蒙特卡罗临界计算全局计数效率新算法研究

Algorithm researches for efficient global tallying in criticality calculation of Monte Carlo method

为提高蒙特卡罗临界计算时全局计数的整体效率,对比分析了新提出的均匀计数密度算法、均匀径迹数密度算法和原有的均匀裂变点算法.以大亚湾核反应堆pin-by-pin模型的全局体平均通量计数和中子沉积能计数为例,前两种算法较均匀裂变点算法都获得了整体效率的提高.上述算法已经在自主开发的并行蒙特卡罗输运程序JMCT上予以实现.

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.