燃耗截面基本库对输运燃耗耦合计算的影响分析

Impact analysis of basic burnup cross section on coupled calculation of neutron transport burnup

基于蒙特卡罗中子输运程序和ORIGEN2点燃耗程序的蒙特卡罗输运燃耗耦合计算方法应用广泛。但现有评价库中子连续截面的核素个数远小于燃耗计算涉及到的核素数量,即通过输运计算得到的燃耗截面不足以完全替代燃耗计算的基本库。采用经过栅元验证的蒙特卡罗燃耗程序MCBMPI,对最新的VERA燃耗计算基准题进行验证计算,对比分析不同的燃耗截面基本库对输运燃耗计算的影响。分析结果表明:1)在实际应用中尽量不要采用典型热中子截面库,会带来较大偏差;2)在燃耗计算核素替换较多的情况下,对该基准题而言,选取典型压水堆基本库还是典型快堆基本库,对结果影响不大,二者keff偏差在8‰以内,燃耗末期235U偏差在4‰以内,135Xe偏差在5‰左右;3)建议选取与研究对象能谱相近的基本库。

[Background]Monte Carlo depletion method is based on Monte Carlo neutron transport code and ORIGEN2 point depletion code,it is widely used for neutron transport burnup.However,the continuous energy neutron cross sections of the nuclides in the current evaluation library are less than the cross section used in the depletion calculation.Therefore,the depletion cross sections generated by Monte Carlo transport calculation cannot replace all the cross sections in the basic depletion library.[Purpose]This study aims to analyse the influence of different basic depletion libraries on coupled calculation of neutron transport burnup.[Methods]The Monte Carlo transport depletion code MCBMPI(Monte Carlo Burnup code in MPI version)was employed to calculate the new VERA(Virtual Environment for Reactor Applications)depletion benchmark.The influence of different basic burnup cross section on the burnup calculation in transportation was compared and analyzed.[Results]Calculation results show that relative error of keff is less than 8‰in all the depletion stages,and the mass variation of 235U and 135Xe are less than 4‰and 5‰respectively,in the last depletion stage.The variation between thermal and fast neutron spectrum cross section library is small in this benchmark.[Conclusions]It is recommended to select a basic depletion cross section library containing similar neutron spectrum instead of a typical thermal neutron cross section library in real application.