Open AccessAlgorithm researches for efficient global tallying in criticality calculation of Monte Carlo method
Author(s) -
Shangguan Danhua,
Li Deng,
Gang Li,
Baoyin Zhang,
Yan Ma,
Y. Fu,
Rui Li,
Xiaoli Hu
Publication year - 2016
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.65.062801
Subject(s) - criticality , computer science , monte carlo method , algorithm , estimator , track (disk drive) , statistical physics , physics , mathematics , nuclear physics , statistics , operating system
Based on the research of the uniform fission site algorithm, the uniform tally density algorithm and the uniform track number density algorithm are proposed and compared with the original uniform fission site algorithm in this paper for seeking high performance of global tallying in Monte Carlo criticality calculation. Because reducing the largest uncertainties to an acceptable level simply by running a large number of neutron histories is often prohibitively expensive, the researches are indispensable for the calculation to reach the goal of practical application (the so called 95/95 standard). Using the global volume-averaged cell flux tally and energy deposition tally of the pin-by-pin model of Dayawan nuclear reactor as two examples, these new algorithms show better results. Although the uniform tally density algorithm has the best performance, the uniform track number density algorithm still has the advantage of being applicable to any type of tally, which is based on the track length estimator without any modification. All the algorithms are realized in a recently developed parallel Monte Carlo particle transport code JMCT.