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.Operator
- n_react (int) – Number of reactions tracked by
openmc.deplete.Operator
- 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 aopenmc.MaterialFilter
- scores (iterable of str) – Reaction identifiers, e.g.
"(n, fission)"
,"(n, gamma)"
, needed for the reaction rate tally.
- materials (iterable of
-
get_material_rates
(mat_index, nuc_index, react_index)[source]¶ Return an array of reaction rates for a material
Parameters: Returns: rates – Array with shape
(n_nuclides, n_rxns)
with the reaction rates in this materialReturn type:
-
nuclides
¶ List of nuclides with requested reaction rates
- n_nucs (int) – Number of burnable nuclides tracked by