OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development
Name
Forget_OpenMC.pdf
Size
651 KB
Format
Adobe PDF
Checksum (MD5)
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Author(s)
Nelson, Adam G.
Romano, Paul Kollath
Horelik, Nicholas Edward
Herman, Bryan R
Forget, Benoit Robert Yves
Smith, Kord S.
Date Issued
June 2014
Journal
SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo
Publisher
EDP Sciences
Citation
Romano, Paul K., Nicholas E. Horelik, Bryan R. Herman, Adam G. Nelson, Benoit Forget, and Kord Smith. “OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development.” Edited by D. Caruge, C. Calvin, C.M. Diop, F. Malvagi, and J.-C. Trama. SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo (2014).
Version
Author's final manuscript
Abstract
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.
MIT Department
Massachusetts Institute of Technology. Department of Nuclear Science and Engineering
Terms of Use
Creative Commons Attribution-Noncommercial-Share Alike
Persistent DSpace Link
DOI of Published Version
https://doi.org/10.1051/snamc/201406016
