Uniform Tally Density–Based Strategy for Efficient Global Tallying in Monte Carlo Criticality Calculation
Authors:
- 1 April 2016
- journal article
- research article
- Published by Taylor & Francis in Nuclear Science and Engineering
- Vol. 182 (4) , 555-562
Abstract
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).Keywords
This publication has 1 reference indexed in Scilit:
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