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Rename participating_nuclides -> nuclides_with_data

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This commit is contained in:
Paul Romano 2018-02-19 13:10:28 -06:00
parent ea335e0696
commit 78e1afb0ff
3 changed files with 60 additions and 64 deletions
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14
openmc/deplete/atom_number.py
View file

@ -96,9 +96,9 @@ class AtomNumber(object):
These indexes can be strings (which get converted to integers via
the dictionaries), integers used directly, or slices.
val : float
The value to set the array to.
"""
The value [atom] to set the array to.
"""
mat, nuc = pos
if isinstance(mat, str):
mat = self.mat_to_ind[mat]
@ -119,10 +119,10 @@ class AtomNumber(object):
Returns
-------
numpy.array
The density indexed.
"""
numpy.ndarray
Density in [atom/cm^3]
"""
if isinstance(mat, str):
mat = self.mat_to_ind[mat]
if isinstance(nuc, str):
@ -140,9 +140,9 @@ class AtomNumber(object):
nuc : str, int or slice
Nuclide index.
val : numpy.array
Array of values to set.
"""
Array of densities to set in [atom/cm^3]
"""
if isinstance(mat, str):
mat = self.mat_to_ind[mat]
if isinstance(nuc, str):

7
openmc/deplete/chain.py
View file

@ -332,10 +332,11 @@ class Chain(object):
root = ET.parse(filename)
except Exception:
if filename is None:
print("No chain specified, either manually or in environment variable OPENMC_DEPLETE_CHAIN.")
msg = ("No chain specified, either manually or in environment "
"variable OPENMC_DEPLETE_CHAIN.")
else:
print('Decay chain "{}" is invalid.'.format(filename))
raise
msg = 'Decay chain "{}" is invalid.'.format(filename)
raise IOError(msg)
for i, nuclide_elem in enumerate(root.findall('nuclide_table')):
nuc = Nuclide.from_xml(nuclide_elem)

103
openmc/deplete/openmc_wrapper.py
View file

@ -116,7 +116,7 @@ class OpenMCOperator(Operator):
The OpenMC geometry object.
number : openmc.deplete.AtomNumber
Total number of atoms in simulation.
participating_nuclides : set of str
nuclides_with_data : set of str
A set listing all unique nuclides available from cross_sections.xml.
chain : openmc.deplete.Chain
The depletion chain information necessary to form matrices and tallies.
@ -126,7 +126,8 @@ class OpenMCOperator(Operator):
Dictionary mapping material ID (as a string) to an index in reaction_rates.
burn_nuc_to_ind : OrderedDict of str to int
Dictionary mapping nuclide name (as a string) to an index in
reaction_rates.
reaction_rates. Consists of all nuclides with neutron data and appearing
in the depletion chain.
burnable_mats : list of str
All burnable material IDs
@ -136,7 +137,6 @@ class OpenMCOperator(Operator):
self.geometry = geometry
self.number = None
self.participating_nuclides = None
self.burn_mat_to_ind = OrderedDict()
self.burn_nuc_to_ind = None
@ -146,7 +146,7 @@ class OpenMCOperator(Operator):
# Clear out OpenMC, create task lists, distribute
if comm.rank == 0:
openmc.reset_auto_ids()
mat_burn_list, volume, nuc_dict = self.extract_mat_ids()
mat_burn_list, volume, nuc_dict = self._extract_mat_ids()
else:
# Dummy variables
mat_burn_list = None
@ -160,10 +160,10 @@ class OpenMCOperator(Operator):
self.burnable_mats = list(chain(*mat_burn_list))
# Load participating nuclides
self.load_participating()
self._load_participating()
# Extract number densities from the geometry
self.extract_number(mat_burn, volume, nuc_dict)
self._extract_number(mat_burn, volume, nuc_dict)
# Create reaction rates array
index_rx = {rx: i for i, rx in enumerate(self.chain.reactions)}
@ -190,7 +190,7 @@ class OpenMCOperator(Operator):
openmc.reset_auto_ids()
# Update status
self.set_density(vec)
self._set_density(vec)
time_start = time.time()
@ -205,7 +205,7 @@ class OpenMCOperator(Operator):
time_openmc = time.time()
# Extract results
op_result = self.unpack_tallies_and_normalize()
op_result = self._unpack_tallies_and_normalize()
if comm.rank == 0:
time_unpack = time.time()
@ -216,7 +216,7 @@ class OpenMCOperator(Operator):
return copy.deepcopy(op_result)
def extract_mat_ids(self):
def _extract_mat_ids(self):
"""Extracts materials and assigns them to processes.
Returns
@ -229,15 +229,15 @@ class OpenMCOperator(Operator):
Volume of each cell
nuc_dict : OrderedDict of str to int
Nuclides in order of how they'll appear in the simulation.
"""
"""
mat_burn = set()
nuc_set = set()
volume = OrderedDict()
# Iterate once through the geometry to get dictionaries
for mat in self.geometry.get_all_materials().values():
for nuclide in mat.get_nuclide_densities():
for nuclide in mat.get_nuclides():
nuc_set.add(nuclide)
if mat.depletable:
mat_burn.add(str(mat.id))
@ -263,7 +263,7 @@ class OpenMCOperator(Operator):
return mat_burn_lists, volume, nuc_dict
def extract_number(self, mat_burn, volume, nuc_dict):
def _extract_number(self, mat_burn, volume, nuc_dict):
"""Construct self.number read from geometry
Parameters
@ -281,10 +281,8 @@ class OpenMCOperator(Operator):
for i, mat in enumerate(mat_burn):
self.burn_mat_to_ind[mat] = i
n_nuc_burn = len(self.chain)
self.number = AtomNumber(self.burn_mat_to_ind, nuc_dict, volume,
n_nuc_burn)
len(self.chain))
if self.settings.dilute_initial != 0.0:
for nuc in self.burn_nuc_to_ind:
@ -294,23 +292,22 @@ class OpenMCOperator(Operator):
# Now extract the number densities and store
for mat in self.geometry.get_all_materials().values():
if str(mat.id) in self.burn_mat_to_ind:
self.set_number_from_mat(mat)
self._set_number_from_mat(mat)
def set_number_from_mat(self, mat):
def _set_number_from_mat(self, mat):
"""Extracts material and number densities from openmc.Material
Parameters
----------
mat : openmc.Materials
mat : openmc.Material
The material to read from
"""
"""
mat_id = str(mat.id)
mat_ind = self.number.mat_to_ind[mat_id]
nuc_dens = mat.get_nuclide_atom_densities()
for nuclide in nuc_dens:
number = nuc_dens[nuclide][1] * 1.0e24
for nuclide, density in mat.get_nuclide_atom_densities().values():
number = density * 1.0e24
self.number.set_atom_density(mat_id, nuclide, number)
def form_matrix(self, y, mat):
@ -344,17 +341,17 @@ class OpenMCOperator(Operator):
if comm.rank == 0:
self.geometry.export_to_xml()
self.settings.settings.export_to_xml()
self.generate_materials_xml()
self._generate_materials_xml()
# Initialize OpenMC library
comm.barrier()
openmc.capi.init(comm)
# Generate tallies in memory
self.generate_tallies()
self._generate_tallies()
# Return number density vector
return self.total_density_list()
return list(self.number.get_mat_slice(np.s_[:]))
def finalize(self):
"""Finalize a depletion simulation and release resources."""
@ -370,7 +367,7 @@ class OpenMCOperator(Operator):
nuclides = []
densities = []
for nuc in number_i.nuc_to_ind:
if nuc in self.participating_nuclides:
if nuc in self.nuclides_with_data:
val = 1.0e-24 * number_i.get_atom_density(mat, nuc)
# If nuclide is zero, do not add to the problem.
@ -395,14 +392,14 @@ class OpenMCOperator(Operator):
mat_internal = openmc.capi.materials[int(mat)]
mat_internal.set_densities(nuclides, densities)
def generate_materials_xml(self):
def _generate_materials_xml(self):
"""Creates materials.xml from self.number.
Due to uncertainty with how MPI interacts with OpenMC API, this
constructs the XML manually. The long term goal is to do this
through direct memory writing.
"""
"""
materials = openmc.Materials(self.geometry.get_all_materials()
.values())
@ -414,12 +411,26 @@ class OpenMCOperator(Operator):
materials.export_to_xml()
def _get_tally_nuclides(self):
"""Determine nuclides that should be tallied for reaction rates.
This method returns a list of all nuclides that have neutron data and
are listed in the depletion chain. Technically, we should tally nuclides
that may not appear in the depletion chain because we still need to get
the fission reaction rate for these nuclides in order to normalize
power, but that is left as a future exercise.
Returns
-------
list of str
Tally nuclides
"""
nuc_set = set()
# Create the set of all nuclides in the decay chain in cells marked for
# burning in which the number density is greater than zero.
for nuc in self.number.nuc_to_ind:
if nuc in self.participating_nuclides:
if nuc in self.nuclides_with_data:
if np.sum(self.number[:, nuc]) > 0.0:
nuc_set.add(nuc)
@ -439,12 +450,10 @@ class OpenMCOperator(Operator):
nuc_list = None
# Store list of tally nuclides on each process
nuc_list = comm.bcast(nuc_list, root=0)
tally_nuclides = [nuc for nuc in nuc_list if nuc in self.chain]
nuc_list = comm.bcast(nuc_list)
return [nuc for nuc in nuc_list if nuc in self.chain]
return tally_nuclides
def generate_tallies(self):
def _generate_tallies(self):
"""Generates depletion tallies.
Using information from the depletion chain as well as the nuclides
@ -465,21 +474,7 @@ class OpenMCOperator(Operator):
tally_dep.scores = self.chain.reactions
tally_dep.filters = [mat_filter]
def total_density_list(self):
"""Returns a list of total density lists.
This list is in the exact same order as depletion_matrix_list, so that
matrix exponentiation can be done easily.
Returns
-------
list of numpy.array
A list of arrays containing total atoms of each material
"""
return list(self.number.get_mat_slice(np.s_[:]))
def set_density(self, total_density):
def _set_density(self, total_density):
"""Sets density.
Sets the density in the exact same order as total_density_list outputs,
@ -495,7 +490,7 @@ class OpenMCOperator(Operator):
for i in range(self.number.n_mat):
self.number.set_mat_slice(i, total_density[i])
def unpack_tallies_and_normalize(self):
def _unpack_tallies_and_normalize(self):
"""Unpack tallies from OpenMC and return an operator result
This method uses OpenMC's C API bindings to determine the k-effective
@ -587,7 +582,7 @@ class OpenMCOperator(Operator):
return OperatorResult(k_combined, rates)
def load_participating(self):
def _load_participating(self):
"""Loads a cross_sections.xml file to find participating nuclides.
This allows for nuclides that are important in the decay chain but not
@ -602,7 +597,7 @@ class OpenMCOperator(Operator):
except KeyError:
filename = None
self.participating_nuclides = set()
self.nuclides_with_data = set()
try:
tree = ET.parse(filename)
@ -624,8 +619,8 @@ class OpenMCOperator(Operator):
for name in mats.split():
# Make a burn list of the union of nuclides in cross_sections.xml
# and nuclides in depletion chain.
if name not in self.participating_nuclides:
self.participating_nuclides.add(name)
if name not in self.nuclides_with_data:
self.nuclides_with_data.add(name)
if name in self.chain:
self.burn_nuc_to_ind[name] = nuc_ind
nuc_ind += 1