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The OpenMC Monte Carlo Code

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.

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.

Recommended publication for citing

Paul K. Romano, Nicholas E. Horelik, Bryan R. Herman, Adam G. Nelson, Benoit Forget, and Kord Smith, "OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development," Ann. Nucl. Energy, 82, 90--97 (2015).

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   Contents
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.. toctree::
    :maxdepth: 1

    quickinstall
    examples/index
    releasenotes/index
    methods/index
    usersguide/index
    devguide/index
    pythonapi/index
    capi/index
    io_formats/index
    publications
    license