WO2017017193A1 - Nouveaux n,n-dialkylamides dissymétriques, leur synthèse et leurs utilisations - Google Patents
Nouveaux n,n-dialkylamides dissymétriques, leur synthèse et leurs utilisations Download PDFInfo
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- C—CHEMISTRY; METALLURGY
- C07—ORGANIC CHEMISTRY
- C07C—ACYCLIC OR CARBOCYCLIC COMPOUNDS
- C07C231/00—Preparation of carboxylic acid amides
- C07C231/02—Preparation of carboxylic acid amides from carboxylic acids or from esters, anhydrides, or halides thereof by reaction with ammonia or amines
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- C—CHEMISTRY; METALLURGY
- C07—ORGANIC CHEMISTRY
- C07C—ACYCLIC OR CARBOCYCLIC COMPOUNDS
- C07C233/00—Carboxylic acid amides
- C07C233/01—Carboxylic acid amides having carbon atoms of carboxamide groups bound to hydrogen atoms or to acyclic carbon atoms
- C07C233/02—Carboxylic acid amides having carbon atoms of carboxamide groups bound to hydrogen atoms or to acyclic carbon atoms having nitrogen atoms of carboxamide groups bound to hydrogen atoms or to carbon atoms of unsubstituted hydrocarbon radicals
- C07C233/04—Carboxylic acid amides having carbon atoms of carboxamide groups bound to hydrogen atoms or to acyclic carbon atoms having nitrogen atoms of carboxamide groups bound to hydrogen atoms or to carbon atoms of unsubstituted hydrocarbon radicals with carbon atoms of carboxamide groups bound to acyclic carbon atoms of an acyclic saturated carbon skeleton
- C07C233/05—Carboxylic acid amides having carbon atoms of carboxamide groups bound to hydrogen atoms or to acyclic carbon atoms having nitrogen atoms of carboxamide groups bound to hydrogen atoms or to carbon atoms of unsubstituted hydrocarbon radicals with carbon atoms of carboxamide groups bound to acyclic carbon atoms of an acyclic saturated carbon skeleton having the nitrogen atoms of the carboxamide groups bound to hydrogen atoms or to acyclic carbon atoms
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- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B3/00—Extraction of metal compounds from ores or concentrates by wet processes
- C22B3/20—Treatment or purification of solutions, e.g. obtained by leaching
- C22B3/26—Treatment or purification of solutions, e.g. obtained by leaching by liquid-liquid extraction using organic compounds
- C22B3/32—Carboxylic acids
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0226—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors
- C22B60/0239—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors nitric acid containing ion as active agent
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0247—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using basic solutions or liquors
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0252—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries
- C22B60/026—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries liquid-liquid extraction with or without dissolution in organic solvents
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- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/04—Obtaining plutonium
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- 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
- Y02P—CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
- Y02P10/00—Technologies related to metal processing
- Y02P10/20—Recycling
Definitions
- the invention relates to novel dissymmetrical ⁇ /, V-dialkylamides as well as to a process for synthesizing them.
- It also relates to a method of treating an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, which makes it possible to extract, separate and decontaminate the uranium and plutonium present in this solution in a single cycle and without resorting to any plutonium reduction operation, and wherein one of these A / V-dialkylamides or a mixture thereof is used as an extractant.
- the invention finds particular application in the treatment of spent nuclear fuel based on uranium (in particular uranium oxides - UOX) or uranium and plutonium (in particular mixed oxides of uranium and plutonium - MOX ).
- uranium in particular uranium oxides - UOX
- uranium and plutonium in particular mixed oxides of uranium and plutonium - MOX
- the PUREX process which is implemented in all the used nuclear fuel treatment plants existing in the world (The Hague in France, Rokkasho in Japan, Sellafield in the UK, etc.), uses tri-n-butyl phosphate (or TBP) as extractant, to recover uranium and plutonium, by liquid-liquid extraction, from aqueous solutions from the dissolution of these fuels in nitric acid.
- TBP tri-n-butyl phosphate
- the TBP is used in 30% (v / v) solution in an organic diluent (hydrogenated tetrapropylene (or TPH) or ⁇ -dodecane).
- organic diluent hydrogenated tetrapropylene (or TPH) or ⁇ -dodecane.
- TPH hydrogenated tetrapropylene
- ⁇ -dodecane This organic solution is commonly called “solvent" in the field under consideration.
- LCUPu aims to decontaminate uranium and plutonium with respect to americium, curium and fission products with a partition of uranium and plutonium into two aqueous streams as soon as first cycle, by reductive desmtraction of plutonium;
- 2CU second uranium purification cycle
- a third plutonium purification cycle (respectively called “2CPu” and “3CPu”), aimed at perfecting the decontamination of plutonium to meet the specifications defined by the ASTM standards for plutonium, the finished product, and to concentrate it before conversion to oxide.
- DBP di-n-butyl phosphate
- the de-extraction of uranium from the solvent in which it was previously extracted is incomplete if it is carried out at room temperature, hence the need to carry out this desextraction at a temperature of 50 ° C. (which corresponds to the temperature maximum permitted by the flash point of the solvent); however, even at this temperature, uranium de-extraction is diluent (the organic / aqueous flow ratio (O / A) being less than 1); the solubility of the TBP, which is not negligible in the aqueous phase (up to 300 mg / L according to the acidity of the aqueous phase), requires the implementation of washing with the organic diluent of the aqueous phases resulting from the different cycles of extraction to recover solubilized TBP in these aqueous phases; and
- the inventors have set themselves the goal of developing a process that , while being as efficient as the PUREX process in terms of recovery and decontamination of uranium and plutonium present in aqueous nitric solutions for the dissolution of spent nuclear fuels, makes it possible to overcome all the limits related to use of the TBP as extractant, and in particular includes only one treatment cycle and is free of any operation of reducing reductive plutonium extraction.
- the inventors have therefore first of all sought to find extractants that have the properties required to make possible the development of such a process.
- N, N-dialkylamides represent a family of extractants which has been widely studied as a possible alternative to TBP in the treatment of spent nuclear fuels, in particular because they generally have a good affinity for the Uranium and plutonium with high acidity, are less soluble than TBP in the aqueous phase, are completely incinerable (CHON principle) and have degradation products less troublesome than those of TBP.
- N, N-dialkylamides extract plutonium less well from a strongly acidic aqueous phase than does the TBP.
- TBP strongly acidic aqueous phase
- the invention thus proposes, in the first place, new ⁇ /, V-dialkylamides which are dissymmetrical and which correspond to formula (I) below:
- R represents a linear or branched alkyl group comprising from 8 to 15 carbon atoms.
- linear alkyl group comprising 8 to 15 carbon atoms means any alkyl group chosen from the ⁇ -octyl, n-nonyl, ⁇ -decyl, n-undecyl, ⁇ -dodecyl or n-tridecyl groups, ⁇ tetradecyl and ⁇ -pentadecyl, while
- branched alkyl group comprising 8 to 15 carbon atoms means any alkyl group comprising 8, 9, 10, 11, 12, 13, 14 or 15 carbon atoms and having one or more branches, identical or different , such as n-heptyl, ⁇ -octyl, n-nonyl, ⁇ -decyl, n-undecyl, ⁇ -dodecyl, n-tridecyl or ⁇ -tetradecyl substituted by a methyl group (such as, for example, a 2- or 4-methylheptyl, 2- or 4-methyloctyl, 2- or 4-methyloctyl, etc.); a ⁇ -hexyl, n-heptyl, ⁇ -octyl, n-nonyl, ⁇ -decyl, n-undecyl, ⁇ -dodecyl or n-tridecyl group substituted with an ethyl group (
- aqueous solution and “aqueous phase” are equivalent and interchangeable, just as the terms “organic solution” and “organic phase” are equivalent and interchangeable.
- the linear or branched alkyl group, represented by R in formula (I) above does not comprise more than 12 carbon atoms, and this, for reasons of viscosity (viscosity A /, / V-dialkylamides increasing, in fact, with the number of carbon atoms that R).
- this group is selected from the groups ⁇ -octyl, n-decyl, ⁇ -dodecyl, 2-ethylhexyl and 2-ethyloctyl, the ⁇ -octyl group being very particularly preferred.
- N, N-dialkylamides defined above are advantageously obtained by reacting a halide of formula (II) below:
- X represents a halogen atom and, preferably, a chlorine atom, with an amine of formula HN (CH 3) R in which R represents a linear or branched alkyl group comprising from 8 to 15 carbon atoms, in presence of a base.
- the subject of the invention is also a process for the synthesis of ⁇ 1 / ⁇ -dialkylamides, which comprises this reaction.
- Said reaction can be carried out either in aqueous solution, in which case the base is, for example, sodium hydroxide or potassium hydroxide, or in an organic solvent such as dichloromethane or diethyl ether, in which case the base is, for example, triethylamine or diisopropylethylamine.
- the N, N-dialkylamides defined above have been found to be capable of very efficiently extracting uranium (VI) and plutonium (IV) from an acidic aqueous solution such as an aqueous nitric solution.
- the subject of the invention is still the use of a ⁇ -, V-dialkylamide or a mixture of ⁇ /, V-dialkylamides as defined above, in order to extract the uranium (VI) and / or plutonium (IV) of an acidic aqueous solution.
- the uranium and / or the plutonium are preferably extracted from the acidic aqueous solution by liquid-liquid extraction, that is to say by bringing this aqueous solution into contact with an organic solution.
- an organic solution comprising the N, N-dialkylamide or the mixture of N, N-dialkylamides in an organic diluent, and then separating the aqueous and organic solutions.
- the organic solution preferably comprises from 1 mol / L to 2 mol / L and, more preferably, from 1.3 mol / L to 1.5 mol / L of ⁇ -N-dialkylamide or mixture of N, N-dialkylamides.
- the acidic aqueous solution is, preferably, an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, that is to say an aqueous solution typically comprising from 3 mol / L to 6 mol / L nitric acid.
- the ⁇ -N-dialkylamides defined above have been found to subsequently make it possible to separate one on the other, the uranium and plutonium thus extracted, without the reduction of plutonium, this separation being able to be:
- uranium and plutonium that is to say in which are obtained two aqueous solutions comprising one a mixture of plutonium and uranium and the other uranium without plutonium.
- the invention also relates to the use of a
- a /, / V-dialkylamide or a mixture of N, N-dialkylamides as previously defined, to totally or partially separate uranium (VI) from plutonium (IV) from an acidic aqueous solution, which use comprises:
- a plutonium dextraction at the oxidation state + IV, of the organic solution resulting from step a), this de-extraction comprising at least one bringing the organic solution into contact with an aqueous solution comprising 0.1 mol / L at 0.5 mol / L of nitric acid, then a separation of the organic and aqueous solutions; and c) an extraction of all or part of the uranium fraction present in the aqueous solution resulting from stage b), this extraction comprising at least one bringing the aqueous solution into contact with an organic solution identical to the organic solution used in step a), followed by separation of the aqueous and organic solutions;
- an aqueous solution comprising plutonium without uranium or a mixture of plutonium and uranium, and an organic solution comprising uranium without plutonium.
- the organic solution used in step a) and, consequently, that used in step c) preferably comprise from 1 mol / l to 2 mol / l and more preferably from 1.3 mol / l to 1.5 mol / L of ⁇ - / V-dialkylamide or mixture /, / V-dialkylamide mixture.
- the acidic aqueous solution from which uranium and plutonium are co-extracted is preferably an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid. that is to say an aqueous solution typically comprising from 3 mol / l to 6 mol / l of nitric acid.
- the uranium present in the organic solution resulting from stage c) can then be extracted from this phase by contacting the organic solution with an aqueous solution comprising at most 0.05 mol / L of nitric acid, and then separation organic and aqueous solutions.
- an aqueous solution comprising at most 0.05 mol / L of nitric acid, and then separation organic and aqueous solutions.
- the ⁇ /, / V-dialkylamides defined above have been found to make it possible to extract uranium (VI) and plutonium (IV) from an aqueous solution resulting from the dissolution of nuclear fuels. used in nitric acid with very high separation factors vis-à-vis the main fission products present in this solution.
- these A /, / V-dialkylamides have made it possible to develop a process for treating an aqueous nitric solution for dissolving a spent nuclear fuel which, while being as efficient as the PUREX process In terms of recovery and decontamination of the uranium and plutonium present in such a solution, it is free of any reductive plutonium de-extraction operation and has only one treatment cycle.
- the subject of the invention is furthermore a method of treating in one cycle an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, the aqueous solution comprising uranium, plutonium, americium, curium and fission products including technetium, the cycle comprising:
- step b) a decontamination of the organic solution resulting from step a) with respect to americium, curium and fission products, this decontamination comprising at least one contacting, in an extractor, of the solution organic with an aqueous solution comprising from 0.5 mol / L to 6 mol / L of nitric acid, then a separation of the organic and aqueous solutions;
- step b) a partition of the uranium and plutonium present in the organic solution resulting from step b) into an aqueous solution comprising either plutonium without uranium or a mixture of plutonium and uranium, and an organic solution comprising uranium without plutonium, this partition comprising: ci) a plutonium dextraction, at the oxidation + IV degree, and a uranium fraction of the organic solution resulting from step b), this de-extraction comprising at least one contacting, in an extractor, the organic solution with an aqueous solution comprising from 0.1 mol / L to 0.5 mol / L of nitric acid, and then a separation of the organic and aqueous solutions;
- step di an extraction of the uranium fraction present in the aqueous solution resulting from step di), this extraction comprising at least one bringing into contact, in an extractor, of the aqueous solution with an organic solution identical to the solution organic used in step a), and then a separation of aqueous and organic solutions;
- step e a removal of the uranium from the organic solution resulting from step di), said de-extraction comprising at least one contacting, in an extractor, of the organic solution with an aqueous solution comprising at most 0.05 mol / L nitric acid, then a separation of organic and aqueous solutions; and f) a regeneration of the organic phase resulting from step e);
- a first and a second decontaminated aqueous solution is obtained with respect to americium, curium and fission products including the technetium, the first aqueous solution comprising plutonium without uranium or a mixture of plutonium and uranium, and the second aqueous solution comprising uranium without plutonium.
- the organic solution used in step a) and hence those used in steps c 2 ) and d 2 ) since the organic solutions used in steps a), c 2 ) and d 2 ) have the same composition preferably comprise from 1 mol / L to 2 mol / L and, more preferably, from 1.3 mol / L to 1.5 mol / L of ⁇ /, -V-dialkylamide or a mixture of ⁇ /, / V-dialkylamides.
- the aqueous solution used in step b) may comprise from 0.5 mol / l to 6 mol / l of nitric acid.
- this aqueous solution comprises from 4 mol / L to 6 mol / L of nitric acid so as to facilitate the extraction of ruthenium and technetium from the organic solution resulting from step a).
- step b) advantageously further comprises a deacidification of the organic solution, this deacidification comprising at least one bringing the organic solution into contact with an aqueous solution comprising from 0.1 mol / l to 1 mol / l and better still, 0.5 mol / L of nitric acid, followed by separation of the organic and aqueous solutions.
- the bringing into contact of the organic and aqueous solutions in the extractor in which stage ci) takes place comprises a circulation of these solutions in a ratio of flow rates O / A which is advantageously greater than 1, of preferably equal to or greater than 3 and, more preferably, equal to or greater than 5 so as to obtain a removal of the concentrating plutonium, that is to say a plutonium desextraction which leads to an aqueous solution in which the plutonium concentration is greater than that presented by this element in the organic solution from which it is extracted.
- the reducing agent (s) present in the aqueous solution used in step di) is (are) preferably chosen from uranose nitrate (also called “U (IV) ) "), Hydrazinium nitrate (also known as” hydrazine nitrate “), hydroxylammonium nitrate (also known as” hydroxylamine nitrate “), acetaldoxime and mixtures thereof such as a mixture of uranous nitrate and hydrazinium nitrate, a mixture uranous nitrate and hydroxylammonium nitrate or a mixture of uranous nitrate and acetaldoxime, preferably given to a mixture of uranous nitrate and hydrazinium nitrate or a mixture of uranous nitrate and hydroxylammonium nitrate than the it is preferentially used at a concentration ranging from 0.1 mol / L to 0.3 mol / L and, typically, from 0.2
- step di which can be carried out at room temperature, is, however, preferably carried out at a temperature ranging from 30 to 40 ° C and, more preferably, from 32 ° C so as to promote the kinetics of de-extraction. technetium while limiting at best the reoxidation phenomena of this element in aqueous phase.
- step d 2 preferably comprises, in addition, an acidification of the aqueous solution resulting from step di), this acidification comprising an addition of nitric acid in the extractor in which takes place step d 2 ) to bring the concentration of nitric acid in the aqueous solution to a value at least equal to 2.5 mol / L.
- Step e) can be performed at room temperature. However, it is preferably carried out at a temperature ranging from 40 ° C to 50 ° C to, again, promote the de-extraction of uranium.
- the extractor in which step e) is carried out is therefore preferably heated to a temperature between 40 ° C and 50 ° C.
- the contacting of the organic and aqueous solutions in the extractor in which this step takes place comprises a circulation of these solutions in a ratio of flow rates O / A greater than 1 so as to obtain a removal of the concentrating uranium, that is to say a uranium extraction which leads to an aqueous solution in which the concentration of uranium is greater than that which this element presents in the organic solution from which it is extracted.
- the process of the invention also comprises a step f) of regeneration of the organic solution resulting from stage e), this regeneration preferably comprising at least one washing of the organic solution with a basic aqueous solution, followed by at least one washing of the organic solution with an aqueous solution of nitric acid.
- the de-extraction of uranium is easier to implement than that of the PUREX process since it can be carried out at room temperature as well as in hot conditions and by using an O / A flow ratio greater than 1, which allows uranium to be extracted in a concentrated manner, which is not possible in the PUREX process;
- plutonium removal is also easier to implement than that of the plutonium.
- PUREX process and can be carried out more concentrically than the latter; these advantages are all the more important as the future spent nuclear fuel treatment plants will have to process plutonium-rich fuels (such as MOX fuels from light water or fast neutron reactors) than the fuels currently treated;
- N, N-dialkylamides are less troublesome than those of TBP because they are soluble in water and do not form complexes capable of retaining plutonium;
- the ⁇ -, V-dialkylamides typically have an aqueous phase solubility that is 100 to 200 times lower than that of the TBP, which makes it possible to envisage the suppression or, at the very least, a reduction in the washing with the organic diluent of aqueous solutions resulting from the process of the invention compared to those provided in the PUREX process;
- FIG. 1 represents, in the form of two lines, the variation of the logarithm of the distribution coefficients, denoted DM, of uranium on the one hand, and plutonium on the other hand, as obtained in tests of extraction carried out with an A / V-dialkylamide of the invention, as a function of the logarithm of the free concentration (in mol / L) of this free ⁇ /, -V-dialkylamide in the organic phase used in these tests. extraction.
- DM logarithm of the distribution coefficients
- FIG. 2 represents a schematic diagram of the process for treating an aqueous nitric solution for dissolving a spent nuclear fuel of the invention
- the rectangles 1 to 7 represent multi-stage extractors such as those conventionally used in the treatment of spent nuclear fuels (mixer-settlers, pulsed columns or centrifugal extractors); the organic phases are symbolized by solid lines whereas the aqueous phases are symbolized by dashed lines.
- R linear or branched C8 to C15 alkyl group.
- MOEHA which corresponds to the formula (I) above, in which R represents a ⁇ -octyl group, is synthesized from 2-ethylhexanoyl chloride and N-methyl-N-octylamine, in the presence of sodium hydroxide in water (reaction scheme A).
- sodium hydroxide (30% NaOH - 112 g - 0.839 mol - 1.19 eq.), Water (100 g) and N-methyl-N-octylamine (100 g. 0.698 mole - 1 eq.)
- the system is stirred, the set point is set at 4 ° C.
- the 2-ethylhexanoyl chloride 136.5 g - 0.839 mol - 1.19 eq.
- the progress of the reaction is monitored and shows the presence of 0.6% residual amine.
- the medium is heated at 50 ° C. for 30 minutes to consume the residual amine.
- the medium is then cooled to 20 ° C. and then decanted.
- the organic phase is washed twice with 100 ml of water to obtain 208 g of crude product.
- the MOEHA is then obtained at a purity of 98.3% (measured by gas chromatography coupled to a flame ionization detector or GC-FID) after two distillations under pressure (respectively at 2 and 8 mbar).
- the DEHA M which corresponds to formula (I) above, in which R represents a ⁇ -decyl group, is synthesized according to reaction scheme (a) from 2-ethylhexanoyl chloride and A / -decyl. - / V-methylamine, in the presence of triethylamine (Et 3 N) in anhydrous dichloromethane (DCM) (reaction scheme A).
- the M DdEHA which corresponds to formula (I) above, in which R represents a ⁇ -dodecyl group, is synthesized from 2-ethylhexanoyl chloride and N-dodecyl- ⁇ -methylamine, presence of Et 3 N in anhydrous DCM (reaction scheme A).
- the M DdEHA is then obtained at a purity of 99.5% (measured by GC-FI D) after a single very low pressure distillation (0.7 mbar).
- EHA which corresponds to formula (I) above in which R represents a 2-ethylhexyl group, is synthesized from 2-ethylhexanoyl chloride and N-methyl-hexanoyl.
- 2-ethylhexanamine in the presence of EtsN in anhydrous DCM (Reaction Scheme A), wherein N-methyl-2-ethylhexanamine was synthesized from 2-ethylhexylamine (Reaction Scheme B).
- the crude oil obtained is purified by chromatography on silica gel (elution: DCM 100%) to give, after concentration under controlled vacuum (0.035 mbar / 40 ° C), the compound 3 of the reaction scheme B in which R is a group 2-ethylhexyl (21 g) as a colorless oil.
- the crude oil obtained is purified by chromatography on silica gel (elution: DCM 100%) to give a 98.5% pure fraction (HPLC at 210 nm) of M (2-EH) EHA (2.6 g) in the form of a pale yellow oil. With the impure fractions recovered, the yield of synthesis from 2-ethylhexanamine is estimated at 71%.
- the crude oil obtained is purified by chromatography on silica (elution: cyclohexane / DCM) to give 2-ethyloctane-nitrile (solution of 24 g at 60% by mass with DCM / cyclohexane, ie 9.8 g estimated) in the form a pale yellow solution.
- the crude oil obtained is purified by chromatography on silica gel (elution: DCM 100%) to give a pure fraction at 98.3% (H PLC at 210 nm) of M (2-EO) EHA (2.6 g ) in the form of a pale yellow oil. With the impure fractions recovered, the yield of the step is estimated at 64%.
- solutions comprising 1.4 mol / l of MOEHA, M DEHA or M DdEHA in TPH; and as aqueous phases: aliquots of an aqueous solution comprising 90 g / l of uranium (VI), approximately 70 mg / l of plutonium (IV) and 4.15 mol / l of HN0 3 .
- organic phases the organic phases obtained after the extraction tests above.
- aqueous phases aliquots of an aqueous solution comprising 0.1 mol / l of HN0 3 .
- Each of these tests is carried out by contacting, in tube and with stirring, an organic phase with an aliquot of aqueous solution for 15 minutes at 25 ° C.
- the volume ratio O / A used is 1 for the extraction tests and 1 for the desextraction tests. Then, these phases are separated from each other after centrifugation.
- the concentrations of uranium and plutonium are measured in the organic and aqueous phases thus separated, by X-ray fluorescence for uranium and by a spectrometry for plutonium.
- Table I shows, for each A /, / V-dialkylamide tested, the concentrations of uranium, denoted [U] or g, as obtained in the organic phases at the end of the tests. extraction, the uranium distribution coefficients, denoted Du, and plutonium, denoted Dp u , as obtained at the end of the extraction and desextraction tests, the concentrations of nitric acid, denoted [HN03 ] aq ., as obtained in the aqueous phases after the extraction and desextraction tests, as well as the U / Pu separation factors, denoted FSu / p u , as obtained at the end of desextraction tests.
- solutions comprising respectively 0.1 mol / L, 0.5 mol / L, 0.75 mol / L, 1.0 mol / L, 1.25 mol / L, 1.5 mol / L and 2 mol / L MOEHA in TPH; and
- aqueous phases aliquots of an aqueous solution comprising 2 g / l of uranium (VI), 1 mol / l of HNO3 and 2 mol / l of UNO3, and aliquots of an aqueous solution comprising 1.7.10 4 mol / L of plutonium (IV), 1 mol / L of HNOs and 2 mol / L of LiNO 3 .
- each organic phase is brought into contact, in a tube and with stirring, with an aliquot of aqueous solution for 15 minutes at 25 ° C, in a volume ratio O / A of 1. Then, these phases are separated. one after the other after centrifugation.
- Uranium concentrations are measured in the aqueous phases by plasma flash atomic emission spectrometry (or ICP-AES) while uranium concentrations in the organic phases are determined by de-extruding these elements into water and measuring by ICP-AES their concentration in the aqueous phases resulting from this desextraction.
- Plutonium concentrations are measured in the aqueous and organic phases by spectrometry a.
- FIG. 1 represents, in the form of two lines, the variation of the logarithm of the distribution coefficients, denoted DM, of uranium on the one hand, and plutonium on the other hand, as a function of logarithm of the free concentration (in mol / L) of MOEHA in organic phase (total concentration of MOEHA corrected for the fraction of nitric acid extracted in the organic phase).
- aqueous phase an aqueous solution previously obtained by dissolving in 5M nitric acid pellets from different irradiated fuels UOX-REB type (Boiling Water Reactor) and UOX-REP (Pressurized Water Reactor).
- This aqueous solution comprises 4.3 mol / l of H N03 and its composition in elements is presented in Table II below.
- the organic phase previously equilibrated with 6 mol / l of HNO 3, is contacted, in tube and with stirring, with the aqueous phase for 15 minutes at 25 ° C., in a volume ratio O / A of 2.5.
- the concentrations of Te, Np, Zr, Mo and Fe could only be measured in aqueous phase by ICP-AES and the concentrations of these elements in organic phase were estimated by difference between the initial concentrations of said elements in aqueous phase. and those measured at equilibrium after extraction.
- solutions comprising 0.5 mol / L of M (2-EH) EHA or of M (2-EO) EHA in TPH;
- the concentration of uranium and the activity of plutonium ( 239 ⁇ 240 ⁇ ) are measured in the organic and aqueous phases thus separated, respectively by ICP-AES and spectrometry a.
- Table IV shows, for each A /, / V-dialkylamide tested, the distribution coefficients of uranium, denoted Du, and plutonium, denoted Dp u , as obtained, as well as the separation factors.
- U / Pu denoted FSu / p u , as obtained for an acidity of 0.5 mol of HN0 3 / L.
- Also shown in this table are the experimental results obtained under the same operating conditions but using as organic phases, aliquots of a solution comprising 0.5 mol / l of MOEHA in TPH.
- FIG. 2 shows a block diagram of the method of treating an aqueous nitric solution for dissolving a spent nuclear fuel of the invention.
- the process comprises 8 steps.
- the first of these steps aims at jointly extracting uranium and plutonium, the first at the oxidation state + VI and the second at the degree of oxidation + IV of the nitric aqueous solution for dissolving spent nuclear fuel.
- Such a solution typically comprises from 3 to 6 mol / l of HNO 3, uranium, plutonium, minor actinides (americium, curium and neptunium), fission products (La, Ce, Pr, Nd, Sm, Eu, Gd, Mo, Zr, Ru, Te, Rh, Pd, Y, Cs, Ba, ...) as well as some corrosion products such as iron.
- the "Co-extraction U / Pu” step is carried out by circulating, in the extractor 1, the counter-current dissolution solution of an organic phase (denoted “PO" in FIG. 2) which comprises from 1 mol / L at 2 mol / L and, more preferably, from 1.3 mol / L to 1.5 mol / L of a ⁇ /, / V-dialkylamide of the invention or a mixture of ⁇ /, / V-dialkylamides of the invention, dissolved in an organic diluent.
- PO organic phase
- This organic diluent is a linear or branched aliphatic hydrocarbon, such as ⁇ -dodecane, TPH or isoparaffinic diluent which is sold by TOTAL under the trade reference Isane IP 185T, preferably being given to TPH.
- the second step of the process aims at extracting from the organic phase resulting from the "Co-extraction U / Pu" the fraction of the fission products that have been extracted from the dissolution solution, jointly with uranium and plutonium.
- the "PF wash” stage comprises one or more operations for washing the organic phase resulting from the "U / Pu co-extraction", each washing operation being carried out by circulating this organic phase, in the extractor 2, against the current of a nitric aqueous solution whose concentration can range from 0.5 mol / L to 6 mol / L of HN03 but is preferably from 4 mol / L to 6 mol / L of HN03 and more preferably from 4 to 5 mol / L of HN03 to facilitate the removal of ruthenium and technetium.
- this step further comprises a deacidification of the organic phase, which is carried out by circulating this organic phase against the current of a weakly acidic aqueous nitric solution, that is to say comprising from 0.1 mol / l to 1 mol / l of HN03 as, for example, an aqueous solution comprising 0.5 mol / L of HNO 3, in order to prevent an excessive amount of acid being entrained towards the extractor devolved to the third stage, denoted "Pu extraction" in Figure 2, and does not disturb the performance of this third step.
- a weakly acidic aqueous nitric solution that is to say comprising from 0.1 mol / l to 1 mol / l of HN03 as, for example, an aqueous solution comprising 0.5 mol / L of HNO 3, in order to prevent an excessive amount of acid being entrained towards the extractor devolved to the third stage, denoted "Pu extraction" in Figure 2, and does not disturb the
- the "Pu extraction” stage which represents the first step of the U / Pu partition, aims to extract the plutonium from the oxidation state + IV, and, therefore, without any reduction of this plutonium, the organic phase resulting from the PF wash.
- the plutonium (IV) desextraction which is carried out at the “Pu Desextraction” stage, is accompanied by a removal of a fraction of the uranium (VI) which is also present in the organic phase resulting from the "PF wash. ".
- the U / Pu partition to lead to an aqueous solution comprising plutonium without uranium, and to an organic solution comprising uranium without plutonium; - the quantity of uranium which makes it possible to obtain, at the end of the " 1st Wash U", an aqueous solution comprising uranium and plutonium in a previously chosen ratio, if it is desired that the partition U / Pu lead to an aqueous solution comprising a mixture of plutonium and uranium in this ratio and an organic solution comprising uranium without plutonium.
- the " 1st Washing U” is carried out by circulating, in the extractor 4, the aqueous phase resulting from the "Pu Desextraction” against the current of an organic phase of composition identical to that of the organic phase used for "Co-extraction U / Pu".
- the quantity of uranium extracted is adjusted by adjusting the ratio of O / A flow rates and the acidity of the aqueous phase, the uranium being, in fact, As much better extracted than the ratio of the organic phase / aqueous phase flow rates and the acidity of the aqueous phase are high.
- a more or less concentrated addition of HNO3 to the aqueous phase circulating in the extractor 4 can therefore be provided depending on the acidity that it is desired to confer on this aqueous phase.
- the fifth step aims to extract the organic phase resulting from the “Pu extraction", the fraction of technetium having been extracted during the "U / Pu co-extraction”. which was not desextracted during the "PF wash”, in order to decontaminate this organic phase with respect to technetium.
- Pro extraction means the fraction of neptunium that was extracted during the "Co-extraction U / Pu” that followed technetium up to “A-Tc dam", as well as traces of plutonium that this organic phase is likely to to contain again.
- uranose nitrate or U (IV)
- hydrazinium nitrate or NH
- hydroxylammonium nitrate or NHA
- acetaldoxime or a mixture of these may be used as reducing agents.
- ci such as a mixture U (IV) / NH, U (IV) / NHA or U (IV) / acetaldoxime, preferably being given to a mixture U (IV) / NH or U (VI) / NHA.
- Gluconic acid may be added to the aqueous solution to reduce the reoxidation phenomena of technetium in the aqueous phase and thus limit the consumption of reducing agent (s).
- This step can be carried out at room temperature (that is to say at 20-25 ° C) but it is preferably carried out at a temperature ranging from 30 ° C to 40 ° C and, better still, 32 ° C. C so as to promote the kinetics of the extraction of technetium while limiting the reoxidation phenomena of technetium in the aqueous phase and, therefore, the risk of seeing the technetium, once extracted, be re-extracted into the organic phase.
- the sixth step is designed to extract from the aqueous phase resulting from the " ⁇ -Tc barrier" the uranium that has been desextracted, together with the technetium, at the previous stage in order to avoid that the " ⁇ -Tc Dam" step results in an excessive loss of uranium in the aqueous phase.
- the seventh step aims to extract the uranium (VI) from the organic phase resulting from the "a-Tc barrier".
- nitric aqueous solution of very low acidity that is to say comprising at most 0 , 05 mol / L of HN03 as, for example for example, an aqueous solution comprising 0.01 mol / l of HNO 3.
- This step can be carried out at room temperature (that is to say at 20-25 ° C.) but it is preferably carried out hot (that is to say typically at a temperature of 40-50 ° C. ) and using an O / A flow ratio greater than 1 for the uranium (VI) to be concentrically desextract.
- two raffinates which correspond to the aqueous phases emerging respectively from the extractors 1 and 6 and which comprise, for the first, fission products as well as americium and curium ("primary raffinate” in FIG. 2) and, for the second, technetium, neptunium and possibly traces of plutonium (“secondary raffinate” in Figure 2);
- the aqueous phase leaving the extractor 4 which comprises either decontaminated plutonium or a mixture of decontaminated plutonium and uranium and which is called "Pu flux” or "Pu + U flux” as the case may be;
- the eighth step aims to regenerate this organic phase by subjecting it to one or more washings with a basic aqueous solution, for example a first wash with an aqueous solution of 0.3 mol / L of sodium carbonate, followed by a second washing with a 0.1 mol / L aqueous solution of sodium hydroxide, followed by one or more washes with an aqueous solution of nitric acid allowing reacidifying it, for example an aqueous solution comprising 2 mol / l of HNO 3, each washing being carried out by circulating said organic phase, in an extractor, against -current of the aqueous washing solution.
- the organic phase thus regenerated can then be returned to extractors 1 and 4 for reintroduction into the treatment cycle.
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GB1801315.1A GB2555362B (en) | 2015-07-29 | 2016-07-28 | Novel dissymmetric N, N-dialkylamides, the synthesis thereof and uses of same |
JP2018504199A JP6775572B2 (ja) | 2015-07-29 | 2016-07-28 | 新規な非対称n,n−ジアルキルアミド、その合成及び使用 |
CN201680044477.7A CN107922314B (zh) | 2015-07-29 | 2016-07-28 | 新型不对称n,n-二烷基酰胺及其合成方法与应用 |
US15/748,030 US10252983B2 (en) | 2015-07-29 | 2016-07-28 | Dissymmetric N,N-dialkylamides, the synthesis thereof and uses of same |
RU2018107117A RU2702739C2 (ru) | 2015-07-29 | 2016-07-28 | Новые асимметричные n,n-диалкиламиды, их синтез и применение |
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FR3063499A1 (fr) * | 2017-03-06 | 2018-09-07 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Procede de recuperation et de purification de l'uranium present dans une boue de diuranate de potassium contaminee par du plutonium, du neptunium et du technetium |
WO2019002788A1 (fr) | 2017-06-29 | 2019-01-03 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Carbamides pour la séparation de l'uranium(vi) et du plutonium(iv) sans réduction du plutonium(iv) |
CN110312702A (zh) * | 2017-01-26 | 2019-10-08 | 原子能和替代能源委员会 | 尤其用于分离铀(vi)和钚(iv)的不对称n,n-二烷基酰胺,其合成及用途 |
WO2023067273A1 (fr) | 2021-10-21 | 2023-04-27 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Monoamides à amine cyclique pour l'extraction de l'uranium(vi) et du plutonium(iv) et pour leur séparation sans réduction du plutonium(iv) |
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CN109207724B (zh) * | 2018-09-12 | 2020-12-29 | 哈尔滨工业大学(威海) | 一种从含钒铬溶液中同时萃取分离钒铬的萃取溶剂及萃取方法 |
CN110144471B (zh) * | 2019-05-15 | 2020-10-09 | 中国原子能科学研究院 | 从核燃料后处理废液中提取锝的方法 |
CN113429312B (zh) * | 2021-06-24 | 2022-12-13 | 福建省长汀金龙稀土有限公司 | 一种n,n-二烃基酰胺羧酸化合物及其制备方法和应用 |
CN114574698B (zh) * | 2022-03-29 | 2023-03-07 | 中国原子能科学研究院 | 一种乏燃料后处理铀纯化方法 |
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FR2591213B1 (fr) | 1985-12-05 | 1988-02-05 | Commissariat Energie Atomique | Procede d'extraction de l'uranium vi et/ou du plutonium iv presents dans une solution aqueuse au moyen de n,n-dialkylamides |
FR2642561A1 (fr) * | 1989-02-01 | 1990-08-03 | Commissariat Energie Atomique | Procede pour separer l'uranium vi du thorium iv presents dans une solution aqueuse au moyen d'un n, n-dialkylamide, utilisable notamment pour separer l'uranium produit par irradiation du thorium |
FR2642562B1 (fr) | 1989-02-01 | 1991-04-05 | Commissariat Energie Atomique | Procede d'extraction de l'uranium vi et/ou du plutonium iv presents dans une solution aqueuse acide au moyen d'un melange de n,n-dialkylamides, utilisable pour le retraitement de combustibles nucleaires irradies |
FR2880180B1 (fr) * | 2004-12-29 | 2007-03-02 | Cogema | Perfectionnement du procede purex et ses utilisations |
FR2901627B1 (fr) | 2006-05-24 | 2009-05-01 | Commissariat Energie Atomique | Procede de retraitement d'un combustible nucleaire use et de preparation d'un oxyde mixte d'uranium et de plutonium |
FR2947663B1 (fr) | 2009-07-02 | 2011-07-29 | Areva Nc | Procede ameliore de traitement de combustibles nucleaires uses |
FR2954354B1 (fr) * | 2009-12-22 | 2012-01-13 | Commissariat Energie Atomique | Procede de purification de l'uranium d'un concentre d'uranium naturel |
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FR3015760B1 (fr) | 2013-12-20 | 2016-01-29 | Commissariat Energie Atomique | Procede de traitement d'un combustible nucleaire use comprenant une etape de decontamination de l'uranium(vi) en au moins un actinide(iv) par complexation de cet actinide(iv) |
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CN110312702A (zh) * | 2017-01-26 | 2019-10-08 | 原子能和替代能源委员会 | 尤其用于分离铀(vi)和钚(iv)的不对称n,n-二烷基酰胺,其合成及用途 |
CN110312702B (zh) * | 2017-01-26 | 2022-10-14 | 原子能和替代能源委员会 | 尤其用于分离铀(vi)和钚(iv)的不对称n,n-二烷基酰胺,其合成及用途 |
FR3063499A1 (fr) * | 2017-03-06 | 2018-09-07 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Procede de recuperation et de purification de l'uranium present dans une boue de diuranate de potassium contaminee par du plutonium, du neptunium et du technetium |
WO2019002788A1 (fr) | 2017-06-29 | 2019-01-03 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Carbamides pour la séparation de l'uranium(vi) et du plutonium(iv) sans réduction du plutonium(iv) |
US11479833B2 (en) | 2017-06-29 | 2022-10-25 | Commissariat A L'Énergie Atomique Et Aux Énergies | Carbamides for separating uranium(VI) and plutonium(IV) without reducing the plutonium(IV) |
WO2023067273A1 (fr) | 2021-10-21 | 2023-04-27 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Monoamides à amine cyclique pour l'extraction de l'uranium(vi) et du plutonium(iv) et pour leur séparation sans réduction du plutonium(iv) |
FR3128460A1 (fr) * | 2021-10-21 | 2023-04-28 | Commissariat A L'energie Atomique Et Aux Energies Alternatives | Monoamides à amine cyclique et leurs utilisations |
GB2626474A (en) * | 2021-10-21 | 2024-07-24 | Commissariat A Lenergie Automique Et Aux Energies Alternatives | Cyclic amine monoamides for the extraction of uranium(V1)and plutonium(1V)and for the separation thereof without reduction of the plutonium(1V) |
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CN107922314B (zh) | 2020-07-07 |
CN107922314A (zh) | 2018-04-17 |
FR3039547B1 (fr) | 2017-08-25 |
US20180222849A1 (en) | 2018-08-09 |
JP2018532691A (ja) | 2018-11-08 |
JP6775572B2 (ja) | 2020-10-28 |
FR3039547A1 (fr) | 2017-02-03 |
GB2555362A (en) | 2018-04-25 |
GB201801315D0 (en) | 2018-03-14 |
RU2702739C2 (ru) | 2019-10-10 |
GB2555362B (en) | 2020-02-19 |
RU2018107117A3 (fr) | 2019-08-28 |
RU2018107117A (ru) | 2019-08-28 |
US10252983B2 (en) | 2019-04-09 |
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