| OpenMC | |
|---|---|
| Developer | Massachusetts Institute of Technology |
| Initial release | December 2012 |
| Written in | C++, Python |
| Operating system | Linux |
| Type | Computational physics |
| Website | openmc |
| Repository | github |
OpenMC is an open source Monte Carlo neutron and photon simulation transport code. Initially developed by the Computational Reactor Physics Group at MIT in 2011 as part of a project to develop scalable parallel algorithms for future exascale supercomputers, it has been contributed to by various universities, laboratories, and other institutions since its release to the public in December 2012.[1][2] Unlike other major Monte Carlo transport codes such as MCNP or Serpent, it is not export controlled. It has been used in labs including the Consortium for Advanced Simulation of LWRs and the ANL Center for Exascale Simulation of Advanced Reactors.[1][3]
Features of OpenMC include the ability to perform fixed source, k-eigenvalue, and subcritical multiplication calculations on models built with Constructive solid geometry or CAD. It also features large Python and C/C++ APIs that expand its features.[2][4]
References
- ^ a b Romano, Paul K.; Horelik, Nicholas E.; Herman, Bryan R.; Nelson, Adam G.; Forget, Benoit; Smith, Kord (2015-08-01). "OpenMC: A state-of-the-art Monte Carlo code for research and development". Annals of Nuclear Energy. Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013, SNA + MC 2013. Pluri- and Trans-disciplinarity, Towards New Modeling and Numerical Simulation Paradigms. 82: 90–97. doi:10.1016/j.anucene.2014.07.048. hdl:1721.1/109853. ISSN 0306-4549.
- ^ a b "OpenMC". openmc.org. Retrieved 2025-07-20.
- ^ "OpenMC". Argonne National Laboratory. Retrieved 2025-07-20.
- ^ "Release Notes — OpenMC Documentation". docs.openmc.org. Retrieved 2025-07-20.
