openmc.deplete.helpers.FluxCollapseHelper

class openmc.deplete.helpers.FluxCollapseHelper(n_nucs, n_reacts, energies, reactions=None, nuclides=None)[source]

Class that generates one-group reaction rates using multigroup flux

This class generates a multigroup flux tally that is used afterward to calculate a one-group reaction rate by collapsing it with continuous-energy cross section data. Additionally, select nuclides/reactions can be treated with a direct reaction rate tally when using a multigroup flux spectrum would not be sufficiently accurate. This is often the case for (n,gamma) and fission reactions.

New in version 0.12.1.

Parameters

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

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

  • energies (iterable of float) – Energy group boundaries for flux spectrum in [eV]

  • reactions (iterable of str) – Reactions for which rates should be directly tallied

  • nuclides (iterable of str) – Nuclides for which some reaction rates should be directly tallied. If None, then reactions will be used for all nuclides.

Variables

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

generate_tallies(materials, scores)[source]

Produce multigroup flux spectrum tally

Uses the openmc.lib module to generate a multigroup flux tally for each burnable material.

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_index, nuc_index, react_index)[source]

Return an array of reaction rates for a material

Parameters

  • mat_index (int) – Index for 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

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()[source]

Reset the cached mean rate and flux tallies. .. note:

This step must be performed after each transport cycle