乏燃料贮运用铝基碳化硼复合材料的屏蔽性能计算

采用蒙特卡罗方法, 利用MCNP程序计算了在中子能量为0.5–20 MeV, ²³⁵U核热中子裂变源条件下, 厚度为3–9 cm、碳化硼含量5%–15%的铝基碳化硼复合材料在空气、水、200–1400 ppm (1 ppm=10^-6) 硼酸溶液介质中的中子透射系数. 结果表明: B₄C/Al复合材料的透射系数随碳化硼含量和材料厚度的增加而减少, 随中子能量的升高而增大, 而硼酸浓度的改变对中子透射系数影响不大. B₄C/Al复合材料在水中比硼酸中更能发挥其屏蔽效果, 在空气中屏蔽特性显现出“反转”现象, 中子能量高于5 MeV时透射系数几乎没有变化. 在裂变源条件下的B₄C/Al复合材料中子透射系数均比稳定源20 MeV 低. 介质的中子屏蔽效果是硼酸溶液>水> 空气, 水介质的中子透射系数与介质厚度呈指数下降关系, 裂变源和稳定源条件下分别近似为e^-0.71x和e^-0.669x, x为厚度(cm). 关键词： 蒙特卡罗 乏燃料设备 中子吸收材料 4C/Al')" href="#">B₄C/Al

Shielding property calculation of B4C/Al composites for spent fuel transportation and storge

MCNP program is used to calculate the neutron transmission coefficient of 3–9 cm-thick neutron absorber material B₄C/Al composite with 5%–15% B₄C content in air, water, 200–1400 ppm (1 ppm=10^-6) H₃BO₃ solution, irradiated by 0.5–20 MeV neutrons and ²³⁵U thermal neutron fission source. The results show that the transmission coefficient of B₄C/Al composite decreases with the increase of the content of B₄C and the thickness of material, but increases with the increase of neutron energy, and has little influence from the variation in H₃BO₃ solution concentration. A better shielding effect of B₄C/Al composite is displayed in water than in H₃BO₃ solution, and a “reversal” phenomenon of the shielding effect occurs in air. The neutron transmission coefficient is almost unchanged with neutron energy when neutron energy is higher than the 5–15 MeV. The neutron transmission coefficient of B₄C/Al composite irradiated under a fission source is lower than under a steady 20 MeV neutron source. Ranking the shielding performances of media, the sequence is H₃BO₃ solution > water > air, and the exponential decay relationships between neutron transmission coefficient and thickness of medium can be expressed as e^-0.71x and e^-0.669x, where x is thickness of medium in cm.
Keywords： Monte-Carlo spent fuel apparatus neutron absorber material 4C/Al')" href="#">B₄C/Al