Open AccessModified uniform-fission-site algorithm in Monte Carlo simulation of reactor criticality problem
Author(s) -
Shangguan Danhua,
Gang Li,
Li Deng,
Baoyin Zhang,
Rui Li,
Fu Yuan-Guan
Publication year - 2015
Publication title -
wuli xuebao
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.199
H-Index - 47
ISSN - 1000-3290
DOI - 10.7498/aps.64.052801
Subject(s) - criticality , monte carlo method , benchmark (surveying) , computer science , neutron transport , algorithm , fission , matching (statistics) , nuclear engineering , statistical physics , physics , nuclear physics , mathematics , neutron , statistics , engineering , geodesy , geography
Because of a very non-uniform power distribution in core region, a very non-uniform distribution of relative uncertainties exists for tallies in Monte Carlo criticality calculations of pin-by-pin reactor model. To make a large part of cells obtain small enough relative uncertainties with reasonable time costs, increasing the total sample scale is not a good choice. By realizing a modified uniform-fission-site algorithm on the basis of source iteration algorithm of parallel Monte Carlo transport code JMCT, we obtain higher efficiencies for tallies in the calculations of pin-by-pin model of the Dayawang reactor plant. This work supplies a useful tool for matching the goal of simulating the benchmark pin-by-pin reactor models with a pre-described standard(the so called Kord-Smith challenge).