openmc.deplete.helpers.DirectReactionRateHelper

class openmc.deplete.helpers.DirectReactionRateHelper(n_nuc, n_react)

Class for generating one-group reaction rates with direct tallies

This class generates reaction rate tallies for each nuclide and transmutation reaction relevant for a depletion calculation.

Parameters

• n_nucs (int) – Number of burnable nuclides tracked by openmc.deplete.CoupledOperator

• n_react (int) – Number of reactions tracked by an instance of openmc.deplete.CoupledOperator

Variables

nuclides (list of str) – All nuclides with desired reaction rates.

generate_tallies(materials, scores)

Produce one-group reaction rate tally

Uses the openmc.lib to generate a tally of relevant reactions across all burnable materials.

Parameters

• materials (iterable of openmc.lib.Material) – Burnable materials in the problem. Used to construct a openmc.lib.MaterialFilter

• scores (iterable of str) – Reaction identifiers, e.g. "(n, fission)", "(n, gamma)", needed for the reaction rate tally.

get_material_rates(mat_index, nuc_index, rx_index)

Return an array of reaction rates for a material

Parameters

• mat_index (int) – Index for the material

• nuc_index (iterable of int) – Index for each nuclide in nuclides in the desired reaction rate matrix

• rx_index (iterable of int) – Index for each reaction scored in the tally

Returns

rates – Array with shape (n_nuclides, n_rxns) with the reaction rates in this material

Return type

numpy.ndarray

property nuclides

List of nuclides with requested reaction rates

property rate_tally_means

The mean results of the tally of every material’s reaction rates for this cycle

reset_tally_means()

Reset the cached mean rate tallies. .. note:

This step must be performed after each transport cycle