Uniform Tally Density–Based Strategy for Efficient Global Tallying in Monte Carlo Criticality Calculation

Based on the inspiration of the uniform fission site (UFS) algorithm, we propose a strategy for biasing fission secondary neutrons using tally density obtained from past cycles in a Monte Carlo criticality calculation when the purpose is to seek high-performance global tallying. Using this strategy for global volume-averaged cell flux and energy deposition tallies when performing criticality calculations on a pin-by-pin model of the Dayawan nuclear power station nuclear reactor yields better performance. All the strategies (including the original UFS algorithm) are implemented in a parallel Monte Carlo particle transport code JMCT (J Monte Carlo Transport), which is recently developed software constructed on the framework of JCOGIN (J COmbinatorial Geometry Monte Carlo transport INfrastructure).