openmc.deplete.FissionYieldDistribution¶
-
class
openmc.deplete.
FissionYieldDistribution
(fission_yields)[source]¶ Energy-dependent fission product yields for a single nuclide
Can be used as a dictionary mapping energies and products to fission yields:
>>> fydist = FissionYieldDistribution{ ... {0.0253: {"Xe135": 0.021}}) >>> fydist[0.0253]["Xe135"] 0.021
Parameters: fission_yields (dict) – Dictionary of energies and fission product yields for that energy. Expected to be of the form
{float: {str: float}}
. The first float is the energy, typically in eV, that represents this distribution. The underlying dictionary maps fission products to their respective yields.Variables: - energies (tuple) – Energies for which fission yields exist. Sorted by increasing energy
- products (tuple) – Fission products produced at all energies. Sorted by name.
- yield_matrix (numpy.ndarray) – Array
(n_energy, n_products)
whereyield_matrix[g, j]
is the fission yield of productj
for energy groupg
.
See also
from_xml_element()
- Construction methodsFissionYield
- Class used for storing yields at a given energy
-
classmethod
from_xml_element
(element)[source]¶ Construct a distribution from a depletion chain xml file
Parameters: element (xml.etree.ElementTree.Element) – XML element to pull fission yield data from Returns: Return type: FissionYieldDistribution
-
to_xml_element
(root)[source]¶ Write fission yield data to an xml element
Parameters: root (xml.etree.ElementTree.Element) – Element to write distribution data to