The OpenMC development team is proud to announce the release of version 0.9.0. Some of the exciting new features in this release include:
- Stochastic volume calculations
- Multi-delayed group cross section generation
- Ability to calculate multi-group cross sections over meshes
- Temperature interpolation on cross section data
- Nuclear data interface in Python API
- Ability to define fuel by enrichment
- Random sphere packing for TRISO particle generation
- Critical eigenvalue search
- Energy function tally filters
- New XML parser, pugixml
- Differential tallies
- Consistent multi-group scattering matrices
Our documentation has also improved substantially and includes a number of new example Jupyter notebooks.
I’d like to thank Will Boyd, Sterling Harper, Colin Josey, Jingang Liang, Adam Nelson, Sam Shaner, Jon Walsh, and especially first-time contributors Amanda Lund and Qingming He for their contributions to this release. The source code can be downloaded at https://github.com/mit-crpg/openmc/releases or the master branch on the GitHub Repository.