def generate_tallies(self, materials, scores):
"""Produce multigroup flux spectrum tally
Uses the :mod:`openmc.lib` module to generate a multigroup flux tally
for each burnable material.
Parameters
----------
materials : iterable of :class:`openmc.Material`
Burnable materials in the problem. Used to construct a
:class:`openmc.MaterialFilter`
scores : iterable of str
Reaction identifiers, e.g. ``"(n, fission)"``, ``"(n, gamma)"``,
needed for the reaction rate tally.
"""
self._materials = materials
# adds an entry for fisson to the dictionary of reactions
self._mts = [REACTION_MT[x] for x in scores]
self._scores = scores
# Create flux tally with material and energy filters
self._flux_tally = Tally()
self._flux_tally.writable = False
self._flux_tally.filters = [
MaterialFilter(materials),
EnergyFilter(self._energies)
]
self._flux_tally.scores = ['flux']
# Create reaction rate tally
if self._reactions_direct:
self._rate_tally = Tally()
self._rate_tally.writable = False
self._rate_tally.scores = self._reactions_direct
self._rate_tally.filters = [MaterialFilter(materials)]
if self._nuclides_direct is not None:
self._rate_tally.nuclides = self._nuclides_direct
def generate_tallies(self, materials, scores):
"""Produce one-group reaction rate tally
Uses the :mod:`openmc.lib` to generate a tally
of relevant reactions across all burnable materials.
Parameters
----------
materials : iterable of :class:`openmc.Material`
Burnable materials in the problem. Used to
construct a :class:`openmc.MaterialFilter`
scores : iterable of str
Reaction identifiers, e.g. ``"(n, fission)"``,
``"(n, gamma)"``, needed for the reaction rate tally.
"""
self._rate_tally = Tally()
self._rate_tally.writable = False
self._rate_tally.scores = scores
self._rate_tally.filters = [MaterialFilter(materials)]
def generate_tallies(self, materials, mat_indexes):
"""Construct the fission rate tally
Parameters
----------
materials : iterable of :class:`openmc.lib.Material`
Materials to be used in :class:`openmc.lib.MaterialFilter`
mat_indexes : iterable of int
Indices of tallied materials that will have their fission
yields computed by this helper. Necessary as the
:class:`openmc.deplete.Operator` that uses this helper
may only burn a subset of all materials when running
in parallel mode.
"""
self._local_indexes = asarray(mat_indexes)
# Tally group-wise fission reaction rates
self._fission_rate_tally = Tally()
self._fission_rate_tally.writable = False
self._fission_rate_tally.scores = ['fission']
self._fission_rate_tally.filters = [MaterialFilter(materials)]