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Published July 25, 2016 | Version v0.8.0
Software Open

openmc: OpenMC 0.8.0

Creators

  • 1. Argonne National Laboratory
  • 2. MIT
  • 3. University of Michigan
  • 4. Massachusetts Institute of Technology
  • 5. University of California, Berkeley
  • 6. SNPTC
  • 7. Teachers Pay Teachers
  • 8. Idaho National Laboratory
  • 9. The University of South Carolina

Description

This release of OpenMC includes a few new major features including the capability to perform neutron transport with multi-group cross section data as well as experimental support for the windowed multipole method being developed at MIT. Source sampling options have also been expanded significantly, with the option to supply arbitrary tabular and discrete distributions for energy, angle, and spatial coordinates.

The Python API has been significantly restructured in this release compared to version 0.7.1. Any scripts written based on the version 0.7.1 API will likely need to be rewritten. Some of the most visible changes include the following:

  • SettingsFile is now Settings, MaterialsFile is now Materials, and TalliesFile is now Tallies.
  • The GeometryFile class no longer exists and is replaced by the Geometry class which now has an export_to_xml()` method.
  • Source distributions are defined using the Source class and assigned to the Settings.source property.
  • The Executor class no longer exists and is replaced by openmc.run()and openmc.plot_geometry() functions.

The Python API documentation has also been significantly expanded.

New Features
  • Multi-group mode
  • Vast improvements to the Python API
  • Experimental windowed multipole capability
  • Periodic boundary conditions
  • Expanded source sampling options
  • Distributed materials
  • Subcritical multiplication support
  • Improved method for reproducible URR table sampling
  • Refactor of continuous-energy reaction data
  • Improved documentation and new Jupyter notebooks
Bug Fixes
  • 70daa76e0e9d0bd8163e5c9d306788dbd7cf30c6 Make sure MT=3 cross section is not used
  • 40b05fe94e16731703ca758a55953d62f353fc7a Ensure source bank is resampled for fixed source runs
  • 9586ed3c0718ce5fbfdacc551966a4de9e64fb42 Fix two hexagonal lattice bugs
  • a855e8f1b04f3983b699422c9d938d21f2f3315c Make sure graphite models don't error out on max events
  • 7294a1363485c64b4ee7cb1a5a8e4f0bb67d8ec7 Fix incorrect check on cmfd.xml
  • 12f2467d6e3f529dba3ad8f88f11fb04588246bd Ensure number of realizations is written to statepoint
  • 0227f4823080a686e8130df065762d8d855d6f3d Fix bug when sampling multiple energy distributions
  • 51deaa7cbf4a5bc06bb9ddc0dd2beef830115333 Prevent segfault when user specifies '18' on tally scores
  • fed74b8d761b5baca0f776cdb61fef15a8f8bdf8 Prevent duplicate tally scores
  • 8467aea4e3e019d13ca8127c2dd84de7858348dc Better threshold for allowable lost particles
  • 493c6fd9ccb9d1e23516ba83b24249fb1d4ccba7 Fix type of return argument for h5pget_driver_f
Contributors

This release contains new contributions from the following people:

  • @wbinventor
  • @friedmud
  • @smharper
  • @cjosey
  • @liangjg
  • @nelsonag
  • @paulromano
  • @kellyrowland
  • @samuelshaner

Files

openmc-v0.8.0.zip

Files (2.9 MB)

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Additional details

Related works

Is supplement to
https://github.com/mit-crpg/openmc/tree/v0.8.0 (URL)