OpenMC Monte Carlo Code
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']
Mesh to Settings.entropy_mesh
Mesh to Settings.ufs_mesh
sourcepoint_* options are now specified in a Settings.sourcepoint dictionarySettings.resonance_scattering
Settings.temperature['multipole'] = True
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.
openmc.data
Settings.cutoff
Material.add_element)openmc.model.pack_trisos
openmc.search_for_keff
openmc.model.Model
openmc.plot_inline
openmc.EnergyFunctionFilter
Geometry.determine_paths
openmc.TallyDerivative)openmc.openmoc_compatible
This release contains new contributions from the following people:
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.GeometryFile class no longer exists and is replaced by the Geometry class which now has an export_to_xml()` method.Source class and assigned to the Settings.source property.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.
This release contains new contributions from the following people:
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.
openmc.mgxs Python module enabling multi-group cross section generationThis release contains new contributions from the following people:
This release contains new contributions from the following people:
This release contains new contributions from the following people:
This release contains new contributions from the following people:
This release contains new contributions from the following people:
This release contains new contributions from the following people: