OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development

• Main Authors: Nelson, Adam G. (Author), Romano, Paul Kollath (Contributor), Horelik, Nicholas Edward (Contributor), Herman, Bryan R (Contributor), Forget, Benoit Robert Yves (Contributor), Smith, Kord S. (Contributor)
• Other Authors: Massachusetts Institute of Technology. Department of Nuclear Science and Engineering (Contributor)
• Format: Article
• Language: English
• Published: EDP Sciences, 2017-06-14T14:29:20Z.
• Subjects: Article
• Online Access: Get fulltext

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.
United States. Department of Energy (Rickover Fellowship Program in Nuclear Engineering)
United States. Department of Energy (Contract No. DE-AC05-00OR22725)
United States. Department of Energy. Office of Advanced Scientific Computing Research (Contract DE-AC02-06CH11357)