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An open source method of characteristics code for solving the 2D neutron distribution in nuclear reactors

Project description

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 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.

Citing OpenMOC

Please cite OpenMOC in your publications if it helps your research:

  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/X license.

Project details

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