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

[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 k_eff 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. (authors)