A method for dehalogenation and vitrification of radioactive metal halide waste

Abstract

The present invention relates to a method for dehalogenating and vitrifying radioactive metal halide waste. The method comprises the following steps: mixing the radioactive metal halide waste with oxalic acid and removing the halogen from the radioactive metal halide waste by heat treatment. The vitrification method comprises the following steps: solidifying the dehalogenated waste treated by the above-mentioned method for dehalogenating radioactive metal halide waste using a vitrification substrate to form a vitrified solid. The method for dehalogenating and vitrifying radioactive metal halide waste provided by the present invention not only has a low dehalogenation temperature and high efficiency, but also has a large solidified waste package capacity. Furthermore, no new substances remain after dehalogenation, making it suitable for existing vitrification processes and having good application prospects and social value.

Description

Technical Field

[0001] The present invention relates to the technical field of radioactive waste treatment, and in particular to a method for dehalogenation and vitrification of radioactive metal halide waste. Background Art

[0002] Nuclear energy is a safe, economical, and efficient clean energy source, and one of the primary energy sources for my country to achieve its carbon peak and carbon neutrality goals. Reprocessing spent fuel, a closed nuclear fuel cycle that recovers and fully utilizes the uranium and plutonium in spent fuel, is crucial for the sustainable development of nuclear energy. Dry spent fuel reprocessing, with its advantages of radiation resistance, low criticality risk, wide reprocessing potential, and minimal radioactive waste, has become one of the most promising technologies for future advanced fuel cycles.

[0003] During the dry reprocessing of spent fuel, halides are often used as diluents, resulting in the production of molten salt waste containing metal halides. For example, molten salt electrolysis uses spent metal fuel as the anode and molten chloride salt as the electrolyte. Actinides and fission elements are separated based on the redox potential difference, and uranium and plutonium are deposited on an inert cathode. However, as fission elements accumulate in the molten salt, the efficiency of separation decreases, affecting the quality of the cathode recovery product. In this case, the molten salt must be unloaded and treated and disposed of as radioactive waste. Molten salt reactors, on the other hand, use molten fluoride mixed salts as fuel carriers and coolants. Dry reprocessing of spent fuel also typically uses fluoride as an oxidant and adsorbent. Uranium, thorium, the molten salt carrier, and fission nuclides are separated through processes such as fluorination volatilization, vacuum distillation, and molten salt extraction. Consequently, various types of fluorine-containing radioactive waste are generated.

[0004] Due to their direct contact with radioactive spent fuel, these radioactive wastes, containing large amounts of halogens (halogen content generally exceeding 40% by weight), are classified as high-level radioactive waste. The currently more established international method for treating high-level radioactive waste is to solidify it using a borosilicate glass matrix. However, the low solubility of halogens in borosilicate glass (generally less than 1% by weight) limits the waste containment capacity of the glass matrix. Furthermore, during the high-temperature melting process, the volatilization of halides can easily cause radionuclide migration. Therefore, existing glass solidification processes are no longer suitable for treating radioactive metal halide waste.

[0005] Given the high halogen content in radioactive metal halide waste, two treatment approaches are currently being proposed internationally. One involves using a material with high halogen solubility as a solidification substrate, such as the glass-bonded sodalite glass-ceramic solidification developed at Argonne National Laboratory in the United States, which has a processing capacity of 300-400 kg / batch (Bateman KJ, Morrisona M.C, Rappleye DS, Simpson MF, Frank SM, Scale up of ceramic waste forms for electrodefiner salts produced during spent fuel treatment [J], Journal of Nuclear Fuel Cycle and Waste Technology, 2015, 13:55). However, this process is complex and the waste bag capacity is low (8-14 wt%). The other approach involves first dehalogenating the waste with a dehalogenating agent before solidification. For example, the Korea Atomic Energy Research Institute used a SAP (silica-alumina-phosphate) composite prepared by a sol-gel method as a dechlorination agent to achieve efficient dechlorination of chloride molten salt waste at 650°C (Park HS, Kim IT, Cho YZ, Eun HC, Lee HS, Stabilization / solidification of radioactive salt waste by using xSiO2-yAl2O3-zP2O5 (SAP) material at molten salt state[J], Environmental Science & Technology, 2008, 42:9357). However, the dechlorinated SAP composite has poor compatibility with the cured substrate.In addition, researchers at the Pacific Northwest National Laboratory in the United States found that NH4H2PO4 can also effectively remove chlorine from chlorides at 600°C, but the dechlorinated waste contains a large amount of phosphate, which needs to be solidified using a phosphate glass matrix (Riley BJ, Peterson JA, Vienna JD, Ebert WL, Frank SM, Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form[J], Journal of Nuclear Materials, 2020, 529:151949). However, phosphate glass is extremely corrosive to furnace refractory materials and metal electrodes, making it difficult to achieve industrial application.

[0006] The two halide waste treatment methods mentioned above each have their advantages and disadvantages. For example, while the glass-bonded sodalite solidification process can accommodate a higher concentration of halogens, it is complex. While the dehalogenation and resolidification process can initially remove the halogens from the waste, the residual substances present after dehalogenation pose new challenges for the subsequent development of the solidified product. Furthermore, the current high dehalogenation temperature (greater than 600°C) poses a risk of high-temperature migration of radioactive volatile nuclides in the waste. Therefore, it is necessary to develop a reliable and simple method for the safe treatment of radioactive metal halide waste. Summary of the Invention

[0007] In view of this, the present invention provides a method for dehalogenation and vitrification of radioactive metal halide waste, which is used to solve the problems in the prior art of radioactive metal halide waste treatment, such as complex process, high dehalogenation temperature, and poor compatibility of the new substances remaining after dehalogenation with the curing substrate.

[0008] A first aspect of the present invention provides a method for dehalogenating radioactive metal halide waste, comprising the following steps:

[0009] The radioactive metal halide waste is mixed with oxalic acid and subjected to heat treatment to remove the halogen in the radioactive metal halide waste.

[0010] A second aspect of the present invention provides a glass solidification method, comprising the following steps:

[0011] The dehalogenated waste treated by the dehalogenation method of radioactive metal halide waste provided by the first aspect of the present invention is solidified by a glass solidification substrate to form a glass solidified body.

[0012] The third aspect of the present invention provides a glass solidified body, which is obtained by the glass solidification method provided by the second aspect of the present invention.

[0013] Compared with the prior art, the present invention has the following beneficial effects:

[0014] (1) The dehalogenation method provided by the present invention has low dehalogenation temperature and high efficiency, is not only energy-saving and easy to control, but also reduces the migration risk of volatile nuclides;

[0015] (2) No new substances remain after dehalogenation treatment. Dehalogenated waste can be solidified using a glass solidification substrate. The waste bag has a large capacity, and the chemical stability of the solidified body meets the requirements for the disposal of high-level glass solidified bodies.

[0016] (3) The present invention proposes to first use oxalic acid for dehalogenation and then adopt a glass solidification method, which makes it possible to treat radioactive metal halide waste using the existing mature glass solidification process. The process is simple and practical, and has good application prospects and social value. BRIEF DESCRIPTION OF THE DRAWINGS

[0017] Figure 1 This is a process flow chart of one embodiment of the method for dehalogenation and vitrification of radioactive metal halide wastes of the present invention;

[0018] Figure 2 The results show the effect of the addition ratio of oxalic acid (H2C2O4) to chlorine atoms on the dechlorination efficiency in Example 1 of the present invention;

[0019] Figure 3 This is the result of the effect of heat treatment temperature on dechlorination efficiency in Example 1 of the present invention;

[0020] Figure 4 This is the result of the effect of holding time on dechlorination efficiency in Example 1 of the present invention;

[0021] Figure 5 is the XRD pattern of the glass solidified body prepared in Example 1 of the present invention;

[0022] Figure 6 3 is the XRD pattern of the glass solidified body prepared in Example 2 of the present invention. DETAILED DESCRIPTION

[0023] In order to make the purpose, technical solutions and advantages of the present invention more clearly understood, the present invention will be further described in detail below with reference to the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain the present invention and are not intended to limit the present invention.

[0024] See also Figure 1The first aspect of the present invention provides a method for dehalogenating radioactive metal halide waste, comprising the following steps: mixing the radioactive metal halide waste with oxalic acid, and removing the halogen in the radioactive metal halide waste by heat treatment.

[0025] In the present invention, the heat treatment temperature is 100-600°C. Further, 250-500°C. Within this temperature range, the dehalogenation efficiency can reach over 90%. Furthermore, within the temperature range of 280-400°C, the lower heat treatment temperature can achieve extremely high dehalogenation efficiency and reduce the risk of migration of volatile nuclides.

[0026] Furthermore, the heat treatment time is 20 to 1000 min, further 60 to 600 min, and further 90 to 300 min.

[0027] In some embodiments of the present invention, the mixture of radioactive metal halide waste and oxalic acid is placed in an environment preheated to a heat treatment temperature for heat treatment. In this process, the heat treatment time is preferably 30 to 500 minutes, more preferably 60 to 300 minutes.

[0028] In some embodiments of the present invention, the mixture of radioactive metal halide waste and oxalic acid is heated in a furnace at a rate of 1 to 20°C / min to a heat treatment temperature. In this process, the heat treatment time includes both a heating time and a holding time. Furthermore, the mixture is heated in a furnace at a rate of 1 to 10°C / min to a heat treatment temperature and then held at that temperature for 0 to 180 minutes.

[0029] In some more specific embodiments of the present invention, the heat treatment process is: heating to 300° C. in a furnace at a rate of 5° C. / min and keeping the temperature for 0 to 120 min.

[0030] In the present invention, the radioactive metal halide waste and oxalic acid are both in solid powder form, which is more conducive to uniform mixing of the two and improves the dehalogenation effect. Furthermore, the average particle size of the oxalic acid and the radioactive metal halide waste is less than 100 mesh.

[0031] In some specific embodiments of the present invention, before the heat treatment, the radioactive metal halide waste is mixed with oxalic acid solid and crushed to an average particle size of less than 100 mesh.

[0032] In the present invention, during the dehalogenation process, the molar ratio of oxalic acid to halogen is greater than 0.5. Furthermore, the molar ratio of oxalic acid to halogen is greater than 0.8. Furthermore, the molar ratio of oxalic acid to halogen is greater than 1. Furthermore, the molar ratio of oxalic acid to halogen is 1.2 to 3. Within this molar ratio range, the dehalogenation efficiency can reach over 90%. Furthermore, the molar ratio of oxalic acid to halogen is 1.5 to 2.5. Within this molar ratio range, the dehalogenation efficiency can reach over 90%, and the amount of oxalic acid used is reduced.

[0033] In some preferred embodiments of the present invention, the molar ratio of oxalic acid to halogen is 2.

[0034] In the present invention, radioactive metal halide waste includes at least one of chloride molten salt waste and fluoride molten salt waste generated during the dry reprocessing of spent fuel. Furthermore, the chloride molten salt waste includes at least one of alkali metal chlorides, alkaline earth metal chlorides, and rare earth metal chlorides; and the fluoride molten salt waste includes at least one of alkali metal fluorides, alkaline earth metal fluorides, and rare earth metal fluorides. Furthermore, the chloride molten salt waste includes LiCl, KCl, NaCl, CsCl, SrCl2, and rare earth metal chlorides; and the fluoride molten salt waste includes LiF, NaF, KF, CsF, MgF2, SrF2, and rare earth metal fluorides.

[0035] A second aspect of the present invention provides a glass solidification method, comprising the following steps:

[0036] The dehalogenated waste treated by the dehalogenation method of radioactive metal halide waste provided by the first aspect of the present invention is solidified by a glass solidification substrate to form a glass solidified body.

[0037] In the present invention, the glass-solidified substrate is borosilicate glass. Further, the chemical composition of the glass-solidified substrate includes: 63-70 wt% SiO2, 17-22 wt% B2O3, 6-8 wt% Al2O3, and 5-10 wt% CaO.

[0038] In the present invention, the glass-solidified matrix accounts for 15 to 35% of the radioactive waste packaging capacity, further 20 to 35%, and even further 25 to 35%, calculated as a percentage by mass of oxides.

[0039] In the present invention, the step of solidifying the radioactive waste treated by the dehalogenation method for radioactive metal halide waste provided by the first aspect of the present invention to form a glass solidified body by using a glass solidified substrate is specifically: after mixing the dehalogenated waste with the glass solidified substrate, heating and melting, and cooling to form a glass solidified body.

[0040] In the present invention, the heating and melting temperature is 1000-1400° C., further 1100-1200° C.; and the melting time is 1-6 hours, further 1-3 hours.

[0041] In some specific embodiments of the present invention, the heating and melting temperature is 1200° C. and the time is 1 to 2 hours.

[0042] The third aspect of the present invention provides a glass solidified body, which is obtained by the glass solidification method provided by the second aspect of the present invention.

[0043] Example 1

[0044] In this example, non-radioactive chlorides were used to simulate the electrolytic refining waste salt produced by the dry reprocessing of spent fuel from molten salt electrolysis. The composition is listed in Table 1. A total of 20 g of oxalic acid and chloride molten salt waste was weighed in varying proportions, thoroughly mixed, and placed in a 100 ml corundum crucible. The crucible was heated to 100-600°C at a furnace rate of 5°C / min, held at this temperature for 0-120 min, and then removed from the crucible and naturally cooled to room temperature to obtain dechlorinated waste. The dechlorination efficiency (CRE) was calculated according to the following formula, where M1 and M2 are the mass of Cl in the original waste and the dechlorinated waste, respectively.

[0045]

[0046] Table 1 Chloride molten salt waste composition (wt%)

[0047]

[0048] The effects of different oxalic acid to chlorine molar ratios, heat treatment temperatures and holding times on dechlorination efficiency are shown in Figure 2. Figures 2-4 .in, Figure 2 The heat treatment temperature is 300℃ and the temperature is kept for 60min; Figure 3 The molar ratio of oxalic acid to chlorine was 2, and the temperature was kept for 0 min; Figure 4 The molar ratio of oxalic acid to chlorine is 2, and the heat treatment temperature is 300℃. Figures 2-4 The optimal dechlorination conditions are: a molar ratio of oxalic acid to chlorine of 2, heating the furnace to 300°C and holding for 60 minutes, achieving a dechlorination efficiency of 99%. The dechlorinated waste obtained under the optimal dechlorination conditions was vitrified using the vitrification formulation described in Table 2 (waste bag volume of 35 wt%). 20 g of the mixture was weighed, thoroughly mixed, and placed in a 50 ml corundum crucible. After holding in a muffle furnace at 1200°C for 1 hour, the molten glass was poured into a preheated copper mold and cooled to obtain a vitrified body.

[0049] Table 2 Example 1 Dechlorination waste and glass solidification substrate in the designed glass solidification formula (wt%)

[0050]

[0051] The XRD diffraction pattern of the solidified body obtained in this embodiment is ( Figure 5 ) is a typical amorphous diffraction peak, proving that the prepared sample is in a glassy state. The density measured by the Archimedean drainage method is 2.58 g / cm 3The chemical stability of the glass solidified body samples was evaluated using PCT-7. Table 3 shows the PCT-7 leaching results of the main elements in this example. The leaching values of each element are all lower than 2 g / m 2 , meeting the international requirements for the chemical stability of high-level glass solidification bodies.

[0052] Table 3 PCT-7 leaching results of main elements in glass solidified body of Example 1

[0053]

[0054]

[0055] Example 2

[0056] In this example, non-radioactive fluoride was used to simulate fluorine-containing waste generated by the dry reprocessing of spent fuel from a molten salt reactor. The composition is listed in Table 4. With a molar ratio of oxalic acid to fluorine of 2, 20 g of oxalic acid and fluoride molten salt waste were weighed, thoroughly mixed, and placed in a 100 ml corundum crucible. The crucible was then heated to 300°C at a rate of 5°C / min and held for 60 minutes. The crucible was then removed and naturally cooled to room temperature to obtain defluorinated waste. The defluorination efficiency (FRE) was calculated according to the following formula, where M3 and M4 are the mass of F in the original waste and the defluorinated waste, respectively.

[0057]

[0058] Table 4 Fluoride molten salt waste composition (wt%)

[0059]

[0060] The above defluorination method can achieve a 91% fluorine removal efficiency in fluoride molten salt waste. The defluorinated waste was vitrified using the vitrification formula (waste bag volume 25 wt%) designed in Table 5. 20 g of the mixture was weighed, thoroughly mixed, and placed in a 50 ml corundum crucible. The mixture was heated in a muffle furnace at 1200°C for 1 hour, then poured into a preheated copper plate mold and cooled to obtain a vitrified body.

[0061] Table 5 Example 2 Defluorination waste and glass curing substrate in the designed glass curing formula (wt%)

[0062]

[0063] The XRD diffraction pattern of the solidified body obtained in this embodiment is ( Figure 6 ) is a typical amorphous diffraction peak, proving that the prepared sample is in a glassy state. The density measured by the Archimedean drainage method is 2.48 g / cm 3The chemical stability of the glass solidified body samples was evaluated using PCT-7. Table 6 shows the PCT-7 leaching results of the main elements in this example. The leaching values of each element are all lower than 2 g / m 2 , meeting the international requirements for the chemical stability of high-level glass solidification bodies.

[0064] Table 6 PCT-7 leaching results of main elements in glass solidified body of Example 2

[0065]

[0066]

[0067] The above description is only a preferred specific embodiment of the present invention, but the scope of protection of the present invention is not limited thereto. Any changes or substitutions that can be easily thought of by any technician familiar with this technical field within the technical scope disclosed by the present invention should be covered by the scope of protection of the present invention.

Claims

1. A method for dehalogenating radioactive metal halide waste, characterized in that: The method comprises the following steps: mixing radioactive metal halide waste with oxalic acid, and removing halogen from the radioactive metal halide waste by heat treatment; The radioactive metal halide waste components include any one or more of alkali metal chloride, alkaline earth metal chloride, rare earth metal chloride, alkali metal fluoride, alkaline earth metal fluoride, rare earth metal fluoride, and CsI.

2. The method for dehalogenating radioactive metal halide waste according to claim 1, wherein: The heat treatment temperature is 100~600 o C.

3. The method for dehalogenating radioactive metal halide waste according to claim 1, wherein: The heat treatment temperature is 280~400 o C, the heat treatment time is 20 to 1000 minutes.

4. The method for dehalogenating radioactive metal halide waste according to claim 1, wherein: The radioactive metal halide waste and the oxalic acid are both in solid powder form.

5. The method for dehalogenating radioactive metal halide waste according to claim 1, wherein: During the dehalogenation process, the oxalic acid and the radioactive metal halide waste are mixed in a ratio of oxalic acid to halogen greater than 0.

5.

6. The method for dehalogenating radioactive metal halide waste according to claim 1, wherein: During the dehalogenation process, the oxalic acid and the radioactive metal halide waste are mixed in a molar ratio of oxalic acid to halogen of 1.2 to 3.

7. A glass solidification method, characterized in that: The following steps are involved: The dehalogenated waste treated by the dehalogenation method of radioactive metal halide waste according to any one of claims 1 to 6 is solidified by a glass solidification substrate to form a glass solidified body.

8. The glass solidification method according to claim 7, wherein: The glass-solidified substrate is borosilicate glass; and the glass-solidified substrate accounts for 15 to 35% of the radioactive waste packaging capacity, calculated as a percentage by mass of oxides.

9. A glass solidified body, characterized in that: The glass solidification body is obtained by the glass solidification method according to claim 7.

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