openmc.data.FissionProductYields¶
- class openmc.data.FissionProductYields(ev_or_filename)[source]¶
Independent and cumulative fission product yields.
- Parameters
ev_or_filename (str of openmc.data.endf.Evaluation) – ENDF fission product yield evaluation to read from. If given as a string, it is assumed to be the filename for the ENDF file.
- Variables
cumulative (list of dict) – Cumulative yields for each tabulated energy. Each item in the list is a dictionary whose keys are nuclide names and values are cumulative yields. The i-th dictionary corresponds to the i-th incident neutron energy.
energies (Iterable of float or None) – Energies at which fission product yields are tabulated.
independent (list of dict) – Independent yields for each tabulated energy. Each item in the list is a dictionary whose keys are nuclide names and values are independent yields. The i-th dictionary corresponds to the i-th incident neutron energy.
nuclide (dict) – Properties of the fissioning nuclide.
Notes
Neutron fission yields are typically not measured with a monoenergetic source of neutrons. As such, if the fission yields are given at, e.g., 0.0253 eV, one should interpret this as meaning that they are derived from a typical thermal reactor flux spectrum as opposed to a monoenergetic source at 0.0253 eV.
- classmethod from_endf(ev_or_filename)[source]¶
Generate fission product yield data from an ENDF evaluation
- Parameters
ev_or_filename (str or openmc.data.endf.Evaluation) – ENDF fission product yield evaluation to read from. If given as a string, it is assumed to be the filename for the ENDF file.
- Returns
Fission product yield data
- Return type