ENDF/B-VII Depletion Library Compression to Optimize the Computational Efficiency in STREAM Code

Abstract

A deterministic neutron transport analysis code, STREAM (Steady state and Transient REactor Analysis code with Method of Characteristics), has been developed to perform whole light-water reactors (LWR) core calculations with the direct transport analysis method and the two-step method. A detailed isotopic transition matrix based on ENDF/B-VII nuclear data and including 3837 nuclides and 43416 transitions can be used in STREAM to solve the Bateman equations that describe the nuclide production, depletion, and decay processes. In this study, a simplified burnup chain model was created out of the detailed matrix to reduce computation time and memory usage for the purpose of calculating effective neutron multiplication factors (keff) and power distributions. With the simplified burnup chain model, the keff variation for an OPR-1000 fuel assembly (FA) depletion calculation (with gadolinium fuel) is reproduced within 30 pcm and the power distribution is almost identical with a maximum discrepancy of 0.1 % at the location of burnable absorber rod compared to the detailed chain model, for a speed-up factor of 40 in depletion calculation and 3 in total simulation. These reductions of computational cost are achieved by removing nuclides (and their transitions) whose contribution to reactivity or anti-reactivity is negligible, thereby condensing the number of selected nuclides to 281 in the optimized depletion chain compared to more than 3800 nuclides in the detailed depletion chain.