class openmc.deplete.helpers.DirectReactionRateHelper(n_nuc, n_react)[source]

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.


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

generate_tallies(materials, scores)[source]

Produce one-group reaction rate tally

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

  • 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)[source]

Return an array of reaction rates for a material

  • 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


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

Return type


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 the cached mean rate tallies. .. note:

This step must be performed after each transport cycle