Method for managing high-level radioactive waste liquid containing americium and curium

Abstract

The invention relates to a method for managing high-level radioactive waste liquid containing americium and curium, which comprises the following steps of: preparing waste liquid containing Am and Cm into a glass solidified body, and if the glass solidified body has no recycling and reusing requirements, keeping the glass solidified body in a stainless steel container for temporary storage; if the glass solidified body has the requirement for recycling and reusing, a mixed solution of dilute nitric acid and hydrogen peroxide is adopted as a leaching agent to leach key nuclides in the glass solidified body; the feasible dual-purpose management method for recovery, reuse and final disposal of the high-level radioactive waste liquid containing americium and curium can be obtained under the current situation that the market demand of isotopes of Am and Cm is small and the production technology is not mature at present.

Description

Technical Field

[0001] The invention belongs to the technical field of spent fuel post-processing, and in particular relates to a method for managing high-level radioactive liquid waste containing americium and curium. Background Art

[0002] Among the high-level waste liquids produced by spent fuel reprocessing, Am and Cm are the most toxic minor actinides that deserve the most attention. If they can be made into isotope products, or made into targets for transmutation, or enter the fast reactor fuel cycle system, their long-term hazards can be reduced or even eliminated. Unfortunately, neither the transmutation technology nor the fast reactor fuel cycle technology is mature at present, the market demand for Am and Cm isotopes is very small, and the production technology is also immature. Therefore, it is possible to consider temporarily storing the waste liquid or products containing Am and Cm for a period of time, and then retrieving and reusing them after the relevant technology matures.

[0003] If the acidic waste liquid containing Am and Cm is temporarily stored in a solution state, such as in a high-magnification tank, there is a risk of leakage and strict supervision is required; if Am and Cm are prepared as oxide products for temporary storage, special technical means are needed to solve the problems of product heat accumulation and increased pressure inside the temporary storage container. In the future, a large amount of solid waste (inner tank and outer tank of the temporary storage container) will be generated, and the compression, filling, capping and decontamination operations are complicated and the operating cost is very high. Summary of the invention

[0004] The purpose of the present invention is to provide a method for managing high-level radioactive liquid waste containing americium and curium, which comprises the following steps: making waste liquid containing Am and Cm into a glass solid body, and if there is no need for recovery and reuse, retaining the glass solid body in a stainless steel container for temporary storage; if there is a need for recovery and reuse, using a leaching agent to leach key nuclides in the glass solid body; and obtaining a feasible dual-purpose management method for recovery and reuse and final disposal of high-level radioactive liquid waste containing americium and curium under the current situation that the market demand for isotopes of Am and Cm is small and the production technology is very immature.

[0005] To achieve the above objectives, the technical solution adopted by the present invention is: a method for managing high-level radioactive liquid waste containing americium and curium, the method comprising the following steps:

[0006] S1. Making a glass solid from the waste liquid containing Am and Cm: Mix and melt the waste liquid containing Am and Cm and a glass substrate in a glass melting furnace, pour the obtained mixed melt into a stainless steel container, and obtain a glass solid after cooling;

[0007] If the glass solidified body has no need for retrieval and reuse, the process proceeds to step S2; if the glass solidified body has a need for retrieval and reuse, the process proceeds to step S3;

[0008] S2. If there is no need to retrieve and reuse the glass solidified body, the glass solidified body is kept in the stainless steel container for temporary storage;

[0009] S3. If the glass solidified body needs to be retrieved and reused, a mixed solution of dilute nitric acid and hydrogen peroxide is used as an extractant to extract the key nuclides in the glass solidified body.

[0010] Furthermore, in step S1, the glass melting furnace is a ceramic electric melting furnace or a cold crucible furnace.

[0011] Furthermore, in step S1, the glass substrate is borosilicate glass or glass-ceramics.

[0012] Furthermore, the crystal phase of the microcrystalline glass is pyrochlore, zircon (Ca, Ln) (Zr, Ln) (Ti, Al) 2 O 7 or peridotite (Ca, Ln) 10 (SiO4)6O2.

[0013] Furthermore, the waste inclusion rate in the glass solidified body is 20% to 40%.

[0014] Furthermore, in step S3, for the glass solid body that needs to be retrieved and reused, the glass solid body is disintegrated and broken into glass solid particles of 100-300 μm by one or more methods of cutting, disintegration, and grinding, and a mixed solution of dilute nitric acid and hydrogen peroxide is used as an extractant to extract the key nuclides in the glass solid body.

[0015] Furthermore, in step S3, the concentration of nitric acid in the leaching agent is 0.5-2 mol / L; and the amount of hydrogen peroxide added is 0.5-2 ml.

[0016] Furthermore, in step S3, the ratio of the leaching agent to the glass solid particles is 10-30 ml of the leaching agent per 1 g of the glass solid particles.

[0017] Furthermore, in step S3, the process of leaching the key nuclides in the glass solid body is carried out under the condition of heating to 80-100°C.

[0018] Furthermore, in step S3, the process of leaching the key nuclides in the glass solidified body is carried out under ultrasonic conditions.

[0019] The beneficial effects of the present invention are: using a high-level waste liquid management method containing americium and curium provided by the present invention, through the following steps: making the waste liquid containing Am and Cm into a glass solid body, if the glass solid body has no need for retrieval and reuse, then the glass solid body is retained in a stainless steel container for temporary storage; if the glass solid body has a need for retrieval and reuse, a mixed solution of dilute nitric acid and hydrogen peroxide is used as an extractant to extract the key nuclides in the glass solid body; under the current situation where the market demand for Am and Cm isotopes is small and the production technology is very immature, a feasible management method for the retrieval and reuse and final disposal of high-level waste liquid containing americium and curium can be obtained. The specific beneficial effects of the present invention are mainly reflected in:

[0020] (1) Using mature ceramic electric melting furnace or cold crucible technology, the waste liquid containing Am and Cm is made into a glass solid body, which not only has a high waste inclusion rate (20% to 40%), but also has chemical and mechanical stability that can meet the performance requirements of the glass solid body and meet the requirements of the final waste disposal for the waste source;

[0021] (2) From the perspective of temporary product storage, this dual-use management technology does not require the development of dedicated temporary product storage containers and packaging systems. Since the thermal conductivity of the glass matrix is ​​much better than that of the oxide form, and the glass matrix is ​​in close contact with the stainless steel product container, the decay heat of the waste metal in the glass matrix can be quickly discharged without heat accumulation and the risk of excessive pressure in the product container; moreover, due to the effective envelope of the glass matrix, the risk of diffusion of alpha nuclides during the temporary storage of glass solid products is very small. For glass solids that have no need for retrieval in the future, they can be directly disposed of geologically, eliminating the preparation operation from "intermediate form in the temporary storage stage" to "final waste form suitable for disposal", which not only saves costs but also reduces the generation of secondary waste;

[0022] (3) For glass solid products that need to be retrieved, after cutting, disassembling, grinding and other processes, a mixed solution of dilute nitric acid and a small amount of hydrogen peroxide is used as an extractant. Under the condition of ultrasonic wave, it only takes a short time to selectively extract Am, Cm and rare earth elements in the glass scraps, and most of the precious metals (Ru, Rh and Pd) and Zr are retained in the glass matrix. Compared with the traditional high-temperature alkali melting method, hydrofluoric acid or concentrated nitric acid corrosion method, the above method can selectively extract certain metal elements in the glass matrix under very mild reaction conditions, which can effectively reduce the long-term risks of equipment, operators, and workplaces. In addition, using dilute nitric acid + a small amount of hydrogen peroxide as an extractant, the nitric acid solution formed can easily connect to the subsequent nuclide purification process. Moreover, since Zr and precious metals are basically retained in the glass matrix, the risk of interface contamination during the subsequent liquid-liquid extraction operation to extract nuclides can also be reduced. BRIEF DESCRIPTION OF THE DRAWINGS

[0023] Figure 1 It is a schematic flow chart of a method for managing high-level radioactive liquid waste containing americium and curium according to an embodiment of the present invention. DETAILED DESCRIPTION

[0024] The technical solutions in the embodiments of the present invention will be further clearly and completely described below in conjunction with the drawings and embodiments. Obviously, the described embodiments are only part of the embodiments of the present invention, rather than all of the embodiments. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without making creative work are within the scope of protection of the present invention.

[0025] The inventors discovered in preliminary process research and calculations that since the thermal power density of the mixed oxide products of Am and Cm will remain at a relatively high level for a long period of time, if Am and Cm are prepared into oxide products for temporary storage, the above oxide products will also face multiple challenges during the temporary storage stage, such as heat extraction, tank pressure control, and effective sealing of α nuclides. There are risks to the safety of long-term temporary storage, which is not only costly, but also requires re-preparation into final waste packages for disposal in the future.

[0026] In the embodiment of the present invention, the inventors have found through research that the thermal conductivity of the glass solidified product prepared from the waste liquid containing americium (Am) and curium (Cm) is 0.9W / m·℃, which is 20 times that of the uncompacted mixed oxide of Am and Cm, and 4 times that of the mixed oxide of Am and Cm in the compacted (20kPa) state. In other words, the glass solidified product as a temporary storage form of americium (Am) and curium (Cm) has better heat dissipation performance than the mixed oxide product.

[0027] At the same time, the inventors found through research that after the separation of high-level radioactive waste, lanthanide elements are the absolute main body in the waste stream containing americium (Am) and curium (Cm), accounting for about 95% of the total metal mass in the waste. Therefore, the key to the glass solidification of the waste stream containing americium (Am) and curium (Cm) is to effectively contain high-content lanthanide elements. Although borosilicate glass formula is the mainstream for solidifying high-level radioactive waste, from the perspective of long-term safety of geological disposal of radioactive waste, it is not the best choice to prepare the waste stream of Am and Cm into an amorphous borosilicate glass solid body. Glass-Ceramics / glass-crystalline is a more ideal waste substrate, and it can be implemented in a short period of time:

[0028] (1) Ready-made glass melting furnaces (such as cold crucibles) can be used, and the equipment is highly mature;

[0029] (2) Designing a suitable glass-ceramic formula, the inclusion rate is usually very high, which can reach 30% to 40% or even higher;

[0030] (3) Controlling the cooling process of the molten glass to meet the requirements of crystal nucleation and growth can form a microcrystalline structure in which the crystal phase (containing Am, Cm and rare earth) is uniformly dispersed in the glass phase (containing other fission products);

[0031] (4) The crystallization products of the silicate system are zircon (Ca, Ln) (Zr, Ln) (Ti, Al) 2 O 7 or peridotite (Ca, Ln) 10 (SiO4)6O2 is a representative example, which is very suitable for waste sources with high lanthanide content, such as waste streams containing americium (Am) and curium (Cm). Using these crystalline phases as the substrate for actinide solidification, its radiation and corrosion resistance is much better than that of borosilicate glass substrates. Moreover, some natural minerals with the same / similar structure (containing rare earth and radioactive uranium and thorium) have existed for millions of years, which shows that microcrystalline glass can effectively contain high content of lanthanides on ultra-long time scales.

[0032] like Figure 1 As shown, this embodiment provides a method for managing high-level radioactive liquid waste containing americium and curium, the method comprising the following steps:

[0033] S1. Making a glass solidified body from the waste liquid containing Am and Cm: Mix and melt the waste liquid containing Am and Cm and a glass substrate in a glass melting furnace, pour the obtained mixed melt into a stainless steel container, and obtain a glass solidified body after cooling; if the glass solidified body has no need for recovery and reuse, proceed to step S2; if the glass solidified body has a need for recovery and reuse, proceed to step S3;

[0034] Optionally, in step S1, the glass melting furnace is a ceramic electric melting furnace or a cold crucible furnace, and the waste liquid containing Am and Cm and the glass substrate are made into a glass solidified body by using mature ceramic electric melting furnace or cold crucible technology.

[0035] Specifically, the waste inclusion rate in the glass solid body is high, which is 20% to 40%; the chemical stability and mechanical stability of the glass solid body can meet the requirements of the final waste disposal for the waste source item.

[0036] Optionally, the glass substrate is borosilicate glass or glass-ceramics.

[0037] In a specific embodiment, the glass melting furnace is a cold crucible furnace, and the glass substrate is made of microcrystalline glass; by controlling the cooling process of the molten glass to meet the requirements of crystal nucleation and growth, a microcrystalline structure in which the crystal phase (containing Am, Cm and rare earth) is uniformly dispersed in the glass phase (containing other fission products) is formed in the glass solid body.

[0038] Optionally, the crystal phase of the microcrystalline glass is pyrochlore, zircon (Ca, Ln) (Zr, Ln) (Ti, Al) 2 O 7 or peridotite (Ca, Ln) 10 (SiO4)6O2. The above-mentioned microcrystalline glass is used as the substrate for solidification of actinides, and its radiation resistance and corrosion resistance are much better than those of borosilicate glass substrates. Moreover, some natural minerals with the same / similar structure (containing rare earth and radioactive uranium and thorium) have existed for millions of years, indicating that the above-mentioned microcrystalline glass can effectively contain high content of lanthanide elements on an ultra-long time scale.

[0039] S2. If there is no need to retrieve and reuse the glass solidified body, the glass solidified body is kept in the stainless steel container for temporary storage;

[0040] Specifically, in step S2, from the perspective of temporary storage of the glass solidified product, the method provided in this embodiment does not require the development of a dedicated temporary storage product container and packaging system. Since the thermal conductivity of the glass matrix is ​​much better than that of the oxide form, and the glass matrix is ​​in close contact with the stainless steel product container, the decay heat of the waste metal in the glass matrix can be quickly extracted, and there is no risk of heat accumulation or excessive pressure in the product container; moreover, due to the effective envelope of the glass matrix, the risk of diffusion of alpha nuclides during the temporary storage of the glass solidified product is very small.

[0041] In a specific embodiment, microcrystalline glass peridotite (Ca, Ln) is used. 10 (SiO4)6O2 is used as the glass substrate, and the waste liquid containing americium (Am) and curium (Cm) is prepared into a glass solid product under the above conditions. Its thermal conductivity is 0.9W / m·℃, which is 20 times that of the uncompacted Am and Cm mixed oxide, and 4 times that of the Am and Cm mixed oxide in the compacted (20kPa) state. That is, the glass solid product as a temporary storage form of americium (Am) and curium (Cm) has better heat dissipation performance than the mixed oxide product.

[0042] Optionally, in step S2, the glass solidified body that is temporarily stored without the need for retrieval can be directly disposed of geologically, eliminating the preparation operation from the "intermediate form in the temporary storage stage" to the "final waste form suitable for disposal", which not only saves costs but also reduces the generation of secondary waste.

[0043] S3. If the glass solidified body needs to be retrieved and reused, a mixed solution of dilute nitric acid and a small amount of hydrogen peroxide is used as an extractant to extract the key nuclides Am and Cm in the glass solidified body.

[0044] Specifically, for glass solid products that need to be retrieved and reused, the glass solid is disassembled and broken into glass solid particles of 100-300 μm in size through cutting, disassembly, grinding and other processes, and a mixed solution of dilute nitric acid and a small amount of hydrogen peroxide is used as an extractant to extract key nuclides in the glass solid.

[0045] Optionally, in the leaching agent, the concentration of nitric acid is 0.5-2 mol / L, and the amount of hydrogen peroxide added is 0.5-2 ml (hydrogen peroxide is a commercial product, in which the mass percentage of H2O2 is 30%).

[0046] In a specific embodiment, the concentration of nitric acid in the leaching agent is 1 mol / L, and the amount of hydrogen peroxide added is 1 ml.

[0047] Optionally, the ratio of the leaching agent to the glass solid particles is 10 to 30 ml of the leaching agent per 1 g of the glass solid particles.

[0048] Optionally, the process of leaching the key nuclides in the glass solid body is carried out under the condition of heating to 80-100°C.

[0049] Optionally, the process of leaching the key nuclides in the glass solid body is carried out under ultrasonic conditions.

[0050] In a specific embodiment, at 90°C, the key nuclides in the glass solid body are leached for 5 hours at a ratio of 20 ml of leaching agent / 1 g of glass solid particles, and almost all of Am, Cm and rare earth metals (99.5%) loaded in the glass solid body, as well as Na, Ca, Al, etc. in the glass solid body can be leached out; while zirconium (Zr) and platinum group metals (Pd, Ru, Rh) that easily cause interface contamination in the subsequent liquid-liquid extraction process are almost completely retained in the glass solid body and will not enter the leaching agent (hydrogen peroxide / nitric acid solution).

[0051] In another specific embodiment, at 90°C, when the key nuclides in the glass solid body are leached with a ratio of 20 ml of leaching agent / 1 g of glass solid particles, ultrasonic waves (frequency of 20 KHz) are added, and the time for completing the leaching of Am, Cm and rare earth metals (99.5%) can be shortened to 3 hours (most of the nuclides are basically leached out within 3 hours).

[0052] Existing methods for recovering metal elements from glass substrates often require very harsh chemical conditions, such as the use of highly corrosive hydrofluoric acid or nearly saturated concentrated nitric acid (16 mol / L) to react for 4 hours at high temperature (115°C) (the Oak Ridge Laboratory in the United States uses this process), which places high demands on the corrosion resistance of the reactor and the protection of the operator. In this embodiment, in the process of leaching the key nuclides in the glass solid body, under the condition of microwave, it only takes a short time to selectively leach Am, Cm and rare earth elements in the glass scraps, and most of the precious metals (Ru, Rh and Pd) and Zr are retained in the glass substrate. Compared with the traditional high-temperature alkali melting method, hydrofluoric acid or concentrated nitric acid corrosion method, the above method can selectively leach certain metal elements in the glass substrate under very mild reaction conditions, which can effectively reduce the long-term risks of equipment, operators, and workplaces. In addition, using dilute nitric acid + a small amount of hydrogen peroxide as a leaching agent, the nitric acid solution formed after the leaching process is easily connected to the subsequent nuclide purification process. Moreover, since Zr and precious metals are basically retained in the glass matrix, the risk of interface contamination during the subsequent liquid-liquid extraction operation to extract nuclides can also be reduced.

[0053] The specific implementation modes of the present invention are further described below through examples.

[0054] Example 1: Making a glass solid from a stripping solution containing Am and Cm, and extracting key nuclides from the glass solid

[0055] S11, preparing a glass solidified body from the stripping waste liquid containing Am and Cm;

[0056] Specifically, microcrystalline glass peridotite (Ca, Ln) is used. 10 (SiO4)6O2 is used as a glass matrix, and the stripping waste liquid containing Am and Cm and the glass matrix are mixed and melted in a cold crucible furnace. The obtained mixed melt is poured into a stainless steel container, and the cooling process of the molten glass is controlled to meet the requirements of crystal nucleation and growth, thereby forming a microcrystalline structure in which the crystal phase (containing Am, Cm and rare earth) is uniformly dispersed in the glass phase (containing other fission products) in the prepared glass solid body.

[0057] S12, leaching key nuclides in the glass solidified body;

[0058] Specifically, under the conditions of 90° C. and ultrasonic wave (frequency of 20 KHz), the key nuclides in the glass solid body are leached for 3 hours at a ratio of 20 ml of leaching agent / 1 g of glass solid particles, and almost all of Am, Cm and rare earth metals (99.5%) loaded in the glass solid body, as well as Na, Ca, Al, etc. in the glass solid body can be leached out; while zirconium (Zr) and platinum group metals (Pd, Ru, Rh) that easily cause interface contamination in the subsequent liquid-liquid extraction process are almost completely retained in the glass solid body and will not enter the leaching agent (hydrogen peroxide / nitric acid solution).

[0059] The method described in the present invention is not limited to the specific embodiments described. The above embodiments are only examples of the present invention. The present invention can also be implemented in other specific ways or other specific forms without departing from the gist or essential features of the present invention. Therefore, the described embodiments should be regarded as illustrative rather than restrictive in any respect. The scope of the present invention should be described by the attached claims, and any changes equivalent to the intent and scope of the claims should also be included in the scope of the present invention.

Claims

1. A method for managing high-level radioactive liquid waste containing americium and curium, characterized in that: The method comprises the following steps: S1. Making a glass solid from the waste liquid containing Am and Cm: Mix and melt the waste liquid containing Am and Cm and a glass substrate in a glass melting furnace, pour the obtained mixed melt into a stainless steel container, and obtain a glass solid after cooling; If the glass solidified body has no need for retrieval and reuse, the process proceeds to step S2; if the glass solidified body has a need for retrieval and reuse, the process proceeds to step S3; S2. If there is no need to retrieve and reuse the glass solidified body, the glass solidified body is kept in the stainless steel container for temporary storage; S3. If the glass solidified body needs to be retrieved and reused, a mixed solution of dilute nitric acid and hydrogen peroxide is used as an extractant to extract the key nuclides in the glass solidified body.

2. A method for managing high-level liquid waste containing americium and curium according to claim 1, characterized in that: In step S1, the glass melting furnace is a ceramic electric melting furnace or a cold crucible furnace.

3. The method for managing high-level liquid waste containing americium and curium according to claim 1, characterized in that: In step S1, the glass substrate is borosilicate glass or glass-ceramics.

4. A method for managing high-level radioactive liquid waste containing americium and curium according to claim 3, characterized in that: The crystal phase of the microcrystalline glass is pyrochlore, zircon (Ca, Ln) (Zr, Ln) (Ti, Al) 2 O 7 or peridotite (Ca, Ln) 10 (SiO4)6O2.

5. The method for managing high-level liquid waste containing americium and curium according to claim 1, characterized in that: The waste inclusion rate in the glass solidified body is 20% to 40%.

6. A method for managing high-level radioactive liquid waste containing americium and curium according to claim 1, characterized in that: In step S3, for the glass solid body that needs to be retrieved and reused, the glass solid body is disintegrated and broken into glass solid particles of 100-300 μm by one or more methods of cutting, disintegration, and grinding, and a mixed solution of dilute nitric acid and hydrogen peroxide is used as an extractant to extract the key nuclides in the glass solid body.

7. The method for managing high-level radioactive liquid waste containing americium and curium according to claim 1, characterized in that: In step S3, the concentration of nitric acid in the leaching agent is 0.5-2 mol / L; the amount of hydrogen peroxide added is 0.5-2 ml.

8. A method for managing high-level radioactive liquid waste containing americium and curium according to claim 6, characterized in that: In step S3, the ratio of the leaching agent to the glass solid particles is 10-30 ml of the leaching agent per 1 g of the glass solid particles.

9. The method for managing high-level radioactive liquid waste containing americium and curium according to claim 1, characterized in that: In step S3, the process of leaching the key nuclides in the glass solid body is carried out under the condition of heating to 80-100°C.

10. The method for managing high-level radioactive liquid waste containing americium and curium according to claim 1, characterized in that: In step S3, the process of leaching the key nuclides in the glass solidified body is carried out under ultrasonic conditions.

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