GB2326268A - Recovery of uranium carbonato complex by ion flotation - Google Patents

Abstract

A process for recovering uranium from an aqueous medium containing the uranium as a dissolved carbonato complex, which process comprises an ion flotation process comprising dissolving in the medium an agent which has surfactant properties and causes the uranium to enter the solid phase, bubbling a gas through the medium to float the solid phase uranium to a surface region of the medium, then separating the solid phase uranium from the bulk of medium.

Description

RECOVERY PROCESS The present invention relates to the recovery of actinides from irradiated nuclear fuel into a dissolver liquor, and more particularly to a process for recovering uranium from such a dissolver liquor.

It is proposed in EP-A-282810 to dissolve MOX fuel [MOX: mixed oxide (UO2 and PuO2)] by treating the fuel with an aqueous solution containing HCO3 and CO32-.

The aqueous solution or dissolver liquid optionally contains an oxidant as a solvent aid. Cerium (IV) is described as a suitable oxidant and presumable serves to oxidise the UO2 into the more soluble ion UO22+.

Ion flotation is a technique in which dissolved ions are contacted with an agent which acts as a surfactant and which forms an insoluble product with one or more selected types of ion. Gas, e.g. nitrogen or air, is bubbled through the solution. The bubbles pick up the solid products as they rise to the top of the solution, where a foam usually forms. The solid is then separated from the solution, typically by means of foam overflowing a weir.

The present invention relates to a process for extracting dissolved uranium from an aqueous carbonate-containing medium, for example a dissolver liquor obtained by following the procedure of EP-A-2828 10 or a similar procedure.

According to the present invention ion flotation is used to recover an actinide carbonato complex from solution, especially as part of a selective process in which uranium is preferentially recovered by ion flotation from a liquid medium containing uranium and plutonium (IV) before the plutonium is extracted by, for example, a known technique, eg solvent extraction.

A process provided by the invention, therefore, is for recovering uranium from an aqueous medium containing uranium as a dissolved carbonato complex, which process comprises an ion flotation process comprising dissotvlY {nSmedium an agent which has surfactant properties and causes the uranium to enter the solid phase, bubbling a gas through the medium to float the solid phase uranium to a surface region of the medium, then separating the solid phase uranium from the medium.

The uranium is usually uranium (VI) and the uranium (VI) carbonato complex typically comprises tricarbonatouranylate. The surfactant agent is normally an alkyltrimethyl-ammonium bromide, preferably in at least stoichiometric quantity.

The extraction from process may be preceded by a process for dissolving uranium ions derived from irradiated nuclear fuel, which process comprises optionally heating the irradiated fuel to oxidise uranium therein and contacting the oxidised fuel with an aqueous carbonate solution. The carbonate is preferably ammonium or sodium carbonate or, less preferably, another alkali metal carbonate or a mixture of any of the aforegoing. A suitable concentration range for ammonium carbonate is 10 to 200 g/l; ammonium carbonate is the most preferred carbonate.

The aqueous carbonate solution may contain one or more oxidising and/or complexing agents. As possible oxidising agents may be mentioned hydrogen peroxide and cerium (IV) salts, optionally used in combination. Suitable complexing agents are, amongst others, nitrate salts, notably ammonium or sodium (or other alkali metal) nitrate.

Normally, the irradiated fuel contains UO2 which is oxidised to U308 during the heating of the fuel as well as, usually, plutonium (IV), the latter of which preferably remains unoxidised by the heating of the fuel but which is dissolved into the aqueous carbonate solution.

The present invention is further described by way of example only with reference to the accompanying drawings, in which: Figure 1 shows the absorption spectra of a mixed uranium/plutonium carbonate solution before and after treatment with a surfactant useful in the invention; Figure 2 is a representative absorption spectrum for uranium (VI) in aqueous carbonate solution; and Figure 3 is a graphical representation of the percentage recovery of uranium with time.

In a preferred embodiment of the invention irradiated UO2 or MOX (UO2 and PuO2) fuel is received in the head-end and transferred to a furnace, where the fuel is heated to a temperature of at least 480 "C in an oxygen-containing gas in order to oxidise the UO2 to U3 08. The plutonium remains as PuO2, In one process, air at a temperature of 500 "C or higher is pumped into the furnace to heat the fuel and, when this temperature has been reached, air at ambient temperature is pumped into the furnace.

The fuel will typically include cladding, and this will rupture as a result of the expansion of the fuel during oxidation. In some processes, however, the cladding is ruptured or removed mechanically before the fuel enters the furnace. The air cools the fuel to between ambient temperature and 100 "C. Of course, another oxygencontaining gas may be used in place of air.

The oxidised fuel is leached out of the cooled fuel by a dissolver liquor comprising aqueous carbonate solution. Preferably, aqueous carbonate solution is sprayed onto the fuel (which is often in the form of ruptured fuel assemblies), typically while the fuel remains in the furnace. The leaching process is usually performed at ambient pressure and using carbonate solution at a temperature of between ambient and 100 "C. The carbonate salt used is preferably ammonium carbonate, normally at a concentration of from 10 to 200 g/l. Alternatively, sodium or another alkali metal salt may be used, or even a mixture of alkali metal salts or a mixture containing both alkali metal and ammonium salts.

The aqueous carbonate solution may contain oxidising and/or complexing agents to assist dissolution of actinides. Exemplary oxidising agents are hydrogen peroxide and cerium (IV) salts, or a mixture thereof. Suitable complexing agents include nitrate salts, for example ammonium or alkali metal (especially sodium) nitrate, or a mixture thereof.

In a modification of the invention, the fuel is not initially heated to oxidise the UO2 the dissolution is carried out at, for example, ambient temperature. In this modification the use of complexing agents and oxidants is normal.

The result of the dissolution is normally a solution containing uranium, plutonium, neptunium, the minor actinides (americium and curium) and a fraction of the fission products. A certain percentage of the fission products will not be dissolved and will be routed to a suitable waste treatment process, for example a treatment process known per se. The concentration of uranium in the dissolver liquor will in many cases be between 10 and 50 g/l, with the concentrations of the other species typically being in proportion to the uranium concentration in dependance upon their relative proportions in the irradiated fuel.

The dissolver liquor from the head-end is fed to an ion flotation process, which is performed in an ion flotation cell. The ion flotation cell in practice consists of a tall vessel which is constantly fed with dissolver liquor. Alternatively, but less desirably, a batch process may be performed. The temperature of the dissolver liquor is maintained in the flotation cell. An agent with surfactant properties and capable of forming a solid adduct with the dissolved uranium is added to the dissolver liquor.

Typically, a surfactant (suitably dissolved in, for example, ethanol, water or a mixture thereof) is fed into the cell continuously and an insoluble adduct of the surfactant with uranium is formed. Any plutonium, and typically the minor actinides and a fraction of the fission products, remain in the solution phase. The surfactant used is suitably an alkyltrimethylammonium bromide and preferably dodecyltrimethyl-ammonium bromide (DTAB) or cetyl trimethyl ammonium bromide (CTAB), of which DTAB is most preferred. Any other chemical with surfactant properties could be used, provided that the added agent caused the uranium to enter the solid phase.

The agent in practice will comprise a cationic species used in at least stoichiometric amount. The stoichiometric amount is determined by the relative charges of the carbonato complex and the surfactant. Thus, four ions CH3(CH2)"NMe3+ stoichiometrically associate with Uo2(Co3)34-. Air (or some other gas, e.g.N2) is pumped into the bottom of the flotation cell and rises up through the liquor as bubbles. The solid adduct of uranium is floated to the surface of the liquor by the bubbles and a foam is formed. The foam is separated from the liquor. In one class of embodiments, the foam overflows a weir at the top of the cell and is collected in a vessel of some kind. The solids in the foam may be dissolved, for example, and either disposed of or processed further to recover the uranium.

Any plutonium remaining in the dissolver liquor may be extracted by, for example, a technique known per se, eg solvent extraction.

It will also be understood that the invention includes a process for recovering plutonium and uranium from irradiated nuclear fuel containing UO2 and Pu2, comprising: (a) optionally heating the fuel in the presence of a gas comprising oxygen to oxidise uranium to U308 without substantially oxidising plutonium; (b) contacting the fuel with aqueous dissolver liquor containing dissolved ammonium or alkali metal carbonate to cause plutonium and uranium to be dissolved in the liquor; (c) passing the resultant liquor into an ion flotation vessel; (d) introducing into the liquor in the ion flotation vessel a surfactant which forms a substantially insoluble product with uranium ions to cause such a product to form; (e) introducing a gas into the flotation vessel at a bottom regionthereof, whereby gas bubbles rise to the top of the liquor to form a foam and, in rising to the top of the liquor, collect the solid product; (f) removing the foam; (g) optionally removing the substantially uranium-free liquor from the ion flotation vessel; and (h) removing plutonium from said liquor.

Uranium and/or plutonium recovered by the invention may be subjected to one or more further processes, for example to form a nuclear fuel product, for example a fuel pellet or fuel rod.

EXAMPLES The behaviour of mixed uranium/plutonium/carbonate solutions in ion flotation has been investigated experimentally, in that the formation of an insoluble adduct of uranium with surfactants in aqueous carbonate solution has been explored. The formation of these insoluble adducts is a precursor to the removal of the elements from the solution phase by ion flotation.

Example Experimental A mixed uranium/plutonium/carbonate solution was prepared by adding 2.5ml of a plutonium (IV) nitrate solution (3 .99g(Pu)/ 1) to a solution of ammonium uranyltricarbonate (5.8g(U)/l). The resultant solution is analogous to that which would be produced in the head-end of the above described process.

The absorption spectrum of the solution was recorded from 400 to 500nm using a Perkin Elmer Lambda 19 UV/vis/nIR spectrophotometer.

An aliquot of a saturated solution of dodecyltrimethylammonium bromide (DTAB) in ethanol was added to the U/Pu solution such that the DTAB/uranium molar ratio was 24. The solution was observed for the formation of precipitate.

After standing for a few minutes, a sample of the supernate was filtered off using Whatman 0.2m polypropylene syringe filtration units. The absorption spectrum of the filtrate was recorded from 400 to 500nm.

Results and Discussion The addition of the solution of DTAB to the mixed U/Pu solution resulted in the formation of a pale yellow, finely divided precipitate. The spectra of the starting solution and the filtrate are shown in Figure 2. The four peaks labelled "U" are the characteristic absorptions of the uranyltricarbonate ion, while the peak labelled "Pu" is characteristic of plutonium (IV) in aqueous carbonate solution. It is apparent from the spectra that the majority of the uranium is removed from the solution when the DTAB is added, but that the bulk of the plutonium remains in solution. The peak to trough heights labelled Au and Apu have previously been shown to be proportional to the concentration of uranium and plutonium respectively. Measurement of these peak to trough heights in the spectra shown in Figure 2 indicates that 88.8% of the uranium has been lost from the solution phase, but only 1.6% of the plutonium has precipitated.

This experiment demonstrates conclusively that a high percentage of uranium can be selectively precipitated from mixed U/Pu-carbonate solutions by the addition of the surfactant DTAB, with almost all of the plutonium remaining in solution.

Example2 Experimental A stock solution of ammonium tricarbonatouranylate, (NH4)4UO2(CO3)3, was prepared by dissolving U03 in 80g/l ammonium carbonate solution. An aliquot of this U(VI) stock solution was diluted with 80g/1 ammonium carbonate to give a solution of 5.84g(U)/1.

1 00my of this diluted U(VI) solution were transferred to a 250ml Drechsel bottle fitted with a porosity 3 sparger. An aliquot of DTAB solution (221.6g/1 in ethanol) was added such that the DTAB/U ratio was 4. The solution was stirred with a magnetic follower for 15 minutes. After this equilibration period, a sample of the solution was withdrawn whilst stirring continued. The air supply was turned on and the time recorded as t=O. The air pressure was adjusted until foam was forced out of the air outlet pipe of the Drechsel bottle. The foam was collected in a measuring cylinder. Stirring of the solution was maintained throughout the experiment.

Additional samples of the liquid phase were taken at recorded time intervals.

All the samples were diluted by half with 80g/1 ammonium carbonate solution to dissolve any precipitated sublate contained therein. The absorption spectra of the solutions were recorded from 400-500nm using a Perkin-Elmere Lambda 19 UV/vis/nIR spectrophotometer.

Results and Discussion A typical absorption spectrum for U(VI) in aqueous carbonate solution is shown in Figure 3. The peak-trough height marked AU is proportional to U(VI) concentration.

Measurement of AU in the spectra of the samples taken during the U(VI) ion flotation experiment allowed the percentage recovery of uranium with time to be calculated. The results of these calculations are presented graphically in Figure 3.

The graph shows that a uranium recovery of almost 40% was achieved after 175 minutes of flotation.

Example3 Experimental 125 ml of 31.7 mg (Pu)/l solution was prepared by diluting an aliquot of plutonium (IV) nitrate solution with 80 g/l ammonium carbonate. The solution was transferred to a 250 ml Drechsel bottle equipped witth a porosity 3 sparger. A 5 ml aliquot of a DTAB solution (22.6 g/l in ethanol) was added. The solution was stirred magnetically for 15 minutes.

A sample of the solution was withdrawn and the air sparge was turned on.

Overflowing foam was collected in a measuring cylinder. Additional samples of the liquid phase were withdrawn at recorded time intervals and submitted for total a and a kick-sort analyses. Stirring was maintained throughout.

Results and Discussion The results of the Pu-active ion flotation experiment are summarised in Table 1. The overall plutonium recovery is minimal, and probably lies within the errors associated with the analyses.

Time (min) Alpha (Bq/ml) Pu239+240 Pu239+240 Pu recovery (%ala) (Bq/ml) (%) o 4.51E+04 56.0 2.53E+04 0.00 20 4.29E+04 57.2 2.45E+04 2.84 38 4.42E+04 56.1 2.48E+04 1.82 58 4.21E+04 57.9 2.44E+04 3.48

Conclusions U(VI) can be recovered from aqueous ammonium carbonate solution by ion flotation using the surfactant DTAB. Under similar conditions, the recovery of Pu (IV) from aqueous ammonium carbonate solution by ion flotation is not possible.

These results imply that a separation of U(VI) from Pu(IV) in aqueous ammonium carbonate solution could be achieved by ion flotation using the surfactant DTAB.

Claims (20)

1. A process for recovering uranium from an aqueous medium containing the uranium as a dissolved carbonato complex, which process comprises an ion flotation process comprising dissolving in the medium an agent which has surfactant properties and causes the uranium to enter the solid phase, bubbling a gas through the medium to float the solid phase uranium to a surface region of the medium, then separating the solid phase uranium from the bulk of medium.

2. A process of claim 1, wherein the uranium is uranium (VI).

3. A process of claim 2, wherein the uranium (VI) carbonato complex comprises tricarbonatouranylate (U02(CO3) 4,-).

4. A process of any of claims 1 to 3, wherein said agent is dodecyl trimethylammonium bromide (DTAB), cetyltrimethylammonium bromide or another alkyltrimethylammonium bromide.

5. A process of any of claims 1 to 4, wherein the aqueous medium contains uranium and plutonium (IV).

6. A process of any of claims 1 to 5, wherein said agent comprises a cationic species dissolved in an amount at least stoichiometric with respect to the dissolved uranium carbonato species.

7. A process of any of claims 1 to 6, wherein the solid phase uranium in the surface region of the medium is separated from the remainder of the medium by means of a weir which the surface region overflows.

8. A process of claim 7, wherein the solid phase uranium in the surface region is in the form of a foam.

9. A process for separating aqueous uranium (VI) from aqueous plutonium (IV), comprising: providing an aqueous medium containing uranium (VI), plutonium (IV), carbonate in an amount sufficient for the uranium to have formed carbonato complexes, and an agent which has surfactant properties and in the presence of which the uranium carbonato complexes but substantially not the plutonium enter the solid phase; bubbling a gas through the medium to float the solid phase uranium to a surface region of the medium; then separating the solid phase uranium from the medium.

10. A process of claim 9, which further includes the feature(s) recited in one or more of claims 3, 4, 6, 7 or 8.

11. A process of claim 9 or claim 10, wherein the aqueous medium is provided by contacting irradiated nuclear fuel with an aqueous carbonate solution to dissolve uranium and plutonium in the solution.

12. A process of claim 11, wherein the irradiated fuel is heated to oxidise the uranium therein before being contacted with the aqueous carbonate solution and/or the aqueous carbonate solution contains an oxidising agent.

13. A process of claim 12, wherein the oxidising agent comprises hydrogen peroxide or a cerium (IV) salt, or both.

14. A process of claim 12 or claim 13, wherein the aqueous carbonate solution further contains a nitrate salt or another complexing agent.

15. A process of any of claims 11 to 14, wherein the aqueous carbonate solution is a solution of ammonium carbonate or an alkali metal carbonate.

16. A process of claim 15, wherein the aqueous carbonate solution is a solution of ammonium carbonate at a concentration of from 10 to 200g/l.

17. The use to extract uranium (VI) from aqueous solution in a carbonatecontaining medium of an ion flotation process involving dodecyltrimethylammonium bromide as surfactant forming a solid adduct containing uranium (VI).

18. A process for recovering plutonium and uranium from irradiated nuclear fuel containing U02 and Pu02 comprising: (a) optionally heating the fuel in the presence of a gas comprising oxygen to oxidise uranium to U308 (b) contacting the fuel with aqueous dissolver liquor containing dissolved ammonium or alkali metal carbonate to cause plutonium and uranium to be dissolved in the liquor; (c) passing the resultant liquor into an ion flotation vessel; (d) introducing into the liquor in the ion flotation vessel a surfactant which forms a substantially insoluble product with uranium species in the liquor to cause such a product to form; (e) introducing a gas into the flotation vessel at a bottom region thereof, whereby gas bubbles rise to the top of the liquor to form a foam and, in rising to the top of the liquor, collect the solid product; and (f) removing the foam.

19. A process of claim 18, wherein the nuclear fuel is contained in a cladding and step (a) is performed, whereby expansion of the uranium resulting from its oxidation causes the cladding to rupture.

20. A process of any of claims 1 to 19, which further comprises subjecting recovered uranium to one or more further processes, for example to form a nuclear fuel product.