RU2384902C1 - Method of purifying uranium oxides from impurities - Google Patents

Abstract

FIELD: physics, nuclear.

SUBSTANCE: invention relates to the technology of producing nuclear fuel for power generation purposes. The method involves dissolving oxides, extracting uranium with 30% tributylphosphate in an organic diluent, washing the uranium extract with nitric acid solution, re-extracting uranium with a weakly acidic aqueous solution and processing the evaporated re-extract into oxides. The dissolving, extraction and washing processes are combined through simultaneous reaction of uranium oxides with tributylphosphate in a hydrocarbon diluent containing nitric aicd, and with aqueous nitric acid solution. The organic solution obtained from the reaction is taken for re-extraction.

EFFECT: optimisation of the process of purifying oxides of natural, regenerated or reusable (weapons-grade) uranium from impurities.

3 cl, 4 tbl, 5 ex, 2 dwg

Description

The invention relates to a technology for producing nuclear fuel for energy purposes, in particular to a process for purifying impurities from oxides of natural, regenerated or return (weapons) uranium.

Known methods for the purification of oxides of natural, regenerated or return (weapons) uranium include operations common to all types of uranium:

dissolving uranium oxides in aqueous nitric acid solutions, extracting uranium from nitric acid solutions with tributyl phosphate (TBP) in a hydrocarbon diluent, washing the uranium extract with a nitric acid solution, and reextracting uranium with water or a slightly acidic solution (B.V. Gromov. Introduction to the chemical technology of uranium. A textbook for universities .- M .: Atomizdat, 1978, p.223) - prototype. The obtained strips are evaporated, processed into oxides, which are sent to subsequent operations to obtain nuclear fuel (fluorination, etc.).

A known method of purification of natural uranium concentrates, which consists in dissolving uranium compounds in a solution of nitric acid with an excess acidity of 20-30 g / l. After filtration, the uranium nitrate solution is sent for extraction processing using a 30% TBP solution in a hydrocarbon diluent as an extractant. The extraction is carried out with a phase ratio of O: B = 2.5: 1; the extract is washed from impurities 1 ÷ 4 mol / l with a solution of nitric acid with a phase ratio of O: B = 10: 1. Reextraction is carried out with water heated to 60 ° C with a phase ratio of O: B = 1: (1 ÷ 2). The obtained re-extract is processed into oxides. (N. S. Turaev, I. I. Zherin. Chemistry and technology of uranium. - M., Publishing House "Ore and Metals", 2006, p. 315-316).

A known method of purification of oxides of return (weapons) highly enriched uranium (HEU) from plutonium with a view to its further use as reactor fuel (RF Patent No. 2107959, publ. 27.03.98, IPC G21C 19/44, 19/46, C01G 43/00 , 56/00). The method includes extraction of uranium from a nitric acid solution with 30% TBP in a hydrocarbon diluent, washing the extract with a nitric acid solution, and stripping uranium with water to produce a stripping that is subjected to further processing.

A known method of purification of oxides of return (weapons) of uranium from impurities in the processing of HEU (Improvement of the extraction and precipitation technology of processing HEU. Abstract. Fifth Scientific and Technical Conference of the Siberian Chemical Combine. October 20-22, 1999 - Seversk, 1999. p. .51-59.). The method includes dissolving triuran octaoxide in a solution of 4.5-5 mol / L nitric acid to obtain a solution with a uranium concentration of 140-180 g / L, introducing hydrazine hydrate and uranium (IV) into the solution, and extracting uranium with 25-30% TBP in a hydrocarbon diluent; reextraction of uranium from organic solutions with weakly acidic solutions. Trieran octaoxide purified from plutonium and other impurities is obtained from reextracts.

Known methods for cleaning oxides of regenerated uranium from technetium (RF Patent No. 2184083, publ. 06/27/2002, IPC C01G 43/00, 57/00; Application for invention No. 2007109748, publ. 09/27/2008, Bull. No. 27, IPC C22B 60 / 02, C01G 43/00, G21C 19/46, B01D 11/04). The methods include dissolving uranium trioxide in nitric acid solutions, introducing hydrazine and uranium (IV) into the solution to restore technetium (VII) to an unextractable TBP valence state, extracting uranium with 30% TBP in a hydrocarbon diluent with an O: B = 4 phase ratio , 6: 1, washing the extracts with washing solutions at a phase ratio of O: B = 10: 1. From the washed extracts, uranium is back-extracted with mildly acidic solutions.

Common disadvantages of these methods are the complexity (multi-stage process) and the relatively large volume of liquid radioactive waste category SAO, due to the volume of the source and wash solutions.

The objective of the invention is to simplify the method, as well as reducing the volume of liquid radioactive waste in the processing of oxides of natural, regenerated and return (weapons) uranium, and thereby reducing the cost of processing uranium-containing materials and the disposal of water-tail solutions, while ensuring the required purification of uranium oxides from impurities .

The problem is solved in that in a method for purification of impurities of oxides of natural, regenerated or weapons-grade uranium, including dissolving oxides, extracting uranium with 30% tributyl phosphate in an organic diluent, washing the uranium extract with a nitric acid solution, stripping uranium with a weakly acidic aqueous solution and processing evaporated re-oxide into oxides , combine the operation of dissolution, extraction and washing by simultaneous interaction of uranium oxides with tributyl phosphate in a hydrocarbon diluent containing nitric acid that, and an aqueous nitric acid solution, and the organic solution obtained after the interaction is sent for re-extraction.

The ratio of tributyl phosphate in a hydrocarbon diluent containing nitric acid and an aqueous nitric acid solution, which is used to interact with uranium oxides, is set to (10 ÷ 50): 1, and the interaction is carried out at a temperature of (18 ÷ 60) ° С.

The preparation of tributyl phosphate in a hydrocarbon diluent containing nitric acid is carried out at a temperature of (18 ÷ 25) ° С by means of countercurrent multistage contacting of tributyl phosphate in a hydrocarbon diluent with a solution of 8.0 mol / L nitric acid with the ratio of the organic solution stream to the aqueous solution stream O: B = (7 ÷ 8): 1, the nitric acid solution obtained after contacting is diluted to a concentration of (0.05 ÷ 0.1) mol / L with water and condensate from evaporation of the back-extract and sent to the back-extraction operation.

Figure 1 presents a diagram of the processing of oxides by the proposed method. Figure 2 is a diagram of the processing of oxides by a known (classical) method.

The proposed method has been tested in a laboratory setting for the processing of natural triuran octoxide, regenerated uranium trioxide and return (weapons) triuran octoxide.

The method is as follows.

An organic solution is prepared - tributyl phosphate in an organic diluent containing nitric acid in a predetermined amount (see example 5). A predetermined amount of uranium oxide, an organic solution of 30% tributyl phosphate in a hydrocarbon diluent containing nitric acid (organic phase), and an aqueous nitric acid solution (aqueous phase) are placed in the reactor. The resulting mixture was thoroughly mixed at a given temperature. As a result of the simultaneous interaction of uranium oxide with organic and aqueous solutions, the dissolution of uranium and impurities occurs, while uranium is absorbed by the organic phase, while the impurities are absorbed by the aqueous solution. In fact, there is a combination of dissolution, extraction and washing. The resulting organic solution was separated from the aqueous solution. Uranium is reextracted from the organic solution with a weak nitric acid solution, the extract is evaporated and processed into uranium oxide, which is sent for further processing into reactor fuel.

1. Purification of concentrates of natural uranium.

Example 1. Purification of a concentrate of natural uranium (triuran octoxide) from impurities that form during the fluorination of volatile and non-volatile fluorides.

Weighed portions of uranium concentrate corresponding in composition to “Technical Requirements for Uranium Concentrate ASTM C967-02”, with a uranium content of 83.7%, each weighing 5.85 g (4.9 g U), were treated with 50 ml of a 30% TBP hydrocarbon diluent containing 1.05 mol / l nitric acid and 1 ml of nitric acid aqueous solution (organic to aqueous phase ratio O: B = 50: 1) at a temperature of (40 ÷ 60) ° С and intensively mixing the mixture with a magnetic stirrer . After dissolving a portion of uranium oxide, organic solutions containing uranium were separated from aqueous solutions, reextracted from organic solutions of uranium with an aqueous solution containing 0.05 mol / L nitric acid, the strips were evaporated, and the impurity content was determined in evaporated strips of uranium. For reextraction, an aqueous solution of nitric acid was used, leaving the saturation unit of the extractant with nitric acid, diluted with water and condensate from evaporation of the reextract.

The experiments showed that the complete dissolution of uranium from concentrates in 50 ml of 30% TBP in a hydrocarbon diluent containing 1.05 mol / L nitric acid and 1 ml of an aqueous nitric acid solution was achieved in (40-120) min. A decrease in temperature below 40 ° C led to an increase in the interaction time of uranium oxide with solutions up to 3-4 hours, i.e. to reduce the performance of the installation. An increase in temperature above 60 ° C led to a sharp increase in the rate of interaction, while the interaction was accompanied by an increase in temperature, foaming of the solution, and the release of gaseous reaction products.

The uranium content in the extracts averaged 90.9 g / l regardless of the dissolution temperature, and the impurity content in the stripped off uranium reextracts met ASTM C787-03 requirements for enrichment of uranium hexafluoride. One stripped off extracts were sent for further processing - obtaining oxides, fluorination, etc.

The cleaning results are shown in table 1.

Table 1 The content of impurities in concentrates and extracts of uranium. Impurity The content of impurities in uranium concentrate,% to U The content of impurities in the reextracts of uranium,% to U Required ASTM C787-03 impurity content,% to U Volatile Fluoride Impurities Si 0.1 ≤0.001 0.01 Mo 0.01 ≤0.0001 0.00014 V 0,03 ≤0.0001 0.00014 AT 0.005 ≤0.00003 0.00010 Ti 0,0003 ≤0.0001 0.00010 That 0,0003 ≤0.0001 0.00010 Sb 0.0010 ≤0.0001 0.00010 W 0,0002 ≤0.0001 0.00014 Non-volatile fluoride impurities Na 0.5 ≤0.002 The total content of impurities should not exceed 500 μg / g U Al 0.2 ≤0,0003 Sa 0.2 0.0010 Th ≤0.03 ≤0.001 Zr 0.1 ≤0.001 Fe ≤1.5 ≤0,0007 Mn 0.3 ≤0.0001 Cr 0.05 ≤0,0003

An increase in the ratio of the volume of organic solution to the volume of water above 50 led to an increase in the content of impurities in reextracts of uranium, in particular molybdenum, to 0.0003% of the content of uranium, and a decrease to a decrease in saturation of the extractant with uranium with a slight increase in the positive effect in terms of purification of uranium from impurities .

2. Purification of regenerated uranium from technetium - 99, thorium - 228 and its radioactive decay products.

During long-term storage of regenerated uranium oxides in the warehouses of manufacturing plants, the products of oxides accumulate radioactive decay of the uranium-232 isotope with high-energy γ-radiation (thorium-228, thallium-208, etc.), which eliminates their direct fluorination to hexafluoride uranium and requires preliminary purification of uranium from them. In addition, technetium-99 may be present in the oxides of regenerated uranium in amounts exceeding the standard ASTM C787-03 for uranium hexafluoride for enrichment. This requires preliminary, before fluorination of uranium oxides, extraction purification of solutions of regenerated uranium from technetium-99. As a rule, the cleaning of uranium from thorium-228 and technetium-99 in the extraction schemes for processing uranium solutions is carried out by maximally saturating the extractant with uranium, introducing complexing agents into the initial and washing solutions that form non-extractable TBP complex compounds with thorium, and complexing agents and reducing agents to convert the extracted tributyl phosphate technetium (VII) into non-extractable TBP technetium (IV).

Example 2. Purification of oxides of regenerated uranium from technetium.

Uranium trioxide powder with a total uranium content of 82.67% and technetium-99 of 10.4 μg / g U was used. According to the content of other impurities - stable impurities, γ- and α-radionuclides - uranium trioxide met the requirements of ASTM C787-03 for uranium hexafluoride for enrichment.

Weighed portions of a powder of uranium trioxide weighing 6.6 g (5.45 g by uranium content) were each treated with 50 ml of 30% TBP in a hydrocarbon diluent containing 0.9 mol / L nitric acid and 1 ml of an aqueous nitric acid solution, in which the reducing agent was introduced - uranium (IV) and the complexing agent - trilon B (sodium salt of ethylenediaminetetraacetic acid). The interaction was carried out in the temperature range (18-40) ° C. The organic solution obtained after the interaction was sent for re-extraction, the re-extract was evaporated, processed into oxides, which were sent for further processing.

The experiments showed that in the indicated temperature range the complete dissolution of uranium trioxide continued for 5-15 minutes. An increase in temperature led to an increase in the reaction reaction rate. The introduction of uranium (IV) and trilon B into aqueous nitric acid solutions led to the purification of uranium from technetium directly in the operation of interaction of uranium oxide with organic and aqueous solutions, i.e. on the operation, which is a combination of dissolution-extraction-washing operations. Then reextracted purified from technetium uranium (VI). The reextract was evaporated, processed into oxides and sent for further processing.

The results of the experiments are shown in table 2.

table 2 Purification of uranium from technetium The content of U (IV) and Trilon B in aqueous solution Uranium (VI) content and technetium in uranium extract The cleaning coefficient (To _Pts ) U from TC U (IV), g / l Trilon B, g / l U, g / l Tc, μg / g U 10 5 96.5 0.16 27.9 10 10 98.9 0.05 89.4 5 5 95.3 0.19 23.5

When the content of uranium (IV) and trilon B in aqueous solutions was 5-10 g / l, the technetium content in uranium extracts was 0.05-0.19 μg / g U, which is lower than the value regulated by ASTM C787-03 for uranium hexafluoride for enrichment (≤0.5 μg / g U).

An increase in the ratio of the volume of organic solution to the volume of nitric acid solution of more than 50 leads to a decrease in the coefficient of purification of uranium from technetium. A decrease of less than 50 increases the volume of aqueous solutions and reduces the saturation of the extractant with uranium, while the increase in the positive effect in terms of purification of uranium from technetium is insignificant, at O: B = 50: 1 the content of technetium is lower than the values regulated by ASTM C787-03.

Example 3. Purification of oxide of regenerated (high-background) uranium from thorium-228 and its radioactive decay products.

A triuran octaoxide powder was used with a uranium-232 content of 1.8 · 10 ^-7% to total uranium with an exposure dose rate (MED) of 0.0033 nA / kg · kg U. Weighed portions of 5.85 g (4.96 g for uranium). The experimental procedure did not differ from that given in Example 1, except that the complexing agent Trilon B was introduced into the aqueous phase (nitric acid solution) based on its content of 10 g / L. In the experiments, the ratio of the volume of the organic phase to the volume of the aqueous phase was varied O: B = (10 ÷ 50): 1. The content of thorium-232 and its radioactive decay products in uranium extracts was determined by the exposure dose rate (DER) of the uranium extract due to their high-energy γ-radiation.

In relation to the purification of uranium from thorium, a decrease in the ratio of the volume of the organic solution to the volume of the aqueous solution led to an increase in the purification of uranium from thorium.

The results of the experiments are shown in table 3.

Table 3. Purification of uranium from thorium-228 and its radioactive decay products. O: B Uranium extract The cleaning coefficient (K _Pts ) U from γ-radionuclides U, g / l DER nA / kg · kg U 10: 1 107.7 <0.00002 > 165 25: 1 101,2 0.000066 50,0 50: 1 99.7 0.00010 33.0 100: 1 95.1 0.001 3

When the ratio of the volume of the organic phase to the volume of water equal to O: B = 10: 1, the EDR of the uranium extract did not exceed the background value. An increase in the ratio of the volume of the organic phase to the volume of the aqueous phase led to an increase in the EDR of the uranium extract due to a decrease in the coefficients of purification of uranium from thorium-228 and its radioactive decay products.

To ensure a normal radiation situation in the site of fluorination of oxides of regenerated uranium, it is sufficient to reduce the content of γ-radionuclides in uranium solutions by 30-50 times, which is achieved by the interaction of uranium oxide with a mixture of organic and aqueous solutions taken in the ratio O: B = (10-50 ):one.

3. Purification of return (weapons) uranium. Example 4. Purification of weapons grade uranium from plutonium. A triuranic octaoxide powder was used with a mass of 5.85 g (4.9 g by uranium content) and a plutonium content of 1631 Bq / g U, obtained by oxidation of metal shavings of return (weapons) uranium. The experimental technique did not differ from that given in Example 1, except that a reducing agent, uranium (IV) and hydrazine nitrate, was introduced into the aqueous phase based on their content in the solution (5-10) g / l and (2-5) g / l, respectively. The complete interaction of weapons-grade uranium oxide with the organic and aqueous phases in the temperature range (40-60) ° C was achieved in (40-120) min. The uranium content in the organic phase was 103.5 g / l, and the plutonium content was 5.4 Bq / g U. The plutonium purification rate from plutonium was ~ 300.

An increase in the ratio of the volume of the organic phase to the volume of the aqueous phase (nitric acid solution) of more than 50 is impractical due to a decrease in the coefficient of purification of uranium from plutonium (for example, at O: B = 100: 1, the plutonium content in evaporated uranium reextracts was 50 Bq / g U), and a decrease in O: B less than 50 increases the volume of aqueous solutions and reduces the saturation of the extractant with uranium, while the increase in the positive effect in terms of purification of uranium from plutonium is insignificant. At O: B = 50: 1, the plutonium content is lower than the values regulated by ASTM C787-03 on uranium hexafluoride for enrichment (the α-activity of plutonium should not exceed 25 Bq / g U).

From the above examples it follows that the proposed method provides the necessary purification of uranium from impurities (radioactive and non-radioactive) in the processing of oxides of natural, regenerated and return (weapons) uranium.

4. Saturation of the extractant with nitric acid (preparation of tributyl phosphate in an organic diluent containing nitric acid).

When saturating the extractant with nitric acid, it should be taken into account that the necessary saturation of the extractant with nitric acid must be ensured with a minimum volume of water-tail solutions formed in this process and a minimum content of nitric acid in them.

The saturation of the organic phase with nitric acid depends on the mass of uranium oxide loaded onto the combined operation of dissolution, extraction and washing, i.e. Installation performance depends. Two other factors (the volume of water-tail solutions formed during the saturation operation and the content of nitric acid in them) affect the balance of the entire scheme with respect to the flows of the aqueous phase. The aqueous nitric acid solution from the operation of saturating the extractant with nitric acid is diluted with water and condensate from evaporation of the stripping and sent to the operation of stripping of uranium.

Example 5. Saturation of the extractant with nitric acid.

The extractant was saturated with a 30% solution of TBP in a hydrocarbon diluent with nitric acid at a temperature of 20 ° C in an installation consisting of 5 centrifugal extractors. In the experiments, the content of nitric acid in the initial aqueous solutions and the ratio of the flow of organic solutions to the flows of aqueous solutions were varied.

When the content of nitric acid in the initial aqueous solution was 8 mol / L and the ratio of the flow of the organic solution to the aqueous flow was 7-8, the saturation of the organic solution with nitric acid was (0.95-1.05) mol / L. This is enough to dissolve uranium oxides with a mass that ensures saturation of the extractant with uranium (90-105) g / l.

The results of the experiments are shown in table 4.

Table 4 Extractant saturation with nitric acid The ratio of the flow of the organic solution to the stream of aqueous solution The initial concentration of nitric acid in an aqueous solution, mol / l The concentration of nitric acid in an equilibrium organic solution, mol / l The concentration of nitric acid in an equilibrium aqueous solution, mol / l 8: 1 8 0.95 0.39 7: 1 8 1.05 0.70 6: 1 8 1,11 1.33

The content of nitric acid in the water-tail solutions from the saturation operation was (0.39-0.70) mol / L at O: B = (7-8): 1, which when diluted with water and / or condensate from evaporation of uranium reextracts 7-8 times allows you to get in a reextracting solution the content of nitric acid necessary for uranium extraction equal to (0.05-0.1) mol / l.

Heating the extraction mixture above ambient temperature is impractical due to a slight change in the distribution coefficients of nitric acid.

For comparison and explanation of the proposed method, figure 1.2 shows the basic scheme of purification (processing) of uranium oxides according to the proposed (figure 1) and classical (figure 2) methods indicating volumes of technological solutions per 1 ton of processed uranium.

As a classical method of purification of uranium oxides, a method is taken that includes leaching (dissolving) of uranium with a solution of nitric acid from uranium concentrates, extraction of uranium with a solution of TBP (30 ± 2)% vol. in a hydrocarbon diluent from solutions containing 350 ÷ 450 g / l of uranium, when the extractant is saturated with uranium 100-120 g / l, washing the uranium extract with a washing solution at a ratio of the flow of the organic solution to the flow of the wash, equal to 10: 1, reextraction of uranium from the organic solution a solution of "weak" nitric acid, evaporation of a uranium re-extract using condensate for uranium re-extraction operations (B.V. Gromov. Introduction to the chemical technology of uranium. Textbook for high schools. - M .: Atomizdat, 1978, p.223).

As can be seen from figures 1 and 2, per 1 ton of processed uranium in the proposed method, there is a decrease in water-tail solutions (VHR), falling into the category of medium-level waste (SAO) and to be disposed of, 3 ÷ 16 times compared with the classical uranium extraction purification scheme (in the classical scheme, the amount of water-tail solutions (water chemistry) for disposal is 3.33 m ³ , and in the proposed one (0.2 ÷ 1.0) m ³ ).

The proposed method provides for the purification of oxides from impurities necessary for their further processing into reactor fuel, and at the same time simplifies the instrumentation and technological scheme for the extraction purification of uranium oxides by combining the operations of dissolution, extraction and washing, reduces the amount of water chemistry for disposal.

Claims (3)

1. The method of purification from impurities of oxides of natural, regenerated or weapons-grade uranium, including the dissolution of oxides, extraction of uranium with 30% tributyl phosphate in an organic diluent, washing the uranium extract with a nitric acid solution, reextracting uranium with a weakly acidic aqueous solution and reprocessing one stripped off reextract into oxides, characterized in that combine the operations of dissolution, extraction and washing by simultaneous interaction of uranium oxides with tributyl phosphate in a hydrocarbon diluent containing nitric acid And aqueous nitrate solution, and after reaction of the resulting organic solution is routed to stripping. 2. The method according to claim 1, characterized in that the ratio of tributyl phosphate in the hydrocarbon diluent containing nitric acid and aqueous nitric acid solution, which are used to interact with uranium oxides, is set to (10 ÷ 50): 1, and the interaction is carried out at a temperature (18 ÷ 60) ° С. 3. The method according to claim 1, characterized in that the preparation of tributyl phosphate in an organic diluent containing nitric acid is carried out by multi-stage countercurrent contacting at a temperature of (18 ÷ 25) ° C of tributyl phosphate in a hydrocarbon diluent with a solution of 8 mol / l nitric acid with a ratio the flow of the organic solution to the flow of aqueous O: B = (7 ÷ 8): 1, the solution of nitric acid after contacting is diluted with water and condensate from evaporation of the back-extract to a concentration of (0.05 ÷ 0.1) mol / L and sent to the back-extraction operation ii.

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