核电站中乏燃料储存格架用到的中子吸收材料需要兼具结构和功能一体化的要求,本文提出用碳纤维Cf增强B4C/Al中子吸收复合材料。利用Monte Carlo方法对碳纤维增强铝基碳化硼中子吸收材料(Cf/B4C/Al)的中子透射率进行模拟计算,研究B4C含量...核电站中乏燃料储存格架用到的中子吸收材料需要兼具结构和功能一体化的要求,本文提出用碳纤维Cf增强B4C/Al中子吸收复合材料。利用Monte Carlo方法对碳纤维增强铝基碳化硼中子吸收材料(Cf/B4C/Al)的中子透射率进行模拟计算,研究B4C含量、Cf含量、不同能量中子入射以及材料厚度变化时对中子透射率的影响,并与B4C/Al材料进行比较。结果表明,在1 e V-0.1 Me V能量范围的中子入射下,当B4C含量小于35%时,加入碳纤维能明显改善B4C/Al材料的中子屏蔽性能;在100 e V中子入射下,材料的中子透射率随B4C含量增加呈现指数下降;且Cf/B4C/Al材料的中子透射率随碳纤维含量增加持续降低;当Cf含量达到10%时,材料中子透射率降至最低,之后趋于平稳。通过模拟计算,得到Cf/B4C/Al材料的各组分的最优配比为35 vol.%B4C和10 vol.%Cf。展开更多
使用MCNP5程序模拟了能量为14.88 Me V的快中子在(W+B_4C)/Al和W/Al复合材料中的输运过程,并与铅硼聚乙烯、聚乙烯以及钨进行对比。计算了各屏蔽材料的中子衰减性能、中子透射能谱以及中子俘获过程中释放γ射线的透射能谱,为中子屏蔽材...使用MCNP5程序模拟了能量为14.88 Me V的快中子在(W+B_4C)/Al和W/Al复合材料中的输运过程,并与铅硼聚乙烯、聚乙烯以及钨进行对比。计算了各屏蔽材料的中子衰减性能、中子透射能谱以及中子俘获过程中释放γ射线的透射能谱,为中子屏蔽材料的选择提供了理论依据。模拟结果和实际结果吻合,证实了蒙特卡罗方法的可靠性。模拟结果也可以得出,对于快中子,高原子序数材料的屏蔽效果要好于低原子序数材料,含钨45%的复合材料的屏蔽性能与商用铅硼聚乙烯的屏蔽效能相近,但是激发γ射线能量低于铅硼聚乙烯,考虑到成分可调控性、使用温度以及力学性能等因素,(W+B_4C)/Al复合材料是一种极具应用潜力的新型中子屏蔽材料。展开更多
This study is a comparison of gamma ray linear attenuation coefficient of two typs of shielding materials made of Saudi white and red sand. Each shield was consisted of one part of cement two parts of sand in addi-tio...This study is a comparison of gamma ray linear attenuation coefficient of two typs of shielding materials made of Saudi white and red sand. Each shield was consisted of one part of cement two parts of sand in addi-tion to water. Different thicknesses were tested. The concentrations of all elements in each shield material were determined by Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The results obtained from the ICP-MS were used in MCNP4B (Monte Carlo N-Particle Transport Computer Code System) [1] to calculate the attenuation coefficient. The theoretical (MCNP4B) and the experimental calculations were found to be in a good agreement. In the casw of the largest thickness used, 28cm, the gamma ray intensity passing through the white sand shield was approximately half of the intensity obtained through the red sand shield. The average linear attenuation coefficients were found to be 0.17cm-1 and 0.15cm-1 for white and red sand shields respectively. The study shows that white sand is better for attenuating gamma ray compared to the red sand.展开更多
文摘核电站中乏燃料储存格架用到的中子吸收材料需要兼具结构和功能一体化的要求,本文提出用碳纤维Cf增强B4C/Al中子吸收复合材料。利用Monte Carlo方法对碳纤维增强铝基碳化硼中子吸收材料(Cf/B4C/Al)的中子透射率进行模拟计算,研究B4C含量、Cf含量、不同能量中子入射以及材料厚度变化时对中子透射率的影响,并与B4C/Al材料进行比较。结果表明,在1 e V-0.1 Me V能量范围的中子入射下,当B4C含量小于35%时,加入碳纤维能明显改善B4C/Al材料的中子屏蔽性能;在100 e V中子入射下,材料的中子透射率随B4C含量增加呈现指数下降;且Cf/B4C/Al材料的中子透射率随碳纤维含量增加持续降低;当Cf含量达到10%时,材料中子透射率降至最低,之后趋于平稳。通过模拟计算,得到Cf/B4C/Al材料的各组分的最优配比为35 vol.%B4C和10 vol.%Cf。
文摘使用MCNP5程序模拟了能量为14.88 Me V的快中子在(W+B_4C)/Al和W/Al复合材料中的输运过程,并与铅硼聚乙烯、聚乙烯以及钨进行对比。计算了各屏蔽材料的中子衰减性能、中子透射能谱以及中子俘获过程中释放γ射线的透射能谱,为中子屏蔽材料的选择提供了理论依据。模拟结果和实际结果吻合,证实了蒙特卡罗方法的可靠性。模拟结果也可以得出,对于快中子,高原子序数材料的屏蔽效果要好于低原子序数材料,含钨45%的复合材料的屏蔽性能与商用铅硼聚乙烯的屏蔽效能相近,但是激发γ射线能量低于铅硼聚乙烯,考虑到成分可调控性、使用温度以及力学性能等因素,(W+B_4C)/Al复合材料是一种极具应用潜力的新型中子屏蔽材料。
文摘This study is a comparison of gamma ray linear attenuation coefficient of two typs of shielding materials made of Saudi white and red sand. Each shield was consisted of one part of cement two parts of sand in addi-tion to water. Different thicknesses were tested. The concentrations of all elements in each shield material were determined by Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The results obtained from the ICP-MS were used in MCNP4B (Monte Carlo N-Particle Transport Computer Code System) [1] to calculate the attenuation coefficient. The theoretical (MCNP4B) and the experimental calculations were found to be in a good agreement. In the casw of the largest thickness used, 28cm, the gamma ray intensity passing through the white sand shield was approximately half of the intensity obtained through the red sand shield. The average linear attenuation coefficients were found to be 0.17cm-1 and 0.15cm-1 for white and red sand shields respectively. The study shows that white sand is better for attenuating gamma ray compared to the red sand.
