OpenMC is a community-developed software project providing a Monte Carlo code for neutron and photon transport. Its solver supports fixed-source, k-eigenvalue, and subcritical multiplication calculations, using models defined with constructive solid geometry or CAD representations. A flexible tally system lets users calculate and analyze a broad range of physical quantities. OpenMC can run in parallel through a hybrid MPI and OpenMP programming model and has been tested on leadership-class supercomputers.

The project also offers extensible Python and C/C++ interfaces for building pre- and post-processing tools, generating multigroup cross sections, automating workflows, performing depletion calculations, coupling multiphysics applications, and visualizing geometry and tally results. Its Python-based nuclear data interface allows users to inspect, modify, and process data in ENDF, ACE, and OpenMC’s native HDF5 formats. Supporting tools and practices include continuous integration, automated critical benchmark simulations, cross-code comparisons, and performance testing to help maintain the software’s quality and accuracy.