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47 lines
1.6 KiB
ReStructuredText
===========================
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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. OpenMC supports either
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continuous-energy or multi-group transport. The continuous-energy particle
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interaction data is based on a native HDF5 format that can be generated from ACE
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files used by the MCNP and Serpent Monte Carlo codes.
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OpenMC was originally developed by members of the `Computational Reactor Physics
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Group <http://crpg.mit.edu>`_ at the `Massachusetts Institute of Technology
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<http://web.mit.edu>`_ starting in 2011. Various universities, laboratories, and
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other organizations now contribute to the development of OpenMC. For more
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information on OpenMC, feel free to send a message to the User's Group `mailing
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list <https://groups.google.com/forum/?fromgroups=#!forum/openmc-users>`_.
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.. admonition:: Recommended publication for citing
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:class: tip
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Paul K. Romano, Nicholas E. Horelik, Bryan R. Herman, Adam G. Nelson, Benoit
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Forget, and Kord Smith, "`OpenMC: A State-of-the-Art Monte Carlo Code for
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Research and Development <https://doi.org/10.1016/j.anucene.2014.07.048>`_,"
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*Ann. Nucl. Energy*, **82**, 90--97 (2015).
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.. only:: html
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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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examples/index
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releasenotes
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methods/index
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usersguide/index
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devguide/index
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pythonapi/index
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capi/index
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io_formats/index
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publications
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license
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developers
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