Journal
ANNALS OF NUCLEAR ENERGY
Volume 82, Issue -, Pages 90-97Publisher
PERGAMON-ELSEVIER SCIENCE LTD
DOI: 10.1016/j.anucene.2014.07.048
Keywords
Monte Carlo; Neutron transport; OpenMC; Parallel; XML; HDF5
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Funding
- Naval Reactors Division of the U.S. Department of Energy
- Consortium for Advanced Simulation of Light Water Reactors, an Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy [DE-AC05-00OR22725]
- Office of Advanced Scientific Computing Research, Office of Science, U.S. Department of Energy [DE-AC02-06CH11357]
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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.
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