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Merge pull request #1894 from pshriwise/model_deplete

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Passing `openmc.Model` to depletion functions.
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Paul Romano 2021-11-09 06:51:09 -06:00 • committed by GitHub
commit 176b148b6b
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7 changed files with 65 additions and 32 deletions
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4
examples/pincell_depletion/run_depletion.py
View file

@ -84,9 +84,11 @@ settings.entropy_mesh = entropy_mesh
# Initialize and run depletion calculation
###############################################################################
model = openmc.Model(geometry=geometry, settings=settings)
# Create depletion "operator"
chain_file = './chain_simple.xml'
op = openmc.deplete.Operator(geometry, settings, chain_file)
op = openmc.deplete.Operator(model, chain_file)
# Perform simulation using the predictor algorithm
time_steps = [1.0, 1.0, 1.0, 1.0, 1.0] # days

58
openmc/deplete/operator.py
View file

@ -10,6 +10,7 @@ densities is all done in-memory instead of through the filesystem.
import copy
from collections import OrderedDict
import os
from typing import Type
import xml.etree.ElementTree as ET
from warnings import warn
from pathlib import Path
@ -65,12 +66,14 @@ class Operator(TransportOperator):
directly. Instead, an instance of this class is passed to an integrator
class, such as :class:`openmc.deplete.CECMIntegrator`.
.. versionchanged:: 0.13.0
The geometry and settings parameters have been replaced with a
model parameter that takes an openmc.Model object
Parameters
----------
geometry : openmc.Geometry
OpenMC geometry object
settings : openmc.Settings
OpenMC Settings object
model : openmc.Model
OpenMC model object
chain_file : str, optional
Path to the depletion chain XML file. Defaults to the file
listed under ``depletion_chain`` in
@ -144,6 +147,8 @@ class Operator(TransportOperator):
Attributes
----------
model : openmc.Model
OpenMC model object
geometry : openmc.Geometry
OpenMC geometry object
settings : openmc.Settings
@ -186,12 +191,21 @@ class Operator(TransportOperator):
"cutoff": FissionYieldCutoffHelper,
}
def __init__(self, geometry, settings, chain_file=None, prev_results=None,
def __init__(self, model, chain_file=None, prev_results=None,
diff_burnable_mats=False, normalization_mode="fission-q",
fission_q=None, dilute_initial=1.0e3,
fission_yield_mode="constant", fission_yield_opts=None,
reaction_rate_mode="direct", reaction_rate_opts=None,
reduce_chain=False, reduce_chain_level=None):
# check for old call to constructor
if isinstance(model, openmc.Geometry):
msg = "As of version 0.13.0 openmc.deplete.Operator requires an " \
"openmc.Model object rather than the openmc.Geometry and " \
"openmc.Settings parameters. Please use the geometry and " \
"settings objects passed here to create a model with which " \
"to generate the depletion Operator."
raise TypeError(msg)
check_value('fission yield mode', fission_yield_mode,
self._fission_helpers.keys())
check_value('normalization mode', normalization_mode,
@ -202,15 +216,24 @@ class Operator(TransportOperator):
fission_q = None
super().__init__(chain_file, fission_q, dilute_initial, prev_results)
self.round_number = False
self.settings = settings
self.geometry = geometry
self.model = model
self.settings = model.settings
self.geometry = model.geometry
# determine set of materials in the model
if not model.materials:
model.materials = openmc.Materials(
model.geometry.get_all_materials().values()
)
self.materials = model.materials
self.diff_burnable_mats = diff_burnable_mats
self.cleanup_when_done = True
# Reduce the chain before we create more materials
if reduce_chain:
all_isotopes = set()
for material in geometry.get_all_materials().values():
for material in self.materials:
if not material.depletable:
continue
for name, _dens_percent, _dens_type in material.nuclides:
@ -232,7 +255,7 @@ class Operator(TransportOperator):
if self.prev_res is not None:
# Reload volumes into geometry
prev_results[-1].transfer_volumes(geometry)
prev_results[-1].transfer_volumes(self.model)
# Store previous results in operator
# Distribute reaction rates according to those tracked
@ -373,7 +396,7 @@ class Operator(TransportOperator):
# Extract all burnable materials which have multiple instances
distribmats = set(
[mat for mat in self.geometry.get_all_materials().values()
[mat for mat in self.materials
if mat.depletable and mat.num_instances > 1])
for mat in distribmats:
@ -411,7 +434,7 @@ class Operator(TransportOperator):
self.heavy_metal = 0.0
# Iterate once through the geometry to get dictionaries
for mat in self.geometry.get_all_materials().values():
for mat in self.materials:
for nuclide in mat.get_nuclides():
model_nuclides.add(nuclide)
if mat.depletable:
@ -463,13 +486,13 @@ class Operator(TransportOperator):
# Now extract and store the number densities
# From the geometry if no previous depletion results
if prev_res is None:
for mat in self.geometry.get_all_materials().values():
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_mat(mat)
# Else from previous depletion results
else:
for mat in self.geometry.get_all_materials().values():
for mat in self.materials:
if str(mat.id) in local_mats:
self._set_number_from_results(mat, prev_res)
@ -609,12 +632,9 @@ class Operator(TransportOperator):
through direct memory writing.
"""
materials = openmc.Materials(self.geometry.get_all_materials()
.values())
# Sort nuclides according to order in AtomNumber object
nuclides = list(self.number.nuclides)
for mat in materials:
for mat in self.materials:
mat._nuclides.sort(key=lambda x: nuclides.index(x[0]))
# Grab the cross sections tag from the existing file
@ -623,9 +643,9 @@ class Operator(TransportOperator):
tree = ET.parse(str(mfile))
xs = tree.find('cross_sections')
if xs is not None:
materials.cross_sections = xs.text
self.materials.cross_sections = xs.text
materials.export_to_xml()
self.materials.export_to_xml()
def _get_tally_nuclides(self):
"""Determine nuclides that should be tallied for reaction rates.

20
openmc/deplete/results.py
View file

@ -9,6 +9,7 @@ import copy
import h5py
import numpy as np
import openmc
from openmc.mpi import comm, MPI
from .reaction_rates import ReactionRates
@ -496,16 +497,23 @@ class Results:
results.export_to_hdf5("depletion_results.h5", step_ind)
def transfer_volumes(self, geometry):
def transfer_volumes(self, model):
"""Transfers volumes from depletion results to geometry
Parameters
----------
geometry : OpenMC geometry to be used in a depletion restart
model : OpenMC model to be used in a depletion restart
calculation
"""
for cell in geometry.get_all_material_cells().values():
for material in cell.get_all_materials().values():
if material.depletable:
material.volume = self.volume[str(material.id)]
if not model.materials:
materials = openmc.Materials(
model.geometry.get_all_materials().values()
)
else:
materials = model.materials
for material in materials:
if material.depletable:
material.volume = self.volume[str(material.id)]

2
openmc/model/model.py
View file

@ -326,7 +326,7 @@ class Model:
with _change_directory(Path(directory)):
with openmc.lib.quiet_dll(output):
depletion_operator = \
dep.Operator(self.geometry, self.settings, **op_kwargs)
dep.Operator(self, **op_kwargs)
# Tell depletion_operator.finalize NOT to clear C API memory when
# it is done

6
tests/regression_tests/deplete/test.py
View file

@ -49,9 +49,11 @@ def test_full(run_in_tmpdir, problem, multiproc):
settings.seed = 1
settings.verbosity = 1
model = openmc.Model(geometry=geometry, settings=settings)
# Create operator
chain_file = Path(__file__).parents[2] / 'chain_simple.xml'
op = openmc.deplete.Operator(geometry, settings, chain_file)
op = openmc.deplete.Operator(model, chain_file)
op.round_number = True
# Power and timesteps
@ -170,7 +172,7 @@ def test_depletion_results_to_material(run_in_tmpdir, problem):
diff_vs_expected = unified_diff(reference_lines, result_file_lines)
# Check all lines match, printing errors along the way
success = True
success = True
for line in diff_vs_expected:
success = False
print(line.rstrip())

5
tests/unit_tests/test_deplete_activation.py
View file

@ -65,7 +65,7 @@ def test_activation(run_in_tmpdir, model, reaction_rate_mode, reaction_rate_opts
# Create transport operator
op = openmc.deplete.Operator(
model.geometry, model.settings, 'test_chain.xml',
model, 'test_chain.xml',
normalization_mode="source-rate",
reaction_rate_mode=reaction_rate_mode,
reaction_rate_opts=reaction_rate_opts,
@ -142,9 +142,10 @@ def test_decay(run_in_tmpdir):
chain.add_nuclide(sr89)
chain.export_to_xml('test_chain.xml')
model = openmc.Model(geometry=geometry, settings=settings)
# Create transport operator
op = openmc.deplete.Operator(
geometry, settings, 'test_chain.xml', normalization_mode="source-rate"
model, 'test_chain.xml', normalization_mode="source-rate"
)
# Deplete with two decay steps

2
tests/unit_tests/test_transfer_volumes.py
View file

@ -40,7 +40,7 @@ def test_transfer_volumes(run_in_tmpdir):
geometry = openmc.Geometry(root)
# Transfer volumes
res[0].transfer_volumes(geometry)
res[0].transfer_volumes(openmc.Model(geometry))
assert mat1.volume == 1.5
assert mat2.volume is None