OpenMC is a community-developed Monte Carlo neutron and photon transport
simulation code. It is capable of performing fixed source, k-eigenvalue, and
subcritical multiplication calculations on models built using either a
constructive solid geometry or CAD representation. OpenMC supports both
continuous-energy and multigroup transport. The continuous-energy particle
interaction data is based on a native HDF5 format that can be generated from
ACE files produced by NJOY. Parallelism is enabled via a hybrid MPI and OpenMP
programming model.