RU2793956C1 - Extraction method for processing uranium-containing solutions - Google Patents

Abstract

FIELD: extraction processing of uranium-containing solutions.

SUBSTANCE: invention can be used in the technology of processing irradiated nuclear fuel of nuclear power plants, as well as various uranium-containing solutions, including recycled materials. The processing method includes extraction of uranium from uranium-containing solutions with an organic solution of tributyl phosphate in an inert diluent, washing the extract with a nitric acid solution. Re-extraction of uranium is carried out with an aqueous solution of 1.0-2.5 mol/l of glycine and 0.03-0.07 mol/l of nitric acid.

EFFECT: invention allows to reduce losses of uranium with the circulating extraction solvent, to increase the technological stability of the process and the possibility of recycling mother liquors after precipitation of uranium from re-extracts.

2 cl, 4 tbl

Description

The invention relates to the extraction processing of uranium-containing solutions and can be used in the technology of processing irradiated nuclear fuel of nuclear power plants (mainly), as well as various uranium-containing solutions, including recycled materials.

In all cases, the extraction processing cycle is completed by the operation of re-extraction of uranium from the organic phase (solution of tributyl phosphate, TBP in a diluent) into the aqueous phase.

Reextraction, as a rule, is carried out with a weak (0.05-0.07 mol / l) solution of nitric acid at a temperature of ~ 60 ° C (Gromov B.V., Savelyeva V.I., Shevchenko V.B. Chemical technology of irradiated nuclear fuel, M. Energoatomizdat, 1983, p. 351). Features of the uranium distribution isotherm under these conditions do not allow obtaining re-extracts with a uranium content of more than 100 g/l, which requires significant energy consumption for re-extract evaporation.

Significantly higher concentrations of uranium are achieved in the processes of uranium stripping with the use of reagent stripping agents. The most concentrated re-extracts, more than 300 g/l for uranium, were obtained using 60-400 g/l urea solutions as a re-extractant, under certain conditions regulating the relationship of such parameters of the re-extraction process as temperature, uranium content in the extract, nitric acid content in extract, the content of carbamide in the re-extractant and the ratio of the flows of the organic and aqueous phases (RF Patent No. 2 170 964 C1, publ. 20.07.2001).

According to the task and the technical nature of this method is closest to the claimed and selected as a prototype.

Analysis of the prototype method revealed a number of significant drawbacks.

1. Even when the re-extraction process is carried out in the prescribed parameters, the residual content of uranium in the circulating extractant is on average 20 mg/l. At the soda regeneration operation, uranium is concentrated by 10-12 times, which creates problems in the disposal of the regenerate.

2. Minor deviations from the regulatory parameters, inevitable in real production conditions, lead (we quote the description of the invention to the patent) “... to a sharp increase in the uranium content in the circulating extractant, a decrease in the uranium content in re-extracts, an increase in the uranium content in carbonate washing solutions, which can lead to to the formation of precipitates of uranium carbonate compounds in them.

3. During the precipitation of ammonium polyuranates from urea re-extracts, and this is the only option for their processing (RF Patent No. 2 114 469 C1, publ. 06/27/1998), a mother liquor is formed with a total salt content (carbamide + ammonium nitrate) up to 500 g/l. There is no technology for the disposal of such a product.

The technical result of the invention is to reduce the loss of uranium with the circulating extractant, increase the technological stability of the process and the possibility of recycling mother liquors after precipitation of uranium from re-extracts.

The result is achieved in the method of processing uranium-containing solutions, including the extraction of uranium, the washing of the extract and the re-extraction of uranium, while the re-extraction is carried out with an aqueous solution of an amino acid (glycine).

The stripping solution contains 1.0-2.5 mol/l of glycine, depending on the required degree of uranium concentration.

Additionally stripping solution contains 0.03-0.07 mol/l nitric acid.

Stripping is carried out at a temperature of 25-60°C.

The efficiency of using aqueous solutions of glycine, NH ₂ CH ₂ COOH as uranium stripping agents from tributyl phosphate extracts is determined by the complexation of uranyl with the carboxyl group and the binding of the resulting nitric acid by the amino group of the glycine molecule.

The introduction of small amounts of nitric acid into the stripping solution prevents the appearance of uranium associated with dibutylphosphoric acid, which is present in small amounts in real TBP solutions, in the recycled extractant.

Uranium peroxide or (with the addition of plutonium) uranium-plutonium peroxide is precipitated from the glycine-containing re-extract, glycine in the mother liquor is decomposed to a residual content of not more than 10 mg/l, which makes it possible to dispose of the mother liquor in the normal mode of processing liquid radioactive waste of an average level of activity (RF Patent No. 2 638 543 C1, published on 12/14/2017).

The proposed method was tested on an installation of 10 small-sized mixing and settling extractors with pulsating mixing and flow transportation. The input streams (extract and re-extractant), if necessary, were heated in heat exchangers before entering the plant.

For all experiments, a 30% solution in isopar M was used and an extract was prepared containing 90 g/L of uranium and 8 g/L of nitric acid. The extract contained 10 mg/l dibutylphosphoric acid.

In preliminary experiments, it was found that the stationary mode in the block at the selected flow rates is established in ~ 4 hours.

Example 1 Composition of the stripping solution: glycine 1.5 mol/l, nitric acid 3.2 g/l, flow rate 69 ml/h. Extract consumption 200 ml/h.

The stationary mode of the block was controlled by the constancy of the reextract composition.

In table. 1 shows data on the content of uranium and nitric acid in the equilibrium phases at the steps of the block. The extract input is in step 1, the re-extractant input is in step 10.

The consumption of the re-extract exceeded the consumption of the re-extractant by 9.3%, the consumption of the circulating extractant was lower than the consumption of the extract by 9%. These changes are associated with the mass transfer of uranium into the aqueous phase.

Example 2 Composition of the stripping solution: glycine 2.5 mol/l, nitric acid 3.2 g/l, flow rate 54 ml/h. The rest of the experimental conditions are unchanged.

Data on the content of uranium and nitric acid in the equilibrium phases at the steps of the block are given in Table. 2.

Example 3. The composition of the stripping solution is the same as in example 1, the flow rate is 54.8 ml/h. The composition and consumption of the extract are the same as in examples 1 and 2.

The input streams (extract and re-extractant) were heated to 50°C before being fed into the block.

Data on the content of uranium and nitric acid in the equilibrium phases at the steps of the block are given in Table. 3.

Example 4. The composition of the stripping solution is the same as in example 2, the flow rate is 40.4 ml/h. The temperature of the working environment is -50°C.

The composition and consumption of the extract are unchanged.

Data on the content of uranium and nitric acid in the equilibrium phases at the steps of the block are given in Table. 4.

Uranium peroxide was precipitated from the re-extract (Example 4), the mother liquor was corrected in composition, and then sent to the operation of the destruction of glycine using a zirconium-platinum catalyst (see reference 4). The residual content of glycine was 8.5 mg/L.

Comparison of technological indicators of the prototype method and the proposed method demonstrates the significant advantages of the proposed method:

- the content of uranium in the circulating extractant in the proposed method is ~ 20 times lower, which eliminates the problem of processing soda regenerates;

- in the process of re-extraction according to the proposed method, a uranium-free zone is formed in the re-extraction unit, which significantly increases the technological stability of the process (shift of the uranium front with fluctuations in flow rates and return of the uranium-free zone to a stationary state when the process stabilizes);

- During the precipitation processing of uranium re-extracts obtained by the proposed method, the mother liquors are disposed of in the normal mode of processing liquid radioactive waste of an average level of activity.

Claims (2)

1. Method for processing uranium-containing solutions, including extraction of uranium, washing of the extract and stripping of uranium, characterized in that stripping of uranium is carried out with an aqueous solution of 1.0-2.5 mol/l of glycine and 0.03-0.07 mol/l of nitric acid . 2. The method according to p. 1, characterized in that the stripping process is carried out at a temperature of 25-60°C.