Pyrometallurgical process for treating and recycling a spent fuel salt of a molten chloride nuclear reactor

Abstract

The invention relates to a pyrometallurgical process for treating and recycling a spent fuel salt of a molten chloride nuclear reactor.

Description

[0001] Description

[0002] Title of the invention: PYROMETALURGICAL PROCESS FOR THE TREATMENT AND RECYCLING OF A SPENT FUEL SALT FROM A MOLTEN CHLORIDE NUCLEAR REACTOR

[0003] technical field

[0004] The invention relates to the field of spent fuel treatment from molten salt reactors or MSRs, also known by the acronym MSR for "Molten Salt Reactors".

[0005] 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.

[0006] Prior art

[0007] Molten salt nuclear reactors are reactors designed to use a liquid fuel which is a molten salt, typically based on fluorides or chlorides.

[0008] First conceived in the 1960s, molten salt reactor technology is now the focus of much attention because these reactors potentially offer many advantages, including high energy efficiency and flexibility of use.

[0009] 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.

[0010] The treatment and recycling of such salt must allow, at a minimum

[0011] - to selectively recover the actinides present in this salt so that they can be reused in the constitution of a new combustible salt, and - to eliminate the lanthanides from the solvent of the salt so that this solvent can also be reused in the constitution of a new combustible salt.

[0012] Ideally, it would be desirable for the processing and recycling of used fuel salt to allow for the recovery of more:

[0013] - 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

[0014] - on the other hand, insoluble fission products that are potentially valuable outside the nuclear sphere, such as platinum group metals and molybdenum.

[0015] Furthermore, it would be desirable for the treatment-recycling to be carried out in a closed or almost closed cycle so as to minimize the number of reagents used and the volume of waste produced during this treatment-recycling.

[0016] The state of the art in the treatment of spent fuel salt from a molten chloride nuclear reactor is extremely limited and unsatisfactory.

[0017] Indeed, in patent application GB 2536857, 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.

[0018] Furthermore, the implementation of this supposedly 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.

[0019] In patent application GB 2554068, 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.

[0020] The only example of this treatment, 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% UCl₂, and 20% PuCl₂. In this example, the plan is to bring the spent salt into contact with a bismuth-calcium alloy to extract all the actinides (i.e., U, Pu, Np, and Am) and approximately one-third of the lanthanides, and then to condition the actinide-depleted salt 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% UCl₂. All the actinides present in the bismuth-calcium alloy, except for uranium, exchange with the UCl₂ from 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.

[0021] Finally, a process for treating a 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.

[0022] Description of the invention

[0023] 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.

[0024] 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:

[0025] a) extraction of actinides from spent fuel salt, the extraction of actinides comprising contacting the salt 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 Redi, alloyed with the metal Mi, 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 Mi, which are separated from each other;

[0026] 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;

[0027] (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

[0028] d) adjusting 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. Thus, according to the invention:

[0029] - Actinides are selectively extracted in alloy form from lanthanides and soluble alkali and alkaline earth fission products, then lanthanides are selectively extracted secondarily, also in alloy form, from soluble alkali and alkaline earth fission products, these extractions being 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), then 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 lanthanide extraction, and

[0030] - 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.

[0031] 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.

[0032] Among these, preference is given to aluminium, the use of which has indeed been shown to lead to particularly high separation factors between actinides and lanthanides.

[0033] 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 Mi, 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.

[0034] For example:

[0035] - if the solvent of the spent combustible salt is of the NaCl type, then the reducing metal Redi is preferentially sodium;

[0036] - 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

[0037] - if the solvent of the spent combustible salt is of the NaCl-MgCb type, then the reducing metal Redi is preferentially sodium or magnesium.

[0038] As an example, the reduction of plutonium, present in plutonium trichloride in the +3 oxidation state, by magnesium alloyed with aluminum leads to the formation of a Pu-Al alloy and magnesium chloride, this reaction being represented by the following equation:

[0039] [Math 1]

[0040] PüCljjgel) + l>5Mg-Al(alloy) — PU-Al(alloy) + l,5MgCl2(salt). Preferably also, step b) is also a reducing extraction.

[0041] To do this, it is possible to use a metal M2 different from the metal Mi, 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 Redi, for example an alkali or alkaline earth metal different from that used as the reducing metal Redi.

[0042] However, to simplify the implementation of the process but also to enable this process to operate in a closed or quasi-closed cycle with, as a result, an optimal reduction in the number of reagents 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 aluminium - and that the reducing metal Red2 be identical to the reducing metal Redi.

[0043] In any case, 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.

[0044] As previously mentioned, the extraction of actinides and lanthanides, instead of being reductive extractions, can be carried out by electrolysis, for example, using a positive electrode, or anode, made of an inert material such as graphite, and, as a negative electrode, or cathode, a medium comprising metal Mi (for the extraction of actinides) and metal M2 (for the extraction of lanthanides), and 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.

[0045] During these electrolysises, dichlorine is produced at the anode which can advantageously be recovered and then used as a source of chlorine for the deextraction of the actinides, i.e. in step c). Indeed, in step c), the contacting of the metallic phase obtained at the end of step a) with a source of chlorine includes, preferably, bubbling - 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.

[0046] Preferably, dichlorine is used as the gaseous source of chlorine.

[0047] As an example, the extraction of plutonium from a metallic phase, in which it is alloyed with aluminum, by bubbling dichlorine through this metallic phase leads to the obtaining of plutonium trichloride on the one hand, and aluminum trichloride on the other.

[0048] 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.

[0049] 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).

[0050] This reduction can be achieved by electrolysis, for example, using a positive electrode made of an inert material such as graphite and a negative electrode made of a medium containing a liquid metal M3, immiscible with the salt phase. In this case, electrolysis produces chlorine at the anode, while the reducing metal Red1 and, if applicable, the reducing metal Red2 combine with the metal M3. Here too, the chlorine thus produced can be recovered and then used as a source of chlorine for the de-extraction of the actinides, i.e., in step c).

[0051] With the aim of ensuring the closed or nearly closed-loop treatment and recycling of spent fuel salt, the metal M3 is preferably the metal Mi from the metallic phase obtained after the actinide deextraction step. This allows the metal Mi, used for actinide extraction, to be reused in a step of the process other than that extraction. Furthermore, if the same metal is used as the reducing metals Redi and Red2, then the metal Mi-reducing metal alloy obtained after electrolysis can be used in further actinide and lanthanide extraction steps.

[0052] According to the invention, the process preferably comprises, before step a): i) a step for 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 for removing insoluble fission products from the spent fuel salt, so named because they are present in this salt in the form of metallic insolubles. This type of fission product is essentially represented by platinum group metals (Pt, Pd, Ru, Rh, ...) and molybdenum.

[0053] As known in itself, step i) may include bubbling an inert gas, such as helium or argon, into the spent combustible salt.

[0054] The volatile fission products thus removed from the spent fuel salt can then be sent to a unit dedicated to the separation of gases and aerosols.

[0055] 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 IV 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 IVU and the said insoluble fission products.

[0056] Advantageously, step ii) is complemented by a distillation step of the metallic phase from step ii) to separate the metal IV from the insoluble fission products. Therefore, the metal IVU is preferably a metal which, in addition to being capable of selectively adsorbing the insoluble fission products, has a sufficiently low boiling point to be easily distilled, as is notably the case for zinc (TE: 907 °C) and cadmium (TE: 767 °C).

[0057] Once separated from the insoluble fission products, the IVU 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.

[0058] If step ii) is not implemented, then the insoluble fission products are extracted from the salt with the actinides during their recovery.

[0059] Typically, the amount of fissile material present in the saline phase obtained at the end of step c) is insufficient to produce a new fuel salt capable of 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.

[0060] The actinide chloride(s) thus supplied may, in particular, have been previously obtained by:

[0061] - Chlorination of the corresponding oxalate or oxalates with hydrogen chloride, - Hydrochlorination of the corresponding oxide or oxides with hydrogen chloride,

[0062] - carbochlorination of the corresponding oxide or oxides with carbon and chlorine, and / or

[0063] - chlorination of the corresponding actinide or actinides in the metallic state with dichlorine.

[0064] The chlorine produced during the various electrolysis steps that may be carried out as part of the implementation of the process, and in particular during the electrolysis that may be carried out to reduce the content of the reducing metal Redi and, where applicable, Red2, can advantageously be used to produce said actinide chloride(s), whether:

[0065] - to produce the hydrogen chloride necessary for the chlorination of the oxalate or corresponding oxalates or for the hydrochlorination of the oxide or corresponding oxides, hydrogen chloride being able, in fact, to be produced by the reaction of chlorine with either hydrogen or methane and oxygen, or

[0066] - to carry out the carbochlorination of the corresponding oxide or oxides or the chlorination of the corresponding actinide or actinides in the metallic state. 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. A new fuel salt of this type is, for example, a salt of formula NaCI-UCI3, NaCI-PuCI3, NaCI-UCI3-PuCI3, NaCI-ThCI4-PuCI3, NaCI-ThCI4-UCI3-PuCI3, NaCI-MgCI2-PuCI3, NaCI-MgCI2-UCI3-PuCI3, NaCI-MgCI2-PuCI3-AmCI3, NaCI-KCI-UCI3or NaCI-KCI-UCI3-PuCI3.

[0067] Other advantages and features of the invention will become apparent from the following non-limiting supplementary description, which refers to the attached figures.

[0068] Brief description of the figures

[0069] [Fig. 1] is a schematic representation of a preferred embodiment of the process of the invention.

[0070] [Fig. 2] illustrates the separation factors between plutonium and cerium (FSp u / ce) on the one hand, and between plutonium and magnesium (FSp u / M g) 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.

[0071] [Fig. 3] illustrates the evolution of the plutonium extraction rate, denoted Ep uand expressed as a percentage, 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.

[0072] [Fig. 4] illustrates the evolution of the extraction rate of neodymium, cerium, lanthanum, and praseodymium, denoted E^s and expressed as percentages, as a function of the plutonium extraction rate, denoted Ep uand also expressed in 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.

[0073] [Fig. 5] illustrates the evolution of the extraction rate of plutonium, neodymium, cerium, lanthanum, praseodymium and samarium, denoted E x 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 in Figure 4.

[0074] [Fig. 6] illustrates the evolution of the extraction rate of neodymium, cerium, and praseodymium, denoted Eins and expressed as percentages, as a function of the extraction rate of plutonium, denoted Ep u 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 aluminium as solvent and magnesium as reducing agent; in this figure, lanthanum and samarium are not present because they have not been extracted.

[0075] [Fig. 7] illustrates the evolution of the extraction rate of plutonium, neodymium, cerium, lanthanum, praseodymium and samarium, denoted E xand expressed as a function of the number of magnesium equivalents used, denoted nbeq. 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.

[0076] [Fig. 8] illustrates the evolution of the cerium extraction rate, noted Ece and expressed in percentages, as a function of the number of magnesium equivalents used, noted nbéq. 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.

[0077] Detailed description of a preferred embodiment of the invention. Reference is made to Figure 1, which schematically illustrates a preferred embodiment of the process of the invention.

[0078] This implementation method was designed to treat and recycle a spent fuel salt from 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.

[0079] Spent fuel salt is a molten salt that typically includes 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 (mainly represented by platinum group metals and molybdenum) and, possibly, volatile fission products (Xe, Kr, ZrC, etc.) if these have not already been volatilized in the reactor.

[0080] As can be seen in Figure 1, the process, which forms a cycle between the extraction of spent fuel salt from the reactor and the introduction of new fuel salt into that reactor, mainly comprises 8 steps.

[0081] These steps are:

[0082] 1) A step marked "Bubbling" in Figure 1, which aims to bubble an inert gas, typically helium or argon, into the spent fuel salt to remove volatile fission products that may still be present in this salt.

[0083] The gases thus recovered can then be processed in a unit dedicated to the separation of aerosols and gases.

[0084] 2) A step labeled "Digestion," which aims to remove insoluble fission products from the spent combustible salt obtained at the end of the "Bubbling" step by contacting this salt with a metallic layer containing a metal in a liquid state, immiscible with salt—which layer is, in the context of this preferred embodiment, a layer of liquid zinc—followed by separation of the spent combustible salt from the metallic layer, thereby adsorbing the insoluble fission products by the zinc. The following are obtained at the end of this step:

[0085] - a saline phase, denoted Si in Figure 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

[0086] - a metallic phase which includes zinc in liquid form and insoluble fission products adsorbed by this zinc.

[0087] 3) A step marked "Distillation", which aims to subject the metallic phase obtained at the end of the "Digestion" step to 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" step of a subsequent cycle.

[0088] 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 packaging in a specific matrix.

[0089] 4) A step marked "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.

[0090] This extraction is a reductive extraction which is carried out by bringing the saline phase Si into contact with a metallic sheet, noted Red-AI in Figure 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.

[0091] 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.

[0092] This reducing agent reduces the actinides present in the saline phase Si 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.

[0093] For the sake of clarity, we will assume in what follows that the solvent of the spent fuel is a NaCl-MgCb solvent and, consequently, that the reducing metal used in the step

[0094] "Extraction Ans" is magnesium.

[0095] After separating the metallic layer from the saline phase, the following is obtained:

[0096] - a metallic phase, labeled Ans-AI in Figure 1, comprising actinide-aluminum alloys, and

[0097] - a saline phase, noted S2 in Figure 1, comprising lanthanides, soluble alkali and alkaline-earth fission products and, possibly, a fraction of actinides that have not been extracted.

[0098] 5) A step marked "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.

[0099] This step also consists of a reducing extraction, which is also carried out by contacting the saline phase S2 with a metallic sheet, this sheet comprising the same metal as the metallic sheet used in the step

[0100] "Extraction Ans", namely aluminum under this preferred method, and the same reducing agent as used in the "Extraction Ans" step, namely magnesium.

[0101] After separating the saline phase from the metallic layer, we obtain the following:

[0102] - a metallic phase, denoted Lns-AI in Figure 1, comprising lanthanide-aluminum alloys and, where appropriate, 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

[0103] - a saline phase S3 which is devoid of:

[0104] * of volatile fission products since they are eliminated in the "Bubbling" step, * of insoluble fission products since they are eliminated in the "Digestion" step, * of actinides since they are extracted in the "Ans Extraction" step as well as in the "Lns Extraction" step in the case where a fraction of actinides remains unextracted after the "Ans Extraction" step, and

[0105] * of lanthanides,

[0106] but which, on the other hand, is loaded with magnesium chloride.

[0107] 6) A step marked "Electrolysis", which aims to reduce, by electrolysis, the magnesium chloride content of the saline phase S3 to bring this content to that which must be presented by the solvent of the new combustible salt obtained at the end of the treatment of the used combustible salt.

[0108] This electrolysis is preferably carried out using a positive electrode, or anode, made of an inert material such as graphite and a negative electrode, or cathode, made of a metallic sheet including 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.

[0109] After separating the saline phase from the metallic layer, we obtain the following:

[0110] - 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

[0111] - a metallic sheet enriched with magnesium.

[0112] By using, for the "Electrolysis" step, a metal sheet comprising the same metal as that of the metal sheet used in the "Ans Extraction" and "Lns Extraction" steps, namely aluminium in the context of this preferred embodiment, this step makes it possible to reform a Red-AI metal sheet usable in "Ans Extraction" and "Lns Extraction" steps of a subsequent cycle.

[0113] 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 "Ans Extraction" step, while transferring them into the saline phase S4 obtained at the end of the "Electrolysis" step. To do this, this step includes contacting the metallic phase obtained at the end of the "Ans Extraction" step with the saline phase S4, as well as bubbling some of the dichlorine produced during the "Electrolysis" step into this metallic phase, and then separating the metallic phase from the saline phase.

[0114] After separating the saline phase from the metallic phase, we obtain the following:

[0115] - a saline phase S5 comprising actinide chlorides in a solvent of the same type as the spent fuel solvent, and

[0116] - a metallic phase which consists only of aluminium, which can be reused in an "Electrolysis" step of a subsequent cycle.

[0117] 8) A step noted as "Fissile material adjustment", which aims to adjust the fissile material content of the saline phase S5.

[0118] This adjustment consists of adding at least one actinide chloride to this saline phase, this chloride being advantageously obtained beforehand – as illustrated in Figure 1 – from the corresponding oxide, denoted AnO x , by :

[0119] - hydrochlorination of this oxide with hydrogen chloride, or

[0120] - carbochlorination of this oxide with carbon and chlorine,

[0121] using part of the chlorine produced during the "Electrolysis" step to produce hydrogen chloride used for hydrochlorination or to carry out carbochlorination.

[0122] The combustible salt obtained at the end of this step can then be used as new fuel salt in the reactor.

[0123] All the steps that have just been 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 550 °C and 850 °C depending on the composition of said saline phases and said metals.

[0124] Validation of the invention process

[0125] I - Reductive extraction of actinides from a molten chloride phase

[0126] The possibility of recovering actinides from a molten chloride phase by reductive extraction was verified by a thermodynamic approach and by various experimental reductive extraction tests. Thermodynamic approach:

[0127] 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 that would include:

[0128] - a molten chloride phase containing plutonium (1 mol%) and cerium (1.5 mol%) as trichlorides in a NaCl-MgCb mixture (67.25 mol%-31.75 mol%), and - a metallic phase containing magnesium in a metal selected from aluminium, gallium, bismuth and zinc as solvent.

[0129] The separation factors between plutonium and cerium, FSp u / ce, and between plutonium and magnesium, FSp u / Mg, thus determined are illustrated in figure 2.

[0130] As this figure shows, the use of aluminum as a solvent for the metallic phase must lead to FSp u / ce and FSp u / M g the highest. Other metals (Ga, Bi and Zn) also allow plutonium to be separated from both cerium and magnesium, but with lower performance.

[0131] Experimental tests:

[0132] 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 molten chloride phase at 600 °C.

[0133] To achieve this, the following were used:

[0134] - a molten chloride phase comprising plutonium trichloride (1.2 mol%) in a NaCl-MgCb mixture (67.4 mol%-31.4 mol%), and

[0135] - a metallic phase comprising magnesium, at a level of 1 to 3 magnesium equivalents, in zinc; the number of equivalents is defined as the molar ratio Mg / PuCb (i.e., Nbeq. Mg = n Mg / np u ci3).

[0136] The two phases were added cold to a graphite crucible, and the mixture was then heated to 600 °C in a well furnace. A minimum of 6 hours separated the magnesium additions, which increased 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 20 °C was reached. Samples of the saline and metallic phases were taken before each magnesium addition. 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). The plutonium extraction rates, Ep uThe results thus obtained are illustrated in Figure 3.

[0137] This figure shows that using 1 equivalent of magnesium allows 67% of the plutonium initially present in the chloride phase to be extracted, while using 3 equivalents of magnesium allows 99.5% of the plutonium initially present in the same phase to be extracted.

[0138] Other tests consisted of tests to separate plutonium from a series of lanthanides, namely neodymium, cerium, lanthanum, praseodymium and samarium, from a molten chloride phase at 600 °C.

[0139] These trials used:

[0140] - a molten chloride phase comprising plutonium, neodymium, cerium, lanthanum, praseodymium and samarium in the form of trichlorides in a NaCl-MgCb mixture, and whose molar composition is given in Table 1 below:

[0141] [Table 1]

[0142]

[0143] And

[0144] - 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 / n Pu ci3).

[0145] These tests were carried out following an operating protocol identical to that described above.

[0146] The results are illustrated in Figure 4, which shows the lanthanide extraction rates, E^s, as a function of the plutonium extraction rate, E Pu , as well as in Figure 5 which shows the extraction rates of plutonium and lanthanides, E x , depending on the number of magnesium equivalents used, Nbéq. Mg.

[0147] As shown in Figure 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 wise to recover less actinides and, in particular, plutonium during the "Ans Extraction" step to minimize the amount of lanthanides likely to be co-extracted with the actinides, knowing that if actinides are not extracted at the "Ans Extraction" step, then they will be extracted jointly with the lanthanides at the "Lns Extraction" step.

[0148] The plutonium extraction rate can be chosen using a number of magnesium equivalents as shown in Figure 5.

[0149] 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.

[0150] Other tests consisted of tests to separate plutonium from a series of lanthanides, namely neodymium, cerium, lanthanum, praseodymium and samarium, from a phase of chlorides molten at 700°C.

[0151] These trials used:

[0152] - a molten chloride phase comprising plutonium, neodymium, cerium, lanthanum, praseodymium and samarium in the form of trichlorides in a NaCl-MgCb mixture, and whose molar composition is given in Table 2 below:

[0153] [Table 2]

[0154]

[0155] And

[0156] - 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 / PuCb (i.e. Nbeq. Mg = nMg / npu).

[0157] The tests were carried out following an operating protocol identical to that described above, except that the temperature in the well furnace was 700 °C and not 600 °C.

[0158] The results of these tests are illustrated in Figure 6, which shows the lanthanide extraction rates, Ei.ns, as a function of the plutonium extraction rate, Ep u , as well as in Figure 7 which shows the extraction rates of plutonium and lanthanides, E x , depending on the number of magnesium equivalents used, Nbéq. Mg.

[0159] 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.

[0160] II - Reductive extraction of lanthanides from a molten chloride phase

[0161] Various reductive extraction tests of lanthanides, complementary to the previous ones, were carried out using molten chloride phases containing cerium but free of any plutonium. 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-MgCb 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 verified that, as with plutonium, cerium is extracted more as the number of equivalents of magnesium used is higher.

[0162] This is illustrated in Figure 8, which shows the cerium extraction rates obtained as a function of the number of magnesium equivalents used.

[0163] Furthermore, tests were also conducted to vary the concentration of cerium trichloride ([CeC]) in the NaCl-MgCb mixture, the mass of aluminum in the metallic phase, and / or the number of magnesium equivalents used (nbeq. Mg). The cerium extraction rates, Ece, thus obtained are presented in Table 3 below.

[0164] [Table 3]

[0165] >

[0166]

[0167] 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 quantity of the reducing metal, in this case magnesium, appears to be the determining factor in the extraction rates obtained.

[0168] References cited[1] GB-A-2536857

[0169] [2] GB-A-2554068

[0170] [3] US-A-2017 / 0301413

Claims

Demands 1. 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 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 contacting the salt 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 Redi, alloyed with the metal Mi, 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 Mi, which are 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 bringing the metallic phase into contact with both the saline phase obtained at the end of step b) and a source of chlorine, thereby obtaining a saline phase comprising actinide chlorides and a metallic phase depleted in actinides, which are separated from each other; d) adjusting 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 process 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 process 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 of the spent fuel salt or one of the alkali or alkaline earth metals that are part of the solvent 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 process 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 an 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 liquid state, immiscible with the saline phase, thereby obtaining the release of chlorine gas at the anode while the reducing metal Redi and, where applicable, the reducing metal Red2 combine with the metal M3.

13. Method according to claim 12, wherein the metal M3 is identical to the metal Mi.

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

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

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

21. A process according to any one of claims 1 to 20, wherein the spent fuel salt is obtained by irradiating a 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.

22. A process according to claim 21, wherein the spent combustible salt is obtained from the irradiation of a combustible salt of formula NaCI-UCI3, NaCI-PuCI3, NaCI-UCI3-PuCI3, NaCI-ThCI4-PuCI3, NaCI-ThC-UCh-PuCh, NaCI-MgCI2-PuCI3, NaCI-MgCI2-UCI3-PuCI3, NaCI-MgCI2-PuCI3-AmCI3, NaCI-KCI-UCI3 or NaCI-KCI-UCI3-PuCI3.

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