A method for solidifying radioactive waste

Abstract

The present invention discloses a method for solidifying radioactive waste, which relates to the field of treating radioactive pollution materials. The method includes: distilling the DMSO washing water waste liquid containing HMX to recover DMSO and obtain a distillation residue; sedimenting and drying the distillation residue to obtain dry uranium powder; mixing a ceramic base material, a lead compound and the dry uranium powder in a ball milling device to obtain a mixed powder, placing the mixed powder in a mold, applying a pressure of 50-200 MPa to press a blank and obtain a green body; pre-sintering the green body at a pre-sintering temperature in an inert atmosphere to obtain a pre-sintered green body; sintering the pre-sintered green body at a temperature of 900-1250 °C in an inert atmosphere, then holding for 2-4 hours, and cooling to room temperature to obtain a solidified uranium-containing ceramic. By adding lead to the sintering raw materials, the shielding effect of ionizing radiation can be improved, the sintering temperature can be reduced, and the chemical dissolution degree can be reduced, which not only ensures environmental safety but also maintains the long-term shielding ability of the material.

Description

Technical Field

[0001] The present invention relates to the field of radioactive contaminated material treatment, and in particular to a radioactive waste solidification method. Background Art

[0002] During the later stages of nuclear weapon charging or assembly, shaped explosives (including octogen) are often assembled with nuclear materials in the same "assembly chamber" or undergo the same production line (e.g., hot pressing, machining, cleaning, etc.). Complete physical isolation of nuclear materials is difficult. Any tiny particle of dust, debris, vapor, or surface deposits may carry radioactive nuclides. Once in contact with octogen-containing components or cleaning solutions, they can introduce radioactivity into the cleaning solution. A few modified explosives, special munitions, or experimental formulations may incorporate compounds containing radioactive isotopes (e.g., trace amounts of radioactive isotopes used for detection, tracing, or initiating special reactions). This practice also generates radioactive waste during subsequent cleaning, recovery, and recrystallization processes.

[0003] To produce octogen, dimethyl sulfoxide (DMSO) is needed to dissolve it. Therefore, radioactive substances, such as residual uranium, are present in the DMSO washing wastewater containing octogen.

[0004] To prevent radioactive materials from escaping and contaminating the environment, existing technologies require solidification of the radioactive materials. This is typically done by cementing them and then dumping them into the sea or burying them underground. However, conventional cement has high porosity and is easily corroded by the seabed or underground environment, causing the radioactive materials to escape into the environment. Summary of the Invention

[0005] To solve the above technical problems, the present invention provides a radioactive waste solidification method, comprising:

[0006] S1, distilling the DMSO washing water waste liquid containing octogen to recover DMSO to obtain a distillation residue;

[0007] S2, settling and drying the distillation residue to obtain dry uranium powder;

[0008] S3, mixing a ceramic base material, a lead compound, and the dry uranium powder in a ball mill to obtain a mixed powder, placing the mixed powder in a mold, and applying a pressure of 50 to 200 MPa to form a green compact to obtain an embryonic body;

[0009] S4. Pre-sintering the green body at a pre-sintering temperature in an inert atmosphere to obtain a pre-sintered green body;

[0010] S5. In an inert atmosphere, sinter the pre-sintered green body at a temperature of 900 - 1250 °C, then keep it warm for 2 - 4 hours. After cooling to room temperature, a solidified uranium-containing ceramic is obtained.

[0011] In some embodiments, in S3, by mass, the ceramic base material includes: 35 - 45 parts of alumina, 25 - 35 parts of silica, 5 - 15 parts of zirconia, 3 - 7 parts of calcium oxide, 3 - 7 parts of iron oxide, and 5 - 15 parts of binder.

[0012] In some embodiments, the binder includes one or more of clay, talcum powder, slag, or fly ash.

[0013] In some embodiments, in S3, by mass, the ratio of the ceramic base material, the lead compound, and the dried uranium powder is (60 - 75):(10 - 25):(5 - 15).

[0014] In some embodiments, the lead compound includes one or more of PbO, PbTiO3, and PbO - SiO2.

[0015] In some embodiments, in S4, the pre-sintering temperature is 600 - 800 °C, and the pre-sintering time is 12 hours.

[0016] In some embodiments, the initial temperature of the sintering is 900 - 950 °C, the heating rate is 3 - 5 °C / min, the holding temperature is 1200 - 1250 °C, and the cooling rate is 2 - 5 °C / min.

[0017] In some embodiments, S1 includes:

[0018] After diluting the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, add a mild oxidant step by step at 30 - 60 °C and stir at a speed of 200 rpm;

[0019] Perform vacuum distillation at a temperature of 40 - 60 °C and -0.08 MPa to -0.09 MPa to recover DMSO and obtain a distillation residue.

[0020] In some embodiments, the mild oxidant includes one or more of ozone, persulfate, and hypochlorite.

[0021] In some embodiments, S2 includes:

[0022] Adjust the pH of the distillation residue to 6 - 7, add phosphate in batches at a temperature of 55 - 65 °C and stir at a speed of 300 rpm. After reacting for 2 hours, centrifuge at a speed of 4000 rpm to obtain a uranium-enriched sediment;

[0023] The uranium-enriched sediment is vacuum-dried for 8 to 24 hours, and the drying temperature is controlled at 80 to 120 °C to obtain dried uranium powder.

[0024] Compared with the prior art, the beneficial effects of the present invention are as follows:

[0025] 1. By drying the uranium solution in the early stage and removing volatile components such as organic solvents and HMX, the gas released by the decomposition of organic matter during sintering is significantly reduced, the porosity of the uranium-containing ceramics is decreased, and thus the migration rates of uranium and lead are reduced.

[0026] 2. By adding lead to the sintering raw materials, when ionizing radiation passes through the solidified body, scattering and energy loss continuously occur between the uranium phase and the lead phase, thereby significantly improving the shielding effect.

[0027] 3. By adding lead to the sintering raw materials, the sintering temperature can be significantly reduced. The path of ionizing radiation is blocked by a large number of grains and high-Z phases, and the attenuation is more thorough.

[0028] 4. Lead can form a stable lead-based composite phase with the ceramic matrix, and its chemical dissolution rate is significantly reduced, which not only ensures environmental safety but also maintains the long-term shielding ability of the material.

[0029] 5. By carefully designing and optimizing the formula of the ceramic matrix and cooperating with a suitable sintering process (including temperature, atmosphere, holding time, heating and cooling rates), a ceramic solidified body with low porosity, high strength and a high fixation effect on uranium can be obtained, significantly reducing the migration risk of radioactive substances in the disposal environment. Description of the Drawings

[0030] In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the following will briefly introduce the drawings required for use in the embodiments. It should be understood that the following drawings only show some embodiments of the present invention and should not be regarded as limiting the scope. For those of ordinary skill in the art, other related drawings can be obtained based on these drawings without creative efforts.

[0031] Figure 1 It is a schematic flow chart of a method for solidifying radioactive waste provided by the present invention;

[0032] Figure 2 It is a schematic flow chart of a method for recovering DMSO provided by the present invention;

[0033] Figure 3 It is a schematic flow chart of a method for preparing dried uranium powder provided by the present invention. Detailed Embodiments

[0034] It should be noted that the following detailed description is illustrative and aims to provide further explanation of the present application. Unless otherwise specified, all technical and scientific terms used herein have the same meaning as commonly understood by those of ordinary skill in the technical field to which the present application pertains.

[0035] It should be noted that the terms used herein are only for describing specific embodiments and are not intended to limit the exemplary embodiments according to the present application. As used herein, unless the context clearly indicates otherwise, the singular forms are also intended to include the plural forms. In addition, it should also be understood that when the terms "comprise" and / or "include" are used in this specification, they specify the presence of features, steps, operations, devices, components, and / or combinations thereof.

[0036] The present invention will be further described in conjunction with specific examples below. The following examples are only for explaining the present invention and do not constitute a limitation to the present invention. The test samples and test procedures used in the following examples include the following content (if the specific experimental conditions are not specified in the examples, they are usually in accordance with conventional conditions or the conditions recommended by the reagent company; the reagents, consumables, etc. used in the following examples can be obtained from commercial channels unless otherwise specified).

[0037] Figure 1 It is a schematic flow diagram of a method for solidifying radioactive waste provided by the present invention.

[0038] As Figure 1 shown, the present invention provides a method for solidifying radioactive waste, including:

[0039] S1. Distill the DMSO washing water waste liquid containing HMX to recover DMSO and obtain distillation residue;

[0040] S2. Sediment and dry the distillation residue to obtain dry uranium powder;

[0041] S3. Mix ceramic base material, lead compound and the dry uranium powder in a ball milling device to obtain a mixed powder. Place the mixed powder in a mold and apply a pressure of 50 - 200 MPa to compact it to obtain a green body;

[0042] S4. Pre-sinter the green body at a pre-sintering temperature in an inert atmosphere to obtain a pre-sintered green body;

[0043] S5. Sinter the pre-sintered green body at a temperature of 900 - 1250 °C in an inert atmosphere, then hold for 2 - 4 hours, and after cooling to room temperature, obtain the solidified uranium-containing ceramic.

[0044] By drying the uranium solution in the early stage and removing volatile components such as organic solvents and HMX, the gas released by the decomposition of organic components during sintering is greatly reduced, and the porosity is controllable.

[0045] During sintering, uranium oxides (UO2, U3O8) or hydroxides UO2(OH)2 are converted into a stable UO2²⁺ solid state at high temperature; these uranium phases can undergo ion substitution or mechanical encapsulation with the ceramic matrix to form a dense "ceramic-lead-uranium" ternary solidified body ceramic. After that, uranium mainly exists at the grain boundaries or inside the crystal lattice, greatly reducing the dissolution and diffusion of uranium by the external leaching solution.

[0046] Lead has a relatively large attenuation coefficient for high-energy photons such as X-rays and gamma rays. The large atomic nucleus of lead and its outer electron cloud greatly increase the probability of Compton scattering and photoelectric effect of photons, thereby reducing radiation penetration. By introducing lead compounds as one of the raw materials for sintering and solidifying uranium, lead-based compounds such as PbO or PbTiO3 are introduced into the ceramic, which can form a "high-Z phase" microscopically, causing the rays to undergo more scattering or absorption inside the material, forming multiple scattering / absorption interfaces, which is equivalent to adding a "shielding microstructure" to the solidified body. Since uranium itself is also a high-Z (92) element, the solidified body already has a certain ray attenuation effect. When the rays pass through the solidified body, they continuously scatter and lose energy between the uranium phase and the lead phase, thus significantly improving the shielding effect.

[0047] At the same time, lead oxides or lead silicate glasses can soften or partially melt in the range of 700-900 °C, significantly reducing the sintering temperature, which is beneficial to achieving overall densification at a relatively low temperature. The molten lead-based phase can wet the surfaces of ceramic aggregates and uranium oxides, better "adhere" and seal the voids. In a dense material, the paths of ionizing radiation are blocked by a large number of grains and high-Z phases, and the attenuation is more thorough.

[0048] In addition, when lead can form a stable lead-based composite phase with the ceramic matrix, its chemical dissolution rate is significantly reduced. This ensures that lead will not dissolve easily in the later stage (such as in seawater, groundwater and other environments), which not only ensures environmental safety but also maintains the long-term shielding ability of the material.

[0049] In some embodiments, in S3, by mass, the ceramic matrix includes: 35-45 parts of alumina, 25-35 parts of silica, 5-15 parts of zirconia, 3-7 parts of calcium oxide, 3-7 parts of iron oxide and 5-15 parts of binder.

[0050] In this formula, alumina can form a dense α-Al2O3 crystal phase after sintering or react with other components to form a spinel phase (such as MgAl2O4), etc., improving the material strength and reducing the porosity. At high temperature, aluminum ions (Al³⁺) and uranium ions (U 4 ⁺, U6 Ion substitution or co - location can occur among certain newly formed crystal phases with metal cations such as Zr⁺, Fe³⁺, Ca²⁺, etc. Silica can jointly form silicate glass or crystal phases such as mullite (3Al₂O₃·2SiO₂) with Al₂O₃, CaO, etc., providing a silicate network that can "wrap" uranium oxide particles and block their direct contact with the external environment. Partially stabilized zirconia (PSZ) has a transformation toughening mechanism, which reduces the overall micro - cracks after sintering, further reducing the porosity. Compared with the pure Al₂O₃ and SiO₂ systems, it can significantly improve the thermal shock resistance and fracture toughness of the ceramic. Calcium oxide is prone to form various calcium aluminosilicates (such as anorthite CaAl₂Si₂O₈, calcium aluminate melilite, etc.) with SiO₂ and Al₂O₃ at high temperatures, which helps liquid - phase sintering, reduces the local melting point, and accelerates the densification of sintering. Iron oxide can improve the oxidation ability of the system, which helps the stable existence or phase transformation of higher - valence uranium (U 5 ⁺ / U 6 ⁺).

[0051] When sintering at high temperatures (900 - 1300 °C), uranium ions (typical valence states are U 4 ⁺ or U 6 ⁺) can partially replace metal cations (such as Zr 4 ⁺, Fe³⁺, Ca²⁺, etc.) in the matrix lattice, entering the lattice points or grain boundaries. Once uranium enters the lattice, it is not easily dissolved or carried away by leaching solutions (such as groundwater, brine, etc.), and the leaching rate is greatly reduced.

[0052] At the same time, during the sintering process, the silicate glass phase or calcium aluminosilicate phase formed by components such as SiO₂ and CaO can "wrap" uranium oxide particles within the glass - ceramic network. The porosity of the dense sintered body is extremely low (usually < 5%, sometimes up to < 2%), and it is difficult for water and other fluids to penetrate or diffuse, greatly weakening the outward migration of radionuclides.

[0053] In addition, the main crystal phases such as Al₂O₃ and ZrO₂ themselves have high chemical inertness, and their dissolution rates after long - term immersion in acid, alkali, or salt solutions are very low. Components such as Fe₂O₃ and CaO will jointly form various composite oxides or crystal phases (uranates, calcium uranates, iron uranates, zirconium uranates, etc.) with uranium ions under appropriate sintering atmospheres. These phases often have stable structures similar to natural minerals (such as perovskite structure, zircon structure, or pyroxene / garnet structure), which can firmly bind uranium.

[0054] In summary, using this ceramic matrix as a fixative for uranium can:

[0055] (1) Improve the sintering density, reduce the porosity, and reduce the leaching channels.

[0056] (2) A stable ceramic crystal phase or glass-ceramic composite phase is formed, which is extremely chemically inert and uranium ions are not easily migrated.

[0057] (3) Can accommodate uranium in various valence states (U 4 ⁺、U 6 ⁺, etc.), by ion replacement or solid solution, the uranium ions are anchored in the ceramic matrix.

[0058] (4) Compared with traditional cement curing, high-temperature ceramicization has advantages in long-term (decades to hundreds of years) stability and lower leaching rate.

[0059] By carefully designing and optimizing the above formula, and combining it with a suitable sintering process (temperature, atmosphere, holding time, heating and cooling rates), a ceramic solid body with low porosity, high strength and a high degree of fixation for uranium can be obtained, greatly reducing the risk of migration of radioactive materials in the disposal environment.

[0060] In some embodiments, the binder includes one or more of clay, talc, slag, or fly ash.

[0061] In some embodiments, in S3, the ratio of the ceramic base material, the lead compound and the dry uranium powder is (60-75): (10-25): (5-15) by mass.

[0062] Sintering is accomplished through a combination of solid diffusion, liquid wetting, and particle rearrangement. When lead compounds form a liquid phase with a certain volume fraction at high temperatures, they can wet and bond the ceramic and uranium particles, acting as a lubricant / bonding agent. If the liquid phase ratio (caused by an excess of lead compounds) far exceeds the optimal value, the sintered body can experience localized micropore collapse or delamination. Uranium can also migrate and become unevenly distributed in the low-viscosity liquid phase. Even at high temperatures, lead can evaporate over large areas or accumulate on the surface, resulting in uneven internal structure and susceptibility to cracking in the finished product.

[0063] However, if the uranium powder content is too high, the skeleton phase of the ceramic base material will be destroyed by the excess uranium phase, forming a porous, low-strength sintered body, while greatly increasing the radiation activity and increasing the risk of secondary contamination.

[0064] Both uranium (Z = 92) and lead (Z = 82) are high-Z elements, which can effectively absorb / scatter gamma rays and X-rays. The ceramic matrix itself (with relatively low Z, mainly Al, Si, O, etc.) can only provide structural support and chemical stability, and its contribution to the shielding of high-energy photons is limited. When the total proportion of lead compounds and uranium powder reaches 25 - 35% by mass percentage, the solidified body has a significant "high-Z phase distribution". This distribution forms multiple attenuation paths in the thickness direction of the material, thereby significantly weakening the penetration of rays. When the ratio of lead to uranium is appropriate (such as (2 - 1):1), "multiple scattering / absorption of lead-uranium-lead-uranium" can occur inside the material, and the energy loss is amplified in segments; if the proportion of either lead or uranium is too low, the attenuation process is incomplete.

[0065] By setting appropriate ratios, a multi-phase structure as uniform as possible can be formed between lead and the ceramic matrix, and between uranium and the ceramic matrix at high temperatures, and lead and uranium can be locked in stable chemical phases respectively, significantly reducing the risk of long-term leaching. At the same time, the "synergistic shielding" of high-Z lead + high-Z uranium significantly reduces the penetration of gamma rays / X-rays.

[0066] In some embodiments, the lead compound includes one or more of PbO, PbTiO3, and PbO - SiO2.

[0067] In some embodiments, in S4, the pre-sintering temperature is 600 - 800 °C, and the pre-sintering time is 12 hours.

[0068] By pre-sintering at this temperature, residual moisture and a small amount of organic binders can be removed, and the porosity of the uranium-containing ceramic can be increased.

[0069] In some embodiments, the initial temperature of the sintering is 900 - 950 °C, the heating rate is 3 - 5 °C / min, the holding temperature is 1200 - 1250 °C, and the cooling rate is 2 - 5 °C / min.

[0070] By controlling the temperature parameters of the sintering, while moderately reducing the sintering energy consumption, the density and uniformity of the ceramic body can be ensured, enabling the internal crystal lattice to transition smoothly during the heating and cooling processes, and reducing the stress concentration caused by too large a temperature gradient.

[0071] Figure 2 It is a schematic flow diagram of the method for recovering DMSO provided by the present invention.

[0072] As Figure 2 shown, in some embodiments, S1 includes:

[0073] S11. Dilute the DMSO washing waste liquid containing HMX to the safety concentration limit of HMX, and then add a mild oxidant step by step under the condition of 30-60 °C, and stir at a speed of 200 rpm.

[0074] S12. Carry out vacuum distillation at a temperature of 40-60 °C and a pressure of -0.08 MPa to -0.09 MPa to recover DMSO and obtain a distillation residue.

[0075] By diluting HMX to the safety concentration limit and adding a mild oxidant step by step under the mild condition of 30-60 °C, the detonation sensitivity of HMX can be reduced, and accidental deflagration during subsequent distillation or heating can be avoided, providing a safer and more stable raw material environment for subsequent processes such as uranium enrichment precipitation and ceramic sintering.

[0076] In some embodiments, the safety concentration limit of HMX is 5% by mass concentration.

[0077] By carrying out distillation under a reduced pressure state of 40-60 °C and -0.08 MPa to -0.09 MPa, DMSO can be efficiently recovered and energy consumption can be reduced, and it is helpful for resource recovery and waste liquid reduction.

[0078] In some embodiments, the mild oxidant includes one or more of ozone, persulfate, and hypochlorite.

[0079] Figure 3 Schematic diagram of the preparation method process of the dry uranium powder provided by the present invention.

[0080] As Figure 3 shown, in some embodiments, the S2 includes:

[0081] S21. Adjust the pH of the distillation residue to 6-7, and at a temperature of 55-65 °C, add phosphate in batches and stir at a speed of 300 rpm. After reacting for 2 hours, centrifuge at a speed of 4000 rpm to obtain uranium-enriched sediment.

[0082] S22. Carry out vacuum drying of the uranium-enriched sediment for 8-24 hours, and control the drying temperature at 80-120 °C to obtain dry uranium powder.

[0083] By adding phosphate in batches and stirring, uranium ions can react with phosphate radicals to form insoluble uranium phosphate salt ((UO2)(HPO4)·xH2O), effectively trapping and enriching uranium. The appropriate drying temperature avoids the decomposition or sublimation of uranium compounds at too high a temperature, and also avoids the risk of HMX residue caused by rapid heating; it provides a safe and uniform uranium source material for subsequent high-temperature ceramic sintering.

[0084] Example 1

[0085] This embodiment provides a method for solidifying radioactive waste. Specifically, it includes the following steps:

[0086] After diluting the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, a mild oxidant is added step by step at 45 °C and stirred at a speed of 200 rpm;

[0087] Under the conditions of a temperature of 50 °C and a pressure of -0.085 MPa, vacuum distillation is carried out to recover DMSO and obtain a distillation residue;

[0088] Adjust the pH of the distillation residue to 6.5. At a temperature of 60 °C, phosphates are added in batches and stirred at a speed of 300 rpm. After reacting for 2 hours, centrifugal separation is carried out at a speed of 4000 rpm to obtain uranium-enriched sediment;

[0089] The uranium-enriched sediment is vacuum dried for 16 hours, and the drying temperature is controlled at 100 °C to obtain dried uranium powder;

[0090] 700 g of ceramic base material, 150 g of PbO and 100 g of dried uranium powder are mixed in a ball milling device to obtain a mixed powder. The mixed powder is placed in a mold and compacted under a pressure of 100 MPa to obtain a green body. Among them, the ceramic base material includes 280 g of alumina, 210 g of silica, 70 g of zirconia, 35 g of calcium oxide, 35 g of iron oxide and 70 g of talc powder;

[0091] In an inert atmosphere, the green body is pre-sintered at a temperature of 700 °C for 12 hours to obtain a pre-sintered green body;

[0092] In an inert atmosphere, the temperature is raised to 1200 °C at a heating rate of 4 °C / min at a temperature of 925 °C, and the pre-sintered green body is sintered, then held for 3 hours, and then cooled to room temperature at a cooling rate of 3 °C / min to obtain a solidified uranium-containing ceramic.

[0093] Example 2

[0094] This embodiment provides a method for solidifying radioactive waste. Specifically, it includes the following steps:

[0095] After diluting the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, a mild oxidant is added step by step at 45 °C and stirred at a speed of 200 rpm;

[0096] Under the conditions of a temperature of 50 °C and a pressure of -0.085 MPa, vacuum distillation is carried out to recover DMSO and obtain a distillation residue;

[0097] Adjust the pH of the distillation residue to 6.5. At a temperature of 60 °C, add phosphate in batches and stir at a speed of 300 rpm. After reacting for 2 hours, centrifuge at a speed of 4000 rpm to obtain uranium-enriched sediment;

[0098] Vacuum-dry the uranium-enriched sediment for 16 hours, with the drying temperature controlled at 100 °C to obtain dried uranium powder;

[0099] Mix 700 g of ceramic base material, 150 g of PbO, and 150 g of dried uranium powder in a ball-milling device to obtain a mixed powder. Place the mixed powder in a mold and apply a pressure of 100 MPa to compact it into a green body. Among them, the ceramic base material includes 280 g of alumina, 210 g of silica, 70 g of zirconia, 35 g of calcium oxide, 35 g of iron oxide, and 70 g of talc powder;

[0100] In an inert atmosphere, pre-sinter the green body at a temperature of 700 °C for 12 hours to obtain a pre-sintered green body;

[0101] In an inert atmosphere, heat the pre-sintered green body from 700 °C to 1200 °C at a heating rate of 4 °C / min, sinter it, then hold for 3 hours, and then cool it to room temperature at a cooling rate of 3 °C / min to obtain a solidified uranium-containing ceramic.

[0102] Example 3

[0103] This example provides a method for solidifying radioactive waste. Specifically, it includes the following steps:

[0104] Dilute the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, and then add a mild oxidant step by step at 45 °C and stir at a speed of 200 rpm;

[0105] Perform vacuum distillation at a temperature of 50 °C and -0.085 MPa to recover DMSO and obtain a distillation residue;

[0106] Adjust the pH of the distillation residue to 6.5. At a temperature of 60 °C, add phosphate in batches and stir at a speed of 300 rpm. After reacting for 2 hours, centrifuge at a speed of 4000 rpm to obtain uranium-enriched sediment;

[0107] Vacuum-dry the uranium-enriched sediment for 16 hours, with the drying temperature controlled at 100 °C to obtain dried uranium powder;

[0108] In a ball milling device, 700 g of ceramic base material, 200 g of PbO, and 150 g of dry uranium powder are mixed to obtain a mixed powder. The mixed powder is placed in a mold and compacted under a pressure of 100 MPa to obtain a green body. Among them, the ceramic base material includes 280 g of alumina, 210 g of silica, 70 g of zirconia, 35 g of calcium oxide, 35 g of iron oxide, and 70 g of talc powder;

[0109] In an inert atmosphere, at a temperature of 700 °C, the green body is pre-sintered for 12 hours to obtain a pre-sintered green body;

[0110] In an inert atmosphere, at a temperature of 925 °C, the temperature is raised to 1200 °C at a heating rate of 4 °C / min, and the pre-sintered green body is sintered, then held for 3 hours, and then cooled to room temperature at a cooling rate of 3 °C / min to obtain a solidified uranium-containing ceramic.

[0111] Comparative Example 1

[0112] This comparative example provides a method for solidifying radioactive waste. Specifically, it includes the following steps:

[0113] After diluting the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, a mild oxidant is added step by step at 45 °C and stirred at a speed of 200 rpm;

[0114] At a temperature of 50 °C and -0.085 MPa, vacuum distillation is carried out to recover DMSO and obtain a distillation residue;

[0115] The pH of the distillation residue is adjusted to 6.5. At a temperature of 60 °C, phosphates are added in batches and stirred at a speed of 300 rpm. After reacting for 2 hours, centrifugal separation is carried out at a speed of 4000 rpm to obtain uranium-enriched sediment;

[0116] The uranium-enriched sediment is vacuum dried for 16 hours, and the drying temperature is controlled at 100 °C to obtain dry uranium powder;

[0117] In a ball milling device, 700 g of ceramic base material and 100 g of dry uranium powder are mixed, but PbO is not added, to obtain a mixed powder. The mixed powder is placed in a mold and compacted under a pressure of 100 MPa to obtain a green body. Among them, the ceramic base material includes 280 g of alumina, 210 g of silica, 70 g of zirconia, 35 g of calcium oxide, 35 g of iron oxide, and 70 g of talc powder;

[0118] In an inert atmosphere, at a temperature of 700 °C, the green body is pre-sintered for 12 hours to obtain a pre-sintered green body;

[0119] In an inert atmosphere, the pre-sintered green body is sintered at a temperature of 925 °C with a heating rate of 4 °C / min up to 1200 °C, then held for 3 hours, and then cooled to room temperature at a cooling rate of 3 °C / min to obtain a solidified uranium-containing ceramic.

[0120] Comparative Example 2

[0121] This comparative example provides a method for solidifying radioactive waste. Specifically, it includes the following steps:

[0122] After diluting the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, a mild oxidant is added step by step at 45 °C and stirred at a speed of 200 rpm;

[0123] At a temperature of 50 °C and -0.085 MPa, vacuum distillation is carried out to recover DMSO, and a distillation residue is obtained;

[0124] The pH of the distillation residue is adjusted to 6.5. At a temperature of 60 °C, phosphates are added in batches and stirred at a speed of 300 rpm. After reacting for 2 hours, centrifugal separation is carried out at a speed of 4000 rpm to obtain a uranium-enriched sediment;

[0125] The uranium-enriched sediment is vacuum dried for 16 hours, and the drying temperature is controlled at 100 °C to obtain a dried uranium powder;

[0126] 700 g of ceramic base material, 150 g of PbO and 100 g of dried uranium powder are mixed in a ball milling device to obtain a mixed powder. The mixed powder is placed in a mold and compacted under a pressure of 100 MPa to obtain a green body. Among them, the ceramic base material includes 280 g of alumina, 210 g of silica, 70 g of zirconia, 35 g of calcium oxide, 35 g of iron oxide and 70 g of talc powder;

[0127] In an inert atmosphere, the green body is sintered at a temperature of 925 °C with a heating rate of 4 °C / min up to 1200 °C, then held for 3 hours, and then cooled to room temperature at a cooling rate of 3 °C / min to obtain a solidified uranium-containing ceramic.

[0128] Experimental detection method

[0129] The uranium-containing ceramics obtained in Examples 1-3 and Comparative Examples 1 and 2 are detected, including porosity measurement, compressive strength measurement, uranium leaching rate and gamma ray attenuation rate.

[0130] Specifically, the detection method is as follows:

[0131] Porosity (apparent porosity) measurement: Deionized water is used for the measurement of apparent porosity. First, measure the dry mass, then measure the impregnated mass after vacuum degassing in water, and finally measure the suspended mass in the water under the water-saturated state. Calculate the apparent porosity and bulk density according to the standard formula, and the standard formula is as follows:

[0132]

[0133] Among them, is the impregnated mass, is the dry mass, is the suspended mass.

[0134] Compressive strength measurement: Apply a load using a universal testing machine at a constant loading rate until the specimen is crushed, record the maximum load, and calculate the compressive strength based on the cross-sectional area of the specimen.

[0135] Uranium leaching rate: After crushing the specimen to meet the particle size requirements with a crusher, mix it with the TCLP leaching solution at a solid-liquid ratio of 1:20 and continuously stir for 18 hours, filter and collect the leaching solution, and use an inductively coupled plasma mass spectrometer to measure the uranium concentration.

[0136] Gamma-ray attenuation rate: Place the ^137Cs radiation source and the scintillation detector at a fixed distance. First, measure the counting rate in the empty field state, then place the specimen between them and measure the counting rate again, and calculate the attenuation rate by comparing the two measurement results.

[0137] Experimental equipment

[0138] Electronic analytical balance for measuring mass: Mettler Toledo-XS205

[0139] Vacuum degassing device: Shanghai Huate Technology HWT-5

[0140] Universal testing machine: Instron 5967

[0141] Crusher: Tianjin Test Instrument FZ-102

[0142] Filter: 0.45μm PES filter membrane of Sartorius and filtration device

[0143] Inductively coupled plasma mass spectrometer: Agilent 7900 ICP-MS

[0144] ^137Cs radiation source: Dongfanghong Source DJG-137

[0145] NaI(Tl) scintillation detector: ORTEC 905 series

[0146] Experimental results

[0147] Table 1 Detection Results of Uranium-containing Ceramics

[0148]

[0149] As shown in Table 1, in Example 2, the ratio of PbO:uranium powder was adjusted from 1.5:1 to 1:1, the total amount of PbO remained unchanged, but the amount of uranium powder increased (150 g), resulting in a slightly increased porosity (3.8%) and a slightly decreased strength (110 MPa). In addition, due to the increased uranium content in the material, the gamma attenuation rate slightly increased (96%), and it synergistically shielded radiation with PbO. Compared with Example 1, by implementing the formulation of Example 2, the ability of the uranium-containing ceramic in ionizing radiation shielding was slightly improved, but the mechanical and leaching properties were slightly sacrificed.

[0150] In Example 3, compared with Example 1, the total proportion of PbO + uranium powder was higher, resulting in a porosity increase to 4.2% and a strength of 105 MPa. However, due to the synergistic shielding of high lead + high uranium (both are high-Z elements), the gamma attenuation rate was further increased (97%), and the uranium leaching rate could also be maintained at a low level (1.0×10 -5 g / m²·d).

[0151] In Comparative Example 1, compared with Example 1, no lead compound (PbO) was added. When there was no PbO, the porosity increased (6.8%), and the uranium leaching rate also increased accordingly (2.0×10 -5 g / m²·d), and the gamma attenuation rate decreased significantly to 82%. This shows that the addition of PbO not only helps with densification (reducing porosity) but also improves the distribution of high-Z (82) elements in the material, enhancing radiation shielding.

[0152] In Comparative Example 2, compared with Example 1, the pre-sintering step was cancelled, and the final porosity was as high as 8.5%, the strength decreased significantly (60 MPa), and the uranium leaching rate and gamma attenuation rate showed worse performance. This shows that the pre-sintering step helps to remove bound water, organic residues, etc., and allows the melt or crystal phase to be more evenly distributed in the subsequent sintering, reducing microcracks and pores.

[0153] In summary, from the table results, it can be seen that reasonable ratios of PbO and uranium powder (such as in Examples 1, 2, and 3) and performing pre-sintering (Examples 1, 2, and 3) can significantly improve the density, compressive strength, uranium fixation, and radiation shielding ability of the solidified body.

[0154] Comparing the results of the two comparative examples (without adding PbO or cancelling the pre-sintering), whether it is porosity, mechanical properties, uranium leaching rate, or gamma attenuation rate, they are all significantly deteriorated, indicating that without these two elements, the material properties have dropped significantly.

[0155] Obviously, the described embodiments are only a part of the embodiments of the present invention, rather than all embodiments. All other embodiments obtained by those of ordinary skill in the art and related fields based on the embodiments of the present invention without creative efforts shall fall within the scope of protection of the present invention. Structures, devices, and operation methods not specifically described and explained in the present invention shall be implemented by conventional means in the art unless otherwise specified and limited.

Claims

1. A method for solidifying radioactive waste, characterized in that, Including: S1. Distill the DMSO washing water waste liquid containing HMX to recover DMSO and obtain a distillation residue. S2. Sediment and dry the distillation residue to obtain dry uranium powder. S3. Mix a ceramic base material, a lead compound, and the dry uranium powder in a ball milling device to obtain a mixed powder. Place the mixed powder in a mold and apply a pressure of 50 - 200 MPa to compact it into a green body. The ceramic base material includes: 35 - 45 parts of alumina, 25 - 35 parts of silica, 5 - 15 parts of zirconia, 3 - 7 parts of calcium oxide, 3 - 7 parts of iron oxide, and 5 - 15 parts of a binder. The lead compound includes one or more of PbO, PbTiO3, and PbO - SiO2. S4. Pre - sinter the green body at a pre - sintering temperature in an inert atmosphere to obtain a pre - sintered green body. S5. Sinter the pre - sintered green body at a temperature of 900 - 1250 °C in an inert atmosphere, then hold for 2 - 4 hours, and after cooling to room temperature, obtain a solidified uranium - containing ceramic.

2. The radioactive waste solidification method according to claim 1, characterized in that The binder includes one or more of clay, talc powder, slag, or fly ash.

3. The radioactive waste solidification method according to claim 1, wherein In S3, by mass, the ratio of the ceramic base material, the lead compound, and the dry uranium powder is (60 - 75):(10 - 25):(5 - 15).

4. The radioactive waste solidification method according to claim 1, characterized in that, In S4, the pre - sintering temperature is 600 - 800 °C, and the pre - sintering time is 12 hours.

5. The radioactive waste solidification method according to claim 1, characterized in that, The initial temperature of the sintering is 900 - 950 °C, the heating rate is 3 - 5 °C / min, the holding temperature is 1200 - 1250 °C, and the cooling rate is 2 - 5 °C / min.

6. The radioactive waste solidification method according to claim 1, wherein, S1 includes: Dilute the DMSO washing water waste liquid containing HMX to the safety concentration limit of HMX, and then add a mild oxidant step - by - step at 30 - 60 °C and stir at a speed of 200 rpm. Perform vacuum distillation at a temperature of 40 - 60 °C and a pressure of - 0.08 MPa to - 0.09 MPa to recover DMSO and obtain a distillation residue.

7. The radioactive waste solidification method according to claim 6, characterized in that The mild oxidant includes one or more of ozone, persulfate, and hypochlorite.

8. The radioactive waste solidification method according to claim 1, characterized in that, S2 includes: Adjust the pH of the distillation residue to 6 - 7, add phosphate in batches at a temperature of 55 - 65 °C and stir at a speed of 300 rpm. After reacting for 2 hours, centrifuge at a speed of 4000 rpm to obtain a uranium - enriched sediment. Vacuum - dry the uranium - enriched sediment for 8 - 24 hours, and control the drying temperature at 80 - 120 °C to obtain dry uranium powder.

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