基于MC法的乏燃料贮存用铝基复合材料屏蔽性能研究

发布时间：2018-06-02 03:39

本文选题：乏燃料贮存 + 中子吸收性能　；参考：《太原理工大学》2015年硕士论文

【摘要】：随着我国核电产业的快速发展，核电站乏燃料的持续安全贮存成为核电技术自主化过程中亟待解决的关键性问题。乏燃料贮存用辐射屏蔽材料需同时满足较高的结构性能、腐蚀性能和屏蔽性能要求。本文基于MonteCarlo方法，以6061铝合金为载体，通过添加碳化硼作为中子吸收组元、铅作为γ辐射屏蔽组元，为设计结构/功能一体化的新型铝基乏燃料贮存用辐射屏蔽材料提供理论依据。本文采用MCNP5.0和Super MC/MCAM软件，研究了碳化硼颗粒增强铝基复合材料的组分配比、厚度、增强相尺寸与中子透射系数的关系，分析了铝基碳化硼材料与中子的相互作用机制；研究了铅-碳化硼铝基复合材料的组分配比、射线能量与中子透射系数、次级γ射线和衰变γ射线吸收率的关系；研究了贮存乏燃料的硼酸溶液硼浓度和乏燃料组件密集化程度对辐射屏蔽性能造成的影响；分析了在乏燃料中235U衰变条件下，铝基辐射屏蔽材料的屏蔽效果。结果表明：由于中子吸收核素10B的存在，B4C含量对中子吸收性能影响很大、使次级γ射线吸收率微弱降低；由于其密度较小使材料整体密度降低，会导致一次γ射线的吸收率略有下降；随着微米级B4C颗粒存在使中子透射系数提高11%以上；B4C颗粒形状改变引起的中子在材料内穿行距离的变化也会对中子吸收性能造成影响。由于γ射线吸收核素Pb具有较大的原子序数、较高γ射线的质量吸收系数和高密度，其含量提高会使铝基屏蔽材料的中子吸收性能有所提高、对次级γ射线与的吸收能力均显著增强。2cm厚、含铅量为25%的铝基复合材料的一次γ射线屏蔽能力与0.5cm铅板相当。入射粒子的能量直接决定着材料的辐射屏蔽效果，这是由于材料的宏观吸收截面通常随射线能量增大而减小。当入射中子的能量低于200eV时，中子吸收性能与次级γ射线屏蔽性能随能量减小而显著提高；当一次γ射线的能量低于660KeV时，对其的屏蔽性能随能量减小而显著提高。由于乏燃料贮存介质硼酸溶液对入射高能中子先“慢化”后“热化”的作用，，铝基屏蔽材料的中子吸收性能随硼原子浓度提高而降低、随组件栅距增大而减小；次级γ射线屏蔽性能随硼原子浓度提高而降低、随组件栅距增大而减小；贮存介质变化对材料吸收乏燃料衰变产生的γ射线的吸收影响不大。在乏燃料组件密集化贮存（组件栅距23cm）、屏蔽层厚度为0.7cm的条件下，乏燃料235U衰变中子经硼浓度2500ppm的硼酸溶液环境充分慢化、热化后，B4C wt.%为30%的B4C/Al复合材料作为屏蔽层时，可以吸收98.04%的中子、45.44%的次级γ射线和20.21%的一次γ射线；B4Cwt.%为30%、Pb wt.%为25%的Pb-B4C/Al复合材料作为屏蔽层时，可以吸收98.82%的中子、61.05%的次级γ射线和47.08%的一次γ射线。
[Abstract]:With the rapid development of nuclear power industry in China, the continuous and safe storage of spent fuel in nuclear power plant has become a key problem to be solved urgently in the process of nuclear power technology autonomy. The radiation shielding material for spent fuel storage needs to meet high structural performance, corrosion performance and shielding performance. Based on the MonteCarlo method, 6061 aluminum alloy is used in this paper. As a carrier, by adding boron carbide as a neutron absorption component and lead as a shielding element of gamma radiation, it provides a theoretical basis for the design of a new type of aluminum based spent fuel storage shielding materials for the design of structural / functional integration.
In this paper, MCNP5.0 and Super MC/MCAM software were used to study the relationship between the group distribution ratio, thickness, the thickness, the phase size of the reinforced aluminum matrix composite and the neutron transmission coefficient, and the interaction mechanism of the aluminum based boron carbide material with the neutron, and the group distribution ratio, the ray energy and the neutron energy of the lead carbon boric aluminum matrix composites were studied. The relationship between the transmission coefficient, the secondary gamma ray and the absorption rate of decay gamma ray, the effect of boron concentration in the boric acid solution and the intensity of the spent fuel assembly on the radiation shielding performance of the spent fuel were studied, and the shielding effect of the aluminum based radiation shielding material under the 235U decay condition in the spent fuel was analyzed.
The results show that because of the existence of the neutron absorption nuclide 10B, the B4C content has a great influence on the absorption properties of the neutron, which makes the secondary gamma ray absorptivity weakly reduced. Because of its smaller density, the overall density of the material decreases, which will lead to a slight decrease in the absorption rate of the primary gamma ray, and the neutron transmission coefficient increases by 11% with the existence of the micron grade B4C particles. On the other hand, the variation of the neutron penetration distance caused by the change of B4C particle shape will also affect the neutron absorption performance.
Because of the large atomic number and high mass absorption coefficient and high density of gamma ray absorption nuclide Pb, the neutron absorption properties of the aluminum based shielding material will be improved and the absorption capacity of the secondary gamma ray and the absorptive capacity of the secondary gamma ray increases significantly, and the primary gamma ray shielding energy of the aluminum matrix composites with 25% lead content is 25%. The force is equivalent to the 0.5cm lead plate.
The energy of the incident particle is directly determined by the radiation shielding effect of the material, because the macroscopic absorption cross section of the material decreases with the increase of the ray energy. When the energy of the incident neutron is lower than 200eV, the absorption and secondary gamma ray shielding performance of the neutron is significantly increased with the decrease of energy; when the energy of the primary gamma ray is lower than 660KeV The shielding performance of the device increases with the decrease of energy.
The neutron absorption performance of the aluminum based shielding material decreases with the increase of boron concentration and decreases with the increase of the component spacing, and the secondary gamma ray shielding performance decreases with the increase of boron concentration, and decreases with the increase of the component spacing. The change of storage medium has little effect on the absorption of gamma ray produced by the absorption of spent fuel decay. Under the condition of the dense storage of the spent fuel assembly (component distance 23cm) and the thickness of the shielding layer of 0.7cm, the spent fuel 235U decay neutrons are fully slowed by the boric acid solution environment of boron concentration 2500ppm, and the B4C wt.% is 30% of the B4C/Al composite after the heating. As a shielding layer, 98.04% neutrons, 45.44% secondary gamma rays and 20.21% gamma rays can be absorbed; when B4Cwt.% is 30% and Pb wt.% is 25% Pb-B4C/Al composite material as shielding layer, it can absorb 98.82% neutrons, 61.05% secondary gamma rays and 47.08% gamma rays.
【学位授予单位】：太原理工大学
【学位级别】：硕士
【学位授予年份】：2015
【分类号】：TB333

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