Hi all,
The OpenMC development team is proud to announce the release of version 0.5.4. This release includes various new features and bug fixes added over the past four months.
New features in version 0.5.4 include:
- Source sites outside geometry are resampled
- XML-Fortran backed replaced by FoX XML
- Ability to write particle track files
- Handle lost particles more gracefully (via particle track files)
- Multiple random number generator streams
- Mesh tally plotting utility converted to use Tkinter rather than PyQt
- Script added to download ACE data from NNDC
- Mixed ASCII/binary cross_sections.xml now allowed
- Expanded options for writing source bank
- Re-enabled ability to use source file as starting source
- S(a,b) recalculation avoided when same nuclide and S(a,b) table are accessed
Bug fixes include:
-
32c03c4
Check for valid data in cross_sections.xml -
c71ef57
Fix bug in statepoint.py -
8884fb9
Check for all ZAIDs for S(a,b) tables -
b38af09
Fix XML reading on multiple levels of input -
d28750c
Fix bug in convert_xsdir.py -
cf567ca
ENDF/B-VI data checked for compatibility -
6b94613
Fix p_valid sampling inside of sample_energy
I’d like to thank Paul Romano, Nick Horelik, Adam Nelson, Jon Walsh, Streling Harper and Tuomas Viitanen for their contributions in this release.The source code can be downloaded at https://github.com/mit-crpg/openmc/releases or the master branch on the GitHub Repository.
Best regards,
Bryan