PYROMETALURGICAL PROCESS FOR THE TREATMENT AND RECYCLING OF SPENT FUEL SALT FROM A MOLTEN CHLORIDE NUCLEAR REACTOR

Abstract

The invention relates to a pyrometallurgical process for the treatment-recycling of a spent fuel salt from a molten chloride nuclear reactor.

Description

Title of the invention: PYROMETALURGICAL PROCESS FOR THE TREATMENT AND RECYCLING OF A SPENT FUEL SALT FROM A MOLTEN CHLORIDE NUCLEAR REACTOR technical field

[0001] The invention relates to the field of spent fuel treatment of molten salt nuclear reactors or MSRs, also designated by the acronym MSR for "Molten Salt Reactors".

[0002] More specifically, the invention relates to a pyrometallurgical process for the treatment and recycling of a spent fuel salt, also called irradiated fuel salt, from a molten chloride nuclear reactor. Prior art

[0003] Molten salt nuclear reactors are reactors designed to use a liquid fuel which is a molten salt, typically based on fluorides or chlorides.

[0004] First conceived in the 1960s, molten salt reactor technology is now the subject of much attention because these reactors potentially offer many advantages, including, in particular, high energy efficiency and flexibility of use.

[0005] Typically, a spent fuel salt from a molten chloride reactor comprises alkali and alkaline earth soluble actinides, lanthanides and fission products which are in the form of molten chlorides, insoluble fission products and, possibly, volatile fission products, in a solvent consisting of one or more alkali or alkaline earth metal chlorides, also in the molten state.

[0006] The treatment and recycling of such a salt must allow, at a minimum: - to selectively recover the actinides present in this salt so that they can be reused in the composition of a new combustible salt, and - to eliminate the lanthanides from the salt solvent so that this solvent can also be reused in the constitution of a new combustible salt.

[0007] Ideally, it would be desirable for the treatment and recycling of spent fuel salt to also allow for the recovery of: - on the one hand, the chlorine present in this salt in the form of chlorides, particularly when the spent fuel salt comes from the irradiation of a new fuel salt enriched in chlorine-37 and is therefore itself rich in chlorine-37, in order to be able to reuse this chlorine, and - on the other hand, insoluble fission products that are potentially valuable outside the nuclear sphere, such as platinum group metals and molybdenum.

[0008] In addition, it would be desirable for the treatment-recycling to be carried out in a closed or quasi-closed cycle so as to reduce as much as possible the number of reagents used and the volume of waste produced during this treatment-recycling.

[0009] The state of the art in the treatment of spent fuel salt from a molten chloride nuclear reactor is extremely limited and unsatisfactory.

[0010] Indeed, in patent application GB 2 536 857, hereinafter referred to as [1], a process for preparing a new fuel salt for molten chloride reactors by converting spent nuclear fuel from a conventional nuclear reactor, i.e. light water or gas-cooled and, therefore, based on uranium oxide, was proposed, on the one hand, and, on the other hand, a process called recycling of spent molten chloride fuel salt but which, in reality, simply consists of adding fissile material to the reactor and removing the noble fission products.

[0011] Furthermore, the implementation of this purported recycling process is subject to conditions, namely that the molten chloride reactor must be an epithermal or fast neutron reactor and that the fuel salt must contain uranium-238 or another neutron-absorbing isotope, preferably fertile but non-fissile, in addition to its fissile isotopes.

[0012] In patent application GB 2 554 068, hereinafter referred to as [2], a process for converting spent nuclear fuel, based on uranium oxide, into a new fuel salt for molten chloride reactors was also proposed, as well as a process for treating the spent fuel salt obtained from the irradiation of this new fuel salt.

[0013] The only example of this treatment, which is described in reference [2], is based on a simulation and concerns a spent salt resulting from the irradiation of a salt comprising 60% NaCl, 20% UCl3 and 20% PuCl3. In this example, it is planned to bring the spent salt into contact with a bismuth-calcium alloy to extract from the salt all of the actinides (i.e. U, Pu, Np and Am) and about 1 / 3 of the lanthanides, and then to condition the salt thus depleted in actinides for storage for a period of 300 years. The actinide-enriched bismuth-calcium alloy is then contacted with a salt containing 60% NaCl and 40% UCl3. All the actinides present in the Bi-Ca alloy, except for uranium, exchange with the UCl3 in the salt and thus end up in this salt, which constitutes the new fuel salt. Therefore, in this process, the uranium and solvent from the spent fuel salt are not recycled, and the lanthanides are only partially removed.

[0014] Finally, a process for treating spent combustible salt with molten chlorides was proposed in US patent application 2017 / 0301413, hereinafter referenced [3]. but which only allows the volatile and insoluble fission products of this salt to be eliminated. Description of the invention

[0015] The invention aims precisely to propose a pyrometallurgical process for the treatment and recycling of a spent fuel salt from a molten chloride nuclear reactor which makes it possible to meet the minimum requirements, mentioned above, which must be met by a process for the treatment and recycling of a spent fuel salt, but also, if desired, to ensure extensive recycling of the constituents of this salt, and this advantageously in a closed or quasi-closed cycle.

[0016] The invention therefore relates to a pyrometallurgical process for the treatment and recycling of spent fuel salt from a molten chloride nuclear reactor, the spent fuel salt being a molten salt comprising actinides, lanthanides, and soluble alkali and alkaline earth fission products in the form of chlorides, insoluble fission products, and, optionally, volatile fission products in a solvent consisting of one or more chlorides selected from alkali and alkaline earth metal chlorides, which process comprises at least the following steps: a) extraction of actinides from the spent fuel salt, the extraction of actinides comprising contacting the salt with a medium comprising a metal Mi in liquid form, immiscible with the salt, and reducing the oxidation state of the actinides by a reducing metal RedB alloyed with the metal M h or by an electric current,whereby a saline phase depleted in actinides and a metallic phase enriched in actinides in alloy form with the metal Mb are obtained, which are then separated from each other; b) extraction of lanthanides from the saline phase obtained at the end of step a), the extraction of lanthanides comprising bringing the saline phase into contact with a medium comprising a metal M2 in liquid form, immiscible with the saline phase, and reducing the degree of oxidation of the lanthanides by a reducing metal Red2, alloyed with the metal M2, or by an electric current, thereby obtaining a saline phase depleted in lanthanides and a metallic phase enriched in lanthanides in alloyed form with the metal M2, which are separated from each other; (c) deextraction of actinides from the metallic phase obtained at the end of step (a), the deextraction comprising contacting the metallic phase with both the saline phase obtained at the end of step (b) and a chlorine source, thereby obtaining a saline phase comprising actinide chlorides and a metallic phase depleted in actinides, which are then separated; and d) adjustment of the fissile material content of the saline phase obtained at the end of step c) by adding at least one actinide chloride to the saline phase, thereby obtaining a new fuel salt for a molten chloride nuclear reactor.

[0017] Thus, according to the invention: - Actinides are extracted selectively from lanthanides and soluble alkali and alkaline earth fission products in their allied form, and then lanthanides are extracted secondarily, also in their allied form, selectively from soluble alkali and alkaline earth fission products. These extractions are carried out either by reductive extraction (if the reduction of their oxidation state is ensured by a reducing agent) or by electrolysis (if the reduction of their oxidation state is ensured by an electric current). - the actinides are transferred, in the form of chlorides, from the metallic phase in which they are present after their extraction to the saline phase resulting from the extraction of the lanthanides, and - the actinide chloride content of this saline phase is adjusted, which leads to the production of a new fuel salt for a molten chloride nuclear reactor.

[0018] According to the invention, the metal Mi used in step a) can in particular be chosen from aluminium, gallium, bismuth, cadmium, tin, lead and zinc.

[0019] Among these, preference is given to aluminium, the use of which has, in fact, been shown to lead to particularly high separation factors between actinides and lanthanides.

[0020] Preferably, step a) is a reductive extraction and, therefore, an extraction in which the reduction of the oxidation state of the actinides is ensured by the reducing metal Redi alloyed with the metal Mb in which case this reducing metal is preferably an alkali or alkaline-earth metal and, better still, the one or one of those which enter(s) into the composition of the solvent of the spent fuel salt, in order to avoid introducing into the saline phase from the extraction of the actinides a metallic element foreign to this solvent.

[0021] Thus, for example: - if the solvent of the spent combustible salt is of the NaCl type, then the reducing metal Redi is preferentially sodium; - if the solvent of the spent combustible salt is of the NaCl-KCl type, then the reducing metal Redi is preferentially sodium or potassium; whereas - if the solvent of the spent combustible salt is of the NaCl-MgCl2 type, then the reducing metal Redi is preferentially sodium or magnesium.

[0022] By way of example, the reduction of plutonium, present in plutonium trichloride in the +3 oxidation state, by magnesium alloyed with aluminium leads to the formation of a Pu-Al alloy and magnesium chloride, this reaction being represented by the following equation:

[0023] [Math.l] PuChf.e;} +• ~+ Pu-A^-^e;f l.SMgChr^îp

[0024] Preferably also, step b) is also a reducing extraction.

[0025] To achieve this, it is possible to use a metal M2 different from the metal Mb, for example gallium, bismuth, cadmium, lead, tin or zinc if the metal Mi is aluminium, as well as a reducing metal Red2 different from the reducing metal Redb, for example an alkali or alkaline earth metal different from that used as the reducing metal Redb

[0026] However, to simplify the implementation of the process and also to enable it to operate in a closed or nearly closed cycle, thereby optimally reducing the number of reactants used and the volume of waste produced during its implementation, it is preferable that the metal M2 be identical to the metal Mi – and therefore, preferably aluminum – and that the reducing metal Red2 be identical to the reducing metal Redb

[0027] In any event, although lanthanides are thermodynamically more stable and, therefore, more difficult to reduce than actinides, the amount of reducing metal used for the extraction of actinides must be judiciously chosen to allow satisfactory extraction of actinides while limiting that of lanthanides or, in other words, to obtain an optimal separation factor between actinides and lanthanides.

[0028] As previously stated, the extraction of actinides and lanthanides, instead of being reductive extractions, can be carried out by electrolysis, for example by using a positive electrode, or anode, made of an inert material such as graphite, and, as a negative electrode, or cathode, the medium comprising metal Mi (for the extraction of actinides) and metal M2 (for the extraction of lanthanides), and by applying a direct electric current between the two electrodes. Here too, the metals Mi and M2 can be different from each other, but it is preferable that they be identical.

[0029] During these electrolysis processes, dichlorine is produced at the anode which can advantageously be recovered and then used as a source of chlorine for the deextraction of actinides, i.e. in step c).

[0030] Indeed, in step c), the contacting of the metallic phase obtained at the end of step a) with a chlorine source preferably includes bubbling or sloshing - the two words being considered here as synonyms - of a gaseous source of chlorine and, in particular, of gaseous chlorine, or dichlorine, or of anhydrous hydrogen chloride, in this metallic phase.

[0031] Preferably, dichlorine is used as the gaseous source of chlorine.

[0032] By way of example, the de-extraction of plutonium from a metallic phase, in which it is alloyed with aluminium, by bubbling dichlorine in this metallic phase leads to the obtaining of plutonium trichloride on the one hand, and aluminium trichloride on the other.

[0033] When steps a) and b) are reducing extractions, then step b) leads to obtaining a saline phase which, in addition to being depleted in actinides and lanthanides, is also enriched in reducing metal Redi and, possibly, in reducing metal Red2 if Red2 is different from Redi.

[0034] In this case, the process preferably includes, between steps b) and c), a step of reducing the content of reducing metal Redi and, where applicable, of reducing metal Red2 of the saline phase obtained at the end of step b).

[0035] This reduction can be obtained by electrolysis, which is carried out, for example, using a positive electrode made of an inert material such as graphite and a negative electrode made of a medium comprising a metal M3 in the liquid state, immiscible with the saline phase, in which case the electrolysis produces chlorine at the anode while the reducing metal Redi and, where applicable, the reducing metal Red2 combine with the metal M3. Here too, the chlorine thus produced is recovered and then used as a source of chlorine for the deextraction of the actinides, i.e. in step c).

[0036] Still with the aim of ensuring the treatment and recycling of spent fuel salt in a closed or nearly closed cycle, the metal M3 is preferably the metal Mi of the metallic phase obtained at the end of the actinide de-extraction step. This allows the metal Mb, which was used for the actinide extraction, to be reused in a step of the process other than this extraction. If, moreover, the same metal is used as the reducing metals Redi and Red2, then the reducing metal alloy Mm obtained at the end of the electrolysis can be used in further steps for the extraction of actinides and lanthanides.

[0037] According to the invention, the method preferably comprises, before step a): (i) a step aimed at removing volatile fission products, such as xenon, krypton and / or zirconium tetrachloride, from the spent fuel salt, which may be present in this salt if these fission products have not already been volatilized in the reactor, and / or ii) a step aimed at removing insoluble fission products from the spent fuel salt, so named because they are present in this salt in the form of insoluble metallic compounds. This type of fission product is mainly represented by platinum group metals (Pt, Pd, Ru, Rh, ...) and molybdenum.

[0038] As known per se, step i) may include bubbling an inert gas, such as helium or argon, into the spent combustible salt.

[0039] The volatile fission products thus eliminated from the spent fuel salt can then be sent to a unit dedicated to the separation of gases and aerosols.

[0040] As for step ii), it may include a digestion of this salt, that is to say, bringing this salt into contact with a medium comprising a metal M4 in a liquid state, immiscible with the salt and capable of selectively adsorbing the insoluble fission products, then a separation of the spent combustible salt from this medium, whereby at the end of this step we obtain, on the one hand, the spent combustible salt, now devoid of the insoluble fission products, and, on the other hand, a metallic phase comprising the metal M4 and the said insoluble fission products.

[0041] Advantageously, step ii) is complemented by a distillation step of the metallic phase from step ii) to separate the metal M4 from the insoluble fission products.

[0042] Therefore, the metal M4 is preferably a metal which, in addition to being able to selectively adsorb insoluble fission products, has a boiling point low enough to be easily distilled, which is in particular the case of zinc (TE: 907 °C) and cadmium (TE: 161 °C).

[0043] Once separated from the insoluble fission products, the M4 metal can be reused later in a new digestion stage while the insoluble fission products can be sent to a processing unit in charge of their recovery or, failing that, of their conditioning in a specific matrix.

[0044] If step ii) is not implemented, then the insoluble fission products are extracted from the salt with the actinides during their recovery.

[0045] Typically, the quantity of fissile material present in the saline phase obtained at the end of step c) is insufficient to produce a new fuel salt suitable for maintaining the criticality of the molten chloride nuclear reactor in which this new fuel salt will be used, hence the fact that step d) consists of an adjustment of the fissile material content of this saline phase by an external supply of one or more actinide chlorides.

[0046] The actinide chloride(s) thus supplied may, in particular, be obtained beforehand by: - Chlorination of the corresponding oxalate or oxalates with hydrogen chloride, - Hydrochlorination of the corresponding oxide or oxides with hydrogen chloride, - carbochlorination of the corresponding oxide or oxides with carbon and chlorine, and / or - chlorination of the corresponding actinide or actinides in the metallic state with dichlorine.

[0047] The chlorine produced during the various electrolysis steps that may be carried out in the implementation of the process, and in particular during the electrolysis that may be carried out to reduce the content of reducing metal Redi and, where applicable, Red2, can advantageously be used to produce said actinide chloride(s), whether: - to produce the hydrogen chloride necessary for the chlorination of the oxalate or the corresponding oxalates or for the hydrochlorination of the oxide or the corresponding oxides, the hydrogen chloride being able, in fact, to be produced by reaction of dichlorine either with dihydrogen or with methane and dioxygen, or - to carry out the carbochlorination of the oxide or the corresponding oxides or the chlorination of the actinide or the corresponding actinides in the metallic state.

[0048] According to the invention, the spent fuel salt is preferably obtained from the irradiation of a new fuel salt comprising at least one actinide chloride selected from uranium trichloride, plutonium trichloride, thorium tetrachloride and americium trichloride, in a solvent consisting of one or more chlorides selected from sodium chloride, potassium chloride and magnesium chloride.

[0049] A new combustible salt of this type is, for example, a salt of formula NaCl-UCl3, NaCl-PuCl3, NaCl-UCl3-PuCl3, NaCl-ThCl4-PuCl3, NaCl-ThCl4-UCl3-PuCl3, NaCl-MgCl2-PuCl3, NaCl-MgCl2-UCl3-PuCl3, NaCl-MgCl2-PuCl3-AmCl3, NaCl-KCl-UCl3 or NaCl-KCl-UCl3-PuCl3.

[0050] Other advantages and features of the invention will become apparent from the following non-limiting supplementary description, which refers to the accompanying figures. Brief description of the figures

[0051] [Fig-1] is a schematic representation of a preferred embodiment of the process of the invention.

[0052] [Fig.2] illustrates the separation factors between plutonium and cerium (FSPu / ce) on the one hand, and between plutonium and magnesium (FSPu / Mg) on ​​the other hand, as obtained by thermodynamic calculations for a reductive extraction of plutonium from a molten chloride phase comprising, in addition to plutonium, cerium, by means of a metallic phase comprising aluminium, gallium, bismuth or zinc as solvent and magnesium as reducing agent; in this figure, the left ordinate axis corresponds to the separation factor between plutonium and cerium while the right ordinate axis corresponds to the separation factor between plutonium and magnesium.

[0053] [Fig.3] illustrates the evolution of the plutonium extraction rate, denoted EPu and expressed in percentages, based on the number of magnesium equivalents used, noted nbeq Mg, as obtained during tests aimed at recovering plutonium by reductive extraction from a molten chloride phase using a metallic phase comprising zinc as a solvent and magnesium as a reducing agent.

[0054] [Fig.4] illustrates the evolution of the extraction rate of neodymium, cerium, and the lanthanum and praseodymium, noted ELns and expressed as percentages, as a function of the plutonium extraction rate, noted EPu and also expressed as percentages, as obtained during tests aimed at recovering plutonium by reductive extraction from a molten chloride phase comprising, in addition to plutonium, neodymium, cerium, lanthanum, praseodymium and samarium, by means of a metallic phase comprising zinc as solvent and magnesium as reducing agent; in this figure, samarium is not present because it has not been extracted.

[0055] [Fig. 5] illustrates the evolution of the extraction rate of plutonium, neodymium, and cerium, lanthanum, praseodymium and samarium, noted Ex and expressed as a function of the number of magnesium equivalents used, noted nbéq Mg, as obtained during the same tests as those whose results are illustrated on [Fig.4].

[0056] [Fig.6] illustrates the evolution of the extraction rate of neodymium, cerium and the praseodymium, noted ELns and expressed as a percentage, as a function of the plutonium extraction rate, noted EPu and also expressed as a percentage, as obtained during tests aimed at recovering plutonium by reductive extraction from a molten chloride phase comprising, in addition to plutonium, neodymium, cerium, lanthanum, praseodymium and samarium, by means of a metallic phase comprising aluminium as solvent and magnesium as reducing agent; in this figure, lanthanum and samarium are not present because they have not been extracted.

[0057] [Fig.7] illustrates the evolution of the extraction rate of plutonium, neodymium, and cerium, lanthanum, praseodymium and samarium, denoted Ex and expressed as a function of the number of magnesium equivalents used, denoted nbéq Mg, as obtained during tests aimed at recovering plutonium by reductive extraction from a molten chloride phase comprising, in addition to plutonium, neodymium, cerium, lanthanum, praseodymium and samarium, by means of a metallic phase comprising aluminium as solvent and magnesium as reducing agent.

[0058] [Fig.8] illustrates the evolution of the cerium extraction rate, denoted ECe and expressed in percentages, depending on the number of magnesium equivalents used, denoted nbeq. Mg, as obtained during tests aimed at recovering cerium by reductive extraction from a molten chloride phase using a metallic phase comprising aluminium as a solvent and magnesium as a reducing agent.

[0059] Detailed description of a preferred embodiment of the invention

[0060] Reference is made to [Fig. 1] which schematically illustrates one method of implementation preferred method of the invention.

[0061] This embodiment was designed to treat and recycle a spent fuel salt from the irradiation in an MSR reactor of a fuel salt comprising at least one actinide chloride selected from uranium trichloride, plutonium trichloride, thorium tetrachloride and americium trichloride, dissolved in a solvent consisting of one or more chlorides selected from sodium chloride, potassium chloride and magnesium chloride.

[0062] Spent fuel salt is a molten salt which typically comprises actinides (including minor actinides such as americium) in the form of chlorides, lanthanides (La, Ce, Pr, Nd, Sm, Eu, etc.) and soluble alkali and alkaline-earth fission products (Cs, Rb, Ba and Sr), also in the form of chlorides, insoluble fission products (essentially represented by platinum group metals and molybdenum) and, possibly, volatile fission products (Xe, Kr, ZrCl4, etc.) if these have not already been volatilized in the reactor.

[0063] As can be seen in [Fig.1], the process, which forms a cycle between the extraction of the spent fuel salt from the reactor and the introduction of a new fuel salt into this reactor, mainly comprises 8 steps.

[0064] These steps are:

[0065] 1) A step labeled "Bubbling" in [Fig. 1], which aims to make an inert gas bubble, typically helium or argon, in the spent fuel salt to remove volatile fission products that may still be present in that salt.

[0066] The gases thus recovered can then be processed in a unit dedicated to the separation of aerosols and gases.

[0067] 2) A step labeled "Digestion", which aims to eliminate spent combustible salt obtained at at the end of the "Bubbling" stage the insoluble fission products by contacting this salt with a metallic sheet comprising a metal in liquid state, immiscible with salt - which sheet is, in the context of the present preferred implementation method, a sheet of liquid zinc - then separation of the spent combustible salt from the metallic sheet, whereby the insoluble fission products are adsorbed by the zinc.

[0068] The following are obtained at the end of this step: - a saline phase, denoted Si on [Fig. 1], which corresponds to the spent combustible salt now devoid of both volatile and insoluble fission products but still containing actinides, lanthanides, and soluble alkali and alkaline earth fission products, and - a metallic phase which includes zinc in liquid form and insoluble fission products adsorbed by this zinc.

[0069] 3) A step labeled "Distillation", which aims to subject the obtained metallic phase at the end of the "Digestion" stage to a distillation, for example at a temperature of 910 °C to 1200 °C, possibly under reduced pressure, to recover the zinc present in this phase, which allows it to be reused in the "Digestion" stage of a subsequent cycle.

[0070] The metallic phase obtained at the end of this step can then be sent to a unit responsible for processing it either for the recovery of all or part of the fission products it contains and, in particular, platinum group metals or for their conditioning in a specific matrix.

[0071] 4) A step labeled “Extraction Ans”, which aims to extract from the saline phase Si the actinides present in this phase but without extracting the lanthanides or the soluble alkali and alkaline-earth fission products also present.

[0072] This extraction is a reducing extraction which is carried out by bringing the saline phase Si into contact with a metallic sheet, noted Red-Al on the [Fig.1], comprising a metal in the liquid state, immiscible with the saline phase - this metal being aluminium in the context of the present preferred embodiment - and a reducing agent which is an alkali or alkaline-earth metal and, preferably, the alkali or alkaline-earth metal which is part of the composition of the solvent of the spent fuel or one of the alkali or alkaline-earth metals which are part of the composition of this solvent.

[0073] Thus, for example, if the solvent of the spent fuel salt is sodium chloride, then the reducing metal is preferably sodium. If the solvent of the spent fuel salt is a mixture of sodium chloride and potassium chloride, then the reducing metal is preferably sodium or potassium, whereas if the solvent of the spent fuel salt consists of a mixture of sodium chloride and magnesium chloride, then the reducing metal is preferably sodium or magnesium.

[0074] This reducing agent allows the actinides present in the saline phase Si to be reduced to their oxidation state 0 and, therefore, to the state of metals, which leads to their transfer, by chemical affinity, into the metallic layer in the form of an aluminum alloy.

[0075] For the sake of clarity, it will be assumed in what follows that the solvent of the spent fuel is a NaCl-MgCl2 solvent and, consequently, the reducing metal used in the "Extraction Ans" step is magnesium.

[0076] After separation of the metallic layer from the saline phase, the following is obtained: - a metallic phase, denoted Ans-Al in [Fig. 1], comprising actinide-aluminum alloys, and - a saline phase, noted S2 on [Fig.1], comprising lanthanides, soluble alkali and alkaline-earth fission products and, possibly, a fraction of actinides that have not been extracted.

[0077] 5) A step labeled "Lns Extraction", which aims to extract from the saline phase S2 the lanthanides and, where applicable, the fraction of actinides still present in this phase but without extracting the soluble alkali and alkaline-earth fission products.

[0078] This step also consists of a reductive extraction which is also carried out by bringing the saline phase S2 into contact with a metallic mat, this mat comprising the same metal as that of the metallic mat used in the "Extraction Ans" step, namely aluminium in the context of the present preferred method of putting, and the same reducing agent as that used in the "Extraction Ans" step, namely magnesium.

[0079] After separation of the saline phase from the metallic layer, the following are obtained: - a metallic phase, denoted Lns-Al in [Fig. 1], which comprises lanthanide-aluminum alloys and, where applicable, a small amount of actinide-aluminum alloys, which metallic phase can be sent to a processing unit dedicated to conditioning these alloys in a specific matrix, and - a saline phase S3 which is devoid of: * of volatile fission products since they are eliminated in the "Bubbling" stage, * of insoluble fission products since they are eliminated during the "Digestion" stage, * of actinides since they are extracted at the "Ans Extraction" stage as well as at the "Lns Extraction" stage in the case where a fraction of actinides remains unextracted at the end of the "Ans Extraction" stage, and * of lanthanides, but which, on the other hand, is loaded with magnesium chloride.

[0080] 6) A step labeled "Electrolysis", which aims to reduce, by electrolysis, the content in magnesium chloride of the saline phase S3 to bring this content to that which should be presented by the solvent of the new combustible salt obtained at the end of the treatment of the used combustible salt.

[0081] This electrolysis is preferably carried out using a positive electrode, or anode, made of an inert material of the graphite type and a negative electrode, or cathode, made of a metallic sheet comprising a metal in the liquid state, in which case the electrolysis produces dichlorine at the anode - which is recovered - while the reducing agent of the saline phase S3 combines with the metal of the cathode.

[0082] After separation of the saline phase from the metallic layer, the following is obtained: - a saline phase S4 which is now composed almost entirely of the solvent from the spent combustible salt and the chlorides of the soluble alkali and alkaline earth fission products, and - a metallic sheet enriched with magnesium.

[0083] By using, for the "Electrolysis" step, a metal sheet comprising the same metal as that of the metal sheet used in the "Extraction Ans" and "Extraction Lns" steps, namely aluminium in the context of this preferred embodiment, this step makes it possible to reform a Red-Al metal sheet usable in "Extraction Ans" and "Extraction Lns" steps of a subsequent cycle.

[0084] 7) A step labeled "Ans Deextraction", which aims to recover, in the form of chlorides, the actinides present, in alloy form, in the metallic phase obtained at the end of the "Extraction Ans" step while transferring them into the saline phase S4 obtained at the end of the "Electrolysis" step.

[0085] To do this, this step includes bringing the metallic phase obtained at the end of the "Extraction Ans" step into contact with the saline phase S4 as well as bubbling some of the dichlorine produced during the "Electrolysis" step into this metallic phase, then separating the metallic phase from the saline phase.

[0086] After separation of the saline phase from the metallic phase, the following is obtained: - a saline phase S5 comprising actinide chlorides in a solvent of the same type as the spent fuel solvent, and - a metallic phase which consists only of aluminium, which can be reused in an "Electrolysis" step of a subsequent cycle.

[0087] 8) A step labeled "Fissil material adjustment", which aims to adjust the content of fissile material from the saline phase S5.

[0088] This adjustment consists of adding at least one actinide chloride to this saline phase, this chloride being advantageously obtained beforehand - as illustrated in [Fig.1] - from the corresponding oxide, noted AnOx, by: - ​​hydrochlorination of this oxide with hydrogen chloride, or - carbochlorination of this oxide with carbon and dichlorine, using part of the dichlorine produced during the "Electrolysis" step to produce the hydrogen chloride used for hydrochlorination or to carry out carbochlorination.

[0089] The combustible salt obtained at the end of this step can then be used as new fuel salt in the reactor.

[0090] All the steps just described are, of course, carried out at temperatures at which the saline phases treated during these steps and the metals used are in a liquid state, that is to say typically temperatures between between 550 °C and 850 °C depending on the composition of said saline phases and said metals. Validation of the invention process

[0091] I - Reductive extraction of actinides from a molten chloride phase

[0092] The possibility of recovering actinides from a molten chloride phase by reductive extraction has been verified by a thermodynamic approach and by various experimental reductive extraction tests. Thermodynamic approach:

[0093] This approach, designed to assess the influence of the type of metallic solvent used to perform the reductive extraction on the separation factor between plutonium (used as an actinide) and cerium (used as a lanthanide), as well as on the separation factor between plutonium and magnesium (used as a reducing agent), focused on a system comprising: - a molten chloride phase containing plutonium (1 mol%) and cerium (1.5 mol%) as trichlorides in a NaCl-MgCl2 mixture (67.25 mol%-31.75 mol%), and - a metallic phase containing magnesium in a metal chosen from aluminium, gallium, bismuth and zinc as solvent.

[0094] The separation factors between plutonium and cerium, FSPu / Ce, and between plutonium and magnesium, FSPu / Mg, thus determined are illustrated in [Fig.2].

[0095] As this figure shows, the use of aluminum as the solvent for the metallic phase should lead to the highest FSPu / Ce and FSPu / Mg values. The other metals (Ga, Bi, and Zn) also allow the separation of plutonium from both cerium and magnesium, but with lower performance. Experimental tests:

[0096] Tests were conducted to study the influence of the molar ratio between magnesium (used as a reducing agent) and plutonium (used as an actinide) on the extraction rate and, therefore, the recovery of plutonium from a phase of molten chlorides at 600 °C.

[0097] To achieve this, the following were used: - a molten chloride phase comprising plutonium trichloride (1.2 mol%) in a NaCl-MgCl2 mixture (67.4 mol%-31.4 mol%), and - a metallic phase comprising magnesium, at a level of 1 to 3 equivalents of magnesium, in zinc; the number of equivalents is defined as the molar ratio Mg / PuCl3 (i.e. Nbeq. Mg = nMg / nPuCl3).

[0098] The two phases were added cold to a graphite crucible, and the assembly was then heated to 600 °C in a well furnace. A minimum of 6 hours separate the additions of magnesium, which allowed for an increase in the number of equivalents. At the end of the tests, the furnace was turned off, and the phases were mechanically separated when a temperature of 200°C was reached. Samples of the saline and metallic phases were taken before each addition of magnesium. The samples were dissolved in a nitric acid solution, and the elements present in solution were quantified by inductively coupled plasma atomic emission spectrometry (ICP-AES).

[0099] The plutonium extraction rates, EPu, thus obtained are illustrated in [Fig.3].

[0100] This figure shows that the use of 1 equivalent of magnesium makes it possible to extract 67% of the plutonium initially present in the chloride phase, while the use of 3 equivalents of magnesium allows the extraction of 99.5% of the plutonium initially present in this same phase.

[0101] Other tests consisted of tests separating plutonium from a series of lanthanides, namely neodymium, cerium, lanthanum, praseodymium and samarium, from a molten chloride phase at 600 °C.

[0102] In these tests, the following were used: - a molten chloride phase comprising plutonium, neodymium, cerium, lanthanum, praseodymium and samarium in the form of trichlorides in a NaCl-MgCl2 mixture, and whose molar composition is given in Table 1 below:

[0103] [Tables1] Chlorides

[0104] and - a metallic phase comprising magnesium in zinc, the number of magnesium equivalents present in this phase being progressively increased during the tests; here too, the number of magnesium equivalents is defined as the molar ratio Mg / PuCl3 (i.e. Nbeq. Mg = nMg / nPuCl3).

[0105] These tests were carried out following an operating protocol identical to that described above.

[0106] The results are illustrated in [Fig.4] which shows the lanthanide extraction rates, ELns, as a function of the plutonium extraction rate, EPu, and in [Fig.5] which shows the plutonium and lanthanide extraction rates, Ex, as a function of the number of magnesium equivalents used, Nbéq Mg.

[0107] As shown in [Fig.4], the extraction rates of neodymium, cerium, lanthanum and praseodymium are higher when the plutonium extraction rate is high, which means that it may be advantageous to recover less actinides and, in particular, plutonium during the "Extraction Ans" step to minimize the amount of lanthanides likely to be co-extracted with the actinides, knowing that if any actinides are not extracted at the "Extraction Ans" step, then they will be extracted jointly with the lanthanides at the "Extraction Lns" step.

[0108] The plutonium extraction rate can be chosen using a number of magnesium equivalents as shown in [Fig.5].

[0109] It should be noted that the samarium was not extracted at all under the conditions under which these tests were carried out, hence its absence in figures 4 and 5.

[0110] Other tests consisted of tests separating plutonium from a series of lanthanides, namely neodymium, cerium, lanthanum, praseodymium and samarium, from a phase of molten chlorides at 700°C.

[0111] In these tests, the following were used: - a molten chloride phase comprising plutonium, neodymium, cerium, lanthanum, praseodymium and samarium in the form of trichlorides in a NaCl-MgCl2 mixture, and whose molar composition is given in Table 2 below:

[0112] [Tableaux2] Chlorides

[0113] and - a metallic phase comprising magnesium in aluminium, the number of magnesium equivalents present in this phase being progressively increased during the tests; here again, the number of magnesium equivalents is defined as the molar ratio Mg / PuCl3 (i.e. Nbéq Mg = nMg / nPuCl3).

[0114] The tests were carried out following an operating protocol identical to that described above, except that the temperature prevailing in the well furnace was 700 °C and not 600 °C.

[0115] The results of these tests are illustrated in [Fig.6] which shows the lanthanide extraction rates, ELns, as a function of the plutonium extraction rate, EPu, and in [Fig.7] which shows the plutonium and lanthanide extraction rates, Ex, as a function of the number of magnesium equivalents used, Nbéq . Mg.

[0116] A comparison of figures 6 and 7 with figures 4 and 5 shows that the use of aluminium as the solvent for the metallic phase, instead of zinc, minimizes the recovery of lanthanides during the reductive extraction of plutonium compared to that obtained with zinc, with the possibility of recovering 95% of the plutonium without lanthanides.

[0117] II - Reductive extraction of lanthanides from a molten chloride phase

[0118] Various reductive extraction tests of lanthanides, complementary to the previous ones, were carried out using molten chloride phases comprising cerium but free of any plutonium.

[0119] By a first series of tests carried out at 700 °C using a molten chloride phase comprising between 1 mol% and 10 mol% of cerium trichloride in a NaCl-MgCl2 mixture (68 mol %-32 mol %) and a metallic phase comprising from 1.5 to 14 equivalents of magnesium for 1 equivalent of cerium, in aluminium, it was possible to verify that, as with plutonium, cerium is extracted more and more as the number of equivalents of magnesium used is high.

[0120] This is illustrated in [Fig.8] which shows the cerium extraction rates obtained as a function of the number of magnesium equivalents used.

[0121] In addition, tests aimed at varying the concentration of cerium trichloride ([CeCl3]) in the NaCl-MgCl2 mixture, the mass of aluminium in the metallic phase and / or the number of magnesium equivalents used (nbeq. Mg) were also carried out.

[0122] The extraction rates of cerium, ECe, thus obtained are presented in Table 3 below.

[0123] [Tables3] [CeCM Mass Ai [g) gS 0' 1 15 7 SB 1 15 3.5 92 1 30 7 99 2 30 33 9 g 1 30 S3 5 75 7 w 10 75 7 993

[0124] As this table shows, the extraction performance appears to be independent of both the concentration of the element to be extracted in the molten chloride phase and the amount of metallic solvent used. Only the amount of reducing metal, in this case magnesium, appears to be a determining factor in the extraction rates obtained. References cited

[0125] [1] GB-A-2 536 857 [2] GB-A-2 554 068 [3] US-A-2017 / 0301413

Claims

1. Demands Pyrometallurgical process for the treatment and recycling of spent fuel salt from a molten chloride nuclear reactor, the spent fuel salt being a molten salt comprising soluble alkali and alkaline earth actinides, lanthanides and fission products in the form of chlorides, insoluble fission products and, optionally, volatile fission products, in a solvent consisting of one or more chlorides selected from alkali and alkaline earth metal chlorides, which process comprises at least the following steps: a) extraction of actinides from spent fuel salt, the extraction of actinides comprising bringing the salt into contact with a medium comprising a metal Mi in liquid form, immiscible with salt, and reducing the degree of oxidation of the actinides by a reducing metal RedB alloyed with the metal Mb or by an electric current, thereby obtaining a saline phase depleted in actinides and a metallic phase enriched in actinides in alloyed form with the metal Mb, which are then separated from each other; b) extraction of lanthanides from the saline phase obtained at the end of step a), the extraction of lanthanides comprising bringing the saline phase into contact with a medium comprising a metal M2 in liquid form, immiscible with the saline phase, and reducing the degree of oxidation of the lanthanides by a reducing metal Red2, alloyed with the metal M2 or by an electric current, thereby obtaining a saline phase depleted in lanthanides and a metallic phase enriched in lanthanides in alloyed form with the metal M2, which are separated from each other; c) deextraction of actinides from the metallic phase obtained at the end of step a), the deextraction comprising contacting the metallic phase with both the saline phase obtained at the end of step b) and a chlorine source, thereby obtaining a saline phase comprising actinide chlorides and a metallic phase depleted in actinides, which are separated from each other; d) adjustment of the fissile material content of the saline phase obtained at the end of step c) by adding at least one actinide chloride to the saline phase, thereby obtaining a new fuel salt for a molten chloride nuclear reactor.

2. A method according to claim 1, wherein each of the metals Mi and M2 is selected from aluminium, gallium, bismuth, cadmium, lead, tin and zinc.

3. A method according to claim 1 or claim 2, wherein step a) comprises a reduction of the degree of oxidation of the actinides by the reducing metal Redi.

4. A method according to any one of claims 1 to 3, wherein step b) comprises a reduction of the degree of oxidation of the lanthanides by the reducing metal Red2.

5. A method according to any one of claims 1 to 4, wherein each of the reducing metals Redi and Red2 is an alkali or alkaline earth metal.

6. A method according to claim 5, wherein each of the reducing metals Redi and Red2 is the alkali or alkaline earth metal that is part of the solvent composition of the spent fuel salt or one of the alkali or alkaline earth metals that are part of the solvent composition of the spent fuel salt.

7. A method according to any one of claims 1 to 6, wherein the metal M2 is identical to the metal Mi and the reducing metal Red2 is identical to the reducing metal Redi.

8. A method according to claim 7, wherein the metal Mi and the metal M2 are aluminium.

9. A method according to any one of claims 1 to 8, wherein the contacting of the metallic phase obtained at the end of step a) with a chlorine source includes bubbling of a gaseous chlorine source into the metallic phase.

10. A method according to claim 9, wherein the chlorine gas source is dichlorine.

11. A process according to any one of claims 1 to 10, comprising, between steps b) and c), a step of reducing the content of reducing metal Redi and, optionally, of reducing metal Red2 if the reducing metal Red2 is different from the reducing metal Redi, of the saline phase obtained at the end of step b).

12. A process according to claim 11, wherein the reduction step comprises electrolysis of the saline phase obtained at the end of step b), the electrolysis comprising the use of an anode made of an inert material and a cathode comprising a metal M3 in the state liquid, not miscible with the saline phase, whereby dichlorine is released at the anode while the reducing metal Redi and, where applicable, the reducing metal Red2 combine with the metal M3.

13. A method according to claim 12, wherein the metal M3 is identical to the metal Mb

14. A process according to claim 12 or claim 13, wherein the dichlorine produced by electrolysis is used as a source of chlorine in step c).

15. A process according to any one of claims 12 to 14, wherein the dichlorine produced by electrolysis is used to produce said at least one actinide chloride, which is added in step d).

16. A process according to any one of claims 1 to 15, further comprising, before step a): i) a step for removing volatile fission products from the spent fuel salt that may be present in the spent fuel salt; and / or ii) a step for removing insoluble fission products from the spent fuel salt.

17. A method according to claim 16, wherein step i) comprises bubbling an inert gas into the spent combustible salt, typically helium or argon.

18. A process according to claim 16, wherein step ii) comprises contacting the spent combustible salt with a medium comprising a liquid metal M4, immiscible with the salt and capable of selectively adsorbing the insoluble fission products, thereby obtaining a spent combustible salt free of insoluble fission products and a metallic phase comprising the metal M4 and the insoluble fission products, which are separated from each other.

19. A process according to claim 18, further comprising a distillation of the metallic phase obtained at the end of step ii) to separate the metal M4 from the insoluble fission products.

20. A method according to claim 18 or claim 19, wherein the metal M4 is zinc or cadmium.

21. A method according to any one of claims 1 to 20, wherein the spent combustible salt is obtained by irradiating a combustible salt comprising at least one actinide chloride selected from uranium trichloride, plutonium trichloride, the thorium tetrachloride and americium trichloride, in a solvent consisting of one or more chlorides selected from sodium chloride, potassium chloride and magnesium chloride.

22. Process according to claim 21, in which the spent fuel salt comes from the irradiation of a fuel salt of formula NaCl-UCl3, NaCl-PuCl3, NaCl-UCl3-PuCl3, NaCl-ThCl4-PuCl3, NaCl-ThCl4-UCl3-PuCl3, NaCl-MgCl2-PuCl3, NaCl-MgCl2-UCl3-PuCl3, NaCl-MgCl2-PuCl3-AmCl3, NaCl-KCl-UCl3 or NaCl-KCl-UCl3-PuCl3.

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