OpenMC version 0.7.1

Hi all,

The OpenMC development team is proud to announce the release of a new version, 0.7.1. Once again, there are an array of exciting new features, including:

  • Support for complex cells (unions and complements)

  • General quadric surface type

  • Improved handling of secondary particles

  • Binary output is now solely HDF5

  • openmc.mgxs Python module enabling multi-group cross section generation

  • Collision estimator for tallies

  • Delayed fission neutron production tallies with ability to filter by delayed group

  • Inverse velocity tally score

  • Performance improvements for binary search

  • Performance improvements for reaction rate tallies

There are a number of bug fixes as well. I’d like to thank Bryan Herman, Colin Josey, Kelly Rowland, Adam Nelson, Sam Shaner, Sterling Harper, Jon Walsh, and Will Boyd for their contributions to this release. The source code can be downloaded at or the master branch on the GitHub Repository.

Best regards,

Thanks to all the developers :slight_smile: you guys rock. keep up the good work

Amazing, good news.
I’m checking the neweast delayed fission neutron fearture :).