利用MCNP模拟气体裂变产物混合源的γ剂量率

Simulation of Gamma Dose Rate of Complicated Fission Gas by MCNP Method

摘要: 西安脉冲反应堆辐照铀靶后，抽取Kr、Xe裂变气体，通过活性炭吸附于气体源盒内。HPGe γ谱仪测量源盒内混合气体活度，塑料闪烁探测器测量γ剂量率。将源盒、塑料闪烁探测器的几何结构、材料作为蒙特卡罗程序（MCNP）输入信息，模拟塑料闪烁探测器对源盒中核素活度与其γ剂量率对应关系，结合HPGe γ谱仪所测活度得到剂量率模拟值，结果与实测值偏差小于6%。该工作说明在已知放射源空间结构、放射性核素种类和活度的情况下，采用MCNP模拟计算复杂气体放射源γ剂量率的方法是可行的。

Abstract: Gamma dose rate of fission gas is calculated by MCNP method and compared with the data measured by plastic scintillation. The fission gas absorbed by active carbon in source vessel was produced by neutron irradiation of uranium in Xi’an Pulsed Reactor. The simulation model is composed of geometry and material of source and plastic scintillation detector as well as the gamma-ray energies and probabilities of ⁸⁵Kr^m, ⁸⁷Kr, ⁸⁸Kr, ¹³⁵Xe^m, ¹³⁵Xe, ¹³⁸Xe and ¹³⁸Cs whose activities were measured by an HPGe γ detector. The presented calculation shows agreement with experiments less than 6% which consequently confirms the reliability of the simulation for gamma dose rate of complicated radioactive gas.