OpenMC

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]

OpenMC
Developer(s)	Massachusetts Institute of Technology
Initial release	December 2012
Written in	C++
Operating system	Linux
Type	Computational physics
Website	openmc.org

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]