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Use get_all_materials() to simplify logic

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This commit is contained in:
Paul Romano 2018-02-19 07:23:26 -06:00
parent 5c4ea0d640
commit 82fea220c4
1 changed files with 21 additions and 50 deletions
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71
openmc/deplete/openmc_wrapper.py
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@ -104,7 +104,7 @@ class OpenMCOperator(Operator):
----------
geometry : openmc.Geometry
The OpenMC geometry object.
settings : OpenMCSettings
settings : openmc.deplete.OpenMCSettings
Settings object.
Attributes
@ -130,8 +130,8 @@ class OpenMCOperator(Operator):
Number of nuclides considered in the decay chain.
mat_tally_ind : OrderedDict of str to int
Dictionary mapping material ID to index in tally.
"""
"""
def __init__(self, geometry, settings):
super().__init__(settings)
@ -170,8 +170,10 @@ class OpenMCOperator(Operator):
# Extract number densities from the geometry
self.extract_number(mat_burn, mat_not_burn, volume, nuc_dict)
# Create reaction rate tables
self.initialize_reaction_rates()
# Create reaction rates array
index_rx = {rx: i for i, rx in enumerate(self.chain.reactions)}
self.reaction_rates = ReactionRates(
self.burn_mat_to_ind, self.burn_nuc_to_ind, index_rx)
def __call__(self, vec, print_out=True):
"""Runs a simulation.
@ -243,35 +245,17 @@ class OpenMCOperator(Operator):
volume = OrderedDict()
# Iterate once through the geometry to get dictionaries
cells = self.geometry.get_all_material_cells()
for cell in cells.values():
if isinstance(cell.fill, openmc.Material):
mat = cell.fill
for nuclide in mat.get_nuclide_densities():
nuc_set.add(nuclide)
if mat.depletable:
mat_burn.add(str(mat.id))
volume[str(mat.id)] = mat.volume
else:
mat_not_burn.add(str(mat.id))
for mat in self.geometry.get_all_materials().values():
for nuclide in mat.get_nuclide_densities():
nuc_set.add(nuclide)
if mat.depletable:
mat_burn.add(str(mat.id))
if mat.volume is None:
raise RuntimeError("Volume not specified for depletable "
"material with ID={}.".format(mat.id))
volume[str(mat.id)] = mat.volume
else:
for mat in cell.fill:
for nuclide in mat.get_nuclide_densities():
nuc_set.add(nuclide)
if mat.depletable:
mat_burn.add(str(mat.id))
volume[str(mat.id)] = mat.volume
else:
mat_not_burn.add(str(mat.id))
need_vol = []
for mat_id in volume:
if volume[mat_id] is None:
need_vol.append(mat_id)
if need_vol:
exit("Need volumes for materials: " + str(need_vol))
mat_not_burn.add(str(mat.id))
# Sort the sets
mat_burn = sorted(mat_burn, key=int)
@ -334,18 +318,13 @@ class OpenMCOperator(Operator):
if self.settings.dilute_initial != 0.0:
for nuc in self.burn_nuc_to_ind:
self.number.set_atom_density(np.s_[:], nuc, self.settings.dilute_initial)
self.number.set_atom_density(np.s_[:], nuc,
self.settings.dilute_initial)
# Now extract the number densities and store
cells = self.geometry.get_all_material_cells()
for cell in cells.values():
if isinstance(cell.fill, openmc.Material):
if str(cell.fill.id) in mat_dict:
self.set_number_from_mat(cell.fill)
else:
for mat in cell.fill:
if str(mat.id) in mat_dict:
self.set_number_from_mat(mat)
for mat in self.geometry.get_all_materials().values():
if str(mat.id) in mat_dict:
self.set_number_from_mat(mat)
def set_number_from_mat(self, mat):
"""Extracts material and number densities from openmc.Material
@ -364,14 +343,6 @@ class OpenMCOperator(Operator):
number = nuc_dens[nuclide][1] * 1.0e24
self.number.set_atom_density(mat_id, nuclide, number)
def initialize_reaction_rates(self):
"""Create reaction rates object. """
# Create dictionary to map reactions to indices
index_rx = {rx: i for i, rx in enumerate(self.chain.reactions)}
self.reaction_rates = ReactionRates(
self.burn_mat_to_ind, self.burn_nuc_to_ind, index_rx)
def form_matrix(self, y, mat):
"""Forms the depletion matrix.