基于随机抽样法的多群核数据不确定性影响分析

基于随机抽样方法，研究多群核数据不确定性对反应堆物理计算的影响。首先利用SCALE软件包中核数据协方差矩阵和自主开发的随机抽样模块SAMP，得到多群微观截面等核数据的抽样值，之后分别使用SCALE/TRITON和PARCS程序进行组件计算及堆芯稳态计算，最后通过统计分析得到组件和堆芯计算结果的不确定度。以Almaraz压水堆核电厂装载的燃料组件和首循环堆芯为对象，研究了不同燃耗下有效增殖因子、动力学参数、核素浓度和双群均匀化宏观截面等组件计算结果，以及堆芯功率分布等堆芯计算结果的不确定度。分析结果表明:组件计算结果不确定度多随燃耗变化，快群宏观截面不确定度总体高于热群；堆芯计算结果受核数据不确定性影响显著，其中稳态径向功率分布的最大不确定度为1.9%左右。

Analysis on influence of multigroup nuclear data uncertainty based on stochastic sampling method

Stochastic sampling method is used in this paper to study the influence of the uncertainty of multigroup nuclear data on reactor physics calculation. Random values for the nuclear data such as multigroup microscopic cross-sections are generated by sampling the AMPX multigroup library using the self-developed SAMP module and the covariance matrix of SCALE software. Lattice calculations and the steady core calculations are then performed by SCALE and PARCS respectively. The uncertainties of lattice and core calculation results are obtained by statistical analysis accordingly. Taking the fuel assembly and first cycle core of Almaraz PWR plant as the objects, the uncertainties of lattice calculation results such as effective multiplication factor, kinetic parameter, nuclide concentration and two-group macroscopic cross section, and the uncertainties of steady core calculation results like core power distribution are analyzed. The results indicate that the assembly parameters are dependent on the fuel burnup, and the maximum uncertainty of steady core radial power distribution is about 1.9%.