A method for reprocessing spent fuel based on uranium clusters

Abstract

This invention belongs to the field of spent fuel reprocessing technology, specifically relating to a spent fuel reprocessing method based on uranium clusters, comprising the following steps: S1, disassembling the spent fuel assembly and cutting it into small fuel rods; S2, dissolving the spent fuel using hydrogen peroxide and an alkaline solution, filtering to obtain a precipitate and filtrate; S3, treating the filtrate obtained in S2 using gel electrophoresis to separate uranium clusters and a solution containing Pu, actinides, and fission products; S4, adding dilute nitric acid and hydrogen peroxide to the uranium clusters obtained in S3 to convert uranium peroxide clusters into uranium filament precipitate, filtering to obtain the precipitate, thus achieving the separation of uranium elements. This invention's spent fuel reprocessing process based on uranium clusters is a spent fuel reprocessing process with advantages such as simple procedures, low corrosiveness, no risk of nuclear proliferation, and better economic benefits compared to the PUREX process.

Description

Technical Field

[0001] This invention belongs to the field of spent fuel reprocessing technology, specifically relating to a spent fuel reprocessing method based on uranium cluster compounds. Background Technology

[0002] Nuclear power, as a safe, clean, and efficient energy source, plays a vital role in meeting people's growing energy demands and improving the ecological environment. However, with the development of nuclear power, the amount of highly radioactive spent fuel is also increasing. Based on the development of nuclear power in my country, the cumulative total of spent fuel is projected to reach 18,500 tU by 2025. If spent fuel is not properly processed, it will pose a significant threat to human health and the environment. Furthermore, the current utilization rate of nuclear fuel is very low (less than 1%), meaning that over 95% of the mass of spent fuel is unreacted uranium. Separating and recycling this unreacted uranium could significantly reduce the amount of spent fuel and improve its utilization rate. Therefore, reprocessing spent fuel is essential.

[0003] Currently, the only industrially viable reprocessing technology is the PUREX (Plutonium Uranium Redox Extraction) process. This process involves dissolving spent fuel with concentrated nitric acid, adjusting the acidity and concentration of the solution to a suitable range, and adding substances such as nitrogen dioxide to adjust the valence state of plutonium ions, ultimately resulting in uranium and plutonium ions in the solution existing in +6 and +4 valence states, respectively. Following this, a co-decontamination and separation cycle is then performed. Specifically, such as... Figure 1As shown, an organic solution composed of the extractant tributyl phosphate (TBP) and a diluent (usually kerosene, n-dodecane, etc.) is mixed with spent fuel dissolution solution. Due to TBP's strong complexing ability with U(VI) and Pu(IV) ions, uranium and plutonium are extracted together from the dissolution solution and introduced into the organic phase (i.e., 1AP solution). Other actinide ions and fission products remain in the dissolution solution and are treated as high-level radioactive waste (1AW solution). This process achieves the separation of uranium and plutonium from other actinide ions and fission products. Then, the 1AP solution is mixed with a solution containing a reducing agent such as ferrous aminosulfonate or U(IV) ions, reducing Pu(IV) to Pu(III) ions, which have a weaker complexing ability with TBP, and back-extracted into the aqueous solution (1BP). This process allows plutonium to leave the organic solution and enter the aqueous phase, achieving the separation of uranium and plutonium. Uranium is then back-extracted into an aqueous solution (1CU) using dilute nitric acid solution. Uranium and plutonium are then separately purified in their respective cycles to further remove impurities, yielding qualified uranium and plutonium products. In the plutonium purification cycle, Pu(III) ions are first oxidized to Pu(IV) ions using substances such as sodium nitrite. Then, utilizing the strong complexing ability of TBP on Pu(IV) ions, plutonium is transferred to the organic phase (2AP), further separating it from fission products and other impurities. Next, Pu(IV) ions are reduced back to Pu(III) ions and back-extracted into an aqueous solution (2BP). Concentrating this solution yields pure plutonium nitrate. Similarly, in the uranium purification cycle, the strong complexing ability of TBP on U(VI) ions is again utilized to transfer uranium to the organic phase (2DU), further separating it from fission products and other impurities. Uranium is then back-extracted into an aqueous solution (2EU) using dilute nitric acid. Concentrating this aqueous solution yields pure uranium nitrate.

[0004] While the PUREX process is relatively mature, it also has several shortcomings. For example, the process is complex, the solutions are highly acidic and corrosive, and it generates large quantities of difficult-to-handle high-level radioactive waste and organic waste. More importantly, the strong radioactivity of spent fuel dissolution liquids causes radioactive decomposition of organic solutions such as TBP and kerosene, forming complex decomposition products including dibutyl phosphate, monobutyl phosphate, butanol, carbonyl compounds, alkyl compounds, and nitroalkanes. These products have a certain complexing ability with fission products, thus reducing separation efficiency and the uranium-plutonium purification coefficient, often requiring multiple separation cycles to obtain qualified uranium and plutonium products. Furthermore, the PUREX process separates plutonium separately, posing a risk of nuclear proliferation. Therefore, we need to further improve the PUREX process or develop new reprocessing procedures. Summary of the Invention

[0005] To address the aforementioned technical problems, this invention provides a spent fuel reprocessing method based on uranium cluster compounds.

[0006] The inventive concept of this invention is that uranyl ions react with peroxide ions under alkaline conditions to form a novel type of uranium material, uranium peroxide clusters. For example... Figure 2 As shown, a uranyl ion coordinates with two peroxide and two hydroxide ions to form a hexagonal bipyramidal polyhedron. Adjacent polyhedra are interconnected by sharing edges to form a novel nanocage structure. Figure 2 The uranium peroxide cluster in the sample contains a total of 60 polyhedra with uranyl ions, hence we call it U0. 60 Its molecular size is 2.5 nm, and its molecular weight is around 20,000. The applicant of this invention has also conducted a series of studies in the field of uranium peroxide clusters in recent years, synthesizing uranium peroxide clusters with a molecular size of 4 nm and a molecular weight of around 45,000. Uranium peroxide clusters exist in solution as giant anions, which are significantly larger than common inorganic ions. Based on the above theory, this invention dissolves spent fuel in hydrogen peroxide and a solution containing hydroxide ions, converting the uranium into uranium peroxide clusters, while allowing plutonium, actinides, and fission products to exist as Plutonium peroxide clusters in the solution. 4+ 、Cs + 、Ba 2+ SeO4 2- It exists in the form of simple inorganic cations or anions. Then, by utilizing the size difference between uranium peroxide clusters and Pu, secondary actinides, and fission product ions, uranium can be separated and extracted, leading to the development of spent fuel reprocessing technologies based on uranium clusters (see...). Figure 3 ).

[0007] The present invention is specifically implemented through the following technical solution.

[0008] A spent fuel reprocessing method based on uranium clusters includes the following steps:

[0009] S1. Disassemble the spent fuel assembly and cut it into small fuel rods;

[0010] S2. Dissolve spent fuel using a mixed solvent consisting of hydrogen peroxide and alkaline solution or carbonate solution, filter, and obtain a precipitate and filtrate containing actinide products and fission products. Store the precipitate as high-level radioactive solid waste.

[0011] S3. The filtrate obtained in S2 is separated by gel electrophoresis to obtain a solution containing uranium clusters and a solution containing Pu, actinides and fission products. The solution containing Pu, actinides and fission products is stored as high-level radioactive waste.

[0012] S4. Add dilute nitric acid and hydrogen peroxide to the solution of the uranium-containing cluster obtained in S3, filter, and obtain filtrate and precipitate. Store the filtrate and the precipitate is the separated uranium ore.

[0013] Preferably, in S2, the concentration of the alkaline solution or carbonate solution is 1-5 mol / L, the mass concentration of hydrogen peroxide is 30%, and the volume ratio of hydrogen peroxide to alkaline solution or carbonate solution is approximately 1:1.

[0014] Preferably, in S2, the concentration of uranyl ions dissolved from UO2 in spent fuel under the oxidation of hydrogen peroxide is 0.5–2 mol / L.

[0015] Preferably, in S2, the alkaline solution includes ammonia water, LiOH solution, NaOH solution or KOH solution.

[0016] Preferably, in S2, the carbonate solution includes Na2CO3 solution, NaHCO3 solution, NH4HCO3 or (NH4)2CO3 solution.

[0017] Preferably, in S3, the gel electrophoresis uses commercial electrophoresis equipment, the gel is agar, the buffer is commercial NaOH-Na2CO3 solution, a voltage of 100V is applied to both electrodes, and uranium clusters are precipitated from the anode.

[0018] Preferably, in S4, the concentration of dilute nitric acid is 0.1 mol / L, the mass concentration of hydrogen peroxide is 30%, the volume ratio of dilute nitric acid to hydrogen peroxide is 1:2, and after adding dilute nitric acid and hydrogen peroxide to the solution containing uranium clusters, the pH of the system solution is 1-3.

[0019] Compared with the prior art, the present invention has the following advantages:

[0020] To more clearly compare the similarities and differences between the spent fuel reprocessing process based on uranium clusters of this invention and the PUREX process, and to highlight the advantages of this invention, we also describe the main steps of the PUREX process through... Figure 5 To demonstrate:

[0021] Similarities: The first step of both processes is the same, both requiring the disassembly and shearing of spent fuel assemblies.

[0022] The differences are as follows:

[0023] (1) The dissolution methods of spent fuel are different (S2). The PUREX process uses concentrated nitric acid to dissolve the spent fuel. Most substances in the spent fuel can be dissolved in this solution, and the chemical composition of the resulting solution is very complex. In addition, concentrated nitric acid is highly corrosive, so the facilities and equipment used in the reprocessing process need to have good corrosion resistance. The process of this invention based on uranium clusters uses mild solutions such as hydrogen peroxide, ammonia (or alkaline solutions such as carbonates and LiOH) to dissolve the fuel. Only UO2, some other actinide products and some fission products can be dissolved in this solution. The composition of the resulting solution is relatively simple and the corrosiveness of the solution is also weak.

[0024] (2) PUREX separates uranium and plutonium from spent fuel liquid by solvent extraction, but the separate separation of plutonium carries the risk of nuclear proliferation; the process based on uranium clusters in this invention separates uranium from the liquid, while plutonium, along with the minor actinide products and fission products, remains in the high-level waste liquid, thus eliminating the risk of nuclear proliferation.

[0025] (3) The purification methods of uranium are different; PUREX further purifies uranium and plutonium through TBP solvent extraction; the process of this invention based on uranium clusters is to purify uranium by adding dilute nitric acid and hydrogen peroxide to convert the uranium clusters into uranium precipitate.

[0026] (4) The preliminary products obtained by the PUREX process are uranyl nitrate and plutonium nitrate, while the preliminary product obtained by the process based on uranium clusters in this invention is uranium ore.

[0027] (5) The PUREX process is relatively complex, while the process of this invention based on uranium cluster compounds is relatively simple and does not use a large amount of organic solvents such as TBP, and there are no problems such as radiation decomposition of organic solvents.

[0028] Therefore, the spent fuel reprocessing process based on uranium clusters of the present invention is a spent fuel reprocessing process with advantages such as simple procedures and low corrosivity. Compared with the PUREX process, it has better economic benefits and is more suitable for large-scale application. Attached Figure Description

[0029] Figure 1 This is a schematic diagram of the PUREX two-loop process;

[0030] Figure 2 Uranium peroxide clusters U0 are represented using a polyhedral model and a ball-and-stick model, respectively. 60 The structure;

[0031] Figure 3 This is a schematic diagram of the spent fuel reprocessing process based on uranium clusters according to the present invention. Figure 1 ;

[0032] Figure 4 This is a schematic diagram illustrating the present invention of converting uranium into uranium peroxy clusters by dissolving spent fuel, and then separating uranium from Pu, actinides and fission products (FP) ions by gel electrophoresis.

[0033] Figure 5 This is a schematic diagram of the main processes in the PUREX workflow. Detailed Implementation

[0034] To enable those skilled in the art to better understand and implement the technical solutions of the present invention, the present invention will be further described below in conjunction with specific embodiments and accompanying drawings, but the embodiments described are not intended to limit the present invention.

[0035] Unless otherwise specified, the experimental and detection methods described in the following embodiments are conventional methods; unless otherwise specified, the reagents and materials are commercially available.

[0036] A spent fuel reprocessing method based on uranium clusters, such as Figure 3 and Figure 4 As shown, it includes the following steps:

[0037] S1. Disassemble the spent fuel assembly and cut it into small fuel rods;

[0038] S2. Use hydrogen peroxide and an alkaline solution (such as ammonia, LiOH, NaOH, or carbonates like Na2CO3 or (NH4)2CO3) to dissolve spent fuel. The dissolving solution is 30% hydrogen peroxide (by mass), and the alkaline or carbonate solution has a concentration of 1–5 mol / L. The two solutions are mixed in a 1:1 volume ratio, with hydrogen peroxide in appropriate excess. The treated spent fuel is then added. The UO2 in the spent fuel dissolves into uranyl ions under the oxidation of hydrogen peroxide, ensuring the uranyl ion concentration is controlled at 0.5–2 mol / L after dissolution. The uranyl ions coordinate with peroxide and hydroxide ions to form uranium clusters existing as giant anions. Pu, some actinides, and fission products are dissolved in this alkaline solution, and then... 4+ 、Cs + MoO4 2- It exists in the form of simple inorganic cations or anions. Insoluble actinide products and fission products are properly stored as high-level radioactive solid waste;

[0039] S3. By utilizing the difference in the direction or speed of movement of uranium cluster ions with Pu, actinides, and fission product ions during gel electrophoresis, uranium is extracted in the form of clusters; Pu, actinides, and fission products are treated together as high-level radioactive waste and properly stored.

[0040] The gel electrophoresis system is a commercially available gel electrophoresis device. Agar is used as the gel, and a commercially available NaOH-Na2CO3 solution is used as the buffer. A suitable amount of spent fuel dissolution solution is slowly added, and a voltage of 100V is applied to both electrodes. Cations in the solution move towards the cathode. Because the precipitation rate of impurity anions is slower than that of uranium cluster anions, the solution containing uranium clusters precipitates first from the anode of the device. The precipitated solution is collected. Repeating the above operation multiple times can achieve the separation of uranium clusters.

[0041] S4. Add 0.1 mol / L dilute nitric acid to the separated uranium cluster solution, and simultaneously add two volumes of 30% hydrogen peroxide. Control the pH of the solution at 1-3. The uranium cluster is converted into uranium hydrothermal ore (UO2)O2(H2O)4 precipitate. A small amount of fission product ions adsorbed on the surface of the uranium peroxide cluster remain in the solution. Separate the precipitate from the solution by filtration and properly store the filtrate. Compared with the uranium content in the spent fuel before dissolution, the recovery efficiency can reach 95%.

[0042] Example 1

[0043] A spent fuel reprocessing method based on uranium clusters includes the following steps:

[0044] S1. Disassemble the spent fuel assembly and cut it into small fuel rods;

[0045] S2. Use hydrogen peroxide and ammonia to dissolve spent fuel. The dissolving solution uses 30% hydrogen peroxide and 4 mol / L ammonia, mixed in a 1:1 volume ratio, with hydrogen peroxide in appropriate excess. Add the treated spent fuel; the UO2 in the spent fuel dissolves into uranyl ions under the oxidation of hydrogen peroxide. Ensure the uranyl ion concentration is controlled at 0.5 mol / L after dissolution. Uranyl ions coordinate with peroxide and hydroxide ions to form uranium peroxide clusters existing as giant anions. After dissolving Pu, some actinides, and fission products in this alkaline solution, Pu... 4+ 、Cs + MoO4 2- They exist in the form of simple inorganic cations or anions, and insoluble actinide products and fission products are properly stored as high-level radioactive solid waste.

[0046] S3. By utilizing the difference in the direction or speed of movement of uranium cluster ions with Pu, actinides and fission product ions during gel electrophoresis, uranium is extracted in the form of clusters. Pu, actinides and fission products are treated together as high-level radioactive waste liquid and stored properly.

[0047] The gel electrophoresis system is a commercially available gel electrophoresis device. Agar is used as the gel, and commercially available NaOH-Na2CO3 solution is used as the buffer. An appropriate amount of spent fuel dissolution solution is slowly added, and a voltage of 100V is applied to both electrodes. Cations in the solution move towards the cathode. Since the precipitation rate of impurity anions is slower than that of uranium cluster anions, the solution containing uranium clusters precipitates first from the anode of the device. The precipitated solution is collected. By repeating the above operation multiple times, the separation of uranium clusters can be achieved.

[0048] S4. Add 0.1 mol / L dilute nitric acid and two volumes of 30% hydrogen peroxide to the separated uranium cluster solution, maintaining the pH at approximately 1. The uranium peroxide clusters are converted into uranium hydrate precipitate (UO2)O2(H2O)4. A small amount of fission product ions adsorbed on the surface of the uranium peroxide clusters remain in the solution. Separate the precipitate from the solution by filtration and properly store the filtrate. The recovery efficiency can reach 95% compared to the uranium content in the spent fuel before dissolution.

[0049] Example 2

[0050] A spent fuel reprocessing method based on uranium clusters includes the following steps:

[0051] S1. Disassemble the spent fuel assembly and cut it into small fuel rods;

[0052] S2. Use hydrogen peroxide and Na2CO3 to dissolve spent fuel. The dissolving solution should be 30% hydrogen peroxide and 3 mol / L Na2CO3 solution, mixed in a 1:1 volume ratio. Hydrogen peroxide can be added in excess. The treated spent fuel is then added. The UO2 in the spent fuel dissolves into uranyl ions under the oxidation of hydrogen peroxide, ensuring the uranyl ion concentration is controlled at 2 mol / L after dissolution. Uranyl ions coordinate with peroxide and hydroxide ions to form uranium peroxide clusters existing as giant anions. Pu, some actinides, and fission products are dissolved in this alkaline solution, and then... 4+ 、Cs + MoO 2- It exists in the form of simple inorganic cations or anions; insoluble actinide products and fission products are properly stored as high-level radioactive solid waste.

[0053] S3. By utilizing the difference in the direction or speed of movement of cluster ions with Pu, actinides, and fission product ions during gel electrophoresis, uranium is extracted in the form of cluster ions; Pu, actinides, and fission products are treated together as high-level radioactive waste and properly stored.

[0054] The gel electrophoresis system is a commercially available gel electrophoresis device. Agar is used as the gel, and a commercially available NaOH-Na2CO3 solution is used as the buffer. A suitable amount of spent fuel dissolution solution is slowly added, and a voltage of 100V is applied to both electrodes. Cations in the solution move towards the cathode. Because the precipitation rate of impurity anions is slower than that of uranium cluster anions, the solution containing uranium clusters precipitates first from the anode of the device. The precipitated solution is collected. Repeating the above operation multiple times can achieve the separation of uranium clusters.

[0055] S4. Add an equal volume of 0.1 mol / L dilute nitric acid to the separated uranium peroxide cluster solution, and simultaneously add two volumes of 30% hydrogen peroxide. Control the pH of the solution at around 3. The uranium peroxide cluster is converted into uranium ore (UO2)O2(H2O)4 precipitate. A small amount of fission product ions adsorbed on the surface of the uranium peroxide cluster remain in the solution. Separate the precipitate from the solution by filtration, properly store the filtrate, and the recovery efficiency can reach 95% compared with the uranium content in the spent fuel before dissolution.

[0056] Example 3

[0057] A spent fuel reprocessing method based on uranium clusters includes the following steps:

[0058] S1. Disassemble the spent fuel assembly and cut it into small fuel rods;

[0059] S2. Use hydrogen peroxide and NaOH to dissolve spent fuel. The dissolving solution should be 30% hydrogen peroxide and 2.5 mol / L NaOH solution, mixed at a volume ratio of approximately 1:1, with hydrogen peroxide in slight excess. Add the treated spent fuel. The UO2 in the spent fuel dissolves into uranyl ions under the oxidation of hydrogen peroxide, ensuring the uranyl ion concentration is controlled at 1 mol / L after dissolution. Uranyl ions coordinate with peroxide and hydroxide ions to form uranium peroxide clusters existing as giant anions. After dissolving Pu, some actinides, and fission products in this alkaline solution, the Pu... 4+ 、Cs + MoO 2- They exist in the form of simple inorganic cations or anions, and insoluble actinide products and fission products are properly stored as high-level radioactive solid waste.

[0060] S3. By utilizing the difference in the direction or speed of movement of cluster ions with Pu, actinides and fission product ions during gel electrophoresis, uranium is extracted in the form of cluster ions. Pu, actinides and fission products are treated together as high-level radioactive waste liquid and stored properly.

[0061] The gel electrophoresis system is a commercially available gel electrophoresis device. Agar is used as the gel, and a commercially available NaOH-Na2CO3 solution is used as the buffer. A suitable amount of spent fuel dissolution solution is slowly added, and a voltage of 100V is applied to both electrodes. Cations in the solution move towards the cathode. Because the precipitation rate of impurity anions is slower than that of uranium cluster anions, the solution containing uranium clusters precipitates first from the anode of the device. The precipitated solution is collected. Repeating the above operation multiple times can achieve the separation of uranium clusters.

[0062] S4. Add an equal volume of 0.1 mol / L dilute nitric acid to the separated uranium peroxide cluster solution, and simultaneously add two volumes of 30% hydrogen peroxide. Control the pH of the solution at around 2. The uranium peroxide cluster is converted into uranium ore (UO2)O2(H2O)4 precipitate. A small amount of fission product ions adsorbed on the surface of the uranium peroxide cluster remain in the solution. Separate the precipitate from the solution by filtration, properly store the filtrate, and the recovery efficiency can reach 95% compared with the uranium content in the spent fuel before dissolution.

[0063] Therefore, the method of this invention can efficiently recover uranium from spent fuel and is a spent fuel reprocessing process with advantages such as simple procedures and low corrosivity. Compared with the PUREX process, it has better economic benefits. Comparing the two processes, the first step is the same, both requiring the disassembly and shearing of spent fuel assemblies, but the remaining steps differ, as shown in:

[0064] The dissolution methods for spent fuel differ (S2). The PUREX process uses concentrated nitric acid, in which most substances in spent fuel are soluble, resulting in a highly complex chemical composition of the solution. Furthermore, concentrated nitric acid is highly corrosive, necessitating facilities and equipment with excellent corrosion resistance in the reprocessing process. In contrast, the process based on uranium cluster compounds uses milder solutions such as hydrogen peroxide, ammonia (or alkaline solutions like carbonates or LiOH), dissolving only UO2, some other actinide products, and some fission products. This results in a relatively simpler composition and less corrosive solution.

[0065] PUREX separates uranium and plutonium from spent fuel dissolution using solvent extraction. Separating plutonium alone carries the risk of nuclear proliferation. This invention, based on a uranium cluster compound process, separates only uranium from the dissolution, leaving plutonium, along with minor actinide products and fission products, in the high-level waste liquid, thus eliminating the risk of nuclear proliferation.

[0066] The purification methods for uranium differ. PUREX further purifies uranium and plutonium through TBP solvent extraction. This invention, based on a process for uranium clusters, proposes to purify uranium by adding dilute nitric acid and hydrogen peroxide to convert the uranium clusters into uranium precipitate.

[0067] The initial products obtained from the PUREX process are uranyl nitrate and plutonium nitrate, while the initial product obtained from the process based on uranium clusters in this invention is uranium hydrothermal ore.

[0068] The PUREX process is relatively complex, while the process based on uranium cluster compounds in this invention is relatively simple, does not use large amounts of organic solvents such as TBP, and does not have problems such as radiation decomposition of organic solvents.

[0069] Obviously, those skilled in the art can make various modifications and variations to this invention without departing from its spirit and scope. Therefore, it is intended to include any modifications and variations that fall within the scope of the claims and their equivalents.

Claims

1. A spent fuel reprocessing method based on uranium cluster compounds, characterized in that, Includes the following steps: S1. Disassemble the spent fuel assembly and cut it into small fuel rods; S2. Dissolve spent fuel using a mixed solvent consisting of hydrogen peroxide and alkaline solution or carbonate solution, filter, and obtain a precipitate and filtrate containing actinide products and fission products. Store the precipitate as high-level radioactive solid waste. S3. The filtrate obtained in S2 is separated by gel electrophoresis to obtain a solution containing uranium clusters and a solution containing Pu, actinides and fission products. The solution containing Pu, actinides and fission products is stored as high-level radioactive waste. S4. Add dilute nitric acid and hydrogen peroxide to the solution of the uranium-containing cluster obtained in S3, filter, and obtain filtrate and precipitate. Store the filtrate and the precipitate is the separated uranium ore. In S2, the concentration of the alkaline solution or carbonate solution is 1~5 mol / L, the mass concentration of hydrogen peroxide is 30%, and the volume ratio of hydrogen peroxide to alkaline solution or carbonate solution is 1:

1. In S2, the concentration of uranyl ions dissolved from UO2 in spent fuel under the oxidation of hydrogen peroxide is 0.5~2 mol / L. Uranyl ions coordinate with peroxide and hydroxide ions to form uranium clusters that exist as giant anions.

2. The spent fuel reprocessing method based on uranium clusters according to claim 1, characterized in that, In S2, the alkaline solution includes ammonia water, LiOH solution, NaOH solution or KOH solution.

3. The spent fuel reprocessing method based on uranium clusters according to claim 1, characterized in that, In S2, the carbonate solution includes Na2CO3 solution, NaHCO3 solution, NH4HCO3 or (NH4)2CO3 solution.

4. The spent fuel reprocessing method based on uranium clusters according to claim 1, characterized in that, In S3, the gel electrophoresis uses commercial electrophoresis equipment, the gel is agar, the buffer is commercial NaOH-Na2CO3 solution, and a voltage of 100V is applied to both electrodes, causing uranium clusters to precipitate from the anode.

5. The spent fuel reprocessing method based on uranium clusters according to claim 1, characterized in that, In S4, the concentration of dilute nitric acid is 0.1 mol / L, the mass concentration of hydrogen peroxide is 30%, and the volume ratio of dilute nitric acid to hydrogen peroxide is 1:

2. After adding dilute nitric acid and hydrogen peroxide to the solution containing uranium clusters, the pH of the system solution is 1~3.

Images