openmc.deplete.helpers.DirectReactionRateHelper

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

Class that generates tallies for one-group rates

Parameters:
Variables:

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.

Parameters:
  • materials (iterable of openmc.Material) – Burnable materials in the problem. Used to construct a openmc.MaterialFilter
  • scores (iterable of str) – Reaction identifiers, e.g. "(n, fission)", "(n, gamma)", needed for the reaction rate tally.
get_material_rates(mat_id, nuc_index, react_index)[source]

Return an array of reaction rates for a material

Parameters:
  • mat_id (int) – Unique ID for the requested material
  • nuc_index (iterable of int) – Index for each nuclide in nuclides in the desired reaction rate matrix
  • react_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