openmc.deplete.abc.NormalizationHelper

class openmc.deplete.abc.NormalizationHelper

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)

Return normalization factor

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

nuclides

List of nuclides with requested reaction rates

prepare(chain_nucs, rate_index)

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()

Reset state for normalization

update(fission_rates)

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()