mit-crpg/openmc: OpenMC 0.9.0
Creators
- 1. Argonne National Laboratory
- 2. MIT
- 3. University of Michigan
- 4. Lawrence Livermore National Laboratory
- 5. Xi'an Jiaotong University
- 6. University of California, Berkeley
- 7. SNPTC
- 8. Teachers Pay Teachers
- 9. Idaho National Laboratory
- 10. The University of South Carolina
Description
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
andcol_spec
>colors
).Plot.colors
now accepts a dictionary mappingCell
orMaterial
instances to RGB 3-tuples or string colors names, e.g.:plot.colors = { fuel: 'yellow', water: 'blue' }
make_hexagon_region
is nowget_hexagonal_prism
Several changes in
Settings
attributes:weight
is now set asSettings.cutoff['weight']
- Shannon entropy is now specified by passing a
Mesh
toSettings.entropy_mesh
- Uniform fission site method is now specified by passing a
Mesh
toSettings.ufs_mesh
- All
sourcepoint_*
options are now specified in aSettings.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 nowSettings.output['path']
All the
openmc.mgxs.Nu*
classes are gone. Instead, anu
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
Files
mit-crpg/openmc-v0.9.0.zip
Files
(3.9 MB)
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Additional details
Related works
- Is supplement to
- https://github.com/mit-crpg/openmc/tree/v0.9.0 (URL)