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WO1999041752A1 - Retraitement de combustible nucleaire - Google Patents

Retraitement de combustible nucleaire Download PDF

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Publication number
WO1999041752A1
WO1999041752A1 PCT/GB1999/000246 GB9900246W WO9941752A1 WO 1999041752 A1 WO1999041752 A1 WO 1999041752A1 GB 9900246 W GB9900246 W GB 9900246W WO 9941752 A1 WO9941752 A1 WO 9941752A1
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WO
WIPO (PCT)
Prior art keywords
fuel
ionic liquid
uranium
cladding
plutonium
Prior art date
Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
Ceased
Application number
PCT/GB1999/000246
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English (en)
Inventor
Robert Charles Thied
Kenneth Richard Seddon
William Robert Pitner
David William Rooney
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Sellafield Ltd
Original Assignee
British Nuclear Fuels PLC
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by British Nuclear Fuels PLC filed Critical British Nuclear Fuels PLC
Priority to JP2000531850A priority Critical patent/JP2002503820A/ja
Priority to KR1020007008690A priority patent/KR20010040796A/ko
Priority to AU25267/99A priority patent/AU2526799A/en
Priority to EP99904948A priority patent/EP1055240A1/fr
Publication of WO1999041752A1 publication Critical patent/WO1999041752A1/fr
Anticipated expiration legal-status Critical
Ceased legal-status Critical Current

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    • C—CHEMISTRY; METALLURGY
    • C07—ORGANIC CHEMISTRY
    • C07F—ACYCLIC, CARBOCYCLIC OR HETEROCYCLIC COMPOUNDS CONTAINING ELEMENTS OTHER THAN CARBON, HYDROGEN, HALOGEN, OXYGEN, NITROGEN, SULFUR, SELENIUM OR TELLURIUM
    • C07F5/00—Compounds containing elements of Groups 3 or 13 of the Periodic Table
    • C07F5/003—Compounds containing elements of Groups 3 or 13 of the Periodic Table without C-Metal linkages
    • 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
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
    • G21C3/42—Selection of substances for use as reactor fuel
    • G21C3/58—Solid reactor fuel Pellets made of fissile material
    • 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 the treating or reprocessing of nuclear fuel using ionic liquids and to a novel form of a uranium compound.
  • the Argonne National Laboratory electrometallurgical treatment process (ANL - EMT) and the Dimitrovgrad SSC - WAR process both use molten salts at high temperatures (773 and 1000K, respectively).
  • the ANL process is fundamentally electrorefining technology, using current flow to ensure the oxidation of a uranium anode to form uranium ions in the molten salt electrolyte. At the cathode the uranium is reduced and electrodeposited as uranium metal.
  • the SSC - RIAR process uses chemical oxidants (chlorine and oxygen gases) to react with powdered UO 2 fuel to form higher oxidation state compounds such as UO 2 Cl 2 which are soluble in the molten salt. At the cathode the uranium compounds are reduced to UO 2 which forms a dendrite deposit.
  • chemical oxidants chlorine and oxygen gases
  • molten salts have been proposed for use in the reprocessing of irradiated fuels from Light Water Reactors (LWRs). These molten salts are typically mixtures of salts which are liquid only at high temperatures and this causes inherent disadvantages in a reprocessing plant.
  • LWRs Light Water Reactors
  • Ionic liquids free of molecular solvents were first disclosed by Hurley and Wier in a series of US patents (24446331, 2446339, 2446350). These ionic liquids comprised aluminium(III) chloride and a variety of N-alkylpyridium halides, and afforded a conducting bath for aluminium electroplating.
  • an ionic liquid is a salt, a mixture of salts, or a mixture of components which produce a salt or salts, which melts below or just above room temperature (in terms of this invention, a salt consists entirely of cationic and anionic species).
  • ionic liquid relates to a salt, a mixture of salts, or a mixture of components which produce a salt or salts, and which melts at a temperature up to 100°C eg from -50 to 100°C.
  • the cation in these ionic liquids is usually organic.
  • Known ionic liquids include aluminium(III) chloride in combination with an imidazolium halide, a pyridinium halide or a phosphonium halide.
  • the halides include 1- ethyl-3-methylimidazolium chloride, N-butylpyridinium chloride and tetrabutylphosphonium chloride.
  • An example of a known ionic liquid system is a mixture of l-ethyl-3-methylimidazolium chloride and aluminium(III) chloride.
  • WO 95/21871, WO 95/21872 and WO 95/21806 relate to ionic liquids and their use to catalyse hydrocarbon conversion reactions (e.g. polymerisation or oligomerisation of olefins) and alkylation reactions.
  • the ionic liquids are preferably 1-(C,-C 4 alkyl)-3-(C 6 -C 30 alkyl) imidazolium chlorides and especially 1- methyl-3-Ci 0 alkyl-imidazolium chloride, or 1-hydrocarbylpyridinium halides, where the hydrocarbyl group is for example ethyl, butyl or other alkyl.
  • WO 98/06106 describes a method of dissolving in an ionic liquid a metal in an initial oxidation state below its maximum oxidation state, wherein the ionic liquid reacts with the metal and oxidises it to a higher oxidation state.
  • the initial metal may be in the form of a compound thereof and may be irradiated nuclear fuel comprising UO 2 and/or PuO 2 as well as fission products.
  • the ionic liquid is nitrate-based, for example a pyridinium or substituted imidazolium nitrate, and may contain a Br ⁇ nsted or Franklin acid. Suitable acids are HNO j , H 2 SO 4 and [NO + ].
  • This International Patent application also describes certain novel ionic liquids, including 1-butylpyridinium nitrate, 1-octylpyridinium nitrate, other nitrate-based ionic liquids whose cation component is not exclusively alkylpyridinium or polymethylenebis (pyridinium), and substituted imidazolium nitrates, especially 1-butyl- 3-methylimidazolium nitrate, l-hexyl-3-methylimidazolium nitrate and l-octyl-3- methylimidazolium nitrate.
  • the oxidising ionic liquids described in WO 98/06106 may i) be intrinsically oxidising, ii) have an oxidising agent added, and/or iii) have a promoting agent of oxidising ability added.
  • the agents may be an oxidant dissolved in a non-oxidising liquid or an auxiliary agent to increase the oxidising reactivity of another oxidising species. If the solvent contains nitrate ions, the agent increases the oxidising reactivity of the medium beyond that which would be provided by the nitrate ions themselves; as described above, such agents include acids and particularly Br ⁇ nsted and Franklin acids.
  • a Br ⁇ nsted acid is a proton donor; conversely, a Br ⁇ nsted base is a proton acceptor.
  • a Franklin acid is a species which will give to a solvent a cation which is characteristic of the solvent system, e.g. [NO] + in the solvent N 2 O 4 ; protons are not Franklin acids.
  • a Lewis acid is any species which is an electron-pair acceptor. There are a wide range of species which can be described as Lewis acids, including BC1 3 FT or transition metal ions.
  • a Lewis base is an electron-pair donor.
  • a superacid is an acidic medium with a Hammet acidity function - H 0 which has a value above 6.
  • the solvent may in principle comprise any ionic liquid but the liquid normally comprises nitrate anions.
  • the cation will in practice comprise one or more organic cations, especially nitrogen heterocycles containing quaternary nitrogen and more especially N-substituted pyridinium or N,N'-disubstituted imidazolium.
  • the substituents are preferably hydrocarbyl and more preferably alkyl, which may be branched, for example.
  • the hydrocarbyl (e.g. alkyl) groups usually contain from 1 to 18 carbon atoms and some usually from 1 to 8 carbon atoms.
  • the cation may therefore be a disubstituted imidazolium ion where the substituent groups take the form C n H 2n+ , for 1 ⁇ n ⁇ 8, and the substituent groups are linear or branched groups.
  • the substituent groups take the form C n H 2n+ , for 1 ⁇ n ⁇ 8, and the substituent groups are linear or branched groups.
  • the cation might be a substituted tetra-alkylammonium ion, where the alkyl groups take the form of C n H 2n+ , for 1 > n ⁇ 6, and are linear or branched groups.
  • a preferred example is tetrabutylammonium.
  • the alkyl groups might be of different lengths.
  • the cation might be a substituted pyridinium ion, where the substituent group also takes the form C n H 2n+1 for 1 ⁇ n ⁇ 8, and the substituent groups are linear or branched groups; suitable substituents include butyl, 2-(2-methyl)propyl, 2- butyl and octyl but straight chain alkyl, especially butyl, is preferred. Of course, minor quantities of contaminants may be present, e.g. protonated - methylimidazole in l-butyl-3-methylimidazolium.
  • the nitrate-based ionic liquids of WO 98/06106 may be prepared by mixing aqueous silver(I) nitrate together with an appropriate organic halide.
  • one such ionic liquid is prepared by mixing together solutions of aqueous silver(I) nitrate and 1- butyl-3-methylimidazolium chloride ([bmim]Cl). Silver chloride is precipitated and the liquid l-butyl-3-methylimidazolium nitrate is formed:
  • the product may be purified by filtration and removing excess water from the filtrate.
  • l-hexyl-3-methylimidazolium nitrate is prepared by a similar method and this material is also a liquid at room temperature.
  • Alternative cations to pyridinium and imidazolium include quaternary phosphonium cations, e.g. tetrahydrocarbylphosphonium. Suitable hydrocarbyl groups are as described above in relation to pyridinium and imidazolium cations. Examples include unsymmetrically substituted phosphonium cations.
  • the agent added to the ionic liquid to enable the oxidising process to occur more efficiently is typically an acid", notably a Br ⁇ nsted acid (e.g. HNO 3 or H 2 SO 4 ) or a Franklin acid, for example [NO + ], serving in either case to make the medium more oxidisingly reactive towards substrates such as, for example, UO 2 or PuO 2 .
  • a Br ⁇ nsted acid e.g. HNO 3 or H 2 SO 4
  • a Franklin acid for example [NO + ]
  • substrates such as, for example, UO 2 or PuO 2
  • one class of oxidising ionic liquids contains an oxidant comprising nitrate and a promoter thereof.
  • the agent when combined with the ionic liquid may react with the ionic liquid to create a new species which is also an ionic liquid.
  • [NO][BF 4 ] is believed to react with the nitrate salts of organic cations to form the tetrafluoroborate(III) salt of the cation.
  • An exemplary reaction is: [Bu-py][NO 3 ] + [NO][BF 4 ] ⁇ N 2 O 4 + [Bu-py][BF 4 ]
  • Bu-py is 1-butylpyridinium
  • [Bu-py][BF 4 ] is an ionic liquid
  • WO 98/06106 Whilst the methods and ionic liquids described in WO 98/06106 are suitable for general use and, in particular, use in nuclear fuel reprocessing, we have now found an advantageous method which is an improvement over the prior art. In particular, WO 98/06106 does not describe clear separation and/or product recovery steps.
  • a method for treating or reprocessing spent nuclear fuel to substantially separate fissile material from other components of irradiated fuel which comprises dissolving the spent fuel or constituent parts of the spent fuel in an ionic liquid.
  • substantially separate is meant partially separate, but preferably completely separate.
  • the method of the invention may preferentially also include the step of treating the ionic liquor, resulting from the step of dissolving spent nuclear fuel in an ionic liquid, by solvent extraction to separate the fissile material from other components of irradiated fuel.
  • the ionic liquid is preferably an oxidising ionic liquid as described in WO 98/06106, which is incorporated herein by reference, particularly one containing an acid, e.g. a Br ⁇ nsted, Lewis, Franklin or superacid; although Br ⁇ nsted and Franklin acids are preferred.
  • an acid e.g. a Br ⁇ nsted, Lewis, Franklin or superacid; although Br ⁇ nsted and Franklin acids are preferred.
  • the process involves oxidation of U(0) or U(IV) (normally as UO 2 ) to U(VI) and usually of Pu(IV) (normally as PuO 2 ) to Pu(VI); for example, uranium dioxide is oxidised to tr ⁇ r ⁇ '-dioxouranium(VI) in complexed form and plutonium dioxide to trora-dioxoplutonium(VI) in complexed form.
  • the dissolution of plutonium and uranium for recycling will normally involve the unavoidable dissolution of other components of the irradiated fuel.
  • the reprocessing methods form a fissile material optionally in the form of an intermediate or final nuclear fuel product, e.g. a gel, a powder, a master batch material, a fuel pellet, a fuel pin or a fuel assembly.
  • the dissolution may involve dissolution of the cladding of a fuel rod, e.g. a stainless steel or magnesium alloy cladding or a zirconium alloy cladding such as that sold under the trade mark Zircaloy; after irradiation of fuel, Zircaloy cladding has a passivating oxide coating and dissolution of Zircaloy cladding therefore involves non- oxidative dissolution of oxidised Zircaloy (zirconium alloy).
  • the dissolution of the cladding of a fuel rod e.g. a stainless steel or magnesium alloy cladding or a zirconium alloy cladding such as that sold under the trade mark Zircaloy
  • Zircaloy cladding has a passivating
  • Zircaloy may be oxidative or non-oxidative.
  • chemical removal of the cladding as for example by dissolution in an ionic liquid, it may be removed prior to the dissolution of the fuel in the ionic liquid, whether by mechanical or chemical means.
  • one class of methods comprises contacting zirconium alloy clad irradiated fuel with an ionic liquid to dissolve the cladding and the fuel.
  • the ionic liquid preferably contains sulfate and may contain sulfuric acid which has been found to react with the zirconium oxide layer on Zircaloy cladding.
  • These methods may include immersing individual fuel pins or fuel assemblies in the ionic liquid within a suitable vessel.
  • a second class of methods includes the step of rupturing the cladding mechanically to expose the fuel pellets to the ionic liquid.
  • the fuel rod is placed initially in a first ionic liquid for dissolving the cladding and subsequently in a second ionic liquid for dissolving the uranium and plutonium or, optionally, compounds thereof.
  • a nitrate ionic liquid is preferred for dissolving fuel rods such as uranium or plutonium.
  • the dissolving of cladding may be carried out as a separate step to the dissolving of fuel, in which case different ionic liquids may be used. If the process of dissolving cladding and fuel is carried out as a single step then a mixed ionic liquid may optionally be used, eg a nitrate and sulfate ionic liquid mixture.
  • a solvent extraction step may be included to remove uranium and/or plutonium.
  • the solution resulting from the dissolution of the fuel is treated to extract other components of irradiated fuel, after which the uranium and plutonium are separated from the ionic liquid.
  • the uranium and plutonium may be removed together or they may be separated, in which latter case the ionic liquid liquor is subjected to a uranium/plutonium split operation.
  • the removal of uranium and plutonium may comprise selective dissolution step. Alternatively a fractional crystallisation may be used to bring about the removal of bulk uranium from the other components of irradiated fuel.
  • the selective dissolution may be conducted before or after extraction of other components of irradiated fuel, however it is preferable that a dissolution step is carried out before an extraction step.
  • the other components of irradiated fuel are suitably removed using a solvent extraction technique, for example one involving the contacting of the ionic liquid solution with a hydrophilic phase, especially an aqueous medium or another ionic liquid into which the other components of irradiated fuel and any Zr are extracted.
  • the ionic liquid solution is contacted with a hydrophobic phase, especially an organic solvent, for example a straight chain hydrocarbon (normally a straight chain alkane) or a mixture of such hydrocarbons, into which the uranium and plutonium are extracted.
  • the solvent extraction technique may involve procedures, such as, for example, oxidation/reduction and/or complexing, to change the solubility properties of one or more selected species in order to control the partition of such species between the two phases.
  • some methods include the complexing of one or more dissolved species to change their relative solubilities in the two solvents.
  • TBP tributyl phosphate
  • oxidation or reduction it may be mentioned that any uranium/plutonium separation may involve the selective reduction of one of the species into an oxidation state where the reduced or non-reduced species may be selectively extracted into another phase, for example an aqueous or organic phase or another ionic liquid.
  • a reagent for the selective reduction of Pu rather than U is stabilised U(IV) ions.
  • An example of this in Purex reprocessing is the use of U(IV) stabilised by hydrazine.
  • the invention includes methods in which there is sequential dissolution of different components of the irradiated fuel. In particular, some processes involve a first ionic liquid dissolving the uranium, a second ionic liquid dissolving the plutonium and a third the fission products, not necessarily in this order.
  • An additional method of bringing about the precipitation of dissolved metals is via a reduction in temperature of the ionic liquid, in which the uranium is dissolved.
  • a separation process for different metals would be via fractional crystallisation.
  • This particular method further distinguishes the use of ionic liquids from conventional molten salts, since it is impracticable to utilise the method with molten salts as such salts would solidify at lower temperatures.
  • the uranium, plutonium and fission products may be recovered from the two or three product streams.
  • the precipitation of dissolved metals can be induced by changing the acidity and basicity of the ionic liquid by adding or subtracting further ionic components from it.
  • precipitation can be brought about by the addition of non-ionic components.
  • the precipitation can be brought about by the addition of organic solvents which are miscible or slightly miscible with the ionic liquid.
  • ethyl acetate can be added to bring about the precipitation of a uranium compound. Ethyl acetate is slightly miscible with the ionic liquid, and at the point when the system just becomes biphasic through addition of ethyl acetate, precipitation is optimised.
  • Volatile non-ionic components added to the ionic liquid can be recovered from the ionic liquid by distillation.
  • the ionic liquid is not volatile. This allows both the organic solvent and the ionic liquid to be recycled to the process.
  • An alternative means of precipitation is the addition to the ionic liquid of an oxidising or reducing agent.
  • the ionic liquid liquor containing dissolved fuel and optionally dissolved cladding is subjected to electrochemical treatment to recover the dissolved uranium and plutonium.
  • the liquor is subjected to electrolysis to deposit the uranium on the cathode as uranium oxide, a uranium compound or uranium metal; the dissolved plutonium may be recovered by a similar route.
  • dissolved plutonium is co-deposited on the cathode with the uranium, irrespective of whether the metals are deposited in the metallic state (in the (0) oxidation state), as complexes or as oxides. Such co-deposition is useful in the manufacture of mixed oxide fuels.
  • Electrode reduction potentials are unique to the element, to the valency of the ion to the solvent and to the presence of other ions or molecules. If a potential is applied across a solution then all metal ions with a more positive potential will be deposited on the cathode. Metal ions with a more negative potential will remain in solution.
  • the electrode can be removed and replaced with a new one, run at a slightly more negative potential, for the deposition of the next metal with a more negative reduction potential. If it is the desire to deposit two metals together, then a potential more negative than the reduction potential for both ions is applied to reduce them together.
  • the cathode material may be selected from known cathode materials, for example.
  • Exemplary materials are carbon, especially glassy carbon, and tungsten.
  • the ionic liquid is optionally subjected to one or more intermediate steps between dissolution of the fuel and electrodeposition of dissolved species; for example, the ionic liquid may be treated to reduce the uranium, either by the addition of a reducing agent or by an additional electroreduction step.
  • Fractional crystallisation of a uranium compound may be a crude purification of the uranium product.
  • the remaining liquor containing fission products, actinides, lanthanides and plutonium can be subjected to further precipitation, or to electrochemical extraction to separate further uranium and/or plutonium.
  • An electrochemical step can be employed on the bulk of the solution to bring about the separation of uranium and plutonium, from the other components of irradiated fuel.
  • the fuel eg uranium or plutonium
  • the fuel may be purified, for example, by recrystallisation from an ionic liquid.
  • the electrochemical step may require that any excess acid is removed from the system or neutralised. If this is not the case, then any electrodeposited product may be re-oxidised and dissolved by the remaining acid present.
  • Methods of removing the acid include one or more of the following:
  • the process illustrated in the flowsheet includes addition of a reductant to the liquor resulting from dissolution and passage of the resultant liquor to an electrochemical apparatus, e.g. an electrochemical bath, where the uranium species is reduced by the application of a current across the electrodes and is deposited at a cathode as uranium metal or another uranium compound.
  • the deposited uranium will be removed from the electrode and passed to a further step for removal of entrapped ionic liquid.
  • a similar process is used to recover dissolved plutonium.
  • the fuel material produced by the method of the invention is novel per se, in that it has a unique dimeric structure, in which two atoms of fuel material form a dimeric species comprising a dicarboxylate bridge.
  • various dicarboxylates may be used, for example, oxalate, malonate, succinate, glutarate, adipate or pimelate; a preferred bridge moiety is oxalate.
  • a fuel material comprising a dimeric species containing two fuel atoms bridged by a dicarboxylate moiety.
  • the preferred fuel is uranium.
  • the preferred fuel material has a structure;
  • Ac is Pu or U, preferably U.
  • the fuel material of the invention also provides a unique X-ray diffraction pattern.
  • a fuel material a molecular structure as shown in Fig. 2 which is derived by X-ray diffraction.
  • 1 -Methylimidazole was distilled under vacuum and stored under dinitrogen prior to use.
  • l-Butyl-3-methylimidazolium or 1-alkylpyridinium salts were prepared by direct reaction of the appropriate alkyl halide with 1-methylimidazole or pyridine, respectively, and recrystallised from ethanenitrile and ethyl ethanoate.
  • Nitrate ionic liquids were all prepared by methods analogous to the following method used to prepare l-butyl-3-methylimidazolium nitrate.
  • l-butyl-3-methylimidazolium chloride (8.04 g, 46.0 m mol) was dissolved in water (15 cm 3 ). To this solution a solution of silver(I) nitrate (7.82 g, 46.0 m mol) in water (20 cm 3 ) was added. A white precipitate (silver(I) chloride) formed immediately. The mixture was stirred (20 min) to ensure complete reaction, and was then filtered twice through a P3 sintered glass funnel to remove the white precipitate (the second filtration was generally necessary to remove the final traces of precipitate). The water was removed on a rotary evaporator, yielding a yellow or brown viscous liquid, sometimes containing small black solid particles.
  • the acetonitrile was removed under vacuum, and the pale yellow ionic liquid product then dried by heating in vacuo (ca. 50 °C, 2-3 d). Some discoloration of the product occurred if the heating was too vigorous. Residual silver(I) chloride as removed by electrolysis; silver metal is electroplated at the cathode; and chlorine gas is evolved at the anode. The resulting ionic liquid was stored under dinitrogen to exclude moisture.
  • a uranyl salt has been identified analysis of which has identified the molecular as C18H30N8O20U2, and the elemental weight percents are: 18.73% C; 2.62% H; 9.71% N; 27.72 % O; and 41.23 % U.
  • Electrochemical reduction of the uranyl salt may produce UO 2 which is insoluble in the ionic liquid and should precipitate from solution. Every mole of uranyl salt is believed to produce two moles of UO 2 . Therefore, complete reduction of 0.23 m mol of the uranyl salt to UO 2 should yield 0.46 m mol UO 2 . Furthermore, since each mole of UO 2 produced requires two mole equivalents of electrons, the total charge required during the production of 0.46 m mol UO 2 should be 88.8 C.
  • An electrolysis cell was set up in a three-electrode cell with the uranyl solution as the bulk solution.
  • the reference electrode was a silver wire immersed in a 0.1 mol L "1 solution of AgNO 3 in [bmim][NO 3 ] separated from the bulk solution in a glass tube with a porous vycor tip.
  • the counter electrode was a platinum coil immersed in pure [bmim][NO 3 ] separated from the bulk solution in a glass tube with a glass frit.
  • the working electrode was a flag (ca 3 cm x 2 cm x 0.1 cm) formed from a sheet of glassy carbon.
  • Electrolysis was carried out on the stirred uranyl solution under a dry nitrogen atmosphere by holding the cathode potential at -1.5 V versus Ag(I)/Ag.
  • the cathode became passivated and electrolysis was halted, perhaps by the adsorption of the UO 2 product.
  • Two techniques were used simultaneously to keep the cathode from becoming passivated. Dry nitrogen was bubbled across the electrode remove passivating material such as adsorbed UO 2 .
  • the working electrode potential was periodically stepped to +1.0 V (where it acts as an anode). Typically, the working electrode was held at 1.5 V for 9.6 s and pulsed to +1.0 V for 0.4 s. Used in conjunction, nitrogen sparging and potential pulsing kept the electrode from passivating.
  • Series 1 shows a voltammogram recorded in the neat solution.
  • the electrochemical window of the ionic liquids is demonstrated as stretching from ca. -2.2V, where the organic cation is reduced, to ca. +1.5V, where nitrate is oxidised.
  • Series 2 shows a voltammogram recorded in the 0.1 mol L "1 uranyl solution prior to electrolysis.
  • the wave with a peak around -0.9 V represents the reduction of uranyl.
  • a much smaller anodic wave can be seen around 0 V.
  • Lack of an anodic wave of equal size during the reverse scan indicates that the reduction is chemically irreversible.
  • the broadness of the cathodic wave indicates slow electron transfer.
  • the solution was diluted with acetone (to make the solution less viscous) and the precipitate was collected by vacuum filtration through size P4 glass frit.

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  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
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  • High Energy & Nuclear Physics (AREA)
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Abstract

L'invention a trait à un procédé de traitement ou de retraitement de combustible nucléaire épuisé, afin de séparer sensiblement la matière fissile des produits de fission, consistant à dissoudre le combustible épuisé ou des parties constitutives du combustible épuisé dans un liquide ionique, et en particulier afin de récupérer l'uranium et/ou le plutonium. L'invention a également trait à une nouvelle structure de cristal.
PCT/GB1999/000246 1998-02-11 1999-02-10 Retraitement de combustible nucleaire Ceased WO1999041752A1 (fr)

Priority Applications (4)

Application Number Priority Date Filing Date Title
JP2000531850A JP2002503820A (ja) 1998-02-11 1999-02-10 核燃料再処理
KR1020007008690A KR20010040796A (ko) 1998-02-11 1999-02-10 핵연료 재처리법
AU25267/99A AU2526799A (en) 1998-02-11 1999-02-10 Nuclear fuel reprocessing
EP99904948A EP1055240A1 (fr) 1998-02-11 1999-02-10 Retraitement de combustible nucleaire

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
GB9802852A GB9802852D0 (en) 1998-02-11 1998-02-11 Nuclear fuel reprocessing
GB9802852.5 1998-02-11

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WO1999041752A1 true WO1999041752A1 (fr) 1999-08-19

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EP (1) EP1055240A1 (fr)
JP (1) JP2002503820A (fr)
KR (1) KR20010040796A (fr)
CN (1) CN1290397A (fr)
AU (1) AU2526799A (fr)
GB (1) GB9802852D0 (fr)
WO (1) WO1999041752A1 (fr)

Cited By (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
WO2001013379A1 (fr) * 1999-08-18 2001-02-22 British Nuclear Fuels Plc Procede de separation de metaux
WO2001015175A3 (fr) * 1999-08-19 2002-02-21 British Nuclear Fuels Plc Procede de recyclage de liquides ioniques
WO2003004727A3 (fr) * 2001-07-06 2003-05-01 Univ Belfast Electrosynthese de composes organiques
WO2004078303A3 (fr) * 2003-03-04 2005-02-03 British Nuclear Fuels Plc Procede pour separer des metaux
RU2253916C1 (ru) * 2004-06-28 2005-06-10 Российский научный центр "Курчатовский институт" Способ переработки облученного ядерного топлива
WO2010105299A1 (fr) * 2009-03-17 2010-09-23 Commonwealth Scientific And Industrial Research Organisation Électro-récupération de métaux
GB2534432A (en) * 2014-06-25 2016-07-27 Hitachi Ltd Method for seperating actinides and apparatus for seperating actinides

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FR2901627B1 (fr) * 2006-05-24 2009-05-01 Commissariat Energie Atomique Procede de retraitement d'un combustible nucleaire use et de preparation d'un oxyde mixte d'uranium et de plutonium
JP2010184902A (ja) * 2009-02-13 2010-08-26 Kri Inc イオン液体の精製方法および回収方法
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WO2001013379A1 (fr) * 1999-08-18 2001-02-22 British Nuclear Fuels Plc Procede de separation de metaux
US6911135B1 (en) 1999-08-18 2005-06-28 British Nuclear Fuels Plc Process for separating metals
WO2001015175A3 (fr) * 1999-08-19 2002-02-21 British Nuclear Fuels Plc Procede de recyclage de liquides ioniques
WO2003004727A3 (fr) * 2001-07-06 2003-05-01 Univ Belfast Electrosynthese de composes organiques
US7972492B2 (en) 2001-07-06 2011-07-05 The Queen's University Of Belfast Electrosynthesis of organic compounds
WO2004078303A3 (fr) * 2003-03-04 2005-02-03 British Nuclear Fuels Plc Procede pour separer des metaux
RU2253916C1 (ru) * 2004-06-28 2005-06-10 Российский научный центр "Курчатовский институт" Способ переработки облученного ядерного топлива
WO2010105299A1 (fr) * 2009-03-17 2010-09-23 Commonwealth Scientific And Industrial Research Organisation Électro-récupération de métaux
AU2010225457B2 (en) * 2009-03-17 2015-08-20 Commonwealth Scientific And Industrial Research Organisation Electrorecovery of metals
US9580772B2 (en) 2009-03-17 2017-02-28 Commonwealth Scientific And Industrial Research Organisation Electrorecovery of metals
GB2534432A (en) * 2014-06-25 2016-07-27 Hitachi Ltd Method for seperating actinides and apparatus for seperating actinides
GB2534432B (en) * 2014-06-25 2017-06-07 Hitachi Ltd Method for seperating actinides and apparatus for seperating actinides

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EP1055240A1 (fr) 2000-11-29
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CN1290397A (zh) 2001-04-04
KR20010040796A (ko) 2001-05-15
AU2526799A (en) 1999-08-30

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