- class openmc.deplete.MicroXS(data=None, index: Axes | None = None, columns: Axes | None = None, dtype: Dtype | None = None, copy: bool | None = None)¶
Microscopic cross section data for use in transport-independent depletion.
New in version 0.13.1.
- classmethod from_array(nuclides, reactions, data)¶
MicroXSobject from arrays.
nuclides (list of str) – List of nuclide symbols for that have data for at least one reaction.
reactions (list of str) – List of reactions. All reactions must match those in
data (ndarray of floats) – Array containing one-group microscopic cross section values for each nuclide and reaction. Cross section values are assumed to be in [b].
- Return type
- classmethod from_csv(csv_file, **kwargs)¶
MicroXSobject from a
- classmethod from_model(model, reaction_domain, chain_file=None, dilute_initial=1000.0, energy_bounds=(0, 20000000.0), run_kwargs=None)¶
Generate a one-group cross-section dataframe using OpenMC. Note that the
openmcexecutable must be compiled.
model (openmc.Model) – OpenMC model object. Must contain geometry, materials, and settings.
chain_file (str, optional) – Path to the depletion chain XML file that will be used in depletion simulation. Used to determine cross sections for materials not present in the inital composition. Defaults to
dilute_initial (float, optional) – Initial atom density [atoms/cm^3] to add for nuclides that are zero in initial condition to ensure they exist in the cross section data. Only done for nuclides with reaction rates.
reactions (list of str, optional) – Reaction names to tally
energy_bound (2-tuple of float, optional) – Bounds for the energy group.
Cross section data in [b]
- Return type