mit-crpg/openmc: OpenMC 0.9.0

This release of OpenMC is the first release to use a new native HDF5 cross section format rather than ACE format cross sections. Other significant new features include a nuclear data interface in the Python API (openmc.data) a stochastic volume calculation capability, a random sphere packing algorithm that can handle packing fractions up to 60%, and a new XML parser with significantly better performance than the parser used previously.

CAUTION: With the new cross section format, the default energy units are now electronvolts (eV) rather than megaelectronvolts (MeV)! If you are specifying an energy filter for a tally, make sure you use units of eV now.

The Python API continues to improve over time; several backwards incompatible changes were made in the API which users of previous versions should take note of:

• Each type of tally filter is now specified with a separate class. For example:

  energy_filter = openmc.EnergyFilter([0.0, 0.625, 4.0, 1.0e6, 20.0e6])

• Several attributes of the Plot class have changed (color -> color_by and col_spec > colors). Plot.colors now accepts a dictionary mapping Cell or Material instances to RGB 3-tuples or string colors names, e.g.:

  plot.colors = { fuel: 'yellow', water: 'blue' }

• make_hexagon_region is now get_hexagonal_prism

• Several changes in Settings attributes:

  • weight is now set as Settings.cutoff['weight']
  • Shannon entropy is now specified by passing a Mesh to Settings.entropy_mesh
  • Uniform fission site method is now specified by passing a Mesh to Settings.ufs_mesh
  • All sourcepoint_* options are now specified in a Settings.sourcepoint dictionary
  • Resonance scattering method is now specified as a dictionary in Settings.resonance_scattering
  • Multipole is now turned on by setting Settings.temperature['multipole'] = True
  • The output_path attribute is now Settings.output['path']

• All the openmc.mgxs.Nu* classes are gone. Instead, a nu argument was added to the constructor of the corresponding classes.

• 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, openmc.data
• Allow cutoff energy via Settings.cutoff
• Ability to define fuel by enrichment (see Material.add_element)
• Random sphere packing for TRISO particle generation, openmc.model.pack_trisos
• Critical eigenvalue search, openmc.search_for_keff
• Model container, openmc.model.Model
• In-line plotting in Jupyter, openmc.plot_inline
• Energy function tally filters, openmc.EnergyFunctionFilter
• Replaced FoX XML parser with pugixml
• Cell/material instance counting, Geometry.determine_paths
• Differential tallies (see openmc.TallyDerivative)
• Consistent multi-group scattering matrices
• Improved documentation and new Jupyter notebooks
• OpenMOC compatibility module, openmc.openmoc_compatible

• c5df6ce146abeee0d83447aa7a1deccf354b9ade Fix mesh filter max iterator check
• 1cfa392bb0580b584643927fb3e57f43f28b9f12 Reject external source only if 95% of sites are rejected
• 3353592cb6092f9d0341e1a0393cedca778aefaf Fix bug in plotting meshlines
• 17c678d9f8f797a2d558162107c8eb49dded714f Make sure system_clock uses high-resolution timer
• 23ec0b89bb58a10017d56e651415886ac35a1c6b Fix use of S(a,b) with multipole data
• 7eefb7306ff0f8745af1447789d8cddc9778617f Fix several bugs in tally module
• 7880d4f2461a945a99ffd4fb5a88a92191a06f5c Allow plotting calculation with no boundary conditions
• ad2d9fff55860625f576b9589bd334b80acee337 Fix filter weight missing when scoring all nuclides
• 59fdcac2a91887cf5ce3960980d79f086ab51ae0 Fix use of source files for fixed source calculations
• 9eff5b8a2ead933a6d9839b8c473fd5c82922f59 Fix thermal scattering bugs
• 7848a97edcdb05bee7b8424568c75a6900b068ab Fix combined k-eff estimator producing NaN
• f139ce8dc12ae036e73ddf46c9ed5ba1a563be1c Fix printing bug for tallies with AggregateNuclide
• b8ddfacaf34aedf689cfef26ae954cf6025e1dda Bugfix for short tracks near tally mesh edges
• ec3cfb5bab75b67f0a86ab63cf08073305cd9663 Fix inconsistency in filter weights
• 5e9b06a861d4f596314eff490ad63c051f833f3a Fix XML representation for verbosity
• c39990accb6d0377fc05f004b0809d08e7a7f384 Fix bug tallying reaction rates with multipole on
• c6b67e64434c15483a26733eadbb7335b10be7ea Fix fissionable source sampling bug
• 48954027704d1413f62addf11bfdd072b33713fc Check for void materials in tracklength tallies
• f0214f4c12450c82788ed0546379d9bcae3174f2 Fixes/improvements to the ARES algorithm

This release contains new contributions from the following people:

• @wbinventor
• @smharper
• @QingmingHe
• @cjosey
• @tjlaboss
• @liangjg
• @amandalund
• @nelsonag
• @paulromano
• @samuelshaner
• @walshjon