RMC – A Monte Carlo code for reactor core analysis Kan Wang a , Zeguang Li b,⇑ , Ding She a , Jin’gang Liang a , Qi Xu a , Yishu Qiu a , Jiankai Yu a , Jialong Sun a , Xiao Fan a , Ganglin Yu a a Department of Engineering Physics, Tsinghua University, Beijing 100084, China b INET, Tsinghua University, Beijing 100084, China article info Article history: Received 29 April 2014 Accepted 21 August 2014 Available online xxxx Keywords: RMC Monte Carlo Reactor analysis Source convergence Burnup calculation 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 calcula- tion, 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. Ó 2014 Elsevier Ltd. All rights reserved. 1. Introduction Generally, neutron transport equations can be solved in two kinds of numerical ways, one of which is the deterministic method and the other one is the stochastic method also called Monte Carlo method (Lux and Koblinger, 1991). Monte Carlo method is flexible in treating complex geometries and energy spectrums, and the time costs are much less dependent on problems dimension. Also it is very easy to perform parallel calculation using Monte Carlo method. However, Monte Carlo method suffers from the disadvan- tages such as more time cost than deterministic method, and random variable results. With the requirement of accurate three- dimensional modeling of the advanced, special or new complex core in reactor neutronics analysis and with the help of great inno- vation of computer technology, Monte Carlo method is becoming a more powerful tool for reactor core analysis and receiving rising attention. With the needs of new conceptual and advanced reactors anal- ysis and to use the advantages of modern computer hardware and software technologies, a new Monte Carlo transport code named RMC (Reactor Monte Carlo Code) has been being developed by Department of Engineering Physics, Tsinghua University, Beijing. The code RMC intends to solve reactor analysis problems, and is able to deal with complex geometry, using continuous energy point-wise cross sections of different materials and temperatures. To meet the requirements of reactor analysis, RMC now has the following functions and special techniques: criticality calculation (Wang and Li, 2011), burnup calculation (She et al., 2013a), parallel calculation (Qiu et al., 2012), fix source calculation and kinetics simulation (Xu et al., 2012), on-the-fly cross-sections processing with temperature (Li and Wang, 2012b), high efficiency searching methods (for cross-sections and geometry) (She et al., 2011), source convergence acceleration (She et al., 2012b), full-core hybrid calculation methods (RMMC method) (Li, 2012), domain decomposition (Liang and Cai, 2012), continuously varying med- ium simulation (Li and Wang, 2011), perturbation and sensitivity analysis (Li et al., 2009), N-TH coupling (Li et al., 2012) and so on. In this paper, parts of the functions and methods implemented in RMC are introduced in Section 2, and some validation results and code performances results are shown in Section 3. 2. Capabilities 2.1. Criticality calculation 2.1.1. Basic physics Criticality calculation of RMC is based on solving neutron trans- port problems using Monte Carlo method (Lieberoth, 1968). The neutron transport method applied in RMC is to simulate neutron histories by tracking each neutron through different regions in the geometry. The distance to next collision site l, is randomly sam- pled according to the total interaction probability in the material http://dx.doi.org/10.1016/j.anucene.2014.08.048 0306-4549/Ó 2014 Elsevier Ltd. All rights reserved. ⇑ Corresponding author. Tel.: +86 10 62783087; fax: +86 10 62782658. E-mail address: lizeguang@tsinghua.edu.cn (Z. Li). Annals of Nuclear Energy xxx (2014) xxx–xxx Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene Please cite this article in press as: Wang, K., et al. RMC – A Monte Carlo code for reactor core analysis. Ann. Nucl. Energy (2014), http://dx.doi.org/10.1016/ j.anucene.2014.08.048