法国专利FR3039547A1 NOVEL DISSYMETRIC N, N-DIALKYLAMIDES, THEIR SYNTHESIS AND USES THEREOF

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专利摘要:
The invention relates to novel asymmetric N, N-dialkylamides which have the following formula (I): wherein R represents a C8-C15 linear alkyl group. It also relates to a process for synthesizing these N, N-dialkylamides as well as their use as extractants, alone or as a mixture, for extracting uranium and / or plutonium from an acidic aqueous solution or to separate totally or partially uranium plutonium from an acidic aqueous solution and, in particular, an aqueous solution from the dissolution of a spent nuclear fuel in nitric acid. It also relates to a method for 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 N, N-dialkylamides or a mixture thereof is used as an extractant. Field of application: treatment of spent nuclear fuels, particularly those based on uranium (for example, UOX) or uranium and plutonium (for example, MOX).
公开号:FR3039547A1
申请号:FR1557264
申请日:2015-07-29
公开日:2017-02-03
发明作者:Manuel Miguirditchian；Pascal Baron；Moreira Sandra Lopes
申请人:Areva NC SA；Commissariat a lEnergie Atomique CEA；Commissariat a lEnergie Atomique et aux Energies Alternatives CEA；
IPC主号:

专利说明:

NOVEL DISSYMMETRICAL Λ /, / V-DIALKYLAMIDES, THEIR SYNTHESIS AND THEIR
USES
DESCRIPTION
TECHNICAL FIELD The invention relates to novel dissymmetrical Λ /, V-dialkylamides as well as to a process for synthesizing them.
It further relates to the use of these Λ /, / V-dialkylamides as extractants, for extracting uranium and / or plutonium from an acidic aqueous solution and, in particular, an aqueous solution resulting from the dissolution of spent nuclear fuel in nitric acid.
It also relates to the use of these A /, A / -dialkylamides as extractants, to separate, totally or partially, uranium from plutonium from an aqueous acid solution and, in particular, an aqueous solution. from the dissolution of spent nuclear fuel in nitric acid.
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, N-dialkylamides or a mixture thereof is used as an extractant. The invention finds particular application in the treatment of uranium-based nuclear fuel (in particular uranium oxide - UOX) or uranium and plutonium (in particular mixed oxides of uranium and plutonium -MOX ).
STATE OF THE PRIOR ART
The PUREX process, which is implemented in all used nuclear fuel treatment plants 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.
In this process, TBP is used in 30% (v / v) solution in an organic diluent (hydrogenated tetrapropylene (or TPH) or n-dodecane). This organic solution is commonly called "solvent" in the field under consideration.
The recovery of uranium and plutonium by the PUREX process is carried out in several cycles: - a first purification cycle of uranium and plutonium (called "lCUPu"), which aims to decontaminate uranium and plutonium against americium, curium and fission products with a partition of uranium and plutonium in two aqueous streams from this first cycle, by reductive desmtraction of plutonium; - a second uranium purification cycle (called "2CU"), which aims to perfect the decontamination of uranium to meet the specifications defined by the ASTM standards for uranium, finished product; and - a second cycle and, in the case of some mills, a third plutonium purification cycle (respectively referred to as "2CPu" and "3CPu"), which aim to complete 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.
The performance of the PUREX process is satisfactory and the experience gained since the start of the plants that implement it is positive. However, there are limits to the use of TBP which preclude the possibility of achieving with this extractant the objectives of simplicity, compactness and improved safety established for the future used nuclear fuel treatment plants, aiming in particular to achieve partition of uranium and plutonium into two aqueous streams without the use of reducing agents.
These limits are as follows: * the decontamination factors of uranium and plutonium with respect to certain fission products (technetium and ruthenium) and transuranics (Np) are insufficient at the end of the first purification cycle, from where an impossibility to achieve with the TBP a scheme that would lead in a single cycle to obtain finished products meeting the above specifications; * the partitioning of uranium and plutonium into two aqueous streams requires the reduction of plutonium (IV) to plutonium (III) (because with TBP, the separation factor between uranium (VI) and plutonium (IV) ) is insufficient, whatever the acidity of the aqueous solution used to make this partition) and, consequently, to use reducing and anti-nitrous agents in large quantities, which generate by degradation unstable and reactive species and are therefore restrictive in terms of safety; * degradation products of TBP affect the performance of the process; in particular, di-n-butyl phosphate (or DBP) leads to the formation of metal complexes some of which are insoluble and can cause retention of plutonium in the solvent, hence the need to carry out an operation called "Pu Dam Which is located downstream of the reductive desextraction of plutonium and which aims to complete this desextraction; the risk of formation of a third phase induced by the presence of plutonium is limiting with respect to the implementation of a plutonium concentrating scheme (for criticality risks) or a scheme allowing the treatment used nuclear fuels with high plutonium content such as MOX fuels from light water reactors or fast neutron reactors; 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 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 organic diluent aqueous phases from different cycles of extraction to recover solubilized TBP in these aqueous phases; and * the incineration of spent TBP and its degradation products generates secondary waste including phosphate solid residues.
Also, with a view to setting up nuclear fuel treatment plants in the future that are simpler and more compact than current plants and with even better safety, 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.
It is found that Λ /, / V-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.
There are two types of Λ /, V-dialkylamides: A /, A / -dialkylamides called "symmetrical" because the two alkyl groups carried by the nitrogen atom are identical; and A /, A / -dialkylamides called "asymmetrical" because the two alkyl groups carried by the nitrogen atom are different.
The symmetrical Λ /, / V-dialkylamides are the first to have been studied. Thus, three French patent applications (FR-A-2 591 213, FR-A-2 642 561 and FR-A-2 642 562, hereinafter references [1], [2] and [3]) relating to the use of symmetrical Λ /, / V-dialkylamides as extractants in the treatment of spent nuclear fuels was notably filed in the 1980s, two of which, namely references [1] and [3], envisage the possibility of carrying out a partition of the uranium and plutonium with these A /, A / -dialkylamides without carrying out a reductive desextraction of the plutonium.
Some of the symmetrical A /, A / -dialkylamides proposed in references [1] and [3] effectively make it possible to co-extract uranium (VI) and plutonium (IV) from a strongly acidic aqueous solution and then to separate them. from each other at lower acidity without having to reduce the plutonium.
However, it turns out that these α, β-dialkylamides extract plutonium less well from a strongly acidic aqueous phase than does the TBP. As a result, in order to obtain a quantitative extraction of the plutonium, it is necessary to increase the number of extraction stages compared to that which is necessary with the TBP, which goes against the desired objective of compactness.
Then, the asymmetrical A /, α-dialkylamides have given rise to a number of studies including those conducted by the Bhabha Atomic Research Center in Bombay (see, for example, the publications by Ruikar et al. , Journal of Radioanalytical and Nuclear Chemistry 1993, 176 (2), 103-111, and Prabhu et al., Radiochemica Acta 1993, 60, 109-114, hereinafter references [4] and [5]) and those conducted by the group led by Guo-Xin Sun of Jinan University (see, for example, the publications in the names of Cui et al., Radiochemica Acta 2005, 93, 287-290, and Sun et al., Journal of Radioanalytical and Nuclear Chemistry 2005, 264 (3), 711-713, hereinafter references [6] and [7]).
However, besides the results of these studies are piecemeal and sometimes contradictory, none of them suggests the possibility of separating uranium and plutonium without reducing the latter.
DISCLOSURE OF THE INVENTION The invention therefore proposes, in the first place, new Λ /, V-dialkyl amides which are asymmetrical and which correspond to formula (I) below:
in which R represents a linear alkyl group comprising from 8 to 15 carbon atoms.
In what precedes and what follows, the expressions "from .... to ....", "from .... to ...." and "understood (e) between .... and. ... "are equivalent and mean to mean that the boundaries are included.
Thus, the term "linear alkyl group comprising from 8 to 15 carbon atoms" means any alkyl group chosen from the n-octyl, n-nonyl, n-decyl, n-undecyl, n-dodecyl and n-tridecyl groups, n-tetradecyl and n-pentadecyl.
In addition, the terms "aqueous solution" and "aqueous phase" are equivalent and interchangeable, just as the terms "organic solution" and "organic phase" are equivalent and interchangeable.
According to the invention, it is preferred that the linear alkyl group, represented by R, does not comprise more than 12 carbon atoms, and this, for reasons of viscosity (the viscosity of Λ /, / V-dialkylamides increasing, in particular). effect, with the number of carbon atoms that R) presents.
More preferably, it is preferred that this group be selected from n-octyl, n-decyl and n-dodecyl, the n-octyl group being very particularly preferred.
The N, N-dialkylamides defined above are advantageously obtained by reacting a halide of formula (II) below:
in which X represents a halogen atom and, preferably, a chlorine atom, with an amine of the formula HN (CH 3) R in which R represents a linear alkyl group comprising from 8 to 15 carbon atoms, in the presence of a base.
Also, the subject of the invention is also a process for synthesizing / V, / V-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 / V, / V-dialkylamides defined above have been found to be capable of very efficiently extracting uranium (VI) and plutonium (IV) from an aqueous acid solution such as an aqueous nitric solution.
Also, the subject of the invention is still the use of a β- / β-dialkylamide or a mixture of Λ /, V-dialkylamides as defined above, in order to extract the uranium (VI) and / or plutonium (IV) of an acidic aqueous solution.
According to the invention, 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. comprising the N, N-dialkylamide or the mixture of N, N-dialkylamides in an organic diluent, and then separating the aqueous and organic solutions.
In which case, 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 the N, N-dialkylamide or A /, A / -dialkylamide mixture.
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.
In addition to being capable of quantitatively extracting uranium (VI) and plutonium (IV) from an aqueous acidic solution, the α, β-dialkylamides defined above have been found to subsequently make it possible to separate one on the other hand, uranium and plutonium thus extracted, without any reduction of plutonium, this separation being either: a total separation of uranium and plutonium, that is to say in which two aqueous solutions including one, plutonium without uranium, and the other, uranium without plutonium; or a partial separation of 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.
Also, the invention also relates to the use of a / V, / V-dialkylamide or a mixture of Λ /, / V-dialkylamides as previously defined, to separate totally or partially the uranium (VI) of plutonium (IV) from an acidic aqueous solution, which use comprises: a) a co-extraction of uranium and plutonium from the aqueous solution, this coextraction comprising at least one in contact with the aqueous solution with an organic solution comprising the β-N-dialkylamide or the mixture of β-N-dialkylamides as extractant in solution in an organic diluent and then a separation of the aqueous and organic solutions; b) 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; whereby an aqueous solution is obtained 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 N, N-dialkylamide or a mixture of N, N-dialkylamides.
As for the acidic aqueous solution from which uranium and plutonium are co-extracted, it 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.
In addition to having the abovementioned properties, the Λ /, / 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.
Given this combination of properties, these Λ /, / V-dialkylamides have made it possible to develop a method for treating a nitric aqueous dissolution solution of 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.
Thus, 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: a) co-extraction of uranium and plutonium from the aqueous solution, the co-extraction comprising at least contacting, in an extractor, the aqueous solution with an organic solution comprising an N, N-dialkylamide or a mixture of N, N-dialkylamides defined above, as extractant, in solution in a diluent organic, then a separation of aqueous and organic solutions; 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; c) 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 state + IV, and a uranium fraction of the organic solution resulting from step b), this de-extraction comprising at least 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, followed by separation of the organic and aqueous solutions; C2) an extraction of all or part of the uranium fraction present in the aqueous solution from Ci), 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; d) a decontamination of the organic solution resulting from step ci) with respect to technetium, the decontamination comprising: di) a technetium extraction, at the oxidation level + IV, of the organic solution resulting from the step ci), this de-extraction comprising at least one contacting, in an extractor, of the organic solution with an aqueous solution comprising from 0.1 mol / L to 3 mol / L of nitric acid and at least one reducing agent capable to reduce the technetium of the degree of oxidation + VII to the degree of oxidation + IV, then a separation of the organic and aqueous solutions; d2) an extraction of the uranium fraction present in the aqueous solution resulting from stage di), this extraction comprising at least one contacting, in an extractor, of the aqueous solution with an organic solution identical to the organic solution used in step a), followed by separation of the aqueous and organic solutions; 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); whereby a first and a second decontaminated aqueous solution is obtained with respect to americium, curium and fission products including 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.
According to the invention, the organic solution used in step a) and hence those used in steps C2) and d2> since the organic solutions used in steps a), C2) and d2) 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 Λ /, -V-dialkylamide or the mixture of / V, / V-dialkylamides.
As previously indicated, the aqueous solution used in step b) may comprise from 0.5 mol / l to 6 mol / l of nitric acid.
However, it is preferred that 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). In which case, 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.
According to the invention, the bringing into contact of the organic and aqueous solutions in the extractor in which step 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 of 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 which is preferably used at a concentration ranging from 0.1 mol / l to 0.3 mol / l and, typically, from 0.2 mol / l.
Furthermore, 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. The extractor in which step di) is therefore preferably is heated to a temperature between 30 ° C and 40 ° C.
According to the invention, step d2) 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 the step d2) 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.
Whatever the temperature at which step e) is carried out, 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.
As previously indicated, 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 process of the invention has, in addition to those already mentioned, the following advantages: the desextraction of uranium is easier to implement than that of the PUREX process since it can be carried out at room temperature as well as at room temperature. while hot and using an O / A flow ratio greater than 1, which makes it possible to extract the uranium in a concentrating manner, which is not possible in the PUREX process; - because it does not involve any plutonium reduction reaction and thus eliminates any risk of re-oxidation of plutonium, 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; the degradation products (by hydrolysis and radiolysis) of β- / V-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, / V-dialkylamides typically have an aqueous phase solubility 100 to 200 times lower than that of the TBP, which makes it possible to envisage the removal or, at the very least, a lightening of the washing with the organic diluent of the solutions. aqueous processes resulting from the process of the invention compared to those provided in the PUREX process; Λ /, / V-dialkylamides and their degradation products comprising only carbon atoms, hydrogen, oxygen and nitrogen, they are totally incinerable and therefore do not produce penalizing secondary waste contrary to the TBP and its degradation products. Other features and advantages of the invention will emerge from the additional description which follows and which refers to the appended figures.
It goes without saying, however, that this additional description is given only as an illustration of the subject of the invention and should in no way be interpreted as a limitation of this object.
BRIEF DESCRIPTION OF THE FIGURES
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 of plutonium on the other hand, as obtained in tests of extraction carried out with a Λ /, / V-dialkylamide of the invention, as a function of the logarithm of the free concentration (in mol / L) of this free Λ /, / V-dialkylamide in the organic phase used in these tests. extraction.
FIG. 2 represents a schematic diagram of the process for treating an aqueous nitric solution for dissolving a spent nuclear fuel of the invention; in this figure, 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.
DETAILED DESCRIPTION OF PARTICULAR EMBODIMENTS I - SYNTHESIS OF Λ /./V-DIALKYLAMIDES OF THE INVENTION: 1.1 - Synthesis of / V-methyl- / V-octyl-2-ethylhexanamide or MOEHA:
MOEHA, which corresponds to formula (I) above, in which R represents an n-octyl group, is synthesized from 2-ethylhexanoyl chloride (136.5 g - 0.839 mol - 1.19 eq.) and β-methyl-β-octylamine (100 g - 0.698 mole -1 eq.) in the presence of 30% sodium hydroxide (112 g - 0.839 mol - 1.19 eq.) in water (100 g).
To do this, sodium hydroxide, water and β-methyl-β-octylamine are introduced into a fully equipped 500 ml reactor. The system is stirred, the set point is set at 4 ° C. The 2-ethylhexanoyl chloride is then poured at a mass temperature between 14 ° C and 17 ° C (casting time: 90 minutes). The progress of the reaction is monitored and shows the presence of 0.6% residual amine. Ninety percent MOEHA was formed and 8.8% was an unknown impurity. 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). 13 C NMR (100 MHz, CDCl 3, 25 ° C) δ (ppm): 3.37 (t, J = 7.0, 2H, 2H <xa); 3.28 (t, J = 7.0, 2H, 2H (xb), 3.00 (s, 3H, CH3a), 2.92 (s, 3H, CH3B), 2.61 - 2.42 (m, 2H); , H2a and H2B); 1.90 - 1.77 (m, 2H, CH2A); 1.71 - 1.35 (m, 10H, 2CH2A and 3CH2B); 1.34 - 1.11 (m, 28H, 7CH2A and 7CH2b) 0.94 - 0.77 (m, 18H, 3CH3A and CH3B) 3H NMR (400MHz, CDCl2S ° C) δ (ppm) 176.2 176.0 (COA and COB) 50.1, 48.1 (CaA and CaB), 43.2, 42.9 (C2A and C2B), 35.6, 33.8 (CH3A and CH3B), 32.8, 32.7, 31, 9, 31.9, 30.1, 30.0, 29.5, 29.4, 29.4, 29.3, 29.2, 27.5, 27.5, 27.0, 26.9; 26.2, 23.1, 23.0, 22.8 (20CH2), 14.2, 14.2, 14.1, 14.1 (2CH3A and 2CH3B), 12.3, 12.2 (CH3A and CH3B) HR-ESI-MS: calculated for [MH] + (C17H35NO) = 269.2714, found = 269.2672 1.2 - Synthesis of / V-decyl- / V-methyl-2-ethylhexanamide or MDEHA:
The MDEHA, which has the formula (I) above, wherein R represents an n-decyl group, is synthesized from 2-ethylhexanoyl chloride (25 g - 0.15 mole - 1.1 eq.) and N-decyl-N-methylamine (24 g - 0.14 mol - 1 eq) in the presence of triethylamine (21.2 g - 0.21 mol - 1.48 eq.) in dichloromethane ( 100 mL).
For this purpose, the dichloromethane, triethylamine and N-decyl- / V-methylamine are introduced into a 500-mL reactor equipped with all the necessary equipment. The system is stirred and cooled to 0 ° C. The 2-ethylhexanoyl chloride is then poured at a mass temperature between 5 ° C and 16 ° C (casting time: 45 minutes). Under stirring, the mass temperature rises gradually to room temperature. After 90 minutes, the progress of the reaction is monitored and shows that there is no longer any starting amine but that 5% of 2-ethylhexanoic chloride remains. 88.9% MDEHA was formed and 4.6% was an unknown impurity. The medium is then washed successively twice with 100 ml of a 10% sodium hydroxide solution and then twice with 100 ml of a 1N hydrochloric acid solution and with 100 ml of a solution of sodium carbonate. 5% sodium. The organic phase is then concentrated under reduced pressure to obtain 43.6 g of an oil. This oil contains 89% MDEHA and 9.46% of the unknown impurity.
The MDEHA is then obtained at a purity of 99.4% (measured by GC-FID) after two distillations under pressure (1.5 mbar). 1.3 - Synthesis of / V-dodecyl- / V-methyl-2-ethylhexanamide or MDdEHA:
MDdEHA, which has the formula (I) above, wherein R represents an n-dodecyl group, is synthesized from 2-ethylhexanoyl chloride (26 g - 0.16 mol - 1.06 eq.) and β-dodecyl-β-methylamine (30 g - 0.15 mol - 1 eq) in the presence of triethylamine (22.7 g - 0.223 mol - 1.49 eq) in dichloromethane (150 mL ).
For this purpose, the dichloromethane, triethylamine and β-dodecyl-β-methylamine are introduced into a 500mL reactor fully equipped. The system is stirred and cooled to 0 ° C. The 2-ethylhexanoyl chloride is then poured at a mass temperature between 0 ° C and 2 ° C (casting time: 40 minutes). Under stirring, the mass temperature rises gradually to room temperature. After 4 hours, the progress of the reaction is monitored and shows that no more amine starting. 97% of MDdEHA was formed. The medium is then washed successively twice with 100 ml of a 10% sodium hydroxide solution and then twice with 100 ml of 1N hydrochloric acid solution and with 1 time of 100 ml of a carbonate solution. 5% sodium. The organic phase is then concentrated under reduced pressure to obtain 55.5 g of an oil.
The MDdEHA is then obtained at a purity of 99.5% (measured by GC-FID) after a single very low pressure distillation (0.7 mbar).
II - EXTRACTANT PROPERTIES OF THE IV / V-DIALKYLAMIDES OF THE INVENTION: II.1 - Acquisition of the Uranium and Plutonium Distribution Coefficients and FSu / Pu Separation Factors from a Synthetic Aqueous Solution uranium and plutonium:
Extraction tests are first performed using: as organic phases: solutions comprising 1.4 mol / L of MOEHA, MDEHA or MDdEHA 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 mol / l of HNO 3.
Then, desextraction tests are carried out using: as organic phases: the organic phases obtained after the extraction tests above; and as aqueous phases: aliquots of an aqueous solution comprising 0.1 mol / l of HNC 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 below shows, for each Λ /, / V-dialkylamide tested, the concentrations of uranium, noted [U] org., As obtained in the organic phases after the extraction tests. , the distribution coefficients of uranium, denoted Du, and plutonium, denoted Dpu, as obtained at the end of the extraction and desextraction tests, the concentrations of nitric acid, denoted by [HNChlaq., as obtained in the aqueous phases at the end of the extraction and desextraction tests, as well as the U / Pu separation factors, denoted FSu / pu, as obtained at the end of the desextraction tests.
This table also shows the experimental results obtained under the same operating conditions but using as organic phases, solutions comprising / V, / V-dialkylamides of the state of the art, namely: a solution comprising 0.9 mol / L of / V, / V-di (2-ethylhexyl) - / sobutanamide (or DEHiBA) and 0.5 mol / L of / V, / V-di (2-ethylhexyl) -n-butanamide (or DEHBA) ) in TPH, these two Λ /, / V-dialkylamides being proposed in reference [3] under the names DOiBA and DOBA; and a solution comprising 1.4 mol / L / V-di (2-ethylhexyl) -3,3-dimethylbutanamide (or DEHDMBA) in TPH, this Λ /, V-dialkylamide being proposed in reference [1] under the name DOTA.
Table I
This table shows that at high acidity, the Λ /, / V-dialkylamides of the invention extract both uranium (VI) (Du (vi)> 2,4) and Λ /, / V-dialkylamides from state of the art but extract more plutonium (IV) (Dpu (iv)> 1.7) than the latter.
It also shows that the plutonium (IV) can then be easily extracted from the organic phase by means of an aqueous nitric solution of low acidity ([HNO 3] = 0.5 M) whereas the uranium is preferentially maintained in this phase. organic (FSu / pu> 14). 11-2 - Study of the stoichiometry of complexes formed by MOEHA with uranium and plutonium:
Extraction tests are carried out using: as organic phases: 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 as aqueous phases: aliquots of an aqueous solution comprising 2 g / l of uranium (VI), 1 mol / l of HNO 3 and 2 mol / l of UNO 3, and aliquots of an aqueous solution comprising 1, 7.10-4 mol / L of plutonium (IV), 1 mol / L of HNO 2 mol / L of UNO3.
To do this, 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.
The results are illustrated in FIG. 1 which represents, in the form of two straight lines, the variation of the logarithm of the distribution coefficients, denoted Dm, of uranium on the one hand, and of 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).
This figure shows that the slope of the line corresponding to the extraction of uranium (VI) is close to 2, confirming the formation of a complex U02 (N03) 2 (M0EHA) 2 which is in accordance with the complexes conventionally observed with / V, / V-dialkylamides.
On the other hand, according to these results, the complex formed by MOEHA with plutonium (IV) would involve three MOEHA molecules for a Pu4 + cation, thus presenting a Pu: MOEHA 1: 3 [Pu (N0 3) 4 ( M0EHA) 3], already observed with other / V, / V-dialkylamides asymmetric (reference [5]). The extraction equilibrium of plutonium (IV) by MOEHA can therefore be written as follows:
Pit4 + + 4N03- + 3MOEHA "-> Pu (N03) 4M0EHA3 11.3 - Acquisition of distribution coefficients of uranium, plutonium and fission products from an aqueous solution resulting from the dissolution of nuclear fuel pellets in HNO3:
Extraction tests are carried out using: as organic phase: a solution comprising 1.4 mol / L of MOEHA in TPH; and as aqueous phase: an aqueous solution previously obtained by dissolving in 5M nitric acid pellets from different irradiated fuels of UOX-REB type (Boiling Water Reactor) and UOX-REP (Pressurized Water Reactor).
This aqueous solution comprises 4.3 mol / l of HNO 3 and its composition in elements is presented in Table II below.
Table II
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.
Then, these phases are separated from each other after centrifugation.
The concentrations of uranium and plutonium on the one hand, and the β-γ-isotope activities on the other, are measured in each of the organic and aqueous phases thus separated, by X-ray fluorescence for uranium and plutonium, and by y spectrometry for β-γ isotopes.
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.
The results obtained in terms of acidity in the aqueous phase, denoted [H +] aq., Concentrations of uranium and plutonium in the aqueous and organic phases, denoted respectively [U] aq., [U] org., [ Pu] aq. and [Pu] org., and distribution coefficients, denoted D, are reported in Table III below.
Also shown in this table are the experimental results obtained under the same operating conditions but using as organic phase, a solution comprising 30% (v / v) of TBP in TPH.
Table III
This table shows that the use of MOEHA as an extractant leads to high distribution coefficients (»1) for ruranium (VI) and plutonium (IV) at an acidity of 5.75 mol / L of HN03, despite the high saturation of the organic phase in uranium (89 g of uranium / L).
It also shows that the use of MOEHA as an extractant also leads to high FSu / pf and FSpu / pf separation factors, particularly with respect to ruthenium 106, since these are all greater than 3,000. FSu / Am and FSpu / Am separation factors are also very high.
These results, which are very similar to those observed under identical conditions but using TBP as extractant, confirm that the Λ /, / V-dialkylamides of the invention make it possible to extract uranium and plutonium quantitatively and quantitatively. selective for americium, curium and the main fission products that may be present in an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, while at the same time allowing partition plutonium and uranium into two aqueous streams including, for the first, plutonium with or without uranium, and, for the second, uranium without plutonium, without having to reduce the plutonium , which is not the case with the TBP. III - PRINCIPLE DIAGRAM OF THE METHOD OF TREATING A NITRIC AQUEOUS DISSOLUTION SOLUTION OF NUCLEAR NUCLEAR FUEL:
Referring to Figure 2 which shows a block diagram of the method of treating an aqueous nitric solution for dissolving a spent nuclear fuel of the invention.
As shown in this figure, the process comprises 8 steps.
The first of these steps, denoted "Co-extraction U / Pu" in FIG. 1, 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 3 to 6 mol / L HNC> 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 step "coextraction U / Pu" is carried out by circulating, in the extractor 1, the solution of dissolution against the current 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, / V-dialkylamide of the invention or a mixture of Λ /, / V-dialkylamides of the invention, dissolved in an organic diluent.
This organic diluent is a linear or branched aliphatic hydrocarbon such as n-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, denoted "PF wash" in FIG. 2, 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.
To do this, 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 HNC> 3 but is preferably from 4 mol / l to 6 mol / l. L> 3 HNC, and more preferably 4 to 5 mol / L HNC> 3 so as to facilitate the removal of ruthenium and technetium.
If the "Wash PF" step is carried out with one or more aqueous solutions of high acidity, that is to say typically equal to or greater than 3 mol / L of HNO 3, then 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 HNC > 3 as, for example, an aqueous solution comprising 0.5 mol / L of HIMCh, in order to prevent an excessive amount of acid being entrained towards the extractor devolved to the third stage, denoted "Dextraction" Pu "in Figure 2, and does not disturb the performance of this third step. 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.
It is carried out by circulating, in the extractor 3, this organic phase against the current of an aqueous solution comprising from 0.1 mol / l to 0.5 mol / l of HNCh and preferably using a ratio of O / A flow rates greater than 1, preferably equal to or greater than 3 and, more preferably, equal to or greater than 5 so that the plutonium (IV) is concentrically desextracting.
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. ".
Also, the fourth step of the process, denoted "1st Washing U" in FIG. 2 and which represents the second stage of the U / Pu partition, is intended to extract from the aqueous phase resulting from the "Pu Desextraction": or all the uranium present in this aqueous phase if it is desired for 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 U wash", an aqueous solution comprising uranium and plutonium in a previously chosen ratio, if it is desired that the U / U partition be Pu lead to an aqueous solution comprising a mixture of plutonium and uranium in this ratio and an organic solution comprising uranium without plutonium.
In both cases, the "1st Washing U" is performed 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 phase organic used for the "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, denoted "α-Tc barrier" in FIG. 2, 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.
It also makes it possible to extract from the organic phase resulting from the "Pu Desextraction" the fraction of neptunium that has been extracted during the "Coextraction U / Pu" which followed the technetium up to the "A-Tc Dam", as well as traces of plutonium that this organic phase is likely to contain again.
It is carried out by circulating, in the extractor 5, the organic phase resulting from the "Pu de-extraction" in countercurrent of a nitric acid solution of low acidity, that is to say comprising 0.1. mol / L at 3 mol / L of HNC> 3 and, better still, 1 mol / L of HNO3, and comprising one or more reducing agents making it possible to reduce the technetium - which is present in the organic phase to the degree of oxidation + VII - in technetium (IV) not extractable by Λ /, / V-dialkylamides, neptunium (VI) in neptunium (IV) or neptunium (V) which are not extractable by / V, / V-dialkylamides low acidity, and the plutonium (IV) plutonium (III) which is less extractable by à /, / V-dialkylamide low acidity than is the plutonium (IV), and without reducing uranium (VI).
As reducing agents, 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, denoted "2nd Washing U" in FIG. 2, is intended to extract from the aqueous phase resulting from the "α-Tc barrier" the uranium that has been desextracted, together with the technetium, in the previous step in order to avoid that the "α-Tc Dam" step results in an excessive loss of uranium in the aqueous phase.
It is carried out by circulating, in the extractor 6, the aqueous phase resulting from the "α-Tc barrier" in counter-current of an organic phase of composition identical to that of the organic phases used for the "Co-extraction U / Pu "and" 1st Washing U ", after an acidification of this aqueous phase by an addition of concentrated nitric acid, for example 10 M, to promote the extraction of uranium.
The seventh step, denoted "Desextraction U" in FIG. 2, aims to extract the uranium (VI) from the organic phase resulting from the "a-Tc barrier".
It is carried out by circulating, in the extractor 7, the organic phase resulting from the "α-Tc barrier" in countercurrent to a nitric aqueous solution of very low acidity, that is to say comprising at most 0 , 05 mol / L of HN03 as, for example, an aqueous solution comprising 0.01 mol / L of HNO3. 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.
At the end of these 7 steps, are obtained: - 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 FIG. 2) and, for the second, technetium, neptunium and, possibly, traces of plutonium ("secondary raffinate" in FIG. 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 aqueous phase exiting the extractor 7, which comprises decontaminated uranium and which is called "U-flow"; and the organic phase leaving the extractor 7, which no longer comprises plutonium or uranium but which may contain a certain number of impurities and degradation products (formed by hydrolysis and radiolysis) of the extractant, which would have accumulated during the previous steps.
Also, the eighth step, denoted "washing PO" in FIG. 2, 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, then one or more washes with an aqueous solution of nitric acid allowing the reacidifying, 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 the current of the aqueous washing solution.
As can be seen in FIG. 2, the organic phase thus regenerated can then be returned to extractors 1 and 4 for reintroduction into the treatment cycle.
REFERENCES CITED
[1] FR-A-2,591,213 [2] FR-A-2,642,561 [3] FR-A-2,642,562 [4] Ruikar et al., Journal of Radioanalytical and Nuclear Chemistry 1993,176 (2) , 103-111 [5] Prabhu et al., Radiochemica Acta 1993, 60, 109-114 [6] Cui et al., Radiochemica Acta 2005, 93, 287-290 [7] Sun et al., Journal of Radioanalytical and Nuclear Chemistry. 2005, 264 (3), 711-713

权利要求:
Claims (22)
[1" id="c-fr-0001]
1. A /, / V-dialkylamide of formula (I) below:

in which R represents a linear alkyl group comprising from 8 to 15 carbon atoms.
[0002]
The β-dialkylamide according to claim 1, wherein the linear alkyl group comprises at most 12 carbon atoms.
[0003]
3. A /, A-dialkylamide according to claim 2, wherein the linear alkyl group is n-octyl, n-decyl or n-dodecyl, preferably n-octyl.
[4" id="c-fr-0004]
4. Process for synthesizing a β-, V-dialkylamide according to any one of claims 1 to 3, which comprises reacting a halide of formula (II) below:

wherein X represents a halogen atom, with an amine of the formula HN (CH3) R wherein R represents a linear alkyl group comprising from 8 to 15 carbon atoms.
[5" id="c-fr-0005]
5. Use of an N, Nd, A, or a mixture of N, N-alkylamides according to any one of claims 1 to 3 for extracting uranium (VI and / or the plutonium (IV) of an acidic aqueous solution.
[6" id="c-fr-0006]
Use according to claim 5, which comprises contacting the acidic aqueous solution with an organic solution comprising the N, N-dialkylamide or the mixture of N, N-dialkyl amides in an organic diluent, followed by separation of the aqueous and organic solutions.
[7" id="c-fr-0007]
The use according to claim 6, wherein the organic solution comprises from 1 mol / L to 2 mol / L and more preferably from 1.3 mol / L to 1.5 mol / L of Λ /, / V- dialkylamide or the mixture of / V, / V-dialkylamides.
[8" id="c-fr-0008]
8. Use of a Λ /, V-dialkylamide or a mixture of Λ /, / V-dialkyl amides according to any one of claims 1 to 3, to separate totally or partially uranium (VI) plutonium (IV) from an acidic aqueous solution, which use comprises: a) co-extraction of uranium and plutonium from the aqueous solution, the coextraction comprising at least one contacting of the solution aqueous solution with an organic solution comprising Λ /, V-dialkylamide or the mixture of Λ /, V-dialkylamides as extractant, in solution in an organic diluent, and then a separation of the aqueous and organic solutions; b) a plutonium dextraction, at the oxidation state + IV, and a fraction of the uranium of the organic solution resulting from step a), the de-extraction comprising at least a contacting of the organic solution with an aqueous solution comprising from 0.1 mol / l to 0.5 mol / l of nitric acid, followed by 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), the extraction comprising at least bringing the aqueous solution into contact with an organic solution identical to the solution organic of a), then a separation of the aqueous and organic solutions; whereby an aqueous solution is obtained comprising plutonium without uranium or a mixture of plutonium and uranium, and an organic solution comprising uranium without plutonium.
[9" id="c-fr-0009]
Use according to claim 8, wherein the organic solution of step a) comprises from 1 mol / L to 2 mol / L and more preferably from 1.3 mol / L to 1.5 mol / L / V, / V-dialkylamide or the mixture of / V, / V-dialkylamides.
[10" id="c-fr-0010]
10. Use according to any one of claims 5 to 9, wherein the acidic aqueous solution is an aqueous solution from the dissolution of spent nuclear fuel in nitric acid.
[11" id="c-fr-0011]
11. 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: a) co-extraction of the uranium and plutonium from the aqueous solution, the coextraction comprising at least one bringing into contact, in an extractor, of the aqueous solution with an organic solution comprising a N, N-dialkylamide or a mixture of N, N-dialkylamides according to any one of Claims 1 to 3, as an extractant, in solution in an organic diluent, and then a separation of the solutions. aqueous and organic; b) a decontamination of the organic solution resulting from step a) with respect to americium, curium and fission products, the 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; c) 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 plutonium-free uranium, the partition comprising: ci) a plutonium dextraction, at the oxidation state + IV, and a uranium fraction of the organic solution resulting from step b), the de-extraction comprising at least 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, followed by separation of the organic and aqueous solutions; C2) an extraction of all or part of the uranium fraction present in the aqueous solution resulting from step c1), the extraction comprising at least one bringing into contact, in an extractor, of the aqueous solution with an organic solution identical to the organic solution of step a), and then a separation of the aqueous and organic solutions; d) a decontamination of the organic solution resulting from stage ci) with respect to technetium, the decontamination comprising: di) a technetium removal at the oxidation level + IV of the organic solution resulting from the step Ci), the de-extraction comprising at least one contacting, in an extractor, of the organic solution with an aqueous solution comprising from 0.1 mol / L to 3 mol / L of nitric acid and at least one reducing agent capable to reduce the technetium of the degree of oxidation + VII to the degree of oxidation + IV, then a separation of the organic and aqueous solutions; d2) an extraction of the uranium fraction present in the aqueous solution resulting from step di), the extraction comprising at least one contacting, in an extractor, of the aqueous solution with an organic solution identical to the solution organic of a), then a separation of the aqueous and organic solutions; e) a removal of the uranium from the organic solution resulting from step di), the de-extraction comprising at least one bringing into contact, 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); whereby a first and a second decontaminated aqueous solution is obtained with respect to americium, curium and fission products including 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.
[12" id="c-fr-0012]
The process according to claim 11, wherein the organic solution of step a) comprises from 1 mol / L to 2 mol / L and more preferably from 1.3 mol / L to 1.5 mol / L / V, N-dialkylamide or a mixture of β-alkylamides.
[13" id="c-fr-0013]
13. The method of claim 11 or claim 12, wherein the aqueous solution of step b) comprises from 4 to 6 mol / L of nitric acid.
[14" id="c-fr-0014]
14. The method of claim 13, wherein step b) further comprises a deacidification of the organic solution, the deacidification comprising at least a contacting of the organic solution with an aqueous solution comprising 0.1 mol / L at 1 mol / L of nitric acid, then a separation of the organic and aqueous solutions.
[15" id="c-fr-0015]
15. A method according to any one of claims 11 to 14, wherein the bringing into contact of the organic and aqueous solutions in the extractor of step Ci) comprises a circulation of organic and aqueous solutions with a ratio of the flow rate of the organic solution at the flow rate of the aqueous solution greater than 1 and, preferably, equal to or greater than 3.
[16" id="c-fr-0016]
The process according to any one of claims 11 to 15, wherein the reducing agent is uranic nitrate, hydrazinium nitrate, hydroxylammonium nitrate, acetaldoxime or a mixture thereof, preferably a mixture of uranous nitrate and hydrazinium nitrate or a mixture of uranous nitrate and hydroxylammonium nitrate.
[17" id="c-fr-0017]
17. A process according to any one of claims 11 to 16, wherein the extractor of step di) is heated to a temperature of 30 ° C to 40 ° C.
[18" id="c-fr-0018]
18. A process according to any one of claims 11 to 17, wherein step d2) comprises acidification of the aqueous solution from step di) to bring the concentration of nitric acid in the aqueous solution to a minimum. value at least equal to 2.5 mol / L, the acidification comprising an addition of nitric acid in the extractor of step d2).
[19" id="c-fr-0019]
19. The method of any one of claims 11 to 18, wherein the extractor of step e) is heated to a temperature of 40 ° C to 50 ° C.
[20" id="c-fr-0020]
20. Process according to any one of claims 11 to 19, wherein the bringing into contact of the organic and aqueous solutions in the extractor of step e) comprises a circulation of the organic and aqueous solutions with a ratio of the flow rate of the organic solution at the flow rate of the aqueous solution greater than 1.
[21" id="c-fr-0021]
21. The method according to any one of claims 11 to 20, wherein the regeneration of the organic solution from step e) comprises at least one washing of the organic solution with a basic aqueous solution, followed by at least one washing the organic solution with an aqueous solution of nitric acid.
[22" id="c-fr-0022]
22. Process according to any one of claims 11 to 21, wherein the organic solution resulting from step f) is divided into a first and a second fraction, the first fraction forming the organic solution of step a) and the second fraction forming the organic solution of step C2).

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FR2551745A1|1985-03-15|PROCESS FOR THE SEPARATION OF NEPTUNIUM FORMING FROM AN ORGANIC PHASE DURING THE RETIREMENT OF IRRADIATED COMBUSTIBLE AND / OR FERTILIZED MATERIALS

FR2748951A1|1997-11-28|PROCESS FOR THE SELECTIVE SEPARATION OF ACTINIDES | AND LANTHANIDES |

RU2514947C2|2014-05-10|Method for re-extraction of plutonium from organic solution of tributyl phosphate

FR2684670A1|1993-06-11|NITROGENIC HETEROCYCLIC SUBSTITUTING AMIDES, PROCESS FOR THEIR PREPARATION AND THEIR USE FOR SELECTIVELY EXTRACTING ACTINIDES | AND SEPARATING THEM IN PARTICULAR FROM LANTHANIDES |.

BE1000098A3|1988-03-08|Method for improving the efficiency of decontamination solution of nuclear fuels and / or materials contaminated with fertile zirconium.

WO2012104246A1|2012-08-09|Process for separating at least one platinoid element from an acidic aqueous solution comprising, besides this platinoid element, one or more other chemical elements

RU2490735C2|2013-08-20|Method for preparation of spent nuclear fuel reprocessing solutions containing complexing agents for extraction of multivalent actinides

FR3063499A1|2018-09-07|PROCESS FOR RECOVERING AND PURIFYING URANIUM PRESENT IN PLUTONIUM, NEPTUNIUM AND TECHNETIUM-CONTAMINATED POTASSIUM DIURANATE SLURRY

同族专利:

公开号 | 公开日

FR3039547B1|2017-08-25|

RU2018107117A3|2019-08-28|

JP2018532691A|2018-11-08|

RU2018107117A|2019-08-28|

RU2702739C2|2019-10-10|

GB2555362A|2018-04-25|

US10252983B2|2019-04-09|

US20180222849A1|2018-08-09|

GB2555362B|2020-02-19|

CN107922314A|2018-04-17|

WO2017017193A1|2017-02-02|

GB201801315D0|2018-03-14|

JP6775572B2|2020-10-28|

CN107922314B|2020-07-07|

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CN110312702A|2017-01-26|2019-10-08|原子能和替代能源委员会|Particularly for the asymmetric N of SEPARATION OF URANIUMand plutonium , N- dialkyl amide, synthesis and purposes|FR2591213B1|1985-12-05|1988-02-05|Commissariat Energie Atomique|PROCESS FOR THE EXTRACTION OF URANIUM VI AND / OR PLUTONIUM IV FROM AN AQUEOUS SOLUTION USING N, N-DIALKYLAMIDES|

FR2642561A1|1989-02-01|1990-08-03|Commissariat Energie Atomique|Process for separating uranium from thorium which are present in an aqueous solution by means of an N,N-dialkylamide, which can be used especially for separating the uranium produced by irradiation of thorium|

FR2642562B1|1989-02-01|1991-04-05|Commissariat Energie Atomique|PROCESS FOR THE EXTRACTION OF URANIUM VI AND / OR PLUTONIUM IV FROM AN ACID AQUEOUS SOLUTION USING A MIXTURE OF N, N-DIALKYLAMIDES, FOR USE IN THE TREATMENT OF IRRADIATED NUCLEAR FUELS|

FR2880180B1|2004-12-29|2007-03-02|Cogema|IMPROVEMENT OF THE PUREX PROCESS AND ITS USES|

FR2901627B1|2006-05-24|2009-05-01|Commissariat Energie Atomique|PROCESS FOR THE REHABILITATION OF USEFUL NUCLEAR FUEL AND THE PREPARATION OF A MIXED OXIDE OF URANIUM AND PLUTONIUM|

FR2947663B1|2009-07-02|2011-07-29|Areva Nc|IMPROVED PROCESS FOR TREATING US NUCLEAR FUELS|

FR2954354B1|2009-12-22|2012-01-13|Commissariat Energie Atomique|PROCESS FOR PURIFYING URANIUM FROM A NATURAL URANIUM CONCENTRATE|

FR2960690B1|2010-05-27|2012-06-29|Commissariat Energie Atomique|PROCESS FOR PROCESSING NUCLEAR FUELS USING NO PLUTONIUM REDUCING EXTRACTION OPERATION|

FR2973377B1|2011-04-01|2013-05-17|Commissariat Energie Atomique|2,9-DIPYRIDYL-1,10-PHENANTHROLINE DERIVATIVES AS LIGANDS OF ACTINIDES, PROCESS FOR THEIR SYNTHESIS AND USES THEREOF|

FR3015760B1|2013-12-20|2016-01-29|Commissariat Energie Atomique|PROCESS FOR TREATING A USE NUCLEAR FUEL COMPRISING A DECONTAMINATION STEP OF URANIUM IN AT LEAST ONE ACTINIDE BY COMPLEXATION OF THIS ACTINIDE |FR3063499A1|2017-03-06|2018-09-07|Commissariat A L'energie Atomique Et Aux Energies Alternatives|PROCESS FOR RECOVERING AND PURIFYING URANIUM PRESENT IN PLUTONIUM, NEPTUNIUM AND TECHNETIUM-CONTAMINATED POTASSIUM DIURANATE SLURRY|

FR3068257B1|2017-06-29|2022-01-14|Commissariat Energie Atomique|CARBAMIDES FOR THE SEPARATION OF URANIUM AND PLUTONIUM WITHOUT PLUTONIUM REDUCTION|

CN109207724B|2018-09-12|2020-12-29|哈尔滨工业大学（威海）|Extraction solvent and extraction method for simultaneously extracting and separating vanadium and chromium from vanadium and chromium-containing solution|

CN110144471B|2019-05-15|2020-10-09|中国原子能科学研究院|Method for extracting technetium from nuclear fuel post-treatment waste liquid|

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优先权:

申请号 | 申请日 | 专利标题

FR1557264A|FR3039547B1|2015-07-29|2015-07-29|NOVEL DISSYMETRIC N, N-DIALKYLAMIDES, THEIR SYNTHESIS AND USES THEREOF|FR1557264A| FR3039547B1|2015-07-29|2015-07-29|NOVEL DISSYMETRIC N, N-DIALKYLAMIDES, THEIR SYNTHESIS AND USES THEREOF|

CN201680044477.7A| CN107922314B|2015-07-29|2016-07-28|Novel asymmetric N, N-dialkyl amide and synthetic method and application thereof|

PCT/EP2016/068016| WO2017017193A1|2015-07-29|2016-07-28|Novel dissymmetric n,n-dialkylamides, the snythesis thereof and uses of same|

JP2018504199A| JP6775572B2|2015-07-29|2016-07-28|New asymmetric N, N-dialkylamides, their synthesis and use|

US15/748,030| US10252983B2|2015-07-29|2016-07-28|Dissymmetric N,N-dialkylamides, the synthesis thereof and uses of same|

RU2018107117A| RU2702739C2|2015-07-29|2016-07-28|Novel asymmetric n,n-dialkylamides, synthesis thereof and use thereof|

GB1801315.1A| GB2555362B|2015-07-29|2016-07-28|Novel dissymmetric N, N-dialkylamides, the synthesis thereof and uses of same|

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