The three dimension multi-group neutron transport Monte-Carlo program MCMG-II and the multi-group point burnup program STEP1.0 based on linear chain method are adopted in the transport-burnup coupling system. MCMG-II has fine geometry description capability and a new collision mechanism about the material except the traditional point wise continuous cross section calculation to speed up. STEP1.0 using backtracking algorithm to build linear chains adaptively and has analytical solutions to avoid time step restriction. The coupling information between the burnup and transport program includes neutron spectrum, source strength, reaction rates, initial nuclide density and its variation and the fission products. The interpreted script language Python is used to implement the coupling system. The coupling information can be transferred accurately and smoothly between the programs by its powerful text processing functions. It can also accomplish the step by step transport-burnup calculations automatically according to the total power, radiation time and timing step numbers with its interpreted, interactive characters and perform the graphic processing of the results. The coupling system can do the high fidelity burnup calculations with more comprehensive reference to the factors such as geometry, neutron spectrum and radiation time.

This content is only available via PDF.
You do not currently have access to this content.