Field of the Invention

[0001] This invention relates to the reprocessing of spent nuclear fuel. In particular, the invention provides a process for the reprocessing or treatment of spent nuclear fuels which combines the advantages of conventional aqueous and high temperature non-aqueous molten salt processes.

Background to the Invention

[0002] Commercial plants for the reprocessing of spent nuclear fuel have traditionally made use of the so-called PUREX process, wherein spent nuclear fuel is dissolved in nitric acid and the dissolved uranium and plutonium are subsequently extracted from the nitric acid solution into an organic phase, typically comprising an organophosphate ester, for example tributyl phosphate (TBP), dissolved in an inert hydrocarbon, such as odourless kerosene. Thereafter, the organic phase is subjected to solvent extraction techniques to partition the uranium from the plutonium.

[0003] This particular processing technology is being further developed to provide treatment and recycle options for spent nuclear fuel produced in advanced reactors, for example, those considered for Generation IV reactor systems. A number of new treatment options are under consideration in this regard; however these systems are required to fulfil demanding criteria and, in particular, they are required to be able to treat fuels with high plutonium and actinide 'loadings', and to provide a processing route which is proliferation resistant.

[0004] Consequently, there is a requirement for a process which produces an output comprising multiple co-separated species suitable for recycling, while fission products and neutron poisons are separated for disposal. It is also necessary that such a process should not be easily modified to enable the re-routing and separation of pure species, for example Pu.

[0005] One such advanced aqueous reprocessing and recycling route which has been developed is based on the PUREX and corresponding UREX processes and leads to a three stage process, the three stages being designated NUEX, TRUEX and TALSPEAK. The functions of these three stages are as follows:

NUEX - This stage is a variation on the PUREX process, in that NUEX performs a similar role to a conventional plant, with the exception that, in this case, only the uranium is separated in a pure form. The process is designed to be proliferation resistant, as no pure Pu stream is created at any stage.

TRUEX - This step facilitates separation of fission products (FPs) from rare earth elements (REs) and actinides (Am, Cm) in the NUEX raffinate.

TALSPEAK - This process (Trivalent Actinide-Lanthanide Separation by

Phosphorus reagent Extraction from Aqueous Komplexes) allows for the separation of Am and Cm from the lanthanides.

[0006] The operation of the NUEX, TRUEX, TALSPEAK process is illustrated in Figure 1 , wherein dissolved spent fuel is introduced to the NUEX process, which allows for uranium and mixed uranium/plutonium streams to be created and for technetium to be separated off to waste. Thereafter, the remaining actinides (americium and curium), together with fission products and rare earth metals are subjected to the TRUEX process, wherein the fission products are separated off as high level waste, following which the TALSPEAK process enables americium and curium to be removed from the rare earths, which are also then separated as high level waste.

[0007] In addition to this aqueous/solvent reprocessing route, various other options have been investigated for the reprocessing or treatment of spent nuclear fuel from Generation IV reactors. These options include separation technology processes based on high temperature non-aqueous melts, based on various types of molten salts, which can be broadly identified in the following categories:

• Fluoride molten salt systems;

• Calcium chloride molten salts;

• Lithium-Potassium chloride molten salts; and

• Low temperature molten salts incorporating organic anions, often termed ionic liquids.

[0008] In the established prior art, two processes have been particularly developed for the treatment of irradiated nuclear fuel from advanced reactors by 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 of up to about 200 °C.

[0009] The first process, developed at Dimitrovgrad SSC-RIAR, makes use of chemical oxidants (chlorine and oxygen gases) to react with powdered uranium dioxide fuel in order to form higher oxidation state compounds such as U0 ₂ CI ₂ which are soluble in the molten salt. In an electrochemical cell, the uranium compounds are reduced to U0 ₂ at the cathode, thereby forming a dendritic deposit.

[0010] The second process, developed by the Argonne National Laboratory (ANL), is fundamentally an electrorefining technology which uses direct current to anodically oxidise uranium to form uranium ions in the molten salt electrolyte. At the cathode, the uranium is reduced and electrodeposited as dendritic uranium metal.

[0011 ] Other alternative molten salts technologies based on fluorine based salts have been investigated, and are currently under development, but they are less advanced. In addition, technology based on elemental fluorine is also being examined in the context of actinide fission product separation, although this is not strictly a molten salt separation process. The use of ionic liquids in advanced recycling processes is even less fully developed, and work in this area is still ongoing.

[0012] Thus, at present, no single separation process has been developed which offers a complete solution to the problem of reprocessing and recycling spent nuclear fuel and, consequently, significant development is still required in order that any of the available option should be available for commercial industrialisation.

[0013] The present inventors, however, have now further investigated this issue, and have developed a process which combines the advantages of the aqueous/solvent process with the benefits of the molten salt approach in order to provide a new fuel cycle. Thus, the benefit of the operating experience, efficiency and historical developments of conventional aqueous reprocessing technology (and that of proposed advanced aqueous processes - NUEX, TRUEX and TALSPEAK) is allied with the positive attributes of other advanced processing technology using molten salts in order to produce an overall process which is tailored to meeting future reprocessing requirements - in particular Global Nuclear Energy Partnership/Advanced Fuel Cycle Initiative (GNEP/AFCI) or any future allied initiative.

Summary of the Invention

[0014] Thus, in accordance with the present invention there is provided a process for the reprocessing of spent nuclear fuel, said process comprising the steps of:

(a) electrorefining the metal to produce a liquid metal cathode comprising the liquid metal alloyed with actinide or fission product metal, or actinide/fission product metal dendrites encased in liquid metal; (b) treating the actinide or fission product metal or actinide/fission product metal dendrites encased in the liquid metal with nitrate in order to provide an actinide nitrate salt or oxide and to isolate and partition chloride ions;

(c) dissolution of the actinide nitrate salt or oxide to provide an aqueous process stream; and

(d) extracting the resulting aqueous process stream with an organic phase and isolating reprocessed fuel from said organic phase.

[0015] Said process may also be applied to the treatment of legacy wastes.

[0016] Optionally, the major part of the liquid metal forming the cathode may, if desired, be removed by means of a conventional distillation process prior to step (c) of the above process.

[0017] When said spent nuclear fuel comprises oxide fuel, said process comprises the additional step of electrochemically treating the spent nuclear fuel in a molten salt electrolyte to produce a metal from said oxide spent fuel. This additional step is carried out prior to step (a) of the above process. In preferred embodiments of the invention, said additional step of electrochemically treating the spent nuclear fuel in a molten salt electrolyte involves electrochemically reducing to metallic form a metal oxide present in the spent nuclear fuel by cathodically electrolysing the oxide in the presence of the molten salt electrolyte, the potential of the cathode being controlled so as to favour oxygen ionisation into solution over deposition of the metal from the cations present in the molten salt.

[0018] In the electrochemical reduction stage, the cathode may be in the form of a mesh basket. The anode may be any inert anode, but is preferably a carbon anode.

[0019] The molten salt electrolyte can be any suitable molten salt, for instance LiCI, LiCI/KCI, CaCI ₂ or BaCI ₂ .

[0020] Optionally, the spent fuel may be first treated mechanically to remove zircalloy cladding before it is added to the electrolytic cell. Alternatively, the zircalloy cladding is first chopped into segments, and these segments are treated by the process of the invention.

[0021 ] After electrochemical reduction the irradiated fuel is in the form of a metallic solid at the cathode, and this metallic solid, containing fission products is then used directly as an anode in the electrorefining process which is carried out in either the same or a separate cell. The cathode in the electrorefiner is a molten metal, into which the actinide and some fission products are encapsulated. Optionally, uranium may be separated on its own by collection at a solid iron cathode. [0022] Treatment of the electrorefined liquid metal in order to remove chloride ions is crucial to the success of the invention, and is carried out as a transitional step between the molten salt stage and the final aqueous/solvent processing. It is most satisfactorily achieved by separation and capture of the electrorefined metal as an alloy or metal dendrite in a liquid cathode, and subsequent extraction into a nitrate phase. Said liquid metal cathode may comprise any suitable metal which is in liquid form at the operating temperature of the process, for example zinc or lead; most preferably, however, said liquid metal cathode comprises cadmium or bismuth. Said nitrate comprises either a molten nitrate non-aqueous melt, an aqueous solution of a nitrate salt, nitric acid, or other nitrate medium or liquor compatible with aqueous reprocessing. The resulting solution is used as a feed stream for the aqueous plant.

[0023] Said nitrate salt may comprise sodium or potassium nitrate, and is in the form of a solution which is compatible with aqueous processing. Preferably, said nitrate salt comprises a non-aqueous melt. Extraction into said non-aqueous nitrate salt melt is typically achieved by means of either electrochemical anodic oxidation or chemical oxidation using, for example, cadmium nitrate or bismuth nitrate, wherein the more electropositive actinides and fission products will react with the cadmium nitrate or bismuth nitrate to form actinide or fission product nitrates, whilst the cadmium and bismuth will reduce to the respective metals. Said nitrate salt may oxidise the actinide or fission product metals to the corresponding oxides, which may subsequently be dissolved using nitric acid.

[0024] Dissolution of the actinide and fission product nitrate salts or oxides thereby produced is most effectively achieved by treatment with solutions of nitric acid.

[0025] Subsequent extraction of the aqueous process stream is then performed using a suitable organic phase. Most conveniently, said organic phase comprises an organophosphate ester dissolved in an inert hydrocarbon. Preferably, said organophosphate ester comprises tributyl phosphate and said inert hydrocarbon comprises odourless kerosene.

[0026] Accordingly, the present invention provides a flexible, simplified process for the reprocessing of a wide range of spent nuclear fuels, including those produced by Generation IV and other reactor systems. The process isolates short lived fission products and partitions them to the molten salt phase. Therefore, this process offers the opportunity to reduce cooling times of spent fuel and, therefore, the inventory of spent fuel prior to reprocessing, which will be of particular importance in the case of advanced reactor fuel cycles. Brief Description of the Drawings

[0027] Embodiments of the invention are further described hereinafter with reference to the accompanying drawings, in which:

Figure 1 is a schematic representation of a prior art NUEX-TRUEX-TALSPEAK aqueous reprocessing method;

Figure 2 is a schematic representation of an advanced reprocessing and recycling method according to the present invention; and

Figure 3 is a schematic representation of the transitional process, carried out between the molten salt and aqueous/solvent stages of the processing, involving treating the electrorefined metal so as to remove chloride ions and provide an aqueous process stream.

Description of the Invention

[0028] The process according to the present invention provides a fully flexible operation, capable of accepting what are considered the most challenging of spent nuclear fuels (for example an inert matrix fuel). This flexibility means that the process may be used to treat or process materials other than Generation IV spent fuel and in fact, could be used to process or condition legacy waste and other legacy materials. It is also possible that short cooled fuel could be reprocessed, reducing the necessity for storage and cool down, in the event that the overall spent nuclear fuel inventory needed to be reduced.

[0029] The process of the invention fundamentally alters the perception and use of molten salts processing technology. Thus, in the claimed process, the molten salt process provides primary and "crude" separation of short lived radionuclides such as Cs, Sr, Ba and the like from Spent Nuclear Fuel (SNF) from any conventional or advanced reactor system. The process can also treat oxide or metal fuels, and could be readily adapted for other advanced fuel types, including nitride or carbide fuels. Removal of these Short Lived Fission Products (SLFPs) allows short cooled SNF from light water reactors (LWR), and other reactor systems, to be processed in well developed aqueous processing plants without the attendant destruction of solvent due to high radiation flux.

[0030] In addition, the primary separation achieved at the molten salt stage removes some lanthanides and, most particularly, noble metals; the latter remain with the anode in the electrorefining unit, and this minimises any demand on aqueous plant centrifuge technology, which is traditionally required to remove these insoluble species. Finally, the use of the disclosed suite of processing options means that only one fully flexible processing line would be required in the foreseeable future. [0031] Referring to the accompanying drawings, the prior art NUEX-TRUEX-TALSPEAK process of Figure 1 has already been discussed. Turning to Figure 2, there is shown a schematic representation of a process according to the invention wherein spent fuel from, for example, a LWR or Advanced Recycle Reactor/ Advanced Reactor (ARR/AR) is subjected to electroreduction and electrorefining in a molten salt. During the molten salt stage of processing, all the short lived FPs, for example Cs, Sr, and Ba will be removed, together with the volatile fission products noble metals and technetium and zirconium. The molten salt process is based on technology developed by ANL and Idaho National Laboratory (INL) but this has been developed by the incorporation of the additional technology of oxide reduction into the overall process, in order that oxide fuels, and potentially carbide and nitride fuels, may also be treated.

[0032] Thus, typically, 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 such as LiCI, CaCI ₂ or BaCI ₂ . A voltage is applied between the cathode and the anode. At the cathode the reaction involves the diffusion of oxygen atoms from the oxide lattice to the oxide/molten salt interface, followed by ionisation according to the reaction:

O + 2e ^" → O ^2-

The oxide ions which are produced dissolve in the electrolyte and are transferred to the anode where they are re-oxidised to produce carbon monoxide, carbon dioxide oxygen gas or a combination of said gases. 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 from the fused salt. Electrolysis at elevated temperatures results in an increased rate of oxygen diffusion, thereby also encouraging ionisation rather than metal deposition. Should a metal be deposited from the electrolyte instead of oxide ions, this is not necessarily deleterious to the process, and this can be accommodated.

[0033] Following electrolysis, the irradiated fuel is left in the form of a metallic solid at the cathode. This metallic solid, which contains fission products, is then used directly as the feed for an electrorefining process which can be carried out in the same or a different cell.

[0034] It should be emphasised that whilst the preferred embodiment of the invention envisages electrochemically reducing to metallic form a metal oxide present in the spent nuclear fuel by cathodically electrolysing the oxide in the presence of the molten salt electrolyte, and subsequently electrorefining the metal, the invention also envisages the use of other molten salt processes, for example, oxide electrowinning, non-aqueous liquid- liquid extraction, and the like, as 'primary separation' processes.

[0035] Returning to Figure 2, following the molten salt stage of processing, there follows the transition step, prior to the aqueous/solvent actinide recovery stages. At this stage, by using a nitrate based aqueous or non-aqueous salt system, soluble species will be produced which will be chemically compatible with the aqueous process. This stage of the process will be further discussed with reference to Figure 3.

[0036] It is believed that the transition step is essential to the overall success of the claimed process. This unit operation is vital to ensure that no chloride is transferred from the molten salt plant to the aqueous processing plant, since such an occurrence could result in corrosion of stainless steel plant components.

[0037] Optional procedures for carrying out this stage of the process could, for example, involve washing the electrorefiner cathode product thoroughly with de-ionised water. In this case, however, there is the potential for chloride to cross the molten salt/ aqueous plant boundary, which would clearly be totally unsatisfactory. Liquid metal distillation or metal ingot production, followed by dissolution in a nitrate solvent, will however provide a well partitioned product.

[0038] Thus, the inventors have developed the process which is depicted schematically in Figure 3, wherein the products of the electrorefining step for onward processing in the aqueous stage of treatment, which include, for example, U, Pu and other minor actinides, plus small quantities of fission products and lanthanides, are initially separated and captured as an alloy or metal dendrite in a liquid cathode such as bismuth or cadmium. In this state, the metal alloy or dendrite is separated from the salt and, therefore, chloride free, provided the dendrite is held within the molten metal matrix.

[0039] Then, from this molten metal matrix, these metals may be extracted into a nonaqueous nitrate salt, for example sodium nitrate, by either electrochemical anodic oxidation or chemical oxidation using for example, cadmium nitrate or bismuth nitrate. The extracted material is then entirely suitable as a feed stream for the aqueous stage of processing, since the product is highly soluble and the carrier salt has negligible interaction with the solvent extraction process, and can be readily recovered. An aqueous nitric acid treatment of the cooled, solidified metal moiety could also be used, if desired.

[0040] If a non-aqueous solvent is used for recovering the actinides and minor fission product metals from the cathode, then the final stage of processing involves dissolution of the non-aqueous nitrate salt in nitric acid and subsequent extraction of the dissolved uranium, plutonium and actinides from the aqueous solution into an organic phase, typically comprising an organophosphate ester, for example tributyl phosphate (TBP), dissolved in an inert hydrocarbon, such as odourless kerosene. Thereafter, the organic phase is typically subjected to solvent extraction techniques to partition the uranium from the plutonium.

[0041 ] Thus, the process of the invention provides a compact and flexible process for the treatment of a "broad" range of nuclear fuels. The use of a molten salt pre-treatment step opens up the possibility of a fully flexible reprocessing or conditioning process for both advanced reactor systems and orphan fuels. The feed from the molten salt pre-treatment step will comprise U, Pu, minor actinides (Am and Cm) and some REs in a highly soluble form. Thus, all the short lived FPs, for example Cs, Sr, and Ba will be removed, together with the volatile fission products, noble metals and zirconium (in the case of an inert matrix fuel) in this pre-treatment step.

[0042] As a consequence of the removal of these materials prior to the aqueous/solvent process, this stage of the procedure is considerably simplified, especially in the case of advanced reactor recycling processes. In such cases, the following advantages would typically accrue:

• Removal of the TRUEX process;

• Reduction in capacity and size of both the NUEX and TALSPEAK plants;

• Significant modification of the NUEX process, with some process steps removed:

o Feedstream treatment ahead of the chemical separation may be removed and replaced by a simple dissolver;

o UP cycle may possibly be reduced in size due to lower uranium content;

o UP2 Tc rejection contactor not required due to prior removal of Tc;

o TP cycle may be reduced to a simple evaporation plant due to no FPs being present from the Np/Pu (and U) stream.

[0043] As previously discussed, the transition step comprises the conversion of any metal chloride to a highly soluble chloride-free feed stream (metal nitrate) for the subsequent aqueous/solvent processing. In conventional aqueous systems, gadolinium is added to the dissolver solution before addition of sheared fuel. Gadolinium acts as a neutron poison and the amount added is calculated on the basis that the fuel is unirradiated, has the design basis maximum enrichment, and is entirely undissolved and present in a shape that is ideal for criticality. In the case of the process of the present invention, it is likely that the better dissolution which is achieved will obviate the requirement for the addition of Gd. [0044] Throughout the description and claims of this specification, the words "comprise" and "contain" and variations of them mean "including but not limited to", and they are not intended to (and do not) exclude other moieties, additives, components, integers or steps. Throughout the description and claims of this specification, the singular encompasses the plural unless the context otherwise requires. In particular, where the indefinite article is used, the specification is to be understood as contemplating plurality as well as singularity, unless the context requires otherwise.

[0045] Features, integers, characteristics, compounds, chemical moieties or groups described in conjunction with a particular aspect, embodiment or example of the invention are to be understood to be applicable to any other aspect, embodiment or example described herein unless incompatible therewith. All of the features disclosed in this specification (including any accompanying claims, abstract and drawings), and/or all of the steps of any method or process so disclosed, may be combined in any combination, except combinations where at least some of such features and/or steps are mutually exclusive. The invention is not restricted to the details of any foregoing embodiments. The invention extends to any novel one, or any novel combination, of the features disclosed in this specification (including any accompanying claims, abstract and drawings), or to any novel one, or any novel combination, of the steps of any method or process so disclosed.

[0046] The reader's attention is directed to all papers and documents which are filed concurrently with or previous to this specification in connection with this application and which are open to public inspection with this specification, and the contents of all such papers and documents are incorporated herein by reference.