A method for separating lanthanides and actinides from high level liquid waste using a combination of extractants

Abstract

The application relates to a method for separating lanthanum and actinium from high-level radioactive waste liquid by using a combined extractant, which comprises the following steps: co-extracting high-level radioactive waste liquid containing Am / Cm and lanthanum series elements by using a combined extractant, shielding fission element by using a shielding agent, and obtaining an extracted organic phase; washing the obtained organic phase by using a buffer solvent, and obtaining a washed organic phase; back-extracting the obtained washed organic phase by using a buffer solvent, and obtaining a back-extraction liquid and a waste organic phase; supplementing the back-extraction liquid, and obtaining an Am / Cm intermediate product liquid; removing trace extractants in the obtained Am / Cm intermediate product liquid by using a diluent, and obtaining an oil-removed Am / Cm intermediate product liquid; evaporating and concentrating the oil-removed Am / Cm intermediate product liquid, and obtaining a target product Am / Cm crude product liquid. Compared with the prior art, the application belongs to a circulating process, and has the advantages of short process, cheap and easily-obtained reagent, and high separation efficiency.

Description

Technical Field

[0001] This invention relates to the field of advanced nuclear fuel cycle technology, and in particular to a method for separating lanthanum and actinium from high-level radioactive waste using a combination of extractants. Background Technology

[0002] The development of nuclear power has generated a large amount of spent fuel. This spent fuel is used to recover uranium and plutonium through the PUREX process, producing a large amount of high-level radioactive waste (HDW). Uranium and plutonium can be made into MOX fuel, which, after being irradiated in a reactor, produces nuclear energy, improving the utilization rate of nuclear fuel. HDW, after treatment and disposal, is ultimately stored in a deep geological disposal facility. To improve the utilization rate of the disposal facility volume and reduce nuclear safety risks, the advanced nuclear fuel cycle separates long-lived, high-heat-releasing secondary actinide nuclides (Am / Cm) from the HDW, converting them into short-lived or stable nuclides through transmutation.

[0003] In the field of advanced nuclear fuel cycles, the first step is to separate Am / Cm from high-level radioactive waste. To reduce the radioactivity of high-level radioactive waste, scholars both domestically and internationally have conducted extensive and in-depth research, developing multiple technical routes for the separation of Am / Cm from high-level radioactive waste. Typical separation processes include the TRUEX process + Advanced TALSPEAK process in the United States, the DIAMX process + SANEX process in France, the TRPO process + Cyanex 301 process in China, the DIDPA process in Japan, and the CTH process in Sweden. These processes are all two-cycle processes, which are relatively long and require specific separation reagents. To simplify the two-cycle separation process, France developed a one-cycle process (1-cycle SANEX process and i-SANEX process) based on the DIAMX and SANEX processes, while the United States developed a one-cycle ALSEP process based on the TRUEX and TLSPEAK processes. Compared to the two-cycle Am / Cm separation process, the one-cycle Am / Cm separation process has significant advantages in terms of technology, economy, and operation. In the one-cycle Am / Cm separation process, the Am / Cm and Ln separation reagents used in the 1-cycle SANEX and i-SANEX processes are not commercially available, are difficult to prepare, have a slow mass transfer rate during extraction, and have a large partition coefficient of Am / Cm by the extractant, making back-extraction of Am / Cm difficult. Summary of the Invention

[0004] The purpose of this invention is to overcome the shortcomings of the prior art by providing a method for separating lanthanum and actinium from high-level radioactive waste liquid using a combination of extractants, which has a short process, high separation efficiency, and low operating cost.

[0005] The objective of this invention can be achieved through the following technical solutions:

[0006] This invention provides a method for separating lanthanum and actinium from high-level radioactive waste using a combination of extractants, comprising the following steps:

[0007] S1: A combination of extractants was used to co-extract high-level radioactive waste liquid containing Am / Cm and lanthanide elements, and a shielding agent was used to shield the fragmented elements to obtain the extracted organic phase;

[0008] S2: The organic phase obtained in step S1 is washed with acetyloxyoxime acid (AHA) buffer solvent to obtain the washed organic phase;

[0009] S3: The washed organic phase obtained in step S2 is back-extracted using diethyltriaminepentaacetic acid (DTPA) buffer solvent to obtain a back-extract solution;

[0010] S4: Use fresh combined extractant to replenish the back-extraction solution obtained in step S3 to obtain Am / Cm intermediate product solution;

[0011] S5: Use a diluent to remove trace amounts of extractant from the Am / Cm intermediate product solution obtained in step S4, and obtain an oil-free Am / Cm intermediate product solution;

[0012] S6: The oil-removed Am / Cm intermediate product liquid obtained in step S5 is evaporated and concentrated to obtain the target product Am / Cm crude product liquid.

[0013] Furthermore, the combined extractant mentioned in step S1 includes amide ether reagents and phosphate ester reagents, with the following dosage relationship: amide ether reagents : phosphate ester reagents = 0.01-0.4M : 0.1-1.0M.

[0014] Furthermore, the amide ether reagent is selected from one of N,N,N',N'-tetra(2-ethylhexyl)-3-oxaprandiamide (T2EHDGA) or N,N,N',N'-tetraoctyl-3-oxaprandiamide (TODGA);

[0015] The phosphate ester extractant is selected from one of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEHEHP or P507) and di(2-ethylhexyl) phosphate (HDEHP or P204).

[0016] Further, the shielding agent mentioned in step S1 is one or more of 1,2-cyclohexanediaminetetraacetic acid (CDTA), hydroxyethyl ethylenediaminetriacetic acid (HEDTA), or ethylenediaminetetraacetic acid (EDTA).

[0017] Furthermore, the concentration range of the shielding agent is 0.01-0.5M.

[0018] Furthermore, the pH of the acetoxyxamic acid buffer solvent in step S2 is 2-6, and the concentration range of acetoxyxamic acid is 0.05-1.0 mol / L.

[0019] Furthermore, the pH of the diethyltriaminepentaacetic acid buffer solvent mentioned in step S3 is 2-6, and the concentration range of diethyltriaminepentaacetic acid is 0.01-0.20 mol / L.

[0020] Furthermore, the waste organic phase mentioned in step S3 is washed with an alkaline solution and recycled.

[0021] Furthermore, the alkaline solution is a sodium hydroxide solution.

[0022] Furthermore, the concentration of the sodium hydroxide solution is 1.0-1.5M.

[0023] Furthermore, the diluent mentioned in step S5 is one or both of kerosene or n-dodecane.

[0024] Furthermore, the evaporation and concentration method described in step S6 is membrane distillation or evaporator concentration, with a concentration factor of 20-50 times.

[0025] Compared with the prior art, the present invention has the following advantages:

[0026] (1) The present invention has the advantages of short process, cheap and readily available reagents, and high separation efficiency.

[0027] (2) The present invention uses this process to extract Am / Cm from high-level radioactive waste liquid, which can not only reduce equipment investment, but also reduce operating costs.

[0028] (3) The present invention separates Am / Cm from high-level radioactive waste liquid, which not only facilitates the vitrification of the separated waste liquid and saves the volume of deep geological disposal facilities, but also reduces the nuclear safety risks associated with the storage of waste liquid. Attached Figure Description

[0029] Figure 1 This is a schematic diagram of the separation process of the present invention. Detailed Implementation

[0030] The present invention will now be described in detail with reference to the accompanying drawings and specific embodiments.

[0031] Example 1

[0032] This embodiment provides a method for separating lanthanum and actinium from high-level radioactive waste using a combination of extractants, including the following steps:

[0033] S1: Prepare 10 mL of simulated high-level radioactive waste solution (specific composition: 0.3 g / L Cs). + 0.2g / L Sr2+ 0.18g / LFe 3+ 0.67 g / L Zr 4+ 0.4 g / L Mo 6+ 0.15g / L La 3+ 0.20 g / L Ce 3+ 0.25g / L Pr 3+ 0.30 g / L Nd 3+ 0.12 g / L Sm 3+ 0.05g / L Am 3+ Add EDTA shielding agent to control the EDTA concentration to 0.1 mol / L, add 10 mL of 0.1 M T2 EHDGA + 0.75 M HDEHP combined extractant to extract the simulated feed solution, and obtain the extracted organic phase and extract.

[0034] S2: Wash the molybdenum ions in the extracted organic phase obtained in step S1 with 10 mL of lactic acid buffer solution containing 0.60 mol / L acetyloxyoxime acid (AHA) to obtain the washed organic phase and washing solution.

[0035] S3: The washed organic phase obtained in step S2 was back-extracted using 10 mL of lactic acid buffer solution containing 0.02 mol / L diethyltriaminepentaacetic acid (DTPA) to obtain back-extract and waste organic phase.

[0036] S4 uses 10 mL of fresh 0.1 M T2EHDGA + 0.75 M HDEHP combined extractant to replenish the back-extraction solution obtained in step S3, and removes lanthanide elements from the back-extraction solution to obtain Am / Cm intermediate product solution;

[0037] S5: 2 mL of n-dodecane was used to remove trace amounts of extractant from the Am / Cm intermediate product solution, yielding a de-oiled Am / Cm intermediate product solution (specific composition: 0.005 g / L La). 3+ 0.008 g / L Ce 3+ 0.013 g / L Pr 3+ 0.032 g / L Nd 3+ 0.014 g / L Sm 3+ 0.045g / L Am 3+ Cs + 、Sr 2+ Fe 3+ Zr 4+ Mo 6+ All less than 1E -3 g / L);

[0038] S6: The Am / Cm intermediate product liquid obtained after degreasing in step S5 is concentrated by a factor of 30 to obtain the Am / Cm crude product liquid; the waste organic phase described in step S3 is washed with 10 mL of 1.5 mol / L sodium hydroxide, and the washed organic phase is recycled.

[0039] Example 2

[0040] S1: Prepare 10 mL of simulated high-level radioactive waste solution (specific composition: 0.3 g / L Cs). + 0.2g / L Sr 2+ 0.18g / LFe 3+ 0.67 g / L Zr 4+ 0.4 g / L Mo 6+ 0.15g / L La 3+ 0.20 g / L Ce 3+ 0.25g / L Pr 3+ 0.30 g / L Nd 3+ 0.12 g / L Sm 3+ 0.05g / L Am 3+ HEDTA shielding agent was added to make its concentration 0.1 mol / L. The simulated feed solution was extracted with 10 mL of 0.05 M MTODGA + 0.5 M HDEHP combined extractant to obtain the organic phase and extract solution after extraction.

[0041] S2: Wash the molybdenum ions in the extracted organic phase obtained in step S1 with 10 mL of citric acid buffer solution containing 0.60 mol / L acetyloxyoxime acid (AHA) to obtain the washed organic phase and washing solution.

[0042] S3: The washed organic phase obtained in step S2 was back-extracted using 10 mL of citric acid buffer solution containing 0.02 mol / L diethyltriaminepentaacetic acid (DTPA) to obtain back-extract and waste organic phase.

[0043] S4: Use 10 mL of fresh 0.05 M TODGA + 0.5 M HDEHP combined extractant to replenish the back-extraction solution obtained in step S3, and remove the lanthanides from the back-extraction solution to obtain the Am / Cm intermediate product solution;

[0044] S5: 2 mL of n-dodecane was used to remove trace amounts of extractant from the Am / Cm intermediate product solution, yielding an oil-free Am / Cm intermediate product solution (specific composition: 0.003 g / L La). 3+ 0.004 g / L Ce 3+ 0.015g / L Pr 3+ 0.028 g / L Nd 3+0.017g / L Sm 3+ 0.040g / L Am 3+ Cs + 、Sr 2+ Fe 3+ Zr 4+ Mo 6+ All less than 1E -3 g / L);

[0045] S6: The Am / Cm intermediate product liquid obtained after degreasing in step S5 is concentrated by a factor of 20 to obtain the Am / Cm crude product liquid; the waste organic phase described in step S3 is washed with 10 mL of 1.0 mol / L sodium hydroxide, and the washed organic phase is recycled.

[0046] Example 3

[0047] S1: Prepare 10 mL of simulated high-level radioactive waste solution (specific composition: 0.3 g / L Cs). + 0.2 g / L Sr 2+ 0.18g / LFe 3+ 0.67 g / L Zr 4+ 0.4 g / L Mo 6+ 0.15g / L La 3+ 0.20 g / L Ce 3+ 0.25g / L Pr 3+ 0.30 g / L Nd 3+ 0.12 g / L Sm 3+ 0.05g / L Am 3+ CDTA shielding agent was added to a concentration of 0.1 mol / L. The simulated feed solution was extracted with 10 mL of a combination extractant of 0.08 M T2 EHDGA and 0.75 M HEHEHP to obtain the organic phase and extract.

[0048] S2: Wash the molybdenum ions in the extracted organic phase obtained in step S1 with 10 mL of citric acid buffer solution containing 0.60 mol / L acetyloxyoxime acid (AHA) to obtain the washed organic phase and washing solution.

[0049] S3: The washed organic phase obtained in step S2 was back-extracted using 10 mL of citric acid buffer solution containing 0.02 mol / L diethyltriaminepentaacetic acid (DTPA) to obtain back-extract and waste organic phase.

[0050] S4 uses 10 mL of fresh 0.08 M T2EHDGA + 0.75 M HEHEHP combined extractant to supplement the back-extraction solution obtained in step S3, and removes lanthanide elements from the back-extraction solution to obtain Am / Cm intermediate product solution;

[0051] S5: 2 mL of n-dodecane was used to remove trace amounts of extractant from the Am / Cm intermediate product solution, yielding a de-oiled Am / Cm intermediate product solution (specific composition: 0.005 g / L La). 3+ 0.008 g / L Ce 3+ 0.015g / L Pr 3+ 0.030 g / L Nd 3+ 0.013g / L Sm 3+ 0.043g / L Am 3+ Cs + 、Sr 2+ Fe 3+ Zr 4+ Mo 6+ All less than 1E -3 g / L);

[0052] S6: The Am / Cm intermediate product liquid obtained after degreasing in step S5 is concentrated by a factor of 25 to obtain the Am / Cm crude product liquid; the waste organic phase described in step S3 is washed with 10 mL of 1.5 mol / L sodium hydroxide, and the washed organic phase is recycled.

[0053] The above description of the embodiments is provided to enable those skilled in the art to understand and use the invention. It will be apparent to those skilled in the art that various modifications can be made to these embodiments, and the general principles described herein can be applied to other embodiments without inventive effort. Therefore, the present invention is not limited to the above embodiments, and any improvements and modifications made by those skilled in the art based on the disclosure of the present invention without departing from the scope of the invention should be within the protection scope of the present invention.

Claims

1. A method for separating lanthanum and actinium from high-level radioactive waste using a combination of extractants, characterized in that, Includes the following steps: S1: A combination of extractants was used to co-extract high-level radioactive waste liquid containing Am / Cm and lanthanide elements, and a shielding agent was used to shield the fragmented elements to obtain the extracted organic phase; The combined extractant comprises amide ether reagents and phosphate ester reagents, with the following ratio: amide ether reagent : phosphate ester reagent = 0.01-0.4 M : 0.1-1.0 M; The amide ether reagent is one of N,N,N',N'-tetra(2-ethylhexyl)-3-oxaprandiamide or N,N,N',N'-tetraoctyl-3-oxaprandiamide; the phosphate ester reagent is one of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester or di(2-ethylhexyl) phosphate ester. The shielding agent is one or more of 1,2-cyclohexanediaminetetraacetic acid, hydroxyethyl ethylenediaminetriacetic acid, or ethylenediaminetetraacetic acid; S2: The organic phase obtained in step S1 is washed with a lactic acid buffer solution or a citrate buffer solution of acetoxyxamic acid to obtain the washed organic phase; S3: The washed organic phase obtained in step S2 is back-extracted using a lactic acid buffer solution or a citric acid buffer solution of diethyltriaminepentaacetic acid to obtain a back-extract and a waste organic phase; S4: Use fresh combined extractant to replenish the back-extract obtained in step S3 to obtain Am / Cm intermediate product liquid; S5: Use a diluent to remove trace amounts of extractant from the Am / Cm intermediate product solution obtained in step S4, and obtain an oil-free Am / Cm intermediate product solution; S6: The oil-removed Am / Cm intermediate product liquid obtained in step S5 is evaporated and concentrated to obtain the target product Am / Cm crude product liquid.

2. The method for separating lanthanum and actinium from high-level radioactive waste liquid using a combined extractant according to claim 1, characterized in that, The concentration range of the shielding agent in step S1 is 0.01-0.5 M.

3. The method for separating lanthanum and actinium from high-level radioactive waste liquid using a combined extractant according to claim 1, characterized in that, The pH of the lactate buffer solution or citrate buffer solution of acetoxyxamic acid mentioned in step S2 is 2-6, and the concentration range of acetoxyxamic acid is 0.05-1.0 mol / L.

4. The method for separating lanthanum and actinium from high-level radioactive waste liquid using a combined extractant according to claim 1, characterized in that, The pH of the lactate buffer solution or citric acid buffer solution of diethyltriaminepentaacetic acid mentioned in step S3 is 2-6, and the concentration range of diethyltriaminepentaacetic acid is 0.01-0.20 mol / L.

5. The method for separating lanthanum and actinium from high-level radioactive waste liquid using a combined extractant according to claim 1, characterized in that, The diluent mentioned in step S5 is one or both of kerosene or n-dodecane.

6. The method for separating lanthanum and actinium from high-level radioactive waste liquid using a combined extractant according to claim 1, characterized in that, The evaporation and concentration method described in step S6 is membrane distillation or evaporator concentration, with a concentration factor of 20-50 times.

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