class openmc.deplete.helpers.ChainFissionHelper[source]

Computes normalization using fission Q values from depletion chain

  • nuclides (list of str) – All nuclides with desired reaction rates. Ordered to be consistent with openmc.deplete.CoupledOperator

  • energy (float) – Total energy [J/s/source neutron] produced in a transport simulation. Updated in the material iteration with update().

prepare(chain_nucs, rate_index)[source]

Populate the fission Q value vector from a chain.

  • chain_nucs (iterable of openmc.deplete.Nuclide) – Nuclides used in this depletion chain. Do not need to be ordered

  • rate_index (dict of str to int) – Dictionary mapping names of nuclides, e.g. "U235", to a corresponding index in the desired fission Q vector.


Update energy produced with fission rates in a material


fission_rates (numpy.ndarray) – fission reaction rate for each isotope in the specified material. Should be ordered corresponding to initial rate_index used in prepare()