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Created Changes since version 0.8.0 (markdown)

Paul Romano 2016-11-09 15:28:24 -06:00
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## Notable changes in user input
- OpenMC now uses an HDF5 format for cross sections rather than ACE libraries. Scripts have been written to make converting ACE libraries easy (see descriptions [here](http://openmc.readthedocs.io/en/latest/usersguide/install.html#cross-section-configuration)).
- Nuclides are no longer identified by cross section "suffixes", e.g. 71c, 80c, 20t. Instead, temperatures can be assigned directly to materials and/or cells and this information is used to load the correct cross sections.
- Nuclides and thermal scattering tables should now use names consistent with the proposed [GND](https://ndclx4.bnl.gov/gf/project/gnd/) format. This means U235 instead of U-235, Am242_m1 instead of Am-242m, and c_H_in_H2O instead of lwtr or hh2o. Note that when ACE libraries are converted, the proper GND names are used (so regardless of source, light water will become c_H_in_H2O).
- All energies are now in **eV** rather than **MeV**. If you create an energy filter in a tally, make sure you account for this.
- Tally filters have been refactored. Rather than creating instances of `openmc.Filter` with a `type=` argument, there are now separate classes for each type of filter, e.g., `openmc.CellFilter`, `openmc.EnergyFilter`, `openmc.MeshFilter`.
- The `cross_sections` attribute has been moved from `openmc.Settings` to `openmc.Materials`. Similarly the `<cross_sections>` element in the XML input should appear in materials.xml now, not settings.xml.
- Support for elements in XML files / Fortran code has been removed. If you want to specify natural elements, you'll need to do so via the Python API. When a material with a natural element is exported to XML, the known natural abundances will be used to create individual isotopes.
## Other enhancements
- For users with multi-temperature libraries, it is possible to use temperature interpolation.
- The Python API now features an `openmc.data` package that can parse ACE and ENDF files and forms the basis for converting ACE to HDF5 files. It also has physical data that may be useful for users, such as isotope natural abundances, and atomic masses.
- If fission energy release data is present in your cross section libary, you can now tally prompt and recoverable fission energy release during a simulation.