removes some or all of the carbon, hydrogen, oxygen, and nitrogen components that may gasify during storage, and turns the uranium dioxide into a fluorite structure containing MA and Ln, resulting in a solidified body with high thermal and chemical stability.

[0023] The storage device 18 stores the stabilized 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 only needs to be able to remove decay heat from the stabilized MA-containing material. Alternatively, the storage section 60 may be installed underground, and the stabilized material may be buried underground.

[0024] The MA-containing material stored in the storage unit 60 of the processing system 10 can be used as fuel for nuclear power generation systems. When using MA as fuel for nuclear transmutation, the stored fluorite-structured solidified body is melted to produce a solution containing U, MA, and Ln (solidified body melting). Next, the resulting solution is subjected to a purification process for U and MA (MA purification). In MA purification, for example, U and MA are separated from Ln in the resulting solution, and highly exothermic MA is separated from MA as needed. Next, the resulting U and MA (Np, Am, etc.) are mixed with U and Pu to obtain mixed oxides, thereby producing fuel (MA fuel production). The resulting fuel is burned in a fast breeder reactor or the like (MA combustion).

[0025] In the processing device 10 of this embodiment, an extractant and a diluent are added to the waste liquid and mixed in the extraction device 12 to perform a separation process to separate MA together with Ln from the waste liquid, 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, thereby producing a solidified body containing U, MA, and Ln. Next, in the processing device 10, an organic component (CHON) is removed from the solidified body containing MA in the stabilization device 16 to form a fluorite structure. The processing device 10 stores the processed product with the fluorite structure in the storage device 18.

[0026] It is preferable that the processing apparatus 10 performs the heat treatment of the solidification and stabilization apparatus 13 in one device. This simplifies the device configuration. Furthermore, since the processing apparatus 10 can perform multiple processes in one device, it is possible to reduce the number of times radioactive materials need to be transported, and the processing can be performed more safely.

[0027] Furthermore, by using a diluent with a boiling point of 30°C or higher and 100°C or lower, or with no flash point or 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.

[0028] The processing device 10 separates, solidifies, and stabilizes MA along with Ln. This allows for efficient extraction of MA from liquid high-level radioactive waste, reducing the waste disposal burden, and solidifying the waste. Furthermore, by extracting MA and Ln and stabilizing the material, MA can be stored stably for reuse as fuel. Furthermore, when reused, the solidified material can be made reusable simply by dissolving it.

[0029] Furthermore, by processing MA and Ln without separating them, the process of separating MA and Ln becomes unnecessary, and the processing load can be reduced.

[0030] Furthermore, the processing device 10 preferably uses an extractant that does not form a third phase when producing an MA extract from highly radioactive materials containing nitric acid-based substances. Specifically, it is preferable to use a DGA-based material. One example is T2EHDGA. Furthermore, it is preferable that the extractant 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. Furthermore, it is preferable that the extractant does not contain any corrosive components.

[0031] Furthermore, the processing apparatus 10 supplies uranium to a liquid containing MA and Ln using 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 with MA and Ln immobilized. This allows the stabilization-treated solid to have a long-term stable crystal structure, enabling stable storage of MA. Specifically, trivalent MA and Ln can be dissolved in the fluorite structure, resulting in a more stable fluorite structure than a mixture of MA and Ln or an oxide structure containing only one of them. Furthermore, by forming a crystal structure using uranium, when a solid containing MA is reprocessed for use, it can be free of unwanted components such as Si, Na, and Ca. This allows for a stabilized substance to be obtained in a state that is easy to use during reprocessing.

[0032] The uranium supply unit 44 preferably supplies uranium as an organic complex. This facilitates dissolution in the diluent. The ligands coordinated to the organic U complex are preferably composed only of carbon, hydrogen, nitrogen, and oxygen. Furthermore, the ligands coordinated to the organic U complex are preferably easy to stabilize (thermally decompose) and do not contain corrosive components. This allows for convenient removal during the stabilization process, enabling MA-containing substances to be stabilized and stored stably.

[0033] In the above embodiment, the extractor 12 produces an MA extract by extracting with an organic solvent and then solidifies the extract. However, this is not limiting. The treatment device may also produce a stripped solution by extracting MA into a liquid phase, solidify it, and stabilize it.

[0034] Fig. 2 is a schematic diagram showing the general configuration of a high-level radioactive material treatment device of another embodiment. A high-level radioactive material treatment device (treatment device) 10A shown in Fig. 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 using high-level radioactive waste (hereinafter also referred to as "HALW") as the high-level radioactive material, similar to the treatment device 10.

[0035] The extractor 12A extracts MA components from waste liquid. The extractor 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 liquid high-level radioactive waste, and supplies it to the MA extract liquid production unit 28.

[0036] The extractant supply unit 24 supplies the extractant to the MA extract solution generator 28. The extractant captures MA and Ln. The extractant is a liquid that transfers to the diluent. 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 a complexing agent that selectively forms a complex with MA. Examples of extractants include n-octyl(phenyl)-N,N'-diisobutylcarbamoylmethylphosphine oxide-tributyl phosphate mixture (CMPO-TBP mixture), diisodecyl phosphoric acid, 6,6'-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-1,2,4-benzotriazin-3-yl)-2,2'-bipyridine (BTBP), and N,N'-dibutyl-N,N'-dimethyltetradecylmalonamide (DMDBTDMA). A DGA-based material (complexing agent) is preferably used as the extractant. Specific examples of the complexing agent include N,N,N',N'-tetraoctyl-3-oxapentanediamide (TODGA), tetra(2-ethylhexyl)diglycolamide (T2EHDGA), etc. The extracting agent may be used alone or in combination of two or more.

[0037] 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 selected appropriately 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 organic solvent may be used alone, or two or more organic solvents may be used in combination. Furthermore, it is preferable that the diluent is separated from the MA in the MA stripping liquid production unit 30 and then converted into a reusable liquid.

[0038] The MA extract production unit 28 is supplied with waste liquid, an extractant, and a diluent. In the MA extract production unit 28, MA and Ln in the waste liquid are brought into contact with the extractant by solvent extraction, resulting in the transfer of MA and Ln to the extractant. The extractant also transfers to the diluent. After the extraction process, the MA extract production unit 28 separates the diluent from the waste liquid to produce an MA extract, which is an extractant containing MA and Ln. The MA extract production unit 28 may be a continuous system in which each material is continuously supplied to produce the MA extract, or a batch system in which each material is intermittently supplied to produce the MA extract. In this embodiment, the extractant supply unit 24 and the diluent supply unit 26 are provided. However, the extractant supply unit 24 and the diluent supply unit 26 may be combined into a single device to supply a liquid organic solvent containing dissolved extractant.

[0039] The MA stripping liquid production unit 30 is supplied with the MA extract from the MA extract production unit 28 and with the stripping agent from the stripping agent supply unit 32 .

[0040] The stripping agent supply unit 32 supplies a substance that transfers MA and Ln from the MA extract to the liquid phase as a stripping agent. The stripping agent is, for example, nitric acid. The stripping agent is diluted with a liquid diluent, such as water. The hydrothermal treatment promoter supply unit 34 supplies the hydrothermal treatment promoter to the MA stripping solution production unit 30. The hydrothermal treatment promoter is a neutral to basic liquid, and is a reducing agent composed of CNOH, such as an aldehyde compound or an amine compound. The preparation of the mixture of the MA extracting agent, stripping agent, and hydrothermal treatment promoter is not limited to this embodiment. The hydrothermal treatment promoter supply unit 34 may supply the stripping agent to the MA extract, and then supply the hydrothermal treatment promoter to the MA stripping solution.

[0041] The MA stripping solution generator 30 brings the MA extract into contact with a diluent containing a stripping agent, transferring the MA and Ln contained in the organic phase MA extract to the diluent containing a stripping agent. The extractor 12A may reuse the organic solvent (extractant and diluent) in the MA extract after treatment. The MA stripping solution generator 30 controls the acid concentration of the liquid and the amount of stripping agent to maintain a predetermined component ratio during solidification treatment 14A.

[0042] The solidification device 14A supplies uranium to the MA stripping solution, removes the liquid components of the mixed liquid, and produces a solidified product. 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 solution in the solidification processing unit 40A. The uranium supply unit 44A supplies a uranium-containing substance that is uniformly soluble in the liquid-phase solvent that constitutes the MA stripping solution. Uranium dissolves in the liquid phase in a tetravalent or hexavalent state.

[0043] The solidification treatment unit 40A includes a hydrothermal treatment unit 45A, a thermal treatment unit 46A, an exhaust gas recovery unit 47A, an oxidizing gas supply unit 48A, and an exhaust gas recovery unit 49A. The hydrothermal treatment unit 45A heats the MA stripping solution containing uranium to, for example, 80 to 150°C, causing hydrothermal treatment and oxide conversion, and evaporating the liquid components. The liquid components also include a hydrothermal treatment accelerator. The thermal treatment unit 46A is an integrated unit with the hydrothermal treatment unit 45A and further heats the target object from which the liquid has been evaporated, for example, to 300°C, to thermally decompose the residue. The exhaust gas recovery unit 47A is connected to the hydrothermal treatment unit 45A and recovers the exhaust gas evaporated by the hydrothermal treatment unit 45A. The oxidizing gas supply unit 48A supplies an oxidizing gas, such as oxygen or air, to the thermal treatment unit 46A. The exhaust gas supply unit 49A recovers the exhaust gas discharged from the thermal treatment unit 46A.

[0044] The solidification unit 40A hydrothermally synthesizes the MA stripping solution mixed with uranium in the hydrothermal treatment unit 45A to form a solidified material containing MA and uranium. The liquid component is removed by solid-liquid separation of the solidified material and the diluent component. The hydrothermal reaction can also be accelerated by adding a hydrothermal treatment accelerator. The solidification unit 40A heats the oxidized target material from which the liquid has been removed in the heat treatment unit 46A while supplying an oxidizing gas from the oxidizing gas supply unit 48A, thereby removing extractant components, specifically, C, N, O, and H, to produce a solidified material. The solidified material contains U, MA, and Ln. Specifically, the solidified material contains UO2 and an M oxide precursor. The M oxide precursor is, for example, MOOH or M(OH)3. M is at least one of MA and Ln. The hydrothermal treatment temperature of the hydrothermal treatment unit 45A is, for example, 80°C to 150°C, and the treatment time is, for example, 0.5 hours to 20 hours. The temperature in the heat treatment section 46 is equal to or higher than the temperature at which thermal decomposition occurs, for example, 300° C. or higher.

[0045] The stabilization device 16A performs a stabilization treatment to remove part 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 treatment unit 50A.

[0046] The stabilization treatment unit 50A includes a reduction unit 52A, a reducing gas supply unit 54A, and an exhaust gas recovery unit 56A. The reduction unit 52A heats the solidified body in a reducing atmosphere to calcinate and sinter it. The reduction unit 52A is an integrated device with the distillation unit 45A and the heat treatment unit 46A. The reducing gas supply unit 54A supplies a reducing gas, such as hydrogen, to the reduction unit 52A. The exhaust gas recovery unit 56A recovers the exhaust gas discharged from the reduction unit 52A.

[0047] The stabilization unit 50A supplies reducing gas from the reducing gas supply unit 54 to the reduction unit 52A, while heating, calcining, and sintering the solidified material in the reduction unit 52A, solidifying the material into a fluorite structure. Specifically, the stabilization unit 50A sinters the solidified material to perform denitration and pyrolysis, removing impurity components (organic components, nitrate components, and moisture) in the solidified material, i.e., CHON components, and then sintering the solidified material. The sintering temperature is, for example, 600°C or higher and 800°C or lower.

[0048] The stabilization treatment section 50A removes impurities from the uranium-containing solidified material, and then sinters it to obtain fluorite-structured uranium oxide with MA and Ln dissolved therein. The stabilization device 16A removes some or all of the carbon, hydrogen, oxygen, and nitrogen components that may gasify during storage, resulting in a fluorite-structured uranium dioxide containing MA and Ln, resulting in a solidified material with high thermal and chemical stability.

[0049] The storage device 18 stores the stabilized material. The storage device 18 includes a storage unit 60. The storage unit 60 stores the stabilized MA-containing material in a solid state. The storage unit 60 is, for example, a cask. Alternatively, the storage unit 60 may be installed underground, and the stabilized material may be buried underground.

[0050] The processing device 10A of this embodiment performs extraction and back-extraction in the extraction device 12A, processes a solution containing MA in the liquid phase, supplies uranium, and then produces a fluorite structure of uranium dioxide with MA and Ln immobilized in the stabilization device 16A.

[0051] It is preferable that the processing apparatus 10A performs the heat treatment of the solidification and stabilization apparatus 13 in one apparatus. This simplifies the apparatus configuration. Furthermore, since the processing apparatus 10A can perform multiple processes in one apparatus, it is possible to reduce the transportation of radioactive materials and perform processing more safely.

[0052] Furthermore, even when back-extraction is performed, by forming a fluorite structure, the solid after stabilization treatment can have a long-term stable crystalline structure, allowing for stable storage of MA. Specifically, trivalent MA and Ln can be dissolved in the fluorite structure, making the stable fluorite structure even more stable. Furthermore, by forming a crystalline structure using uranium, when a solid containing MA is reprocessed for use, it is possible to obtain a structure that is free of unnecessary components such as Si, Na, and Ca. This allows for a substance that is stabilized and easy to use during reprocessing. Furthermore, the processing device 10A can achieve various effects similar to those of the processing device 10.

[0053] The uranium supply unit 44 of the processing equipment 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 equipment, thereby providing a structure that is more stable than the fluorite structure of the solid material during storage.

[0054] As described above, the processing apparatus 10, 10A allows MA to have a long-term stable crystal structure, preferably a fluorite structure, thereby enabling long-term stable storage of MA. The MA-containing substance produced by the processing apparatus 10, 10A can be made more stable by setting the stabilization treatment conditions and storage conditions to predetermined conditions. Specifically, a more stable state can be achieved by achieving a predetermined balance in the molar ratio of MA to U and O (oxygen). The molar ratio of MA to U and O (oxygen) can be set within a predetermined range during the stabilization treatment by adjusting the amount of U added relative to MA and the treatment atmosphere.

[0055] Next, with reference to FIGS. 3 to 8, suitable conditions for the processing method in the processing apparatus 10 will be described. 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 923 K (650 °C). FIG. 7 is an explanatory diagram showing a ternary phase diagram at 773 K (500 °C). FIG. 8 is an explanatory diagram showing a ternary phase diagram at 623 K (350 °C).

[0056] FIGS. 3 to 8 are diagrams showing crystal structures formed by the molar ratios of three elements of oxygen (O), neodymium (Nd), and uranium (U) at 1 atm (1 atmosphere) and each temperature. Niobium becomes a simulated MA that simulates minor actinides. That is, by performing the treatment with the same amount of MA as Nd, the crystal structures shown in FIGS. 3 to 8 can be produced. In FIGS. 3 to 8, the atmosphere conditions were adjusted by controlling the supply of an inert gas containing a trace amount of hydrogen.

[0057] As shown in FIG. 3, when the temperature is 1523 K, that is, 1250 °C, by setting the molar ratio in region 104A, a crystal containing a fluorite structure can be produced. Specifically, in region 104A, a substance in which a fluorite structure, U3O8, rhombohedron, etc. are mixed is generated. Also, when the temperature is 1523 K, that is, when the temperature is 1250 °C, by setting region 104A, a stabilized crystal structure can be obtained. Specifically, in region 104A, in addition to the fluorite structure, a substance in which uranium单质, rhombohedron are mixed is generated.

[0058] 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, when the molar amount of uranium is set to 1, it is preferable that the molar amount X of the radioactive substance satisfies 0 < X ≦ 7. Also, when the molar amount of uranium is set to 1, it is preferable that the molar amount Y of oxygen (O) satisfies 1.9 ≦ Y ≦ 12.

[0059] As shown in Fig. 4, when the temperature is 1273 K, that is, 1000 °C, by setting the molar ratio of Region 100B, a substance with only a fluorite structure can be produced. In Region 100B, all the included crystals have a fluorite structure. Also, as shown in Fig. 4, when the temperature is 1273 K, that is, 1000 °C, by setting it to Region 104B, a substance containing a stabilized fluorite structure can be produced.

[0060] As shown in Fig. 4, when the temperature is 1273 K, that is, 1000 °C, it is preferable to set the conditions of the region of Region 104B, and more preferably to set the conditions of Region 100B. As an example, when the molar amount of uranium is set to 1, the molar amount X of the radioactive substance preferably satisfies 0 < X ≤ 7, and more preferably 0.08 ≤ X ≤ 0.5. Also, when the molar amount of uranium is set to 1, the molar amount Y of oxygen preferably satisfies 1.9 ≤ Y ≤ 12, and more preferably 2.2 ≤ Y ≤ 3.2.

[0061] As shown in Fig. 5, when the temperature is 1073 K, that is, 800 °C, by setting the molar ratio of Region 100C, a substance with only a fluorite structure can be produced, and by setting the molar ratio of Region 104C, a substance containing a stabilized fluorite structure can be produced.

[0062] As shown in Fig. 5, when the temperature is 1073 K, that is, 800 °C, it is preferable to set the conditions of the region of Region 104C, and more preferably to set the conditions of Region 100C. As an example, when the molar amount of uranium is set to 1, the molar amount X of the radioactive substance preferably satisfies 0 < X ≤ 7, and more preferably 0 < X ≤ 0.6. Also, when the molar amount of uranium is set to 1, the molar amount Y of oxygen preferably satisfies 1.9 ≤ Y ≤ 12, and more preferably 2.0 ≤ Y ≤ 3.2.

[0063] As shown in Fig. 6, when the temperature is 923 K, that is, 650 °C, by setting the molar ratio of Region 100D, a substance with only a fluorite structure can be produced, and by setting the molar ratio of Region 104D, a substance containing a stabilized fluorite structure can be produced.

[0064] As shown in Fig. 6, when the temperature is 923 K, that is, 650 °C, it is preferable to set the conditions for the region of region 104D, and more preferably to set the conditions for region 100C. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive substance preferably satisfies 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 preferably satisfies 1.9 ≤ Y ≤ 12, and more preferably 2.2 ≤ Y ≤ 3.2.

[0065] As shown in Fig. 7, when the temperature is 773 K, that is, 500 °C, by setting the molar ratio of region 100E, a substance with only a fluorite structure can be produced, and by setting the molar ratio of region 104E, a substance containing a stabilized fluorite structure can be produced.

[0066] As shown in Fig. 7, when the temperature is 773 K, that is, 500 °C, it is preferable to set the conditions for the region of region 104E, and more preferably to set the conditions for region 100E. As an example, when the molar amount of uranium is 1, the molar amount X of the radioactive substance preferably satisfies 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 preferably satisfies 1.9 ≤ Y ≤ 12, and more preferably 2.3 ≤ Y ≤ 3.2.

[0067] As shown in Fig. 8, when the temperature is 623 K, that is, 350 °C, by setting the molar ratio of region 100F, a substance with only a fluorite structure can be produced, and by setting the molar ratio of region 104F, a substance containing a stabilized fluorite structure can be produced.

[0068] As shown in Fig. 8, when the temperature is 623 K, that is, 350 °C, it is preferable to set the conditions for the region 104F, and more preferably to set the conditions for the region 100F. As an example, when the molar amount of uranium is set to 1, the molar amount X of the radioactive substance is preferably 0 < X ≤ 6.3, and more preferably 0.2 ≤ X ≤ 0.4. Also, when the molar amount of uranium is set to 1, the molar amount Y of oxygen is preferably 1.9 ≤ Y ≤ 12, and more preferably 2.3 ≤ Y ≤ 3.

[0069] From the above, when the processing device is performing the processing in the stabilization processing unit, when the molar amount of the radioactive substance (MA) is X with the molar amount of uranium being 1 in the solidified body, it is preferably 0 < X ≤ 7, and more preferably 0 < X ≤ 6.3. By setting the molar ratio of uranium to MA within the above range, a substance containing a stabilized fluorite structure in the regions 104A to 104F can be produced, and the crystal can be stabilized. Also, the processing devices 10 and 10A can suppress the change in the crystal state during long-term storage by producing a crystal that satisfies the conditions of the region 104F. Further, in the stabilization processing unit, when the molar amount X of the radioactive substance (MA) with the molar amount of uranium being 1 in the solidified body is 0 < X ≤ 7, and the molar amount Y of oxygen with the molar amount of uranium being 1 is 1.9 ≤ Y ≤ 12, it is preferable. By setting the molar ratio of uranium, MA, and oxygen within the above range, a substance containing a stabilized fluorite structure in the regions 104A to 104F can be produced, and the crystal can be stabilized.

[0070] Further, in the stabilization treatment unit, when the molar amount of the radioactive substance (MA) is X and the molar amount of uranium is set to 1 in the solidified body, it is preferable that 0 < X ≤ 6.3, and it is more preferable that the temperature of the stabilization treatment is 600°C or higher and 1000°C or lower. Thereby, stable crystals can be formed with a high probability during the stabilization treatment. When the temperature of the stabilization treatment is 600°C or higher and 1000°C or lower, the stabilization treatment unit more preferably satisfies 0.08 ≤ X ≤ 0.5. Further, by setting the ternary system range to the ratio that satisfies the condition where only the fluorite structure is formed under the condition of 500°C, even when the temperature of the stabilization treatment is 600°C or higher and 1000°C or lower, crystals that can be stored for a long time can be obtained. Further, by being able to form only a substance with a fluorite structure, it can be stably stored even when the storage temperature rises to 500°C.

[0071] In the stabilization treatment unit, it is preferable that the molar amount X of the radioactive substance (MA) when the molar amount of uranium in the solidified body is 1 satisfies 0 < X ≤ 7, more preferably 0 < X ≤ 6.3, and even more preferably 0.1 ≤ X ≤ 0.5. Specifically, further, the stabilization treatment unit preferably satisfies the conditions of region 104E of the ternary system phase diagram at 773 K (500°C) shown in FIG. 7, and more preferably satisfies the conditions of region 100E. Thereby, when the temperature of the manufactured substance rises to 500°C during long-term storage, it is possible to suppress a change in the crystal structure, and it can be stably stored for a long time. In the stabilization treatment unit, it is preferable that the molar amount X of the radioactive substance (MA) when the molar amount of uranium in the solidified body is 1 satisfies 0.1 ≤ X ≤ 0.5, and the molar amount Y of oxygen when the molar amount of uranium is set to 1 satisfies 2.3 ≤ Y ≤ 3.2. Thereby, only a fluorite structure can be preferably produced.

[0072] As described above, the present invention has been described with reference to the embodiments, but the present disclosure is not limited to the above embodiments. Each configuration and their combinations in the above embodiments are examples, and additions, omissions, substitutions, and other changes in the configuration are possible without departing from the spirit of the present invention.

[0073] The present disclosure discloses the following inventions. Note that it is not limited to the following. (1) A uranium supply unit that supplies uranium to a liquid from which a radioactive substance containing at least one of minor actinoids and lanthanoids has been extracted; A solidification treatment unit that heats the liquid to which uranium has been supplied and from which the minor actinoids and the lanthanoids have been extracted, evaporates a part of the liquid, and then further heats it to produce a solidified body; A stabilization treatment unit that sinters the produced solidified body, removes some or all of the carbon, hydrogen, oxygen, and nitrogen components, and generates a fluorite structure of uranium dioxide incorporating the minor actinoids and the lanthanoids therein; and A high-level radioactive substance treatment system in which, in the stabilization treatment unit, the molar amount X of the radioactive substance when the molar amount of uranium is taken as 1 in the solidified body satisfies 0 < X ≤ 7.

[0074] (2) The high-level radioactive substance treatment system according to (1), in which, in the stabilization treatment unit, the molar amount Y of oxygen when the molar amount of uranium is taken as 1 satisfies 1.9 ≤ Y ≤ 12.

[0075] [[ID=1,6]](3) The high-level radioactive substance treatment system according to (1) or (2), in which, in the stabilization treatment unit, the molar amount X of the radioactive substance when the molar amount of uranium is taken as 1 in the solidified body satisfies 0 < X ≤ 6.3, and the temperature of the stabilization treatment is 600°C or higher and 1000°C or lower.

[0076] (4) The high-level radioactive substance treatment system according to any one of (1) to (3), in which, in the stabilization treatment unit, the molar amount X of the radioactive substance when the molar amount of uranium is taken as 1 in the solidified body satisfies 0.1 ≤ X ≤ 0.5.

[0077] (5) The high-level radioactive substance treatment system according to (4), in which, in the stabilization treatment unit, the molar amount Y of oxygen when the molar amount of uranium is taken as 1 satisfies 2.3 ≤ Y ≤ 3.2.

[0078] (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 the stabilization processing, and the amount of oxygen is controlled by the amount of reducing gas supplied.

[0079] (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 lanthanoids from a liquid containing high-level radioactive materials, extracts the organic solvent containing the minor actinides and the lanthanoids, and supplies a diluting liquid to produce a diluted liquid, wherein the uranium supply unit adds a uranium complex that dissolves in the organic solvent to the liquid.

[0080] (8) A high-level radioactive material treatment system according to (7), wherein the diluting liquid has a boiling point of 30°C or higher and 100°C or lower.

[0081] (9) A high-level radioactive material processing system according to (7) or (8), wherein the diluted liquid has no flash point or a flash point equal to or higher than the decomposition temperature of the extractant.

[0082] (10) An extraction device that uses an organic solvent liquid to extract the minor actinides and the lanthanoids from a liquid containing high-level radioactive materials, mixes the extracted organic solvent with a diluent containing a liquid-phase stripping agent, and generates a liquid in which the minor actinides and the lanthanoids have been transferred to the diluent containing the liquid-phase stripping agent, The high-level radioactive material processing system according to any one of (1) to (6), wherein the uranium supply unit adds uranium ions that dissolve in the liquid phase to the liquid.

[0083] (11) A high-level radioactive material processing system according to (10), wherein the solidification device produces a solidified body by heat treatment.

[0084] A step of supplying uranium at a ratio such that, when the molar amount of the radioactive substance in the liquid from which at least one of minor actinides and lanthanoids has been extracted is X with the molar amount of uranium being 1, 0 < X ≤ 7; A step of producing a solidified body by heating the liquid supplied with uranium and from which the minor actinides and the lanthanoids have been extracted to evaporate a part of the liquid and then further heating to solidify it; A step of sintering the produced solidified body, removing part or all of carbon, hydrogen, oxygen, and nitrogen components, and performing a stabilization treatment for generating a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanoids therein in one apparatus, the method for treating high-level radioactive substances including these steps.

Explanation of Signs

[0085] 10, 10A Treatment system (high-level radioactive substance treatment system) 12, 12A Extraction device 14, 14A Solidification device 16, 16A Stabilization device 18 Storage device 22 Waste liquid supply unit 24 Extractant supply unit 26 Diluent supply unit 28 MA extraction liquid generation unit 30 MA back-extraction liquid generation unit 34 Hydrothermal treatment accelerator supply unit 40, 40A Solidification treatment unit 44, 44A Uranium supply unit 45, 45A Distillation unit 46, 46A Heat treatment unit 47, 47A Diluent recovery unit 48, 48A Oxidizing gas supply unit 49, 49A Exhaust gas recovery unit 50, 50A Stabilization treatment unit 52, 52A Reduction unit 54, 54A Reduction gas supply unit 56, 56A Exhaust gas recovery unit 60 Storage unit

Claims

1. a uranium supply unit that supplies uranium to the liquid from which radioactive materials containing at least one of minor actinides and lanthanides have been extracted; a solidification processing unit that supplies uranium, heats the liquid from which the minor actinides and the lanthanoids have been extracted, evaporates 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 part or all of the carbon, hydrogen, oxygen, and nitrogen components, thereby producing a fluorite structure of uranium dioxide incorporating the minor actinides and the lanthanoids therein; The stabilization treatment unit is a high-level radioactive material treatment system in which the molar amount X of the radioactive material in the solidified body, when the molar amount of uranium is 1, is 0<X≦7.

2. 2. The high-level radioactive material processing system according to claim 1, wherein the stabilization treatment section has a molar amount Y of oxygen in the range of 1.9≦Y≦12 when the molar amount of uranium is taken as 1.

3. 2. The high-level radioactive material processing system according to claim 1, wherein the stabilization treatment 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, where X is 0<X≦6.3, and the stabilization treatment temperature is 600°C or higher and 1000°C or lower.

4. 2. The high-level radioactive material processing system according to claim 1, wherein the stabilization treatment unit has a molar amount X of the radioactive material in the solidified body, where X is 0.1≦X≦0.5, where X is the molar amount of uranium in the solidified body.

5. 5. The high-level radioactive material processing system according to claim 4, wherein the stabilization treatment section has a molar amount Y of oxygen in the range of 2.3≦Y≦3.2 when the molar amount of uranium is 1.

6. 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 the stabilization processing, and the amount of oxygen is controlled by the amount of reducing gas supplied.

7. an extraction device that uses an organic solvent liquid to extract the minor actinides and the lanthanoids from a liquid containing a high-level radioactive material, extracts the organic solvent containing the minor actinides and the lanthanoids, and supplies a dilution liquid to produce a diluted liquid; 2. 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.

8. 8. 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.

9. 8. The high-level radioactive material processing system according to claim 7, wherein the diluting liquid has no flash point or a flash point equal to or higher than the decomposition temperature of the extractant.

10. an extraction device that extracts the minor actinides and the lanthanoids from a liquid containing high-level radioactive materials using an organic solvent liquid, and then mixes the extracted organic solvent with a diluent containing a liquid-phase stripping agent to produce a liquid in which the minor actinides and the lanthanoids have been transferred into the diluent containing the liquid-phase stripping agent; 2. 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.

11. The high-level radioactive material processing system according to claim 10, wherein the solidification device produces a solidified body by a heat treatment.

12. supplying uranium to a liquid from which radioactive materials containing at least one of minor actinides and lanthanides have been extracted at a rate such that the molar amount X of the radioactive material satisfies the relationship 0≦X≦7, where X is the molar amount of uranium taken as 1; a step of supplying uranium, heating the liquid from which the minor actinides and the lanthanoids 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 incorporating the minor actinides and the lanthanoids therein.

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