BE569879A - - Google Patents

Description

       

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   The invention relates, in general, to the recovery of uranium from processing solutions, and more particularly to the recovery of uranium uranium by extraction with an alkyl phosphate and precipitation in the state of tetrafluoride. purified uranium.



   Uranium is efficiently recovered from low grade ores, mineral washing solutions, various processing solutions, or other uranium-containing materials by various processes, such as ion exchange, " resin in paste ", and solvent extraction, applied heretofore, in which the uranium is selectively adsorbed out of an impure solution by contact with various ion exchange resins, for example, disposed in a column of ion exchange, and the uranium is selectively adsorbed out of an ore slurry or paste by various ion exchange resins disposed in situ therein, and in which a phase of extraction containing various organic phosphate esters in an organic solvent is employed to leach or extract uranium out of unwanted material,

   respectively. Uranium is obtained, in the final stages of the above processes, in the form of an enriched elution or extraction solution in which the uranium is dissolved in solutions of mineral acid or salt having a acidity varying between that of neutral solutions and that of strongly acidic solutions of 8 to 10 M and even more.



   It has now been discovered that valuable uranium-containing products can be selectively extracted from these elution solutions and extraction solutions with high efficiency, economically, and at a particularly high state of purity. by a treatment comprising the reduction of the uranium to oxidation state 4, the extraction with an extraction phase comprising an alkyl phosphate and a solvent-diluent, and the precipitation of the values extracted in the state exceptionally pure uranous tetrafluoride from this extraction phase.

   Alkyl phosphate, i.e., the preferred compound tributyl phosphate, hereinafter referred to as TBP, has been found to be surprisingly selective for uranium uranium and generally capable of to reject most of the other metal ions found in appreciable amounts with the uranium in such reduced elution and extraction solutions. In addition, the reduced uranium extracted in the T.B.P. is finally recovered from it in the form of exceptionally pure uranous tetrafluoride by simple precipitation directly from this phase by the addition of an aqueous solution of hydrofluoric acid.



   These highly purified products of uranous tetrafluoride are quite essential and useful in the atomic energy industry, as their purity is sufficient for transformation into uranium hexafluoride used as feed material in the diffusion process. gas in which the uranium is separated into isotopic components to produce uranium enriched in U 235 and even metallic uranium which is used in nuclear reactors. Due to the detrimental sensitivity of nuclear reactors, even to very small amounts of neutron-absorbing impurities, the raw materials from which nuclear fuels are derived must, necessarily, be extremely pure in order to produce nuclear fuels of the necessary purity in the nuclear industry. this technique.



   The subject of the invention is therefore a process for the recovery of uranium in the very pure state from elution solutions obtained in the processes by ion exchange.



   It also relates to a process for the recovery of highly purified uranium from the elution solutions obtained in the “resin in paste” processes.



   . It also relates to a process for the recovery of highly purified uranium from the acid extraction solutions obtained in

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 solvent extraction processes.



   The invention also provides a process for the separation of uranium from associated metal impurities in acidic extraction solutions and elution solutions from the processing of various uranium ores and other uranium recovery processes.



   The invention also provides a process for separating uranium from associated metal impurities by reduction in an acidic solution or an eluting solution thereof and then by contacting this solution or elution with an extraction phase. immiscible comprising TBP and an organic solvent-diluent.



   Finally, a subject of the invention is a process for the preparation of high purity uranium tetrafluoride by extracting uranium uranium from extraction solution and elution solutions reduced by an organic phase comprising =: TBP and a solvent-diluent and then precipitation of the urancux fluoride out of this organic phase with aqueous hydrofluoric acid.



   Other objects and advantages of the invention will emerge from the following description, with reference to the appended drawings, in which: FIG. 1 is a flow diagram illustrating the method of the invention; - Fig. 2 is a graph of the uranium extraction coefficient as a function of the acid concentration for the extraction of tetravalent uranium by an extraction phase of the invention; - Fig. 3 is a graph of the electrolytic reduction of ions in a strong acidic extraction solution; - Fig. 4 is a graphical representation of the electrolytic reduction of ions in an ion exchange eluting solution; - Fig. 5 is a graphical representation of the electrolyte reduction of ions in a "resin-in-paste" eluting solution.



   In general, the process of the invention relates to the recovery of uranium in the form of precipitated or solid UF, extremely pure, out of strong acid extraction solutions obtained in the final operation in various processing methods. Solvent extraction described in the patent application of Robert R. Grinstead, filed in the United States of America on June 29, 1956 under No. SN 696,034, as well as in certain other patent applications mentioned herein. Reference is made below to appropriate parts of these applications. These applications include those of Richard H. Bailes and Ray S. Long, No. SN 335,276 filed February 5, 1953; Ray S. Long, N SN 491,798 filed March 2, 1955; Ray S. Long, N SN 502,253 filed April 18, 1955; R.

   Grinstead, No. 3N 527,428 filed Aug 9, 1955; and Robert R. Grinstead, No. SN 527.429 filed August 9, 1955. Likewise, uranium can be recovered by the process of the invention from various uranium-containing effluent liquids obtained in exchange processes. ions, as described in U.S. Patent to RS Long, No. 2,770,520 granted November 13, 1956, and in an article by Grins- tead, Ellis and Olsen published in the Proceedings of the International Conference on the Peaceful Uses of atomic energy (Geneva, August 1955), volume 8, pages 49-53.



   Likewise, uranium can be recovered by the process of the invention from various effluent liquids containing uranium finally obtained in the "resin in paste" process as described in an article by Hollis and Mc Arthur in the Accounts. proceedings of the International Conference on the Peaceful Uses of Atomic Energy (Geneva, August 1955), volume 8, pages 54-63.

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   In short, according to the invention, whether it is an extraction solution, an ion exchange elution, a "resin in paste" elution or another solution. any analogue, the uranium in this solution is reduced to oxidation state 4, the acidity is set to about 8-10 M and the uranium is extracted from the solution with an organic phase comprising alkyl phosphate and an organic solvent-diluent. The uranium in the organic extract phase is then precipitated out thereof in the form of high purity uranium tetrafluoride by adding an aqueous solution of hydrofluoric acid.



   The following is a summary of the prior methods for producing the feed solution used in the method of the invention. In solvent extraction processes, a solvent extraction phase comprising an organic solvent-diluent and between less than 1 and 20 to 30% or more of mono- and di-alkylated derivatives of phosphoric acid is employed, for example. esters of alkyl-ortophosphoric, alkyl-pyrophosphoric and alkyl-orthophosphoric acids, for washing or extracting uranium from mineral acid solutions, solids or mixtures of solids in the form of slurry.

   The solvent-diluent can be any of a large number of water-immiscible organic fluids, such as aromatic, aliphatic, petroleum-based solvents, etc .; Uranium is extracted from the various leach solutions , neutral or acidic, obtained by leaching ores or other solids with mineral acids, and from aqueous sludges of solids. Extraction products can also be used to leach uranium directly from solids such as high lime ores in the presence of concentrated mineral acid. The phase ratios, concentrations of the extraction product, the manner of bringing into contact and the other variables of the extraction are adjusted so as to obtain the optimum degree of recovery, to carry out the purification, to adjust the selectivity of the extract. extraction, etc.

   Uranium is efficiently separated from these extracts by contact, preferably, with hydrochloric acid at a concentration of at least 8-10 molar. The acidic separation solution contains uranium in the form of UO ++ (uranyl) or U + (ura-
 EMI3.1
 neux) together with A 1 +++ Fe +++ vanadium and other materials which are mined at least as efficiently as uranium. H2SO4 with a concentration greater than about 8 M is used in the same way.



   In the ion exchange process for the recovery of uranium, the leaching of the ores is done with acid or carbonate solutions, so the uranium is dissolved to be then purified and concentrated. In the case of acid washing, the crushed ore is treated with sulfuric acid; this treatment solubilizes the uranium and various other valuable metals, and thus separates them from the waste material or gangue. In the case of carbonate treatment, the crushed ore is washed with about 5-10% sodium carbonate solution, usually at elevated temperature; Washing is generally preceded by roasting of the crushed ore and in some cases the presence of an oxidizing agent is beneficial.

   In all cases, the uranium and various other valuable metals are solubilized and thus separated from the waste material or gangue. To be usefully used in the ion exchange process, the wash solution should be clear and contain no suspended solids. By flocculation, precipitation and filtration of the suspended solids by well known techniques, a clarified uranium leach solution is obtained for further treatment by ion exchange. These leaching and clarification treatments are carried out for example by percolation, washing or by stirring, thickening and filtration, etc.

   In any case, a clear washing solution is obtained, which flows through columns containing fixed beds of strong base anion exchange resin, in which:
 EMI3.2
 uranium, in the form of the complex anion [ÚO; / S04) 3 4 d to ns ".1; e" itī: 1, ...: 1 1 J 't "', - .., . 1, i -, '.1 ...;, f,; rf "v. ,

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   in the case of sulfuric acid wash solutions, and as [UO2 (CO3) 3-4 in the case of carbonate wash solutions, is adsorbed onto the resin and thus removed from the wash solution. Strongly basic anion exchange resins of the type comprising quaternary amine substituents on a styrene matrix. divinyl-benzene are well suited for this use in exchange columns.

   The elution of uranium can be carried out with a large number of eluents comprising solutions of salts and / or acids; however, the preferred solution is roughly 1 M in 01 :; and about 0.1 M in HCl. Such an acidified saline solution flows through the adsorbent resin bed, and thus uranium and various metallic impurities, if any, are eluted out of the resin and appear in the eluted solution. .



   The "resin-in-paste" uranium recovery process is most advantageously used for the treatment of ores or other uranium-containing products in which a washing solution or sludge which cannot be subjected to water is obtained. to effective clarification. In the ion exchange process, the wash solution arriving at the resin batches in the exchange columns should not contain suspended solids, as then the columns quickly oolog and become ineffective.

   However, in the "resin-in-paste" process, uranium is efficiently recovered directly from the ore wash paste by contacting this paste with large amounts of suitable ion exchange resins which are suitable. placed in stainless steel wire baskets, by immersion in this paste. Uranium and various metal ion impurities are selectively adsorbed out of the wash paste by strongly basic anion exchange resins similar in composition to those employed in the above ion exchange process. but having physical dimensions and properties suitable for the special conditions of loading, sieving and elution cycles of the "resin in pate" process.

   After the adsorption cycle, the resin and the spent paste are separated, and the uranium is then separated by elution from the resin, with a solution chosen from a large number of salt and / or acid solutions, including Preferred solutions are about 1 M in salt, eg NHO, and 0.1 M in acid, eg NHO. In any case, a chosen eluent and the absorbate are brought into contact, and the uranium is separated by elution from the resin and appears in the effluent eluting solution.



  These uranium-containing solutions obtained in the separation processes by solvent extraction, by elution after ion exchange and by elution in the "resin-in-paste" process are well suited for use as starting solutions for the preparation. performing the method of this invention.



  Solvent extraction solutions, when they have a concentration of 8 M or more in mineral acid, are used without subsequent adjustment of the acid content; however, ion exchange and "resin-in-paste" eluting solutions, normally weakly acidic, are treated with concentrated acid, preferably hydrochloric acid, to increase the acid concentration of these solutions to. at least 8-10M. Acid H2SO4 of a concentration greater than about 8 M and other mineral acids can be used, but give less desirable results.

   Although the solutions obtained by elution in the ion exchange and "resin-in-paste" processes have lower acidity, and therefore require acidification, they are nevertheless suitable for treatment by the process of the invention.



  To obtain the best advantages of the invention, it is necessary, as indicated below, to extract the reduced uranium out of a solution having an acid concentration greater than about 8 M and, therefore, to such neutral or weakly acidic elution solutions are treated with a concentrated acid prior to the extraction of uranium.



   The reduction of the metal ions, comprising the uranium and the impurities, contained in the solution is also carried out before the extraction operation and can be done either on the solution with low acidity, or after

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 acidification, depending on the economics of reduction '.' When the solution obtained by elution is reduced by adding an agent, for example aluminum, zinc, etc., it is more economical to reduce the solution at low acidity, because there is relatively little metal. lost in reaction with acid, as would be the case in strongly acidic solution.

   However, in case the electrolytic reduction is applied, it is more advantageous to acidify before reduction, since the increase in conductivity of the high acidity solution facilitates the passage of electric current and the resulting reduction. - aunt of uranium and other ions. In any case, a concentrated mineral acid, preferably hydrochloric acid, is added to the elution solution to bring the acid concentration of this solution to at least 8 M, the impurities and products of uranium value. contained therein being reduced before or after the acidity adjustment.



   The reducing agent used in the process of the invention is chosen for its optimum reducing power of uranium, its non-contamination characteristics, its ease of removal and its economy. The greatest importance is attached to the ability of the agent to completely reduce uranium in solution and to the formation of products which are not mined together with the uranium and appear in the fluoride product. uranium in the form of contaminants.



  Due to the need to obtain a product of high purity, the reducing agent must be carefully chosen. Certain granular metals, preferably aluminum, are suitable. Such a metal is reasonably priced and can be obtained very pure. In addition, the metal oxidation product, the A1 +++ ion, is excluded from the extraction phases of the invention.



  The reduction efficiency of $ (aluminum, relative to its weight, is also very advantageous, since the metal gives three electrons per atom oxidized. Other metals can also be used in the reduction of uranium, for example eg zinc and magnesium, however, in general, taking into consideration reduction efficiency, cost and non-contamination, aluminum is the preferred reducing agent.



   In addition, uranium is reduced with the same efficiency and, in some cases preferably, by electrolysis. This electrolytic reduction is effected by means well known in the art whereby an electric current is passed through the solution, this solution serving as electrolysis in a compartmentalized cell. By passing electric current, the solution in the cathode compartment is reduced. Uranium is most efficiently reduced in such an electrolytic cell, especially in the presence of strongly acidic extraction solutions from the aforementioned solvent extraction processes.

   Similarly, ion exchange and resin-in-paste elution solutions are reduced, but with less efficiency than solvent extraction solutions.



   The feed solution, that is to say the solutions obtained by solvent extraction, by ion exchange elution and by the "resin in paste" process, in which the acid content is set to at least 8 at 10 M, can be more effectively reduced by electrolysis. Other ionic impurities in feed solutions are: similarly reduced along with uranium. Unwanted ions such as ferric ion are thus reduced to the ferrous state, a state in which these ions are effectively rejected by the T. B.P. extraction phase during the subsequent extraction operation.



  Typical energy requirements for this electrolytic reduction are shown in Example VIII below. Electrolytic reduction has the advantage of effectively reducing uranium in solution without adding any material to the solution that could be carried throughout the process and contaminate the desired high purity UF4 product.



   With uranium in the +4 oxidation state and an acid content of at

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 minus 8-10 M, the feed solution is then subjected to the extraction operation. In accordance with the invention, uranium uranium is extracted from such a feed solution concentrated in mineral acid by contact with an immiscible extraction phase comprising T.B.P. mixed with an organic solvent-diluent. In practice, it has been found that this extraction phase shows a striking extraction efficiency for uranium +4, but exhibits almost complete extractive selectivity towards the other metal ions present, including at least Fe ++, Al +++, Mo +++, and V ++ ions.



   Conventional apparatus and methods are used for contacting the reduced acidified feed solution with the TBP extraction phase. This apparatus can be adapted for single phase equilibration or for extraction. against a current; It may consist, for example, of continuous mixer-settler apparatus, extraction pulse columns and other apparatus effectively applicable to contacting and separating non-target phases.



   It has been found that the extraction phases containing from about 10 and up to 70% T, B.P. are convenient for the extraction of reduced uranium.



  With higher concentrations of T.B.P., there is an increase in the loss of extractant in the acid phase, and with lower concentrations, an increase in the loss of acid in the extraction phase.



  It should be noted that the solutions of T.B.P. of lower concentration and therefore of lower extraction efficiency can be used effectively in the process of the invention by employing multi-stage contacting and stage separation and recycling operations. multiples; Thus, the lower extraction efficiency of these solutions is effectively utilized.



  In order to determine the concentrations of the extraction product, it should be noted that the increase in the concentration of the extraction product compensates for the decrease in the extraction capacity obtained when the acid concentration is lower, the concentration in uranium is weaker, with a shorter contact time, with a less efficient contact method or less favorable phase ratios. However, in general, it can be said that: it has been found that the T.B.P. between about 20 and 40% by weight are the most satisfactory, because within these concentration limits, the uranium extraction coefficient is generally sufficient and the losses of the extraction product in the acid phase are low.

   Ratios of about 1: 1 between T.B.P. extractor product, and feed solution phases are suitable in single stage operation; however, ratios from less than 1/1 to about 4/1 can be used in a multi-stage operation.



   With regard to the tributyl phosphate (TBP) used in the extraction phase of the invention, it should be noted that this product is well known in the chemical art and is readily available from a large number of industrial sources. . However, if desired, the tributyl phosphate can be prepared by dissolving phosphorus oxy-trichloride in a suitable solvent, for example xylene or toluene, and then contacting the solution with a solvent. stoichiometric excess of butyl alcohol; thus tri-butyl phosphate is formed, mixed with the solvent. Usually, normal butyl alcohol is used; however isomers and mixtures thereof should behave in a similar manner.

   The T.B.P. obtained can be isolated from the solvent by distilling the latter; or if the solvent is the same as the solvent-diluent used in the uranium extraction, then the tributyl phosphate-solvent reaction solution can be used directly in the extraction operation when suitably diluted with this solvent-diluent.



   With regard to the solvent-diluent for the T.B.P. extraction phase, it has been established that the choice of the diluent solvent is not critical from the standpoint of extractability. In addition, the loss or decomposition of solvent-diluent is insignificant in the equilibria in a strongly acidic medium; by c o n s e q u e n t o n can say

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 the choice of organic solvent-diluent is determined by simple factors such as the solubility of T.B.P. in the solvent, the price of the solvent, the safety risks incurred in the use of the solvent, and the handling behavior of the solvent-diluent in an apparatus. particular extraction.



  The solvents which) are used in the process of the invention include fluid hydrocarbons, halogenated hydrocarbons, aliphatics, aromatics and others. More particularly, solvents such as benzene, ethyl-benzene, chlorobenzene, toluene, xylene, kerosene and other fluid solvents derived from petroleum, and carbon tetrachloride have been found satisfactory. Toluene, because of its low volatility, low cost and other generally desirable factors, is a preferred solvent-diluent.



   The transfer rate of uranium from the feed solution phase to the organic phase is very high, with virtually all of the uranium being transferred within a few minutes of contact time. There is no advantage in extending the contact time beyond a total of about five minutes, since prolonging the contact does not result in an increase in the amount of uranium mined. In fact, increasing the contact time slightly increases the loss of acid from solution in the T phase. B.P.

   What is even more important is that a long contact time exposes the reduced feed solution to the oxidizing action of air, and as a result various metal ion impurities can be oxidized. in a state easier to extract and, therefore, appear with the uranium in the organic phase, and possibly contaminate the final UF4 product. Therefore, equilibration can be carried out at a high rate with efficient transfer of uranium and a resulting gain in the extraction and purity of the final product.



   It follows from what has been said above that all precautions must be taken, following reduction of the feed solution, to prevent excessive contact with air or another oxidant to avoid reoxidation. , either uranium or metallic impurities .- This reoxidation can be avoided by reducing and circulating the feed solution in enclosures isolated from the atmosphere or by carrying out successive reduction operations and extracting the invention immediately one after the other or in an inert atmosphere.



   Phase separation by conventional methods gives a spent acidic feed solution which is recycled for further uranium enrichment is an organic extract phase from which the uranium is recovered. The extract usually contains uranium as a complex compound of T. B.P.-tevalent uranium. Hexavalent uranium may be present if it existed in the reduced feed solution; this hexavalent uranium is recovered by conventional means when the concentration is high enough to make recovery practical. Complete reduction of the feed solution and careful handling keeps the hexavalent uranium content to a minimum.

   Very small amounts of other metal ions, such as Fe and Mo +++, in addition to small amounts of the strong acid, may also appear in the extract; these materials are effectively separated from the uranium in the subsequent precipitation operation.



   Finally, the uranium is precipitated out of the organic extract phase as UF4. Due to the particular efficiency of the combination of operations of the process of the invention, this precipitate of UF. is obtained in an exceptionally pure state, so that it can be used directly or with only% of minimal modifications as a feed material for various processes for the production of uranium metal of extreme purity, as a feed product for uranium isotope separation processes, and any other application requiring uranium of the highest purity.



   More precisely, the T.B.P. loaded with uranium is contacted with a reagent containing a source of available fluoride ions.

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  The tetravalent uranium in the extract phase reacts with these fluoride ions to produce insoluble uranous tetrafluoride salt which precipitates out of solution ';: As UF4 is essentially completely insoluble, uranium is quantitatively precipitated out of, l 'extracted and is recovered by a simple operation of filtration or centrifugation. As indicated above, any source of fluoride ion may be suitable for use as a precipitant of UF4; however, the preferred precipitant is hydrofluoric acid in aqueous solution.

   Concentrated or 48% hydrofluoric acid is normally used, although in many cases more dilute HF solutions can also be used, provided that dilution does not increase the tendency to water. emulsification.



   Although anhydrous GF can be used to precipitate uranium out of the extraction product, its use is impractical, due to the reaction of this reagent with the T.B.P. itself, and the resulting destruction of the extraction properties of TBP with aqueous solutions of HF, the extraction phase is hardly affected, and it can be recycled over and over to extract more uranium. out of other parts of the reduced feeding solution.



   Certain other materials can also be used effectively as precipitants of uranium, these materials being various soluble fluorides, for example NaF and NH4F. Such fluoride salt solutions, when brought into contact with the extract, precipitate uranium in the state of an insoluble double salt, for example NaUF5 or NH4UF5 To obtain very pure uranium fluoride, these salts Doubles are normally not desirable because they introduce a second contaminating cation into the precipitate. For feed materials, these cations are unnecessary and undesirable. However, for some uses such double salts are suitable and so such a process is occasionally used when extreme purity is not required.

   It should also be noted that a salt such as NH4UF5 can be easily transformed into UF4 by a simple ignition operation, in which the precipitate is heated to remove the volatile ammonium ion in the form of ammonia gas and ammonium fluoride leaving the stable tetrafluoride behind.



   Concentrated aqueous hydrofluoric acid (48%), the preferred precipitant of the invention, precipitates the uranium from the T. B.P. extraction phase in a very efficient and economical manner, giving a very pure uranous tetrafluoride.



  This aqueous solution of HF precipitates 99% of the uranium in the extract, when present in an amount in excess of about 10% of the stoichiometric amount and, in greater excess, precipitation of 1 is obtained. Almost quantitative uranium (99.77%) The UF precipitate is then filtered or centrifuged to separate it from solution, collected, dried and packaged for use as above.



   The following examples give details of the process, concerning the preparation of the very pure UF4 precipitate by extraction of uranium +4 with an extraction phase of TBP, of a solvent extraction solution, of solutions of uranium +4. elution by ion exchange and elution solutions by "resin in paste", in accordance with the invention. These examples are given by way of illustration of the invention and are not limiting.



  EXAMPLE 1.-
A series of experiments were carried out to determine the distribution of uranium +4 between HCl and an extraction phase having a concentration of 20% T.B.P. in toluene, depending on the concentration of HCl.



  The uranium in the +6 oxidation state dissolved in 0.5 M HOI is reduced by metallic zinc and the HCl concentration is adjusted in several portions by increasing it to values of 4.6 at the , 5 M prior to extraction. The electromotive potential of the reduced solution measured between a normal calomel electrode and platinum is approximately +0.4 volts at each level of increase.



  At the above potential, a trace of uranium +3 is present initially, but

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 this material is rapidly oxidized by air to the +4 state. The individual parts
 EMI9.1
 they of the reduced solution, after adjusting the acid concentration, are stirred with an equal amount of extractant under a nitrogen atmosphere for a period of two minutes.



  The distribution of tetravalent uranium between phases is determined by phase analysis, and extraction coefficients are calculated; obtains the results shown graphically in FIG. 2. As can be seen
 EMI9.2
 in Fïg2, which represents the extraction coefficient (KO, A depending on the acid concentration, the extraction coefficient rises slowly until an acid concentration of about 7 M is reached, after which the coefficient rises rapidly to a value of 560 for 10.5 '-of HCl concentration, which shows the remarkable efficiency of the extraction according to the invention.



    EXAMPLE 2. -
In a further series of tests to determine the suitability for extractio
 EMI9.3
 uranium at '7' tatr'd, 'aoxidation + 4 from solutions of chloride with a 20% tributyl phosphate in a toluene extraction phase, the
 EMI9.4
 chloride concentration by adding A101 3 to a solution of HCl containing uranium. In some cases the uranium is reduced by the zinc to the +4 state before the chloride adjustment and, in other cases, the reduction takes place after the chloride adjustment. The results of these tests are shown in Table 1 below.



   TABLE I.
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  Concen- Total Conc. HC1 Report Conc. final U308 gel Coefficient
 EMI9.6
 
<tb> tration <SEP> cI <SEP> 14 <SEP> M <SEP> of extraction <SEP>
<tb>
 
 EMI9.7
 initial phases equous Kola u308 gli A / 0 O {: n1- aqueous 12.5 4.6 0.2 1.0 0.13 12e37 0.011
 EMI9.8
 
<tb> 10.0 <SEP> 4.7 <SEP> 1.0 <SEP> 1.0 <SEP> 0.08 <SEP> 9.9 <SEP> o, oo8
<tb>
<tb> 10.0 <SEP> 5.6 <SEP> 1.0 <SEP> 1.0 <SEP> 0.41 <SEP> 9.6 <SEP> 0.045
<tb>
<tb> 10.0 <SEP> 6.5 <SEP> 1.0 <SEP> 1.0 <SEP> 2.47 <SEP> 7.5 <SEP> 0.33
<tb>
 
 EMI9.9
 12, 5 6, 7 0, 2 1, 0 9.35 3.08 3.0 1 olO 7.4 1.0 1.0 8.04 1.96 4.1
 EMI9.10
 
<tb> 12.5 <SEP> 4.6 <SEP> 4.6 <SEP> 0.4 <SEP> 0.13 <SEP> 9.7 <SEP> 0.013
<tb>
<tb> 12.5 <SEP> 4.6 <SEP> 4.6 <SEP> 1.0 <SEP> 0.13 <SEP> 9.9 <SEP> 0.013
<tb>
<tb> 12.5 <SEP> 6.7 <SEP> 6.7 <SEP> 1.0 <SEP> 9.07 <SEP> 0.93 <SEP> 9.8
<tb>
 
 EMI9.11
 10, 0 8, 8 8, 8 0, 4 3.99 0,

  016 250
 EMI9.12
 
<tb> 10.0 <SEP> 8.8 <SEP> 8.8 <SEP> 1.0 <SEP> 9.97 <SEP> 0.029 <SEP> 340
<tb>
<tb>
<tb> 5.0 <SEP> 10.5 <SEP> 10.5 <SEP> 1.0 <SEP> 4.99 <SEP> 0.009 <SEP> 560
<tb>
<tb>
<tb>
<tb> 5, <SEP> 0 <SEP> 1.0, <SEP> 5 <SEP> 10.5 <SEP> 2.5 <SEP> 12.47 <SEP> 0.014 <SEP> 890
<tb>
 
It is evident from the data shown in Table 1 that the uranium extraction coefficient + 4 in the TBP-toluene extractant becomes very high for a chloride concentration greater than 8 M and increases at a high speed for at least 10.5 M.



  EXAMPLE 3.-
The relative efficiency of extraction of the various metallic impurities most often present with uranium in solutions subjected to treatment by the process of the invention has been established. A solution containing uranium, iron, molybdenum and vanadium is obtained and reduced by metallic aluminum to a potential of + 0.4 volts (see above) and concentrated HCl is added to form a solution 10 M in HCl. This reduced solution contains:

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   @
17.0 g / l U3O8 (U + 4)
0.4 g / l Fe (Fe + 2)
0.4 g / l V2O5 (V + 3)
0.18 g / 1 MB (Mo + 3)
Aliquots of the above solution were diluted with water to form 7.8 and 9 M HCl solutions, respectively.

   These samples at the adjusted acid concentrations are then brought into equilibrium with T.B.P. 20% in toluene with aqueous / organic phase ratios of 1/1. A part of each organic extract is then washed with an equal volume of Hcl of a concentration equal to that used in the corresponding extraction.



  The extraction coefficient and the washing losses are given in Table II below.



   TABLE II.
 EMI10.1
 
<tb>



  Conc. <SEP> HCl <SEP> Coefficient <SEP> of extra <SEP> and <SEP> ion <SEP> KO / A <SEP> Loss <SEP> at <SEP> washing <SEP>% <SEP>
<tb>
<tb> M <SEP> U3O8 <SEP> Fe <SEP> Mo <SEP> U3O8 <SEP> Fe <SEP> Mo
<tb>
<tb>
<tb> 10 <SEP> 1000 <SEP> 0.16 <SEP> 1 <SEP>, <SEP> 24 <SEP> 0.10 <SEP> 2-
<tb>
<tb>
<tb> 9 <SEP> 530 <SEP> 0.13 <SEP> 1 <SEP>, <SEP> 70 <SEP> 0.15 <SEP> 5 <SEP> approx. 10
<tb>
<tb>
<tb> 8 <SEP> 300 <SEP> 0.07 <SEP> 1.09 <SEP> 0.42 <SEP> 10 <SEP> -
<tb>
<tb>
<tb> 7 <SEP> 50 <SEP> 0.03 <SEP> 1.38 <SEP> 2.65 <SEP> 29 <SEP> -
<tb>
 
The above results show that, under the conditions indicated, uranium is extracted with an efficiency which increases greatly when the HCl concentrations of the solutions are increased, while the ferrous ion is extracted more efficiently, but with an increase not as strong as that of uranium.



  The molybdenum extraction coefficient remains practically constant, and, although this is not indicated in the table, the extraction of vanadium is negligible, as no vanadium was detected in the extract phases.



  EXAMPLE 4.-
The ability of aluminum to reduce uranium from +6 oxidation state to +4 oxidation state in HCl solutions of various concentrations has been investigated. The degree of reduction of uranium in dilute HCl was determined from the potential of the solution, as measured between a saturated calcium electrode and a platinum electrode. A potential of + 0.3 volts is easily reached, at which point a trace of uranium in the +3 oxidation state is present. The presence of uranium +3 is indicated by a red coloration which develops in the solution when the HCl concentration increases to about 10 M; however uranium +3 is immediately oxidized to uranium +4 by short exposure to air without appreciable oxidation, at the same time, of tetravalent uranium.

   In strong HCl solutions, the above potential reading process is of no value, as the potential remains strongly negative throughout the reduction. Therefore, the degree of reduction of uranium in 10 M HCl with given amounts of granulated aluminum is estimated as follows. Under identical conditions (concentrations of HCl, T.B.P. and U), the extraction coefficient for uranium +4 is categorically greater than that for uranium +6. This difference in the possibilities of extraction is used to determine the amount of each type present in the solution after reduction with a known amount of aluminum.

   The extraction coefficient for each type is determined by averaging the results of several tests carried out under the same conditions. A solution containing 10.0 g / l U3O8 and 10 M HCl was prepared, and 25 cm3 portions were stirred for two minutes with equal volumes of T.B.P. at 20% in toluene. An average

 <Desc / Clms Page number 11>

 analyzes carried out on duplicate tests give a value of 37.4 for the extraction coefficient (K O / A) for uranium +6. The solution of uranium +4 is prepared by first reducing the uranium, and then by adjusting the HCl concentration to 10 M. The extraction coefficient for uranium +4 is thus found to be 770.
 EMI11.1
 



  Portions of the above solution (10.0 gy1 of U308en u: + were treated with known amounts of aluminum; after which, an aliquot of each, 25 cm3, was extracted by stirring for two minutes with a 25 cm3 portion of TBP at the $ in toluene as extractor product The depleted aqueous phases obtained are then analyzed to determine the total uranium (+6 and +4) and the organic concentration is calculated from these data. corresponding, Knowing the total amount of uranium in each phase and the value of K 0 / A for each oxidation state, we calculate the amount of uranium +4 in each system, and, from this, the degree of Reduction The results of this reduction process are shown in Table III below.



   TABLE III.
 EMI11.2
 



  Al uses% U308 in u3 08in K 0 / A Reduction (stoechiom.), F of 11% (stoechiom.) Aqueous gl organic ¯¯¯¯¯¯¯¯¯ ¯¯¯¯¯¯¯¯ 1 ¯¯¯¯¯ ¯¯¯¯¯¯ 'oo, 26 9 74 37, 4 P 300 0.23 9, 77 42, 5 11 900 eye 981 5, 6 28 1500 0, r 4 986 70.5 48 260 0.013 9, 9 770 100 Uranium reduced prior to the addition of HCl for 10 M.



  EXAMPLE 5.-
In order to determine the uranium purification carried out by the reduction stage of the process according to the invention, the distribution of an impurity, molybdenum, between the HCl solutions and the T.B.P. extraction phases was determined. Molybdenum exists in some Colorado ores and accompanies uranium in the primary extraction process in the HCl extraction solution. It is necessary to find a way to reject molybdenum in the purification process to ensure high purity of the UF4 product. The behavior of molybdenum after various degrees of reduction was therefore investigated, during the following extraction tests with the T.B.P. extraction solution.



   A solution containing molybdenum in the form of the molybdate ion and hydrochloric acid is used. Portions of the solution are reduced with known amounts of granulated aluminum, adjusted to the desired HCl concentration with concentrated HCl, and then extracted with an equal volume of T.B.P. 20% in toluene; the results are shown in Table IV below.



  The amounts of aluminum are given as a percentage of the stoichiometric amount for reducing molybdenum to the trivalent state following the reaction:
 EMI11.3
 In practice, part of the aluminum is used to react
 EMI11.4
 with HCl and is therefore not available for molybdenum.

 <Desc / Clms Page number 12>

 
 EMI12.1
 
<tb> Equivalent <SEP> Conc.

   <SEP> in <SEP> Mo <SEP> in <SEP> Mo <SEP> in <SEP> K <SEP> O / A <SEP>
<tb>
<tb> stoichiome - <SEP> HCl <SEP> organic <SEP> aqueous
<tb>
<tb> bar <SEP> of <SEP> M <SEP> g / l <SEP> g / l
<tb>
<tb> A1%
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 200 <SEP> 6 <SEP> 0.18 <SEP> 1.32 <SEP> 0.14
<tb>
<tb>
<tb>
<tb> 200 <SEP> 8 <SEP> 0.12 <SEP> 1.38 <SEP> 0.09
<tb>
<tb>
<tb> 200 <SEP> 10 <SEP> 0.14 <SEP> 1.36 <SEP> 0.10
<tb>
<tb>
<tb>
<tb> 59 <SEP> 8 <SEP> 2.56 <SEP> 0.36 <SEP> 7.1
<tb>
<tb>
<tb> 118 <SEP> 8 <SEP> 2.20 <SEP> 0.82 <SEP> 2.7
<tb>
<tb>
<tb>
<tb> 177 <SEP> 8 <SEP> 0.71 <SEP> 2.24 <SEP> 0.32
<tb>
<tb>
<tb> 236 <SEP> 8 <SEP> 0.34 <SEP> 2.60 <SEP> 0.13
<tb>
<tb>
<tb>
<tb> 0 <SEP> 10 <SEP> 1.85 <SEP> not <SEP> found-
<tb>
<tb>
<tb> 64 <SEP> 1'0 <SEP> 1.78 <SEP> 0.07 <SEP> 25.4
<tb>
<tb>
<tb>
<tb> 128 <SEP> 10 <SEP> 1.20 <SEP> 0.65 <SEP> 1,

  8
<tb>
 
The above results show that as the amount of the reducing agent increases, the extraction of molybdenum decreases. The extraction of molybdenum (+3) reduced in the extractor product T.B.P. is therefore very low and thus by reduction of the uranium solution the separation of uranium and molybdenum which is an impurity is carried out.



  EXAMPLE 6.-
The efficiency with which the uranous ion is extracted from solutions in hydrochloric acid of various acidities was determined with T.B.P. as extractor. in .toluene. The distribution of uranium +4 between hydrochloric acid solution and T.B.P. at 20% in toluene is shown in Table V below.



   TABLE V.
 EMI12.2
 
<tb>



  Cone. <SEP> HCl <SEP> Concentration <SEP> Ratio <SEP> U3O8 <SEP> in <SEP> U3O8 <SEP> in <SEP> K <SEP> 0 / A <SEP>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> M <SEP> initial <SEP> phases <SEP> organic <SEP> aqueous
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> U308aqueous <SEP> g / l <SEP> Aqu / Org. <SEP> g / l <SEP> g / l
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 3 <SEP> 320 <SEP> 0.26 <SEP> --- <SEP> 320 <SEP> very <SEP> low <SEP>
<tb>
<tb>
<tb>
<tb> 6 <SEP> 35 <SEP> 1.0 <SEP> 28.6 <SEP> 6.4 <SEP> 4.5
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 7 <SEP> 11.9 <SEP> 1.0 <SEP> 11.7 <SEP> 0.23 <SEP> 50
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 7, <SEP> 6 <SEP> 87 <SEP> 0.5 <SEP> 43 <SEP> 0.37 <SEP> 116
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 9 <SEP> 35 <SEP> 1.0 <SEP> 34.8 <SEP> 0.15 <SEP> 230
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 9 <SEP> 42, <SEP> 8 <SEP> 1.0 <SEP> 42, <SEP> 6 <SEP> 0,

  18 <SEP> 240
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 9 <SEP> 80 <SEP> 0, <SEP> 5 <SEP> 39, <SEP> 9 <SEP> 0.14 <SEP> 290
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 9 <SEP> 15.3 <SEP> 1.0 <SEP> 15.3 <SEP> 0, <SEP> 029 <SEP> 530
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 10 <SEP> 10 <SEP> 1.0 <SEP> 10.0 <SEP> 0.013 <SEP> 770
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 10 <SEP> 17 <SEP> 1.0 <SEP> 17.0 <SEP> 0.018 <SEP> 950
<tb>
 The effect of the acid concentration is evident from the results.

 <Desc / Clms Page number 13>

 The states indicated, i.e., the extraction increases with increasing acid concentration, the extraction efficiency becoming remarkably high at HCl concentrations of 9 M and above.



  EXAMPLE 7.-
Research has been done on the recovery of uranous ions from charged organic extractants. Uranium tetrafluoride is precipitated from the extract by the addition of 48% hydrofluoric acid. Various stoichiometric percentages of hydrofluoric acid are added to determine whether the precipitation of UF4 is complete. The results of this research are shown in Table VI below.



   TABLE VI.
 EMI13.1
 
<tb>



  Solution <SEP> Concentration <SEP> HF <SEP> 48% <SEP> Equivalent <SEP> Concentration <SEP> Recovery
<tb>
 
 EMI13.2
 flight. em3 U 0 gli added, stoechiom. final U 0 UF g F? i.x gl 3 4
 EMI13.3
 
<tb> 25 <SEP> 34.8 <SEP> 0.442 <SEP> 85 <SEP> 9.31 <SEP> 73.2
<tb>
<tb>
<tb>
<tb> 25 <SEP> 34.9 <SEP> 0, <SEP> 530 <SEP> 102 <SEP> 0.96 <SEP> 97.2
<tb>
<tb>
<tb>
<tb>
<tb> 50 <SEP> 43.0 <SEP> 1.324 <SEP> 104 <SEP> 1.13 <SEP> 97.4
<tb>
<tb>
<tb>
<tb>
<tb> 25 <SEP> 42.6 <SEP> 0.707 <SEP> 111 <SEP> 0.39 <SEP> 99.1
<tb>
<tb>
<tb>
<tb>
<tb> 50 <SEP> 43.0 <SEP> 1.565 <SEP> 123 <SEP> 0.12 <SEP> 99.7
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> 25 <SEP> 28.6 <SEP> 0.609 <SEP> 143 <SEP> 2.37 <SEP> 91.7
<tb>
<tb>
<tb>
<tb>
<tb> 25 <SEP> 42.6 <SEP> 1.073 <SEP> 170 <SEP> 0.13 <SEP> 99.7
<tb>
<tb>
<tb>
<tb>
<tb> @
<tb>
 
As the results shown above show,

   it suffices for hydrofluoric acid in excess of 10% over the stoichiometric amount to precipitate 99% of the uranium out of solution as UF4, and even larger excesses result in a recovery of up to minus 99.7% of the uranium EXAMPLE 8. -
The electrolytic reduction of uranium present in several types of processing solutions has been studied.

   An 8M HCl extraction solution and elution solutions from ion exchange and "resin in paste" plants were obtained, solutions having the following partial analyzes:
 EMI13.4
 
<tb> Installation <SEP> Installation <SEP> Solution
<tb>
<tb> driver <SEP> by <SEP> exchange <SEP> of elut <SEP> ion <SEP>
<tb>
<tb>
<tb> HCl. <SEP> ground. <SEP> of ions. <SEP> "resin
<tb>
<tb>
<tb> extraction <SEP> Kerr-Ma <SEP> Gee <SEP> in <SEP> paste "
<tb>
<tb>
<tb> U.S.B.M.

   <SEP> Monticello
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb>
<tb> g / l <SEP> U3O8 <SEP> 47.5 <SEP> 13.13 <SEP> 6.70
<tb>
<tb>
<tb>
<tb> g / l <SEP> Fe <SEP> 3.28 <SEP> 1.17 <SEP> 0.12
<tb>
<tb>
<tb>
<tb> g / l <SEP> Mo <SEP> 1.11 <SEP> 0.016 <SEP> 0.007
<tb>
<tb>
<tb>
<tb> 'equivalent / 1 <SEP> H + <SEP> 8.43 <SEP> 0, <SEP> 20 <SEP> 0.16
<tb>
 
The reduction is carried out in a sealed graphite cuvette serving as a cathode and having a total surface area of about 400 cm 2. A graphite rod of about 1.6 am in diameter serves as the anode and forms a separate compartment

 <Desc / Clms Page number 14>

 cathode with an alundum cuvette 2.54 cm in diameter. 600 cc of the test solution is contained in the cathode compartment and 15 ³m3 in the anode compartment.

   Samples of the catholyte are taken at various times during the reduction and analyzed for ferrous and ferric iron.



  The reduction of uranium is determined by precipitation of uranium to the state of UF by HF and analysis of the filtrate for unreduced uranium. Figs. 3, 4 and 5 of the accompanying drawings graphically illustrate the reduction of iron and uranium in the 8M HCl extraction solution, the ion exchange elution solution and the resin-in-paste elution solution. "respectively, depending on the amount of current passed.



   Examination of these figures shows that in each case the reduction of iron occurs before the reduction of uranium. Uranium is almost completely reduced by the time the theoretical electrochemical equivalent of U3O8 and Fe has been applied to the 8M HCl extraction section. The reduction efficiency is not as high for the eluting solutions. by ion exchange and by "resin in paste", wherein about 90% of the uranium in the ion exchange eluting solution and 80% of the uranium in the "resin in paste" eluting solution. paste "are reduced by the theoretical electrochemical equivalent.

   The emf in the cases indicated above, is between 1.5 and 3.0 volts; this variation gives only negligible effects. However, above about 3.0 volts the evolution of hydrogen is evident and so the uranium reduction efficiency drops. The energy required for the reduction of uranium in the above solutions for 3 volts is 0.3 KwH / 0.454 kg of U3O8 for the 8 M HCl extraction solution and about 0.5 KwH + 0.454 kg of U3O8 for the elution solutions by ion exchange and "resin in paste". Electrolytic reduction has the obvious advantage of not introducing other contaminants.



   Legend of figures.



   A - Solvent extraction process.



   B - "" by ion exchange.



   C - "" by "resin in paste".



   D - Acidification.,
E - Reduction.



   F - Extraction.



   G - Precipitation. a - 8-10 M Uranium +6 acid feed solution (Extraction solution). b - Acid feed solution 8-10 M Uranium +6 c- HCl 8-10 M, Uranium +4, Impurities. d - Extraction product T.B.P. - recycled solvent. e - Uranium extracted U +4. f - Electrochemical equivalent of U3O8 + Fe. g - Ampere-hours / liter. h - Reduced quantity%. i - M concentration in HCl. j - Extraction coefficient K 0 / A.



   1 - Concentrated acid.



   2 - Extraction product.



   3 - Raffinate.

 <Desc / Clms Page number 15>

 



     4 - Aqueous solution of HF.


    

Claims (1)

ABSTRACT. The subject of the invention is a process for the recovery of uranium which is remarkable in particular by the following characteristics, considered separately or in combination: a) - applied to the recovery of values of tetravalent uranium from an aqueous solution with a concentration of mineral acid greater than approximately 8 M, it comprises bringing this solution into contact with an extraction phase comprising an organic solvent-diluent and tributyl phosphate (T. B.P.) to extract the uranium values; b) - the uranium is then recovered from the extraction phase; c) - uranium is precipitated from the extract in the form of fluoride; d) - for the recovery of uranium values after production of a solution of said values in a concentrated mineral aid, the process comprises the reduction of the uranium values in the +4 oxidation state, and placing in contact from said solution with an extraction phase of tributyl phosphate to extract uranium +4; e) - the solution containing impurities at the same time as values; uranium, the uranium values of the solution are reduced to the +4 oxidation state, the concentrated acid solution is brought into contact with an extraction phase comprising tributyl phosphate dissolved in an organic solvent-diluent to selectively extract the reduced uranium without the impurities, and recovering the purified uranium values from the tributyl phosphate extraction phase; f) - consisting in recovering uranium values from a material, the process comprises the production of a solution of these values in a mineral acid with a concentration greater than about 8 M, the treatment of this solution by a reducing agent to bring the uranium to the tetravalent state, bringing the solution into contact with an extraction phase of T. B.P. in an organic solvent-diluent to extract the uranium and recover the tetravalent uranium from the extract; g) - said extract is treated with an aqueous fluoride solution to precipitate, out of this extract, a uranous fluoride salt; h) - the uranium values to be recovered being placed in aqueous solution, the acidity of this solution is adjusted to at least 8 M by addition of concentrated mineral acid, the uranium values in the acidified solution are reduced to 1 oxidation 'state' +4, then the solution is brought into contact with an extraction phase comprising tributyl phosphate dissolved in an organic solvent-diluent to extract the uranium values +4, and the uranium is recovered outside the extraction phase; i) - uranium values are reduced by bringing their solution into contact with a metallic reducing agent; j) - uranium values are reduced by electrolysis; k) - said mineral acid comprises hydrochloric acid; 1) - said mineral acid comprises sulfuric acid; m) - the product consisting in preparing extremely pure uranous tetrafluoride from a solution, obtained from a solvent extraction process, loaded with impure uranium and which has a concentration of at least 8 M in mineral acid, this solution is treated with a reducing agent to reduce the uranium <Desc / Clms Page number 16> in this solution the +4 oxidation state, the extraction solution is brought into contact with an extraction phase comprising about 10 to 70% T.B.P. in an organic solvent-diluent selected from the group consisting of liquid petroleum solvents, chlorinated aliphatic, aromatic, aliphatic * and chlorinated aromatic hydrocarbons, to extract the uranium, the extraction phase is treated with an aqueous solution of fluoride to precipitating the uranium tetrafluoride, and the uranous tetrafluoride is separated from the phases; n) - said starting solution is an elution solution loaded with uranium, originating from an ion exchange process; o) - said starting solution is an elution solution loaded with uranium originating from a “resin in paste” process; p) - the process consisting in recovering uranium values from a mixture containing impurities, a concentrated mineral acid solution of these uranium values and impurities comprising Fe, Mo and V is produced, the solution with a reducing agent for reducing uranium, to the tetravalent oxidation state; Since impurities are thus also reduced, the solution is then brought into contact with an extraction phase comprising 10 to 70% of TB P. in an organic solvent-diluent in order to selectively extract the uranium and separate it from said impurities, and the liquid is recovered. uranium purified from the extract; q) - the extract is treated with an aqueous fluoride solution to precipitate a uranous fluoride, and the uranous fluoride precipitate is separated from the extraction and solution phases; r) - the solution is treated with aqueous hydrofluoric acid to precipitate the uranium tetrafluoride, and the uranium tetrafluoride is separated from the extraction and acid phases; s) - said solvent-diluent is a material selected from the group consisting of liquid petroleum solvents, aliphatic, aromatic, chlorinated aliphatic and chlorinated aromatics.