Separation method for uranium-233 in thorium oxide target after reactor irradiation
By employing a dual-column combined method and nitric acid solution rinsing technology, uranium-233 was successfully separated and purified from thorium oxide targets after reactor irradiation. This solved the problem of separation difficulties in existing technologies, achieving efficient uranium-233 separation and thorium matrix recovery with good radiation resistance, thus meeting the requirements for high purity.
Patent Information
- Authority / Receiving Office
- CN · China
- Patent Type
- Applications(China)
- Current Assignee / Owner
- Filing Date
- 2025-11-11
- Publication Date
- 2026-03-27
AI Technical Summary
Existing technologies are difficult to efficiently separate and purify trace amounts of uranium-233 from thorium oxide targets after reactor irradiation. Furthermore, the separation process is hampered by problems such as dissolution difficulties, high radioactivity, high thorium matrix content, and low uranium-233 content, making it difficult to meet high purity requirements.
A dual-column method using methyltrioctylammonium chloride extraction resin and dipentyl phosphate extraction resin was employed to achieve efficient separation and purification of uranium-233 through target dissolution, column separation, and purification, while recovering the thorium matrix. Nitric acid solution was used for rinsing and desorption, and thorium oxide was treated by calcination.
It achieves efficient separation and purification of trace uranium-233 from a large amount of thorium matrix and fission products, with a thorium removal rate of not less than 99.99% and a total uranium-233 recovery rate of not less than 90%, reducing the amount of radioactive waste and meeting high purity requirements.
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Figure CN121737490A_ABST
Abstract
Description
Technical Field
[0001] This invention relates to the fields of radiochemistry, nuclide separation and nuclear measurement technology, and specifically to a method for separating uranium-233 from a thorium oxide target after reactor irradiation. Background Technology
[0002] Uranium-233 is a synthetic isotope of fissile uranium with a half-life of 1.592 × 10⁻⁶. 5 Nuclear energy is an important strategic resource, mainly used in fields such as nuclear industry, geological exploration, nuclear optical clock manufacturing, and nuclear medicine.
[0003] Uranium-233, as a key isotope diluent in isotope dilution mass spectrometry (IDMS), plays an irreplaceable role in the precise and accurate measurement of uranium. It is widely used in burnup analysis of irradiated nuclear fuel and U-Pb dating in geological sciences. IDMS is a high-precision chemical analysis method widely used for the quantitative analysis of elements and compounds. Its core principle is to add a known amount of a concentrated isotope of the analyte (the diluent) to the sample, mixing it homogeneously with the sample's isotopes to change the isotopic abundance of the analyte. Mass spectrometry is then used to determine the proportion of that isotope abundance in the mixed sample. Finally, the concentration of the analyte in the sample can be calculated using the IDMS formula.
[0004] Furthermore, thorium-229 nuclear optical clocks, based on the time-frequency standard of thorium-229 nuclear transitions, provide a higher-precision time-frequency standard applicable to the study of fundamental physical laws. They have wide-ranging applications in civilian and military fields such as high-precision navigation and positioning, precision strikes, quantum computing, structural detection of materials and their surfaces, solar cells, and geogravimetry and seismic detection. Uranium-233, as a key material for preparing thorium-229 nuclear optical clocks, can produce thorium-229 through the alpha decay of uranium-233. The thorium-229 ions produced by the decay of exempt-grade uranium-233 can meet the research requirements for ion-state thorium nuclear optical clocks. Increasing the yield of uranium-233 to the milligram level could meet the development needs of solid-state nuclear optical clocks.
[0005] Meanwhile, uranium-233 can also be used to prepare uranium-233 activity standard sources for gamma spectrometer measurements, as a raw material for the production of actinium-225, a raw material for medical isotope generators (used for targeted therapy of prostate cancer), as a key core material for thorium-based molten salt reactors, and for the development of new nuclear energy.
[0006] Uranium-233 is primarily produced through artificial nuclear reactions. It is generated by irradiating thorium-232 in a reactor; the thorium-232 captures neutrons and undergoes two beta decays. The nuclear reaction formula is as follows:
[0007] Thorium targets irradiated by reactors are characterized by their difficulty in dissolution, high radioactivity, high thorium matrix content, low uranium-233 content, and the presence of various fission products. The separation and purification of uranium-233 is therefore complex and challenging. Uranium separation and purification can generally be achieved through methods such as ion exchange and dissolution extraction to remove impurity nuclides. The type of separation material and the separation conditions are crucial to successful separation. Although similar studies have been conducted previously, most have not been validated through thermal experiments, and the separation performance of the system under strong radioactivity conditions requires further verification. Furthermore, some applications demand higher uranium-233 purity, placing even more stringent requirements on the separation and purification process.
[0008] Therefore, there is an urgent need to establish a method that has good separation effect, good radiation resistance, low cost, and sufficient application verification to solve the problem of separating and preparing uranium-233 from reactor irradiated thorium oxide targets. Summary of the Invention
[0009] The present invention aims to solve at least one of the technical problems existing in the prior art or related art.
[0010] The purpose of this invention is to propose a highly efficient, radiation-resistant method for separating and purifying trace amounts of uranium-233 in a high-concentration thorium matrix by means of target dissolution, column separation and purification, and to achieve effective recovery of the thorium matrix, targeting uranium-233 in thorium oxide targets after reactor irradiation.
[0011] Therefore, the present invention provides a method for separating uranium-233 from a thorium oxide target after reactor irradiation, comprising the following steps: Dissolution of thorium oxide target: The thorium oxide target is placed in a dissolution vessel, and nitric acid solution, hydrofluoric acid and aluminum nitrate are added in sequence. Then the mixture is heated until the thorium oxide target is completely dissolved to obtain the sample solution. Resin column pretreatment: The first resin column was packed with methyl trioctyl ammonium chloride extraction resin, and the second resin column was packed with dipentyl pentyl phosphate extraction resin. The first resin column and the second resin column were pretreated respectively. Target solution loading: The sample solution is loaded onto the first resin column, and the effluent is loaded onto the second resin column; Purification of uranium-233: The second resin column was rinsed with a nitric acid solution of the first concentration to remove impurities and thorium, and then rinsed with a nitric acid solution of the second concentration to desorb uranium-233. The rinsing solution containing uranium-233 was collected.
[0012] Further, the dissolution step of the thorium oxide target includes: The mass of the thorium oxide target is 100mg~500mg; The nitric acid solution has a volume of 10 mL to 50 mL and a concentration of 2 mol / L to 5 mol / L. The volume of the hydrofluoric acid is 10 μL to 50 μL; The mass of the aluminum nitrate is 0.01g~0.05g; The heating temperature is 120℃~160℃.
[0013] Further, the resin column pretreatment step includes: The first resin column is made of polypropylene or quartz glass with an inner diameter of 6 mm to 15 mm. It is packed using a wet packing method, filled with methyltrioctylammonium chloride extraction resin, and the column height is controlled to be 10 cm to 20 cm. After packing, the resin column is rinsed with 10 mL to 50 mL of nitric acid solution with a concentration of 2 mol / L to 5 mol / L at a flow rate of 1 mL / min to 3 mL / min. Then, the resin column is rinsed with deionized water at a flow rate of 1 mL / min to 3 mL / min until neutral. The second resin column is made of polypropylene or quartz glass with an inner diameter of 6mm to 15mm. It is packed using a wet packing method, filled with diammonium pentyl phosphate extraction resin, and the column height is controlled to be 5cm to 15cm. After packing, the resin column is rinsed with 10mL to 50mL of nitric acid solution with a concentration of 2mol / L to 5mol / L at a flow rate of 1mL / min to 3mL / min. Then, the resin column is rinsed with deionized water at a flow rate of 1mL / min to 3mL / min until neutral.
[0014] Furthermore, the resin column pretreatment step further includes: Before filling with resin, the methyl trioctyl ammonium chloride extraction resin and the dipentyl phosphate extraction resin were soaked in 2 mol / L to 5 mol / L nitric acid solution for 12 h to 24 h respectively.
[0015] Further, the target material dissolution solution loading step includes: The sample solution is loaded onto the first resin column at a flow rate of 0.5 mL / min to 1.0 mL / min, and then the effluent is loaded onto the second resin column at a flow rate of 0.5 mL / min to 1.0 mL / min.
[0016] Furthermore, the target material dissolution solution loading step also includes: Before loading the sample solution, the first resin column and the second resin column are rinsed with 10 mL to 50 mL of nitric acid solution with a concentration of 2 mol / L to 5 mol / L at a flow rate of 0.8 mL / min to 1.2 mL / min.
[0017] Furthermore, the purification steps of the uranium-233 include: The first concentration of nitric acid solution is 1 mol / L to 3 mol / L, the volume is 80 mL to 180 mL, and the rinsing flow rate is 0.8 mL / min to 1.5 mL / min, used to remove impurities and thorium; The second concentration of nitric acid solution is 0.005 mol / L to 0.05 mol / L, the amount used is 20 mL to 80 mL, the rinsing flow rate is 0.5 mL / min to 1.0 mL / min, and it is used to desorb uranium-233.
[0018] Furthermore, the separation method further includes, after the purification step of uranium-233, the separation and recovery of thorium; The separation and recovery of thorium includes: The first and second resin columns were connected in series and rinsed with 5 mL to 10 mL of 2 mol / L to 5 mol / L nitric acid solution at a flow rate of 0.8 mL / min to 1.5 mL / min to remove metal ions that were not firmly adsorbed. Then, the first and second resin columns were rinsed with 80 mL to 150 mL of 1 mol / L to 4 mol / L nitric acid solution at a flow rate of 0.8 mL / min to 1.5 mL / min to remove impurities and uranium-233. The first resin column and the second resin column were separated. The first resin column was rinsed with 0.05 mol / L to 0.2 mol / L nitric acid solution at a flow rate of 0.5 mL / min to 1.0 mL / min to recover the thorium sample solution. The thorium sample solution was evaporated and dried, and then calcined to obtain thorium oxide solid.
[0019] Furthermore, the drying process includes drying at 100℃~120℃ for 7h~10h.
[0020] Furthermore, the calcination includes: heating to 600℃~800℃ and calcining for 10h~14h to obtain thorium oxide solid.
[0021] Furthermore, the separation method achieves a thorium removal rate of no less than 99.99%, a total recovery rate of uranium-233 of no less than 90%, a total recovery rate of thorium of no less than 90%, and a total removal rate of γ nuclides of no less than 99.99%.
[0022] Compared with the prior art, the present invention has at least the following beneficial effects: The present invention provides a method for separating uranium-233 from thorium oxide targets after reactor irradiation. The method employs a dual-column system of methyl trioctyl ammonium chloride extraction resin and dipentyl phosphate extraction resin to separate and purify uranium-233 from thorium oxide targets after reactor irradiation. This method can efficiently separate and purify trace amounts of uranium-233 from a large amount of thorium matrix and various fission products, with a thorium removal rate of not less than 99.99% and a total recovery rate of uranium-233 after separation of not less than 90%. Attached Figure Description
[0023] To more clearly illustrate the technical solutions in the embodiments of the present invention, the accompanying drawings used in the description of the embodiments will be briefly introduced below. Obviously, the accompanying drawings described below are only some embodiments recorded in the embodiments of the present invention. For those skilled in the art, other drawings can be obtained based on these drawings.
[0024] Figure 1 This is a flow chart of the uranium-233 separation and purification process provided in an embodiment of the present invention. Detailed Implementation
[0025] To better understand the above technical solutions, the technical solutions of the embodiments of this application will be described in detail below with reference to the accompanying drawings and specific embodiments. It should be understood that the embodiments of this application and the specific features in the embodiments are detailed descriptions of the technical solutions of the embodiments of this application, rather than limitations on the technical solutions of this application. In the absence of conflict, the embodiments of this application and the technical features in the embodiments can be combined with each other.
[0026] This invention provides a method for separating uranium-233 from a thorium oxide target after reactor irradiation, comprising the following steps: Dissolution of thorium oxide target: Place the thorium oxide target in a dissolution vessel, add nitric acid solution, hydrofluoric acid and aluminum nitrate in sequence, and then heat until the thorium oxide target is completely dissolved to obtain the sample solution; Resin column pretreatment: The first resin column was packed with methyl trioctyl ammonium chloride extraction resin, and the second resin column was packed with diammonium pentyl phosphate extraction resin. The first and second resin columns were pretreated respectively. Target solution loading: The sample solution is loaded onto the first resin column, and the effluent is loaded onto the second resin column; Purification of uranium-233: The second resin column was rinsed with a nitric acid solution of the first concentration to remove impurities and thorium, and then rinsed with a nitric acid solution of the second concentration to desorb uranium-233. The rinsing solution containing uranium-233 was collected.
[0027] The method for separating uranium-233 from thorium oxide targets after reactor irradiation provided in this invention employs a dual-column separation process. For trace uranium in a large thorium matrix and various fission products, a polyacrylate backbone resin is used, loaded with methyltrioctylammonium chloride (A336) and dipentyl phosphate (DAAP) extractants to prepare an extraction resin for the separation and purification of uranium-233, ultimately yielding a high-purity uranium-233 product. Specifically, the dual-column combination of methyltrioctylammonium chloride and dipentyl phosphate extraction resins for the separation and purification of uranium-233 from thorium oxide targets after reactor irradiation can efficiently separate and purify trace uranium-233 from a large thorium matrix and various fission products, with a thorium removal rate of not less than 99.99% and a total uranium-233 recovery rate of not less than 90%.
[0028] In some embodiments, the thorium oxide target dissolution step includes: Addition of co-solvent: Add 100mg~500mg of irradiated high-purity ThO2 target material to the sample dissolution vessel, and then add 10mL~50mL of 2mol / L~5mol / L nitric acid solution, 10μL~50μL of hydrofluoric acid and 0.01g~0.05g of aluminum nitrate in sequence; Heating and dissolving: Place the dissolving vessel on an electric heating plate and heat it at a temperature of 120℃~160℃ until ThO2 is completely dissolved. This is the sample dissolving solution Y1.
[0029] Specifically, due to the extremely stable chemical properties of thorium oxide, it exhibits strong resistance to most common acids, making it difficult to dissolve. In this embodiment of the invention, a 2 mol / L to 5 mol / L nitric acid solution is used, with the addition of hydrofluoric acid and aluminum nitrate as co-solvents. This achieves rapid and complete dissolution of thorium oxide, ensuring the acidity of the solution matches that of the separation system. This avoids the need for evaporation and system transfer under high radioactivity conditions, simplifying the operation and reducing personnel radiation exposure. The volume of the solution and the amounts of hydrofluoric acid and aluminum nitrate should generally be as small as possible, while ensuring complete dissolution of the target material, to avoid affecting subsequent separation results.
[0030] In some embodiments, the resin column pretreatment step includes: Resin pretreatment: A336 extraction resin and DAAP extraction resin were soaked in 2 mol / L to 5 mol / L nitric acid solution for 12 h to 24 h respectively.
[0031] First resin column packing: The first resin column is made of polypropylene or quartz glass with an inner diameter of 6mm~15mm; wet packing is used, and methyltrioctylammonium chloride extraction resin is packed, with the column height controlled at 10cm~20cm; after packing, the resin column is washed with 10mL~50mL of 2mol / L~5mol / L nitric acid solution at a flow rate of 1mL / min~3mL / min, and then washed with deionized water at a flow rate of 1mL / min~3mL / min until neutral; this is the first resin column, which is kept for later use.
[0032] Second resin column packing: The second resin column is made of polypropylene or quartz glass with an inner diameter of 6mm~15mm; wet packing is used, filling it with diammonium pentyl phosphate extraction resin, and controlling the column height to be 5cm~15cm; after packing, the resin column is washed with 10mL~50mL of 2mol / L~5mol / L nitric acid solution at a flow rate of 1mL / min~3mL / min, and then washed with deionized water at a flow rate of 1mL / min~3mL / min until neutral; this is the second resin column, which is kept for later use.
[0033] It is understandable that by controlling the concentration of nitric acid and the rinsing rate used in resin pretreatment, impurities in the resin can be effectively removed and exchange sites activated.
[0034] In some embodiments, the target solution loading step includes: Resin column pre-equilibration: Before loading the sample solution, rinse the first and second resin columns with 10 mL to 50 mL of nitric acid solution with a concentration of 2 mol / L to 5 mol / L at a flow rate of 0.8 mL / min to 1.2 mL / min.
[0035] Sample loading: The sample solution is loaded into the first resin column at a flow rate of 0.5 mL / min to 1.0 mL / min, and then the eluent is loaded into the second resin column at a flow rate of 0.5 mL / min to 1.0 mL / min.
[0036] Specifically, the resin column is pre-equilibrated before loading the solution to ensure that the acid system inside the resin column is consistent with the subsequent loading solution, thus avoiding dilution of the sample during loading. The first and second resin columns are connected in series. By controlling the flow rate of the sample loading, thorium in the sample solution is adsorbed onto the first resin column, while uranium-233 in the sample solution is adsorbed onto the second resin column.
[0037] In some embodiments, the purification step of uranium-233 includes: The first concentration of nitric acid solution is 1 mol / L to 3 mol / L, the volume is 80 mL to 180 mL, and the rinsing flow rate is 0.8 mL / min to 1.5 mL / min, used to remove impurities and thorium; The second concentration of nitric acid solution is 0.005 mol / L to 0.05 mol / L, the amount used is 20 mL to 80 mL, and the rinsing flow rate is 0.5 mL / min to 1.0 mL / min, used for desorption of uranium-233.
[0038] Specifically, the first and second resin columns are separated, and the second resin column is eluted to further purify the crude uranium-233 product. Experiments have demonstrated that the target metal ions can only be effectively eluted from the resin column at the aforementioned acidity and flow rate. The amount of eluent must be precisely controlled: too little will result in incomplete elution, while too much will dilute the product and increase the burden on subsequent processing.
[0039] In some embodiments, the separation method further includes the separation and recovery of thorium after the uranium-233 purification step; Thorium separation and recovery, including: The first and second resin columns were connected in series and eluted with 5-10 mL of 2-5 mol / L nitric acid solution at a flow rate of 0.8-1.5 mL / min to remove unadsorbed metal ions. Then, the first and second resin columns were eluted with 80-150 mL of 1-4 mol / L nitric acid solution at a flow rate of 0.8-1.5 mL / min to remove impurities and uranium-233. The first and second resin columns were separated. The first resin column was eluted with 0.05 mol / L to 0.2 mol / L nitric acid solution at a flow rate of 0.5 mL / min to 1.0 mL / min to recover the thorium sample solution. Thorium sample solution was evaporated and dried, then calcined to obtain thorium oxide solid; wherein, drying included drying at 100℃~120℃ for 7h~10h; calcination included heating to 600℃~800℃ and calcining for 10h~14h to obtain thorium oxide solid.
[0040] Specifically, the purpose of thorium separation and recovery is to purify and reuse unreacted thorium oxide from the irradiation process, while simultaneously reducing radioactive waste. The main steps include: removal of residual impurities and uranium from the first resin column, thorium leaching, and the conversion of thorium nitrate to thorium oxide. Experimental verification has shown that thorium oxide recovery can only be achieved under the aforementioned acidity and flow rate conditions. The conversion of thorium nitrate to thorium oxide requires calcination at 600℃–800℃. This temperature range ensures complete decomposition of thorium nitrate, and the resulting thorium oxide exhibits high purity and stable morphology.
[0041] In summary, the method for separating uranium-233 from a thorium oxide target after reactor irradiation provided by the embodiments of the present invention has the following beneficial effects: (1) Using A336 extraction resin and DAAP extraction resin in a dual-column configuration, uranium-233 in reactor irradiated thorium oxide targets can be separated and purified efficiently from a large amount of thorium matrix and various fission products. The thorium removal rate is not less than 99.99%, and the total recovery rate of uranium-233 after separation is not less than 90%.
[0042] (2) By using a dual-column separation process of A336 extraction resin and DAAP extraction resin, a large amount of thorium matrix can be successfully recovered and the thorium raw material can be purified. The total recovery rate of thorium is not less than 90%, which effectively reduces the amount of radioactive waste and makes reserves for subsequent raw material recycling.
[0043] (3) The dual-column separation process using A336 extraction resin and DAAP extraction resin can effectively remove γ nuclides, with a total γ nuclide removal rate of not less than 99.99%.
[0044] Example 1: A method for separating uranium-233 from a thorium oxide target after reactor irradiation. See Figure 1 It includes the following steps: Step 1: Dissolution of thorium oxide target 1.1 Addition of co-solvent: Add 100 mg of irradiated ThO2 to the sample dissolution vessel, and then add 10 mL of 2 mol / L nitric acid solution, 10 μL of hydrofluoric acid, and 0.01 g of aluminum nitrate in sequence.
[0045] 1.2 Heating and dissolving: Place the dissolving vessel on an electric heating plate and heat it to 135℃ until ThO2 is completely dissolved. This is the sample solution Y1.
[0046] Step 2: Resin Column Pretreatment 2.1 Resin pretreatment: A336 extraction resin and DAAP extraction resin were used for dual-column separation, and the resins were soaked in 2 mol / L nitric acid solution for 12 h.
[0047] 2.2 First Column Packing: The resin column has an inner diameter of 8 mm and a length of 20 cm, made of PP or quartz glass. A336 extraction resin was packed using a wet method, with the column height controlled at 10 cm. The resin column was rinsed with 20 mL of 2 mol / L nitric acid solution, and finally rinsed with deionized water until neutral. The flow rate was adjusted to 1.0 mL / min. This is the first resin column, which was kept for later use.
[0048] 2.3 Second Column Packing: The resin column has an inner diameter of 8 mm and a length of 20 cm, made of PP or quartz glass. A wet packing method is used for DAAP extraction resin packing, maintaining a column height of around 10 cm. The resin column is rinsed with 20 mL of 2 mol / L nitric acid, and finally rinsed with deionized water until neutral. The flow rate is adjusted to 1 mL / min. This is the second resin column, which is then stored for later use.
[0049] Step 3: Loading the target material solution onto the column 3.1 Resin column pre-equilibration: Before loading the sample, the first and second resin columns were rinsed with 20 mL of 2 mol / L nitric acid, and the flow rate was adjusted to 1.0 mL / min.
[0050] 3.2 Sample loading: After the nitric acid solution has just dried out, connect the first resin column and the second resin column in series, then add the sample solution Y1 to be separated into the resin column, adjust the flow rate to 0.5 mL / min, and replace the receiving bottle at the same time.
[0051] Step 4: Purification of Uranium-233 First, separate the first resin column and the second resin column.
[0052] 4.1 The second resin column was rinsed with 100 mL of 2 mol / L nitric acid solution to remove impurities and thorium from the second resin column. The flow rate was adjusted to 1.0 mL / min.
[0053] 4.2 The second resin column was rinsed with 25 mL of 0.01 mol / L nitric acid solution to wash down the uranium-233 on the second resin column. The flow rate was adjusted to 0.8 mL / min to obtain the uranium-233 sample solution Y3.
[0054] Step 5: Thorium separation and recovery 5.1 After the uranium-233 sample solution Y3 has just dried out, connect the first resin column and the second resin column in series, add 5 mL of 2 mol / L nitric acid solution to continue rinsing the first resin column and the second resin column to wash away the metal ions that have not been firmly adsorbed, and adjust the flow rate to 1.0 mL / min.
[0055] 5.2 After the nitric acid solution from step 5.1 has dried, add 80 mL of 2 mol / L nitric acid solution to rinse the first and second resin columns to remove impurities and uranium-233. Adjust the flow rate to 0.8 mL / min.
[0056] 5.3 Separate the first resin column and the second resin column. Elute the first resin column with 25 mL of 0.05 mol / L nitric acid solution to wash down Th. Adjust the flow rate to 1.0 mL / min, preserve and recover to obtain the recovered thorium sample solution Y2.
[0057] 5.4 After the recovered thorium sample solution Y2 was heated to dryness, it was first placed in an oven at 100℃ for 8 hours. Then, the solid was transferred into a muffle furnace and slowly heated to 700℃ for calcination for 12 hours to obtain the recovered thorium oxide solid.
[0058] Step 6: Sample Measurement 6.1 The concentrations of uranium and thorium in samples Y1, Y2 and Y3 were measured using inductively coupled plasma atomic emission spectrometry (ICP-AES), and the results are shown in Table 1.
[0059] 6.2 The abundance of uranium-233 in sample Y3 was determined by inductively coupled plasma mass spectrometry, and the results are shown in Table 2.
[0060] 6.3 The activity of γ nuclides in samples Y1 and Y3 was measured using a high-purity germanium γ-ray spectrometer, and the results are shown in Table 3.
[0061] Table 1. Uranium and thorium concentrations in each solution during the separation process measured by plasma atomic emission spectrometry.
[0062] Table 2. Uranium abundance in uranium-233 product solution Y3 measured by inductively coupled plasma mass spectrometry.
[0063] Table 3. Gamma nuclide removal rate in purified uranium-233 product
[0064] As shown in Table 1, after separation, the removal rate of thorium was greater than 99.99%, the total recovery rate of uranium-233 was 90.24%, and the recovery rate of thorium was 90.00%.
[0065] As can be seen from Table 2, the abundance of uranium-233 in the uranium-233 product solution is 82.125%.
[0066] As can be seen from Table 3, the removal rate of γ nuclides in the purified uranium-233 product is greater than 99.99%.
[0067] The above are merely preferred embodiments of the present invention and are not intended to limit the present invention. Various modifications and variations can be made to the present invention by those skilled in the art. Any modifications, equivalent substitutions, improvements, etc., made within the spirit and principles of the present invention should be included within the scope of protection of the present invention.
[0068] In view of the detailed description above, these and other changes can be made to these embodiments, and this written description includes embodiments of the best mode that disclose the invention. The patent scope of the invention is defined by the claims, which are not limited by this disclosure. The scope of protection of the invention is not limited thereto. Any equivalent substitutions or modifications made by those skilled in the art within the scope of the invention disclosed herein, based on the technical solutions and concepts of the invention, are within the scope of protection of the invention.
Claims
1. A method for separating uranium-233 from a thorium oxide target after reactor irradiation, characterized in that, Includes the following steps: Dissolution of thorium oxide target: The thorium oxide target is placed in a dissolution vessel, and nitric acid solution, hydrofluoric acid and aluminum nitrate are added in sequence. Then the mixture is heated until the thorium oxide target is completely dissolved to obtain the sample solution. Resin column pretreatment: The first resin column was packed with methyl trioctyl ammonium chloride extraction resin, and the second resin column was packed with dipentyl pentyl phosphate extraction resin. The first resin column and the second resin column were pretreated respectively. Target solution loading: The sample solution is loaded onto the first resin column, and the effluent is loaded onto the second resin column; Purification of uranium-233: The second resin column was rinsed with a nitric acid solution of the first concentration to remove impurities and thorium, and then rinsed with a nitric acid solution of the second concentration to desorb uranium-233. The rinsing solution containing uranium-233 was collected.
2. The separation method according to claim 1, characterized in that, The dissolution step of the thorium oxide target includes: The mass of the thorium oxide target is 100mg~500mg; The nitric acid solution has a volume of 10 mL to 50 mL and a concentration of 2 mol / L to 5 mol / L. The volume of the hydrofluoric acid is 10 μL to 50 μL; The mass of the aluminum nitrate is 0.01g~0.05g; The heating temperature is 120℃~160℃.
3. The separation method according to claim 1, characterized in that, The resin column pretreatment step includes: The first resin column is made of polypropylene or quartz glass with an inner diameter of 6 mm to 15 mm. It is packed using a wet packing method, filled with methyltrioctylammonium chloride extraction resin, and the column height is controlled to be 10 cm to 20 cm. After packing, the resin column is rinsed with 10 mL to 50 mL of nitric acid solution with a concentration of 2 mol / L to 5 mol / L at a flow rate of 1 mL / min to 3 mL / min. Then, the resin column is rinsed with deionized water at a flow rate of 1 mL / min to 3 mL / min until neutral. The second resin column is made of polypropylene or quartz glass with an inner diameter of 6mm to 15mm. It is packed using a wet packing method, filled with diammonium pentyl phosphate extraction resin, and the column height is controlled to be 5cm to 15cm. After packing, the resin column is rinsed with 10mL to 50mL of nitric acid solution with a concentration of 2mol / L to 5mol / L at a flow rate of 1mL / min to 3mL / min. Then, the resin column is rinsed with deionized water at a flow rate of 1mL / min to 3mL / min until neutral.
4. The separation method according to claim 3, characterized in that, The resin column pretreatment step further includes: Before filling with resin, the methyl trioctyl ammonium chloride extraction resin and the dipentyl phosphate extraction resin were soaked in 2 mol / L to 5 mol / L nitric acid solution for 12 h to 24 h respectively.
5. The separation method according to claim 1, characterized in that, The target solution loading process includes: The sample solution is loaded onto the first resin column at a flow rate of 0.5 mL / min to 1.0 mL / min, and then the effluent is loaded onto the second resin column at a flow rate of 0.5 mL / min to 1.0 mL / min.
6. The separation method according to claim 5, characterized in that, The target material dissolution solution loading step further includes: Before loading the sample solution, the first resin column and the second resin column are rinsed with 10 mL to 50 mL of nitric acid solution with a concentration of 2 mol / L to 5 mol / L at a flow rate of 0.8 mL / min to 1.2 mL / min.
7. The separation method according to claim 1, characterized in that, The purification steps for the uranium-233 include: The first concentration of nitric acid solution is 1 mol / L to 3 mol / L, the volume is 80 mL to 180 mL, and the rinsing flow rate is 0.8 mL / min to 1.5 mL / min, used to remove impurities and thorium; The second concentration of nitric acid solution is 0.005 mol / L to 0.05 mol / L, the amount used is 20 mL to 80 mL, the rinsing flow rate is 0.5 mL / min to 1.0 mL / min, and it is used to desorb uranium-233.
8. The separation method according to claim 1, characterized in that, The separation method further includes, after the purification step of uranium-233, the separation and recovery of thorium; The separation and recovery of thorium includes: The first and second resin columns were connected in series and rinsed with 5 mL to 10 mL of 2 mol / L to 5 mol / L nitric acid solution at a flow rate of 0.8 mL / min to 1.5 mL / min to remove metal ions that were not firmly adsorbed. Then, the first and second resin columns were rinsed with 80 mL to 150 mL of 1 mol / L to 4 mol / L nitric acid solution at a flow rate of 0.8 mL / min to 1.5 mL / min to remove impurities and uranium-233. The first resin column and the second resin column were separated. The first resin column was rinsed with 0.05 mol / L to 0.2 mol / L nitric acid solution at a flow rate of 0.5 mL / min to 1.0 mL / min to recover the thorium sample solution. The thorium sample solution was evaporated and dried, and then calcined to obtain thorium oxide solid.
9. The separation method according to claim 8, characterized in that, The drying process includes drying at 100℃~120℃ for 7h~10h; the calcination process includes heating to 600℃~800℃ and calcining for 10h~14h to obtain thorium oxide solid.
10. The separation method according to claim 1, characterized in that, The separation method has a thorium removal rate of not less than 99.99%, a total recovery rate of not less than 90% for uranium-233, a total recovery rate of not less than 90% for thorium, and a total removal rate of not less than 99.99% for γ nuclides.