CN115806386A - Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body - Google Patents
Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body Download PDFInfo
- Publication number
- CN115806386A CN115806386A CN202211340346.7A CN202211340346A CN115806386A CN 115806386 A CN115806386 A CN 115806386A CN 202211340346 A CN202211340346 A CN 202211340346A CN 115806386 A CN115806386 A CN 115806386A
- Authority
- CN
- China
- Prior art keywords
- sulfur
- waste liquid
- radioactive waste
- solidified body
- level radioactive
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
- 239000007788 liquid Substances 0.000 title claims abstract description 70
- 229910052717 sulfur Inorganic materials 0.000 title claims abstract description 55
- 239000011593 sulfur Substances 0.000 title claims abstract description 55
- 239000005388 borosilicate glass Substances 0.000 title claims abstract description 29
- 238000000034 method Methods 0.000 title claims abstract description 24
- 239000002699 waste material Substances 0.000 title claims description 14
- 239000011521 glass Substances 0.000 claims abstract description 47
- 239000002927 high level radioactive waste Substances 0.000 claims abstract description 38
- 238000001354 calcination Methods 0.000 claims abstract description 21
- 238000001704 evaporation Methods 0.000 claims abstract description 18
- 239000000654 additive Substances 0.000 claims abstract description 14
- 230000000996 additive effect Effects 0.000 claims abstract description 14
- 239000011324 bead Substances 0.000 claims abstract description 13
- 238000001816 cooling Methods 0.000 claims abstract description 11
- 238000002156 mixing Methods 0.000 claims abstract description 11
- 239000011259 mixed solution Substances 0.000 claims description 22
- 230000008020 evaporation Effects 0.000 claims description 17
- 238000005352 clarification Methods 0.000 claims description 10
- VWDWKYIASSYTQR-UHFFFAOYSA-N sodium nitrate Chemical compound [Na+].[O-][N+]([O-])=O VWDWKYIASSYTQR-UHFFFAOYSA-N 0.000 claims description 10
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims description 8
- 230000004992 fission Effects 0.000 claims description 8
- 229910017604 nitric acid Inorganic materials 0.000 claims description 8
- 239000003758 nuclear fuel Substances 0.000 claims description 8
- 230000008569 process Effects 0.000 claims description 7
- AYJRCSIUFZENHW-UHFFFAOYSA-L barium carbonate Chemical compound [Ba+2].[O-]C([O-])=O AYJRCSIUFZENHW-UHFFFAOYSA-L 0.000 claims description 6
- IWOUKMZUPDVPGQ-UHFFFAOYSA-N barium nitrate Chemical compound [Ba+2].[O-][N+]([O-])=O.[O-][N+]([O-])=O IWOUKMZUPDVPGQ-UHFFFAOYSA-N 0.000 claims description 6
- 239000006060 molten glass Substances 0.000 claims description 6
- 238000000137 annealing Methods 0.000 claims description 5
- 238000000265 homogenisation Methods 0.000 claims description 5
- 239000011734 sodium Substances 0.000 claims description 5
- 235000010344 sodium nitrate Nutrition 0.000 claims description 5
- 239000004317 sodium nitrate Substances 0.000 claims description 5
- 229910018068 Li 2 O Inorganic materials 0.000 claims description 4
- 229910004298 SiO 2 Inorganic materials 0.000 claims description 4
- 239000002245 particle Substances 0.000 claims description 4
- 238000004519 manufacturing process Methods 0.000 claims description 3
- 239000000203 mixture Substances 0.000 claims description 2
- NINIDFKCEFEMDL-UHFFFAOYSA-N Sulfur Chemical compound [S] NINIDFKCEFEMDL-UHFFFAOYSA-N 0.000 abstract description 8
- 239000002901 radioactive waste Substances 0.000 abstract description 7
- 159000000009 barium salts Chemical class 0.000 abstract description 6
- TZCXTZWJZNENPQ-UHFFFAOYSA-L barium sulfate Chemical compound [Ba+2].[O-]S([O-])(=O)=O TZCXTZWJZNENPQ-UHFFFAOYSA-L 0.000 abstract description 3
- 238000002360 preparation method Methods 0.000 abstract 1
- PMZURENOXWZQFD-UHFFFAOYSA-L Sodium Sulfate Chemical compound [Na+].[Na+].[O-]S([O-])(=O)=O PMZURENOXWZQFD-UHFFFAOYSA-L 0.000 description 6
- 239000010808 liquid waste Substances 0.000 description 6
- 229910052938 sodium sulfate Inorganic materials 0.000 description 6
- 235000011152 sodium sulphate Nutrition 0.000 description 6
- 238000012360 testing method Methods 0.000 description 6
- QAOWNCQODCNURD-UHFFFAOYSA-L Sulfate Chemical compound [O-]S([O-])(=O)=O QAOWNCQODCNURD-UHFFFAOYSA-L 0.000 description 4
- 230000002285 radioactive effect Effects 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 229910018072 Al 2 O 3 Inorganic materials 0.000 description 3
- 239000013078 crystal Substances 0.000 description 3
- 230000008023 solidification Effects 0.000 description 3
- 238000007711 solidification Methods 0.000 description 3
- 238000002425 crystallisation Methods 0.000 description 2
- 230000008025 crystallization Effects 0.000 description 2
- 238000002844 melting Methods 0.000 description 2
- 230000008018 melting Effects 0.000 description 2
- 238000005191 phase separation Methods 0.000 description 2
- 239000000941 radioactive substance Substances 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 description 1
- 229910002651 NO3 Inorganic materials 0.000 description 1
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 1
- CZMRCDWAGMRECN-UGDNZRGBSA-N Sucrose Chemical compound O[C@H]1[C@H](O)[C@@H](CO)O[C@@]1(CO)O[C@@H]1[C@H](O)[C@@H](O)[C@H](O)[C@@H](CO)O1 CZMRCDWAGMRECN-UGDNZRGBSA-N 0.000 description 1
- 229930006000 Sucrose Natural products 0.000 description 1
- 239000005354 aluminosilicate glass Substances 0.000 description 1
- 230000009286 beneficial effect Effects 0.000 description 1
- 230000005587 bubbling Effects 0.000 description 1
- 239000000919 ceramic Substances 0.000 description 1
- 239000003638 chemical reducing agent Substances 0.000 description 1
- 239000000599 controlled substance Substances 0.000 description 1
- 230000007547 defect Effects 0.000 description 1
- 238000013461 design Methods 0.000 description 1
- 238000001514 detection method Methods 0.000 description 1
- 238000011038 discontinuous diafiltration by volume reduction Methods 0.000 description 1
- 239000000383 hazardous chemical Substances 0.000 description 1
- 230000006872 improvement Effects 0.000 description 1
- XEEYBQQBJWHFJM-UHFFFAOYSA-N iron Substances [Fe] XEEYBQQBJWHFJM-UHFFFAOYSA-N 0.000 description 1
- 229910052742 iron Inorganic materials 0.000 description 1
- 229910000398 iron phosphate Inorganic materials 0.000 description 1
- WBJZTOZJJYAKHQ-UHFFFAOYSA-K iron(3+) phosphate Chemical compound [Fe+3].[O-]P([O-])([O-])=O WBJZTOZJJYAKHQ-UHFFFAOYSA-K 0.000 description 1
- 238000002386 leaching Methods 0.000 description 1
- 238000010907 mechanical stirring Methods 0.000 description 1
- 238000010309 melting process Methods 0.000 description 1
- 238000012986 modification Methods 0.000 description 1
- 230000004048 modification Effects 0.000 description 1
- 150000002823 nitrates Chemical class 0.000 description 1
- 238000011056 performance test Methods 0.000 description 1
- 239000005365 phosphate glass Substances 0.000 description 1
- 238000001556 precipitation Methods 0.000 description 1
- 230000005855 radiation Effects 0.000 description 1
- 230000009467 reduction Effects 0.000 description 1
- 150000003839 salts Chemical class 0.000 description 1
- 239000005368 silicate glass Substances 0.000 description 1
- 229910052708 sodium Inorganic materials 0.000 description 1
- 239000000243 solution Substances 0.000 description 1
- 239000002915 spent fuel radioactive waste Substances 0.000 description 1
- 239000005720 sucrose Substances 0.000 description 1
- 229910021653 sulphate ion Inorganic materials 0.000 description 1
Images
Landscapes
- Glass Compositions (AREA)
Abstract
The invention discloses a method for preparing a borosilicate glass solidified body of high-sulfur high-radioactive waste liquid, and relates to the technical field of nuclear facility retirement and radioactive waste treatment. The preparation method comprises the following steps: mixing the high-sulfur high-level radioactive waste liquid with an additive, and then evaporating, denitrating, calcining, clarifying, homogenizing and cooling to obtain the high-sulfur high-level radioactive waste liquid borosilicate glass solidified body. Soluble barium salt is added into the high-sulfur high-level radioactive waste liquid, and the barium salt can form stable barium sulfate salt with sulfate radical and then is melted with glass beads to form uniform and stable glass bodies. The solubility of sulfur in the borosilicate glass solidified body is increased from less than or equal to 1 percent to more than or equal to 2 percent, the wrapping capacity of the glass solidified body to sulfur is effectively improved, the generation of yellow phase is avoided, and the prepared glass solidified body has relatively excellent performance.
Description
Technical Field
The invention relates to the technical field of nuclear facility retirement and radioactive waste treatment, in particular to a method for preparing a borosilicate glass solidified body of high-sulfur high-radioactivity waste liquid.
Background
The high-level radioactive waste liquid generated in the post-treatment process of the nuclear fuel contains more than 97% of total fission products in the irradiated nuclear fuel, mainly comprises nitric acid, sodium nitrate, fission products, transuranic elements and other components, and has the characteristics of strong radioactivity, high heat release rate, strong corrosivity and the like. These high levels of radioactive waste must pose serious environmental hazards once introduced into the environment, and must be carefully managed and properly disposed of. The glass solidification is the only international industrially applied high level liquid waste treatment method, and the liquid waste is concentrated and calcined to convert the salt in the liquid waste into oxide, which is then melted together with glass additive to finally form stable glass solidified body. At present, silicate glass solidified bodies, iron phosphate glass solidified bodies, aluminosilicate glass solidified bodies, ceramic bodies and microcrystalline glass have been developed at home and abroad. Among them, borosilicate glass is the first choice in many countries for curing high-level radioactive liquid waste glass because of its good radiation resistance, chemical stability and water resistance.
The high-sulfur high-level radioactive waste liquid is a high-radioactivity waste liquid containing sulfur which is introduced in the post-treatment process of spent fuel, and SO in the high-level radioactive waste liquid in China 3 The content is > 4%, due to the low solubility of the sulphate in borosilicate glass (SO) 3 Less than or equal to 1 percent), SO that the problem of 'yellow phase' is encountered when borosilicate glass is used for curing high-sulfur high-level radioactive waste liquid, namely, a separated yellow second phase (the main component is SO) is generated in the glass melting process 3 、Na 2 O、MoO 3 、Cr 2 O 3 And CaO). The yellow phase is easy to adsorb nuclides such as 90Sr and 137Cs and is easy to dissolve in water, so that the leaching resistance of the glass solidified body is obviously reduced, and therefore, the generation of the yellow phase in the glass solidification process must be avoided.
The 'yellow phase' problem is a difficult problem encountered in the treatment of high-sulfur high-level radioactive waste liquid in many countries, and at present, the following solutions are mainly provided for the 'yellow phase' problem at home and abroad: (1) Adding reducing agent such as carbon powder, sucrose, etc. to decompose sulfate into gas (SO) 2 ) Volatilization, but the method increases the burden of tail gas treatment and secondary waste liquid treatment, and nitrate in the waste liquid can be reduced and decomposed; (2) By increasing the melting temperature, prolonging the melting time and increasing SO 2 The volatile amount is reduced, but the service life of the furnace is shortened in the process, and the easy volatile is increasedThe volatilization amount of radioactive elements (such as Ru) is reduced, and the production capacity of the furnace is reduced; (3) The pack capacity is reduced, but this is not preferable in terms of improvement of the volume reduction ratio, reduction of cost, and the like; (4) The method is characterized in that borosilicate basic glass components are optimized, or the sulfate-containing capacity of borosilicate glass is improved by adopting measures such as bubbling and mechanical stirring, but the measures are limited for improving the sulfate-containing capacity of borosilicate glass.
Disclosure of Invention
The invention aims to overcome the defects of the prior art and provide a method for preparing a borosilicate glass solidified body of high-sulfur high-emission waste liquid, and the prepared glass solidified body has high sulfur content in the high-sulfur high-emission waste liquid and can effectively solve the problem of yellow phase precipitation.
In order to achieve the purpose, the technical scheme adopted by the invention is as follows:
a method for preparing a high-sulfur high-level radioactive waste liquid borosilicate glass solidified body comprises the following steps:
s1: mixing 1L of high-sulfur high-level radioactive waste liquid with 2-4 g of additive to obtain mixed liquid;
s2: and (3) evaporation: taking 60ml of the mixed solution obtained in the step S1, uniformly mixing the mixed solution with 45-55 g of glass beads, and completing the evaporation and salinization of the high-sulfur high-level radioactive waste liquid at the temperature of 100-200 ℃;
s3: denitration: carrying out denitration twice on the mixed solution after the evaporation and the salinization, wherein the denitration for the first time is calcining for 1-2 hours at the temperature of 200-600 ℃; the second denitration is calcining for 2 to 3 hours at the temperature of 600 to 800 ℃;
s4: and (3) calcining: calcining the denitrated mixed solution at 800-1000 ℃ for 1-2 h to obtain molten liquid;
s5: clarification and homogenization: clarifying and homogenizing the molten liquid to obtain molten glass, wherein the temperature in a clarification tank is 1000-1150 ℃;
s6: and (3) cooling: and annealing the clarified and homogenized glass liquid at the temperature of 500-700 ℃ for 1-2 h, and naturally cooling to room temperature to obtain the high-sulfur high-level radioactive waste liquid borosilicate glass solidified body.
Further, the additive amount in step S1 was 2g.
Further, the addition amount of the glass beads in step S2 was 50g.
Further, the high-sulfur high-level radioactive waste liquid generated in the nuclear fuel post-treatment process in the step S1 contains 97% of total fission products in the irradiated nuclear fuel, and mainly contains nitric acid, sodium nitrate and fission products.
Further, the additive in the step S1 is one or more of barium nitrate, barium carbonate and nitric acid.
Further, the glass beads in the step S2 have a particle size of 1 to 3mm and a composition comprising 50 to 53.44w% of SiO 2 12.5-14.6 w% of B 2 O 3 4.25-5.21 w% of Na 2 O, 2.21-3.16 w% Li 2 O, 4.4-6.12 w% Al 2 O 3 2.21-3.42w% CaO, 1.05-2.10w% MgO.
Compared with the prior art, the invention has the following beneficial effects:
(1) Because the high-sulfur high-level radioactive waste liquid contains a certain amount of sodium sulfate, the sulfate radicals can be separated out from the glass in the form of sodium sulfate at a high temperature, and the density of the sodium sulfate is lower than that of the glass, so that the glass is subjected to phase separation, and a yellow phase appears. Soluble barium salt is added into the high-sulfur high-level radioactive waste liquid, and the barium salt can form stable barium sulfate salt with sulfate radical and then is melted with glass beads to form uniform and stable glass bodies. The solubility of sulfur in the borosilicate glass solidified body is increased from less than or equal to 1% to more than or equal to 2%, the sulfur capacity of the glass solidified body is effectively improved, the generation of a yellow phase is avoided, and the prepared glass solidified body has excellent performance.
(2) The invention is also provided with a denitration step, and the denitration is carried out in two sections, the temperature of the first section is low and is mainly used for removing the bound water and completing the volatilization of the nitric acid in the waste liquid, the temperature of the second section is slightly higher than that of the first section, and the denitration step is mainly used for decomposing the sodium nitrate and other nitrates in the waste liquid into oxides and NO X . The denitration is carried out in two sections, and the two sections have different temperatures, so that the denitration speed can be well controlled, and radioactive substances in the waste liquid are prevented from being brought away by gas, and further radioactive substances are prevented from being damagedNamely loss of glass components.
Drawings
FIG. 1 is a graph showing the results of crystal testing of a glass solidification body according to the present invention;
FIG. 2 shows thermogravimetric test data of glass of the present invention.
Detailed Description
The present invention will be further described with reference to the following description and examples, which include but are not limited to the following examples.
The invention provides a method for preparing a borosilicate glass solidified body of high-sulfur high-level radioactive waste liquid, which comprises the following steps:
s1: mixing 1L of high-sulfur high-level radioactive waste liquid with 2-4 g of additive to obtain mixed liquid;
s2: and (3) evaporation: taking 60ml of the mixed solution obtained in the step S1, uniformly mixing the mixed solution with 45-55 g of glass beads, and completing the evaporation and salinization of the high-sulfur high-level radioactive waste liquid at the temperature of 100-200 ℃;
s3: denitration: carrying out denitration twice on the mixed solution after the evaporation and the salinization, wherein the denitration for the first time is calcining for 1-2 hours at the temperature of 200-600 ℃; the second denitration is calcining for 2 to 3 hours at the temperature of 600 to 800 ℃;
s4: and (3) calcining: calcining the denitrated mixed solution at 800-1000 ℃ for 1-2 h to obtain molten liquid;
s5: clarification and homogenization: clarifying and homogenizing the molten liquid to obtain molten glass, wherein the temperature in a clarification tank is 1000-1150 ℃;
s6: and (3) cooling: and annealing the clarified and homogenized glass liquid at the temperature of 500-700 ℃ for 1-2 h, and naturally cooling to room temperature to obtain the high-sulfur high-level radioactive waste liquid borosilicate glass solidified body.
The high-sulfur high-level radioactive waste liquid generated in the post-treatment process of the nuclear fuel contains 97 percent of total fission products in the irradiated nuclear fuel and mainly contains nitric acid, sodium nitrate and fission products; the additive is one or more of barium nitrate, barium carbonate and nitric acid; the glass beads have a particle size of 1 to 3mm and contain 50 to 53.44w% of SiO as a main component 2 12.5-14.6 w% of B 2 O 3 4.25-5.21 w% of Na 2 O, 2.21-3.16 w% of Li 2 O, 4.4-6.12 w% Al 2 O 3 CaO 2.21-3.42w%, mgO 1.05-2.10 w%.
Because the high-sulfur high-level radioactive waste liquid contains a certain amount of sodium sulfate, the sulfate radicals can be separated out from the glass in the form of sodium sulfate at a high temperature, and the density of the sodium sulfate is lower than that of the glass, so that the glass is subjected to phase separation and a yellow phase is generated. In this embodiment, soluble barium salt (additive) is added to the high-sulfur high-level radioactive waste liquid, and the barium salt can form stable barium sulfate salt with sulfate radical, and then is melted with glass beads to form a uniform and stable glass body. In addition, caO and MgO can follow SiO 2 In the embodiment, the additive can reduce the content of CaO and MgO in the waste liquid, further reduce the generation rate of the perovskite and further reduce the crystallization of the glass solidified body.
The high-sulfur high-level radioactive waste liquid borosilicate glass solidified body prepared by the method is characterized in that: the content of oxide contained in the solidified body is less than 19%, the capacity of Huang Xiangbao reaches 2%, the viscosity of the solidified body is less than 800mPas at 1150 ℃, the liquidus temperature reaches 900 ℃, the crystallization temperature is 750-800 ℃, and the main component of the oxide is SiO 2 ：45.90～46.7％、B 2 O 3 ：10.90～12.90％、Na 2 O：10.00～11％、Li 2 O：2.00％、Al 2 O 3 ：3.80～5.00％、CaO：1.00-2.00％、MgO：0.50～1.5.0％、BaO：1.50～3.00％、V 2 O 5 ：1.50％、Fe 2 O 3 :3.70% and transuranic elements: 1.05 to 3.02 percent of Cr 2 O 3 ：0.50％、MoO 3 :0.40 percent of sulfate radical and 0 to 2.02 percent of sulfate radical.
Example 1
The embodiment provides a method for preparing a borosilicate glass solidified body with high-sulfur and high-radioactive waste liquid, which comprises the following steps:
s1: mixing 1L of high-sulfur high-level radioactive waste liquid with 2.5g of additive to obtain a mixed solution;
s2: and (3) evaporation: taking 60ml of the mixed solution obtained in the step S1, uniformly mixing the mixed solution with 45g of glass beads, and finishing the evaporation and salinization of the high-sulfur high-level radioactive waste liquid at the temperature of 100-200 ℃;
s3: denitration: carrying out denitration twice on the mixed solution after the evaporation and the salinization, wherein the denitration for the first time is calcining for 1.5 hours at the temperature of 400 ℃; the second denitration is calcination for 2 hours at the temperature of 600 ℃;
s4: and (3) calcining: calcining the denitrated mixed solution at 950 ℃ for 1.5h to obtain molten liquid;
s5: clarification and homogenization: clarifying and homogenizing the molten liquid to obtain molten glass, wherein the temperature in a clarification tank is 1050 ℃;
s6: and (3) cooling: and (3) annealing the clarified and homogenized molten glass at the temperature of 700 ℃ for 1.5h, and naturally cooling to room temperature to obtain the high-sulfur high-level radioactive liquid waste glass solidified body.
Example 2
The embodiment provides a method for preparing a borosilicate glass solidified body with high-sulfur and high-radioactive waste liquid, which comprises the following steps:
s1: mixing 1L of high-sulfur high-level radioactive waste liquid with 3g of additive to obtain a mixed solution;
s2: and (3) evaporation: taking 60ml of the mixed solution obtained in the step S1, uniformly mixing the mixed solution with 50g of glass beads, and finishing the evaporation and salinization of the high-sulfur high-level radioactive waste liquid at the temperature of 100-200 ℃;
s3: denitration: carrying out denitration twice on the mixed solution after the evaporation and the salinization, wherein the denitration for the first time is calcination for 2 hours at the temperature of 450 ℃; the second denitration is calcination for 2 hours at the temperature of 700 ℃;
s4: and (3) calcining: calcining the denitrated mixed solution at 1000 ℃ for 2h to obtain molten liquid;
s5: clarification and homogenization: clarifying and homogenizing the molten liquid to obtain molten glass, wherein the temperature in a clarification tank is 1150 ℃;
s6: and (3) cooling: and (3) annealing the clarified and homogenized glass liquid at the temperature of 500 ℃ for 2h, and naturally cooling to room temperature to obtain the high-sulfur high-level radioactive liquid waste glass solidified body.
When the evaporation salinization is carried out, the temperature is controlled to be 100-200 ℃, the temperature is not suitable to be too high, because water vapor can take away a large amount of particles in the waste liquid during high-temperature evaporation, and the volatilization of substances in the radioactive waste liquid is accelerated. The high-sulfur high-level radioactive waste liquid borosilicate glass solidified body prepared by the invention is subjected to performance test, crystal test and thermogravimetric test respectively, and the test results are shown in fig. 1 and fig. 2. As can be seen from the test results of fig. 1, no crystals were generated in the glass-solidified body, and the glass-solidified body was excellent in performance; as can be seen from FIG. 2, the glass-solidified body has good thermogravimetric properties and can withstand high temperatures, and the liquidus temperature reaches 900 ℃. The invention also obtains that the solubility of sulfur in the borosilicate glass solidified body is more than or equal to 2 percent through detection, and the sulfur containing capacity of the glass solidified body prepared by the prior art is less than or equal to 1 percent.
The above-mentioned embodiment is only one of the preferred embodiments of the present invention, and should not be used to limit the scope of the present invention, but all the insubstantial modifications or changes made within the spirit and scope of the main design of the present invention, which still solve the technical problems consistent with the present invention, should be included in the scope of the present invention.
Claims (6)
1. A method for preparing a high-sulfur high-level radioactive waste liquid borosilicate glass solidified body is characterized by comprising the following steps:
s1: mixing 1L of high-sulfur high-level radioactive waste liquid with 2-4 g of additive to obtain mixed liquid;
s2: and (3) evaporation: taking 60ml of the mixed solution obtained in the step S1, uniformly mixing the mixed solution with 45-55 g of glass beads, and completing the evaporation and salinization of the high-sulfur high-level radioactive waste liquid at the temperature of 100-200 ℃;
s3: denitration: carrying out denitration twice on the mixed solution after the evaporation and the salinization, wherein the denitration for the first time is calcining for 1-2 hours at the temperature of 200-600 ℃; the second denitration is calcining for 2 to 3 hours at the temperature of 600 to 800 ℃;
s4: and (3) calcining: calcining the denitrated mixed solution at 800-1000 ℃ for 1-2 h to obtain molten liquid;
s5: clarification and homogenization: clarifying and homogenizing the molten liquid to obtain molten glass, wherein the temperature in a clarification tank is 1000-1150 ℃;
s6: and (3) cooling: and annealing the clarified and homogenized glass liquid at the temperature of 500-700 ℃ for 1-2 h, and naturally cooling to room temperature to obtain the high-sulfur high-level radioactive waste liquid borosilicate glass solidified body.
2. The method for producing a high-sulfur high-level radioactive waste borosilicate glass cured body according to claim 1, wherein the additive is added in an amount of 2g in step S1.
3. The method for producing a high-sulfur high-level radioactive waste borosilicate glass cured body according to claim 1, wherein the addition amount of the glass beads in step S2 is 50g.
4. The method for preparing the high-sulfur high-level radioactive waste liquid borosilicate glass solidified body according to claim 1, wherein the high-sulfur high-level radioactive waste liquid generated in the nuclear fuel post-treatment process in the step S1 contains 97% of total fission products in irradiated nuclear fuel, and the main components of the high-sulfur high-level radioactive waste liquid are nitric acid, sodium nitrate and fission products.
5. The method for preparing the high-sulfur high-level radioactive waste borosilicate glass solidified body according to claim 1, wherein the additive in the step S1 is one or more of barium nitrate, barium carbonate and nitric acid.
6. The method for preparing a high-sulfur high-level emission waste liquid borosilicate glass solidified body according to claim 1, wherein the particle size of the glass beads in the step S2 is 1 to 3mm, and the composition comprises 50 to 53.44w% of SiO 2 12.5-14.6 w% of B 2 O 3 4.25-5.21 w% of Na 2 O, 2.21-3.16 w% Li 2 O、4.4～6.12w％Al of (2) 2 O 3 CaO 2.21-3.42w%, mgO 1.05-2.10 w%.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202211340346.7A CN115806386A (en) | 2022-10-28 | 2022-10-28 | Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202211340346.7A CN115806386A (en) | 2022-10-28 | 2022-10-28 | Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CN115806386A true CN115806386A (en) | 2023-03-17 |
Family
ID=85482959
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202211340346.7A Pending CN115806386A (en) | 2022-10-28 | 2022-10-28 | Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN115806386A (en) |
Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN107622806A (en) * | 2017-09-22 | 2018-01-23 | 绵阳科大久创科技有限公司 | A kind of high activity liquid waste glass solidification system and its curing |
| CN109273130A (en) * | 2018-08-07 | 2019-01-25 | 西南科技大学 | A kind of preparation method of the high sodium high activity liquid waste glass ceramics solidified body of high-sulfur |
| CN112624608A (en) * | 2020-12-18 | 2021-04-09 | 中国建筑材料科学研究总院有限公司 | Borosilicate glass curing substrate and preparation method and application thereof |
-
2022
- 2022-10-28 CN CN202211340346.7A patent/CN115806386A/en active Pending
Patent Citations (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN107622806A (en) * | 2017-09-22 | 2018-01-23 | 绵阳科大久创科技有限公司 | A kind of high activity liquid waste glass solidification system and its curing |
| CN109273130A (en) * | 2018-08-07 | 2019-01-25 | 西南科技大学 | A kind of preparation method of the high sodium high activity liquid waste glass ceramics solidified body of high-sulfur |
| CN112624608A (en) * | 2020-12-18 | 2021-04-09 | 中国建筑材料科学研究总院有限公司 | Borosilicate glass curing substrate and preparation method and application thereof |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| US4097401A (en) | Thermodynamically stable product for permanent storage and disposal of highly radioactive liquid wastes | |
| JP5768977B2 (en) | Alumino-borosilicate glass for containment of radioactive liquid waste, and method for treating radioactive liquid waste | |
| US4514329A (en) | Process for vitrifying liquid radioactive waste | |
| US4772431A (en) | Process for the immobilization of nuclear waste in a borosilicate glass | |
| US4797232A (en) | Process for the preparation of a borosilicate glass containing nuclear waste | |
| CN109273130B (en) | Preparation method of high-sulfur high-sodium high-emission waste liquid glass ceramic solidified body | |
| WO2020000097A1 (en) | Increased molybdenum and sulfur solubility in aluminoborosilicate glasses with added phosphorus | |
| CN115806386A (en) | Method for preparing high-sulfur high-radioactivity waste liquid borosilicate glass solidified body | |
| CN114180834B (en) | Iron-containing low-phosphate glass, preparation method and application thereof | |
| CN115583796B (en) | Glass solidified body for improving molybdenum element solubility and preparation method thereof | |
| WO2010150822A1 (en) | Sorption and filtration material | |
| CN114105472B (en) | Iron-containing high-phosphate glass, preparation method and application thereof | |
| JP3864203B2 (en) | Solidification method for radioactive waste | |
| JPS5858499A (en) | Method of processing radioactive waste liquid | |
| CN112645608B (en) | Silicon-free salt bath purification additive material and application method thereof | |
| CN111916246B (en) | Synergistic curing SrF 2 And glass solidified body of high-sodium high-iron waste material and preparation method thereof | |
| US7241932B2 (en) | Encapsulation of radioactive waste using a sodium silicate based glass matrix | |
| Rudolph et al. | Lab-scale R+ D work on fission product solidification by vitrification and thermite processes | |
| CN109721242B (en) | Low-melting-point glass for curing volatile nuclide Tc/Re and preparation and use methods thereof | |
| KR101309579B1 (en) | Treatment method of mercury-containing waste | |
| Holton et al. | Alternatives for vitrification of existing commercial high-level waste by spray calcination/in-can melting | |
| RU2232440C2 (en) | Monolithic silicate glass block for immobilizing liquid radioactive wastes and its manufacturing process | |
| Hamodi et al. | Immobilization of spent ion exchange resin arising from nuclear power plants: an introduction | |
| Kobelev et al. | Full‐Scale Cold Crucible Test on Vitrification of Savannah River Site SB4 HLW Surrogate | |
| JPS6231318B2 (en) |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| PB01 | Publication | ||
| PB01 | Publication | ||
| SE01 | Entry into force of request for substantive examination | ||
| SE01 | Entry into force of request for substantive examination |