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(11) | EP 0 066 988 A2 |
| (12) | EUROPEAN PATENT APPLICATION |
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| (54) | Method of recovering uranium |
| (57) Method of recovering uranium from a silicate-uranium coprecipitate by leaching the
coprecipitate with a leachate having a pH of from 2 to 3, followed by filtering the
coprecipitate. The uranium is dissolved in the filtrate but the silicate is insoluble.
The leachate is preferably an aqueous solution of nitric acid which is also used to
wash the coprecipitate. The uranium can be extracted from the filtrate using an organic
extractant containing DEPA-TOPO. |
[0002] In an ammonium diuranate conversion process for producing uranium oxide powder, a waste stream is produced which contains uranium, fluoride, ammonium, and nitrate ions. To recover the ammonia and lower the fluoride levels, a calcium hydroxide or lime slurry is added which precipitates calcium fluoride. The ammonium diuranate waste stream is processed in an ammonia stripping column and the calcium fluoride slurry is sent to settling lagoons were excess water is decanted and run off. Some of the uranium remains in the calcium fluoride slurry as insoluble calcium uranate.
[0003] The calcium uranate waste creates an expensive disposal problem and is a loss of a valuable resource. The quantity of uranium contaminated waste can be reduced by adding sodium silicate to the uranium waste stream prior to the addition of calcium hydroxide or lime as is described in Japanese Patent Specification No. 48-38320. This results in a silicate-uranium coprecipitate of significantly less volume than the calcium fluoride precipitate. The silicate-uranium coprecipitate is then disposed of by storage in drums.
[0004] Accordingly, the present invention resides in a method of recovering uranium from a silicate-uranium coprecipitate which comprises leaching said coprecipitate with a leachate having a pH of from 2 to 3 followed by filtering said coprecipitate.
[0005] Not only is uranium recovered that would otherwise be wasted, but it is recovered in a form which can be processed in a standard solvent extraction cycle. Moreover, if desired, sufficient uranium can be removed from the coprecipitate to permit its disposal as a non-nuclear waste material.
[0006] A silicate-uranium coprecipitate is typically formed by adding a solution of water glass to an ammonium fluoride solution containing uranium in order to concentrate the uranium in the coprecipitate.- A typical coprecipitate, for example, may contain about 15% (all percentages herein are by weight) by weight sodium silicate, about 85% water, and from 1,000 to 30,000 parts per million (ppm) of uranium, probably in the form of a uranyl silicate. The typical coprecipitate is a toothpaste-like solid containing large amounts of bound water. The invention, however, will work with any silicate-uranium coprecipitate containing virtually any amount of uranium. It is preferable to wash the coprecipitate first in order to minimize the fluoride content in it as fluorides tend to attack silicates and increase their dissolution, thus reducing the degree of separation of the uranium.
[0007] The first step involves lowering the pH of the coprecipitate to from 2 to 3. It has been found that this pH range is essential to the successful operation of the invention because below a pH of 2 the silica begins to dissolve which interferes with the subsequent solvent extraction of the uranium from the filtrate, and above a pH of 3, the uranium is not solubilized. The optimum pH has been found to be about 2.3. The leachate is an inorganic acid which can solubilize the uranium within the specified pH range. Nitric acid is the preferred leachate as it has been found to work very well and it is compatible with the processes which follow the process of this invention, but hydrochloric or sulfuric acid could also be used if desired. If a nitric acid leachate is used, it is preferably from 5 to 20% aqueous nitric acid. The weight ratio of coprecipitate to leachate is preferably from 1:1 to 3:1.
[0008] After the leachate has been mixed with the coprecipitate, it is desirable to filter as soon as possible, preferably within one-half hour, as the silica will gradually dissolve in the leachate and silica which does not dissolve will become hydrolyzed and difficult to filter.
[0009] Once the leachate has been filtered from the coprecipitate, it is preferable to wash the coprecipitate several times in order to maximize the recovery of the uranium that is present in it. The wash is preferably performed with from 5 to 20% aqueous nitric acid. From 4 to 7 washes are usually satisfactory as fewer washes will leave uranium behind in the coprecipitate and more washes will dilute the filtrate and dissolve more silica. The weight ratio of coprecipitate to each wash is preferably from 0.5:1 to 1.5:1 as less wash will not recover all the uranium that could be recovered and more wash will dilute the filtrate.
[0010] The uranium in the filtrate can be recovered by any of a variety of well-known techniques. The preferred technique is solvent extraction using an organic extractant containing di-2-ethylhexyl phosphoric acid and trioctyl phosphine oxide (DEPA-TOPO) as that method is very effective.
[0011] If desired, further washes or higher concentrations of acids in the leachate can be used to reduce the uranium in the coprecipitate to a level sufficient for disposal as a non-nuclear waste. However, it may be less expensive to dispose of the small quantity of coprecipitate as a nuclear waste than to reduce its uranium content to such a low level.
EXAMPLE 1
[0013] A silicate uranium coprecipitate was mixed with a nitric acid leachate in a leach tank. After filtering, for example, in a pressurized rotary filter, a wash was used which resulted in a liquid filtrate, leaving behind the solids.
EXAMPLE 2
[0014] A silicate uranium coprecipitate was prepared by mixing two compositions. The first composition contained 2% ammonium fluoride, 4% ammonia (added as ammonium hydroxide), 91% water, and 15 ppm uranium (added as uranyl nitrate). The second composition contained 6% silica (added as sodium silicate) and the remainder water. Ninety-nine parts of the first composition were mixed with one part of the second composition and the mixture was stirred for one minute and filtered.
[0015] In these experiments, various concentrations of a nitric acid leachate were poured over samples of the coprecipitate, stirred, and quickly filtered. The following table gives the results of three experiments where the concentration of silica leached by the nitric acid was determined.
[0016] The above table shows that only small quantities of silica were leached using 10 wt.% nitric acid and that much larger quantities of silica were leached using 20 wt.% nitric acid. It has been experimentally determined for the particular coprecipitate being tested that 10 wt.% nitric acid results in a pH of 2.3 (the pH obtained using 10 wt.% nitric acid, however, will depend on the particular composition of the coprecipitate used.)
[0017] In the next series of experiments, the amount of uranium leached under various conditions was determined. The following table gives the conditions and results.
[0018] Test No. 1 shows that this process will work well at nitric acid levels as high as 20%. However, since satisfactory results are obtained at 10% nitric acid (Test No. 4), the added nitric acid expense is not economically justified.
[0019] Test No. 2 shows that not all of the uranium is recovered when the pH is too high. Test No. 3 shows that a low recovery is obtained with too much nitric acid, and the pH is too low. Test No. 4 shows that better results are obtained when the coprecipitate is washed to remove fluoride first. Test No. 5 shows that good results can be obtained even without washing to remove-fluoride if the nitric acid concentration is not too high. Test No. 6 shows that an increased leach solution to cake ratio dilutes the final uranyl nitrate stream.
1. A method of recovering uranium from a silicate-uranium coprecipitate characterized
by leaching said coprecipitate with a leachate having a pH of from 2 to 3 followed
by filtering said coprecipitate.
2. A method according to claim 1, characterized in that the leachate is aqueous nitric
acid.
3. A method according to claim 2 characterized in that the nitric acid is from 5 to
20%.
4. A method according to claim 2 or 3, characterized in that, after leaching, the
coprecipitate is washed with from 5 to 20% nitric acid.
5. A method according to claim 4, characterized in that the coprecipitate is washed
4 to 7 times.
6. A method according to any of claims 1 to 5, claim 1 characterized in that the weight
ratio of the coprecipitate to the leachate is from 0.5 to 1 to 1.5 to 1.
7. A method according to any of claims 1 to 5, characterized in that the weight ratio
of the coprecipitate to the leachate is from 1 to 1 to 3 to 1.
8. A method according to any of the preceding claims, characterized in that the coprecipitate
is filtered within one-half hour after the leaching.
9. A method according to any of the preceding claims, characterized in that the coprecipitate
is formed by the addition of water glass to an ammonium fluoride solution containing
uranium.
10. A method according to any of the preceding claims, characterized in that in an
additional last step uranium in the filtrate is extracted with a DEPA-TOPO extractant.