Solidification method of radionuclide technetium in solution and application thereof

Abstract

The invention discloses a method for solidifying radionuclide technetium in a solution and application of the method, and belongs to the technical field of spent fuel post-treatment. The curing method comprises the following steps: firstly, acidifying a technetium-containing solution by using nitric acid, then adding the acidified technetium-containing solution into a sodium metaaluminate solution, and reacting under a low-temperature condition to ensure that technetium is covalently bonded with high-activity mu < 1->-OH in a curing product in a form of pertechnetate (TcO4 <->) and is cured in pseudo-boehmite-like mixed crystals with stable crystal structures; the curing method is mild in operation condition and simple in technological process, overcomes the defects that in a traditional cement curing and glass curing method, the size of a cured product is increased, technetium is prone to volatilization and the like, has the advantages of being small in cured product size, high in curing efficiency, low in cost and environmentally friendly, the maximum curing amount can reach 100 mg / g, and the curing method is suitable for industrial production. The method is suitable for treating technetium-containing acidic nucleus waste liquid generated in the spent fuel post-treatment process.

Description

Technical Field

[0001] This invention belongs to the field of spent fuel reprocessing technology, specifically relating to a method for solidifying the radioactive nuclide technetium in solution and its application. Background Technology

[0002] Nuclear power, as an important component of clean energy, has developed rapidly in recent years, with installed capacity and power generation showing an upward trend. However, the sustainable development of nuclear power is always accompanied by a crucial issue: the safe handling and disposal of spent fuel. Spent fuel reprocessing is a key link in closing the nuclear fuel cycle and realizing resource reuse, but this process generates a large amount of complex and highly radioactive liquid waste, namely high-level radioactive waste. Among them, the radioactive nuclide technetium-99 (… 99 The challenges in handling Tc are particularly prominent.

[0003] Technetium is a typical long-half-life nuclide found in spent fuel. In spent fuel reprocessing, it is primarily produced using pertechnetate ions (TcO4). - The TcO4 is introduced into acidic nuclear waste liquid and stored in underground tanks for a long period, awaiting further treatment. - It has high solubility and easy migration in aquatic environments, and once it enters the aquatic environment, it will pose a long-term and serious potential threat to human living environment and health. Currently, it contains TcO4. - Acidic nuclear waste is typically stored temporarily in underground tanks. However, long-term reliance on these tanks not only occupies enormous space and places continuous storage and management pressure on nuclear facilities, but also poses a significant risk of leakage due to tank aging, corrosion, or extreme natural disasters. Therefore, developing a method to store TcO4... - The technology of efficiently and stably separating and solidifying substances from the liquid phase, transforming them into a stable solid form with low solubility and low migration rate, has urgent practical significance for reducing tank pressure, ensuring environmental safety, and promoting the green and sustainable development of the nuclear power industry.

[0004] Currently targeting TcO4 in nuclear waste liquid - For the solidification treatment of TcO4, the internationally accepted technical routes mainly include cement solidification and glass solidification, but both have obvious limitations and inherent defects. Cement solidification is a relatively simple and low-cost solidification technology. Its basic process involves mixing radioactive waste liquid with cement, additives, etc., and forming a cement solidified body after a hydration reaction. However, when applied to the solidification of TcO4... - At that time, TcO4 - The chemical bonding with cement hydration products is very weak; it is mainly trapped inside the solidified body through physical encapsulation. Under long-term groundwater leaching conditions, TcO4... -It easily migrates from the interior of the solidified body to the surface and dissolves, resulting in high permeability and insufficient long-term stability. In addition, in order to meet the requirements of mechanical strength and leaching rate, a large amount of cementitious base material needs to be added, which leads to a significant increase in the volume of the final solidified product compared to the original waste liquid volume, greatly increasing the space occupied by subsequent transportation and disposal facilities and the overall cost.

[0005] Vitrification involves melting nuclear waste liquid with a glass-forming agent at high temperatures, followed by cooling to form a stable vitrified body. Vitrified bodies possess a dense microstructure and excellent chemical stability, making them ideal containment matrices for most nuclides. However, during the high-temperature melting process, TcO4… - In the absence of a reducing atmosphere, the unstable material will transform into volatile technetium heptoxide (Tc2O7). A large amount of Tc2O7 will vaporize and escape from the molten glass into the gas phase. This not only significantly reduces the inclusion capacity of technetium in the final solidified body and greatly reduces the curing efficiency, but also causes the volatilized technetium-containing radioactive gas to condense at the cold end of the curing equipment, resulting in radioactive contamination of the equipment. This makes subsequent equipment maintenance and secondary waste treatment extremely complex and costly.

[0006] To address the aforementioned shortcomings, some novel TcO4... - Treatment methods have emerged accordingly. For example, Chinese patent publication number CN112851573A, filed on December 25, 2020, discloses a method for recovering technetium from nuclear fuel reprocessing waste liquid. This method employs a large cation precipitation method to selectively precipitate pertechnetate anions (TcO4) from the nuclear fuel reprocessing waste liquid. - This method yields a relatively pure technetium product, greatly simplifying the technetium extraction process. However, this method focuses on "recovery," and the high-purity Tc-99 obtained through precipitation is still high-level radioactive waste that needs to be disposed of.

[0007] Chinese patent CN110144471A, filed on May 15, 2019, discloses a method for extracting technetium from nuclear fuel reprocessing wastewater. It uses tris(N'N-dialkylacetamamido)amine as the extractant. Tris(N'N-dialkylacetamamido)amine exhibits high selectivity for extracting technetium in nitric acid media, but its extraction ability for fragmented elements such as strontium, cesium, and ruthenium, as well as trivalent, pentavalent, and hexavalent actinides, is weak. Combined with the special complexing and back-extraction effects of oxalic acid and EDTA on tetravalent actinides loaded in the organic phase of tris(N'N-dialkylacetamamido)amine, the extraction and purification of technetium from nuclear fuel reprocessing nitric acid wastewater is effectively achieved. However, organic materials are prone to degradation under strong radiation fields, leading to extraction performance failure. Therefore, the long-term operational stability and the cost of replenishing and replacing the extractant need to be considered.

[0008] In summary, existing technologies for processing the radionuclide technetium all face problems such as "low solidification efficiency, poor product stability, and large volume increase ratio" or "harsh process conditions, high risk of secondary contamination, and high cost." Therefore, there is an urgent need in this field to develop a method that can operate under mild conditions and efficiently solidify TcO4. - A novel solidification method that can minimize waste and avoid the risk of secondary pollution. Summary of the Invention

[0009] To address the problems existing in the prior art, this invention provides a method for curing the radioactive nuclide technetium in solution and its application. The curing method has mild operating conditions, a simple process flow, and the advantages of high curing efficiency, small volume of cured product, low cost, and environmental friendliness.

[0010] The technical solution of the present invention is as follows: One of the objectives of this invention is to provide a method for curing the radioactive nuclide technetium in solution. Nitric acid is added to a technetium-containing solution to carry out an acidification reaction, and then the acidified technetium-containing solution is added to a sodium aluminate solution to carry out a curing reaction. After the reaction is completed, the solid and liquid are separated and dried to obtain a technetium-containing cured product.

[0011] Furthermore, the concentration of technetium in the technetium-containing solution is 1-300 mg / L. Furthermore, the pH value of the acidified technetium-containing solution is in the range of 1-3, and the concentration of nitric acid is 0.1-1 mol / L. Furthermore, the molar ratio of sodium aluminate to nitric acid in the acidified technetium-containing solution is 1:1-1.4. Furthermore, the concentration of the sodium aluminate solution is 0.1-1 mol / L. Furthermore, the acidified technetium-containing solution was added to sodium aluminate solution at a constant and slow rate, with the flow rate controlled at 1.0-2.0 mL / min. Furthermore, the curing temperature is 10-50℃, and the curing time is 10-50 min. Furthermore, the solid-liquid separation is achieved by membrane filtration or centrifugation.

[0012] Furthermore, filtration is performed using 0.45µm or 0.22µm filter membranes.

[0013] Furthermore, the filtered residue was dried in a 60°C forced-air drying oven for 12 hours to obtain the solidified product. Furthermore, the technetium-containing cured product is a pseudo-boehmite mixed crystal, wherein technetium is cured through covalent bonding.

[0014] The second objective of this invention is to provide a method for solidifying the radioactive nuclide technetium in solution and its application in the treatment of technetium-containing acidic nuclear waste liquid.

[0015] Compared with the prior art, the beneficial effects of the present invention are as follows: 1. Unlike TcO4 in conventional nuclear waste liquid - The present invention proposes for the first time a curing method that utilizes the reaction of sodium aluminate with an acidic technetium-containing solution to cure technetium. The curing mechanism involves the removal of easily migratable TcO4. - It covalently bonds with the highly active μ1-OH in the solidified product, capturing and firmly embedding itself in the pseudo-boehmite mixed crystal mineral structure. This not only overcomes the defects of easy leaching caused by the physical encapsulation of cement solidification and the risk of high-temperature volatilization caused by glass solidification, but is also more direct, pure and reliable than the extraction or precipitation methods in the prior art that involve multiple steps of separation and are prone to generating secondary organic waste.

[0016] 2. The method for curing the radioactive nuclide technetium provided by this invention comprises only two steps: "acidification" and "mixed curing." The operating conditions are mild, requiring no complex equipment or high-temperature melting devices. However, this curing method features a small volume of cured product and high curing efficiency, with a maximum curing amount reaching 100 mg / g. Example data shows that for technetium-containing solutions with an initial concentration as high as 200 mg / L, after treatment by the method of this invention, the technetium concentration in the filtrate can be reduced to undetectable levels, and the curing amount of technetium can reach approximately 40-59 mg / g, demonstrating extremely high curing efficiency.

[0017] 3. The innovative solidification method of this invention involves only inorganic reagents such as nitric acid and sodium aluminate, eliminating the generation of organic radioactive secondary waste. Its low-temperature operation significantly reduces energy consumption, making it particularly suitable for nuclear waste liquid scenarios with large processing volumes and high processing standards. Furthermore, the final solidified product produced by this method is a mineral-like substance with a stable crystal structure. It is small in size, chemically inert, and has strong resistance to leaching, directly meeting the long-term safety and stability standards required by international standards for high-level radioactive waste geological repositories. This is an environmentally friendly, low-cost, and simple spent fuel reprocessing technology that can alleviate the pressure on spent fuel reprocessing at nuclear power plants and has significant value for the sustainable development of the nuclear energy industry. Attached Figure Description

[0018] Figure 1 The XRD patterns of the cured product obtained in this invention and the blank control sample are shown. Detailed Implementation

[0019] The present invention will be further described below with reference to preferred embodiments. The endpoints and any values ​​of the ranges disclosed in the present invention are not limited to the precise ranges or values. These ranges or values ​​should be understood to include values ​​close to these ranges or values. For numerical ranges, the endpoint values ​​of the various ranges, the endpoint values ​​of the various ranges and individual point values, and individual point values ​​can be combined with each other to obtain one or more new numerical ranges, which should be regarded as specifically disclosed herein. Unless otherwise specified, the experimental methods in the following embodiments are conventional methods, performed in accordance with the techniques or conditions described in the literature in this field or in accordance with the product instructions.

[0020] Unless otherwise specified, all materials and reagents used in the following examples are commercially available. Example 1 This embodiment provides a method for solidifying the radioactive nuclide technetium in solution, including the following steps: S1, using nitric acid to acidify TcO4 - A technetium-containing solution with a concentration of 200 mg / L was prepared until its pH reached 2; S2. Take 30 mL of 0.167 mol / L sodium aluminate solution and place it in a 30℃ constant temperature water bath. Add the acidified technetium-containing solution to the sodium aluminate solution at a flow rate of 1.2 mL / min and solidify at 30℃ for 50 min. S3. Filter the above solution through a 0.45 µm filter membrane, and dry the resulting filter residue in a 60℃ forced-air drying oven for 12 h to obtain the solidified product. Determine the TcO4 content in the filtrate above. - The concentration of TcO4 in the filtrate was determined, and the results showed that the concentration of TcO4 in the filtrate was [not specified]. - The concentration was 0 mg / L, and the cured product had a significant effect on TcO4. - The curing amount was 39.84 mg / g.

[0021] Example 2 This embodiment provides a method for solidifying the radioactive nuclide technetium in solution, including the following steps: S1, using nitric acid to acidify TcO4 - A technetium-containing solution with a concentration of 300 mg / L was prepared until its pH reached 1; S2. Take 30 mL of 0.167 mol / L sodium aluminate solution and place it in a 40℃ constant temperature water bath. Add the acidified technetium-containing solution to the sodium aluminate solution at a flow rate of 1.0 mL / min and solidify at 40℃ for 40 min. S3. Filter the above solution through a 0.22 µm filter membrane, and dry the resulting filter residue in a 60℃ forced-air drying oven for 12 h to obtain the solidified product; Determine the TcO4 content in the filtrate above. - The concentration of TcO4 in the filtrate was determined, and the results showed that the concentration of TcO4 in the filtrate was [not specified]. - The concentration was 6.66 mg / L, and the cured product had a significant effect on TcO4. - The curing amount was 58.18 mg / g.

[0022] Example 3 This embodiment provides a method for solidifying the radioactive nuclide technetium in solution, including the following steps: S1, using nitric acid to acidify TcO4 - A technetium-containing solution with a concentration of 300 mg / L was prepared until its pH reached 3; S2. Take 30 mL of 0.167 mol / L sodium aluminate solution and place it in a 50℃ constant temperature water bath. Add the acidified technetium-containing solution to the sodium aluminate solution at a flow rate of 2.0 mL / min and solidify at 50℃ for 10 min. S3. Filter the above solution through a 0.45 µm filter membrane, and dry the resulting filter residue in a 60℃ forced-air drying oven for 12 h to obtain the solidified product. Determine the TcO4 content in the filtrate above. - The concentration of TcO4 in the filtrate was determined, and the results showed that the concentration of TcO4 in the filtrate was [not specified]. - The concentration was 24.14 mg / L, and the cured product had a significant effect on TcO4. - The curing amount was 58.67 mg / g.

[0023] Performance testing 1. Curing test 30 mL of a 0.167 mol / L sodium aluminate solution was placed in a 30℃ constant temperature water bath. 0.1 mol / L nitric acid was added to the sodium aluminate solution at a flow rate of 1.2 mL / min. After reacting for 50 min, the solution was filtered through a 0.45 µm filter membrane. The resulting filter residue was dried in a 60℃ forced-air drying oven for 12 h to obtain a blank control sample of the cured product. The curing effect of the cured product obtained in Example 1 and the blank control sample of the cured product were tested respectively. The test results are as follows: Figure 1 As shown.

[0024] like Figure 1 As shown, during the crystal growth process of pseudoboehmite, TcO4 - By forming covalent bonds and embedding themselves within its crystal lattice, a pseudo-boehmite mixed crystal is formed, which in turn causes changes in the position of the characteristic peaks in its XRD diffraction pattern.

[0025] The above description is merely an embodiment of the present invention and does not limit the patent scope of the present invention. Any equivalent modifications made based on the content of the present invention specification and drawings, or direct or indirect applications in related technical fields, are similarly included within the patent protection scope of the present invention.

Claims

1. A method for solidifying the radioactive nuclide technetium in solution, characterized in that, Nitric acid was added to the technetium-containing solution to carry out an acidification reaction. Then, the acidified technetium-containing solution was added to a sodium aluminate solution to carry out a curing reaction. After the reaction was completed, the solid and liquid were separated and dried to obtain the technetium-containing cured product.

2. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The concentration of technetium in the technetium-containing solution is 1-300 mg / L.

3. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The pH range of the acidified technetium-containing solution is 1-3, and the concentration of nitric acid is 0.1-1 mol / L.

4. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The molar ratio of sodium aluminate to nitric acid in the acidified technetium-containing solution is 1:1-1.

4.

5. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The concentration of the sodium aluminate solution is 0.1-1 mol / L.

6. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The acidified technetium-containing solution was added to sodium aluminate solution at a constant and slow rate of 1.0-2.0 mL / min.

7. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The curing temperature is 10-50℃ and the curing time is 10-50 min.

8. The method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The solid-liquid separation is achieved by membrane filtration or centrifugation.

9. A method for solidifying the radioactive nuclide technetium in solution according to claim 1, characterized in that, The technetium-containing cured product is a pseudo-boehmite mixed crystal, wherein technetium is cured through covalent bonding.

10. The application of a method for solidifying the radionuclide technetium in a solution according to any one of claims 1 to 9 in the treatment of technetium-containing acidic nuclear waste liquid.

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