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WO2015059777A1 - Procédé de séparation d'actinides et dispositif de traitement de combustible usagé - Google Patents

Procédé de séparation d'actinides et dispositif de traitement de combustible usagé Download PDF

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Publication number
WO2015059777A1
WO2015059777A1 PCT/JP2013/078638 JP2013078638W WO2015059777A1 WO 2015059777 A1 WO2015059777 A1 WO 2015059777A1 JP 2013078638 W JP2013078638 W JP 2013078638W WO 2015059777 A1 WO2015059777 A1 WO 2015059777A1
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Prior art keywords
spent fuel
ionic liquid
fluoride
solution
dissolving
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Ceased
Application number
PCT/JP2013/078638
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English (en)
Japanese (ja)
Inventor
大輔 渡邉
祐子 可児
笹平　朗
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Hitachi Ltd
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Hitachi Ltd
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Priority to PCT/JP2013/078638 priority Critical patent/WO2015059777A1/fr
Publication of WO2015059777A1 publication Critical patent/WO2015059777A1/fr
Anticipated expiration legal-status Critical
Ceased legal-status Critical Current

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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G21C19/42—Reprocessing of irradiated fuel
    • G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/04—Treating liquids
    • G21F9/06—Processing
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
    • G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
    • G21F9/28—Treating solids
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
    • Y02W30/00—Technologies for solid waste management
    • Y02W30/50—Reuse, recycling or recovery technologies

Definitions

  • the present invention relates to a method and apparatus for separating spent fuel components into uranium and plutonium, minor actinides and fission products, that is, a separation method and spent fuel processing apparatus for separating actinides.
  • the spent fuel discharged from the nuclear power plant is planned to be processed into vitrified material and disposed of after the nuclear fuel material is recovered by reprocessing.
  • the vitrified material contains minor actinides, which are radionuclides with a very long half-life, and therefore the vitrified material can reach hundreds of thousands of years by geological disposal. It is said that it needs to be kept stable.
  • the actinide means an element having an atomic number of 89 to 103, and corresponds to an element group called uranium (U), plutonium (Pu), or a minor actinide used as nuclear fuel.
  • the minor actinide refers to an element obtained by removing Pu from a transuranium element among actinide elements, and corresponds to neptunium, americium, curium, and the like.
  • the minor actinides are separated from the high-level radioactive liquid waste generated during the reprocessing of spent fuel, and the neutrons are irradiated to the minor actinides to give a short half-life radionuclide.
  • the technology of converting to is being studied.
  • the high level radioactive liquid waste is a liquid after separation of uranium (U) and Pu from a solution in which spent fuel is dissolved in nitric acid.
  • the high level radioactive liquid waste mainly contains fission products and minor actinides. Is dissolved.
  • uranium and plutonium are separated by reprocessing spent fuel, and minor actinides are separated by a minor actinide separation process from high-level radioactive liquid waste.
  • Each product is separated.
  • Various methods for separating minor actinides from high-level radioactive liquid waste have been studied. For example, methods for selectively separating minor actinides from high-level radioactive liquid waste using an extractant have been studied.
  • an ionic liquid which is a substance consisting only of ions and has a property of becoming a liquid near room temperature, has attracted attention.
  • the ionic liquid can change the combination of the cation and anion which comprise it according to various uses.
  • Typical cations include imidazolium, pyridinium, pyrrolidinium, piperidinium, ammonium, and phosphonium.
  • Representative anions include halide ions (Cl ⁇ , Br ⁇ , I ⁇ ), tetrafluoroborate (BF 4 ⁇ ), hexafluorophosphate (PF 6 ⁇ ), bis (trifluoromethylsulfonyl) amide ( C 2 F 6 NO 4 S 2 ⁇ ), trifluoromethanesulfonate (CF 3 O 3 S ⁇ ), trifluoroacetate (CF 3 COO ⁇ ) and the like.
  • Japanese Patent Publication No. 2002-503820 discloses a method of separating U and Pu by dissolving a spent fuel or a substance containing a spent fuel component in an ionic liquid.
  • Japanese Patent Publication No. 2001-516871 describes a method of regenerating used metal salt generated when reprocessing irradiated fuel with molten salt with an ionic liquid.
  • Non-Patent Document 1 describes a method of separating various elements by supplying components contained in spent fuel such as uranium oxide to an ionic liquid, blowing chlorine gas into the ionic liquid, and performing electrolysis. .
  • the ionic liquid when reprocessing spent fuel or handling radionuclide-containing substances, the ionic liquid contains hydrogen atoms that tend to decelerate neutrons compared to water-based solvents such as nitric acid solutions. Since the amount of ionic liquid used as a solvent is smaller than the method using water as a solvent, it is less likely to generate decelerated neutrons that contribute to the criticality, that is, it is easier to manage operations without causing criticality. There are benefits.
  • Japanese Patent Publication No. 2002-503820 discloses a method for dissolving a spent fuel or a substance containing its component in an ionic liquid, but data such as the amount of the spent fuel component dissolved in the ionic liquid is shown. Absent. In [Non-Patent Document 1], it is necessary to blow chlorine gas, which is corrosive gas, into the ionic liquid in order to dissolve and separate spent fuel components. In this method, chlorine gas is treated. It is expected that the equipment will be complicated, such as the need to install off-gas treatment equipment. Moreover, since chlorine gas is used in [Non-Patent Document 1], it is expected that a normal spent fuel whose chemical form is an oxide is hardly dissolved in an ionic liquid as it is. Easily dissolving the spent fuel component in the ionic liquid is also considered as a particular problem of the method of separating the spent fuel component using the ionic liquid.
  • spent fuel is reacted with a fluorinating agent to produce solid fluoride
  • the produced solid fluoride is dissolved in an ionic liquid
  • U and Pu fission from the solution in which the solid fluoride is dissolved in the ionic liquid
  • the present invention also provides a fluorination treatment apparatus for producing a solid fluoride by reacting a spent fuel with a fluorinating agent, a dissolution tank for producing a solution in which the solid fluoride is dissolved in an ionic liquid, and the solution.
  • components of spent fuel can be efficiently dissolved in an ionic liquid, and facilities for separating spent fuel components into U and Pu, fission products, and minor actinides are provided. It can be a simple facility. In the present invention, separating U and Pu, fission products, and minor actinides will be referred to as separating actinides.
  • the present inventors have newly found through experiments that a compound having a chemical form of fluoride has high solubility in an ionic liquid.
  • Table 1 shows the amount of fluoride dissolved in the ionic liquid obtained by this experiment.
  • the vertical axis in Table 1 represents tetrafluoroborate (BF 4 ⁇ ), hexafluorophosphate (PF 6 ⁇ ), bis (trifluoromethylsulfonyl) amide (C 2 F 6 NO 4 S 2 ⁇ ) as anions.
  • Trifluoromethanesulfonate (CF 3 O 3 S ⁇ ), chloride ion (Cl ⁇ ) were used in the experiment.
  • anion species are arranged in the Lewis basic order of anions, and trifluoroacetate (CF 3 COO ⁇ ) is described as a reference.
  • cerium fluoride was typically supplied to each ionic liquid, dissolved at a temperature of 100 ° C. with stirring for 1 hour, and then the amount of cerium dissolved in the solution was measured. . It is said that ionic liquids have been developed since the 1990s, and there are almost no examples in which the amount of inorganic compounds dissolved in such ionic liquids has been reported.
  • an ionic liquid having an imidazolium cation has a strong Lewis basic ionic liquid and tends to have a high dissolution amount, and an anion of BF 4 ⁇ or C 2 F 6 NO 4 S 2 ⁇ .
  • An ionic liquid having a low Lewis acidity tends to increase the amount of dissolution. Therefore, it is considered that fluoride is best dissolved in an ionic liquid composed of a cation with weak Lewis acidity and an anion with strong Lewis basicity.
  • the weakest BF 4 cation and the Lewis basicity of the Lewis acid strongest imidazolium - even ionic liquids composed of a combination of anion, since it is possible to dissolve the fluoride, at least Table 1 It is considered that the fluoride can be dissolved if the ionic liquid is a combination of an anion and a cation shown in FIG.
  • the present inventors have devised a process for separating actinides using the newly acquired knowledge that fluoride dissolves in ionic liquids.
  • the spent fuel is reacted with fluorine gas, the spent fuel is converted into a form of fluoride that is easily dissolved in the ionic liquid, and the fluoride residue containing the actinide is dissolved in the ionic liquid,
  • the dissolved elements are separated by operations such as solvent extraction.
  • the apparatus for reacting the spent fuel with fluorine gas is referred to in the flame furnace system referred to in Japanese Patent Application Laid-Open Nos. 2004-233066 and 2012-47546, and in Japanese Patent Application Laid-Open No. 2013-101666.
  • a batch type fluorination apparatus can be used. By using such a fluorination apparatus, it is possible to react spent fuel with fluorine gas and convert it into fluoride.
  • Example 1 A first embodiment of the present invention will be described with reference to FIG.
  • FIG. 1 is a flowchart of the present embodiment showing the steps until the actinide is separated from the spent fuel.
  • a fluorination step 1 for fluorinating the spent fuel 4 a dissolution step 2 for dissolving the fluoride residue (solid fluoride) 7 obtained in the fluorination step 1 with an ionic liquid 8, and dissolution
  • a separation step 3 for separating U and Pu10, fission product 11, and minor actinide 12 from the solution 9 obtained in step 2 is provided.
  • the spent fuel 4 discharged from the nuclear power plant is stored in a cladding tube, it is sheared and decoated by an existing method implemented in a reprocessing plant.
  • the spent fuel 4 thus obtained is reacted with the fluorine gas 5 in the fluorination step 1.
  • a flame furnace type or batch type fluorination apparatus can be used.
  • the fluorine gas 5 is very reactive, and almost all the components of the spent fuel 4 react with the fluorine gas 5 and the chemical form is converted from oxide to fluoride.
  • the spent fuel 4 contains various elements and the boiling points of the fluorides differ greatly, in the fluorination step 1, mainly volatile uranium hexafluoride (UF 6 gas 6) and fluoride are used. It is divided roughly into residue 7.
  • the fluoride residue 7 contains Pu, fission products, minor actinides, and small amounts of U.
  • the fluoride residue 7 is collected, and the fluoride residue 7 is dissolved in the ionic liquid 8 in the dissolution step 2 to obtain a solution 9.
  • the ionic liquid 8 any of ionic liquids in which cations and anions shown in Table 1 are combined can be used.
  • Table 1 shows the amount of cerium fluoride dissolved in each ionic liquid as a representative example, but the actinide element, which is an element for separation in this example, is an element of the same group as cerium.
  • the actinide fluoride is considered to dissolve in the ionic liquid as well.
  • the elements dissolved in the solution 9 are separated by various separation methods using differences in chemical properties of the elements and the size of the compounds, and U and Pu10, fission products 11, Each minor actinide 12 can be obtained.
  • the separation method an extraction separation method in which an element is selectively recovered using an extractant, an electrolysis method, an adsorption method in which an element is selectively adsorbed on an adsorbent, and the like are effective.
  • U and Pu10 are reused as nuclear fuel through treatment such as oxide conversion.
  • the fission product 11 is vitrified.
  • Minor actinides 12 are transmutated by neutron irradiation or the like, converted into nuclides with a short half-life, and then solidified in the same manner as fission products.
  • U and Pu10 and minor actinide 12 are separated. However, according to the usage of these elements, Pu and minor actinide 12 may be collected together without being separated. .
  • the fluorine gas 5 is used as a fluorinating agent for converting the spent fuel 4 into fluoride.
  • boron fluoride, carbon fluoride, nitrogen fluoride, chlorine fluoride, bromine fluoride, fluorine The same effect can be obtained by using a fluorinating agent such as iodine fluoride.
  • the spent fuel component can be dissolved by a simple method that does not require an operation such as blowing a gas into the ionic liquid by converting the spent fuel component into a chemical form of fluoride.
  • an operation such as blowing a gas into the ionic liquid by converting the spent fuel component into a chemical form of fluoride.
  • the spent fuel processing apparatus of the present embodiment that separates minor actinides from spent fuel will be described with reference to FIG.
  • the spent fuel processing apparatus of the present embodiment includes a flame furnace 20, a residue receiving tank / dissolution tank 21, a valve A22, a valve B23, a valve C24, a valve D25, a valve E27, a pump 26, and a separation device 28.
  • a valve A22 and a valve B23 are installed between the flame furnace 20 and the residue receiving and melting tank 21. By opening the valve A22 and the valve B23, the flame furnace 20 and the residue receiving tank / dissolving tank 21 are connected, and by closing the valve A22 and the valve B23, the pipe between the frame furnace 20 and the residue receiving tank / dissolving tank 21 is closed.
  • a valve C24 is installed between a collection container (not shown) for collecting the UF 6 gas 6 and the residue receiving tank / dissolution tank 21. By opening the valve C24, the gas UF 6 gas 6 is transferred to the outside of the residue receiving tank / dissolving tank 21 and collected in a recovery container. By closing the valve C24, the UF 6 gas is transferred to the residue receiving tank / dissolving tank 21. From being transferred to the collection container.
  • a valve D25 is installed between a tank (not shown) in which the ionic liquid 8 is stored and the residue receiving tank / dissolving tank 21. By opening the valve D25, the ionic liquid 8 is injected into the residue receiving tank / dissolving tank 21.
  • valve D25 By closing the valve D25, the injection of the ionic liquid 8 into the residue receiving tank / dissolving tank 21 is stopped.
  • a valve E27 is installed between the pump 26 and the residue receiving / dissolving tank 21. By opening the valve E27 while the pump 26 is driven, the solution 9 in which the fluoride residue 7 is dissolved in the ionic liquid 8 can be taken out of the residue receiving tank / dissolving tank 21. By closing the valve E27, the removal of the solution 9 from the residue receiving tank / dissolving tank 21 is stopped.
  • a separation device 28 is connected to the subsequent stage of the pump 26.
  • UF 6 gas 6 and fluoride residue 7 are generated by fluorination treatment of the spent fuel 4. Since the fluoride residue 7 is solid, it falls from the lower part of the frame furnace 20 and is collected in the residue receiving tank / dissolving tank 21. Since the UF 6 gas 6 is a gas, it passes through the frame furnace 20 and the residue receiving tank / dissolving tank 21 and is transferred to the outside of the residue receiving tank / dissolving tank 21 through the valve C24. The UF 6 gas 6 is collected separately, purified and concentrated, and then reused as nuclear fuel.
  • a stirring device for stirring the solution 9 may be installed in the residue receiving and dissolving tank 21. By providing the stirring device, it is possible to shorten the time for the components contained in the fluoride residue 7 to dissolve in the ionic liquid 8.
  • a heat exchanger for adjusting the temperature of the ionic liquid 8 or the solution 9 may be installed in the residue receiving tank / dissolution tank 21.
  • the heat exchanger By heating the temperature of the solution 9 in which the ionic liquid 8 or the fluoride residue 7 is dissolved by the heat exchanger, the time during which the fluoride residue 7 is dissolved in the solution 9 can be shortened.
  • the operation of transferring the solution 9 in the residue receiving and dissolving tank 21 to the separation device 28 is performed.
  • the solution 9 in the residue receiving tank / dissolving tank 21 is transferred to the separation device 28.
  • Separation device 28 separates minor actinide 12, uranium and plutonium 10 and fission product 11 from solution 9.
  • the receiving tank for the fluoride residue 7 and the dissolving tank are the same equipment, so that there is no need to recover the fluoride residue 7 from the receiving tank and transfer it to another dissolving tank. The benefits can be obtained.
  • the spent fuel processing apparatus of the present embodiment is a very simple facility because it is not necessary to blow chlorine gas or the like in the dissolving step 2 for dissolving the fluoride residue 7 obtained from the spent fuel into the ionic liquid. be able to.
  • An example of recovering nuclear fuel material by the PUREX method after reacting spent fuel with fluorine gas as in JP-A-2002-257980 can be considered.
  • An example in which this method is combined with a method of selectively separating minor actinides from high-level radioactive liquid waste using an extractant is also considered as Comparative Example 1.
  • the dissolution process 2 of the present example does not use water, so that the effect that the critical control is easy can be obtained.
  • Example 2 a method in which spent fuel is reacted with fluorine gas and solid fluoride is dissolved in a molten salt and elements are separated by electrolysis as disclosed in Japanese Patent Application Laid-Open No. 2010-127616.
  • the temperature of the dissolution step generally needs to be several hundred degrees Celsius. Since it can melt
  • Example 2 A second embodiment of the present invention will be described with reference to FIG.
  • a flame furnace 20 and a residue receiving tank / dissolving tank 21 are connected via a valve A22 and a valve B23, a recovery container for recovering the UF 6 gas 6, the recovery container and the residue receiving tank was configured to include a valve C24 which is installed between and dissolving tank 21, but spent fuel processing apparatus of the present embodiment is a configuration without a collecting container for collecting the frame oven and UF 6 gas 6.
  • the spent fuel processing apparatus of the present embodiment will be described below with a focus on the configuration different from that of the first embodiment.
  • the spent fuel processing apparatus of this embodiment includes a dissolution tank 30, a heat exchanger 33, a pump 26, a valve E27, a valve F31, and a valve G32.
  • a valve F31 is installed between the fluoride residue container (not shown) in which the fluoride residue 7 is stored and the dissolution tank 30.
  • the fluoride residue container and the dissolution tank 30 are connected by opening the valve F31, and the piping between the fluoride residue container and the dissolution tank 30 is closed by closing the valve F31.
  • a heat exchanger 33 is installed in the dissolution tank 30.
  • the spent fuel is fluorinated in a separately installed flame furnace.
  • the fluorination step 1 is performed using a flame furnace
  • the fluorination step may be performed using a batch type fluorination apparatus.
  • the generated fluoride residue 7 is collected in advance and stored in a fluoride residue container.
  • the valve E27 is closed, the valve G32 is opened, and the ionic liquid 8 is supplied to the dissolution tank 30 from a pipe connected to the valve G32.
  • a stirring device for stirring the solution 9 may be installed in the dissolution tank 30. By providing the stirring device, it is possible to shorten the time for the components contained in the fluoride residue 7 to dissolve in the ionic liquid 8.
  • a heating device that adjusts the temperature of the ionic liquid 8 or the solution 9 may be installed in the dissolution tank 30. Since the spent fuel processing apparatus of the present embodiment includes the heat exchanger 33 in the dissolution tank 30, the solution 9 can be heated using the heat exchanger 33. Since the heat exchanger 33 heats the solution 9 in which the ionic liquid 8 or the fluoride residue 7 is dissolved, the time for which the fluoride residue 7 is dissolved in the solution 9 can be shortened.
  • the operation of transferring the solution 9 to the separation device 28 is performed.
  • the solution 9 in the dissolution tank 30 is transferred to the separation device 28.
  • Separation device 28 separates minor actinide 12, uranium and plutonium 10 and fission product 11 from solution 9.
  • the refrigerant 34 may be supplied to the heat exchanger 33.
  • An example of supplying the refrigerant 34 to the heat exchanger 33 will be described.
  • an abnormality such as an earthquake occurs during the operation of dissolving the fluoride residue 7 in the ionic liquid 8 in the dissolution tank 30 and the pipe connected to the dissolution tank 30 is broken, the solution 9 is considered to be at risk of leaking.
  • the ionic liquid of a combination of imidazolium and chloride ions having the highest solubility of cerium fluoride has a melting point near room temperature.
  • the solution 9 is cooled and solidified by supplying the refrigerant 34 to the heat exchanger 33 when an earthquake occurs. Even when an abnormality such as a broken pipe is detected, the solution 9 can be retained in the dissolution tank 30 and leakage to the outside can be prevented.
  • the spent fuel processing apparatus of the present embodiment is configured to include the heat exchanger 33 in the dissolution tank 30, the solution 9 can be solidified in an emergency and the risk of liquid leakage can be eliminated.
  • the spent fuel processing apparatus of this embodiment converts the spent fuel into fluoride and dissolves the spent fuel component in the ionic liquid by dissolving the fluoride in the ionic liquid. It can be made easy.
  • the spent fuel processing apparatus of the present embodiment is a very simple facility because it is not necessary to blow chlorine gas or the like in the dissolving step 2 for dissolving the fluoride residue 7 obtained from the spent fuel into the ionic liquid. be able to.
  • An example of recovering nuclear fuel material by the PUREX method after reacting spent fuel with fluorine gas as in JP-A-2002-257980 can be considered.
  • An example in which this method is combined with a method of selectively separating minor actinides from high-level radioactive liquid waste using an extractant is also considered as Comparative Example 1.
  • the dissolution process 2 of the present example does not use water, so that the effect that the critical control is easy can be obtained.
  • the temperature of the dissolution step generally needs to be several hundred degrees Celsius. Since it can melt

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  • Engineering & Computer Science (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Plasma & Fusion (AREA)
  • Extraction Or Liquid Replacement (AREA)

Abstract

La présente invention porte sur un procédé de séparation d'actinides d'un combustible usagé utilisant un liquide ionique et utilisant un dispositif ayant une configuration d'équipement simple. Le procédé comprend la réaction d'un combustible usagé avec le gaz de fluor pour convertir les formes chimiques des actinides en fluorures qui sont solubles dans un liquide ionique, la dissolution des résidus de fluorures à teneur en actinides dans le liquide ionique et puis la séparation et la récupération des actinides dissous. Le problème de la présente invention peut être résolu à l'aide ce dernier.
PCT/JP2013/078638 2013-10-23 2013-10-23 Procédé de séparation d'actinides et dispositif de traitement de combustible usagé Ceased WO2015059777A1 (fr)

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Cited By (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2016008891A (ja) * 2014-06-25 2016-01-18 株式会社日立製作所 アクチニドの分離方法およびアクチニドの分離装置
CN111584111A (zh) * 2020-05-15 2020-08-25 中国原子能科学研究院 用于乏燃料元件的溶解器及溶解液的处理方法
WO2021178751A3 (fr) * 2020-03-06 2021-12-02 The Board Of Regents Of The Nevada System Of Higher Education On Behalf Of The University Of Nevada, Las Vegas Récupération stoechiométrique d'uf4 à partir d'uf6 dissous dans des liquides ioniques
CN116665942A (zh) * 2023-05-29 2023-08-29 西安交通大学 一种乏燃料核素预分离方法
US11760654B2 (en) 2019-03-29 2023-09-19 The Board Of Regents Of The Nevada System Of Higher Education On Behalf Of The University Of Nevada, Las Vegas Conversion of uranium hexafluoride and recovery of uranium from ionic liquids

Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2002257980A (ja) * 2001-03-02 2002-09-11 Tokyo Electric Power Co Inc:The 使用済原子燃料の再処理方法
JP2011107156A (ja) * 2011-01-24 2011-06-02 Japan Atomic Energy Agency イオン液体を用いたウランの回収方法
JP2013101066A (ja) * 2011-11-09 2013-05-23 Hitachi-Ge Nuclear Energy Ltd ジルコニウムを含む使用済燃料の処理装置および方法

Patent Citations (3)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2002257980A (ja) * 2001-03-02 2002-09-11 Tokyo Electric Power Co Inc:The 使用済原子燃料の再処理方法
JP2011107156A (ja) * 2011-01-24 2011-06-02 Japan Atomic Energy Agency イオン液体を用いたウランの回収方法
JP2013101066A (ja) * 2011-11-09 2013-05-23 Hitachi-Ge Nuclear Energy Ltd ジルコニウムを含む使用済燃料の処理装置および方法

Cited By (6)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2016008891A (ja) * 2014-06-25 2016-01-18 株式会社日立製作所 アクチニドの分離方法およびアクチニドの分離装置
US11760654B2 (en) 2019-03-29 2023-09-19 The Board Of Regents Of The Nevada System Of Higher Education On Behalf Of The University Of Nevada, Las Vegas Conversion of uranium hexafluoride and recovery of uranium from ionic liquids
WO2021178751A3 (fr) * 2020-03-06 2021-12-02 The Board Of Regents Of The Nevada System Of Higher Education On Behalf Of The University Of Nevada, Las Vegas Récupération stoechiométrique d'uf4 à partir d'uf6 dissous dans des liquides ioniques
CN111584111A (zh) * 2020-05-15 2020-08-25 中国原子能科学研究院 用于乏燃料元件的溶解器及溶解液的处理方法
CN116665942A (zh) * 2023-05-29 2023-08-29 西安交通大学 一种乏燃料核素预分离方法
CN116665942B (zh) * 2023-05-29 2024-01-23 西安交通大学 一种乏燃料核素预分离方法

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