The present invention relates to a method for reducing the amount of uranium in a solution containing uranium and an excess of anions, apparatus for the performance of the method and a precipitate formed by such performance.

Uranium is an important naturally occurring radioactive element and is widely distributed in most rock types, ground water, ocean water, living matter and meteorites. Natural uranium is a mixture of three isotopes, 0.00572 U ²³ \ 99.27392 U ²³⁸ and 0.7 042 U ²³⁵ . Uranium is widely used in nuclear research establishments, in the fuel industry and for other purposes associated with its high density. Enriched uranium, in which the amount of fissible isotope U ²³⁵ has been increased, is particularly suitable for use in civilian nuclear power reactors and military weapons. Nuclear fuel reactors utilise low enriched uranium, in which the U ²³⁵ content has been raised from 3 to 52 while highly enriched weapons-grade uranium contains more than 902 U ²³⁵ .

Uranium mining, transport, utilisation, storage and disposal must all be monitored and controlled because of the radioactive properties of the metal. Disposal of uranium and its compounds is often a problem, particularly the disposal of by products and contaminated waste solutions generated during the use of uranium. Even methods of measuring and monitoring uranium use, purity and disposal generate contaminated solutions which need to be disposed of safely. One method of disposal is to cement and store the solutions, which has the major drawbacks of expense and an ever increasing storage requirement. The disposal of natural and depleted uranium is less expensive than disposal of enriched uranium because the levels of radioactivity associated with them are much lower than those of enriched uranium.

Accurate methods of measuring the purity of uranium are also important

to ensure the efficiency and safety of its uses, this is particularly true for enriched uranium. One widely used method of measuring uranium purity is the Davies and Gray titration which results in a contaminated waste solution containing a known amount of uranium and other ions added during titration, in a concentrated phosphoric acid solution. Disposal of this waste solution is a particular problem in the nuclear industry because standard methods for the extraction of uranium cannot be employed due to the high ratio of phosphate ions to uranium. The current practice is therefore to cement and store the solutions but although each titration results in a limited volume of waste this must be stored indefinitely because of the high half life of uranium. Such practice results in continually increasing requirements for storage facilities and associated cost increases.

Despite these problems, the relatively small volumes of such solutions does not justify the development of a system of treatment requiring large, elaborate or costly equipment or which would have to be performed by particularly skilled personnel.

It is an aim of the present invention to provide a method by which the uranium can be removed from solutions containing high levels of anions such as phosphates. It is a further aim of this invention to produce solutions which may be treated by standard decontamination techniques from a solution containing high levels of anions such as phosphates by the method described herein. Yet a further aim of the invention is to provide material comprising uranium containing a precipitate obtainable by any one of the methods described herein. Furthermore it is an aim of the invention to provide methods of treating waste solutions for the above purposes which may be performed using standard laboratory equipment by relatively unskilled personnel. A still further aim of this invention is to provide apparatus suitable for the performance of the aforesaid methods.

A first aspect of the invention provides methods for reducing the

amount of uranium in a solution containing uranium and an excess of anions with a pH value of less than 0.5 comprising adding a quantity of ammonium salt to produce a molar excess of salt with respect to anions and increasing the pH to between 1.5 and J .0 to precipitate the uranium, wherein the concentration of uranium in solution is initially 0.042 moles or less per litre.

A solution containing uranium and an excess of anions will be understood to be any solution of uranium in which anions are also present at a concentration greater than that of the uranium. Preferably the solution treated comprises an excess of anions to uranium greater than 800:1. The solutions may further comprise additional ions eg. iron, vanadium, chromium and molybdenum ions. One preferred example of a solution that may be treated by the method of the present invention is a 3 solution of phosphoric acid. A preferred solution to be treated is that resulting from a Davies and Gray titration used to determine uranium purity. One stage of this titration involves the use of a concentrated solution of phosphoric acid resulting in a waste solution with a high ratio of phosphate ions to uranium. Such solutions are difficult to treat by conventional methods because of this high phosphate to uranium ratio.

The concentration of uranium in the solution to be treated is initially 0.042 moles or less per litre, preferably 0.0084 moles or less per litre and more preferably 0.0042 moles or less per litre. The solution to be treated has a pH value of 0.5 or less before treatment, preferably 0.1 or less.

The quantity of uranium remaining in solution after the treatment will suitably be 152 or less the quantity of uranium in solution before treatment, preferably, 12J. or less, 52 or less, or most preferably 2% or less the quantity of uranium in solution before treatment.

Preferred methods of the invention are those in which the pH of the

solution is raised to pH 1.5 - 7-00, preferably between the ranges 1.8 -2.5 and 2.6 - 6.5. more preferably to pH 2.0 or to pH 3.0 - 6.0. Examples of bases which may be added to alter the pH will be apparent to the person skilled in the art, eg. sodium hydroxide.

The ammonium salt added may be any suitable salt, preferred salts being ammonium chloride and ammonium sulphate. Ammonium salt is added to provide a molar excess of salt added to anions of at least 1.6, preferably at least 2.0 and most preferably at least 2.5-

It will also be apparent to the person skilled in the art that ammonium salt and base added to increase the pH may be added in either liquid phase or solid phase, and may be in the same or different phases, subject to practical considerations. Eg. when utilising ammonium chloride and sodium hydroxide both phases are suitable. Although it is advantageous to keep the volume of solutions to a minimum it is likely to be impractical to add both base and salt in the solid state, as to do so is likely to result in heat generation causing water evaporation and crystallisation upon cooling. One embodiment of the method of the present invention comprises adding sodium hydroxide in liquid phase, a more preferred embodiment comprises adding ammonium salt in solid phase and sodium hydroxide in liquid phase.

In order to allow the precipitate to form the solution is allowed to stand for at least 3 days, preferably 4 days, more preferably 7 days and most preferably 14 days. The method may also be performed by raising the pH in two steps, eg. an initial increase in pH to 5-0 and then an increase to pH 10.0. The method may also be performed in three steps, eg. an initial increase to pH 2.1, followed by an increase to 5-0 and a final increase to 10.0.

A second aspect of the present invention provides a material comprising uranium containing a precipitate obtainable by the

performance of any one of the methods described in the first aspect of the invention. This aspect provides a material comprising a uranium containing precipitate obtainable by reducing the amount of uranium in a solution containing uranium and an excess of anions with a pH value of less than 0.5 by adding a quantity of ammonium salt to the solution to produce a molar excess of salt with respect to anions and increasing the pH to between 1.5 and 7-0 to precipitate the uranium, wherein the concentration of uranium in solution is initially 0.042 moles or less per litre.

The material is preferably obtainable from a solution of phosphoric acid, more preferably 3M phosphoric acid, and the concentration of uranium is initially 0.042 moles or less per litre. The material is preferably obtainable by raising the pH of the solution to 1.5 - 7.0, more preferably by raising it to 1.8 - 2.5 or 2.5 - 6.5. The material is preferably obtainable from the solution by the addition of ammonium chloride or ammonium sulphate, preferably providing a molar excess of salt to anions of at least 1.6. A material as claimed in any one of claims 20 to 28 obtainable from The material is preferably obtainable from the solution by the addition of sodium hydroxide in liquid phase, more preferably the ammonium salt is added in solid phase and the pH is increased by the addition of sodium hydroxide in liquid phase. The precipitates formed can preferably be redissolved in 5-5M nitric acid and the uranium quantitatively recovered by tributyl phosphate reversed phase extraction.

A third aspect of the present invention provides apparatus adapted to perform the methods of the first aspect of the invention.

The methods of the present invention will now be illustrated by way of example only with reference to the following non-limiting examples and Figures. Further embodiments of the invention will occur to those skilled in the art in the light of these.

FIGURES

Fig 1. shows a graph of Uranium recovery as a function of ammonium chloride concentration in the preferred method of the invention.

Fig 2. shows an X-Ray Diffraction analysis of the pH 2.0 precipitate.

Fig 3- shows a graph of the composition of the filtrate as a function of pH; precipitation scheme 1: pH 2.1, 5-0, 10.0 using the method of Example 5>

Fig 4. shows a graph of the composition of the filtrate as a function of pH; precipitation scheme 2: pH 5-0, 10.0 using the method of Example 5-

EXAMPLE 1: PRECIPITATION AT PH 2

20cm ³ 5M ammonium chloride was added to 10cm ³ of a solution resulting from a Davies and Gray titration, performed on depleted uranium, comprising approximately lg/1 uranium, 1.5g/l iron and O.lg/1 vanadium, chromium and molybdenum in a medium of 3 phosphoric acid. 5M Sodium hydroxide was then added to raise the pH from the initial value of approximately 0 until the solution turned cloudy at pH 2.0 due to the precipitation of iron phosphate. This required the addition of 1.6g sodium hydroxide. The solution was then made slightly more acidic by the addition of a drop of nitric acid in order to redissolve the iron phosphate and left to stand for 14 days after which a yellow precipitate was formed, the eluent being green.

The resulting solution was filtered using Whatman 5 l paper and the yield of uranium recovered determined from the ratio of uranium in the precipitate to that remaining in solution. The uranium content of the precipitate was determined by dissolving the precipitate in nitric acid, measuring the uranium concentration using Inductively Coupled

Plasma Mass Spectroscopy (ICPMS) and correcting the results for differences in volume.

2.5g of precipitate was produced per litre of waste solution, which was found to consist of 302 uranium. 88% of the uranium originally in the solution was removed by the process leaving a more dilute waste solution suitable for treatment by conventional methods. The 2.5 g/1 precipitate was suitable for either disposal by conventional methods or the subsequent extraction and utilisation of the uranium.

The experiment was repeated varying the amount of ammonium chloride added and determining the amount of uranium precipitated at the maximum pH prior to iron precipitation. As can be seen from the results, shown below in Table 1, an increase in the amount of ammonium chloride added resulted in a lowering of the pH at which the iron precipitated. The optimum concentration of ammonium chloride added was taken to be 20cm ³ of 5M ammonium chloride per 10cm ³ of waste, which equates to 400g/litre of waste.

No precipitation occurred at pH 2 in the absence of ammonium chloride even after allowing the solution to stand for 60 days. The experiment was repeated replacing ammonium chloride with ammonium sulphate and the yield of uranium appeared to be unaltered. However experiments using potassium chloride in the place of the ammonium salts at pH 2.1 produced significantly decreased yields of uranium (232) •

TABLE 1: The percentage of uranium precipitated as a function of the volume of ammonium chloride added to 10cm ³ of DG solution at pH 2.0:

pH Volume 5M Mole excess % Uranium Mole ratio NH^Cl/cm ³ NH„C1 to U precipitated NH _ή Cl:P0

2 0 0 0 0.0

2 2 240 0 0.33

2 5 600 40 0.83

2 10 1200 70 1.67

1.9 10 1200 86 1.67

- 15 1800 87* 2.50

1.7 20 2400 88 3-33

- 25 3000 88 ^» 4.17

1.3 40 4800 88 6.67

* Highly enriched uranium data, pH not recorded directly, solution at maximum pH prior to iron precipitation.

EXAMPLE 2: PRECIPITATION AT PH ABOVE 2

The method was performed as described in Example 1 except that the pH was raised above 2 by the addition of further sodium hydroxide. A larger volume of precipitate and a clear eluant were obtained. Table

2 below shows the amount of uranium precipitated at various pH values after the solution was allowed to stand for 14 days. A greater percentage of uranium was removed from the waste solution than at the lower pH of Example 1. The eluent produced contained only 2% of the uranium of the original waste solution and so was easier and cheaper to dispose of. The amount of precipitate formed was greater than that produced by the method of Example 1, however it was still considerably easier to dispose of than the original waste solution.

Almost quantitative uranium precipitation occurred between pH 4-6, even for a 10 fold increase in solution volume. As the pH increased one set of results showed a reduction in the percentage of uranium precipitated. At alkaline pH the precipitation efficiency was reduced because of a reduced amount of ammonium chloride (mole ratio 1.67) and ammonia production. Under these alkaline conditions the precipitate contained an increasing amount of clear crystals, thought to be phosphate, and yellow/orange filtrates were produced.

TABLE 2: The percentage uranium recovery as a function of pH for a 2.5 mole excess of ammonium chloride to DG waste (ie 15cm ³ ammonium chloride and 10cm ³ DG waste) :

PH Volume of DG % Uranium solution/cm ³ precipitated

1.4 - 10 73

2.0 10",10 87".87

3.0 100",10 95".96

4.0 100",10 98".99

5.0 100",10 98".99

6.0 10 99

7.0 10 99.88 ^β ,65 ^*

9.6 10 98, 31 ^*

* denotes a I.67 mole excess of ammonium chloride over DG waste ** denotes experiments performed on highly enriched uranium

EXAMPLE : PRECIPITATION OF URANIUM FROM DAVIES AND GRAY SOLUTIONS AT PH .O USING BOTH SOLID AND LIQUID AMMONIUM CHLORIDE AND SODIUM HYDROXIDE

Similar experiments to those described in Examples 1 and 2 were performed altering the phases of ammonium chloride and sodium hydroxide added. The experiments were carried out at pH 4 using 100cm ³ samples of Davies and Gray solution. In order to add ammonium chloride in the liquid phase 150mls of 5M solution were added, to add it in the solid phase 39-8g was added. To add sodium hydroxide in the

liquid phase δOmls of M solution were added, to add it in the solid phase l6g was added. The volume of solution after treatment was meεisured and the dilution factor calculated, to indicate the increase in volume of solution due to treatment, a dilution factor of 2 indicating a two-fold increase. From the results obtained, shown below in Table 3. i can be seen that at pH 4 the phase of ammonium chloride or sodium hydroxide used does not have a significant effect on the amount of uranium precipitated. The most advantageous way of reducing the volume of the solution being found to be the use of solid ammonium chloride and liquid sodium hydroxide, the use of both as solids being less desirable because of consequent heat generation.

TABLE : Precipitation experiments at pH 4.0 involving 100cm ³ of DG solution and both solid and liquid ammonium chloride and sodium hydroxide:

Volume of Dilution Phase of Phase of 2 uranium Effective Degree of solution factor ammonium sodium precip'd phosphate crystal- (cm ³ ) chloride hydroxide cone.** -isation

325 3.25 liquid liquid 98.1 0.92M none

210 2.1 solid liquid 97-0 1.43M significant

250 2.5 liquid solid 97.6 1.20M little

125* 1.25 solid solid 98.9 2.40M extensive

* upon standing overnight the solution had almost completely crystalised

** initial solution 3M phosphoric acid

EXAMPLE 4: CHARACTERISATION OF THE PRECIPITATES

Tables 4 and 5 below show how the composition of the precipitates varied according to the pH at which they were obtained. As can be seen from the results the method carried out at pH 2 produces a smaller amount of precipitate than at higher values of pH, but removes less of the uranium from the solution. However the uranium remaining in solution is low enough for standard methods to be utilised to further treat the waste. At pH 4 the eluent contains a very low amount of uranium but a larger quantity of the precipitate was formed, containing more of the other ions from the Davies and Gray titration, such as vanadium and chromium.

TABLE 4: The mass of precipitate as a function of pH for a 2.5 mole excess of ammonium chloride to phosphate and a three fold dilution:

PH Volume of Mass of Mass of DG solution precipitate precipitate(g) (cm ³ ) (g) per litre of DG solution

2 100, 200 0.25, 0.47 2.5

2.1* 200 0.9. 1-3 5-5

_4#» 200 3.0 15.0

10** 200 60 300

* precipitation just apparent

** extensive drying required to remove water from gelatinous precipitate

TABLE 5

pH 2 pH 2.1 pH 5

Colour of precipitate yellow light dark green green green

Colour of filtrate green green faint green

Description of precipitate fine fine gelatinous powder powder

Weight % of uranium in precipitate ¹ 26% 18* e%

Mole ratio of phosphate : uranium ¹ 3-4 : 1 12 : 1 8 : 1

Mole ratio of iron : uranium ¹ 0.8 : 1 2 : 1 4 : 1

Percentage U in precipitate ² 88* 932 992

Percentage Fe in precipitate ² 902 992

Percentage V in precipitate ² * 44* 782

Percentage Cr in precipitate ² * 272 922

Percentage Mo in precipitate ² 362 952

* small amounts of vanadium and chromium were detected in the precipitate by X-Ray diffraction measurement

¹ X-Ray Diffraction measurement (Cu-K-alpha radiation)

² Inductively Coupled Plasma - Optical Emission Spectra (ICP-OES) measurement

Percentages referred to in the above Table refer to the percentage of the total element present in the DG waste prior to treatment that is precipitated.

A sample of the pH 2 precipitate was analysed by X-Ray diffraction (Cu-K-alpha radiation) to determine whether a major crystal phase or a mixture was present. The data was compared to a database of materials, as shown in Fig. 2. The results obtained indicated that the precipitate was unlikely to contain a mixture of compounds and was likely to comprise either chernikovite [H ₂ (U0 ₂ ) ₂ (P0 ₄ ) ₂ ,8H ₂ 0] or uramphite [NH _i U0 ₂ P0 _i ,3H ₂ 0]. The fact that ammonium ions were essential for precipitation at pH 2 but not at pH 4 or 5 indicated the production of uramphite at pH 2 and a mixture at pH 4 or 5> Experiments were performed on a solution of 3M phosphoric acid and uranium and a uranium precipitate formed at pH 5 in the absence of ammonium chloride after 14 days. This showed that precipitation at pH 5 is not governed by ammonium chloride or co-precipitation with the transition metals but is pH dependant. It was thereby determined that at pH 2 the precipitate is most likely uramphite and at higher pH chernikovite.

EXAMPLE 5: PRECIPITATE AND FILTRATE ANALYSIS AS A FUNCTION OF pH

To determine the influence of the rate of precipitation on the composition of the filtrates and precipitates experiments were performed using two precipitation schemes. The first involved precipitation for 7 days at each of the following pH values: 2.1, 5»0 and 10.0; the second involved precipitation for 7 days at pH 5-0 followed by 7 days at pH 10.0. The results are shown in Table 6 below and in Figs 3 and 4. The precipitation at pH 2.1 is more selective at removing uranium than at pH 5. although the uranium yield is slightly reduced. Increasing the pH from 2.1 to 5«0 resulted in an increase in the amount of all ions precipitated, notably iron, chromium and molybdenum. In both cases an increase to pH 10.0 resulted in no

further metal ion precipitation from solution. The precipitate at pH 10 is translucent and soluble in water and is thought to be ammonium phosphate.

TABLE 6: The filtrate composition at pH 10.0 (including 2.5 fold dilution) and the percentage removal of metal ions following precipitations at (1) pH 2.1, 5.0. 10.0 and (2) pH 5-0. pH 10.0:

Precipitation mass of Concentration in filtrate/ppm and experiment ppte* percentage removal from DG waste

U Fe V Cr Mo

(1) pH 2.1 1.2g 34ppm 6lppm 36ppm 5lppm 35ppm 932 902 44* 27* 362

(1) pH 5-0 3-lg 28ppm 2ppm 23ppm 20ppm 13ppm 942 99.62 632 71* 752

(1) pH 10.0 55.9g 23ppm 0.7ppm 23ppm 13ppm 13ppm 942 99-92 60* 792 742

(2) pH 5-0 lOg 5.3ppm 7.2ppm 13.4ppm 5.2ppm 2.7ppm 992 992 78* 922 952

(2) pH 10.0 40.2g 5•5ppm 7.lppm 13-5PPm 4.9ppm 2.8ppm 992 992 78* 922 952

* Experiments performed on 200cm ³ DG waste

EXAMPLE 6: PRECIPITATION OF DAVIES AND GRAY TITRATION REAGENTS FROM PHOSPHORIC ACID AND AMMONIUM CHLORIDE

A series of experiments was performed to determine whether the reagents used in the Davies Gray titration would precipitate from a solution comprising phosphoric acid and ammonium chloride, at the concentrations used to treat the waste. As can be seen from the results obtained, shown below in Table 7. the iron and vanadium reagents precipitate reasonably closely to pH 2.1 but molybdenum and chromium reagents do not. It is probable that the molybdenum, chromium, and to a lesser extent iron and vanadium, co-precipitate with the uranium.

TABLE 7: The precipitation of Davies and Gray titration reagents from phosphoric acid (3M) and ammonium chloride (400g/l phosphoric acid ) as a function of pH (sodium hydroxide):

Reagent added to the phosphoric acid/ ammonium Comments chloride solution

None: blank No precipitation up to pH 7-0

Ferrous sulphate Precipitate forms at pH 3-2

Vanadyl sulphate Precipitate forms at pH 3>9

Molybdate reagent No precipitation up to pH 7>0

Dichromate reagent No precipitation up to pH 7>0

EXAMPLE 7; RECOVERY OF URANIUM FROM THE PRECIPITATES

Reversed phase tributyl phosphate (TBP) extractions were performed on precipitates formed by the methods described earlier at pH 5 and pH 2. As can be seen from the results shown below in Table 8 an almost quantitative uranium separation occurs for both precipitates. The X-Ray Diffraction results indicate that any phosphate present does not interfere in the separation of the other metals from uranium. It was found from previous TBP separation experiments, after the addition of a large excess of aluminium ions or iron ions to DG solutions that there was no free phosphate in the water (uranium containing) fraction. Similarly it is assumed that no phosphate is present in this water fraction as the uranium recovery is high, this fraction could therefore be heated/oxidised (4θO-6θO°C in air) to produce uranium oxide and subsequently recovered by conventional means.

The precipitate formed at pH 10.0 is predominantly phosphate and has a very low level of uranium and other metal ion contamination. Dissolution in water would produce a phosphate waste which should be suitable for standard treatment.

Although the uranium phosphate precipitate produced at pH 2.1 was contaminated with significant amounts of vanadium, chromium, molybdenum and iron, this did not affect the efficiency of separation (TBP) as shown in Table 8. It is therefore possible that the precipitate could be combined with uranium oxide from other analytical processes which is to be converted into uranium metal and the mixture purified by a TBP process prior to metal production. The purification of the precipitate before such mixing would be unnecessary because any impurities would be insignificant compared to those in the larger quantity of uranium oxide with which it would be combined. This would result in an overall reduction in the cost of the treatment.

IΔBIJL-----: Analysis from a TBP separation of precipitates produced at pH 2.0 and pH 5.0:

ELEMENT Nitric acid rinse Water elute fraction / counts fraction / counts

pH 5.0 ppte, U (HEU ¹ ) 409 35.269

pH 2.0 ppte, U (DU ² ) 0 985

Fe ² 221 8

Mo ² 58* 0

V ² 5.8 0

Cr ² 2.3 1.2

¹ liquid scintillation counting

² X-ray fluorescence analysis

* bright yellow precipitate formed with nitric acid possibly phosphomolybdate

EXAMPLE 8: TREATMENT OF FILTRATES

If desired the filtate may be treated to selectively extract the uranium. The composition of the filtrate at pH 10.0 is dependant on its preparation, as shown in Table 6. The slower more selective pH 2.1 precipitation produces a different final solution to that of the quicker pH 5-0 precipitation. If treatment of the filtrate is desired it is more practical to perform the initial precipitation at pH 5-0 prior to pH 10.0 to prevent significant precipitation of metal ions with the phosphate. Precipitation at pH 10.0 produces a phosphate solid, with little contamination of uranium or other metals, that can be easily dissolved in water and easily disposed of and a filtrate low in phosphate content and potential interfering metal ions.

EXAMPLE Q: PREPARATION OF A FILTER TO PERFORM THE PRECIPITATION METHODS

A 1 litre beaker was modified by a glassblower to incorporate a filter (porosity number 0) and tap arrangement attached to the bottom of the beaker wall. This arrangement facilitates the separation of the filtrate from the precipitate and the precipitate can conveniently be dissolved with nitric acid and flushed through the tap, if not required, or slurried with water, poured out and dried before being further analysed or treated.