bluemira.codes.openmc.tallying
==============================

.. py:module:: bluemira.codes.openmc.tallying

.. autoapi-nested-parse::

   Functions for creating the openmc tallies.



Functions
---------

.. autoapisummary::

   bluemira.codes.openmc.tallying.csg_filter_cells
   bluemira.codes.openmc.tallying.dagmc_tallys


Module Contents
---------------

.. py:function:: 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.


.. py:function:: dagmc_tallys(material_list, model: openmc.Geometry, mesh_shape: tuple[float, Ellipsis] = (100, 100, 100))

   DAGMC default mesh tallys