WO2015135017A1

WO2015135017A1 - Method for the recovery of uranium from a strong sulfuric acid loaded strip or eluate

Info

Publication number: WO2015135017A1
Application number: PCT/AU2015/000129
Authority: WO
Inventors: Areski Rezkallah; Grenvil Marquis Dunn
Original assignee: A-CAP RESOURCES Ltd
Current assignee: A-CAP RESOURCES Ltd
Priority date: 2014-03-11
Filing date: 2015-03-09
Publication date: 2015-09-17
Anticipated expiration: 2016-09-11

Abstract

A method for the recovery of uranium from a Pregnant Strip Solution (PSS) (201) emanating from an upstream uranium recovery process is described. The method comprises the steps of: extracting the uranium values quantitatively from the PSS (201) by loading an ion exchange resin in a separate ion exchange extracting step (200). The barren liquor (202/122) from the extracting step is then returned to the upstream uranium recovery process with substantially all of the acid that was in the PSS. Preferably the method further comprises the step of washing the loaded resin with a suitable wash liquid (211) in an ion exchange wash step (210). Advantageously the method further comprises the step of quantitatively eluting the loaded resin with a reagent in an ion exchange eluting step (220) wherein, when the uranium has been removed, the resin can be recycled to the ion exchange extracting step (200) with minimal loss in its activity.

Description

"METHOD FOR THE RECOVERY OF URANIUM FROM A STRONG SULFURIC ACID LOADED STRIP OR ELUATE"

Field of the Invention

The present invention relates to a method for the recovery of uranium from a strong sulfuric acid loaded strip or eluate and relates more particularly, but not exclusively, to the upgrade or recovery of uranium from an acidic loaded strip liquor emanating from a uranium solvent extraction circuit or from an acidic eluate from a uranium ion exchange circuit.

Background to the Invention

Solvent extraction (SX) and ion exchange (IX) are the preferred processes for the recovery of uranium from acid leach solutions. Either SX or IX could be used in a particular process, depending on the specific flowsheet; however normally when the uranium concentration in the acid leach solution is high, for example >800-900 mg/L U, SX is preferred over IX in terms of costs.

In cases where a tertiary amine is employed as the extractant in a uranium SX process, an approximate 4 molar sulfuric acid strip feed liquor can be employed to remove the uranyl sulfate anion from the extractant. The load chemistry for a tertiary amine type extractant may be shown by the following relationships:

H₄U0₂(S0₄)₃(aq) + 4R₃NH₂S0₄(o)→ [R₃NH]₄U0₂(S0₄)₃(o) + 4H₂S0₄(aq)

When the extractant is stripped with the strong (~4 Molar) sulfuric acid the chemistry could be reflected by the following relationship:

(R₃NH₄)₄ U0₂{SO₄)3(0)+4H₂SO₄(aq)→ H₄U0₂(S0₄)₃(aq)+4(R₃NH)HS0₄(o) In order to facilitate the precipitation of a diuranate intermediate, in excess of 75% of the contained sulfate will need to be neutralised, and removed in order to hydrolyse the uranium efficiently. The current practice therefore for treating the strong acid loaded strip (sometimes referred to as strip raffinate) from a uranium SX step appears to be one of partially neutralising the acid content of this stream with limestone, lime or suitable equivalent alkali.

If the loaded strip were to be processed to an intermediate sodium diuranate, for example, the following chemistry would apply:

H₂S0₄(aq) + CaC0₃(s) + 2H₂0(aq)→ CaS0₄.2H.O(s) + H₂0(aq) + C0₂(g)

Then follows a separation of the gypsum (CaS0₄.2H₂0) solid. Thereafter the filtrate or aqueous fraction containing only very small quantities of free sulfuric acid, sulfate and soluble uranium is treated with sodium hydroxide to produce a sodium diuranate intermediate for sale or for on-refining. H₂S0₄(aq) + 2NaOH{aq)→ Na₂S0₄(aq) + 2H₂0(aq)

2H₄U0₂{S0₄)3(aq)+14NaOH(aq)→Na₂U₂07.6H₂0(s)+6Na₂S0₄(aq)+H₂0(aq)

This process can be costly especially in cases where the acid and alkali are not readily available and have to be imported. In addition, the precipitate in the case of a limestone based neutralisation step may drag out soluble uranium which will need to be recycled for recovery. Furthermore the limestone may introduce impurities to the uranium liquor.

Alternatively, an ion-exchange (IX) circuit can be employed to recover uranium, for example, from acid leach liquors. The uranium is often eluted in a process employing a strong sulfuric acid (1-2 molar). The eluate containing low concentrations of uranium and minor impurity elements does not normally have adequate purity to convert directly to an intermediate uranium oxide of the type sodium diuranate. The load chemistry for a strong base anionic resin may be shown by the following relationship:

H₄U0₂(S0₄)₃(aq) + Res N₄(S0₄)₂{r)→ Res N₄U0₂(S0₄)₃(r) + 2H₂S0₄(aq) The following chemistry is understood to explain the process when the resin is eluted with a strong (1 to 2 molar) sulfuric acid:

ResN₄U0₂(S0₄)₃(r) + 2H₂S0₄(aq)→ ResN₄(S0₄)₂(r) + H₄U0₂(S0₄)₃(aq)

Strong Base Anionic (SBA) resins are also known to load small quantities of ferric iron, molybdenum, vanadium, thorium, etc. and these are also eluted by the strong sulfuric acid eluant and may report to the intermediate diuranate precipitate.

As in the case of the treatment of the loaded strip from an SX circuit, a large quantity of alkali is required to neutralise the free acid if a diuranate intermediate was to be produced from the eluate. The preference in the industry would be to employ the acidic eluate as a feed liquor to a tertiary amine SX circuit; however, the loading of uranium in this solvent extraction circuit is rarely greater than 10 g/L and the solvent extraction circuit is:

• prone to crud formation,

· organic degradation by oxidants to secondary amine which are often less specific towards impurities and can be carcinogenic, and

• require organic removal steps to reduce organic losses and cross contamination, etc.

A benefit of the tertiary amine SX circuit to process the strong acid eluate is that the raffinate from the extract circuit can be returned to the IX circuit as the eluant, but not normally without entrained organic removal. The present invention was developed with a view to providing a method for the recovery of uranium from an acidic loaded strip liquor emanating from a uranium SX circuit or from an acidic eluate from a uranium IX circuit which is less susceptible to the disadvantages of the prior art noted above. References to prior art in this specification are provided for illustrative purposes only and are not to be taken as an admission that such prior art is part of the common genera! knowledge in Australia or elsewhere.

Summary of the Invention

The present invention relates to a process for treating either the strong sulfuric acid strip raffinates (from a SX circuit) or the strong sulfuric acid eluates (from an IX circuit) - hereafter called a Pregnant Strip Solution (PSS). The method of the invention may also be applicable to other types of strong sulfuric acid loaded strip solutions. According to one aspect of the present invention there is provided a method for the recovery of uranium from a Pregnant Strip Solution (PSS) emanating from an upstream uranium recovery process, the method comprising the steps of: extracting the uranium values quantitatively from the PSS by loading an ion exchange resin in a separate ion exchange extracting step; and, returning the barren liquor from the extracting step to the upstream uranium recovery process with substantially all of the acid that was in the PSS.

Preferably the method further comprises the step of washing the loaded resin with a suitable wash liquid in an ion exchange wash step. Advantageously the method further comprises the step of quantitatively e!uting the loaded resin with a reagent in an ion exchange eluting step wherein, when the uranium has been removed, the resin can be recycled to the ion exchange extracting step with minima! loss in its activity. Preferably the resin is loaded in the extracting step to in excess of 30 g/L uranium. Preferably the ion exchange resin is a chelating resin of the type containing, for example, an amino phosphonic or a phosphonic/sulfonic functional group. Typically the chemistry in the extracting step can be shown as:

H₄U0₂(S0₄)₃(aq) + ResP0₃H₂(r)→ (ResP0₃)₂U0₂(r) + 3H₂S0₄(aq)

Preferably the PSS contains strong sulfuric acid at concentrations in excess of 50 g/L and below 600 g/L. More preferably the PSS contains free sulfuric acid at concentrations of between 50 g/L and 400 g/L. Typically an eluant stream applied to the resin in the eluting step is at least an aqueous solution of sodium carbonate. Preferably the eluant stream is a blend of sodium carbonate and sodium bicarbonate. In a preferred embodiment the eluant stream is at least sodium carbonate consisting primarily of a recycled solution and some make-up. Preferably the concentrations of sodium carbonate in the eluant stream are chosen to optimise the concentration of uranium in the e!uate and facilitate its complete removal from the resin.

The elution chemistry could typically be that reflected in:

(ResP0₃)₂U0₂(r) + 3Na₂C0₃(aq)→ ResP0₃Na₂(r) + Na₄U0₂(C0₃)₃(aq) 2ResP0₃H{r) + 2Na₂C0₃(aq)→ 2ResP0₃.Na(r) + 2NaHC0₃

2NaHC0₃(aq) + 2NaOH(aq)→ 2Na₂C0₃(aq) + 2H₂0(aq)

Preferably the eluted resin from the eluting step is washed with a dilute sulfuric acid, to produce an acid washed resin and a wash barren liquor. Advantageously the acid washed resin is recycled to the ion exchange extracting step where the load-wash-elution-wash cycle is repeated.

Preferably the eluate from the eluting step is fed batch or continuously to an intermediate uranium product recovery step. Typically the eluate is a rich blend of sodium uranyl tricarbonate, sodium carbonate and sodium bicarbonate, in a preferred embodiment the intermediate uranium product recovery step is a sodium diuranate precipitation step. Preferably the sodium diuranate precipitation step is performed by employing sodium hydroxide and maintaining an excess of sodium hydroxide concentration of approximately 1 to 2%. Preferably the wash liquid employed in the ion exchange wash step is at a temperature similar to that of the loaded resin advancing to the ion exchange wash step. Typically the wash liquid is a displacement wash of a strip raffinate feed in the extracting step that is in contact with the resin at the time the loaded resin is advanced to the wash step. Advantageously the spent wash from the wash step can be recycled to the PSS. Preferably the washed resin from the wash step is subject to a deprotonation step in which protons, on any cationic functional groups on the resin, are removed by an alkaline fluid. Typically the washed resin is deprotonated by a weak sodium hydroxide containing solution. In a preferred embodiment the sodium diuranate precipitate from the sodium diuranate precipitation step is separated in a separation step using a filter or some equivalent process equipment. Preferably a solid fraction of the sodium diuranate is appropriately washed with wash water to remove most of the aqueous solute. Advantageously the solute in the separated filtrate contains sodium hydroxide and sodium carbonate and is recycled as a primary component of the sodium carbonate eluant stream applied to the resin in the e!uting step.

Throughout the specification, unless the context requires otherwise, the word "comprise" or variations such as "comprises" or "comprising", will be understood to imply the inclusion of a stated integer or group of integers but not the exclusion of any other integer or group of integers. Likewise the word "preferably" or variations such as "preferred", will be understood to imply that a stated integer or group of integers is desirable but not essentia! to the working of the invention. Brief Description of the Drawings

The nature of the invention will be better understood from the following detailed description of a preferred embodiment of the method for the recovery of uranium, given by way of example only, with reference to the accompanying drawings, in which:

Figure 1 shows the overall flow-sheet of a preferred embodiment of the method for the recovery of uranium according to the present invention.

Detailed Description of Preferred Embodiments

A preferred embodiment of the method for the recovery of uranium from a Pregnant Strip Solution (PSS) emanating from an upstream uranium recovery process in accordance with the invention, will now be described in more detail with reference to Figure 1 , in which three recovery steps are illustrated:

• Step 1 : will be described in general terms as it pertains to a solvent extraction (SX) process employing a tertiary amine as the active extractant. However, Step 1 could alternately be an ion exchange (IX) process employing a SBA resin. · Step 2: will be described in more detail as it pertains more particularly to key steps in the method of the present invention.

• Step 3: will be described in general terms as it pertains to a classical sodium diuranate process. As will become apparent from the more detailed description below, the interactions between Step 1 and Step 2, and between Step 2 and Step 3, are particularly advantageous in the preferred embodiment of the present invention. Step 1 in Figure 1

The pregnant solution from the leach [101] is fed to an extraction step [100] in a solvent extraction circuit where the uranium is adsorbed by the extractant [104] to produce a loaded extractant [105] and a raffinate or barren leach solution [102]. The loaded extractant [105] is then scrubbed [1 0] to remove any mechanically transferred pregnant solution from leach and also remove any weakly loaded impurities. Water [11 1 ] can be employed as the scrub feed liquor. The products from the scrub are a scrubbed extractant [1 13] and a scrub raffinate [1 12]. The scrub raffinate [1 12] and the extract raffinate [102] can be blended to produce a blended raffinate [ 03] which can then be returned to the leach.

The scrubbed extractant [113] is then stripped [120] using a strong sulfuric acid (4M) strip feed liquor [121] as a supplement to a recycled strong sulfuric acid stream [122]. The blended strip feed [123] is employed to remove the uranium {and some impurities) from the extractant. A first product of the Strip step [120] is a strip raffinate [124] sometimes referred to as the loaded strip. A second product from the Strip step [120] is a strip organic [125] which is returned to the extraction step [100].

The strip raffinate [124] is the aqueous feed liquor or Pregnant Strip Solution (PSS) which is fed to Step 2 (which pertains to key steps in the method of the present invention). As noted above, the PSS could equally be an eluate stream from a strong acid strip ion exchange process. The strip raffinate [124] is temporarily stored in the strip raffinate surge tank [126] before being used as the PSS to Step 2.

Step 2 in Figure 1

The PSS from an upstream uranium recovery process (Step 1) is fed to an Ion Exchange Extract step [200] as an aqueous feed liquor [201]. Typically the feed liquor [201 ] contains strong sulfuric acid, typically at concentrations in excess of 50 g/L and below 600 g/L. More preferably the feed liquor [201] contains free sulfuric acid at concentrations of between 50 g/L and 400 g/L.

Preferably the feed liquor [201] is free of any entrained organic liquors or solid particles. The feed liquor [201] is preferably processed at a temperature between 1 °C and 100°C, more preferably at a temperature between 20°C to 40°C, in the Ion Exchange Extract step [200]. Preferably the feed liquor [201] is fed continuously to the Ion Exchange (IX) Extract step [200], in which uranium in solution is loaded onto resin beads. The resin employed is preferably a chelating resin of the type containing, for example, an amino phosphonic or a phosphonic/sulfonic functional group.

Preferably a washed ion exchange resin [204] is recycled to the IX Extraction step [200]. Typically the uranium concentration in the washed ion exchange resin [204] advanced to the IX Extraction step [200] is below 5 g/L. More typically the washed ion exchange resin [204] advanced to the IX Extraction step [200] has a uranium concentration below 100 mg/L uranium. Additionally, the uranium contained in the feed liquor [201] loads to in excess of 30 g/L uranium, but typically to concentrations of 60 g/L or more on the ion exchange resin [205].

A barren stream [202] from the IX Extract step [200] is collected in an Ion Exchange Barren Tank [203] from which a barren liquor [122] is returned to the Step 1 solvent extraction strip feed [123]. Typically the aqueous barren stream [202] from the IX Extract step [200] contains less than 50 mg/L (more typically less than 5 mg/L) of uranium. The sulfuric acid concentration in the barren stream [202] is similar in magnitude to that of the PSS feed liquor [201] to the IX Extract step [200],

Advantageously the barren stream [202] is not contaminated with any organic phase since the method of the invention incorporates an ion exchange resin and not a solvent extraction process.

The loaded ion exchange resin [205] is then preferably washed in an Ion Exchange (IX) Wash step [210] employing a wash liquid [21 1] stream. Preferably the wash liquid [21 1] is at a temperature similar to that of the loaded resin advancing to IX Wash step [210]. Preferably the wash liquid [21 1 ] is a displacement wash of a strip raffinate feed in the IX Extract Step [200] that is in contact with the resin beads at the time the loaded resin is moved on to the IX Wash step [210]. The wash liquid [21 1] is preferably any liquid compatible with the overall aims and objectives of the ion Exchange circuit [200/210/220]. Typically the wash liquid [211] is wash water.

Preferably the spent wash [212] from the IX Wash step [210] is recycled to the feed liquor [201] of the IX Extract step [200]. Preferably the IX Wash step [210] is made continuously. Preferably the washed resin [213] from the IX Wash step [210] is subject to a deprotonation step in which protons, on any cationic functional groups on the resin, are removed by an alkaline fluid. Typically the washed resin [213] is deprotonated by a weak sodium hydroxide containing solution. The washed ion exchange resin [213] is then preferably eluted in an Ion Exchange (IX) Elution step [220] to produce an eluate [224]. Preferably an eluant stream [223] applied to the resin in the IX Elution step [220] is at least an aqueous solution of sodium carbonate. Optionally, the eluant stream [223] is a blend of sodium carbonate and sodium bicarbonate. Advantageously the eluant stream [223] is at least sodium carbonate consisting primarily of a recycled solution [222] and some make-up [221 ]. Preferably the concentrations of sodium carbonate in the eluant stream [223] are chosen to optimise the concentration of uranium in the eluate [224] and facilitate its complete removal from the resin [220]. The elution chemistry is typically that reflected in:

(ResPO₃)₂U0₂(r) + 3Na₂C0₃(aq)→ ResPO₃Na₂(r) + Na₄UO₂(CO₃)₃(aq) 2ResPO₃H(r) + 2Na₂CO₃(aq)→ 2ResPO₃.Na(r) + 2NaHCO₃

2NaHCO₃{aq) + 2NaOH(aq)→ 2Na₂CO₃(aq) + 2H₂O(aq)

Optionally, the IX Elution step [220] can be performed in two or more separate process steps. Typically the IX Elution step [220] is a continuous process. Alternately, the elution process in IX Elution step [220] is a batch- continuous process to assist in maximising the concentrations of uranium in the eluate [224]. Preferably the elution process in the IX Elution step [220] is conducted at temperatures between 20°C and 60°C. The eluate [224] is preferably received by an Elution tank [230] from which a stream [231] is fed batch or continuously to an intermediate uranium product [232] recovery step (Step 3 describe below).

Typically the eluted resin [225] from the IX Elution step [220] is finally washed with a dilute sulfuric acid [226] to produce a wash barren liquor [227], Optionally it may be necessary to bleed the acidic solution employed in the IX Extract Step [200] to remove any impurity build up. Preferably the acidic bleed [226] derived from the bleed above is diluted with water.

Advantageously the acid washed resin [228] (also marked [204]) is recycled to the IX Extract circuit [200] where the load-wash-elution-wash cycle is repeated.

Step 3 in Figure 1

The eluate [231] from Elution Tank [230] is fed batch or continuously to an intermediate uranium product recovery step [232], and is processed either continuously or on a batch basis. The eluate is typically a rich blend of sodium uranyl tricarbonate, sodium carbonate and sodium bicarbonate. Small quantities of sulfate may also be present. The intermediate uranium product recovery step is typically a sodium diuranate precipitation step [232]. The sodium diuranate precipitation step [232] is typically performed by employing sodium hydroxide [233] and maintaining an excess of sodium hydroxide of approximately 1 to 2%. The sodium diuranate precipitation step [232] is preferably done by employing a seed recycle process.

The precipitation of sodium diuranate requires a free sodium hydroxide concentration of approximately 0.5 to 1%, and produces an aqueous blend of sodium carbonate and sodium hydroxide which is what the eluant stream [223] is composed of. 2Na₄U0₂(C0₃)₃(aq)+6NaOH(aq)+3H₂0(aq)→Na₂U₂0₇.6H₂0(s)+ Na₂C0₃(aq)

Consequently, the present invention is considerably more efficient in the employment of reagents required for the recovery of uranium. The sodium diuranate precipitate [234] from the sodium diuranate precipitation step [232] is separated in a Separation step [236] using a filter or some equivalent process equipment. The solid fraction [237] is appropriately washed with wash water [238] to remove most of the aqueous solute. The solute in the separated filtrate or aqueous [239] contains recyclable sodium hydroxide and sodium carbonate which is recycled [222] as the primary component of the sodium carbonate eluant stream [223] applied to the resin in the IX Elution step [220] in Step 2.

A minor bleed stream [240] is required to remove impurities and unwanted ions from the circuit. The sodium diuranate product can be sold or processed further to a uranium oxide concentrate.

The preferred embodiment of the method of the invention will be better understood from the following illustrative examples:

EXAMPLE 1

The strip raffinate [124] from the solvent extraction process (Step 1) had the following assay as summarised in Table 1 below:

Table 1

alL

u 5.878

Fe 0.24

Ca 0.26

Na 0.46

H₂S0₄ 387 This strip raffinate [124] became the feed to an ion exchange resin that had been conditioned in strong sulfuric acid rendering the resin in the protonated form. The resin employed in this example was the Lanxess TP260 resin (aminomethyl phosphonic functional group). The feed rate was between 1 and 2 bed volume per hour (BVH). The temperature of the extraction step was 25 - 30°C. The uranium assay of the barren liquor emanating the first and second column in a multi column lead-lag arrangement was:

Table 2

Cumulative Column 1 Exit Column 2 Exit

Bed Volumes U mg/L U mg/L

0

4 0.0950 0.0150

4 0.0200 0.0050

4 0.0050 0.0200

5 0.0150 0.0050

6 637* 0.0200

7 284* 0.0250

7 123* 0.0050

8 23.1* 0.0150

9 9.01^* 0.0300

9 3.59 0.0600

10 2.69 0.0250

10 11.3* 0.0250

1 1 3.17 2.1600

1 1 79.1* 0.1150

12 8.13 0.1100

12 6.10 0.0500

13 22.7 0.0300

13 268 0.0 50

14 1047 0.0200

15 2680 0.0200

15 174 0.3250

16 206 0.9100 16 1541 0.4400

17 5001 0.5300

18 5338 15200

19 5171 0.4700

19 5745 0.2200

20 6080 0.5250

Possibly attributable to channelling

The composite barren liquor sample assay from Column 2 is given in Table 3 below:

Table 3

When the first column had been fully loaded, it was taken out of service, washed and then eluted. The eluant [223] was 300 g/L sodium carbonate liquor and the elution temperature was 25 to 30°C.

The elution profile is detailed in Table 4 below:

Table 4

Cumulative U Concentrate

Bed Volumes mg/L

0

1 55360

1 70399 1 69422

1 53052

2 29362

3 6371

4 8733

5 134

6 125

On completion of the elution, the resin was washed and returned to service. The composition of the concentrated eluate [224] is given in Table 5 below:

Table 5

EXAMPLE 2

The strip raffinate [124] from an integrated solvent extraction and ion exchange operation had the following composition as detailed in Table 6 below:

This strip raffinate [124] was fed to the ion exchange columns arranged in series (Lead, Trail 1 and Trail 2). The resin had been conditioned in strong sulfuric acid. The chelating resin TP260 was again employed in this example. After the load step the resin was:

• washed

• deprotonated

• washed prior to elution

· eluted in two stages

• washed free of eluant

• acid wash in preparation for the load step.

Table 7 provides more detail on each of the above operating steps.

The composition of the streams from the process steps in Table 7 is given below in Table 8.

Table 8

Fe Mo Na P S U

Loading + Wash 1 Mg/L Mg/L Mg/L Mg/L Mg/L Mg/L

Feed (bulk loaded strip D) 9.5 14 14 4.0 123, 194 4,165

Bulk Barren, #2 <1 <1 17 1.0 108, 159 <1

IX Barren after Lead (4.5h) <1 <1 16 1.3 1 17,268 <1

IX Barren after Lead (12h) 9.2 4.3 5.6 1.7 125, 140 3,649

IX Barren after Lag I (12h) 2.0 <1 15 1.6 122,938 <1

Spent wash 1 A (Lead + Lag I) 1.6 <1 15 1.4 106,838 <1

Deprotonation

Spent iiq. after deprotonation <1 1.4 568 <1 724 5.9

Elution/Wash

Elution Stage 1 4.8 133 39,508 51 55 27,836

Elution Stage 2 <1 1.7 70,210 6.6 7.5 466

Spent protonation liquor <1 <1 1 1 ,679 1.9 15,571 <1 The eluate stream [224] from Stage 1 contained approximately 28g/L of U and this was the feed to the recovery circuit (Step 3), The resin at different stages in the ion exchange process was analysed and found to contain the 5 elements shown in Table 9 below.

pro ona on

Quite clearly the elution process is adequate to remove the uranium and essentially all known impurities to very low levels thus rendering the resin 10 recyclable and presentable to the extraction step.

Now that a preferred embodiment of a method for the recovery of uranium from a Pregnant Strip Solution (PSS) has been described in detail, it will be apparent that the described embodiment provides a number of advantages 15 over the prior art, including the following:

(i) The PSS barren liquor after loading the uranium on the ion exchange resin contains a majority of the sulfuric acid that was originally present in the feed liquor and can be recycled to the upstream uranium recovery process (solvent extraction or ion exchange).

20 (ii) The PSS barren liquor is not contaminated with any organic phase since the method of the invention incorporates an ion exchange resin and not a solvent extraction process.

(iii) The method is considerably more efficient in the employment of reagents required for the recovery of uranium.

25 It will be readily apparent to persons skilled in the relevant arts that various modifications and improvements may be made to the foregoing embodiment, in addition to those already described, without departing from the basic inventive concepts of the present invention. Therefore, it will be appreciated that the scope of the invention is not limited to the specific embodiments described.

Claims

1. A method for the recovery of uranium from a Pregnant Strip Solution (PSS) emanating from an upstream uranium recovery process, the method comprising the steps of: extracting the uranium values quantitatively from the PSS by loading an ion exchange resin in a separate ion exchange extracting step; and, returning the barren liquor from the extracting step to the upstream uranium recovery process with substantially all of the acid that was in the PSS.

2. A method for the recovery of uranium as defined in claim 1 , further comprising the step of washing the loaded resin with a suitable wash liquid in an ion exchange wash step.

3. A method for the recovery of uranium as defined in claim 2, further comprising the step of quantitatively eluting the loaded resin with a reagent in an ion exchange eluting step wherein, when the uranium has been removed, the resin can be recycled to the ion exchange extracting step with minimal loss in its activity.

4. A method for the recovery of uranium as defined in any one of the preceding claims, wherein the resin is loaded in the extracting step to in excess of 30 g/L uranium.

5. A method for the recovery of uranium as defined in any one of the preceding claims, wherein the ion exchange resin is a chelating resin of the type containing, for example, an amino phosphonic or a phosphonic/sulfonic functional group.

6. A method for the recovery of uranium as defined in claim 5, wherein the chemistry in the extracting step can be shown as:

H₄U0₂(S0₄)3(aq) + ResP0₃H₂(r)→ (ResP0₃)₂U0₂(r) + 3H₂S0₄(aq)

7. A method for the recovery of uranium as defined in claim 1 , wherein the PSS contains strong sulfuric acid at concentrations in excess of 50 g/L and below 600 g/L.

8. A method for the recovery of uranium as defined in claim 7, wherein the PSS contains free sulfuric acid at concentrations of between 50 g/L and

400 g/L.

9. A method for the recovery of uranium as defined in claim 3, wherein an eluant stream applied to the resin in the eluting step is at least an aqueous solution of sodium carbonate. 10. A method for the recovery of uranium as defined in claim 9, wherein the eluant stream is a blend of sodium carbonate and sodium bicarbonate.

1 1. A method for the recovery of uranium as defined in claim 10, wherein the eluant stream is at least sodium carbonate consisting primarily of a recycled solution and some make-up. 12. A method for the recovery of uranium as defined in any one of claims 9 to 11 , wherein the concentrations of sodium carbonate in the eluant stream are chosen to optimise the concentration of uranium in the eluate and facilitate its complete removal from the resin.

13. A method for the recovery of uranium as defined in claim 12, wherein the elution chemistry is that reflected in:

(ResP0₃)₂U0₂(r) + 3Na₂C0₃(aq)→ ResP0₃Na₂(r) + Na₄U0₂{C0₃)₃(aq) 2ResP0₃H{r) + 2Na₂C0₃(aq)→ 2ResP0₃.Na(r) + 2NaHC0₃

2NaHC0₃(aq) + 2NaOH(aq)→ 2Na₂CO_s(aq) + 2H₂0(aq)

14. A method for the recovery of uranium as defined in claim 3, wherein the eluted resin from the eluting step is washed with a dilute sulfuric acid, to produce an acid washed resin and a wash barren liquor.

15. A method for the recovery of uranium as defined in claim 14, wherein the acid washed resin is recycled to the ion exchange extracting step where the !oad-wash-elution-wash cycle is repeated. 6. A method for the recovery of uranium as defined in claim 12, wherein the eluate from the eluting step is fed batch or continuously to an intermediate uranium product recovery step. 7. A method for the recovery of uranium as defined in claim 6, wherein the eluate is a rich blend of sodium uranyl tricarbonate, sodium carbonate and sodium bicarbonate. 18. A method for the recovery of uranium as defined in claim 16, wherein the intermediate uranium product recovery step is a sodium diuranate precipitation step.

19. A method for the recovery of uranium as defined in claim 8, wherein the sodium diuranate precipitation step is performed by employing sodium hydroxide and maintaining an excess of sodium hydroxide concentration of approximately 0.5 to 2%.

20. A method for the recovery of uranium as defined in claim 2, wherein the wash liquid employed in the ion exchange wash step is at a temperature similar to that of the loaded resin advancing to the ion exchange wash step. 21. A method for the recovery of uranium as defined in claim 20, wherein the wash liquid is a displacement wash of a strip raffinate feed in the extracting step that is in contact with the resin at the time the loaded resin is advanced to the wash step.

22. A method for the recovery of uranium as defined in claim 21 , wherein the spent wash from the wash step is recycled to the PSS.

23. A method for the recovery of uranium as defined in claim 2, wherein the washed resin from the wash step is subject to a deprotonation step in which protons, on any cationic functional groups on the resin, are removed by an alkaline fluid.

24. A method for the recovery of uranium as defined in claim 23, wherein the washed resin is deprotonated by a weak sodium hydroxide containing solution.

25. A method for the recovery of uranium as defined in claim 16 or claim 17, wherein the sodium diuranate precipitate from the sodium diuranate precipitation step is separated from a filtrate in a separation step using a filter or some equivalent process equipment.

26. A method for the recovery of uranium as defined in claim 23, wherein a solid fraction of the sodium diuranate is appropriately washed with wash water to remove most of the aqueous solute.

27. A method for the recovery of uranium as defined in claim 26, wherein a solute in the separated filtrate contains sodium hydroxide and sodium carbonate and is recycled as a primary component of the sodium carbonate eiuant stream applied to the resin in the eluting step.