RMC – A Monte Carlo code for reactor core analysis. (August 2015)

• Record Type: Journal Article
• Authors: Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin
• Is Part Of: Annals of nuclear energy. Volume 82(2015:Aug.)
• Page Start: 121
• Page End: 129
• Publication Date: 2015-08
• Subjects: RMC -- Monte Carlo -- Reactor analysis -- Source convergence -- Burnup calculation; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2014.08.048
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: A Monte Carlo reactor analysis code RMC is developed by Tsinghua University. The criticality benchmark problems have been taken out to test RMC. Full reactor core benchmarks have been taken out. A 2-D pin-by-pin full core burnup problem has been taken out. RMC shows good performances in criticality, burnup and parallel calculations. Abstract: A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the efficiency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.