无源符合中子计数器探测效率与泄漏增殖系数的模拟计算

Simulation Calculation of Detection Efficiency and Leakage Multiplication Coefficient of Passive Coincidence Neutron Counter

无源符合中子法通常用来确定被测源项中自发裂变的钚质量,需要事先对关键参数,如探测效率和增殖泄漏因子等,开展全面深入研究和计算,然后根据探测到的符合中子计数,利用相关公式计算得到钚含量。对探测器的效率进行模拟时,采用^(252)Cf标准源充当测量样本,自发裂变中子能谱接近纯裂变谱,因此适用于快中子的被动测量,即在“快中子的主动测量”模式的基础上,去除Am-Li中子源,保留聚乙烯慢化体和镉内衬等结构;在对泄漏增殖系数M进行模拟时,MCNP输入卡中的探测器模型和进行探测器效率模拟的“快中子的主动测量”模型相同,对结果的准确性并无影响。为了得到符合中子计数器的探测效率与泄漏增殖系数,通过MATLAB编写脚本,实现了针对AWCC中子计数器的MCNP自动建模功能,仅需输入结构参数,即可快速生产AWCC中子计数器的MCNP输入文件,实现相同结构、不同尺寸的AWCC中子计数器的蒙卡模拟,进行探测器效率、中子与3 He反应率以及测定含Pu废料时泄漏增殖系数M的计算。理论探测效率为24%,本次模拟计算出的探测器效率为23.27%,两者结果相差不大,可判定模拟结果可信度较高;泄漏增殖系数M的计算结果随材料质量的增大而增大,与参考文献[4]结果的变化趋势一致。本工作为探测器的实际应用奠定了理论基础。

The passive coincidence neutron method is usually used to determine the mass of plutonium spontaneously fission in the measured source term.It needs to conduct comprehensive and in-depth research and calculation on key parameters,such as detection efficiency and leakage multiplication coefficient,and then count the detected coincidence neutrons and calculate the plutonium content using relevant formulas.In this paper,while simulating the efficiency of the detector,^(252)Cf standard source was used as the measurement sample,and the spontaneous fission neutron energy spectrum was close to the pure fission spectrum.Therefore,it is suitable for passive measurement of fast neutrons,that is,on the basis of active measurement of fast neutrons,the Am-Li neutron source is removed,and the structures such as polyethylene moderator and cadmium liner are retained;while simulating the leakage multiplication coefficient M,the detector model in the MCNP input card is the same as the active measurement of fast neutrons model used to simulate the detector efficiency,which has no impact on the accuracy of the results.In order to obtain the detection efficiency and leakage multiplication coefficient of the coincidence neutron counter,the MCNP automatic modeling function for the AWCC neutron counter was realized by writing a script in MATLAB.The MCNP input file of the AWCC neutron counter can be quickly produced only by inputting structural parameters,and the simulation of the AWCC neutron counter with the same structure and different sizes can be realized to carry out the detector efficiency calculation of reaction rate between neutron and 3 He and leakage multiplication coefficient M in measurement of waste containing Pu.The theoretical detection efficiency was 24%,and the detector efficiency calculated by this simulation was 23.27%.There is little difference between the two results,so it can be judged that the reliability of the simulation results is high.The calculation results of the leakage multiplication coefficient M increase with the increase of the material quality,which is consistent with the change trend of the results in reference[4].This work will lay a theoretical foundation for the practical application of the detector.