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35 lines
1.2 KiB
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===========================
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The OpenMC Monte Carlo Code
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===========================
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OpenMC is a Monte Carlo particle transport simulation code focused on neutron
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criticality calculations. It is capable of simulating 3D models based on
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constructive solid geometry with second-order surfaces. The particle interaction
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data is based on ACE format cross sections, also used in the MCNP and Serpent
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Monte Carlo codes.
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OpenMC was originally developed by members of the `Computational Reactor Physics
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Group`_ at the `Massachusetts Institute of Technology`_ starting
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in 2011. Various universities, laboratories, and other organizations now
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contribute to the development of OpenMC. For more information on OpenMC, feel
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free to send a message to the User's Group `mailing list`_.
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.. _Computational Reactor Physics Group: http://crpg.mit.edu
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.. _Massachusetts Institute of Technology: http://web.mit.edu
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.. _mailing list: https://groups.google.com/forum/?fromgroups=#!forum/openmc-users
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--------
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Contents
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--------
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.. toctree::
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:maxdepth: 1
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quickinstall
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releasenotes/index
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methods/index
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usersguide/index
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devguide/index
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publications
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license
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developers
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