openmoc 0.1.4b

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.

@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}
}