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OpenMC 0.7.1
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@paulromano paulromano released this 23 Dec 21:39
v0.7.1
97fd5e1

This release of OpenMC provides some substantial improvements over version 0.7.0. Non-simple cell regions can now be defined through the | (union) and ~ (complement) operators. Similar changes in the Python API also allow complex cell regions to be defined. A true secondary particle bank now exists; this is crucial for photon transport (to be added in the next minor release). A rich API for multi-group cross section generation has been added via the openmc.mgxs Python module.

Various improvements to tallies have also been made. It is now possible to explicitly specify that a collision estimator be used in a tally. A new delayedgroup filter and delayed-nu-fission score allow a user to obtain delayed fission neutron production rates filtered by delayed group. Finally, the new inverse-velocity score may be useful for calculating kinetics parameters.

Caution! In previous versions, depending on how OpenMC was compiled binary output was either given in HDF5 or a flat binary format. With this version, all binary output is now HDF5 which means you must have HDF5 in order to install OpenMC. Please consult the user's guide for instructions on how to compile with HDF5.

New Features

  • Support for complex cell regions (union and complement operators)
  • Generic 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

Bug Fixes

  • 2993228 Bug with material filter when void material present
  • d748406 Fix triggers on tallies with multiple filters
  • c29a811 Correctly handle maximum transport energy
  • 3edc238 Fixes in the nu-scatter score
  • 629e3b2 Assume unspecified surface coefficients are zero in Python API
  • 5dbe8b7 Fix energy filters for openmc-plot-mesh-tally
  • ff66f41 Fixes in the openmc-plot-mesh-tally script
  • 441fd4f Fix bug in kappa-fission score
  • 7e5974a Allow fixed source simulations from Python API

Contributors

This release contains new contributions from the following people:

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