Classifications

• • G—PHYSICS
  • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
  • G21C—NUCLEAR REACTORS
  • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
  • G21C19/42—Reprocessing of irradiated fuel
  • G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
  • G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
• • C—CHEMISTRY; METALLURGY
  • C01—INORGANIC CHEMISTRY
  • C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
  • C01G56/00—Compounds of transuranic elements
  • C01G56/001—Preparation involving a liquid-liquid extraction, an adsorption or an ion-exchange
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
  • Y02E30/00—Energy generation of nuclear origin
  • Y02E30/30—Nuclear fission reactors
• • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
  • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
  • Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
  • Y02W30/00—Technologies for solid waste management
  • Y02W30/50—Reuse, recycling or recovery technologies

Definitions

• the present invention relates to a process for the treatment of spent nuclear fuel which has, among other advantages, that of preventing the storage on the site where these fuels are treated of purified plutonium (that is to say totally decontaminated to fission products), whether or not mixed with uranium or uranium and neptunium.
• This process is particularly applicable in the treatment of uranium oxide fuels and mixed oxide fuels of uranium and plutonium.
• the used nuclear fuel treatment plants currently use the PUREX (Plutonium Uranium Refining by Extraction) process to recover the uranium and plutonium present in these fuels.
• PUREX Plutonium Uranium Refining by Extraction
• the extractant used is tri-n-butyl phosphate, which has a particular affinity for uranium and plutonium.
• the PUREX process as implemented in the UP2-800 factory in La Hague, France, comprises schematically three cycles, namely:
• a first cycle that aims to jointly decontaminate uranium and plutonium vis-à-vis two actinides (III), americium and curium, as well as most of the fission products, and to achieve a partition of uranium and plutonium in two streams; and
• second cycle uranium and “second cycle plutonium” and which purify separately uranium and plutonium after their partition.
• COEX TM process for co-extraction
• the COEX TM process provides, after uranium and plutonium decontamination operations similar to those implemented in the first cycle of the PUREX process, to partition the uranium and plutonium so that obtain a first stream containing plutonium, uranium and, possibly, neptunium, and a second stream containing uranium and possibly neptunium but not containing plutonium.
• the "second plutonium cycle" of the PUREX process is replaced by a cycle that aims to purify plutonium and uranium and, where appropriate, neptunium present in the first stream from the partition. fission products likely to be still present in this stream.
• the COEX TM process includes a storage step.
• This storage is located either between the cycle to purify the plutonium and uranium and, where appropriate, the neptunium present in the first stream from the partition and the co-conversion, just after the co-conversion.
• the first case it is a mixture of plutonium, uranium and possibly neptunium purified in aqueous solution that is stored, while in the second case, it is a mixture of plutonium, uranium and optionally purified neptunium in solid form which is stored.
• this decontamination comprising at least one co-extraction operation of the uranium in the oxidation state VI, the plutonium in the oxidation state IV, and the neptunium in the oxidation state VI, in a solvent phase, immiscible with water and containing at least one extractant in an organic diluent, and at least one washing operation of the solvent phase obtained at the end of this operation coextraction, with a nitric aqueous phase;
• step b) the partitioning of the uranium, plutonium and neptunium present in the solvent phase obtained at the end of step a) into a first and a second aqueous phase, the first aqueous phase containing either plutonium without uranium or neptunium, or a mixture of plutonium and uranium without neptunium, or a mixture of plutonium, uranium and neptunium, and the second aqueous phase containing either a a mixture of uranium and neptunium without plutonium, that is, uranium without plutonium nor neptunium;
• step d) the purification of the plutonium or the mixture of plutonium and uranium, or the mixture of plutonium, uranium and neptunium, present in the first aqueous phase obtained at the end of step c) with respect to -vis fission products still in this phase, this purification comprising at least one addition of uranium to obtain, after this purification, an aqueous solution containing either a mixture of plutonium and uranium, or a mixture of plutonium, uranium and neptunium; and
• step d) the co-conversion of the mixture of plutonium and uranium, or of the mixture of plutonium, uranium and neptunium present in the aqueous phase obtained at the end of step d) into a mixed oxide.
• the plutonium is never left alone since, up to the purification step d), it is at least associated with fission products, or even with uranium or with a mixture of uranium and neptunium according to the manner in which the partitioning step b) is carried out, while starting from the purification step d), which aims at ridding it completely of fission products, it is at least associated with uranium, or even a mixture of uranium and neptunium;
• step d) of purification makes it possible to produce an aqueous solution containing a mixture of plutonium, uranium and, optionally, neptunium free of fission products, suitable for being converted into a mixed oxide;
• a storage step which lies between step b) of partition and step d) of purification so that the The aqueous phase which is stored contains plutonium which has not yet been purified with respect to the fission products.
• the method of the invention thus allows to avoid that is present on the site where the spent nuclear fuel is treated, purified plutonium, even mixed with uranium or uranium and neptunium.
• step b) is carried out so that the first aqueous phase contains plutonium but without uranium or neptunium, and the second aqueous phase contains uranium. and neptunium but without plutonium.
• step b) preferably comprises:
• step a a plutonium extraction operation in the solvent phase obtained at the result of step a), this plutonium being deextracted in the oxidation state III by means of a nitric aqueous phase containing a reducing agent for reducing plutonium (IV) to plutonium (III) and neptunium (VI) to neptunium (IV) without reducing uranium, for example uranous nitrate - or uranium (IV) - stabilized with an antinitrous agent, for example hydrazinium nitrate;
• step b 3 a washing operation of the aqueous phase obtained after the operation bi) to remove from this phase the uranium and neptunium fractions having followed the plutonium in aqueous phase during said operation bi), this washing being carried out using a solvent phase of the same composition as that used in step a).
• step c) preferably comprises:
• Ci an oxidation operation for bringing the plutonium (III) present in the aqueous phase obtained after the operation b 3 ) to the oxidation state IV;
• step d) preferably comprises:
• step c di) a plutonium (IV) extraction operation present in the aqueous phase obtained after step c), this extraction being carried out using a solvent phase of the same composition as that used in step at) ;
• aqueous phase which contains plutonium (III) and uranium (IV) but which no longer contains fission products and which can therefore be subjected to step e) of co-conversion to obtain of a mixed oxide (U, Pu) O 2 .
• step b) is carried out so that the first aqueous phase contains plutonium and uranium but without neptunium, and that the second aqueous phase contains uranium and neptunium but without plutonium.
• step b) preferably comprises:
• step a bi) an operation for the extraction of the plutonium and a fraction of the uranium present in the solvent phase obtained at the end of step a), this plutonium and this uranium being respectively extracted in the oxidation state III and VI using a nitric aqueous phase containing a reducing agent for reducing plutonium (IV) to plutonium (III) and neptunium (VI) to neptunium (IV) without reducing uranium, for example uranium (IV) stabilized with hydrazinium nitrate;
• step c) preferably comprises:
• Ci an oxidation operation to bring the plutonium (III) present in the aqueous phase obtained after the operation b 3 ) to the oxidation state IV and, if uranium (IV) is present in this phase, to bring this uranium (IV) to oxidation state VI, uranium being, indeed, as the plutonium more stable in aqueous solution in the oxidized state than in the reduced state;
• step d) preferably comprises:
• step c di) a co-extraction operation of plutonium (IV) and uranium (VI) present in the aqueous phase obtained at the end of step c), this coextraction being carried out by means of a phase solvent of the same composition as that used in step a);
• a plutonium de-extraction operation present in the solvent phase thus washed this plutonium being deextracted in the oxidation state III by means of a nitric aqueous phase containing a reducing agent for reducing plutonium (IV) to plutonium (III), for example nitrate stabilized hydroxyl ammonium nitrate hydrazinium; and 4) a washing step of the aqueous phase obtained at the end of step d 3) to remove from this phase the uranium (VI) having followed the plutonium (III) in the aqueous phase during said operation d 3 ), this washing being carried out using a solvent phase of the same composition as that used in step a) and comprising at least one addition of uranium (IV) to said aqueous phase, preferably at the end of operation d 4 ).
• an aqueous phase which contains plutonium (III) and uranium (IV) but which no longer contains fission products and which can therefore be subjected to step e) of co- conversion for obtaining a mixed oxide (U, Pu) O 2 .
• step b) is carried out so that the first aqueous phase contains both plutonium, uranium and neptunium, and that the second aqueous phase contains uranium but no plutonium or neptunium.
• step b) preferably comprises:
• step c) preferably comprises:
• Ci an oxidation operation to bring the plutonium (III) and the neptunium (V) present in the aqueous phase obtained at the end of the operation b 2 ) respectively in the oxidation state IV and VI;
• step d) preferably comprises:
• step c di) a co-extraction operation of plutonium (IV), uranium (VI) and neptunium (VI) present in the aqueous phase obtained at the end of step c), this co-extraction being carried out by means of a solvent phase of the same composition as that used in step a); d 2 ) a washing operation of the solvent phase obtained after the operation di) to remove from this phase the fission products that followed the plutonium, uranium and neptunium in the solvent phase during said operation di), this washing being carried out by means of a nitric aqueous phase;
• V by means of a nitric aqueous phase containing a reducing agent for reducing plutonium (IV) to plutonium (III) and neptunium (VI) to neptunium (V), for example hydroxylammonium nitrate stabilized with hydrazinium nitrate; and
• d 4 a washing operation of the aqueous phase obtained after the operation d 3 ) to remove from this phase the uranium (VI) which followed the plutonium (III) and the neptunium (V) in aqueous phase during said operation d 3 ), this washing being carried out using a solvent phase of the same composition as that used in step a) and comprising at least one addition of uranium (IV) to said aqueous phase, preference at the end of the operation d 4 ).
• step e) is preferably carried out as described in the international application.
• the plutonium in the oxidation state III by stabilizing the plutonium in the oxidation state III, the uranium in the oxidation state IV, and, if applicable, the neptunium in the oxidation state IV, by a monocharged cation consisting solely of atoms selected from oxygen, carbon, nitrogen and hydrogen atoms such as the hydrazinium cation;
• the mixed oxide thus obtained which is in the form of a powder, can then be used directly for the manufacture of pellets of a MOX nuclear fuel, for example by the MIMAS process (for MIcronized MASter Blend).
• this mixed oxide preferably has a U / Pu mass ratio equal to or substantially equal to 50/50 when it does not contain neptunium, and a U / mass ratio.
• Pu / Np equal or substantially equal to 49/49/2 when it contains neptunium.
• step d) and step e it is possible to adjust, either between step d) and step e), or at the beginning of step e), just before stabilizing the plutonium, uranium and, the where appropriate, neptunium, the mass ratio U / Pu or U / Pu / Np of the aqueous phase obtained at the end of step d) to that of the mixed oxide that it is desired to obtain.
• the aqueous phase obtained after step d) immediately has a mass ratio U / Pu or U / Pu / Np consistent with that of this mixed oxide or requiring only a very low subsequent adjustment.
• the uranium (IV) which is added to the plutonium during step d) is preferably in an amount such that the aqueous phase obtained at the end of this step has a mass ratio U / Pu or U / Pu / Np in accordance with that of the mixed oxide obtained in step e).
• the duration of step c) is at least fifteen days and preferably between one month and twelve months, knowing that in this step, the operations ci) and C 2 ) have negligible durations compared to the duration of the operation C 3 ) storage.
• the aqueous phase stored during operation C 3 has a plutonium or a mixture of plutonium and uranium content or still in a mixture of plutonium, uranium and neptunium of 200 to 250 g / L.
• the extractant of the solvent phase used in step a) and hence in steps b) and d) is, preferably, selected from the extractants which more strongly complex actinides in the oxidation state IV and / or VI than the actinides in the oxidation state III and / or V.
• This extractant may especially be a tri-alkyl phosphate such as tri-n-butyl phosphate (or TBP), tri-isobutyl phosphate (or TiBP) or tri-isoamyl phosphate.
• TBP tri-n-butyl phosphate
• TiBP tri-isobutyl phosphate
• TBP tri-n-butyl phosphate
• TiBP tri-isobutyl phosphate
• tri-isoamyl phosphate tri-isoamyl phosphate
• the organic diluent in which this extractant is found can itself be chosen from the various hydrocarbons proposed for liquid-liquid extractions such as toluene, xylene, t-butylbenzene, tri-isopropylbenzene, kerosene and linear dodecanes. or branched, such as n-dodecane or hydrogenated tetrapropylene (or TPH).
• solvent phases containing tri-n-butyl phosphate in a dodecane and in a volume ratio equal to or substantially equal to 30/70 are preferred to use, as in the PUREX process, solvent phases containing tri-n-butyl phosphate in a dodecane and in a volume ratio equal to or substantially equal to 30/70.
• step a) preferably comprises:
• the nitric aqueous phase used for the bi operations preferably contains from 0.05 to 1 mol / L of nitric acid, whereas the aqueous phase used during the operations b 2 ) preferably contains 0 at 0.05 mol / L of nitric acid; the nitric aqueous phase used for the operations d 2 ) preferably contains from 1 to
• aqueous nitric phase used for the operations 3 preferably contains from 0.05 to 2 mol / L of nitric acid.
• the method of the invention also comprises uranium purification operations present in the second aqueous phase obtained at the end of step b) to complete its decontamination into fission products and, optionally, to separate it from neptunium. if said step b) is carried out so that this second aqueous phase contains both uranium and neptunium.
• uranium purification operations can be carried out as in any conventional PUREX process (see, for example, article BN 3,650 of the "Nuclear Engineering” Treatise on Engineering Techniques,
• FIG. 1 represents a block diagram of a first mode of implementation of the method of the invention, designed to obtain a mixed oxide (U / Pu) O 2 .
• FIG. 2 represents a block diagram of a second mode of implementation of the method of the invention, also designed to obtain a mixed oxide (U / Pu) O 2 .
• FIG. 3 represents a block diagram of a third mode of implementation of the method of the invention, designed to obtain a mixed oxide (U / Pu / Np) O 2 .
• rectangles referenced 1 to 13 represent multi-stage extractors such as those conventionally used in the treatment of irradiated nuclear fuels (mixer-settlers, pulsed columns, centrifugal extractors).
• FIG. 1 represents a schematic diagram of a first embodiment of the method of the invention, designed to obtain a mixed oxide of uranium and plutonium, which can be used directly for manufacturing a MOX nuclear fuel, from a dissolution liquor previously obtained by dissolving spent nuclear fuel, for example UO 2 , in nitric acid and clarifying the resulting mixture.
• Such a solution liquor typically contains from 200 to 300 g / l of uranium for 2 to 3 g / l of plutonium. It also contains neptunium, americium, curium and fission products. Its acidity is generally of the order of 3 M.
• the process according to the invention has, for a first step, a step which aims at decontaminating uranium, plutonium and neptunium with respect to actinides (III), ie americium and curium, and vis-à-vis most of the fission products.
• this decontamination step comprises:
• washing PF an operation, called "washing PF", which aims to remove from the solvent phase the fission products, in particular ruthenium and zirconium, having been extracted during the "co-extraction [// Pu / Np", by contacting the solvent phase resulting from this co-extraction with a nitric aqueous phase of moderate acidity, for example a 1 to 3 M nitric acid solution;
• washing Tc an operation, called "washing Tc", which aims to remove from the solvent phase technetium that has been extracted during the "co-extraction U / Pu / Np" by contacting the solvent phase resulting from “washing PF "with a nitric aqueous phase of moderate acidity but higher than that of the nitric aqueous phase used for the "washing PF", for example a solution of nitric acid 3 to 5 M; and
• This last solvent phase is directed to a series of extractors (5-8) in which is carried out the second stage of the process, namely the partitioning of uranium, plutonium and neptunium in two aqueous phases.
• this partition is made in the same way as in the PUREX process implemented in the UP2-800 factory in La Hague.
• Pu desextraction which aims to extract the plutonium from the solvent phase resulting from the "Tc wash", by bringing this phase into contact with a nitric aqueous phase of low acidity, for example an acid solution, 0.05 to 2 M nitric acid containing a reducing agent which reduces Pu (IV) to Pu (III) and Np (VI) to Np (IV) (the latter being extractable by TBP) without reducing the uranium, and an antinitrile agent whose role is to stabilize the reducing agent, Pu (III) and Np (IV) by destroying the nitrous acid which tends to form, this reducing agent is, for example, Uranium (IV) while the antinitrous agent is, for example, hydrazinium nitrate;
• dam Pu an operation, called" dam Pu ", which aims to complete the plutonium desextraction by contacting the solvent phase resulting from the" Pu desextraction "with a nitric aqueous phase of low acidity, for example an acid solution 0.05 to 1 M nitric acid, containing the same reducing agent and the same antinitrous agent as those used for "Pu desextraction”;
• an operation called” U / Np desextraction ", which aims to extract the solvent phase from the" Pu dam "uranium and neptunium, by contacting this phase with an aqueous phase, for example a solution of nitric acid of molarity not exceeding 0.05 M, and "an operation, called” washing U / Np ", which aims to remove from the aqueous phase resulting from the" Pu desextraction "fractions of uranium and neptunium that followed the plutonium during the" Pu desextraction ", by setting in contact with this phase with a solvent phase consisting of 30% (v / v) TBP in TPH.
• the oxidation operation is, for example, carried out by circulating this phase under a stream of nitrogen oxides NO x so as to destroy the antinitrous agent it contains - which allows the nitrous acid to reform and re-oxidize the plutonium (III) to plutonium (IV) - and then remove the excess nitrous acid by decomposition of this acid to NO and NO 2 and degassing nitrogen oxides thus formed.
• the concentration operation is, for example, carried out by evaporation, preferably until an aqueous phase containing from 200 to 250 g / l of plutonium is obtained.
• the storage operation it is, for example, carried out in tubular tubs, for a period of at least fifteen days and up to twelve months, which ensures a functional decoupling between the workshops dealing with the treatment of used nuclear fuels in charge of upstream operations (shearing of fuel rods, dissolution of fuels, clarification of solutions, decontamination and partitioning) and the workshops in charge of operations downstream (purification, co -conversion, manufacture of MOX fuel).
• the aqueous phase is directed to a series of extractors (9-12) in which is carried out the fourth stage of the process, namely the purification of plutonium vis-à-vis the traces of fission products still present in this phase.
• This purification comprises:
• Pu extraction an operation, called “Pu extraction”, which aims to extract the plutonium, in the IV oxidation state, from the aqueous phase resulting from storage, by contacting this phase with a solvent phase consisting of % (v / v) of TBP in TPH;
• Pu desextraction which aims to extract the plutonium from the solvent phase resulting from the "PF wash", by bringing this phase into contact with a nitric aqueous phase of low acidity, for example an acid solution; 0.05 to 2 M nitric acid containing a reducing agent for reducing Pu (IV) to Pu (III), for example hydroxylammonium nitrate (or NHA), stabilized with a hydrazinium nitrate antititrous agent; , and which comprises adding uranium (IV) to said aqueous phase, preferably just before it leaves the extractor 11, in an amount such that it may have, at its output from the extractor, a mass ratio U (IV) / Pu (III) in accordance with that of the mixed oxide (U, Pu) O 2 that it is desired to manufacture, for example 50/50 or substantially equal to 50/50.
• a mass ratio U (IV) / Pu (III) in accordance with that of the mixed oxide (U, Pu) O 2 that
• this co-conversion is preferably carried out by the method described in reference [2] above, that is to say by co-precipitation, by means of oxalic acid or one of its salts or one of its derivatives, uranium (IV) and plutonium (III), previously stabilized by a monocharged cation consisting solely of atoms selected from oxygen, carbon, nitrogen atoms and hydrogen, such as the hydrazinium cation, or a compound, such as a salt, capable of forming such a cation, and then calcining the resulting co-precipitate, preferably under an inert gas or very slightly oxidant, for example a gas comprising predominantly argon.
• the mixed oxide (U / Pu) O 2 thus obtained which is in the form of a powder, can then be used directly for the manufacture of pellets of an MOX nuclear fuel, for example by a method of the MIMAS type. in which case this powder is sieved, mixed with uranium oxide and possibly scrap pellets in the form of chamotte, then the resulting mixture is subjected to pelletizing, then sintering.
• FIG. 2 represents a block diagram of a second embodiment of the method of the invention, designed as the preceding one to obtain a mixed oxide of uranium and plutonium, directly usable for the manufacture of a nuclear fuel MOX, from a dissolution liquor of a spent nuclear fuel, for example of UO 2 .
• an operation, called” Pu / U desextraction " which aims to extract plutonium and a fraction of uranium from the solvent phase resulting from "Washing Tc", by bringing this phase into contact with a nitric aqueous phase of low acidity, for example a 0.05 to 2 M nitric acid solution, containing a reducing agent which reduces Pu (IV) to Pu ( III) and Np (VI) to Np (IV) without reducing uranium (VI), for example uranium (IV) stabilized by an antinitrous agent, for example hydrazinium nitrate;
• a nitric aqueous phase of low acidity for example a 0.05 to 2 M nitric acid solution
• a reducing agent which reduces Pu (IV) to Pu ( III) and Np (VI) to Np (IV) without reducing uranium (VI), for example uranium (IV) stabilized by an antinitrous agent, for example hydrazinium
• dam Pu which aims to complete the plutonium de-extraction, by contacting the solvent phase resulting from the" Pu / U desextraction "with a nitric aqueous phase of low acidity, for example a solution of d 0.05 to 1 M nitric acid, containing the same reducing agent and the same antinitrile agent as those used for "Pu / U de-extraction”;
• an operation, called” washing Np which aims to remove from the aqueous phase resulting from the" Pu / U desextraction "the neptunium fraction that followed the plutonium and uranium during this desextraction, by putting in contact with this phase with a solvent phase containing 30% (v / v) of TBP in TPH, and which comprises the addition of uranium (IV) or uranium (VI) to said aqueous phase, preferably just before it leaves the extractor 11, to compensate for the fraction of uranium likely to have followed neptunium in the solvent phase during this washing.
• the aqueous phase resulting from the "U / Np desextraction" which contains 99 to 99.9% of the uranium and 70 to 80% of the neptunium initially present in the dissolution liquor and which is directed to a series of extractors (not shown in Figure 1) in which the uranium and neptunium will be separated from each other and the uranium will be purified vis-à-vis the fission products; and
• This last aqueous phase is then sent to a unit where it is successively subjected to an oxidation operation to reduce the plutonium (III) and the uranium (IV) respectively to the oxidation state IV and VI, to a concentration operation and an operation storage, which are, for example, carried out as previously described.
• the aqueous phase resulting from the storage is directed towards a series of extractors (9-13) in which the step of purification of plutonium and uranium is carried out with respect to the traces of fission products still present. in this phase.
• this purification comprises:
• washing PF an operation, called "washing PF", which aims to remove the solvent phase resulting from the "co-extraction Pu / U" fission products that were extracted during this extraction and which is carried out as in the first mode implementation previously described;
• Pu desextraction which aims to extract the plutonium from the solvent phase resulting from the "PF wash", by bringing this phase into contact with a nitric aqueous phase of low acidity, for example an acid solution; 0.05 to 2 M nitric acid containing a reducing agent for reducing Pu (IV) to Pu (III) without reducing uranium, for example NHA, stabilized with a hydrazinium nitrate antititrous agent; and • an operation, called “washing U”, which aims to remove from the aqueous phase resulting from the "Pu desextraction” the uranium (VI) which followed the plutonium (III) during this desextraction, by bringing into contact this phase with a solvent phase consisting of 30% (v / v) of TBP in TPH, and which comprises the addition of uranium (IV) to said aqueous phase, preferably just before it leaves the extractor 13 , in an amount such that it can present,
• FIG. 3 diagrammatically illustrates a third mode of implementation of the method of the invention which, unlike the two previous ones, is designed to obtain a mixed oxide of uranium, plutonium and neptunium.
• This mode of implementation differs from the second embodiment previously described in that the partition is performed as in the implementation mode of the COEX TM process which is illustrated in FIG. 5 of reference [1] and in FIG. that the aqueous phase which is subjected to the subsequent stages of storage, purification and co-conversion contains not only plutonium and uranium but also neptunium.
• The" Pu / U desextraction "operation of the second embodiment described above is replaced by an operation called" Pu / U / Np desextraction ", which aims to extract plutonium, neptunium and a fraction of the uranium of the solvent phase resulting from the "Tc wash", by bringing this phase into contact with a nitric aqueous phase of low acidity, for example a 0.05 to 2 M nitric acid solution, containing a reducing agent which reduces the Pu (IV) in Pu (III) and Np (VI) in Np (V) (the latter being very little extractable by TBP) without reducing uranium (VI), for example NHA stabilized with an antinitrile agent, for example hydrazinium nitrate; the "wash Np" operation of the second embodiment described above is eliminated;
• the aqueous phase resulting from the "Pu / U / Np desextraction" contains plutonium (III), uranium (VI), neptunium (V) and trace fission products.
• This phase is then subjected to the oxidation, concentration and storage operations as previously described and then directed to a series of extractors (9-13) in which the purification step is carried out.
• This step is carried out in the same way as the purification step of the second embodiment previously described, except that, since the aqueous phase resulting from the storage contains neptunium:
• the "Pu stripping" operation is replaced by an operation called "Pu / Np stripping", which aims to extract the plutonium, in the oxidation state IV, and the neptunium, in the oxidation state III, from the solvent phase resulting from the "washing PF", by contacting this phase with a nitric aqueous phase of low acidity, for example a 0.05 to 2 M nitric acid solution, containing a reducing agent which makes it possible to reduce the Pu (IV) in Pu (III) and Np (VI) in Np (V) without reducing uranium, for example NHA, stabilized with a antinitrous agent of the hydrazinium nitrate type; while
• the amount of uranium (IV) added during the "washing U” (which has the effect of reducing Np (V) to Np (IV)) is adjusted to obtain a mass ratio U (IV) / Pu (III) / Np (IV) conforming to that of the mixed oxide (Pu, U, Np) O 2 which it is desired to manufacture, for example 49/49/2 or substantially equal to 49/49/2 .

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