A method for separating strontium from TODGA extract in a high-level radioactive waste liquid treatment process

Abstract

This invention relates to a method for separating strontium from TODGA extract in a high-level radioactive wastewater treatment process. The method includes the following steps: washing: washing the strontium-containing TODGA extract with dilute acid to reduce the concentration of hydrogen ions in the organic phase; back-extraction: back-extracting the washed organic phase with dilute acid to extract strontium from the organic phase into the dilute acid solution, thus separating strontium from TODGA. The concentration of the dilute acid is 0.001 mol / L. During washing, the volume ratio of dilute acid to extract is 1:1. After washing, the concentration of hydrogen ions in the organic phase is not higher than 0.01 mol / L. Compared with the prior art, this invention achieves the separation of strontium from TODGA with a strontium yield greater than 99%.

Description

Technical Field

[0001] This invention relates to the field of spent fuel reprocessing, and more specifically to a method for separating strontium from TODGA extract in a high-level radioactive waste liquid treatment process. Background Technology

[0002] In the spent fuel reprocessing process, high-level radioactive waste liquid is a waste liquid with high radioactivity and extremely high toxicity, which concentrates more than 95% of the radioactivity in spent fuel, posing a potential hazard to humans and the environment. Its treatment and disposal has become one of the key factors restricting the sustainable development of nuclear energy. 90 Sr and its daughters 90 Y is a pure β emitter, making it an ideal material for manufacturing β radiation sources and nuclear batteries. It has special applications in fields such as national defense, environmental science, and biomedicine, and has high practical value and economic significance.

[0003] TODGA (N,N,N′,N′-tetraoctyl-3-oxopramethylenediamide) exhibits a high partitioning ratio for strontium, enabling its application in high-level radioactive waste liquids. 90 To obtain strontium nitrate product, strontium extracted into TODGA needs to be separated. Summary of the Invention

[0004] The purpose of this invention is to overcome at least one of the defects of the prior art by providing a method for separating strontium from TODGA extract in a high-level radioactive waste liquid treatment process.

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

[0006] A method for separating strontium from TODGA extract in a high-level radioactive waste liquid treatment process, the method comprising the following steps:

[0007] Washing: The strontium-containing TODGA extract was washed with dilute acid to reduce the concentration of hydrogen ions in the organic phase;

[0008] Back-extraction: The washed organic phase is back-extracted using dilute acid, and strontium is extracted from the organic phase into the dilute acid solution, thus separating strontium from TODGA.

[0009] Furthermore, the dilute acid mentioned is dilute nitric acid.

[0010] Furthermore, the concentration of the dilute acid is not higher than 0.001 mol / L.

[0011] Furthermore, the concentration of the dilute acid is 0.001 mol / L.

[0012] Furthermore, during washing, the volume ratio of dilute acid to extract is (0.8-1.2):1.

[0013] Furthermore, during washing, the volume ratio of dilute acid to extract is 1:1.

[0014] Furthermore, after washing, the concentration of hydrogen ions in the organic phase is no higher than 0.01 mol / L.

[0015] Furthermore, during back-extraction, the volume ratio of dilute acid to organic phase is (0.8-1.2):1.

[0016] Furthermore, during back-extraction, the volume ratio of dilute acid to organic phase is 1:1.

[0017] Furthermore, during washing and back-extraction, extraction equipment with a capacity of 5 or higher is used.

[0018] Compared with the prior art, the present invention...

[0019] (1) In this invention, the yield of strontium is as high as 99% or more;

[0020] (2) Because the partition coefficient of Sr in the TODGA-TBP-OK extraction system is proportional to the acidity, the higher the acidity, the more favorable the extraction, and vice versa, the more favorable the back-extraction. Therefore, washing is an essential process. Detailed Implementation

[0021] The embodiments of the present invention are described in detail below. These embodiments are implemented based on the technical solution of the present invention, and provide detailed implementation methods and specific operation processes. However, the scope of protection of the present invention is not limited to the following embodiments.

[0022] A method for separating strontium from TODGA extract in a high-level radioactive waste liquid treatment process, the method comprising the following steps:

[0023] (1) Reagent preparation

[0024] Calculate the required amount of 65% nitric acid according to formula (1), and calculate the required amount of deionized water according to formula (2) to prepare a 0.001 mol / L dilute nitric acid solution.

[0025]

[0026] V2 = V - V1 Equation (2)

[0027] V represents the volume of 0.001 mol / L dilute nitric acid to be prepared, V1 represents the volume of 65% nitric acid to be added, and V2 represents the volume of deionized water to be added. The unit is liters (L).

[0028] (2) Washing

[0029] First, fill the extraction equipment with more than 5 stages with a prepared 0.001 mol / L dilute nitric acid solution, introduce TODGA that has been extracted with strontium nitrate, and wash with the two-phase flow ratio set to 1:1 to reduce the hydrogen ion concentration in the organic phase to below 0.01 mol / L.

[0030] (3) Back-extraction

[0031] First, fill the extraction equipment with more than 5 stages with a prepared 0.001 mol / L dilute nitric acid solution, then introduce TODGA that has been extracted with strontium nitrate, and set the two-phase flow ratio to 1:1 for back-extraction.

[0032] (4) Sample Analysis

[0033] After the reaction reaches equilibrium, samples are taken for analysis, and the yield of strontium is calculated according to equation (3).

[0034]

[0035] p – Strontium yield;

[0036] C1 — Strontium concentration in the extractant, expressed in grams per liter (g / L);

[0037] V3 – Extractant flow rate, in liters per hour (L / h);

[0038] C2 — Strontium concentration in the back-extraction solution, expressed in grams per liter (g / L);

[0039] V4 – Back-extraction solution flow rate, in liters per hour (L / h).

[0040] The separation of strontium in TODGA was achieved with a strontium yield greater than 99%.

[0041] Example 1

[0042] In this embodiment, the concentration of Sr in the high-level radioactive waste liquid was 2 g / L, and the acidity was 3 mol / L. Extraction was performed using a 6-stage mixed clarifier tank. After extraction with 0.2 mol / L TODGA and 0.5 mol / L TBP-OK, the concentration of Sr in the extraction residue was 0.004 g / L, and the acidity was 2.83 mol / L. The organic phase containing Sr was washed and back-extracted. After back-extraction, the acidity in the aqueous phase was <0.01 mol / L, and the concentration of Sr was 1.986 g / L. Calculations showed that…

[0043]

[0044] The yield of Sr was 99.49%.

[0045] The above description is merely a preferred embodiment of the present invention and is not intended to limit the invention in any other way. Any person skilled in the art may make changes or modifications to the above-disclosed technical content to create equivalent embodiments. However, any simple modifications, equivalent changes, and modifications made to the above embodiments based on the technical essence of the present invention without departing from the scope of the present invention shall still fall within the protection scope of the present invention.

Claims

1. A method for separating strontium from TODGA extract in a high-level radioactive wastewater treatment process, characterized in that, The method includes the following steps: (1) Washing A 6-stage mixing and clarification tank was filled with 0.001 mol / L dilute nitric acid solution. TODGA with extracted strontium nitrate was introduced and the two-phase flow ratio was set to 1:1 for washing to reduce the hydrogen ion concentration in the organic phase to below 0.01 mol / L. (2) Back-extraction A 6-stage mixing and clarifying tank was filled with a 0.001 mol / L dilute nitric acid solution. TODGA, which had been extracted with strontium nitrate, was then introduced, and the two-phase flow ratio was set to 1:1 for back-extraction. (3) Sample Analysis After the reaction reaches equilibrium, samples are taken for analysis, and the yield of strontium is calculated. The high-level radioactive waste liquid contained 2 g / L of Sr and had an acidity of 3 mol / L. Extraction was performed using a 6-stage mixed clarification tank, followed by extraction with 0.2 mol / L TODGA and 0.5 mol / L TBP-OK. The Sr concentration in the extraction residue was 0.004 g / L, and the acidity was 2.83 mol / L. The Sr-containing organic phase was washed and back-extracted. After back-extraction, the acidity in the aqueous phase was <0.01 mol / L, and the Sr concentration was 1.986 g / L, resulting in a Sr recovery rate of 99.49%.

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