The OpenMC Monte Carlo Code

OpenMC is a Monte Carlo particle transport simulation code focused on neutron criticality calculations. It is capable of simulating 3D models based on constructive solid geometry with second-order surfaces. OpenMC supports either continuous-energy or multi-group transport. The continuous-energy particle interaction data is based on ACE format cross sections, also used in the MCNP and Serpent Monte Carlo codes.

OpenMC was originally developed by members of the Computational Reactor Physics Group at the Massachusetts Institute of Technology starting in 2011. Various universities, laboratories, and other organizations now contribute to the development of OpenMC. For more information on OpenMC, feel free to send a message to the User’s Group mailing list. Documentation for the latest developmental version of the develop branch can be found on Read the Docs.