阅读说明：本技术 一种钚回收离子液体萃取剂及其从含钚废液中萃取分离钚的方法 (Plutonium recovery ionic liquid extractant and method for extracting and separating plutonium from plutonium-containing waste liquid ) 是由 占佳 徐辉 沈忠 马特奇 张�林 张春 梁威 于 2020-12-14 设计创作，主要内容包括：本发明公开了一种钚回收离子液体萃取剂及其从含钚废液中萃取分离钚的方法。具体是以离子液体为萃取剂,或将离子液体与传统有机萃取剂复配形成复合萃取剂,用于分离回收含钚废液中的钚离子,具体方法为：取含钚废液加入浓硝酸,调整废液中硝酸浓度为3-8mol/L,再将废液中钚的价态调节为四价,然后加入离子液体萃取剂萃取分离废液中的钚,最后通过反萃来回收离子液体萃取剂中的钚。本发明将离子液体用于废液中钚的分离回收,减少了有毒、可燃、易挥发有机溶剂的使用。(The invention discloses an ionic liquid extractant for plutonium recovery and a method for extracting and separating plutonium from plutonium-containing waste liquid. The method specifically uses ionic liquid as an extracting agent, or compounds the ionic liquid and a traditional organic extracting agent to form a composite extracting agent, and is used for separating and recovering plutonium ions in plutonium-containing waste liquid, and the specific method comprises the following steps: adding concentrated nitric acid into plutonium-containing waste liquid, adjusting the concentration of the nitric acid in the waste liquid to be 3-8mol/L, adjusting the valence state of plutonium in the waste liquid to be tetravalent, adding an ionic liquid extractant to extract and separate the plutonium in the waste liquid, and finally recovering the plutonium in the ionic liquid extractant by back extraction. The invention uses the ionic liquid for separating and recovering plutonium in the waste liquid, and reduces the use of toxic, combustible and volatile organic solvents.)

1. The plutonium recovery ionic liquid extractant is characterized by comprising the following components in percentage by volume:

80-100% of component A

0 to 20 percent of component B

The sum of the two components is 100 percent;

wherein the component A is composed of an ionic liquid [ C_nmim][NTf₂]And [ C_nmim][PF₆]N are the same or different and are selected from 2,4,6,8,10, 12;

the component B is composed of one or more of tributyl phosphate, thenoyl trifluoroacetone, octyl (phenyl) -N, N-diisobutyl amine formyl methyl phosphine oxide, N, N-dimethyl-N, N-dioctyl-2- (2-hexylethoxy) malonamide and N, N, N 'N' -tetra-N-octyl-3-oxaglutaramide.

2. A method for extracting and separating plutonium from a plutonium-containing waste liquid, characterized in that the ionic liquid extractant of claim 1 is used, and comprises the following steps:

adding concentrated nitric acid into the plutonium-containing waste liquid to enable the concentration of the nitric acid in the waste liquid to be 3-8mol/L, further adjusting the valence state of plutonium in the waste liquid to be tetravalent, and then adding an ionic liquid extractant to perform extraction, namely finishing the extraction and separation of the plutonium in the waste liquid.

3. The method according to claim 2, further comprising adding a stripping agent to the ionic liquid extractant containing plutonium obtained after the extraction to perform stripping, and recovering plutonium from the ionic liquid extractant.

4. The method for extracting and separating plutonium from a plutonium-containing waste liquid according to claim 2, wherein the adjustment of the valence state of plutonium to tetravalent state in the plutonium-containing waste liquid is achieved by adding hydrogen peroxide or a sodium nitrite solution to the plutonium-containing waste liquid.

5. The method for extracting and separating plutonium from a plutonium-containing waste liquid according to claim 4, wherein the hydrogen peroxide solution has a mass fraction of 30% and the sodium nitrite solution has a mass concentration of 1 mol/L; the usage amount of the hydrogen peroxide or sodium nitrite solution is 1/50-1/20 of the plutonium-containing waste liquid.

6. A method for the extractive separation of plutonium from plutonium-containing effluents according to claim 2, characterized in that said extraction time is comprised between 15 and 60min and the number of extractions is comprised between 1 and 5.

7. A method of extracting plutonium from plutonium-containing waste liquor according to claim 2, characterised in that the volume ratio of ionic liquid extractant to plutonium-containing waste liquor is from 1:2 to 4: 1.

8. A method for extracting and separating plutonium from a plutonium-containing waste liquor according to claim 3, characterised in that the stripping time is 15-60min and the number of stripping times is 1-5.

9. A method of extracting plutonium from a plutonium-containing waste liquid in accordance with claim 3, wherein the stripping agent is one of oxalic acid, ammonium carbonate or sodium carbonate; the mass concentration of the oxalic acid is 0.05-0.3mol/L, the mass concentration of the ammonium carbonate is 0.1-0.8mol/L, and the mass concentration of the sodium carbonate is 0.1-0.8 mol/L.

10. A method of extracting plutonium from a plutonium-containing waste liquor according to claim 3, characterised in that the plutonium-containing ionic liquid extractant and stripping agent obtained after the extraction have a volume ratio of 2:1 to 1: 5.

Technical Field

The invention belongs to the technical field of radioactive waste liquid treatment, and particularly relates to an ionic liquid extracting agent for plutonium recovery and a method for extracting and separating plutonium from plutonium-containing waste liquid.

Background

The chemical leaching is a technology for reducing and decontaminating the soil polluted by plutonium with high efficiency-cost ratio, and the technology mainly breaks physical or chemical adsorption of soil components to the plutonium through the action of an eluent, dissolves the plutonium adsorbed on the surface of a mineral or combined in a mineral crystal lattice, realizes the transfer of the plutonium from a soil solid phase to a leaching liquid phase, or effectively converts the existing form of the plutonium, and realizes the stabilization of the plutonium in the soil. Although the chemical leaching technology has the advantages of simple and convenient operation, high decontamination efficiency, rapidness, thoroughness and the like, a large amount of plutonium-containing waste liquid is generated after the soil is leached. Therefore, solving the problem of washing waste liquid treatment is the key for determining the application of the technology, and an applicable treatment technology needs to be established for the plutonium-containing waste liquid, so that the resource recovery of plutonium is realized while the amount of secondary waste in soil washing decontamination is effectively controlled.

The chemical leaching decontamination of the plutonium polluted soil can use various reagents such as inorganic acid, organic acid, complexing agent, oxidation-reduction agent and the like, the generated waste liquid not only has high plutonium content and high acidity, but also has very complex matrix components and can contain a large amount of metal ions (such as Fe)³⁺、Al³⁺、Ca²⁺、Mg²⁺Etc.), inorganic acid radical ions (e.g., Cl)^-、NO₃ ^-、SO₄ ^2-Etc.), organic acid radical ion (C)₂O₄ ^2-、C₆H₅O₇ ^3-Humic acid, etc.), complexing agent, oxidant, etc., so the treatment difficulty of the leaching waste liquid is far greater than that of the common radioactive waste liquid.

Although there are many alternative methods for treating radioactive waste liquid, such as chemical precipitation, ion exchange and adsorption separation, evaporation concentration, electrodialysis, membrane separation, microbial adsorption, etc., these conventional methods are somewhat unsuitable or limited for treating waste liquid from leaching of plutonium contaminated soil, for example, chemical precipitation, ion exchange and adsorption separation can generate a large amount of secondary waste, which is not beneficial to reducing the volume of contaminants; the evaporation concentration and electrodialysis method have large energy consumption and higher treatment cost; the membrane separation technology is greatly influenced by the components and the content of the waste liquid, and the waste liquid is generally required to be pretreated by combining other methods; the microbial adsorption technology is also influenced by coexisting ions in the waste liquid, and radiation-resistant microorganisms with high adsorption selectivity on plutonium are not reported, but the acquisition and large-scale culture of the microorganisms are difficult. In order to solve the problems, a plurality of more applicable technologies need to be searched, and the key ring of waste liquid treatment in chemical leaching decontamination of plutonium contaminated soil is broken through.

The ionic liquid is a new 'green solvent', is a molten salt system which is completely composed of specific cations and anions and is in a liquid state at room temperature or near room temperature, and generally consists of organic cations containing nitrogen and phosphorus and large inorganic anions. The ionic liquid has many advantages different from the traditional organic solvent, such as no toxicity, no flammability, lower vapor pressure, high thermal stability and chemical stability, easy recovery, designability and the like, so the ionic liquid is widely concerned in nuclear fuel circulation, radioactive waste liquid treatment and radiochemical analysis. However, at present, research on direct application of the ionic liquid to plutonium extraction and separation of the soil leaching waste liquid is not reported, and if an applicable ionic liquid can be obtained, a special ionic liquid extraction and back extraction method is established, and process parameters and conditions are optimized, so that the technology becomes an environment-friendly, safe, economical and efficient separation and recovery technology for plutonium in the soil leaching waste liquid, and further the problem of waste liquid treatment of the chemical leaching decontamination technology for soil polluted by plutonium is solved.

Disclosure of Invention

The invention aims to solve the problem that a large amount of toxic, combustible and volatile organic solvents are used in a traditional solvent extraction method, and provides a method for separating and recovering plutonium from plutonium-containing waste liquid by using an ionic liquid extractant aiming at plutonium-polluted soil leaching waste liquid with high plutonium content, high acidity and complex matrix components.

The invention firstly uses an ionic liquid extractant to extract and separate plutonium from the plutonium-containing waste liquid, and then realizes the recovery of the plutonium in the ionic liquid through back extraction.

In order to solve the technical problems, the invention provides the following technical scheme:

an ionic liquid extractant for plutonium recovery comprises the following components in percentage by volume:

component A80-100%

Component B0-20%

The sum of the two components is 100 percent;

wherein the component A is composed of an ionic liquid [ C_nmim][NTf₂]And [ C_nmim][PF₆]N are the same or different and are selected from 2,4,6,8,10, 12; the component B is composed of one or more of tributyl phosphate, thenoyl trifluoroacetone, octyl (phenyl) -N, N-diisobutyl amine formyl methyl phosphine oxide, N, N-dimethyl-N, N-dioctyl-2- (2-hexylethoxy) malonamide and N, N, N 'N' -tetra-N-octyl-3-oxaglutaramide.

The invention also provides a method for extracting and separating plutonium from plutonium-containing waste liquid, which utilizes the ionic liquid extractant and comprises the following steps:

adding concentrated nitric acid into the plutonium-containing waste liquid to enable the concentration of the nitric acid in the waste liquid to be 3-8mol/L, further adjusting the valence state of plutonium in the waste liquid to be tetravalent, and then adding an ionic liquid extractant to perform extraction, namely finishing the extraction and separation of the plutonium in the waste liquid.

Preferably, the method for extracting and separating plutonium from a plutonium-containing waste liquid further comprises adding a stripping agent to the plutonium-containing ionic liquid extractant obtained after the extraction, and performing stripping to recover plutonium in the ionic liquid extractant.

Preferably, the adjustment of the valence state of plutonium to quadrivalence state in the plutonium-containing waste liquid is realized by adding hydrogen peroxide or sodium nitrite solution into the plutonium-containing waste liquid.

Preferably, the mass fraction of the hydrogen peroxide is 30%, and the mass concentration of the sodium nitrite solution is 1 mol/L; the usage amount of the hydrogen peroxide solution or the sodium nitrite solution is 1/50-1/20 of the volume of the plutonium waste liquid.

Preferably, the extraction time is 15-60min, and the extraction times are 1-5.

Preferably, the volume ratio of the ionic liquid extractant to the plutonium-containing waste liquid is 1:2-4: 1.

Preferably, the back extraction time is 15-60min, and the back extraction times are 1-5.

Preferably, the stripping agent is one of oxalic acid, ammonium carbonate or sodium carbonate; the mass concentration of the oxalic acid is 0.05-0.3mol/L, the mass concentration of the ammonium carbonate is 0.1-0.8mol/L, and the mass concentration of the sodium carbonate is 0.1-0.8 mol/L.

Preferably, the volume ratio of the ionic liquid extractant to the back extractant obtained after extraction is 2:1-1: 5.

Compared with the prior art, the invention has the following beneficial effects:

1. the ionic liquid is used as an extracting agent, or the ionic liquid and the traditional organic extracting agent are compounded to form the composite extracting agent, so that the composite extracting agent is used for separating and recovering plutonium ions in plutonium-containing waste liquid, the selectivity is high, the distribution rate is good, the ionic liquid is non-toxic and non-flammable, safe and environment-friendly, and unsafe factors and environmental risks in the extraction process are reduced. Meanwhile, the used ionic liquid extractant can be reused through back extraction, so that the economic cost is saved, the generation of secondary waste is reduced, and the resource recovery of nuclides can be realized, thereby showing great application value in the field of radioactive waste liquid treatment.

2. The extraction rate of the ionic liquid extractant for plutonium in the leaching waste liquid of the plutonium polluted soil is over 90 percent, and the extracted ionic liquid extractant can obtain the back extraction rate of over 90 percent after back extraction. By using the method, more than 85% of plutonium in the washing waste liquid of the plutonium contaminated soil can be recovered, the treated waste liquid can be recycled, and the ionic liquid after back extraction can be reused, so that the generation of secondary waste is effectively reduced.

Detailed Description

Reference will now be made in detail to various exemplary embodiments of the invention, the detailed description should not be construed as limiting the invention but as a more detailed description of certain aspects, features and embodiments of the invention.

It is to be understood that the terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the invention. Further, for numerical ranges in this disclosure, it is understood that each intervening value, between the upper and lower limit of that range, is also specifically disclosed. Every smaller range between any stated value or intervening value in a stated range and any other stated or intervening value in a stated range is encompassed within the invention. The upper and lower limits of these smaller ranges may independently be included or excluded in the range.

Unless defined otherwise, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this invention belongs. Although only preferred methods and materials are described herein, any methods and materials similar or equivalent to those described herein can be used in the practice or testing of the present invention. All documents mentioned in this specification are incorporated by reference herein for the purpose of disclosing and describing the methods and/or materials associated with the documents. In case of conflict with any incorporated document, the present specification will control.

It will be apparent to those skilled in the art that various modifications and variations can be made in the specific embodiments of the present disclosure without departing from the scope or spirit of the disclosure. Other embodiments will be apparent to those skilled in the art from consideration of the specification. The specification and examples are exemplary only.

As used herein, the terms "comprising," "including," "having," "containing," and the like are open-ended terms that mean including, but not limited to.

Example 1

Adding 100mL of 1mol/L nitric acid solution into 10g of plutonium-polluted soil, stirring and leaching for 6h at normal temperature, performing solid-liquid separation, collecting supernatant as plutonium-containing waste liquid, and performing ultra-low background liquid scintillation spectrometer (LSC) with Quantum 1220 type) Analyzing the content of plutonium-239 in the waste liquid; measuring the pH of the waste liquid, calculating the hydrogen ion concentration, adding concentrated nitric acid to adjust the nitric acid concentration of a waste liquid system to be 6mol/L, adding 3mL of 30% hydrogen peroxide, standing for 2h, and heating to remove redundant hydrogen peroxide; after cooling, 20mL of plutonium-containing waste liquid were taken out of the tube and introduced into a 100mL plastic centrifuge tube, and 16mL of the tube [ C ] was added₆mim]NTf₂Oscillating and mixing the ionic liquid and 4mL of tributyl phosphate for 30min at normal temperature, standing for layering, respectively collecting the ionic liquid and the waste liquid, and adding 16mL of [ C ] into the waste liquid₆mim]NTf₂The extraction was repeated once with the ionic liquid and 4mL of tributyl phosphate. And (3) measuring the content of the plutonium-239 in the waste liquid after extraction by using the LSC, and then calculating the plutonium-239 extraction rate according to the formula (1).

In the formula: e is plutonium-239 extraction rate,%; [ Pu ] A]_aq,initAnd [ Pu ] in]_aq,eqThe content of plutonium-239 in the waste liquid before and after extraction, Bq/mL, respectively.

And combining the ionic liquids collected by two extractions, adding 40mL0.8mol/L ammonium carbonate into the ionic liquids, oscillating and mixing the ionic liquids at normal temperature for 60min, standing and layering the ionic liquids, respectively collecting the ionic liquids and a back-extraction agent, and adding 40mL0.8mol/L ammonium carbonate into the ionic liquid extraction agent to repeatedly back-extract the ionic liquids once. And combining the back-extraction agents collected by the two back-extractions, measuring the content of the plutonium-239 by using the LSC, and calculating the back-extraction rate of the plutonium-239 according to the formula (2).

In the formula: e' is plutonium-239 back extraction rate,%; [ Pu ] A]'_aq,eqThe content of plutonium-239 in the combined and collected stripping agent after stripping, Bq/mL; [ Pu ] A]_aq,initAnd [ Pu ] in]_aq,eqThe content of plutonium-239 in the waste liquid before and after extraction, Bq/mL respectively; v₁Combining the collected stripping agent volumes; v₀Is the volume of waste liquid.

The results were: e94.14%, E96.02%.

And (3) testing the repeated extraction effect of the ionic liquid extractant after back extraction: and (3) extracting the waste liquor by using the ionic liquid extractant after back extraction according to the same operation steps, and obtaining plutonium-239 with an extraction rate of 92.33%.

The effect that reaches: the recovery rate of plutonium-239 is 90.39%, and the ionic liquid after back extraction can be reused without generating secondary waste.

Example 2

The difference from example 1 is that the ionic liquid extractant is: 15mL of [ C ]₈mim][PF₆]2mL of tributyl phosphate and 1mL of thenoyltrifluoroacetone; the stripping agent is 0.1mol/L oxalic acid.

By detection, the E is 93.31%, and the E' is 94.09%.

And (3) testing the repeated extraction effect of the ionic liquid extractant after back extraction: and (3) extracting the waste liquid by taking the ionic liquid after back extraction according to the same operation steps to obtain plutonium-239 with an extraction rate of 92.38%.

The effect that reaches: the recovery rate of plutonium-239 is 87.80%, and the ionic liquid after back extraction can be reused without generating secondary waste.

Example 3

The difference from example 1 is that the ionic liquid extractant is: 20mL of component A and 4mL of component B, wherein component A comprises 10mL of [ C₄mim][NTf₂]And 10mL of [ C₆mim][PF₆](ii) a The component B is thenoyl trifluoroacetone.

By detection, E is 94.02%, E' is 92.71%.

And (3) testing the repeated extraction effect of the ionic liquid extractant after back extraction: and (3) extracting the waste liquid by taking the ionic liquid after back extraction according to the same operation steps to obtain plutonium-239 with the extraction rate of 92.23%.

The effect that reaches: the recovery rate of plutonium-239 is 87.17%, and the ionic liquid after back extraction can be reused without generating secondary waste.

Example 4

The difference from example 1 is that the ionic liquid extractant is: 25mL of component A and 5mL of component B, wherein component A comprises 10mL of [ C [ ]₄mim][PF₆]、10mL[C₆mim][PF₆]And 5mL of [ C₈mim][PF₆](ii) a The component B is 5mLN, N, N 'N' -tetra-N-octyl-3-oxaglutaramide.

The detection results show that E is 93.76%, and E' is 92.09%.

And (3) testing the repeated extraction effect of the ionic liquid extractant after back extraction: and (3) extracting the waste liquid by taking the ionic liquid after back extraction according to the same operation steps to obtain plutonium-239 with the extraction rate of 91.21%.

The effect that reaches: the recovery rate of plutonium-239 is 86.34%, and the ionic liquid after back extraction can be reused without generating secondary waste.

Example 5

The difference from example 1 is that the volume ratio of ionic liquid extractant to waste liquid is 2: 1.

The detection results show that E is 94.56% and E' is 94.08%.

Example 6

The same as example 1 except that the extraction was repeated 3 times.

By detection, E ═ 96.12% and E ═ 95.02% were determined.

Example 7

The same as example 1, except that the nitric acid concentration of the waste liquid system was adjusted to 8 mol/L.

The result of the detection is that E is 91.23%, and E' is 90.12%.

Example 8

The difference from example 1 is that the ionic liquid extractant is [ C ]₈mim]NTf₂Does not comprise component B; the stripping agent is 0.1mol/L oxalic acid.

The results of the detection show that E is 92.40%, and E' is 93.82%.

And (3) testing the repeated extraction effect of the ionic liquid extractant after back extraction: and (3) extracting the waste liquid by taking the ionic liquid after back extraction according to the same operation steps to obtain plutonium-239 with the extraction rate of 91.57%.

The effect that reaches: the recovery rate of plutonium-239 is 86.69%, and the ionic liquid after back extraction can be reused without generating secondary waste.

The above description is only for the purpose of illustrating the preferred embodiments of the present invention and is not to be construed as limiting the invention, and any modifications, equivalents and improvements made within the spirit and principle of the present invention are intended to be included therein.