bluemira.codes.openmc.tallying

Functions for creating the openmc tallies.

Functions

csg_filter_cells(material_list, csg_model)

Create scores and the filter for the scores. Give them names.

dagmc_tallys(material_list, model[, mesh_shape])

DAGMC default mesh tallys

Module Contents

bluemira.codes.openmc.tallying.csg_filter_cells(material_list, csg_model: bluemira.codes.openmc.make_csg.CellStage)

Create scores and the filter for the scores. Give them names.

Returns:

  • TBR – Achieved by (n,Xt) reaction, which counts the number of tritium-producing nuclear reactions per neutron emitted at the source.

    We used the (n,Xt) score because the Lithium produces a maximum of 1 Tritium per reaction, so there won’t be any concerns about uncer-counting the TBR.

  • Powers – Measures the nuclear heating in various locations and materials, and interpret this as power. “damage-energy” is given by eV per source neutron. Multiply by neutron source rate, and then divide by (number of atoms and threshold displacement energy) to get the DPA.

  • Fluence – Measures # of neutrons streaming through. “flux” is given in # per source particle, so multiply by # of source neutrons to get the total fluence over the simulation. Divide by area to get fluence in unit: cm^-2.

Parameters:

csg_model (bluemira.codes.openmc.make_csg.CellStage)

bluemira.codes.openmc.tallying.dagmc_tallys(material_list, model: openmc.Geometry, mesh_shape: tuple[float, Ellipsis] = (100, 100, 100))

DAGMC default mesh tallys

Parameters:

  • model (openmc.Geometry)

  • mesh_shape (tuple[float, Ellipsis])