openmc (conda): 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.
Keywords: computational-physics|high-performance-computing|monte-carlo-simulation|neutron-transport|neutronics|nuclear-data|nuclear-energy|nuclear-engineering|nuclear-fusion|openmc|particle-transport|photon-transport|radiation-transport