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Welcome to the OpenMOC repository! OpenMOC is a simulation tool for
solving for the flux, power distribution, and multiplication factor
within a nuclear reactor. The code employs the deterministic method
of characteristics, with support for both fixed source and eigenvalue
calculations. The OpenMOC project aims to provide a simple-to-use
Python package bound to a back-end of source code written in C/C++
and CUDA. It includes support for constructive solid geometry and 2D
ray tracing for fully heterogeneous multi-group calculations.
Development of OpenMOC began at MIT in 2012 and is spearheaded by
several graduate students in the
Nuclear Science & Engineering Department
_.
Complete documentation on OpenMOC is hosted at
https://mit-crpg.github.io/OpenMOC/. If you would like to
contribute to the OpenMOC project, please contact
_ the
development team.
For a guided example, see a demonstration IPython Notebook
_.
Detailed installation instructions
_ can be found in the
User's Guide.
Join the OpenMOC users group
_ to ask questions and discuss
methods and simulation workflows.
Please cite OpenMOC in your publications if it helps your research:
.. code-block:: latex
@article{openmoc2014,
author = {Boyd, William and Shaner, Samuel and Li, Lulu and Forget, Benoit and Smith, Kord},
journal = {Annals of Nuclear Energy},
title = {The OpenMOC Method of Characteristics Neutral Particle Transport Code},
volume = {68},
pages = {43--52},
year = {2014}
}
OpenMOC is approved for distribution under the MIT license_.
.. _installation instructions: https://mit-crpg.github.io/OpenMOC/usersguide/install.html .. _license: https://mit-crpg.github.io/OpenMOC/license.html .. _Nuclear Science & Engineering Department: http://web.mit.edu/nse/ .. _IPython Notebook: http://nbviewer.ipython.org/gist/anonymous/abbce6824bceda49a615 .. _contact: https://mit-crpg.github.io/OpenMOC/developers.html .. _users group: https://groups.google.com/forum/#!forum/openmoc-users