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US20040244533A1 - Actinide production - Google Patents

Actinide production Download PDF

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Publication number
US20040244533A1
US20040244533A1 US10/479,730 US47973004A US2004244533A1 US 20040244533 A1 US20040244533 A1 US 20040244533A1 US 47973004 A US47973004 A US 47973004A US 2004244533 A1 US2004244533 A1 US 2004244533A1
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United States
Prior art keywords
uranium
process according
metals
metal
oxide
Prior art date
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Abandoned
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US10/479,730
Inventor
Rober Lewin
Robert Thied
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National Nuclear Laboratory Ltd
Sellafield Ltd
Original Assignee
Individual
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Filing date
Publication date
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Assigned to BRITISH NUCLEAR FUELS PLC reassignment BRITISH NUCLEAR FUELS PLC ASSIGNMENT OF ASSIGNORS INTEREST (SEE DOCUMENT FOR DETAILS). Assignors: LEWIS, ROBERT GLYNN, THIED, ROBERT CHARLES
Publication of US20040244533A1 publication Critical patent/US20040244533A1/en
Assigned to BNFL (IP) LIMITED reassignment BNFL (IP) LIMITED ASSIGNMENT OF ASSIGNORS INTEREST (SEE DOCUMENT FOR DETAILS). Assignors: BRITISH NUCLEAR FUELS PLC
Assigned to NEXIA SOLUTIONS LTD. reassignment NEXIA SOLUTIONS LTD. ASSIGNMENT OF ASSIGNORS INTEREST (SEE DOCUMENT FOR DETAILS). Assignors: BNFL (IP) LIMITED
Abandoned legal-status Critical Current

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Classifications

    • 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
    • G21C19/48—Non-aqueous processes
    • 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

  • This invention relates to methods for the production of metals from oxides present in spent nuclear fuels and is particularly applicable to the production of actinides, specifically uranium and metals more noble than uranium, from actinide oxides which are present in irradiated nuclear fuels.
  • Methods of the present invention can be used in the treatment of irradiated fuels for producing actinides in metallic form suitable for use as feeds in subsequent electrorefining processes.
  • molten salts is intended to cover salts such as lithium chloride which melts at an elevated temperature and also ionic liquids which typically are liquid at room temperature or which melt at a temperature up to about 100° C.
  • the Dimitrovgrad SSC-RIAR process makes use of chemical oxidants (chlorine and oxygen gases) to react with powdered uranium dioxide fuel to form higher oxidation state compounds such as UO 2 Cl 2 which are soluble in the molten salt.
  • chemical oxidants chlorine and oxygen gases
  • UO 2 Cl 2 powdered uranium dioxide fuel
  • the second process developed by the Argonne National Laboratory (ANL) is fundamentally an electrorefining technology which uses current flow to anodically oxidise uranium to form uranium ions in the molten salt electrolyte. At the cathode the uranium is reduced and electrodeposited as uranium metal.
  • the ANL process requires a metal feed. If oxide fuels are to be treated, it is necessary to reduce the uranium oxide (usually UO 2 pellets) to the metal. This reduction process is carried out chemically, using lithium metal in a LiCl or LiCl/KCl molten salt at 500 to 600° C. Alternatively, a salt transport process can be used involving a Cu—Mg—Ca alloy and molten CaCl 2 salt. However, in both reduction methods the by-products, Li 2 O and CaO respectively, need to be recovered from the molten salt phase by an electrolysis step. Effectively this means a two stage process.
  • a disadvantage of the lithium reduction process for producing a metallic feed from an oxide is the production of Li 2 O by-product. This requires recycle to make the process economic, and this is done by an electrolytic recovery of lithium metal. Hence this is a two stage process, comprising a reduction step followed by a lithium recovery stage.
  • the process thereby involves the use of a single electrochemical process to reduce the metal oxide fuel to a metallic form, with oxygen produced as the only by-product.
  • the potential of the cathode is maintained and controlled so that only oxygen ionisation occurs and not the deposition of the cations (eg Ca ions) in the fused salt.
  • the oxide comprises an actinide oxide, such as uranium oxide or irradiated uranium oxide.
  • the present invention seeks to provide a method for the treatment of irradiated fuel which allows for the separation of uranium, and metals more noble than uranium, from such mixtures as found in spent nuclear fuel, and to provide these metals in a form suitable for use as the feed in a molten salt electrorefining process, whilst ensuring that other, more electropositive, metals remain in the form of oxides.
  • a process for reducing to metallic form oxides of uranium, or metals more noble than uranium, present in spent nuclear fuel comprising a mixture of metal oxides comprising cathodically electrolysing the oxide in the presence of a molten salt electrolyte, the potential of the cathode being controlled so as to favour oxygen ionisation over deposition of metal from the cations present in the molten salt, and to ensure that reduction of metals other than uranium or metals more noble than uranium does not occur.
  • the invention provides a single electrochemical process to reduce the metal oxide fuel to a metallic form, with oxygen produced as the only by-product.
  • the potential of the cathode is maintained and controlled so that only oxygen ionisation occurs and not the deposition of the cations (eg Ca ions) in the fused salt, and also to ensure that, whilst reduction of uranium or metals more noble than uranium occurs smoothly, the less noble metals are not reduced and remain in the anode as oxides.
  • the mixture of oxides includes an actinide oxide, such as uranium oxide or irradiated uranium oxide, or mixed uranium/plutonium oxides.
  • the uranium oxide is commonly uranium dioxide.
  • the oxide may be in any physical form, and this is generally dependent on the particular chemical nature of the spent nuclear fuel and the processing to which the material has previously been subjected.
  • the fuel may comprise a powder, an amorphous mass, or a dense solid agglomerate.
  • the material may be treated according to the method of the present invention by connection to an electrical circuit such that it serves as the cathode during electrolysis. Connection to the circuit may be effected by any of the standard means well known to those skilled in the art.
  • the oxide fuel is in contact with the cathode of an electrochemical cell.
  • the cathode could be in the form of a mesh basket.
  • the molten salt electrolyte may be any suitable molten salt or mixture of such salts, for instance chloride salts, preferably CaCl 2 and/or BaCl 2 .
  • the anode may be any suitable inert anode, such as carbon.
  • the oxide fuel may be first treated mechanically to remove its zircaloy cladding before it is added to the electrolytic cell.
  • the zircaloy cladding may be treated with the fuel.
  • the fuel may require to be sheared into sections of small length prior to treatment in order to expose the oxide fuel to the molten salt.
  • an electrolytic cell which has a carbon anode and a mesh basket cathode. Irradiated oxide fuel is placed in the mesh basket.
  • the electrolyte consists of a molten salt or a mixture of such salts comprising, for example, chloride salts such as CaCl 2 or BaCl 2 .
  • a voltage is applied between the cathode and the anode. At the cathode the reaction involves the diffusion of oxygen atoms to the surface of the solid, followed by ionisation according to the reaction:
  • the oxide ions which are produced dissolve in the electrolyte and are transferred to the anode where they are re-oxidised to produce oxygen gas.
  • the potential at the cathode is controlled, via a third reference electrode, to ensure that the reaction occurring at the cathode is oxygen ionisation and not deposition of the cations in the fused salt, and that only selected metal oxides—specifically those of uranium and of metals more noble than uranium—are reduced. Electrolysis at elevated temperatures results in an increased rate of oxygen diffusion, thereby also encouraging ionisation rather than metal deposition.
  • the irradiated fuel is left in the form of a metal/metal oxide solid mixture at the cathode, with uranium and more noble metals having been reduced to the metallic form, whilst the less noble metals remain in the form of their oxides.
  • This metallic/metal oxide product which contains fission products, can be removed and used directly as the feed for an electrorefining process. The remaining components of the cell may be re-used immediately without the need for any cleaning.
  • a potential is applied to the metal/metal oxide mixture at the anode such that only uranium metal enters the salt, whilst the less noble metals remain behind as oxides. Insufficient potential is applied to encourage the dissolution of metals more noble than uranium.
  • the advantage of the process of the present invention is that it is effectively a single stage process. It is used for the treatment of irradiated mixed metal oxide nuclear fuel, possibly in the form of pellets and, most particularly, is applied to fuels which contain uranium oxide, and mixed uranium and plutonium fuels.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Electrolytic Production Of Metals (AREA)
  • Manufacture And Refinement Of Metals (AREA)

Abstract

The invention provides a process for reducing to metallic form oxides of uranium, or metals more noble than uranium, present in spent nuclear fuel comprising a mixture of metal oxides, the process comprising cathodically electrolysing the oxide in the presence of a molten salt electrolyte, the potential of the cathode being controlled so as to favour oxygen ionisation over deposition of the metal from the cations present in the molten salt, and to ensure than reduction of metals other than uranium or metals more noble than uranium does not occur. The invention allows for the reduction of uranium or metals more than uranium present in spent nuclear fuel comprising mixed metal oxides to a metallic form by the use of a single electrochemical process, with oxygen being produced as the only by-product.

Description

    FIELD OF THE INVENTION
  • This invention relates to methods for the production of metals from oxides present in spent nuclear fuels and is particularly applicable to the production of actinides, specifically uranium and metals more noble than uranium, from actinide oxides which are present in irradiated nuclear fuels. Methods of the present invention can be used in the treatment of irradiated fuels for producing actinides in metallic form suitable for use as feeds in subsequent electrorefining processes. [0001]
    • BACKGROUND TO THE INVENTION
  • In the established art, two processes have been developed for the treatment of irradiated nuclear fuel making use of molten salts. As used herein, the term “molten salts” is intended to cover salts such as lithium chloride which melts at an elevated temperature and also ionic liquids which typically are liquid at room temperature or which melt at a temperature up to about 100° C. [0002]
  • The Dimitrovgrad SSC-RIAR process makes use of chemical oxidants (chlorine and oxygen gases) to react with powdered uranium dioxide fuel to form higher oxidation state compounds such as UO[0003] 2Cl2 which are soluble in the molten salt. In an electrochemical cell the uranium compounds are reduced to UO2 at the cathode, forming a dendritic deposit. This process has both technical and environmental limitations.
  • The second process, developed by the Argonne National Laboratory (ANL) is fundamentally an electrorefining technology which uses current flow to anodically oxidise uranium to form uranium ions in the molten salt electrolyte. At the cathode the uranium is reduced and electrodeposited as uranium metal. [0004]
  • The ANL process requires a metal feed. If oxide fuels are to be treated, it is necessary to reduce the uranium oxide (usually UO[0005] 2 pellets) to the metal. This reduction process is carried out chemically, using lithium metal in a LiCl or LiCl/KCl molten salt at 500 to 600° C. Alternatively, a salt transport process can be used involving a Cu—Mg—Ca alloy and molten CaCl2 salt. However, in both reduction methods the by-products, Li2O and CaO respectively, need to be recovered from the molten salt phase by an electrolysis step. Effectively this means a two stage process.
  • A disadvantage of the lithium reduction process for producing a metallic feed from an oxide is the production of Li[0006] 2O by-product. This requires recycle to make the process economic, and this is done by an electrolytic recovery of lithium metal. Hence this is a two stage process, comprising a reduction step followed by a lithium recovery stage.
  • In co-pending patent application no. PCT/GB00/04604 there is disclosed a single step process for reducing to metallic form a metal oxide present in spent nuclear fuel, the process comprising cathodically electrolysing the oxide in the presence of a molten salt electrolyte, the potential of the cathode being controlled so as to favour oxygen ionisation over deposition of the metal from the cations present in the molten salt. [0007]
  • The process thereby involves the use of a single electrochemical process to reduce the metal oxide fuel to a metallic form, with oxygen produced as the only by-product. The potential of the cathode is maintained and controlled so that only oxygen ionisation occurs and not the deposition of the cations (eg Ca ions) in the fused salt. Typically, the oxide comprises an actinide oxide, such as uranium oxide or irradiated uranium oxide. [0008]
  • In the nuclear industry, however, it is often necessary to separate metals from a mixture of metal oxides such as occurs in spent nuclear fuel. Thus mixtures of uranium and plutonium oxides, together with the oxides of other actinide metals, may additionally be contaminated with oxides of other, chemically active, metals such as, for example, those associated with zircalloy cladding. The present invention seeks to provide a method for the treatment of irradiated fuel which allows for the separation of uranium, and metals more noble than uranium, from such mixtures as found in spent nuclear fuel, and to provide these metals in a form suitable for use as the feed in a molten salt electrorefining process, whilst ensuring that other, more electropositive, metals remain in the form of oxides. [0009]
    • STATEMENTS OF INVENTION
  • Thus, according to the present invention there is provided a process for reducing to metallic form oxides of uranium, or metals more noble than uranium, present in spent nuclear fuel comprising a mixture of metal oxides, the process comprising cathodically electrolysing the oxide in the presence of a molten salt electrolyte, the potential of the cathode being controlled so as to favour oxygen ionisation over deposition of metal from the cations present in the molten salt, and to ensure that reduction of metals other than uranium or metals more noble than uranium does not occur. [0010]
  • The invention provides a single electrochemical process to reduce the metal oxide fuel to a metallic form, with oxygen produced as the only by-product. The potential of the cathode is maintained and controlled so that only oxygen ionisation occurs and not the deposition of the cations (eg Ca ions) in the fused salt, and also to ensure that, whilst reduction of uranium or metals more noble than uranium occurs smoothly, the less noble metals are not reduced and remain in the anode as oxides. Typically, the mixture of oxides includes an actinide oxide, such as uranium oxide or irradiated uranium oxide, or mixed uranium/plutonium oxides. The uranium oxide is commonly uranium dioxide. [0011]
  • The oxide may be in any physical form, and this is generally dependent on the particular chemical nature of the spent nuclear fuel and the processing to which the material has previously been subjected. For example, the fuel may comprise a powder, an amorphous mass, or a dense solid agglomerate. In any event, the material may be treated according to the method of the present invention by connection to an electrical circuit such that it serves as the cathode during electrolysis. Connection to the circuit may be effected by any of the standard means well known to those skilled in the art. [0012]
  • Preferably the oxide fuel is in contact with the cathode of an electrochemical cell. The cathode could be in the form of a mesh basket. The molten salt electrolyte may be any suitable molten salt or mixture of such salts, for instance chloride salts, preferably CaCl[0013] 2 and/or BaCl2.
  • The anode may be any suitable inert anode, such as carbon. In a process of the present invention the oxide fuel may be first treated mechanically to remove its zircaloy cladding before it is added to the electrolytic cell. Alternatively, the zircaloy cladding may be treated with the fuel. The fuel may require to be sheared into sections of small length prior to treatment in order to expose the oxide fuel to the molten salt.[0014]
    • DETAILED DESCRIPTION OF THE INVENTION
  • In order to carry out an embodiment of the present invention, an electrolytic cell is assembled which has a carbon anode and a mesh basket cathode. Irradiated oxide fuel is placed in the mesh basket. The electrolyte consists of a molten salt or a mixture of such salts comprising, for example, chloride salts such as CaCl[0015] 2 or BaCl2. A voltage is applied between the cathode and the anode. At the cathode the reaction involves the diffusion of oxygen atoms to the surface of the solid, followed by ionisation according to the reaction:
  • O+2e −→O2−.
  • The oxide ions which are produced dissolve in the electrolyte and are transferred to the anode where they are re-oxidised to produce oxygen gas. The potential at the cathode is controlled, via a third reference electrode, to ensure that the reaction occurring at the cathode is oxygen ionisation and not deposition of the cations in the fused salt, and that only selected metal oxides—specifically those of uranium and of metals more noble than uranium—are reduced. Electrolysis at elevated temperatures results in an increased rate of oxygen diffusion, thereby also encouraging ionisation rather than metal deposition. [0016]
  • After electrolysis the irradiated fuel is left in the form of a metal/metal oxide solid mixture at the cathode, with uranium and more noble metals having been reduced to the metallic form, whilst the less noble metals remain in the form of their oxides. This metallic/metal oxide product, which contains fission products, can be removed and used directly as the feed for an electrorefining process. The remaining components of the cell may be re-used immediately without the need for any cleaning. [0017]
  • In the electrorefining step, a potential is applied to the metal/metal oxide mixture at the anode such that only uranium metal enters the salt, whilst the less noble metals remain behind as oxides. Insufficient potential is applied to encourage the dissolution of metals more noble than uranium. [0018]
  • Thus, only uranium and metals with similar electronegativities enter the salt and the uranium deposited at the cathode is of a higher purity. Contamination of the cathode product with fission product impurities is also substantially reduced. [0019]
  • In an alternative embodiment in accordance with the present invention the electrolytic ionisation of oxygen and the electrorefining processes are carried out in the same cell and the same salt system. [0020]
  • It is to be emphasised that the advantage of the process of the present invention is that it is effectively a single stage process. It is used for the treatment of irradiated mixed metal oxide nuclear fuel, possibly in the form of pellets and, most particularly, is applied to fuels which contain uranium oxide, and mixed uranium and plutonium fuels. [0021]

Claims (12)

1. A process for reducing to metallic form oxides of uranium, or metals more noble than uranium, present in spent nuclear fuel comprising a mixture of metal oxides, comprising:
cathodically electrolysing the metal oxides in the presence of a molten salt electrolyte; and
controlling the potential of a cathode so as to favour oxygen ionisation over deposition of the metal from the cations present in the molten salt, and to ensure that reduction of metals other than uranium or metals more noble than uranium does not occur.
2. A process according to claim 1 wherein the metal oxides comprise fuel pellets which include uranium oxide, irradiated uranium oxide and/or of mixed uranium/plutonium oxide.
3. A process according to claim 2 wherein the uranium oxide comprises uranium dioxide.
4. A process according to claim 1 wherein the metal oxides are located in a mesh basket which forms the cathode.
5. A process according to claim 1 wherein the molten salt electrolyte comprises at least one chloride salt.
6. A process as claimed in claim 5 wherein the chloride salt is CaCl2 or BaCl2.
7. A process according to claim 1 wherein an anode is a carbon anode.
8. A process according to claim 1 wherein the fuel is treated together with its cladding.
9. A process according to claim 8 wherein the cladding is removed from the fuel prior to treatment.
10. A process according to claim 1 wherein a metal/metal oxide mixture resulting from the process is used as the feed for an electrorefining process.
11. A process according to claim 1 wherein the potential applied during the electrorefining process is such that only uranium metal enters the salt.
12. A process according to claim 11 wherein the electrorefining process is carried out in the same electrolytic cell as an electrolytic reduction process.
US10/479,730 2001-06-06 2002-06-06 Actinide production Abandoned US20040244533A1 (en)

Applications Claiming Priority (3)

Application Number Priority Date Filing Date Title
GB0113749A GB0113749D0 (en) 2001-06-06 2001-06-06 Actinide production
GB0113749.6 2001-06-06
PCT/GB2002/002402 WO2002099815A2 (en) 2001-06-06 2002-06-06 Actinide production

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US20040244533A1 true US20040244533A1 (en) 2004-12-09

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US (1) US20040244533A1 (en)
EP (1) EP1393324B1 (en)
JP (1) JP2004528584A (en)
AU (1) AU2002257945A1 (en)
GB (1) GB0113749D0 (en)
WO (1) WO2002099815A2 (en)

Cited By (4)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
CN103680653A (en) * 2012-09-13 2014-03-26 通用电气-日立核能美国有限责任公司 Methods of fabricating metallic fuel from surplus plutonium
US9845542B2 (en) * 2012-06-15 2017-12-19 Kabushiki Kaisha Toshiba Method of recovering nuclear fuel material
WO2019150099A3 (en) * 2018-02-03 2019-10-03 Ian Richard Scott Continuous reprocessing of spent nuclear fuel
CN114349059A (en) * 2021-12-22 2022-04-15 南华大学 Preparation and application of novel uranium-fixed solid oxide fuel cell cathode material

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GB2395958A (en) * 2002-12-05 2004-06-09 British Nuclear Fuels Plc Electrolytic separation of metals
GB0304884D0 (en) * 2003-03-04 2003-04-09 British Nuclear Fuels Plc Process for separating metals
FR2992330B1 (en) * 2012-06-26 2014-08-08 Commissariat Energie Atomique PROCESS FOR SEPARATING AT LEAST ONE FIRST E1 CHEMICAL ELEMENT OF AT LEAST ONE SECOND E2 CHEMICAL ELEMENT INVOLVING THE USE OF A MEDIUM COMPRISING A SPECIFIED MELT SALT
US10221499B2 (en) * 2015-06-25 2019-03-05 Ge-Hitachi Nuclear Energy Americas Llc Nuclear fuel structure and method of making a nuclear fuel structure using a detachable cathode material
US10622112B2 (en) 2016-03-16 2020-04-14 Ian Richard Scott Conversion of spent uranium oxide fuel into molten salt reactor fuel
CN116390829B (en) * 2020-11-06 2024-03-19 山崎马扎克公司 Additive manufacturing device and control method thereof, storage medium and composite processing device

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Cited By (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US9845542B2 (en) * 2012-06-15 2017-12-19 Kabushiki Kaisha Toshiba Method of recovering nuclear fuel material
US10323330B2 (en) 2012-06-15 2019-06-18 Kabushiki Kaisha Toshiba Method of recovering nuclear fuel material
CN103680653A (en) * 2012-09-13 2014-03-26 通用电气-日立核能美国有限责任公司 Methods of fabricating metallic fuel from surplus plutonium
US10280527B2 (en) * 2012-09-13 2019-05-07 Ge-Hitachi Nuclear Energy Americas Llc Methods of fabricating metallic fuel from surplus plutonium
WO2019150099A3 (en) * 2018-02-03 2019-10-03 Ian Richard Scott Continuous reprocessing of spent nuclear fuel
CN111655905A (en) * 2018-02-03 2020-09-11 伊恩·理查德·斯科特 Continuous reprocessing of spent nuclear fuel
US11211176B2 (en) 2018-02-03 2021-12-28 Ian Richard Scott Continuous reprocessing of spent nuclear fuel
CN114349059A (en) * 2021-12-22 2022-04-15 南华大学 Preparation and application of novel uranium-fixed solid oxide fuel cell cathode material

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AU2002257945A1 (en) 2002-12-16
WO2002099815A2 (en) 2002-12-12
EP1393324B1 (en) 2007-01-03
EP1393324A2 (en) 2004-03-03
WO2002099815A3 (en) 2003-04-03
JP2004528584A (en) 2004-09-16
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