- 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.
n_nucs (int) – Number of burnable nuclides tracked by
n_react (int) – Number of reactions tracked by
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
reactionswill be used for all nuclides.
nuclides (list of str) – All nuclides with desired reaction rates.
- generate_tallies(materials, scores)[source]¶
Produce multigroup flux spectrum tally
openmc.libmodule to generate a multigroup flux tally for each burnable material.
materials (iterable of
openmc.Material) – Burnable materials in the problem. Used to construct a
scores (iterable of str) – Reaction identifiers, e.g.
"(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
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