FR3063499A1 - PROCESS FOR RECOVERING AND PURIFYING URANIUM PRESENT IN PLUTONIUM, NEPTUNIUM AND TECHNETIUM-CONTAMINATED POTASSIUM DIURANATE SLURRY - Google Patents

Abstract

The invention relates to a process for the recovery and purification of uranium present in a potassium diuranate sludge which results from the fluorination of uranium tetrafluoride produced from uranyl nitrate partially derived from the refining of yellow cake. and partially from the treatment of spent nuclear fuel and which is, therefore, poorly contaminated with plutonium, neptunium and technetium. Field of application: treatment of waste produced during the upstream nuclear cycle with a view to recovering the uranium contained in this waste.

Description

Holder (s): COMMISSIONER OF ATOMIC ENERGY AND ALTERNATIVE ENERGIES Public establishment.

Extension request (s)

Agent (s): BREVALEX Limited liability company.

PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM PRESENT IN A MUD OF POTASSIUM DIURANATE CONTAMINATED WITH PLUTONIUM, NEPTUNIUM AND TECHNETIUM.

FR 3 063 499 - A1) 6 /) The invention relates to a process for the recovery and purification of uranium present in a potassium diuranate mud which results from the fluorination of uranium tetrafluoride produced from nitrate uranyl partially from the refining of yellow cake and partially from the processing of spent nuclear fuels and which is therefore slightly contaminated with plutonium, neptunium and technetium.

Field of application: treatment of waste produced during the upstream nuclear cycle with a view to recovering the uranium contained in this waste.

i

PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM PRESENT IN A MUD OF POTASSIUM DIURANATE CONTAMINATED WITH PLUTONIUM, NEPTUNIUM AND TECHNETUM

DESCRIPTION

TECHNICAL AREA

The invention relates to the field of treatment of waste produced during the upstream phase of the nuclear cycle with a view to the recovery of the uranium contained in this waste.

More specifically, it relates to a process for recovering and purifying the uranium present in a potassium diuranate mud which results from the fluorination of uranium tetrafluoride produced from uranyl nitrate originating partially from the refining of yeilow cake and partially from the processing of spent nuclear fuels and which is therefore only slightly contaminated with plutonium, neptunium and technetium.

PRIOR STATE OF THE ART

After its extraction from the mines, the uranium ore is processed to extract the uranium, purify it and obtain a uranium concentrate called "yeilow cake", which contains 60 to 80% of uranium in the form of trluranium octaoxide (U ₃ Og).

The yeilow cake is then refined by dissolution in nitric acid and by liquid / liquid extraction to obtain uranyl nitrate (UO ₂ (NO ₃ ) ₂ ) of high purity (> 99.95%).

Uranyl nitrate is then converted by hydrofluorination to uranium tetrafluoride (or UF ₄ ), which is in turn converted by fluorination to uranium hexafluoride (or UF _ë ) to be able to enrich the uranium in ²³⁵ Uranium,

As the conversion of i'lfF ₄ to UFe is not complete, solid reaction residues, called "Ultimated Fluorination" (or IUF) are accumulated during this operation. These residues, which contain uranium and many non-volatile fluorinated impurities, are stabilized in the form of potassium diuranate (K2U2O7 or KDU) and are in the form of so-called “KDU sludge”.

Typically, the uranium present in this sludge is recovered and purified by dissolving the sludge in nitric acid and by hydrometallurgical treatment of the aqueous nitric solution resulting from this dissolution, this treatment being based on the use of phosphate of tri n-butyie phosphate (or T8P) as an extractant. The urum thus recovered and purified is then recycled.

There are ultimate unburnt fluorides of another type than those which have just been described, which are called MUR (for Revalorizable Uranium Matter) and which correspond to residues of the fluorination of UF ₄ produced from nitrate uranyl partially from the refining of yellow cake and partially from the processing of spent nuclear fuels. The KDU sludge from passive MURs is therefore slightly contaminated with plutonium, neptunium and technetium, these elements being present in trace amounts in uranyl nitrate from the processing of spent nuclear fuels.

I! As a result, the treatment of KDU sludge from passive MURs is much more restrictive than that of KDU sludge from IUFs which are derived solely from yellow cake and which are free of any radioactive impurity.

The challenge is therefore to develop a process which makes it possible to recover quantitatively Uranium from a KDU sludge with a passive MUR knowing that this type of residue has high concentrations of fluorides (approximately 15 to 20 g / L) which are known to strongly compress uranium (Vl) in an aqueous medium - while purifying the uranium in a sufficiently selective manner vis-à-vis the impurities present in this sludge and, in particular, vis-à-vis plutonium, neptunium and technetium, so that the recovered uranium can meet the purity specifications imposed by ASTM C 788 for the recycling of uranium in the manufacture of new nuclear fuels.

The invention aims precisely to propose such a method.

STATEMENT OF THE INVENTION

The subject of the invention is therefore a process for recovering and purifying the uranium present in a potassium diuranate mud comprising fluorides, plutonium, neptunium and technetium, which process comprises

Its next steps:

a) a decontamination of uranium with respect to plutonium and neptunium, the decontamination comprising an extraction of uranium, at the oxidation state + VI, from a first aqueous solution resulting from a dissolution of the mud in nitric acid, by bringing the first aqueous solution into contact with an organic solution comprising at least one W, / V-dialkyiamide lipophyie as extractant, in solution in an organic diluent, then separation of the aqueous solutions and organic;

b) decontamination of the uranium with respect to the technetium, the decontamination comprising a de-extraction of the technetium from the organic solution resulting from step a), by bringing the organic solution into contact with a second solution aqueous comprising nitric acid and at least one reducing agent capable of reducing technetium to technetium not extractable by W, W-dialkyl3mide without reducing uranium (VI}, then separation of the organic and aqueous solutions; and

c) recovery of the uranium, the recovery of the uranium comprising a de-extraction of the uranium, at the oxidation state + VI, from the organic solution resulting from step b), by contacting in an extractor organic solution with a third aqueous solution, then separation of the organic and aqueous solutions, whereby an aqueous solution is obtained comprising uranium in the form of uranyie nitrate, purified vis-à-vis plutonium, neptunium and technetium.

In what follows, the terms “aqueous solution” and “aqueous phase” are equivalent and interchangeable, just as the terms “organic solution” and “organic phase” are equivalent and interchangeable.

Furthermore, the expressions "from .... to ....", "ranging from .... to ...." and "between (e) between .... and ...." are equivalent and mean that the terminals are included.

As previously indicated, the extractant which is used in step a) is a / V, W ~ diaIkylamide lipophïle, that is to say a Λ /, / V-dtalkylamide having a total number of carbon atoms greater than 16.

This W, / V-dialkylamide is advantageously chosen from your Λ /, / V-diaikylamides which correspond to the formula (!) Below:

RLCfOJ-NR ^ ³ (I) in which:

R ¹ represents a linear or branched alkyl group in alpha or beta of the carbonyl group, comprising from 3 to 12 carbon atoms; and

R ² and R ³ represent, independently of one another, a linear or branched alkyl group comprising from 6 to 12 carbon atoms.

Among these W, W-dalkylamtdes, it is preferred to use those in which R ² and R ³ both represent a 2-ethyl group such as W, W- (di-2-ethyhexyi) -butanamide (or DEHBA) in which R ¹ represents a group -C3H7, le / V, W ~ di- (2éthyfhexyQ-isobutanamide (or DEHîBA) in which R ¹ represents a group -CH- (CH3} j, W, W-df ~ (2-ethylhexyl) -2,2-methylpropanamide (or DEHDMPA) in which R ¹ represents a group -C- (CH3) 3, or alternatively A /, iV-di- (2-ethylhexyl} -3,3-dimethyl ~ butanamide) ( or DEHDMBA) in which R ¹ represents a group -CH2C (CH ₃ ) ₃ .

DËHiBA is particularly preferred.

As for the organic diluent, it is preferably a tsoparaffin or a mixture of isoparaffins of the type of those sold by Total under the commercial references Isane IP 175 and IP 185,

However, it can also be another aliphatic hydrocarbon, such as n-dodecane, or a mixture of aliphatic hydrocarbons such as kerosene or hydrogenated tetrapropylene (or TPH).

Anyway, the concentration of Λ /, / V-dalkylamide in the organic diluent is typically between 1 mol / L and 2 mol / L and, better still, between 1.5 mol / L and 1.8 mol / L, preferably given at a concentration of 1.75 months / L

Furthermore, the concentration of nitric acid in the first aqueous solution is preferably equal to or greater than 3 mol / L in order to reduce the complexation of uranium by the fluoride ions present in this aqueous solution and thereby even, favor the extraction of uranium by the organic solution.

The reducing agent which is used in step b) is preferably uranous nitrate (also called “U (IV)”) which is advantageously combined with an anti-nitrous agent to stabilize it in aqueous solution, such as hydrazine nitrate (also called "hydrazine nitrate"), hydroxylammonium nitrate (also called "hydroxyiamine nitrate") or acetaldoxime (also called "acetaldehyde oxlroe"), preferably being given to acetaldoxime.

In which case, the concentration of uranous nitrate in the second aqueous solution can range from 1 g / L to 10 g / L and is preferably 5 g / L, while the concentration of acetaldoxime in this same solution can range from 0.1 mol / L to 0.5 mol / L and is preferably located at 0.3 mol / L.

In which case also, the concentration of nitric acid in the second aqueous solution is less than 3 months / L to prevent the uranous nitrate from oxidizing to uranium (Vi) which is not a reducing agent of technetium (Vll). This concentration is preferably between 0.5 mol / L and 2 mol / L and is, for example, 1 month / L.

Furthermore, step b) can be carried out at ambient temperature, that is to say at 20-25 ° C. However, it is preferably carried out at a temperature ranging from 30 ° C. to 50 ° C. so as to accelerate the kinetics of technetium extraction while at the same time limiting the phenomena of reoxidation of this element in aqueous solution. The extractor in which step b) takes place is therefore preferably heated to a temperature of between 30 ° C. and SOT and, better still, between 40 ° C. and 45 ° C.

According to the invention, the third aqueous solution which is used in step c) can be water, preferably distilled, or an aqueous solution of very low concentration nitric acid, that is to say comprising preferably not more than 0.01 mo / L of nitric acid.

Step c) can be carried out at room temperature. However, it is preferably carried out at a temperature ranging from 40’C to 55’C to, again, accelerate the kinetics of uranium stripping. The extractor in which step c) takes place is therefore preferably heated to a temperature of between 40’C and 55’C.

Whatever the temperature at which step c) is carried out, bringing the organic and aqueous solutions into contact in the extractor in which this step takes place comprises circulating these solutions in an O / A flow rate ratio greater than 1 so as to obtain a de-extraction of the concentrating uranium, that is to say a de-extraction of the uranium which leads to an aqueous solution in which the concentration of uranium is higher than that presented by this element in the organic solution from which it is extracted.

In step b), the de-extraction of technetium from the organic solution From step a) is accompanied by a de-extraction of a fraction of the uranium also present in this organic solution.

Also, so that the process can lead to a quantitative recovery of ruranium initially in the first aqueous solution, this process further comprises:

- An extraction of this uranium fraction by contacting in an extractor of the aqueous solution from step b) with an organic solution identical to - that is to say of the same qualitative and quantitative composition as - the organic solution used to extract the uranium in step a), then separation of the aqueous and organic solutions; and

an introduction of the organic solution resulting from this extraction in which stage a) takes place.

Advantageously, the process of the invention further comprises a regeneration of the organic solution resulting from step c), this regeneration preferably comprising at least one washing of the organic solution with a basic aqueous solution, followed by at least minus a wash of the organic solution with an aqueous nitric acid solution.

Advantageously, the organic solution thus regenerated 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 which is used to extract the fraction of uranium present in the aqueous solution from step b).

Other characteristics and advantages of the invention will emerge from the additional description which follows and which refers to the attached figures.

It goes without saying that this additional description is given only by way of illustration of the object of the invention and should in no case be interpreted as a limitation of this object.

BRIEF DESCRIPTION OF THE FIGURES

FIG. 1 illustrates a block diagram of a preferred embodiment of the method of the invention; in this figure, the rectangles 1 to 4 represent multi-stage extractors such as those conventionally used in the field of industrial liquid-liquid extractions (mixed with decanters, pulsed columns, centrifugal extractors, etc.).

FIG. 2 illustrates the results of tube tests aimed at assessing the impact, on the one hand, of the acidity of an aqueous solution resulting from the dissolution of a KOU sludge representative of an ultimate unburnt out of fluorination of passive MUR type in nitric acid and, on the other hand, the concentration of DEHiBA in the organic phase on the extraction of uranium (VÎ) from this aqueous solution.

FIG. 3 schematically illustrates the installation and the operating conditions having been used for a pilot test aimed at validating the feasibility on an industrial scale of the mode of implementation illustrated in FIG. 1.

In FIGS. 1 and 3, the organic phases entering and leaving the extractors are shown by solid lines while the aqueous phases entering and leaving the extractors are shown by dotted lines.

DETAILED PRESENTATION OF SPECIFIC MODES OF IMPLEMENTATION

I - DIAGRAM OF PRINCIPLE OF A PREFERRED IMPLEMENTATION MODE

First of all, reference is made to FIG. 1 which represents a block diagram of a preferred embodiment of the method of the invention.

As can be seen in this figure, the method of the invention is implemented in the form of a cycle which comprises 5 steps.

The first of these steps, denoted "Extraction U" in FIG. 1, consists in extracting uranium to the oxidation state + IV from an aqueous solution ~ called below "MUR solution" - previously obtained by dissolving in nitric acid from a KDU sludge corresponding to an ultimate fluoride burnt of the passive MUR type, to decontaminate it vis-à-vis the plutonium and neptunium also present in this solution.

Indeed, such a solution includes uranium, fluoride tones and many impurities including plutonium, neptunium, technetium - these three elements being present in trace amounts - corn also iron, nickel, copper, chromium, sodium, potassium, calcium and magnesium. It can include from 1 moI / L to 3 mol / L of nitric acid depending on the degree of concentration of the nitric acid that was used to dissolve the mud.

"Extraction U" is carried out by circulating, in extractor 1, the MUR solution against the current of an organic phase, denoted "PO" in FIG. 1, which comprises a W, W-alkylamide Hpophile such as previously defined, in solution in an organic diluent, this W, W-dtalkyiamide preferably being DEHiBA of formula: {CH ₃ ) ₂ -CH-C (O) -N- (CH ₂ -CH (C ₂ H ₅ ) C ₄ H _g h.

The organic diluent is preferably an isoparaffin or a mixture of isoparaffins of the type of those sold by Total under the commercial references Isane IP 175 and IP 185.

However, another aliphatic hydrocarbon or mixture of aliphatic hydrocarbons such as n-docedane, kerosene or TPH can also be used.

The concentration of / V, / V-dialky! Amide in the organic diluent is typically between 1 mol / L and 2 mol / L and, better still, between 1.5 mol / L and 1.8 mol / L, preferably being given at a concentration of 1.75 mol / L.

If, on entering the extractor 1, the MUR solution has a nitric acid concentration 5 of less than 3 mol / L, then the step “Extraction U” preferably further comprises an acidification of this solution within the extractor 1 so as to bring its concentration of nitric acid to a value at least equal to 3 mol / L, and this, to reduce the complexation of uranium by fluoride ions in the aqueous phase and, thereby, promote transfer uranium in the organic phase. This acidification is carried out by adding 1 highly concentrated nitric acid to the extractor, for example in the form of an aqueous solution comprising S mol / L or more of nitric acid, such as, for example, an aqueous solution containing 7.5 mol / L of nitric acid.

The second stage of the process, denoted “Washing Te” in FIG. 1, consists in de-extracting from the organic phase resulting from the stage “Extraction U” the technetium, which was extracted in organic phase jointly with uranium during this extraction, to decontaminate uranium from technetium. It also makes it possible to perfect the decontamination of uranium with respect to plutonium and neptunium which may be present (in minute quantities) in the organic phase resulting from extractor 1.

This second step is carried out by circulating in the extractor 2 the organic phase from the extractor 1 against the flow of an aqueous nitric solution which comprises one or more reducing agents making it possible to reduce the technetium - which is present in the organic phase with oxidation state + VII - in technetium which cannot be extracted with W, W -alkylamides, without reducing uranium (VI).

The reducing agent is preferably uranous nitrate (or U (IV)) which is stabilized by means of an anti-nitrous agent such as hydrazinium nitrate, hydroxylammonium nitrate or, better still , acetaldoxîme, In this case, the concentration of uranous nitrate in the aqueous solution is typically between 1 g / L and 10 g / L, for example 5 g / L, while the concentration of acetaldoxîme is, it , typically between 0.1 mol / L and 0.5 mol / L, for example 0.3 mol / L In this case also, the concentration of nitric acid in the aqueous solution should be less than 3 mol / L to prevent uranous nitrate from oxidizing to uranium (VI), This concentration is preferably between 0.5 mol / L and 2 mol / L and is, for example, 1 mol / L

The “Te washing” step can be carried out at room temperature (that is to say at 20 ° -25 ° C.) but elf is preferably carried out at a temperature ranging from 30 ° C. to SOUT and, better still, from 40 ° C to 45 ° C so as to accelerate the kinetics of reduction and, therefore, of extraction of technetium while limiting the phenomena of reoxidation of reduced technetium in the aqueous phase and, therefore, the risk of seeing technetium, once extract, be re-extracted in organic phase.

The third stage of the process, denoted “Complementary extraction Ü” in FIG. 1, consists in extracting from the aqueous solution resulting from the stage “Washing Te” the fraction of uranium which has been de-extracted in the aqueous phase together with technetium during of this washing to recover this fraction of uranium in the organic phase.

This third step is carried out by circulating in the extractor 3 the aqueous phase coming from the extractor 2 against the flow of an organic phase, also denoted “PO” in FIG. 1, of qualitative and quantitative composition identical to that of the organic phase used for the "Extraction U" step.

Again, the step “Additional extraction U” preferably further comprises acidification of the aqueous phase within extractor 3 to bring its nitric acid concentration to a value at least equal to 3 mol / L, and thus favor the extraction of uranium in the organic phase. This acidification is carried out by adding a very concentrated nitric acid extractor 3, for example in the form of an aqueous solution comprising 5 mol / L or more of nitric acid such as, for example, an aqueous solution containing 7 .5 mol / L nitric acid

As visible in FIG. 1, the organic phase leaving the extractor is re-introduced into one of the last stages of extractor 1 and added to the organic phase circulating in this extractor before the latter enters the 'extractor

2.

he

The fourth step of the process, denoted “Unextraction U” in FIG. 1, consists in de-extracting ruranium (Vî) from the organic phase resulting from the “Washing Te” step to recover it in aqueous phase.

It is carried out by circulating, in extractor 4, the organic phase 5 originating from extractor 2 against the flow of water or of an aqueous nitric solution of very low acidity, that is to say comprising at most 0.05 mol / L of NHO ₃ such as, for example, an aqueous solution comprising 0.01 ml / L of NHO ₃ .

This fourth step may be performed at room temperature {that is to say at 20-25 C ^e), but is preferably performed at a temperature of 40 ° C to 50 "C to accelerate the kinetics of stripping of 'uranium and using an O / A flow ratio greater than 1 so that rurantum (Vi) is de-extracted in a concentrated manner.

At the end of these 4 stages, are obtained:

- two refinins, which correspond to the aqueous phases leaving the extractors 1 and 3 respectively and which comprise for the first (known as “primary raffinate” in FIG. 1), all or almost all of the plutonium and neptunium initially present in the MUR solution, a tiny amount of the technetium initially present in the MUR solution, as well as the non-radioactive impurities (Fe, Ni, Cu, Cr, etc.) initially present in the MUR solution and, for the second (known as "secondary raffinai" on FIG. 1), almost all of the technetium initially present in the MUR solution and, optionally, a minute amount of the plutonium and neptunium initially present in the MUR solution;

the aqueous phase leaving the extractor 4, which comprises uranium in the form of uranyie nitrate, purified with respect to plutonium, neptunium and technetium but also other impurities initially present in the solution WALL ; and

- the organic phase leaving extractor 4, which no longer comprises uranium, plutonium, neptunium, technetium or only in infinitesimal quantities but which may contain degradation products of Λ /, / V-dîaikyiamide which have formed by hydrolysis and radioiyse during the previous steps.

Also, the fifth step, denoted “PO washing” in FIG. 1, consists of regenerating this organic phase by subjecting it to one or more washes with a basic aqueous solution, for example a first washing with an aqueous solution at 0, 3 mol / L of sodium carbonate, followed by a second wash with an aqueous solution of 0.1 mol / L of sodium hydroxide, then with one or more washes with an aqueous solution of nitric acid allowing it to be re-acidified , for example an aqueous solution comprising 2 mol / L of nitric acid, each washing being carried out by circulating said organic phase, in an extractor, against the current of the aqueous washing solution.

As shown in Figure 1, the organic phase thus regenerated can then be returned to extractors 1 and 3 for its reintroduction in a new cycle.

It - EXPERIMENTAL VALIDATION

IL1— Tube tests

A real aqueous solution of passive MUR - hereinafter called "solution

MUR ”- is prepared by dissolving in nitric acid a sample of a potassium diuranate mud representative of an ultimate unburnt fluorination of the passive MUR type and, therefore, contaminated with plutonium, neptunium and technetium . The mud is a mud of medium composition, the KDU sludges having, in fact, a certain variability in their composition.

The composition of this solution is presented in Table I below.

Table I

Concentrations Activities (Bq / gU) {g / L) mol / L u (vp F 155.6 17 239 _pu 19 ²³⁷ Np 16 TC 74 K 32.6 N / A 0.5 It 0.02 Mg 0.01 Fe 0.5 Or 2.7 Cu 1.4 Cr 0.03 no ₃ - 2.3 H ⁺ 1.6

First series of tests;

A sample of the MUR solution is acidified to 3 mol / L of nitric acid then it is subjected to an extraction-washing-extraction procedure in tubes.

Extraction: the sample of the MUR solution is brought into contact, volume by volume, with an organic phase comprising 1.5 mol / L of DEHiBA in IP 185 fane for 10 minutes at 20 ° C. After centrifugation, the organic and aqueous phases are separated from each other.

Washing: the organic phase obtained at the end of the extraction is brought into contact 3 times successively, volume by volume, with an aqueous solution comprising 2 g / L of U (IV), 73 g / L of UfV! ), 0.1 mol / L of aetaldoxime and 1 mol / L of nitric acid for 30 minutes at 45 ° C. After centrifugation, the organic and aqueous phases are separated from each other and the washing is repeated twice, each time renewing the aqueous solution, to simulate 3 stages of washing the organic phase.

Extraction: the organic phase obtained at the end of the third washing is brought into contact, volume by volume, with distilled water for 10 minutes at 45 ° C.

The aqueous phase obtained at the end of this extraction is analyzed to determine its uranium concentration and the activities of plutonium, neptunium and technetium in order to verify whether the specifications of the ASTM C 788 standard are reached.

The results are reported in Table 2 below.

Table II ^{10 * * * * 15 * * * * 20 * * * *}

[U] g / L p ⁹ Pu] Bq / gU t ^ Np] Bq / gU lTc] Bq / gU MUR solution sample 135 16.6 14.0 64.5 acidified to 3 mol / L of NHO ₃ Aqueous phase resulting from the 44 <1 * <4 * <4 * extraction Specifications <5 <1 <50

* Limits of detection

The values of the residual activities measured in ²³⁹ Pu, ⁹⁹ Tc and ²³⁷ Np show that the uranium has been very well decontaminated by the extraction-washing-extraction extraction protocol used in tubes. The specifications for plutonium and technetium (<1 Bq / gU and <50 Bq / gU) have been reached, whereas it is difficult to conclude for neptunium due to an insufficient detection limit.

Second series of tests:

A second series of tests is carried out in order to verify that the high fluoride content (17 g / L, ie 0.55 mol / L) of the MUR solution is not too penalizing to quantitatively extract the uraniumfVl ) and to assess the impact of the concentration of nitric acid in the MUR solution and of the concentration of DEHiBA in the organic phase used to quantitatively extract the uranium fVI) from this solution. Indeed, the purpose of the process being to recover uranium from passive MURs, it is desirable to limit as much as possible the uranium losses during its extraction from the MUR solution.

To do this, samples of the MUR solution, acidified or not acidified to 3 months / L of nitric acid, are brought into contact, volume by volume, with an organic phase comprising either 1.5 mol / L of DEHiBA or 1, 75 mol / L of DEHiBA in IP 1S5 Herbal Tea for 10 minutes at 20 ° C.

After centrifugation, the organic and aqueous phases are separated

Tune each other and the extraction is repeated four to six times, each time renewing the organic phase, and this, to simulate four to six stages of extraction.

The concentration of uranium (VI) is measured in the aqueous phases after each contact with an organic phase.

The results of these measurements are illustrated in FIG. 2 in the form of curves showing revolution of the concentration of uranium in the aqueous phase, noted [U] _sq and expressed in g / L, as a function of the number of contacts, noted N , with the organic phase.

In this figure:

- curve A corresponds to the results obtained for the sample couple of non-acidified MUR solution / organic phase at 1.5 mol / L of DEHiBA;

- curve B corresponds to the results obtained for the sample sample of non-acidified MUR solution / organic phase at 1.75 mol / L of DEHiBA;

- curve C corresponds to the results obtained for the sample couple of acidified MUR solution / organic phase at 1.5 mol / L of DEHiBA; while

- curve D corresponds to the results obtained for the sample couple of acidified MUR solution / organic phase at 1.75 mol / L of DEHiBA.

This figure shows that it is desirable to acidify the MUR solution to 3 mol / L or more of nitric acid (combined effect of the increase in the extracting power of DEHiBA with the concentration of salting out nitrate ions and of the reduction in the complexing effect of fluoride ions vis-à-vis uranyl ions UOs ²⁺ with high acidity) and to increase the concentration of DEHÎBA in the organic phase to 1.75 mol / L if it is desired to limit the losses of uranium during its extraction from the MUR solution.

In fact, the acidification of the MUR solution with 3 mol / L of nitric acid combined with the use of DEHiBA up to 1.75 mol / L in the organic phase makes it possible to obtain less than 7.8 g / L d uranium in the aqueous phase after the fifth contact with the organic phase, which corresponds to recovery of more than 94% of the uranium initially present in the MUR solution in 5 contacts.

If these conditions are not used, the retention of uranium in the aqueous phase will be greater and will require the use of more extraction stages, thereby reducing the compactness of the process of the invention.

H.2 - Pilot test in mixer-settlers

A pilot test in mixer-settlers was carried out in order to validate the feasibility of the process of the invention on an industrial scale.

In this test, the method of the invention is applied to an actual aqueous solution of passive MUR - also referred to below as "MUR solution" previously obtained by dissolving a sample of a diuranate mud in nitric acid. of potassium representative of an ultimate unburnt fluorination of the passive MUR type and, therefore, contaminated with plutonium, neptunium and technetium.

The composition of uranium, radioactive impurities, fluoride ions and protons of this aqueous solution is presented in Table III below.

Ifl array

Concentrations Activities g / L mol / L Bq / gU Bq / L U (VÎ) 155.6 | | F 17.5 I ²³⁹ Pu 1 day 25 3,890 ²³⁷ Np ί ⁹ 1400 ^S9 Tc 1 1 ⁷⁸ 12,137 H i i.ô |

The installation and the operating conditions used for this pilot test are illustrated in FIG. 3.

As can be seen in this figure, the installation includes:

- a first battery with 16 stages of mixer-settlers for the "Extraction U" stage;

- a second battery with 8 stages of mixer-settlers for S the “Washing Te” stage;

- a third battery with 4 stages of mixer-settlers for the step “Additional extraction U”; and

- a fourth battery with 8 stages of mixer-settlers for the step "U extraction".

The organic phases (PO) used in the steps “Extraction U” and “Additional extraction U” comprise 1.75 mol / L of DEHiBA in IP 185 Isane.

An aqueous solution comprising 7.5 months / L of nitric acid is added to the MUR solution circulating in the extractors of the battery dedicated to “Extraction U” to increase the acidity of this solution and thus reduce the complexation of the uranium by the fluoride ions in the aqueous phase but this addition is carried out at 8 ^emo floor of this battery to avoid that the organic phase flowing against the current of URM solution extracts too acid.

An aqueous solution comprising 7.5 mol / L of nitric acid is also added to the aqueous phase circulating in the extractors of the battery dedicated to “Additional extraction U” to, here too, increase the acidity of this phase and thus favor the extraction of uranium from the organic phase. This addition is made on the 4 ^th stage of the battery dedicated to “Additional extraction U”.

In the “Te washing” step, the reducing agent used is uranous nitrate (5 g / L) stabilized by acetaldoxime (0.3 mol / L) in an aqueous solution comprising mol / L of nitric acid. . The temperature of the 8 stages of the battery dedicated to "Te washing" is fixed at 42 ° C to accelerate the kinetics of reduction of Tc (Viî) in reduced Te

Likewise, the temperature of the 8 stages of the battery dedicated to the “U extraction” is fixed at 50 ° C. to accelerate the kinetics of uranium extraction. The O / A ratio is 1.66 in this battery.

The test is carried out continuously for 80 hours including 29 hours with a battery supply dedicated to f “Extraction U” in MUR solution. The good progress of the process is followed by an online analysis of uranium by UV- spectrophotometry. vîstbîe. At the end of the test, the various aqueous and organic phases leaving the extractor batteries are collected and analyzed to assess the performance of the process.

The results of these analyzes are reported in Table fV below.

Table IV fX "J

LO

Be

READ

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61352 FR SL-P

These results show that more than 99.9% of the uranium has been recovered from the MUR solution and that its decontamination with respect to plutonium, neptunium and technetium meets the purity specifications imposed by the AS TM C 788 standard for the recycling of uranium in the manufacture of new nuclear fuels (see Table V below).

TableV

Radioelements Could Np Tc Residual content (Bq / gU) <1 <1 <1.7 Specifications (Bq / gU) <5 <1 <50

S 61352 FR SL-P

Claims (16)

1. Process for the recovery and purification of uranium present in a potassium diuranate mud comprising fluorides, plutonium, neptunium and technetium, which process comprises the following steps: 5 a) a decontamination of uranium with respect to plutonium and neptunium, the decontamination comprising an extraction of uranium, at the oxidation state + VI, from a first aqueous solution resulting from a dissolution of the sludge in nitric acid, by bringing the first aqueous solution into contact with an organic solution comprising at least one lipophilic N, Nd \ aIkylamide As an extractant, in solution in an organic diluent, then separation of the aqueous and organic solutions; b) a decontamination of uranium with respect to technetium, the decontamination comprising a de-extraction of technetium, at the oxidation state + IV, from the organic solution resulting from step a), by contacting, in an extractor, Organic solution with a second aqueous solution comprising nitric acid and at least one reducing agent capable of reducing technetium to technetium which is not extractable by / V, / V-dialkylamide without reducing uranium (VI), then separation of organic and aqueous solutions; and c) recovery of uranium, recovery of uranium 20 comprising a de-extraction of the uranium, at the oxidation state + VI, from the organic solution resulting from step b), by bringing the organic solution into contact with an third aqueous solution, then separation organic and aqueous solutions, whereby an aqueous solution is obtained comprising purified uranium vis-à-vis plutonium, neptunium and technetium. 2. Method according to claim 1, in which the / V, / V-dialkylamide corresponds to the formula (I) below: R ¹ -C (O) -NR ² R ³ (I) in which: S 61352 FR SL-P R ¹ represents a linear or branched alkyl group in alpha or beta of the carbonyl group, comprising from 3 to 12 carbon atoms; and R ² and R ³ represent, independently of one another, a linear or branched alkyl group comprising from 6 to 12 carbon atoms. 3. The method of claim 2, wherein R ² and R ³ each represent a 2-ethylhexyl group. 4. Method according to claim 3, in which N, N- (d \ -210 ethylhexyl) -n-butanamide, le / V, / V-di- (2-ethylhexyl) - / sobutanamide, N, Nd \ - (2ethylhexyl) -2,2-dimethylpropanamide or alternatively / V, / V-di- (2-ethylhexyl) -3,3dimethylbutanamide. 5. The method of claim 4, wherein the / V, / \ / - dialkylamide is / V, / V-di- (2-ethylhexyl) - / sobutanamide. 6. Method according to any one of claims 1 to 5, in which the concentration of / V, / \ / - dialkylamide in the organic diluent is between 1 mol / L and 2 mol / L and, better still, between 1 , 5 mol / L and 1.8 mol / L. 7. Method according to any one of claims 1 to 6, wherein the concentration of nitric acid in the first aqueous solution is equal to or greater than 3 mol / L. 25 8. Method according to any one of claims 1 to 7, in which the reducing agent is uranous nitrate. 9. The method of claim 8, wherein the uranous nitrate is stabilized by acetaldoxime. S 61352 FR SL-P 10. The method of claim 8 or claim 9, wherein the concentration of nitric acid in the second aqueous solution is less than 3 mol / L. 5 11. Method according to any one of claims 1 to 10, in which the extractor of step b) is heated to a temperature between 30 ° C and 50 ° C and, better still, between 40 ° C and 45 ° C. 12. A method according to any one of claims 1 to 11, wherein the third aqueous solution is water or an aqueous solution of nitric acid comprising not more than 0.01 mol / L of nitric acid. 13. Method according to any one of claims 1 to 12, wherein the extractor of step c) is heated to a temperature between 40 ° C and 15 55 ° C. 14. Method according to any one of claims 1 to 13, in bringing the organic and aqueous solutions into contact in the extractor of step c) comprises a circulation of the organic and aqueous solutions with a ratio of the flow rate of 20 the organic solution at a flow rate of the aqueous solution greater than 1. 15. Method according to any one of claims 1 to 14, which further comprises: - contacting in an extractor the aqueous solution from step b) with an organic solution identical to the organic solution used to extract the uranium in step a), then separation of the aqueous and organic solutions; and - Introducing the organic solution thus separated into the extractor of step a). S 61352 FRSL-P 16. Method according to any one of claims 1 to 15, which further comprises a regeneration of the organic solution resulting from step c). S.61352 1/3