OA20559A - Process for stripping uranium contained in a solvent. - Google Patents

Abstract

The present invention relates to a process for the de-extraction, in other words the elution, of uranium contained in an organic phase in the presence of an aqueous solution comprising a mixture of carbonate ions and hydroxide ions.

Description

PROCESS FOR DE-EXTRACTION OF URANIUM CONTAINED IN A SOLVENT

DESCRIPTION

The present invention relates to a process for back-extraction (or “stripping” according to Anglo-Saxon terminology), in other words elution, of the uranium contained in an organic phase. In particular, the process is useful for the treatment of organic phases resulting from a solvent extraction process for uranium liquors resulting from the acid leaching of uranium ores, the solvent extraction process using a compound extracting amine type uranium.

State of the art

Liquid-liquid extraction techniques are well known and have long been used for the recovery of uranium from uranium ores. In a typical extraction process for the recovery of uranium, the uranium ore is subjected to a leaching stage during which the ore is brought into contact with a mineral acid, generally sulfuric acid, so as to pass in solution the uranium and recover a leach liquor charged with uranium (in the +VI oxidation state). The leaching liquor is then treated in a so-called extraction stage which is operated in such a way as to recover the uranium as selectively as possible.

Three main extraction processes are currently used for this purpose: ion exchange, ion exchange with a resin in the pulp and solvent extraction. This latter process quickly conquered the industry since its implementation in the 1950s due to its high yield and the high purity of the uranium produced.

The liquid/liquid solvent extraction process is based on the property possessed by certain extradants, which are compounds immiscible with water, to form soluble complexes with uranium salts. When the leaching solution rich in uranium (+VI), preferably after a clarification step, is brought into contact with the organic solvent, the uranium is partitioned between the aqueous phase and the organic phase. Under appropriate operating conditions, it is thus possible to extract the uranium almost quantitatively in the organic phase while most of the other constituents of the leaching liquor remain in the aqueous phase. After having mixed the aqueous and organic phases to ensure contact, the mixture of the solutions is introduced into a device, for example into a settling tank, to separate the lighter organic phase charged with uranium from the aqueous phase.

The next step in the extraction process consists in selectively recovering the uranium contained in the organic phase. According to a first method, the uranium is directly precipitated from the organic phase with an aqueous liquid containing hydroxide ions. The uranium precipitated in the form of diuranate is separated and then treated by methods well known in the industry.

Another method consists in depleting the uranium (+VI) from the organic phase by bringing the organic phase into contact with an aqueous phase having a composition which modifies the partition coefficient of the uranium (+VI) in favor of the phase watery. It is thus possible to exhaust uranium, from certain organic solutions (for example tributyl phosphate), with essentially pure water. It is also possible to use other organic solutions such as amines but which require the presence of an anion (such as nitrates, chlorides or carbonates) in the aqueous phase to break the uranium-organic compound complex. In this method, the uranium which has been passed into the aqueous phase is then recovered by precipitation, for example in the form of diuranate, by providing a source of hydroxide ions in the aqueous liquid loaded with uranium.

Document US 2,877,250 is known in the state of the art, which describes a process for recovering uranium from a leaching liquor containing sulphate ions by bringing it into contact with an organic phase containing an amine-type compound as an extractant. . The organic phase loaded with uranium is then depleted of uranium by bringing it into contact with an aqueous solution containing carbonate ions so as to produce an aqueous solution containing uranium in solution in the form of uranyl tricarbonate UO ₂ (CO3) ₃ ⁴ -.

Without being bound by any theory, the following equations make it possible to illustrate the chemical reactions which are involved in a process for the treatment of a sulfated acid leach liquor by solvent extraction using a tertiary amine as an extractant. On contact with the H ⁺ ions present in the leaching liquor, the amine extractant R ₃ N of the organic phase protonates to form a complex with the sulphate ions according to the reaction (equation 1):

2(R ₃ NH) ⁺ + SO4 ² ' ----------► (R3NH) ₂ SO4

This complex is capable of fixing uranyl ions by complexation.

A property of uranium (+VI) which is used in the treatment is its ability to form sulphate complexes, in particular uranyl trisulifate, when the medium is loaded with sulphate ions (typically from 5g/L of SO?').

Uranyl trisulfate ions (302(804)3 ^4- react with amines to form a complex [amine-uranyl trisulfate] according to the reaction (equation 2):

2(R ₃ NH) ₂ SO4+UO ₂ (SO4) ₃ ⁴ - ----------► (R ₃ NH)4UO ₂ (SO4)3 + 2 8Ο ₄ ² '

The uranium is thus transferred from the sulphated aqueous phase into the organic phase.

After a stage of separation of the organic and aqueous phases, the organic phase loaded with uranium is brought into contact with an aqueous back-extraction solution comprising carbonate ions, for example a solution of sodium or ammonium carbonate, capable of complexing the uranium so as to transfer it into the aqueous back-extraction phase (or solution).

In theory, this back-extraction step is based on a mechanism of deprotonation of the amine which causes the latter to lose its ability to complex uranium and on the ability of uranyl ions to form complexes with carbonate ions, according to the reaction (equation 3):

(R ₃ NH) ₄ UO ₂ (SO ₄ )3 + 7CO ₃ ² - --------► 4R3N + υθ2(ΟΟ3)3 ⁴ + 4 HCO3- + 3 SO4 ² In the reaction mechanism described above above, the back-extraction of uranium (in the +VI oxidation state) from the organic phase involves an excess of carbonate ions relative to the uranium. According to equation 3, seven moles of carbonate ions per mole of uranium are required, of which four moles of carbonate ions per mole of uranium are involved to deprotonate the amine and three moles to complex the uranium under form of uranyl tricarbonate. Note that the carbonate ion complexation reaction produces four moles of bicarbonate (also known as hydrogen carbonate) per mole of uranium.

The aqueous phase enriched in uranium which is collected after the phase separation step is then treated in a uranium precipitation reaction, for example in the form of sodium diuranate Na ₂ U2O7 (SDU) by means of a solution of sodium hydroxide. At the end of the precipitation step, the suspension of sodium diuranate is subjected to a solid/liquid separation which produces a recoverable uranium precipitate (also called "Yellow Cake" according to the Anglo-Saxon terminology) and a liquor carbonated.

In order to limit the consumption of hydroxide ions involved in the uranium precipitation step, it is recommended to proceed, upstream of this step, to a step of elimination of part of the bicarbonate ions contained in the eluate. stripping aqueous solution. The partial elimination of bicarbonates consists, for example, in neutralizing the bicarbonate ions using dilute sulfuric acid and/or heating the solution to degas the carbon dioxide produced. However, this method has the disadvantage of generating a greenhouse gas effluent.

An object of the invention is to provide a process for recovering the uranium present in an organic phase loaded with uranium which generates less gaseous effluent loaded with carbon dioxide and less liquid effluent to be treated, which consumes less of carbonate and therefore also more economical.

Summary of the invention

In the context of the present invention, the terms “aqueous solution” and “aqueous phase” are equivalent and interchangeable, just as the terms “organic solution”, “organic phase” and “solvent” are equivalent and interchangeable.

The invention therefore relates to a process for the back-extraction of uranium present in an organic phase containing an amine compound or a mixture of amine compounds of formula R ₃ N, R being chosen from hydrogen and a C6 to optionally substituted C18 and with at least one R group being an optionally substituted C6 to C18 alkyl group, the amine compound being capable of forming a complex with uranyl sulphate, the method comprising the following steps:

a) the organic phase is brought into contact in a contacting reactor with an aqueous uranium back-extraction solution comprising carbonate ions so as to transfer the uranium into the aqueous solution by complexation, the pH of the aqueous solution back-extraction being chosen so that the uranium is in soluble form;

b) a liquid/liquid separation step is carried out, whereby an organic phase depleted in uranium and an aqueous back-extraction solution loaded with uranium in the form of uranyl tricarbonate are obtained.

The method is characterized in that the aqueous back-extraction solution also contains hydroxide ions which are added to this solution.

The process according to the invention therefore differs from known processes in that the uranium back-extraction step is carried out with an aqueous back-extraction solution (also called stripping solution) comprising not only carbonate ions useful for ensuring the complexation of uranium (+VI) but also of hydroxide ions which have the role of partly converting the bicarbonate ions into carbonates; bicarbonate ions from the deprotonation reaction of the complex, for example [uranyl aminetrisulfate]. This in situ conversion of the bicarbonate ions makes it possible to limit the quantity of bicarbonate present in the aqueous back-extraction solution (exhaustion solution) loaded with uranium collected at the end of stage b) of separation and, consequently, to reduce the amount of carbon dioxide that is degassed before proceeding with the precipitation of uranium. Thanks to this regeneration of carbonate ions, the process also has the advantage of reducing the additional carbonates necessary for the treatment of the organic phase and therefore contributes to a reduction in operating costs.

According to the invention, the complex [amine-uranyl sulphate] includes in particular the complexes [amine-uranyl bisulphate] and [amine-uranyl trisulphate]. However, it should be noted that the process according to the invention is particularly suitable for treating a solvent containing the [amine-uranyl trisulfate] complex.

The method according to the invention may also comprise the following steps:

c) the aqueous solution laden with uranium resulting from step b) is brought into contact with an aqueous solution containing hydroxide ions so as to produce a suspension containing a uranium precipitate;

d) a solid/liquid separation is carried out on the aqueous suspension, whereby a solid phase containing the uranium precipitate and a carbonated liquid effluent is obtained.

Advantageously, the process according to the invention comprises a step e) in which all or part of the liquid effluent from step d) is recycled to step a). Thus the aqueous effluent resulting from the separation of the uranium concentrate which contains carbonate ions is advantageously reused in the back-extraction stage. This partial recycling thus contributes to limiting the consumption of carbonates in the process.

The process may also have one or more of the characteristics below considered individually or in all technically possible combinations:

According to a preferred embodiment, step a) of back-extraction is carried out by replacing a fraction of the carbonate ions with hydroxide ions. When step a) involves a total of n moles of carbonate ions per mole of uranium and of which m moles of carbonate ions per mole of uranium are useful for the deprotonation of the complex, e.g. the complex [amine-trisulfate uranyl], the stripping solution may contain between (0.25*m) moles and (0.75*m) moles of hydroxide ions and between (n-(0.25*m)) moles and (n -(0.75*m)) moles of carbonate ions.

Thus, according to the invention, the back-extraction solution involved may comprise 1 equivalent to (m-1) equivalents of hydroxide ions and (n-1) equivalents to (n-m+1) equivalents of carbonate ions. This way of operating makes it possible to control the pH in order to avoid any excessive increase which would result in direct precipitation of uranium in the back-extraction solution and/or at the organic phase and aqueous phase interface.

Care will preferably be taken to ensure that the pH of the aqueous uranium stripping solution (eluate) resulting from step b) is between 7 and 10, preferably between 7.3 and 8.

The process according to the invention is in particular applicable for the treatment of an organic phase loaded with uranium in which, for step a), a reaction mechanism involves a total of 7 moles of carbonate ions per mole of uranium (n=7 ) and of which 4 moles of carbonate ions per mole of uranium (m = 4) are involved in the deprotonation of the [amine-uranyl trisulfate] complex. According to a very preferred embodiment, 1 equivalent to 3 equivalents of carbonate ions are replaced by 1 equivalent to 3 equivalents of hydroxide ions, that is to say that in step a) a total of 6 , 5 or 4 equivalents of carbonate ions and respectively 1.2 or 3 equivalents of hydroxide ions.

In the context of the invention, the hydroxide ions are added to step a) by an aqueous solution of hydroxide of an alkali metal or of ammonium hydroxide.

The injection of the aqueous solution containing the hydroxide ions into the contacting reactor can be carried out via a dedicated injection line independent of the injection line of the aqueous carbonate solution or else the two solutions are mixed beforehand and said mixture is then added to the contacting reactor.

The water-immiscible amine or amine mixture of the organic phase is a C8 to C18 aliphatic primary, secondary or tertiary amine, preferably C8 to C12, the aliphatic chain of which is optionally substituted. Preferably, the tertiary amine is chosen from trioctylamine, tridodecylamine. The amine compound can also be a primary amine such as tridecylamine or a secondary amine such as didodecylamine, dioctadecylamine and dihexylamine. In particular, a mixture of Alamine® 336 or Adogen® 364 amines can be used.

Preferably, the organic phase further comprises a branched or linear hydrocarbon diluent immiscible with water. The diluent can be chosen from aliphatic or aromatic hydrocarbons such as for example kerosene, benzene, toluene, naphtha. The organic phase may also include a phase modifying agent, such as isotridecanol, to facilitate phase separation.

The process according to the invention can also include a step b′) in which part of the bicarbonate ions is removed from the aqueous de-extraction solution charged with uranium (resulting from step b) of separation of the organic and aqueous phases) by a degassing in the form of CO2.

Step a) of bringing the aqueous and organic phases into contact is carried out in mixer-settlers or stirred or pulsed columns in a continuous or discontinuous mode with co-current or counter-current circulation of the aqueous and organic phases. Preferably step a) implements several mixer-settlers in series, for example 3 or 4 mixer-settlers with counter-current circulation of the aqueous and organic phases.

In one embodiment, the supply of the solution containing hydroxide ions is carried out at the level of the first back-extraction stage, for example at the level of the mixer-settler into which the organic phase charged with uranium is introduced and from which is extracted the uraniferous eluate (aqueous phase). According to this embodiment, which could be described as mixed counter-current and co-current circulation of the carbonate solution, a fraction of the carbonate solution necessary for the back-extraction is introduced at the level of the first back-extraction stage ( e.g. in the first mixer-settler) concomitantly with the addition of the solution containing hydroxide ions while the complementary fraction of the carbonate solution is sent to the last extraction stage (e.g. in the last mixer-settler). This embodiment has the advantage of making it easier to maintain the pH of the aqueous phase at a value greater than 7 in order to avoid precipitation of uranium at the organic phase/aqueous phase interface, but also to limit the consumption of carbonates.

The process according to the invention is capable of providing a uranium precipitate in the form of sodium diuranate when the precipitation step c) uses a base such as sodium hydroxide.

In one embodiment, the hydroxide ions are produced by causticizing part of the carbonated liquid effluent from step d) and said sodium hydroxide solution is sent to step a).

Description of embodiments

Other characteristics and advantages of the invention will emerge from the detailed description given below, by way of indication and in no way limiting, with reference to the appended figure:

• Figure 1 is a block diagram representing the uranium back-extraction process according to the invention;

• Figure 2 is a block diagram representing a conventionally implemented configuration of a uranium back-extraction unit;

• Figures 3 and 4 are block diagrams illustrating configurations for implementing the method according to the invention.

In what follows, the terms “solvent” and “organic phase” are equivalent and the same applies to the terms “aqueous solution” and “aqueous phase”.

The process shown in Figure 1 is intended for the treatment of an organic phase loaded with uranium with a view to re-extracting the uranium it contains.

The organic phase is typically a product from a solvent extraction unit which treats an acid and sulphate leaching liquor, resulting from the attack of a uranium ore by sulfuric acid. The extraction unit which manages this type of leachate uses an organic solvent comprising as an extractant molecule a compound of amine type, immiscible with water, of formula R3N in which the R group is chosen from hydrogen or a C6 to C18 alkyl radical, it being understood that at least one R group is an alkyl group.

Preferably, the extractant is an aliphatic tertiary amine R3N in which R is chosen from alkyl radicals, optionally substituted, having from 6 to 18 carbon atoms (C6 to 018) and preferably having from 9 to 12 carbon atoms. In a concentrated sulfuric acid medium, these amines have a very strong affinity for UO2(SO4)3 ^3- ions and weak for SO ₄ ² ' and HSO4· ions. Mention may be made, for example, as compound of amine type, of trinormaloctylamine, triisooctylamine and trilaurylamine and their mixture.

The solvent may comprise, where appropriate, an organic diluent in which the amine is dissolved in order in particular to reduce its viscosity and thus facilitate its use. Among the diluents used, mention may be made of aliphatic or aromatic hydrocarbons such as kerosene, benzene, toluene, naphtha. The preferred diluent is kerosene. The concentration of the extractant (R ₃ N amine) in the diluent can vary within wide limits of between 0.05 mol/L and the pure extractant. From a practical point of view, it is preferred to use an organic solution containing between 0.1 and 0.2 mol/L of extractant.

The uraniferous aqueous solution (or leachate) is brought into contact with the extraction solvent, discontinuously or continuously, in co-current or counter-current circulation with the aqueous and organic phases. The ratio of aqueous phase to organic phase is chosen so that the removal of uranium is most effective. Ratios of the organic phase to the aqueous phase of 1:10 to 2:1 are suitable to achieve the desired extraction efficiency, which is at least 99%. From a practical point of view, the phase ratio is adjusted so that the uranium content in the deuranized solvent is less than or equal to 1 ppm.

The contacting of the phases is usually carried out in devices called mixer-settlers although there are many other types of suitable devices such as, for example, stirred columns or pulsed columns.

In a "mixer-settler" type system, one phase is dispersed in the other by stirring or by another suitable form of agitation. The extraction solvent then forms a complex with the uranium which passes into the organic phase. The dispersion of the two phases then flows into the decanter where phase separation occurs on standing. Generally, the extraction is carried out between 20 and 60° C., preferably between 20 and

45°C and with a volume ratio of the organic phase to the aqueous phase (O/A ratio) generally between 1/10 and 2.

The organic phase which is then loaded with uranium extracted from the uraniferous acid leaching solution is then treated by contact with an aqueous back-extraction solution containing an entraining agent (extradient) capable of carrying out the transfer of uranium from said organic phase. in the aqueous back-extraction phase.

According to the invention, the aqueous back-extraction phase is a solution of sodium or ammonium carbonate. Preferably, a sodium carbonate solution is used compared to ammonium carbonate because it makes it possible to provide a completely soluble back-extraction concentrate solution with a much higher concentration of uranium.

Referring to Figure 1, the uraniferous organic phase 1 is sent to a back-extraction unit 2 which may comprise a plurality of batteries of mixer-settlers in series, in which it is brought into contact with an aqueous solution of sodium carbonate which is provided by line 3.

According to the invention, the back-extraction step also involves hydroxide ions which are supplied via line 4 in the form of an aqueous solution of hydroxide of an alkali metal, preferably a solution of sodium hydroxide . In the context of the invention, it is entirely possible to mix the carbonated and hydroxide solutions of an alkali metal in line before sending said mixture to the back-extraction unit 2.

The process according to the invention is based on the principle according to which the carbonate ions participate in the back-extraction of the complexed uranium on the one hand by reacting with the complex, in particular with the [amine-uranyl trisulfate] complex, by deprotonating the amine to release the uranyl trisulfate and, on the other hand, by complexing the uranyl (UO2 ² *) in the form of uranyl tricarbonate, which is soluble in an aqueous medium. In general, the reaction mechanism involves a total of n moles of carbonate ions per mole of uranium, of which m moles of carbonate ions are involved in the deprotonation of the amine and (π-m) moles per mole of uranium participate in the complexation of uranium. For example the numbers n and m are respectively equal to 7 and 4.

This back-extraction step generates bicarbonate ions (see equation 3) which, according to the invention, are partly converted by the hydroxide ions present in the back-extraction solution, thus leading to in situ regeneration of the carbonate ions; that is to say within the stripping unit 2 itself. The regenerated carbonate ions are available to participate again in the uranium stripping mechanism. The process thus responds to the problem of providing a process that uses less carbonate and emits less carbon dioxide into the atmosphere while maintaining a high uranium back-extraction yield.

According to a preferred embodiment of the invention, the quantity of hydroxide ions which is introduced with the carbonated back-extraction solution serves to replace a fraction of the carbonate ions which take part in the back-extraction process. Preferably, the number of moles of hydroxide ions is between 25% and 75% of the number of moles m of carbonate ions S per mole of uranium which are theoretically involved in the deprotonation reaction of the amine of the complex [amine -uranyl sulphate] (for example the complex [amine-uranyl bisulphate] or [amine-uranyl trisulphate]. This quantity is chosen in particular to maintain the buffering capacity of the aoido-basic carbonate/bicarbonate couple so as to guarantee a pH of the back-extraction solution at a value between 7 and 10 11 in order to prevent the precipitation of uranium in the aqueous phase.

By way of illustration, when step a) of back-extraction involves a total of 7 moles of carbonate ion per mole of complex [amine-uranyl trisulfate] and of which 4 moles are involved in the deprotonation reaction of the complex [amine-uranyl trisulfate], it is possible within the scope of the invention to replace 1, 2 or 3 moles of carbonate ions by respectively 1, 15, 2 or 3 moles of hydroxide ions to achieve partial regeneration of the carbonate ions by converting the bicarbonate ions produced during the deprotonation of the amine from the [amine-uranyl trisulfate] complex.

Step a) of bringing the aqueous and organic phases into contact can be carried out discontinuously, continuously in co-current or preferably continuously in counter-current.

Preferably, the operation is carried out in a continuous mode with counter-current circulation of the aqueous and organic phases in a back-extraction unit 2 comprising a battery of “mixer/settler” type devices arranged in series, each mixer/settler providing the equivalent of a contact stage. In a mixer, the aqueous phase is dispersed in the organic phase so as to promote the transfer of uranium into the carbonated aqueous solution. The dispersion then flows into a decanter where phase separation takes place.

Generally, step a) is carried out at a temperature between 20 and 60°C.

In general, the volume ratio (O/A) of the organic phase to the aqueous phase is generally between 4 and 10, but is not limited to this range. Preferably, the operation is carried out with a ratio (O/A) of between 5 and 8.

As indicated in FIG. 1, an aqueous phase 5 (also called by the term “eluate”) enriched in uranium (+VI) (which is in the form of a uranyl bicarbonate complex ) and an organic phase 6 depleted in uranium which can be recycled to the uranium extraction step, optionally after treatment.

Referring to Figure 1, the aqueous eluate 5 loaded with uranium and whose pH is generally between 7 and 10 (preferably between 7.3 and 8), is sent to the next step of precipitation of the uranium, for example in the form of an alkali metal diuranate, preferably in the form of sodium diuranate.

To this end, the eluate 5 is brought into contact in a precipitation unit 7 with a precipitation agent provided by line 8. The precipitation agent is generally a solution of hydroxide of an alkali metal, preferably a sodium or potassium hydroxide solution. In a preferred embodiment, a solution of sodium hydroxide will be used.

This diuranate precipitation step is carried out in a stirred reactor well known to those skilled in the art with an addition of a quantity of hydroxide ions sufficient to reach a pH of between 11 and 13. Preferably, a solution of hydroxide as concentrated as possible in order to limit the phenomenon of dilution.

The reactions that take place within the uranium precipitation reactor are the uranyl tricarbonate decomplexation reaction and the sodium diuranate precipitation according to equation 4:

UO ₂ (CO ₃ ) ₃ ⁴ ' + 6 OH- + 2 Na ⁺ ------► Na ₂ U ₂ O ₇ + 3 H ₂ O + 6 CO ₃ ^a '

As indicated in FIG. 1, in order to reduce the consumption of hydroxide ions at the precipitation stage, it is advantageous upstream of the precipitation stage of the uranium concentrate to pretreat the eluate 5 in a conversion unit 9 hydrogen carbonate ions. This operation consists in particular in heating the uranium eluate in order to promote the following reaction (equation 5):

HCO ₃ - ______► CO ₃ ² - + CO ₂ , H ₂ O which thus makes it possible to convert the bicarbonate ions into CO ₂ and also to generate carbonate ions. At the end of this treatment, a partially decarbonated uranium eluate 5′ is thus obtained which serves as feed in the precipitation unit 7.

It will be noted that even if bicarbonate ions are converted into carbon dioxide at this stage, the quantity of CO ₂ released is however less compared to the method of the prior art in which regeneration is not carried out. in situ carbonate from bicarbonate.

The aqueous suspension of sodium diuranate which is extracted from the precipitation unit 7 via line 10 is sent to a solid/liquid separation device 11, from which a wet cake 12 of sodium diuranate and a liquid effluent 13 essentially containing sodium carbonate. In the context of the invention, any solid/liquid separation technique can be used to recover the solid from the uranium concentrate, such as filtration or centrifugation.

The liquid effluent 13 which is a solution containing carbonate ions is advantageously recycled via a line 14 to the back-extraction unit 2. The recycling of the liquid effluent can be partial in order to limit the quantity of undesirable metals back-extracted with the uranium (such as molybdenum and zirconium) that are returned to the stripping stage. There is thus provided a line 15 connected to line 14 allowing part of the liquid effluent to be purged. Generally, the liquid effluent recycling rate (recycled volume/total volume) is between 30 and 50%.

According to one embodiment, the carbonated liquid effluent 15 which is not recycled in the back-extraction unit is advantageously used for the production of soda in a so-called caustification unit 16. In such a caustification unit 16, the solution containing carbonate ions are treated in the presence of a solution of calcium hydroxide (or milk of lime) whereby a suspension is obtained comprising solid calcium carbonate and sodium hydroxide.

After a solid/liquid separation, a calcium carbonate cake 17 is recovered on the one hand and a sodium hydroxide lye on the other hand which can be recycled in the back-extraction unit 2 via lines 18 and 4.

Example 1 (comparative)

Figure 2 shows the configuration of a back-extraction unit conventionally adopted for the back-extraction of uranium by bringing an aqueous solution of sodium carbonate into contact, countercurrently, with a solvent loaded with uranium.

The stripping unit comprises three mixer-settlers 20, 21 and 22 arranged in series. The uranium-laden solvent feeds the first mixer/settler 20 via line 23 while the aqueous phase for back-extraction of uranium is supplied to the last mixer/settler 22 via line 27 and circulates counter-current to the organic phase. to the mixer-settlers 21 then 20. The organic phase depleted in uranium is drawn off from the mixer/settler 22 via line 26 while the aqueous eluate loaded with uranium is taken from the mixer/settler 20 via line 30.

The uranium-laden solvent used in this example comes from a process for the liquid/liquid extraction of leachate from a uranium ore from the Somaïr mining site in Niger, using four mixer-settlers in series with circulation at against the current of the organic phase and the leachate and with a ratio (O/A) of 0.25 (vol/vol). The organic solvent used is a mixture comprising 8% v/v Alamine 336, 7% v/v Isodecanol, diluted in an Isane IP 185® diluent (mixture of C11 to C13 isoparaffins).

The uranium-laden solvent is further washed with water with a ratio (O/A) of 2 (vol/vol) at 40°C.

Table 1 gives the composition of the uranium solvent after washing with water:

Element Content (mg/kg) U 8030 Fe < 1 Mo 558 You < 1 V < 1 Zr 153

Table 1

In each of the back-extraction stages, 20 mL of aqueous and organic phases are sampled, filtered and analyzed by atomic emission spectrometry - inductively coupled plasma (ICP-AES according to the English terminology). For the analysis of the organic phases, a filtration is first carried out on a “Phase Separator Whatman® 1PS” filter, which makes it possible to eliminate any traces of water possibly co-extracted. The solvents are mineralized, taken up in HNO ₃ medium, then analyzed by ICP-AES

The back-extraction unit is operated in accordance with the prior art, that is to say by using as aqueous back-extraction solution a solution of sodium carbonate at 150 g/L, which is injected into the mixer-settler 27.

The flow rates of the aqueous and organic phases are set at 0.065 L/h and 0.4 L/h respectively so as to impose an O/A ratio of around 6 (vol/vol) at 40°C and with an aqueous phase flow between mixers/decanters of 0.2 L/h.

Table 2 groups the compositions of aqueous and organic phases which are collected respectively by lines 30 and 26:

Element Aqueous phase (line 30) Organic phase (line 26) U 44g/L 6.6mg/kg Fe 267 pg/L 1.2mg/kg

Mo 3.4g/L < 1mg/kg You 8.1mg/L < 1mg/kg V 4.5mg/L < 1mg/kg Zr 757mg/L 2.1mg/kg

Table 2

It can be determined from the analyzes of the incoming and outgoing flows that the uranium back-extraction yield is greater than 99.9% for a recovery rate of approximately 2.5 g of uranium/h.

Example 2 (according to the invention)

The test was carried out with the same organic phase charged with uranium as Example 1, the composition of which is given in Table 1.

The back-extraction unit implemented in this example according to the invention is represented in FIG. 3 which differs from that of FIG. 2 in that it comprises a line 31 intended to introduce a soda solution, which is connected to the mixer-settler 20 into which the organic phase charged with uranium is injected.

In this example, the quantity of OH ions introduced is adjusted so as to replace one equivalent of carbonate ions out of the seven equivalents normally involved in the back-extraction process.

A single supply of aqueous sodium carbonate solution at 150 g/L is used to supply the back-extraction solution to the mixer-settler 22 (via line 27). The back-extraction solution thus circulates against the current of the organic phase which is introduced into the mixer-settler 20 via line 23.

The sodium carbonate feed rate is gradually reduced, in order to replace the CO ₃ ^2- ions with OH ^- ions supplied via line 31. The addition of sodium hydroxide is thus carried out at the level of the first stage where the ions carbonates are mainly consumed and transformed into hydrogen carbonate ions.

In the stationary state of the system, the soda solution at 58 g/L is sent to the mixer/settler 20 at a flow rate of approximately 9 mL/h while the sodium carbonate solution at 150 g/L is injected in the mixer/settler 22 at a flow rate of approximately 56 mL/h so as to maintain an O/A phase ratio equal to 6 (vol/vol).

The acid-base potentiometric titrations carried out on the eluates demonstrate the absence of free OH ^- ions, confirming that these are quantitatively consumed to regenerate the CO ₃ ² ' ions in situ.

Uranium analysis by ICP-AES of the uranium-laden eluate withdrawn from mixer-settler 20 indicates a content of 45.4 g/L of uranium.

Thus, the substitution of an equivalent of carbonate ions by an equivalent of hydroxide ions makes it possible to achieve a back-extraction yield of uranium greater than 99.9%.

By comparing the experimental results of examples 1 and 2, it can be seen that the uranium content in the eluate remains identical, which makes it possible to conclude that the partial replacement of the carbonate ions by hydroxide ions ensures the same back-extraction performance of the uranium compared to using a carbonate solution alone. Moreover, it is observed that the disengagement of the organic and aqueous phases is not modified because the aqueous phases remain clear. Finally, the acid-base assays carried out on the eluates demonstrated the absence of residual hydroxide ions, thus confirming the total consumption of OH by the hydrogen carbonate ions.

Example 3 (according to the invention)

The back-extraction unit used for Example 3 is shown in Figure 4, it differs from that of Figure 3 in that it includes a line 32 for supplying a solution of sodium carbonate and hydroxide of sodium in the mixer-settler 20 into which the uraniferous organic phase to be treated is introduced.

In the configuration tested, two supplies of aqueous solutions of sodium carbonate are put in place; a first injection (via line 32) of a carbonate solution with a Na ₂ CO ₃ concentration of 180 g/L is carried out in the mixer/settler 20 and a second injection (line 27) is implemented in the mixer/ decanter 22 with a Na ₂ CO ₃ solution at a concentration of 70 g/L. Thus, the feeding of the Na ₂ CO ₃ solutions at 70 g/L and at 180 g/L are done counter-current and co-current with the solvent respectively.

It should be noted that in this configuration the ratios of the organic and aqueous phases are different within the stripping mixers/settlers. Thus for the mixers/settlers 21 and 22, the phase ratio O/A is 10 (imposed by the flow rate of the Na ₂ CO3 solution at 70 g/L), whereas within the mixer/settler 20 the O/A phase ratio is 5 which is imposed by the sum of the two feed rates of the sodium carbonate solutions at 70 g/L and at 180 g/L.

Initially, the back-extraction unit is operated at a temperature of 35°C without injection of OH ^- ions and with flow rates for sodium carbonate solutions at 180 g/L (sent to mixer/settler 20) and at 70 g/L (sent to mixer/settler 22) which are 47 mL/h and 40 mL/h respectively.

In the configuration of Example 3, two modes of partial substitution of COs ^2- ions by OH' ions were tested, namely by replacing one equivalent then two equivalents of carbonate ions by OH ^- ions. To this end, a solution of a mixture of sodium hydroxide and sodium carbonate respectively at 10 g/L and 153 g/L is introduced via line 32 into the mixer-settler 20 where the carbonate ions are mainly consumed. To reduce the number of CO3 ² · ion equivalents and compensate for them with OH' ions, the concentration of the sodium carbonate solution initially set at 180 g/L is reduced while at the same time the concentration of the sodium carbonate solution is increased in the mixture which is injected into the mixer/settler 20.

It should be noted that in order to prevent uranium precipitation at the organic phase/aqueous phase interface, the pH of the aqueous phase was maintained above 7.

o Table 3 below groups together the operating conditions applied in steady state for the configuration described in figure 4.

Na ₂ C0s /NaOH (5 equivalents/2 equivalents) NasCOs /NaOH (6 equivalents/1 equivalent) NasCOs (7 equivalents) Test 4.6 4.6 4.6 O/A ratio in mixer/settler 20 O 10 O O/A ratio in mixer/settler 22 I ______ 130 g/L Na ₂ CO ₃ / 20 g/L NaOH 153 g/L Na ₂ CO ₃ / 10 g/L NaOH i 180 g/L Concentration of the NasCOa/NaOH solution sent to the mixer/settler 20 47mL/hr 47mL/hr 47mL/hr Flow rate of the Na ₂ CO3/NaOH solution sent to the mixer/settler 20 70g/L 70g/L O Concentration of the Na ₂ CO3 solution sent to the mixer/settler 22 I! 40mL/h 40mL/h 40mL/h Flow rate of the Na^COs solution sent to the mixer/settler 22 M 7-7.8 7-7.6 pH of the eluate extracted from the mixer/settler 20

The different fluxes from the tests with and without partial substitution of carbonate ions by hydroxide ions were collected and their uranium contents, determined by ICPAES and according to the Davies & Gray method on the uranium-concentrated eluates, are given in table 4.

Configuration tested Uranium content in the organic phase loaded with uranium (line 23) Uranium content in the eluate (line 30) Uranium content in the organic phase mixer/settler outlet 20 (line 24) Uranium content of the organic phase mixer/settler outlet 21 (line 25) Uranium content of the organic phase mixer/settler outlet 22 (line 26) NaiCOa (7 equivalents) 8.96g/L 37.1g/L 3.8g/L 172mg/L < 5mg/L Na^COs /NaOH (θ equivalents/1 equivalent) 8.96g/L 41.3g/L 6.48g/L 302mg/L < 5mg/L NasCOs /NaOH (5 equivalents/2 equivalents) 8.96g/L 39.2g/L 5.3g/L 754mg/L < 5mg/L

Table 4

It is thus found that the uranium content of the eluates recovered after back-extraction in the presence of carbonate ions and hydroxide ions exhibit a value comparable to that of an eluate collected after back-extraction of the solvent with a carbonate solution alone. Uranium recovery yields are around 99.9%. These tests confirm that it is possible to conduct uranium back-extraction with in situ regeneration of COa ^2- ions by OH ^- without harming the efficiency of uranium recovery.

Example 4

This example makes it possible to check whether the quality of the eluates generated in example 4 makes it possible to recover uranium in solid form and in particular in the form of sodium diuranate.

To this end, we proceed as follows:

a) the concentrated uranium eluate is brought into contact with a 150 g/L sodium hydroxide solution for 6 hours at 70°C in order to precipitate sodium diuranate. The suspension after a flocculation step is filtered in order to recover a sodium diuranate cake;

b) the cake is resuspended in a sulfuric acid solution at a pH equal to approximately 3 - 3.5 so as to solubilize the uranium while maintaining the zirconium in the form of a colloidal precipitate;

c) the acidified suspension is then filtered to recover a uranium filtrate;

d) the filtrate is brought back into contact with a 150 g/L sodium hydroxide solution for 6 h at 70° C. in order to precipitate sodium diuranate.

e) the sodium diuranate suspension is filtered to separate the precipitate which is then dried in order to provide Yellow Cake.

Table 5 lists the uranium contents of the precipitate after drying.

Eiuat retrieved from Example 3 U content in dried Yellow Cake (% mass) Stripping with a solution NazCOs alone 68.7 Back-extraction with a Na ₂ CO ₃ /NaOH solution (6 equivalents / 1 equivalent) 68.4 Back-extraction with a Na ₂ CO ₃ /NaOH solution (5 equivalents! 2 equivalents) 67.5

Table 5

It is observed that the contents of Yellow Cake are almost identical. Moreover, no difference was noted concerning the precipitation kinetics insofar as all the precipitation reactions are quantitative for the imposed residence time. Finally, the filtrations took place correctly, for the solids from the three configurations.

Claims (14)

1. Process for the back-extraction of uranium present in an organic phase charged with uranium and containing an amine compound or a mixture of amine compounds of formula R ₃ N, R being chosen from hydrogen and an alkyl group from C6 to C18 optionally substituted and with at least one R group being an optionally substituted C6 to C18 alkyl group, the amine compound being capable of forming a complex with uranyl sulphate, the process comprising the following steps: a) the organic phase is brought into contact with an aqueous uranium back-extraction solution comprising carbonate ions so as to transfer the uranium into the aqueous solution by complexation, the pH of the aqueous back-extraction solution being chosen so that the uranium remains in soluble form; b) a liquid/liquid separation step is carried out, whereby an organic phase depleted in uranium and an aqueous back-extraction solution loaded with uranium in the form of uranyl tricarbonate are obtained; the method being characterized in that the aqueous back-extraction solution further contains hydroxide ions supplied to the aqueous back-extraction solution. 2. Method according to claim 1, comprising the following steps: c) the aqueous solution resulting from step b) is brought into contact with an aqueous solution containing hydroxide ions so as to produce a suspension containing a uranium precipitate; d) a solid/liquid separation is carried out on the aqueous suspension, whereby a solid phase of uranium precipitate and a liquid effluent containing carbonate ions are obtained. 3. Method according to claim 2, characterized in that it comprises a step e) in which all or part of the liquid effluent from step d) is recycled to step a). 4. Method according to one of the preceding claims, in which step a) theoretically involves n moles of carbonate ions per mole of uranium, of which m moles of carbonate ions per mole of uranium are useful for the deprotonation of a uranyl amine sulfate complex and wherein the stripping solution contains between (0.25*m) moles and (0.75*m) moles of hydroxide ions and between (n-(0.25* m)) moles and (n-(0.75*m)) moles of carbonate ions. 5. Process according to claim 4, in which the back-extraction solution comprises from 1 equivalent to (m-1) equivalents of hydroxide ions and from (n-1) equivalents to (nm+1) equivalents of carbonate ions. 6. Process according to claims 4 or 5, in which n is equal to 7 and m is equal to 4. 7. Method according to one of the preceding claims, wherein in step a) an aqueous solution of alkali metal hydroxide or ammonium hydroxide is added. 8. Process according to claim 7, in which the aqueous solution of hydroxide of an alkali metal is a solution of sodium hydroxide. 9. Method according to one of the preceding claims, in which the amine compound or the mixture of amine compounds of formula R ₃ N is a C8 to C12 aliphatic tertiary amine. 10. Method according to one of the preceding claims, in which step a) of bringing the aqueous and organic phases into contact is carried out in mixer-settlers, pulsed columns or stirred columns according to a mode of co-current circulation or counter-current of the aqueous and organic phases. 11. Method according to one of the preceding claims, in which the organic phase further comprises a water-immiscible hydrocarbon diluent. 12. Method according to one of the preceding claims, in which the organic phase loaded with uranium is obtained by bringing a sulphated uranium-bearing leaching solution into contact with an organic phase containing the amine compound or a mixture of amine compounds of formula R ₃ N. 13. Method according to one of the preceding claims, in which the pH of the aqueous back-extraction solution loaded with uranium resulting from step b) is between 7 and 10. 14. Method according to one of claims 2 to 13, in which the hydroxide ions are produced by caustification of part of the liquid effluent from step d).