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 ith dictionary corresponds to the ith 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 ith dictionary corresponds to the ith 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: openmc.data.FissionProductYields