OpenMC: A state-of-the-art Monte Carlo code for research and development

被引：591

作者：

Romano, Paul K. ^{[1
]}

Horelik, Nicholas E. ^{[1
]}

Herman, Bryan R. ^{[1
]}

Nelson, Adam G. ^{[2
]}

Forget, Benoit ^{[1
]}

Smith, Kord ^{[1
]}

机构：

[1] MIT, Dept Nucl Sci & Engn, Cambridge, MA 02139 USA

[2] Univ Michigan, Dept Nucl Engn & Radiol Sci, Ann Arbor, MI 48104 USA

来源：

ANNALS OF NUCLEAR ENERGY | 2015年 / 82卷

关键词：

Monte Carlo; Neutron transport; OpenMC; Parallel; XML; HDF5; TRANSPORT CODE; PERFORMANCE; SIMULATIONS; ALGORITHMS;

D O I：

10.1016/j.anucene.2014.07.048

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes. (C) 2014 Elsevier Ltd. All rights reserved.

引用

页码：90 / 97

页数：8

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