CN108335771B - Neutron shielding material and preparation method thereof - Google Patents
Neutron shielding material and preparation method thereof Download PDFInfo
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- CN108335771B CN108335771B CN201711442865.3A CN201711442865A CN108335771B CN 108335771 B CN108335771 B CN 108335771B CN 201711442865 A CN201711442865 A CN 201711442865A CN 108335771 B CN108335771 B CN 108335771B
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- shielding material
- flame retardant
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F1/00—Shielding characterised by the composition of the materials
- G21F1/02—Selection of uniform shielding materials
- G21F1/10—Organic substances; Dispersions in organic carriers
- G21F1/103—Dispersions in organic carriers
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F1/00—Shielding characterised by the composition of the materials
- G21F1/02—Selection of uniform shielding materials
- G21F1/10—Organic substances; Dispersions in organic carriers
- G21F1/103—Dispersions in organic carriers
- G21F1/106—Dispersions in organic carriers metallic dispersions
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Abstract
The invention provides a neutron shielding material and a preparation method thereof, and the neutron shielding material comprises 80-120 parts of epoxy resin, 1-90 parts of curing agent, 180 parts of flame retardant 130-. Through changing the median particle diameter of fire retardant and density increasing agent, avoid being drawn into gas at the in-process of pouring, control curing temperature and curing time for the solidification rate of this mixed system, improve the solidification effect, guarantee that neutron shielding material has good fire resistance, the heat conductivity, mechanical properties and inside gas pocket are few etc. good performance, it is relatively poor to have solved current neutron shielding material heat conduction and fire behaviour, the formability is poor, the viscosity is not good, be unsuitable for pouring scheduling problem, improve neutron radiation's shielding effect.
Description
Technical Field
The invention relates to a material for shielding spent fuel generated by a nuclear power station, in particular to a neutron shielding material and a preparation method thereof.
Background
Neutrons are one of the basic particles constituting atomic nuclei, and are uncharged, but because of their high speed, they can ionize substances to generate radioactive radiation, and are widely used in the fields of national defense, scientific research, radiotherapy and detection. Neutrons are indispensable in developing nuclear power by nuclear fission, and provide a large amount of clean energy for human beings. However, a large amount of secondary particles generated by neutron ionization interact with tissue cells to cause great harm to human health, and meanwhile, point defects and dislocation can be generated inside the material by neutron irradiation, so that the material performance is degraded. After the occurrence of the nuclear accident of the Japanese Fudao, the safety problem of nuclear energy is widely concerned by the society, and the shielding material which is light, convenient, efficient and excellent in comprehensive performance has important significance for the safe operation of nuclear power equipment. Along with the operation of more and more nuclear power generating units in China, a large number of spent fuel assemblies are generated, and the spent fuel assemblies need to be subjected to spent fuel post-treatment through storage, transportation and transfer. In the process of storage and transportation, it is very important to contain radioactive substances such as neutrons, gamma rays and the like. The spent fuel has a certain radioactivity and continuously generates decay heat. In order to ensure the safety of the container in the use process of the spent fuel storage and transportation container, the heat needs to be discharged in time, so that the radiation protection material is required to have good heat conduction capability, and meanwhile, in order to prevent the damage of fire accidents to the radioactive containment, the radiation protection material is required to have good flame retardant property. However, the existing shielding material has the conditions of poor heat conduction and flame retardant property, poor forming property, poor viscosity, unsuitability for casting and the like.
Therefore, the formula of the shielding material needs to be improved to obtain a neutron shielding material with excellent performance.
Disclosure of Invention
The invention aims to provide a neutron shielding material and a preparation method thereof to overcome the problems of poor heat conduction and flame retardant properties, poor formability, poor viscosity, unsuitability for casting and the like.
In order to achieve the purpose, the invention provides a neutron shielding material which comprises, by weight, 80-120 parts of epoxy resin, 1-90 parts of curing agent, 180 parts of flame retardant 130-one, 5-20 parts of density increasing agent and 1-5 parts of neutron absorber, wherein the median particle size of the flame retardant is 0.5-30 μm, and the median particle size of the density increasing agent is 1-10 μm.
The invention also provides a preparation method of the neutron shielding material, which comprises the following steps:
(1) mixing the epoxy resin and part of the flame retardant in vacuum;
(2) adding the other part of flame retardant, density increaser and neutron absorber into the mixture obtained in the step (1), and carrying out vacuum mixing and defoaming treatment;
(3) adding a curing agent into the mixture obtained in the step (2) for vacuum mixing to obtain a mixture;
(4) pouring the mixture in the step (3);
(5) and (4) carrying out curing treatment.
Compared with the prior art, the neutron shielding material provided by the invention comprises 80-120 parts of epoxy resin, 1-90 parts of curing agent, 180 parts of flame retardant 130-, wherein the median particle diameter of the flame retardant is 0.5-30 μm, the median particle diameter of the density increasing agent is 1-10 μm, for the neutron shielding material, the median particle size of the flame retardant is in a certain range, the larger the median particle size of the flame retardant is, the more the additive amount is, the larger the volume fraction is, the higher the thermal conductivity coefficient of the neutron shielding material is, the smaller the median particle size of the flame retardant is, the better the flame retardant property of the system is, 180 portions of flame retardant 130-; on the premise of ensuring that the density meets the requirement, the density increaser can adjust the uniformity, viscosity and curing process of the mixed system by changing the content and the median particle diameter, so that the viscosity of the system is kept in a proper range, gas is prevented from being involved in the pouring process, the curing temperature and the curing time can be controlled, the curing speed of the mixed system is accelerated, the curing effect is improved, and the neutron shielding material is ensured to have good flame retardance, thermal conductivity, mechanical property, less internal pores and other excellent properties.
In the preparation method, the two-time addition of the flame retardant is beneficial to reducing the agglomeration phenomenon, increasing the uniformity of the mixture and improving the heat conductivity coefficient and the flame retardant property of the neutron shielding material.
Detailed Description
In order to explain the technical contents and structural features of the present invention in detail, the following description is given with reference to the embodiments.
The neutron shielding material comprises, by weight, 80-120 parts of epoxy resin, 1-90 parts of curing agent, 180 parts of flame retardant 130-one, 5-20 parts of density increaser and 1-5 parts of neutron absorber, wherein the median particle size of the flame retardant is 0.5-30 mu m, the median particle size of the density increaser is 1-10 mu m, the median particle size of the flame retardant can be 0.5-5 mu m, 9 mu m, 12 mu m, 15 mu m, 20 mu m, 25 mu m and 30 mu m, and the median particle size of the density increaser can be 1-3 mu m, 5 mu m, 7 mu m, 9 mu m and 10 mu m.
Preferably, the curing agent is selected from one or two of amine curing agent and anhydride curing agent, wherein the amine curing agent is preferably modified cyclic amine, diethylenetriamine and triethylene tetramine, and the anhydride curing agent is preferably methyl tetrahydrophthalic anhydride, methyl hexahydrophthalic anhydride and methyl nadic anhydride.
Preferably, the curing agent further comprises an imidazole curing accelerator, the imidazole curing accelerator is selected from one or more of 2-methylimidazole, 1-benzyl-2-ethylimidazole, 2-ethyl-4-methylimidazole and 1-aminoethyl-2-methylimidazole, the imidazole curing accelerator is selected from 2-methylimidazole, 1-benzyl-2-ethylimidazole, 2-ethyl-4-methylimidazole and 1-aminoethyl-2-methylimidazole, an imidazole substance is added to promote curing of the epoxy resin-curing agent system, the gel curing time of the mixture is properly shortened, sedimentation of the solid mixture in a liquid state is reduced, component uniformity of the neutron shielding material is improved, and the imidazole substance can also improve the heat-resistant long-term performance of the neutron shielding material through a molecule which forms a cross-linked structure through reaction with the epoxy resin.
Preferably, the epoxy resin is selected from one or more of bisphenol a epoxy resin, hydrogenated bisphenol a epoxy resin, cycloaliphatic epoxy resin, novolac epoxy resin, and triglycidyl para-aminophenol.
Preferably, the hydrogen content in the epoxy resin is 7% -11%, the hydrogen content in the curing agent is 6.5% -13%, and the neutron retarding effect is improved.
Preferably, the flame retardant is selected from one or more of nitrogen-containing intumescent flame retardants, magnesium hydroxide, calcium hydroxide, aluminium hydroxide, zinc borate and nickel hydroxide.
Preferably, the neutron absorber is selected from one or more of gadolinium oxide, dysprosium titanate, boron carbide, zirconium diboride and titanium diboride, and the median particle size of the neutron absorber may be 1 μm, 3 μm, 5 μm, 7 μm, 9 μm, 10 μm.
Preferably, the density increasing agent is selected from one or more of tungsten powder, lead powder, copper powder, iron powder, tungsten oxide, lead oxide, iron oxide, zinc zirconium, zinc powder, zirconium oxide powder and zinc oxide powder.
The preparation method for preparing the neutron shielding material comprises the following steps:
(1) mixing the epoxy resin and part of the flame retardant in vacuum, wherein the vacuum mixing time is selected from 10-30min according to specific conditions;
(2) adding another part of flame retardant, density increaser and neutron absorber in the step (1), and carrying out vacuum mixing and defoaming treatment, wherein the defoaming treatment time and the mixing temperature are respectively selected from 2-5h and 30-50 ℃ according to specific conditions;
(3) adding a curing agent into the mixture obtained in the step (2) for vacuum mixing to obtain a mixture;
(4) pouring the mixture in the step (3) into a vacuum mold;
(5) and (4) putting the vacuum mould into an oven for curing treatment, wherein the curing temperature and the curing time are respectively selected from 30-50 ℃ and 24-28h according to specific conditions.
The neutron shielding material of the present invention will be described in detail below with reference to examples.
The components, parts by weight and median particle size for preparing the neutron shielding material are shown in the table below, and all the raw materials in this example are commercially available products.
Weighing the raw materials according to the weight parts and the median particle diameters of the epoxy resin, the curing agent, the flame retardant, the neutron absorber and the density enhancer of examples 1 to 5 and comparative examples 1 to 3 of the neutron shielding material in the table 1, and preparing the neutron shielding material according to the following steps:
(1) mixing the epoxy resin and part of the flame retardant (half of the flame retardant) in vacuum for 10 min;
(2) adding the other part of flame retardant, density increaser and neutron absorber into the mixture obtained in the step (1), and carrying out vacuum mixing and defoaming treatment, wherein the defoaming treatment time and the mixing temperature are respectively 5h and 50 ℃;
(3) adding a curing agent into the mixture obtained in the step (2) for vacuum mixing to obtain a mixture;
(4) pouring the mixture in the step (3) into a vacuum mold;
(5) and (4) placing the vacuum mold into an oven for curing treatment, wherein the curing temperature and the curing time are respectively 50 ℃ and 28 h.
TABLE 1 neutron shielding material composition, parts by weight, median particle size
The neutron shielding materials prepared in examples 1-5 and comparative examples 1-3 are tested for thermal conductivity and flame retardant property, and the test method is as follows:
heat conductivity coefficient testing method
λ ═ ρ · c · α, where: λ is the thermal conductivity, in units of W/(m.K); rho is the bulk density in kg.m-3(ii) a c is specific heat capacity, and the unit is J/(kg. K); alpha is the thermal diffusion coefficient and is given in m2/s。
The heat conductivity coefficient lambda is obtained by detecting the volume density, the thermal diffusion coefficient and the specific heat capacity of the sample and calculating.
(the volume density is measured by an Archimedes drainage method, the specific heat capacity is measured by a DSC sapphire method, and the thermal diffusivity is measured by a laser flash method.)
Flame retardant property test method
The flame-retardant property of the material is expressed by oxygen index, which means that a material (test piece) with a certain size is put into a test device, mixed gas of oxygen and nitrogen is introduced under specified conditions, the test piece is ignited by an igniter, and the minimum oxygen concentration (expressed by percentage) necessary for maintaining continuous combustion like wax is measured. An oxygen index higher than 21% means that the material is not easily combustible, and an oxygen index lower than 21% means that the material is easily combustible.
After the neutron shielding material is subjected to the test of the heat conductivity coefficient and the flame retardant property, the test result is shown in table 2:
TABLE 2 thermal conductivity and flame retardancy of examples 1 to 5 and comparative examples 1 to 3
As can be seen from the comparison of the thermal conductivity and the flame retardant performance of the examples 1-5 and the comparative examples 1-3 in the table 2 above, the thermal conductivity and the flame retardant performance of the examples 1-5 are better than those of the comparative examples 1-2, because for the neutron shielding material, the median particle size of the flame retardant is within a certain range, the larger the median particle size of the flame retardant is, the more the additive amount is, the larger the additive volume fraction is, the higher the thermal conductivity of the neutron shielding material is, the smaller the median particle size of the flame retardant is, the better the flame retardant performance of the system is, and when the median particle size is 0.5 μm-30 μm in the neutron shielding material, the best flame retardant performance and thermal conductivity can be achieved; on the premise of ensuring that the density meets the requirement, the density increaser can adjust the uniformity, viscosity and curing process of the mixed system by changing the content and the median particle diameter, so that the viscosity of the system is kept in a proper range, gas is prevented from being involved in the pouring process, the curing temperature and the curing time can be controlled, the curing speed of the mixed system is accelerated, the curing effect is improved, and the prepared neutron shielding material has good flame retardance, thermal conductivity, mechanical property, less internal pores and other excellent properties.
The above disclosure is only a preferred embodiment of the present application and should not be taken as limiting the scope of the present application, so that the claims of the present application are covered by the appended claims.
Claims (7)
1. A preparation method of a neutron shielding material of the neutron shielding material is characterized by comprising the following steps:
(1) mixing 80-120 parts of epoxy resin and part of flame retardant in vacuum;
(2) adding another part of the flame retardant, 5-20 parts of density increaser with the median particle size of 1-10 mu m and 1-5 parts of neutron absorber into the mixture obtained in the step (1), and performing vacuum mixing and defoaming treatment, wherein the total part of the flame retardant and the other part of the flame retardant is 130-180 parts, and the median particle size of the flame retardant is 0.5 mu m;
(3) adding 1-90 parts of curing agent into the mixture obtained in the step (2) to carry out vacuum mixing to obtain a mixture;
(4) pouring the mixture in (3);
(5) carrying out curing treatment after the step (4);
the flame retardant is selected from one or more of nitrogen-containing intumescent flame retardants, magnesium hydroxide, calcium hydroxide, aluminum hydroxide, zinc borate and nickel hydroxide, the epoxy resin is selected from one or more of bisphenol A epoxy resin, hydrogenated bisphenol A epoxy resin, alicyclic epoxy resin, novolac epoxy resin and triglycidyl p-aminophenol, and the hydrogen content in the epoxy resin is 7-11%;
the contents of the above substances are all calculated by weight parts.
2. The method for preparing the neutron shielding material according to claim 1, wherein the curing agent is one or two selected from amine curing agents and anhydride curing agents.
3. The method for preparing the neutron shielding material according to claim 2, wherein the curing agent further comprises an imidazole curing accelerator, and the imidazole curing accelerator is one or more selected from 2-methylimidazole, 1-benzyl-2-ethylimidazole, 2-ethyl-4-methylimidazole and 1-aminoethyl-2-methylimidazole.
4. The method for preparing the neutron shielding material according to claim 1, wherein the curing agent has a hydrogen content of 6.5% -13%.
5. The method of claim 1, wherein the neutron absorber is selected from one or more of gadolinium oxide, dysprosium titanate, boron carbide, zirconium diboride, and titanium diboride.
6. The method of producing the neutron shielding material of claim 1, wherein the neutron absorber has a median particle size of 1 μm to 10 μm.
7. The method of claim 1, wherein the density increasing agent is selected from one or more of tungsten powder, lead powder, copper powder, iron powder, tungsten oxide, lead oxide, iron oxide, zinc zirconium, zinc powder, zirconium oxide powder, and zinc oxide powder.
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| Publication number | Priority date | Publication date | Assignee | Title |
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| CN109545415A (en) * | 2018-11-12 | 2019-03-29 | 东莞理工学院 | A kind of radiation protection material |
| CN109575606A (en) * | 2019-01-16 | 2019-04-05 | 东莞理工学院 | A kind of neutron shielding material and preparation method thereof |
| CN110965004B (en) * | 2019-09-04 | 2022-05-24 | 扬州斯帕克实业有限公司 | Preparation method of high-temperature corrosion resistant Mg-Pb-Al-B nuclear shielding material |
| CN110619969B (en) * | 2019-09-23 | 2022-10-21 | 中国核动力研究设计院 | Radiation shielding container and preparation method thereof |
| CN111430057B (en) * | 2020-03-18 | 2021-06-08 | 张云逢 | High radioactive nuclear waste container |
| CN113583385A (en) * | 2021-08-03 | 2021-11-02 | 中国核动力研究设计院 | High-density resin-based shielding plate and preparation method thereof |
| CN117049880A (en) * | 2023-08-14 | 2023-11-14 | 西安交通大学 | High-entropy carbide ceramic for spent fuel shielding material and preparation method thereof |
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