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

Danhua ShangGuan, Gang Li, Baoyin Zhang, Li Deng, Yan Ma, Yuanguan Fu, Rui Li,Xiaoli Hu

Nuclear Science and Engineering / Volume 182 / Number 4 / April 2016 / Pages 555-562

Technical Paper /

First Online Publication:March 10, 2016
Updated:April 1, 2016

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).