def __init__(self, filename):
# A user may not have h5py, but they can still use the rest of the
# Python API so we'll only try to import h5py if the user actually inits
# a Summary object.
import h5py
if h5py.__version__ == '2.6.0':
raise ImportError("h5py 2.6.0 has a known bug which makes it "
"incompatible with OpenMC's HDF5 files. "
"Please switch to a different version.")
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
self._openmc_geometry = None
self._opencg_geometry = None
self._read_metadata()
self._read_nuclides()
self._read_geometry()
self._read_tallies()
self._f.close()
def pincell_model():
"""Set up a model to test with and delete files when done"""
openmc.reset_auto_ids()
pincell = openmc.examples.pwr_pin_cell()
pincell.settings.verbosity = 1
# Add a tally
filter1 = openmc.MaterialFilter(pincell.materials)
filter2 = openmc.EnergyFilter([0.0, 1.0, 1.0e3, 20.0e6])
mat_tally = openmc.Tally()
mat_tally.filters = [filter1, filter2]
mat_tally.nuclides = ['U235', 'U238']
mat_tally.scores = ['total', 'elastic', '(n,gamma)']
pincell.tallies.append(mat_tally)
# Add an expansion tally
zernike_tally = openmc.Tally()
filter3 = openmc.ZernikeFilter(5, r=.63)
cells = pincell.geometry.root_universe.cells
filter4 = openmc.CellFilter(list(cells.values()))
zernike_tally.filters = [filter3, filter4]
zernike_tally.scores = ['fission']
pincell.tallies.append(zernike_tally)
# Add an energy function tally
energyfunc_tally = openmc.Tally()
energyfunc_filter = openmc.EnergyFunctionFilter([0.0, 20e6], [0.0, 20e6])
energyfunc_tally.scores = ['fission']
energyfunc_tally.filters = [energyfunc_filter]
pincell.tallies.append(energyfunc_tally)
# Write XML files in tmpdir
with cdtemp():
pincell.export_to_xml()
yield
def test_init(run_in_tmpdir, pin_model_attributes, mpi_intracomm):
mats, geom, settings, tals, plots, _, _ = \
pin_model_attributes
openmc.reset_auto_ids()
# Check blank initialization of a model
test_model = openmc.Model()
assert test_model.geometry.root_universe is None
assert len(test_model.materials) == 0
ref_settings = openmc.Settings()
assert sorted(test_model.settings.__dict__.keys()) == \
sorted(ref_settings.__dict__.keys())
for ref_k, ref_v in ref_settings.__dict__.items():
assert test_model.settings.__dict__[ref_k] == ref_v
assert len(test_model.tallies) == 0
assert len(test_model.plots) == 0
assert test_model._materials_by_id == {}
assert test_model._materials_by_name == {}
assert test_model._cells_by_id == {}
assert test_model._cells_by_name == {}
assert test_model.is_initialized is False
# Now check proper init of an actual model. Assume no interference between
# parameters and so we can apply them all at once instead of testing one
# parameter initialization at a time
test_model = openmc.Model(geom, mats, settings, tals, plots)
assert test_model.geometry is geom
assert test_model.materials is mats
assert test_model.settings is settings
assert test_model.tallies is tals
assert test_model.plots is plots
assert test_model._materials_by_id == {1: mats[0], 2: mats[1], 3: mats[2]}
assert test_model._materials_by_name == {
'UO2': {mats[0]},
'Zirc': {mats[1]},
'Borated water': {mats[2]}
}
# The last cell is the one that contains the infinite fuel
assert test_model._cells_by_id == \
{2: geom.root_universe.cells[2], 3: geom.root_universe.cells[3],
4: geom.root_universe.cells[4],
1: geom.root_universe.cells[2].fill.cells[1]}
# No cell name for 2 and 3, so we expect a blank name to be assigned to
# cell 3 due to overwriting
assert test_model._cells_by_name == {
'fuel': {geom.root_universe.cells[2]},
'': {geom.root_universe.cells[3], geom.root_universe.cells[4]},
'inf fuel': {geom.root_universe.cells[2].fill.cells[1]}
}
assert test_model.is_initialized is False
# Finally test the parameter type checking by passing bad types and
# obtaining the right exception types
def_params = [geom, mats, settings, tals, plots]
for i in range(len(def_params)):
args = def_params.copy()
# Try an integer, as that is a bad type for all arguments
args[i] = i
with pytest.raises(TypeError):
test_model = openmc.Model(*args)
def __init__(self, filename):
# A user may not have h5py, but they can still use the rest of the
# Python API so we'll only try to import h5py if the user actually inits
# a Summary object.
import h5py
if h5py.__version__ == '2.6.0':
raise ImportError("h5py 2.6.0 has a known bug which makes it "
"incompatible with OpenMC's HDF5 files. "
"Please switch to a different version.")
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
self._openmc_geometry = None
self._opencg_geometry = None
self._read_metadata()
self._read_nuclides()
self._read_geometry()
self._read_tallies()
self._f.close()
def setup_regression_test(request):
# Reset autogenerated IDs assigned to OpenMC objects
openmc.reset_auto_ids()
# Change to test directory
olddir = request.fspath.dirpath().chdir()
try:
yield
finally:
olddir.chdir()
def model(request):
openmc.reset_auto_ids()
marker = request.node.get_closest_marker("surf_source_op")
surf_source_op = marker.args[0]
openmc_model = openmc.model.Model()
# Materials
# None
# Geometry
# Concentric void spheres
# - Innermost sphere to bank surface sources
# - Second shell to tally cell flux
# - Outermost sphere as vacuum boundary
sph_1 = openmc.Sphere(r=1.0) # Surface to bank/write sources.
sph_2 = openmc.Sphere(r=2.0)
sph_3 = openmc.Sphere(r=2.5)
sph_4 = openmc.Sphere(r=4.0, boundary_type='vacuum')
cell_1 = openmc.Cell(region=-sph_1)
cell_2 = openmc.Cell(region=+sph_1 & -sph_2)
cell_3 = openmc.Cell(region=+sph_2 & -sph_3) # Cell to tally flux.
cell_4 = openmc.Cell(region=+sph_3 & -sph_4)
root = openmc.Universe(cells=[cell_1, cell_2, cell_3, cell_4])
openmc_model.geometry = openmc.Geometry(root)
# Settings
openmc_model.settings.run_mode = 'fixed source'
openmc_model.settings.particles = 1000
openmc_model.settings.batches = 10
openmc_model.settings.seed = 1
if surf_source_op == 'write':
point = openmc.stats.Point((0, 0, 0))
pt_src = openmc.Source(space=point)
openmc_model.settings.source = pt_src
openmc_model.settings.surf_source_write = {
'surface_ids': [1],
'max_particles': 1000
}
elif surf_source_op == 'read':
openmc_model.settings.surf_source_read = {
'path': 'surface_source_true.h5'
}
# Tallies
tal = openmc.Tally()
cell_filter = openmc.CellFilter(cell_3)
tal.filters = [cell_filter]
tal.scores = ['flux']
openmc_model.tallies.append(tal)
return openmc_model
def pincell_model_w_univ():
"""Set up a model to test with and delete files when done"""
openmc.reset_auto_ids()
pincell = openmc.examples.pwr_pin_cell()
clad_univ = openmc.Universe(cells=[openmc.Cell(fill=pincell.materials[1])])
pincell.geometry.root_universe.cells[2].fill = clad_univ
pincell.settings.verbosity = 1
# Write XML files in tmpdir
with cdtemp():
pincell.export_to_xml()
yield
def __call__(self, vec, power):
"""Runs a simulation.
Simulation will abort under the following circumstances:
1) No energy is computed using OpenMC tallies.
Parameters
----------
vec : list of numpy.ndarray
Total atoms to be used in function.
power : float
Power of the reactor in [W]
Returns
-------
openmc.deplete.OperatorResult
Eigenvalue and reaction rates resulting from transport operator
"""
# Reset results in OpenMC
openmc.lib.reset()
# If the source rate is zero, return zero reaction rates without running
# a transport solve
if power == 0.0:
rates = self.reaction_rates.copy()
rates.fill(0.0)
return OperatorResult(ufloat(0.0, 0.0), rates)
# Prevent OpenMC from complaining about re-creating tallies
openmc.reset_auto_ids()
# Update status
self.number.set_density(vec)
# Update material compositions and tally nuclides
self._update_materials()
nuclides = self._get_tally_nuclides()
self._rate_helper.nuclides = nuclides
self._energy_helper.nuclides = nuclides
self._yield_helper.update_tally_nuclides(nuclides)
# Run OpenMC
openmc.lib.run()
openmc.lib.reset_timers()
# Extract results
op_result = self._unpack_tallies_and_normalize(power)
return copy.deepcopy(op_result)
def __init__(self, filename):
# A user may not have h5py, but they can still use the rest of the
# Python API so we'll only try to import h5py if the user actually inits
# a Summary object.
import h5py
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
self.openmc_geometry = None
self.opencg_geometry = None
self._read_metadata()
self._read_geometry()
self._read_tallies()
def __init__(self, filename):
# A user may not have h5py, but they can still use the rest of the
# Python API so we'll only try to import h5py if the user actually inits
# a Summary object.
import h5py
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
self._openmc_geometry = None
self._opencg_geometry = None
self._read_metadata()
self._read_geometry()
self._read_tallies()
def test_import_properties(run_in_tmpdir, mpi_intracomm):
"""Test importing properties on the Model class """
# Create PWR pin cell model and write XML files
openmc.reset_auto_ids()
model = openmc.examples.pwr_pin_cell()
model.init_lib(output=False, intracomm=mpi_intracomm)
# Change fuel temperature and density and export properties
cell = openmc.lib.cells[1]
cell.set_temperature(600.0)
cell.fill.set_density(5.0, 'g/cm3')
openmc.lib.export_properties(output=False)
# Import properties to existing model
model.import_properties("properties.h5")
# Check to see that values are assigned to the C and python representations
# First python
cell = model.geometry.get_all_cells()[1]
assert cell.temperature == [600.0]
assert cell.fill.get_mass_density() == pytest.approx(5.0)
# Now C
assert openmc.lib.cells[1].get_temperature() == 600.
assert openmc.lib.materials[1].get_density('g/cm3') == pytest.approx(5.0)
# Clear the C API
openmc.lib.finalize()
# Verify the attributes survived by exporting to XML and re-creating
model.export_to_xml("with_properties")
# Load model with properties and confirm temperature/density changed
model_with_properties = openmc.Model.from_xml(
'with_properties/geometry.xml', 'with_properties/materials.xml',
'with_properties/settings.xml')
cell = model_with_properties.geometry.get_all_cells()[1]
assert cell.temperature == [600.0]
assert cell.fill.get_mass_density() == pytest.approx(5.0)
def __init__(self, filename):
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
cv.check_filetype_version(self._f, 'summary', _VERSION_SUMMARY)
self._geometry = openmc.Geometry()
self._fast_materials = {}
self._fast_surfaces = {}
self._fast_cells = {}
self._fast_universes = {}
self._fast_lattices = {}
self._materials = openmc.Materials()
self._nuclides = {}
self._read_nuclides()
self._read_geometry()
def __init__(self, filename):
openmc.reset_auto_ids()
if not filename.endswith(('.h5', '.hdf5')):
msg = 'Unable to open "{0}" which is not an HDF5 summary file'
raise ValueError(msg)
self._f = h5py.File(filename, 'r')
cv.check_filetype_version(self._f, 'summary', _VERSION_SUMMARY)
self._geometry = openmc.Geometry()
self._fast_materials = {}
self._fast_surfaces = {}
self._fast_cells = {}
self._fast_universes = {}
self._fast_lattices = {}
self._materials = openmc.Materials()
self._nuclides = {}
self._read_nuclides()
self._read_geometry()
def test_unstructured_mesh(test_opts):
openmc.reset_auto_ids()
# skip the test if the library is not enabled
if test_opts['library'] == 'moab' and not openmc.lib._dagmc_enabled():
pytest.skip("DAGMC (and MOAB) mesh not enbaled in this build.")
if test_opts['library'] == 'libmesh' and not openmc.lib._libmesh_enabled():
pytest.skip("LibMesh is not enabled in this build.")
# skip the tracklength test for libmesh
if test_opts['library'] == 'libmesh' and \
test_opts['estimator'] == 'tracklength':
pytest.skip("Tracklength tallies are not supported using libmesh.")
### Materials ###
materials = openmc.Materials()
fuel_mat = openmc.Material(name="fuel")
fuel_mat.add_nuclide("U235", 1.0)
fuel_mat.set_density('g/cc', 4.5)
materials.append(fuel_mat)
zirc_mat = openmc.Material(name="zircaloy")
zirc_mat.add_element("Zr", 1.0)
zirc_mat.set_density("g/cc", 5.77)
materials.append(zirc_mat)
water_mat = openmc.Material(name="water")
water_mat.add_nuclide("H1", 2.0)
water_mat.add_nuclide("O16", 1.0)
water_mat.set_density("atom/b-cm", 0.07416)
materials.append(water_mat)
materials.export_to_xml()
### Geometry ###
fuel_min_x = openmc.XPlane(-5.0, name="minimum x")
fuel_max_x = openmc.XPlane(5.0, name="maximum x")
fuel_min_y = openmc.YPlane(-5.0, name="minimum y")
fuel_max_y = openmc.YPlane(5.0, name="maximum y")
fuel_min_z = openmc.ZPlane(-5.0, name="minimum z")
fuel_max_z = openmc.ZPlane(5.0, name="maximum z")
fuel_cell = openmc.Cell(name="fuel")
fuel_cell.region = +fuel_min_x & -fuel_max_x & \
+fuel_min_y & -fuel_max_y & \
+fuel_min_z & -fuel_max_z
fuel_cell.fill = fuel_mat
clad_min_x = openmc.XPlane(-6.0, name="minimum x")
clad_max_x = openmc.XPlane(6.0, name="maximum x")
clad_min_y = openmc.YPlane(-6.0, name="minimum y")
clad_max_y = openmc.YPlane(6.0, name="maximum y")
clad_min_z = openmc.ZPlane(-6.0, name="minimum z")
clad_max_z = openmc.ZPlane(6.0, name="maximum z")
clad_cell = openmc.Cell(name="clad")
clad_cell.region = (-fuel_min_x | +fuel_max_x |
-fuel_min_y | +fuel_max_y |
-fuel_min_z | +fuel_max_z) & \
(+clad_min_x & -clad_max_x &
+clad_min_y & -clad_max_y &
+clad_min_z & -clad_max_z)
clad_cell.fill = zirc_mat
if test_opts['external_geom']:
bounds = (15, 15, 15)
else:
bounds = (10, 10, 10)
water_min_x = openmc.XPlane(x0=-bounds[0],
name="minimum x",
boundary_type='vacuum')
water_max_x = openmc.XPlane(x0=bounds[0],
name="maximum x",
boundary_type='vacuum')
water_min_y = openmc.YPlane(y0=-bounds[1],
name="minimum y",
boundary_type='vacuum')
water_max_y = openmc.YPlane(y0=bounds[1],
name="maximum y",
boundary_type='vacuum')
water_min_z = openmc.ZPlane(z0=-bounds[2],
name="minimum z",
boundary_type='vacuum')
water_max_z = openmc.ZPlane(z0=bounds[2],
name="maximum z",
boundary_type='vacuum')
water_cell = openmc.Cell(name="water")
water_cell.region = (-clad_min_x | +clad_max_x |
-clad_min_y | +clad_max_y |
-clad_min_z | +clad_max_z) & \
(+water_min_x & -water_max_x &
+water_min_y & -water_max_y &
+water_min_z & -water_max_z)
water_cell.fill = water_mat
# create a containing universe
geometry = openmc.Geometry([fuel_cell, clad_cell, water_cell])
### Tallies ###
# create meshes and mesh filters
regular_mesh = openmc.RegularMesh()
regular_mesh.dimension = (10, 10, 10)
regular_mesh.lower_left = (-10.0, -10.0, -10.0)
regular_mesh.upper_right = (10.0, 10.0, 10.0)
regular_mesh_filter = openmc.MeshFilter(mesh=regular_mesh)
if test_opts['holes']:
mesh_filename = "test_mesh_tets_w_holes.e"
else:
mesh_filename = "test_mesh_tets.e"
uscd_mesh = openmc.UnstructuredMesh(mesh_filename, test_opts['library'])
uscd_filter = openmc.MeshFilter(mesh=uscd_mesh)
# create tallies
tallies = openmc.Tallies()
regular_mesh_tally = openmc.Tally(name="regular mesh tally")
regular_mesh_tally.filters = [regular_mesh_filter]
regular_mesh_tally.scores = ['flux']
regular_mesh_tally.estimator = test_opts['estimator']
tallies.append(regular_mesh_tally)
uscd_tally = openmc.Tally(name="unstructured mesh tally")
uscd_tally.filters = [uscd_filter]
uscd_tally.scores = ['flux']
uscd_tally.estimator = test_opts['estimator']
tallies.append(uscd_tally)
### Settings ###
settings = openmc.Settings()
settings.run_mode = 'fixed source'
settings.particles = 1000
settings.batches = 10
# source setup
r = openmc.stats.Uniform(a=0.0, b=0.0)
theta = openmc.stats.Discrete(x=[0.0], p=[1.0])
phi = openmc.stats.Discrete(x=[0.0], p=[1.0])
space = openmc.stats.SphericalIndependent(r, theta, phi)
energy = openmc.stats.Discrete(x=[15.e+06], p=[1.0])
source = openmc.Source(space=space, energy=energy)
settings.source = source
model = openmc.model.Model(geometry=geometry,
materials=materials,
tallies=tallies,
settings=settings)
harness = UnstructuredMeshTest('statepoint.10.h5', model,
test_opts['inputs_true'],
test_opts['holes'])
harness.main()
def model():
openmc.reset_auto_ids()
model = openmc.Model()
# materials (M4 steel alloy)
m4 = openmc.Material()
m4.set_density('g/cc', 2.3)
m4.add_nuclide('H1', 0.168018676)
m4.add_nuclide("H2", 1.93244e-05)
m4.add_nuclide("O16", 0.561814465)
m4.add_nuclide("O17", 0.00021401)
m4.add_nuclide("Na23", 0.021365)
m4.add_nuclide("Al27", 0.021343)
m4.add_nuclide("Si28", 0.187439342)
m4.add_nuclide("Si29", 0.009517714)
m4.add_nuclide("Si30", 0.006273944)
m4.add_nuclide("Ca40", 0.018026179)
m4.add_nuclide("Ca42", 0.00012031)
m4.add_nuclide("Ca43", 2.51033e-05)
m4.add_nuclide("Ca44", 0.000387892)
m4.add_nuclide("Ca46", 7.438e-07)
m4.add_nuclide("Ca48", 3.47727e-05)
m4.add_nuclide("Fe54", 0.000248179)
m4.add_nuclide("Fe56", 0.003895875)
m4.add_nuclide("Fe57", 8.99727e-05)
m4.add_nuclide("Fe58", 1.19737e-05)
s0 = openmc.Sphere(r=240)
s1 = openmc.Sphere(r=250, boundary_type='vacuum')
c0 = openmc.Cell(fill=m4, region=-s0)
c1 = openmc.Cell(region=+s0 & -s1)
model.geometry = openmc.Geometry([c0, c1])
# settings
settings = model.settings
settings.run_mode = 'fixed source'
settings.particles = 500
settings.batches = 2
settings.max_splits = 100
settings.photon_transport = True
space = Point((0.001, 0.001, 0.001))
energy = Discrete([14E6], [1.0])
settings.source = openmc.Source(space=space, energy=energy)
# tally
mesh = openmc.RegularMesh()
mesh.lower_left = (-240, -240, -240)
mesh.upper_right = (240, 240, 240)
mesh.dimension = (3, 5, 7)
mesh_filter = openmc.MeshFilter(mesh)
e_bnds = [0.0, 0.5, 2E7]
energy_filter = openmc.EnergyFilter(e_bnds)
particle_filter = openmc.ParticleFilter(['neutron', 'photon'])
tally = openmc.Tally()
tally.filters = [mesh_filter, energy_filter, particle_filter]
tally.scores = ['flux']
model.tallies.append(tally)
return model
def __init__(self,
geometry,
settings,
chain_file=None,
prev_results=None,
diff_burnable_mats=False,
energy_mode="fission-q",
fission_q=None,
dilute_initial=1.0e3,
fission_yield_mode="constant",
fission_yield_opts=None):
if fission_yield_mode not in self._fission_helpers:
raise KeyError(
"fission_yield_mode must be one of {}, not {}".format(
", ".join(self._fission_helpers), fission_yield_mode))
if energy_mode == "energy-deposition":
if fission_q is not None:
warn("Fission Q dictionary not used if energy deposition "
"is used")
fission_q = None
elif energy_mode != "fission-q":
raise ValueError(
"energy_mode {} not supported. Must be energy-deposition "
"or fission-q".format(energy_mode))
super().__init__(chain_file, fission_q, dilute_initial, prev_results)
self.round_number = False
self.prev_res = None
self.settings = settings
self.geometry = geometry
self.diff_burnable_mats = diff_burnable_mats
# Differentiate burnable materials with multiple instances
if self.diff_burnable_mats:
self._differentiate_burnable_mats()
# Clear out OpenMC, create task lists, distribute
openmc.reset_auto_ids()
self.burnable_mats, volume, nuclides = self._get_burnable_mats()
self.local_mats = _distribute(self.burnable_mats)
# Generate map from local materials => material index
self._mat_index_map = {
lm: self.burnable_mats.index(lm)
for lm in self.local_mats
}
if self.prev_res is not None:
# Reload volumes into geometry
prev_results[-1].transfer_volumes(geometry)
# Store previous results in operator
# Distribute reaction rates according to those tracked
# on this process
if comm.size == 1:
self.prev_res = prev_results
else:
self.prev_res = ResultsList()
mat_indexes = _distribute(range(len(self.burnable_mats)))
for res_obj in prev_results:
new_res = res_obj.distribute(self.local_mats, mat_indexes)
self.prev_res.append(new_res)
# Determine which nuclides have incident neutron data
self.nuclides_with_data = self._get_nuclides_with_data()
# Select nuclides with data that are also in the chain
self._burnable_nucs = [
nuc.name for nuc in self.chain.nuclides
if nuc.name in self.nuclides_with_data
]
# Extract number densities from the geometry / previous depletion run
self._extract_number(self.local_mats, volume, nuclides, self.prev_res)
# Create reaction rates array
self.reaction_rates = ReactionRates(self.local_mats,
self._burnable_nucs,
self.chain.reactions)
# Get classes to assist working with tallies
self._rate_helper = DirectReactionRateHelper(
self.reaction_rates.n_nuc, self.reaction_rates.n_react)
if energy_mode == "fission-q":
self._energy_helper = ChainFissionHelper()
else:
score = "heating" if settings.photon_transport else "heating-local"
self._energy_helper = EnergyScoreHelper(score)
# Select and create fission yield helper
fission_helper = self._fission_helpers[fission_yield_mode]
fission_yield_opts = ({} if fission_yield_opts is None else
fission_yield_opts)
self._yield_helper = fission_helper.from_operator(
self, **fission_yield_opts)
def clean_up_openmc():
""" Resets all automatic indexing in OpenMC, as these get in the way. """
openmc.reset_auto_ids()
def main():
parser = argparse.ArgumentParser()
parser.add_argument('-l', '--benchmarks', type=pathlib.Path,
default=pathlib.Path('benchmarks/lists/pst-short'),
help='List of benchmarks to run.')
parser.add_argument('-c', '--code', choices=['openmc', 'mcnp'],
default='openmc',
help='Code used to run benchmarks.')
parser.add_argument('-x', '--cross-sections', type=str,
default=os.getenv('OPENMC_CROSS_SECTIONS'),
help='OpenMC cross sections XML file.')
parser.add_argument('-s', '--suffix', type=str, default='70c',
help='MCNP cross section suffix')
parser.add_argument('-p', '--particles', type=int, default=10000,
help='Number of source particles.')
parser.add_argument('-b', '--batches', type=int, default=150,
help='Number of batches.')
parser.add_argument('-i', '--inactive', type=int, default=50,
help='Number of inactive batches.')
parser.add_argument('-m', '--max-batches', type=int, default=10000,
help='Maximum number of batches.')
parser.add_argument('-t', '--threshold', type=float, default=0.0001,
help='Value of the standard deviation trigger on '
'eigenvalue.')
parser.add_argument('-o', '--output-name', type=str,
help='Base filename for plot.')
parser.add_argument('-f', '--output-format', type=str, default='png',
help='File format for plot.')
args = parser.parse_args()
# Create timestamp
current_time = time.localtime()
timestamp = time.strftime("%Y-%m-%d-%H%M%S", current_time)
# Check that executable exists
executable = 'mcnp6' if args.code == 'mcnp' else 'openmc'
if not shutil.which(executable, os.X_OK):
msg = f'Unable to locate executable {executable} in path.'
raise IOError(msg)
# Create directory and set filename for results
os.makedirs('results', exist_ok=True)
outfile = f'results/{timestamp}.csv'
# Get a copy of the benchmarks repository
if not pathlib.Path('benchmarks').is_dir():
repo = 'https://github.com/mit-crpg/benchmarks.git'
subprocess.run(['git', 'clone', repo], check=True)
# Get the list of benchmarks to run
if not args.benchmarks.is_file():
msg = f'Unable to locate benchmark list {args.benchmarks}.'
raise IOError(msg)
with open(args.benchmarks) as f:
benchmarks = f.read().split()
# Prepare and run benchmarks
for benchmark in benchmarks:
path = pathlib.Path('benchmarks') / benchmark
if args.code == 'openmc':
openmc.reset_auto_ids()
# Remove old statepoint files
for f in path.glob('statepoint.*.h5'):
os.remove(f)
# Modify settings
settings = openmc.Settings.from_xml(path / 'settings.xml')
settings.particles = args.particles
settings.inactive = args.inactive
settings.batches = args.batches
settings.keff_trigger = {'type': 'std_dev',
'threshold': args.threshold}
settings.trigger_active = True
settings.trigger_max_batches = args.max_batches
settings.output = {'tallies': False}
settings.export_to_xml(path)
# Set path to cross sections XML
materials = openmc.Materials.from_xml(path / 'materials.xml')
materials.cross_sections = args.cross_sections
materials.export_to_xml(path / 'materials.xml')
# Run benchmark
openmc.run(cwd=path)
# Read k-effective mean and standard deviation from statepoint
filename = list(path.glob('statepoint.*.h5'))[0]
with openmc.StatePoint(filename) as sp:
mean = sp.k_combined.nominal_value
stdev = sp.k_combined.std_dev
elif args.code == 'mcnp':
# Read input file
with open(path / 'input', 'r') as f:
lines = f.readlines()
# Update criticality source card
line = f'kcode {args.particles} 1 {args.inactive} {args.batches}\n'
for i in range(len(lines)):
if lines[i].strip().startswith('kcode'):
lines[i] = line
break
# Update cross section suffix
match = '(7[0-4]c)|(8[0-6]c)'
if not re.match(match, args.suffix):
msg = f'Unsupported cross section suffix {args.suffix}.'
raise ValueError(msg)
lines = [re.sub(match, args.suffix, x) for x in lines]
# Write new input file
with open(path / 'input', 'w') as f:
f.write(''.join(lines))
# Remove old MCNP output files
for f in ('outp', 'runtpe', 'srctp'):
try:
os.remove(path / f)
except OSError:
pass
# Run benchmark and capture and print output
arg_list = [executable, 'inp=input']
p = subprocess.Popen(
args=arg_list, cwd=path, stdout=subprocess.PIPE,
stderr=subprocess.STDOUT, universal_newlines=True
)
while True:
line = p.stdout.readline()
if not line and p.poll() is not None:
break
print(line, end='')
# Read k-effective mean and standard deviation from output
with open(path / 'outp', 'r') as f:
line = f.readline()
while not line.strip().startswith('col/abs/trk len'):
line = f.readline()
words = line.split()
mean = float(words[2])
stdev = float(words[3])
# Write results
words = benchmark.split('/')
name = words[1]
case = words[3] if len(words) > 3 else ''
line = '{}, {}, {}, {}'.format(name, case, mean, stdev)
if benchmark != benchmarks[-1]:
line += '\n'
with open(outfile, 'a') as f:
f.write(line)
plot(outfile, output_name=args.output_name,
output_format=args.output_format)
def main():
parser = argparse.ArgumentParser()
parser.add_argument('-l',
'--benchmarks',
type=Path,
default=Path('benchmarks/lists/pst-short'),
help='List of benchmarks to run.')
parser.add_argument('-c',
'--code',
choices=['openmc', 'mcnp'],
default='openmc',
help='Code used to run benchmarks.')
parser.add_argument('-x',
'--cross-sections',
type=str,
default=os.getenv('OPENMC_CROSS_SECTIONS'),
help='OpenMC cross sections XML file.')
parser.add_argument('-s',
'--suffix',
type=str,
default='70c',
help='MCNP cross section suffix')
parser.add_argument('-p',
'--particles',
type=int,
default=10000,
help='Number of source particles.')
parser.add_argument('-b',
'--batches',
type=int,
default=150,
help='Number of batches.')
parser.add_argument('-i',
'--inactive',
type=int,
default=50,
help='Number of inactive batches.')
parser.add_argument('-m',
'--max-batches',
type=int,
default=10000,
help='Maximum number of batches.')
parser.add_argument(
'-t',
'--threshold',
type=float,
default=0.0001,
help='Value of the standard deviation trigger on eigenvalue.')
parser.add_argument(
'--mpi-args',
default="",
help="MPI execute command and any additional MPI arguments")
args = parser.parse_args()
# Create timestamp
timestamp = time.strftime("%Y-%m-%d-%H%M%S")
# Check that executable exists
executable = 'mcnp6' if args.code == 'mcnp' else 'openmc'
if not shutil.which(executable, os.X_OK):
msg = f'Unable to locate executable {executable} in path.'
raise IOError(msg)
mpi_args = args.mpi_args.split()
# Create directory and set filename for results
results_dir = Path('results')
results_dir.mkdir(exist_ok=True)
outfile = results_dir / f'{timestamp}.csv'
# Get a copy of the benchmarks repository
if not Path('benchmarks').is_dir():
repo = 'https://github.com/mit-crpg/benchmarks.git'
subprocess.run(['git', 'clone', repo], check=True)
# Get the list of benchmarks to run
if not args.benchmarks.is_file():
msg = f'Unable to locate benchmark list {args.benchmarks}.'
raise IOError(msg)
with open(args.benchmarks) as f:
benchmarks = [Path(line) for line in f.read().split()]
# Set cross sections
if args.cross_sections is not None:
os.environ["OPENMC_CROSS_SECTIONS"] = args.cross_sections
# Prepare and run benchmarks
for i, benchmark in enumerate(benchmarks):
print(f"{i + 1} {benchmark} ", end="", flush=True)
path = 'benchmarks' / benchmark
if args.code == 'openmc':
openmc.reset_auto_ids()
# Remove old statepoint files
for f in path.glob('statepoint.*.h5'):
os.remove(f)
# Modify settings
settings = openmc.Settings.from_xml(path / 'settings.xml')
settings.particles = args.particles
settings.inactive = args.inactive
settings.batches = args.batches
settings.keff_trigger = {
'type': 'std_dev',
'threshold': args.threshold
}
settings.trigger_active = True
settings.trigger_max_batches = args.max_batches
settings.output = {'tallies': False}
settings.export_to_xml(path)
# Re-generate materials if Python script is present
genmat_script = path / "generate_materials.py"
if genmat_script.is_file():
subprocess.run(["python", "generate_materials.py"], cwd=path)
# Run benchmark
proc = subprocess.run(
mpi_args + ["openmc"],
cwd=path,
stdout=subprocess.PIPE,
stderr=subprocess.STDOUT,
universal_newlines=True,
)
# Determine last statepoint
t_last = 0
last_statepoint = None
for sp in path.glob('statepoint.*.h5'):
mtime = sp.stat().st_mtime
if mtime >= t_last:
t_last = mtime
last_statepoint = sp
# Read k-effective mean and standard deviation from statepoint
if last_statepoint is not None:
with openmc.StatePoint(last_statepoint) as sp:
mean = sp.k_combined.nominal_value
stdev = sp.k_combined.std_dev
else:
# Read input file
with open(path / 'input', 'r') as f:
lines = f.readlines()
# Update criticality source card
line = f'kcode {args.particles} 1 {args.inactive} {args.batches}\n'
for i in range(len(lines)):
if lines[i].strip().startswith('kcode'):
lines[i] = line
break
# Update cross section suffix
match = '(7[0-4]c)|(8[0-6]c)'
if not re.match(match, args.suffix):
msg = f'Unsupported cross section suffix {args.suffix}.'
raise ValueError(msg)
lines = [re.sub(match, args.suffix, x) for x in lines]
# Write new input file
with open(path / 'input', 'w') as f:
f.write(''.join(lines))
# Remove old MCNP output files
for f in ('outp', 'runtpe', 'srctp'):
try:
os.remove(path / f)
except OSError:
pass
# Run benchmark and capture and print output
arg_list = mpi_args + [executable, 'inp=input']
proc = subprocess.run(arg_list,
cwd=path,
stdout=subprocess.PIPE,
stderr=subprocess.STDOUT,
universal_newlines=True)
# Read k-effective mean and standard deviation from output
with open(path / 'outp', 'r') as f:
line = f.readline()
while not line.strip().startswith('col/abs/trk len'):
line = f.readline()
words = line.split()
mean = float(words[2])
stdev = float(words[3])
# Write output to file
with open(path / f"output_{timestamp}", "w") as fh:
fh.write(proc.stdout)
if proc.returncode != 0:
mean = stdev = ""
print()
else:
# Display k-effective
print(f"{mean:.5f} ± {stdev:.5f}")
# Write results
words = str(benchmark).split('/')
name = words[1]
case = '/' + words[3] if len(words) > 3 else ''
line = f'{name}{case},{mean},{stdev}\n'
with open(outfile, 'a') as f:
f.write(line)
def reset():
openmc.reset_auto_ids()
def __init__(self,
geometry,
settings,
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_value('fission yield mode', fission_yield_mode,
self._fission_helpers.keys())
check_value('normalization mode', normalization_mode,
('energy-deposition', 'fission-q', 'source-rate'))
if normalization_mode != "fission-q":
if fission_q is not None:
warn("Fission Q dictionary will not be used")
fission_q = None
super().__init__(chain_file, fission_q, dilute_initial, prev_results)
self.round_number = False
self.settings = settings
self.geometry = geometry
self.diff_burnable_mats = diff_burnable_mats
# Reduce the chain before we create more materials
if reduce_chain:
all_isotopes = set()
for material in geometry.get_all_materials().values():
if not material.depletable:
continue
for name, _dens_percent, _dens_type in material.nuclides:
all_isotopes.add(name)
self.chain = self.chain.reduce(all_isotopes, reduce_chain_level)
# Differentiate burnable materials with multiple instances
if self.diff_burnable_mats:
self._differentiate_burnable_mats()
# Clear out OpenMC, create task lists, distribute
openmc.reset_auto_ids()
self.burnable_mats, volume, nuclides = self._get_burnable_mats()
self.local_mats = _distribute(self.burnable_mats)
# Generate map from local materials => material index
self._mat_index_map = {
lm: self.burnable_mats.index(lm)
for lm in self.local_mats
}
if self.prev_res is not None:
# Reload volumes into geometry
prev_results[-1].transfer_volumes(geometry)
# Store previous results in operator
# Distribute reaction rates according to those tracked
# on this process
if comm.size == 1:
self.prev_res = prev_results
else:
self.prev_res = ResultsList()
mat_indexes = _distribute(range(len(self.burnable_mats)))
for res_obj in prev_results:
new_res = res_obj.distribute(self.local_mats, mat_indexes)
self.prev_res.append(new_res)
# Determine which nuclides have incident neutron data
self.nuclides_with_data = self._get_nuclides_with_data()
# Select nuclides with data that are also in the chain
self._burnable_nucs = [
nuc.name for nuc in self.chain.nuclides
if nuc.name in self.nuclides_with_data
]
# Extract number densities from the geometry / previous depletion run
self._extract_number(self.local_mats, volume, nuclides, self.prev_res)
# Create reaction rates array
self.reaction_rates = ReactionRates(self.local_mats,
self._burnable_nucs,
self.chain.reactions)
# Get classes to assist working with tallies
if reaction_rate_mode == "direct":
self._rate_helper = DirectReactionRateHelper(
self.reaction_rates.n_nuc, self.reaction_rates.n_react)
elif reaction_rate_mode == "flux":
if reaction_rate_opts is None:
reaction_rate_opts = {}
# Ensure energy group boundaries were specified
if 'energies' not in reaction_rate_opts:
raise ValueError(
"Energy group boundaries must be specified in the "
"reaction_rate_opts argument when reaction_rate_mode is"
"set to 'flux'.")
self._rate_helper = FluxCollapseHelper(self.reaction_rates.n_nuc,
self.reaction_rates.n_react,
**reaction_rate_opts)
else:
raise ValueError("Invalid reaction rate mode.")
if normalization_mode == "fission-q":
self._normalization_helper = ChainFissionHelper()
elif normalization_mode == "energy-deposition":
score = "heating" if settings.photon_transport else "heating-local"
self._normalization_helper = EnergyScoreHelper(score)
else:
self._normalization_helper = SourceRateHelper()
# Select and create fission yield helper
fission_helper = self._fission_helpers[fission_yield_mode]
fission_yield_opts = ({} if fission_yield_opts is None else
fission_yield_opts)
self._yield_helper = fission_helper.from_operator(
self, **fission_yield_opts)