Abstract class for obtaining normalization factor on tallies

This helper class determines how reaction rates calculated by an instance of openmc.deplete.Operator should be normalized for the purpose of constructing a burnup matrix. Based on the method chosen, the power or source rate provided by the user, and reaction rates from a ReactionRateHelper, this class will scale reaction rates to the correct values.

Variables:nuclides (list of str) – All nuclides with desired reaction rates. Ordered to be consistent with openmc.deplete.Operator

Return normalization factor

Parameters:source_rate (float) – Power in [W] or source rate in [neutron/sec]
Returns:Normalization factor for tallies
Return type:float

List of nuclides with requested reaction rates

prepare(chain_nucs, rate_index)[source]

Perform work needed to obtain energy produced

This method is called prior to the transport simulations in openmc.deplete.Operator.initial_condition(). Only used for energy-based normalization.

  • chain_nucs (list of str) – All nuclides to be tracked in this problem
  • rate_index (dict of str to int) – Mapping from nuclide name to index in the fission_rates for update().

Reset state for normalization


Update the normalization based on fission rates (only used for energy-based normalization)

Parameters: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()