# openmc.deplete.abc.NormalizationHelper¶

class openmc.deplete.abc.NormalizationHelper[source]

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
factor(source_rate)[source]

Return normalization factor

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

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.

Parameters: 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()[source]

Reset state for normalization

update(fission_rates)[source]

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()