Comparison of Different Strategies for Global Tallying in Monte Carlo Criticality Calculation

Abstract

We propose the uniform tally density algorithm and the uniform track number density algorithm for biasing the fission secondary neutron number in active cycles of the Monte Carlo criticality calculation when the target is seeking high performance of some global tally and compare these strategies with the original uniform fission site algorithm. Using the global volume averaged cell flux tally and the global energy deposition tally of the pin-by-pin model of Dayawan nuclear reactor as examples, the efficiencies of these strategies are compared carefully. All the strategies are realized in a recently developed parallel Monte Carlo particle transport code JMCT.