CN114574698B

CN114574698B - Spent fuel post-treatment uranium purification method

Info

Publication number: CN114574698B
Application number: CN202210324005.4A
Authority: CN
Inventors: 王辉; 王均利; 晏太红; 刘方; 曹智; 申震; 黄小红
Original assignee: China Institute of Atomic of Energy
Current assignee: China Institute of Atomic of Energy
Priority date: 2022-03-29
Filing date: 2022-03-29
Publication date: 2023-03-07
Anticipated expiration: 2042-03-29
Also published as: CN114574698A

Abstract

The disclosure provides a uranium purification method for spent fuel post-treatment. The method comprises the steps of carrying out acidity adjustment on a uranium primary product liquid 1CU to form a uranium purified feed liquid 2DF, and carrying out extraction removal on impurities containing tetravalent neptunium and plutonium in the organic phase 2DF in an extractor by using an extracting agent and a diluent, wherein the extracting agent is a compound shown in a formula (A), wherein R represents a straight-chain or branched-chain alkyl group with 4-12 carbon atoms, and R' represents a straight-chain or branched-chain alkyl group with 1-12 carbon atoms H, methyl or ethyl.

Description

Spent fuel post-treatment uranium purification method

Technical Field

The disclosure belongs to the technical field of spent fuel reprocessing, and particularly relates to a uranium purification method in a Plutonium and Uranium Reduction Extraction (PUREX) process.

Background

The PUREX process is a process flow for separating and recovering uranium (U) and plutonium (Pu) from reactor spent fuel by using tributyl phosphate (TBP) as an extracting agent. Uranium products after spent fuel post-treatment can be used as raw materials of Mixed Oxide (MOX) fuels, and can also be subjected to fluorination and return to isotope centrifugal separation to recycle uranium-235 in the uranium products. Both use scenarios require adequate decontamination of the uranium product, particularly for the deep decontamination of the non-uranium-derived alpha-radioactive components therein. The non-uranium alpha activity of uranium products is mainly contributed by plutonium and neptunium (Np), and therefore uranium products require adequate decontamination of Pu and Np.

The PUREX process consists mainly of a co-decontamination separation cycle, a uranium purification cycle and a plutonium purification cycle. The main task of the uranium purification cycle is to remove further traces of plutonium, neptunium and residual fission products from the uranium primary product liquor (1 CU) coming from the co-decontamination separation cycle, in order to obtain a purer uranium product liquor. 1CU contains a trace amount of plutonium and a certain amount of neptunium, and the valence states of Pu and Np in the feed liquid of 1CU are both +4 and the valence states of uranium are +6 due to the strong reduction section and the back extraction section. In the related art, such as CN102290111, the uranium purification cycle is to first subject the feed to a thermal oxidation pretreatment (oxidation of Np (IV) in lCU to Np (V) which is not 30% tbp/kerosene extracted) or also to thermal evaporative concentration to form uranium purified feed (2 DF); and extracting uranium in the 2DF by using tributyl phosphate (TBP), and purifying a uranium product by using a reduction washing mode. The method needs heating or evaporating operation on a large amount of 1CU, and extraction and back extraction operation on a large amount of uranium, and the process is complex. Moreover, this method, which uses a reducing wash, requires a large amount of reducing agents that generate unstable and reactive substances by degradation and is therefore limited in safety. In addition, the products of the radiation decomposition of TBP may cause plutonium and neptunium to be irremovable during the reduction washing process, and then enter the uranium product in the low-acid stripping process.

There is therefore a need to provide an improved uranium purification process.

Disclosure of Invention

In view of the above, the main objective of the present invention is to provide a uranium purification method that can solve at least some of the technical problems in the PUREX process.

The invention provides a method for purifying uranium in a PUREX flow, which comprises the steps of carrying out acidity adjustment on a uranium primary product liquid 1CU to form a uranium purified feed liquid 2DF, carrying out extraction removal on impurities containing tetravalent neptunium and plutonium in organic phase 2DF in an extractor by using an extracting agent and a diluent, wherein the extracting agent is a compound shown in a formula (A), wherein R represents a straight chain or branched chain alkyl group with 4-12 carbon atoms, R' represents a straight chain or branched chain alkyl group with 1-12 carbon atoms,

according to one embodiment of the invention, R represents a linear or branched alkyl group having 6 to 10 carbon atoms and R' represents a linear or branched alkyl group having 1 to 4 carbon atoms

According to one embodiment of the invention, the concentration of the extractant in the organic phase is between 0.05 and 0.2mol/L.

According to one embodiment of the invention, the 1CU feed solution, after being subjected to acidity adjustment, contains nitric acid at a concentration of 0.5 to 1mol/L.

According to one embodiment of the invention, the extraction uses hydrogenated kerosene or n-dodecane as diluent.

According to an embodiment of the invention, the extraction step further comprises a counter-current contact washing of a washing agent with the organic phase at the outlet of the organic phase of the extractor.

According to one embodiment of the invention, the detergent is 0.5 to 2mol/L nitric acid.

According to one embodiment of the present invention, the number of extraction stages and the number of washing stages of the extraction are 4 to 8 stages, respectively.

According to one embodiment of the invention, the flow ratio of the extraction is 2DF to organic phase to detergent =5 (0.5-1) to (0.15-0.3).

According to one embodiment of the invention, the flow ratio of the extraction is 2DF to organic phase to detergent = 5.

According to the invention, after the acidity of the 1CU feed liquid is adjusted, the di (N, N-dialkyl acetamide) -alkylamine compound is used as an extracting agent in the 2D tank to directly extract and remove impurities Pu (IV) and Np (IV), so that an indirect impurity removal process of extracting a large amount of U (VI) by using TBP and keeping the impurities in a water phase is avoided, and the following remarkable progress is realized.

(1) Compared with the prior art that uranium is extracted through TBP to indirectly remove neptunium and plutonium, the new extracting agent is directly extracted and removed aiming at trace plutonium and neptunium, and the flow of the used new extracting agent is less than 25% of that of TBP which is used as the extracting agent, so that the using amount of the extracting agent is reduced.

(2) The link of heat preservation pretreatment or evaporative concentration of the 1CU feed liquid is omitted, and the process is simplified.

(3) Compared with a typical PUREX uranium purification cycle, the method adopts the nitric acid concentration of 2-4 mol/L for extraction, the method adopts the nitric acid concentration of about 0.5-1 mol/L for extraction, and the use amount of nitric acid and the acid content of final waste liquid can be reduced.

(4) The addition of a reducing agent when the U (VI) extracted organic phase is washed is omitted, and the aftertreatment and safety limit of the corresponding reducing agent are also omitted.

(5) The link of back extraction of uranium from the extracted U (VI) organic phase is omitted, and the process is simplified.

(6) Compared with the prior art, the uranium has the plutonium removal separation coefficient SF _Pu/U Separation coefficient SF for removing neptunium from uranium _Np/U Are all less than 10 ³ Plutonium removal separation factor SF in uranium according to the method _Pu/U Separation coefficient SF for removing neptunium from uranium _Np/U Are all greater than 10 ⁵ And the yield of uranium is the same as that of the prior art.

Drawings

Fig. 1 is a schematic diagram of a uranium purification process in the related art;

fig. 2 is a schematic diagram of a uranium purification method proposed by the present disclosure.

Detailed Description

The technical solutions in the embodiments of the present disclosure will be clearly and completely described below with reference to the embodiments of the present disclosure and the accompanying drawings, and it is obvious that the described embodiments are only some embodiments of the present disclosure, not all embodiments. All other embodiments obtained by a person of ordinary skill in the art based on the embodiments in the present disclosure without any creative effort belong to the protection scope of the present application.

As mentioned above, the present disclosure is intended to provide a new uranium purification method, in particular a method that simultaneously removes small amounts of Pu (IV) and Np (IV) from a U (VI) -containing aqueous solution without having to previously thermally oxidize or concentrate the aqueous solution, and without having to reduce Pu (IV) and Np (IV). According to an embodiment of the present disclosure, a method for uranium purification in a PUREX process is provided, which includes the steps of adjusting acidity of a 1CU feed liquid, and further extracting Pu (IV) and Np (IV) in a 2D tank by using a bis (N, N-dialkylacetamide) -alkylamine compound as an extracting agent.

Bis (N, N-dialkylacetamido) -alkylamines

Bis (N, N-dialkylacetamide) -alkylamines are a new class of extractants having 3 coordinatable atoms per molecule (O atoms on the two carbonyl groups and the central N atom) as a 3-dentate ligand, where the N atom ligand is a softer ligand designed to extract technetium (Tc) produced by uranium-235 fission in spent fuel. After spent fuel is dissolved by nitric acid, technetium is dissolved in the aqueous solution as a metal oxide salt (TcO) ₄ ^- ) The form exists. It was found that at low nitric acid concentrations, the extractant exhibited a high partition ratio to Tc (VII).

The inventors have found that bis (N, N-dialkylacetamide) -alkylamines have unexpectedly high partition ratios for impurity amounts of Pu (IV) and Np (IV) in solutions in the presence of large amounts of uranium at appropriate nitric acid concentrations, and have redesigned the uranium purification scheme to yield the uranium purification method of the present disclosure.

According to an embodiment, the extractant of the present disclosure is a di (N, N-dialkylacetamide) -alkylamine compound, as represented by formula (a), wherein R represents a linear or branched alkyl group containing 4 to 12 carbon atoms, and R' represents a linear or branched alkyl group containing 1 to 12 carbon atoms. Preferably, said R represents a linear or branched alkyl group containing 6 to 10 carbon atoms, and said R' represents a linear or branched alkyl group containing 1 to 4 carbon atoms. More preferably, said R represents a linear alkyl group containing from 6 to 10 carbon atoms and said R' represents an alkyl group of 1 or 2 carbon atoms. According to a preferred embodiment, the extractant is di (N, N-di-N-octylacetamide) -Methylamine (MIDOA), said R is N-octyl and said R' is methyl, as shown in formula (B).

When the concentration of the extractant is 0.05mol/L, the plutonium removal separation coefficient SF in uranium can be already obtained _Pu/U Separation coefficient SF for removing neptunium from uranium _Np/U Are all greater than 10 ⁵ And the extraction effect is better when the concentration is higher; in view of the consumption of the extractant and the phase stability, the preferred concentration is not more than 0.2mol/L.

Uranium purification process

The uranium purification method of the related art, as shown in fig. 1, requires two extractors, one for uranium extraction and reduction scrubbing and the other for uranium stripping. Specifically, 1CU is pretreated by heating or evaporation, and then nitric acid is added to adjust the prepared 2DF. The step needs to consume more energy, and the concentration of nitric acid is required to reach 2 to 4mol/L. Then, 2DF enters the middle part of the first extractor and is in countercurrent contact with an organic phase consisting of TBP and a diluent to extract uranium; most of the impurities remain in the aqueous phase and are discharged from the aqueous phase outlet of the extractor; the organic phase is contacted with a washing agent containing a large amount of reducing agent in a counter-current manner at an organic phase outlet from the extractor, and is further washed to remove residual impurity elements. The number of extraction stages and the number of washing stages were 8 stages. The flow ratio (volume ratio of the streams entering the extractor) was 2DF: organic phase: detergent = 1.5. This step is a step in which the reducing agent generates unstable and reactive substances by degradation, and is therefore limited in safety. And then, the washed organic phase enters another extractor, dilute nitric acid is used for back extraction of uranium to obtain a 2EU product, and the organic phase solvent is washed and recycled.

The uranium purification method proposed by the present disclosure is shown in fig. 2. This process differs from the previous one in that, instead of extracting the uranium from the uranium product by means of an organic phase, impurities, in particular Pu and Np, are extracted, the uranium remaining in aqueous solution. Therefore, the method disclosed by the invention only needs one extractor for extracting and washing Pu and Np, so that the link of back extraction is omitted, and the process is simplified. Specifically, firstly, the 1CU feed liquid is directly added with nitric acid to be adjusted into 2DF without being heated or subjected to evaporation pretreatment. This 2DF contains a small amount of neptunium and plutonium impurities, both mainly present in tetravalent form. Preferably, the concentration of nitric acid in the feed liquid after the acid adjustment is 0.5 to 1mol/L, and more preferably 0.8 to 1mol/L. The concentration of the nitric acid used in the method is obviously lower than that of the nitric acid used in the related technology, and the use amount of the nitric acid and the acid content of the final waste liquid are reduced. The organic phase is then prepared by mixing the bis (N, N-dialkylacetamide) -alkylamine extractant with a diluent such as hydrogenated kerosene or N-dodecane, passed to an extractor and extracted only for traces of plutonium and neptunium. And (3) washing the organic phase containing the plutonium and the neptunium obtained after extraction with nitric acid (without adding a reducing agent, and simultaneously, saving the post-treatment and safety limit of the corresponding reducing agent), so that a small amount of uranium extracted from the organic phase is washed and enters the water phase, and a 2EU product is obtained.

Preferably, the number of extraction stages and the number of washing stages are 4 to 8, and more preferably, both 8. The flow ratio is 2DF to organic phase to detergent =5, (0.5-1) to (0.15-0.3), and the optimal ratio is 5. Since only traces of plutonium and neptunium are extracted and the extractant is preferably in a concentration of 0.05 to 0.2mol/L, the flow rate of the di (N, N-dialkylacetamide) -alkylamine is less than 25% of the flow rate of TBP as extractant, so that the extractant dosage is reduced. Finally, the organic phase can be recycled after washing.

The disclosure is further illustrated with reference to specific examples.

Example 1 distribution ratio test of MIDOA extracts Pu (IV) and Np (IV) at different concentrations

Preparing a simulated 2DF feed liquid (containing 64g/L UO) ₂ ²⁺ 、25mg/L Pu ⁴⁺ 、33mg/L Np ⁴⁺ 0.5mol/L nitric acid), a di (N, N-di-N-octylacetamide) -Methylamine (MIDOA) extracting agent with the concentration shown in the table 1 and hydrogenated kerosene as a diluent are prepared into an organic phase. The process run was carried out in a 16-stage mixer-settler with a mixing chamber of 5mL and a clarification chamber of 12.5 mL. Wherein, UO ₂ ²⁺ The concentration is measured by an L-edge densitometer, pu ⁴⁺ And Np ⁴⁺ The concentration was measured by a liquid scintillation counter.

Distribution ratio D of a substance _ion Is defined as the ratio of the concentration of a substance in an organic phase to that in an aqueous phase when it reaches an extraction equilibrium in the two immiscible phases. From the measured concentrations of the metal ions, distribution ratios of the metal ions when different concentrations of MIDOA were used as the extracting agent were obtained as shown in Table 1 below.

TABLE 1 distribution ratio D of different metal ions for MIDOA extraction at different concentrations _ion

As can be seen from the data in Table 1, the MIDOA can extract Pu (IV) and Np (IV) from uranium with high selectivity at the concentration of 0.05-0.2 mol/L, and the extraction effect is better at higher concentration.

Example 2 Effect of MIDOA extraction of Pu (IV) and Np (IV) at different nitric acid concentrations

A simulated 1CU feed solution (containing 64g/L UO) was prepared ₂ ²⁺ 、25mg/L Pu ⁴⁺ 、33mg/L Np ⁴⁺ 、2.2mg/LTcO ₄ ^- ) The nitric acid concentration shown in Table 2 was adjusted to the acid. 0.05mol/L di (N, N-di-N-octylacetamide) -Methylamine (MIDOA) is used as an extracting agent, and N-dodecane is used as a diluting agent to prepare an organic phase. The process run was carried out in a 16-stage mixer-settler with a mixing chamber of 5mL and a clarification chamber of 12.5 mL. Wherein, UO ₂ ²⁺ The concentration is measured by an L-edge densitometer, pu ⁴⁺ 、Np ⁴⁺ And TcO ₄ ^- Is measured by a liquid scintillation counter.

Coefficient of separation SF for a substance _M/U The ratio of the content of impurities relative to uranium in the initial feed liquid to the content of impurities relative to uranium in the product feed liquid is defined, the distribution ratio of each metal ion under different nitric acid concentrations can be obtained according to the measured concentrations of the metal ions, and the single-stage extraction separation coefficient of uranium to the metal ions is shown in table 2.

TABLE 2 distribution ratio D of different metal ions for MIDOA extraction _ion And separation factor SF for uranium _M/U

As can be seen from the data in Table 2, MIDOA can extract Pu (IV) and Np (IV) from uranium with high selectivity under the condition that the nitric acid concentration is 0.5 to 1mol/L.

Example 3 experiment of uranium purification method of the present disclosure

The 1CU contains 60g/L uranium concentration, 12mg/L Pu (IV) concentration and 32mg/L Np (IV) concentration. After acid adjustment, the concentration of nitric acid is 1mol/L, and 2DF is obtained. The extractant is MIDOA in 0.05mol/L n-dodecane, and the detergent is 1mol/L nitric acid. The process conditions of the extractor for extracting and washing Pu and Np are as follows: grade 8 extraction, grade 8 washing, flow ratio 2DF extractant detergent = 5. The uranium yield of the experiment is more than 99.99 percent according to the calculation of the analysis result of the sample, and the separating coefficients of removing Pu and removing Np in the uranium are respectively 2.11 multiplied by 10 ⁵ And 1.34X 10 ⁵ Are all greater than 10 ⁵ Compared with the prior art, the method improves two orders of magnitude, and the recovery rate of uranium is 99.99 percent, which is the same as the prior art.

The above description is only a preferred embodiment of the present disclosure, and is not intended to limit the claimed scope of the present disclosure, and all modifications, substitutions, or other related technical fields that can be directly or indirectly applied to the contents of the present specification and the drawings under the inventive concept of the present disclosure are included in the claimed scope of the present disclosure.

Claims

1. A method for purifying uranium in a PUREX flow comprises the following steps:

adjusting the acidity of the primary uranium product liquid 1CU to form purified uranium feed liquid 2DF,

the impurities with tetravalent neptunium and plutonium are removed by extraction in an extractor with an organic counter 2DF consisting of an extractant and a diluent,

wherein the extractant is a compound represented by the formula (A), wherein R represents a linear or branched alkyl group having 4 to 12 carbon atoms, R' represents a linear or branched alkyl group having 1 to 12 carbon atoms,

2. the method according to claim 1, wherein R represents a linear or branched alkyl group having 6 to 10 carbon atoms, and R' represents a linear or branched alkyl group having 1 to 4 carbon atoms.

3. The process of claim 1, wherein the concentration of the extractant in the organic phase is from 0.05 to 0.2mol/L.

4. The process as claimed in claim 1, wherein the 1CU feed solution after acidity adjustment contains nitric acid at a concentration of 0.5 to 1mol/L.

5. The process of claim 1, wherein the extraction employs hydrogenated kerosene or n-dodecane as a diluent.

6. The process of any one of claims 1 to 5, wherein the extraction step further comprises counter-current contact washing of a detergent with the organic phase at an organic phase outlet of the extractor.

7. The method of claim 6, wherein the detergent is 0.5 to 2.0mol/L nitric acid.

8. The method of claim 6, wherein the number of extraction stages and the number of washing stages of the extraction are 4 to 8 stages, respectively.

9. The method of claim 6, wherein the flow ratio of the extraction is 2DF to organic phase to detergent =5 (0.5-1) to (0.15-0.3).

10. The process of claim 9, wherein the flow ratio of the extraction is 2DF to organic phase to detergent = 5.