from openmc.examples import slab_mg
if __name__ == '__main__':
model = slab_mg(as_macro=False)
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = [1, 1, 10]
mesh.lower_left = [0.0, 0.0, 0.0]
mesh.upper_right = [10, 10, 5]
# Instantiate some tally filters
energy_filter = openmc.EnergyFilter([0.0, 20.0e6])
energyout_filter = openmc.EnergyoutFilter([0.0, 20.0e6])
energies = [1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6]
matching_energy_filter = openmc.EnergyFilter(energies)
matching_eout_filter = openmc.EnergyoutFilter(energies)
mesh_filter = openmc.MeshFilter(mesh)
mat_filter = openmc.MaterialFilter(model.materials)
nuclides = model.xs_data
scores = {False: ['total', 'absorption', 'flux', 'fission', 'nu-fission'],
True: ['total', 'absorption', 'fission', 'nu-fission']}
for do_nuclides in [False, True]:
t = openmc.Tally()
t.filters = [mesh_filter]
def test_mg_tallies():
create_library()
model = slab_mg()
# Instantiate a tally mesh
mesh = openmc.RegularMesh(mesh_id=1)
mesh.dimension = [10, 1, 1]
mesh.lower_left = [0.0, 0.0, 0.0]
mesh.upper_right = [929.45, 1000, 1000]
# Instantiate some tally filters
energy_filter = openmc.EnergyFilter([0.0, 20.0e6])
energyout_filter = openmc.EnergyoutFilter([0.0, 20.0e6])
energies = [0.0, 0.625, 20.0e6]
matching_energy_filter = openmc.EnergyFilter(energies)
matching_eout_filter = openmc.EnergyoutFilter(energies)
mesh_filter = openmc.MeshFilter(mesh)
mat_filter = openmc.MaterialFilter(model.materials)
nuclides = model.xs_data
scores_with_nuclides = [
'total', 'absorption', 'fission', 'nu-fission', 'inverse-velocity',
'prompt-nu-fission', 'delayed-nu-fission', 'kappa-fission', 'events',
'decay-rate'
]
scores_without_nuclides = scores_with_nuclides + ['flux']
for do_nuclides, scores in ((False, scores_without_nuclides),
(True, scores_with_nuclides)):
t = openmc.Tally()
t.filters = [mesh_filter]
t.estimator = 'analog'
t.scores = scores
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mesh_filter]
t.estimator = 'tracklength'
t.scores = scores
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
# Impose energy bins that dont match the MG structure and those
# that do
for match_energy_bins in [False, True]:
if match_energy_bins:
e_filter = matching_energy_filter
eout_filter = matching_eout_filter
else:
e_filter = energy_filter
eout_filter = energyout_filter
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'analog'
t.scores = scores + ['scatter', 'nu-scatter']
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'collision'
t.scores = scores
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter]
t.estimator = 'tracklength'
t.scores = scores
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
t = openmc.Tally()
t.filters = [mat_filter, e_filter, eout_filter]
t.scores = ['scatter', 'nu-scatter', 'nu-fission']
if do_nuclides:
t.nuclides = nuclides
model.tallies.append(t)
harness = MGXSTestHarness('statepoint.10.h5', model)
harness.main()
def __init__(self, *args, **kwargs):
super().__init__(*args, **kwargs)
# Set settings explicitly
self._model.settings.batches = 3
self._model.settings.inactive = 0
self._model.settings.particles = 100
self._model.settings.source = openmc.Source(
space=openmc.stats.Box([-160, -160, -183], [160, 160, 183]))
self._model.settings.temperature['multipole'] = True
filt_mats = openmc.MaterialFilter((1, 3))
filt_eout = openmc.EnergyoutFilter((0.0, 0.625, 20.0e6))
# We want density derivatives for both water and fuel to get coverage
# for both fissile and non-fissile materials.
d1 = openmc.TallyDerivative(derivative_id=1)
d1.variable = 'density'
d1.material = 3
d2 = openmc.TallyDerivative(derivative_id=2)
d2.variable = 'density'
d2.material = 1
# O-16 is a good nuclide to test against because it is present in both
# water and fuel. Some routines need to recognize that they have the
# perturbed nuclide but not the perturbed material.
d3 = openmc.TallyDerivative(derivative_id=3)
d3.variable = 'nuclide_density'
d3.material = 1
d3.nuclide = 'O16'
# A fissile nuclide, just for good measure.
d4 = openmc.TallyDerivative(derivative_id=4)
d4.variable = 'nuclide_density'
d4.material = 1
d4.nuclide = 'U235'
# Temperature derivatives.
d5 = openmc.TallyDerivative(derivative_id=5)
d5.variable = 'temperature'
d5.material = 1
derivs = [d1, d2, d3, d4, d5]
# Cover the flux score.
for i in range(5):
t = openmc.Tally()
t.scores = ['flux']
t.filters = [filt_mats]
t.derivative = derivs[i]
self._model.tallies.append(t)
# Cover supported scores with a collision estimator.
for i in range(5):
t = openmc.Tally()
t.scores = [
'total', 'absorption', 'scatter', 'fission', 'nu-fission'
]
t.filters = [filt_mats]
t.nuclides = ['total', 'U235']
t.derivative = derivs[i]
self._model.tallies.append(t)
# Cover an analog estimator.
for i in range(5):
t = openmc.Tally()
t.scores = ['absorption']
t.filters = [filt_mats]
t.estimator = 'analog'
t.derivative = derivs[i]
self._model.tallies.append(t)
# Energyout filter and total nuclide for the density derivatives.
for i in range(2):
t = openmc.Tally()
t.scores = ['nu-fission', 'scatter']
t.filters = [filt_mats, filt_eout]
t.nuclides = ['total', 'U235']
t.derivative = derivs[i]
self._model.tallies.append(t)
# Energyout filter without total nuclide for other derivatives.
for i in range(2, 5):
t = openmc.Tally()
t.scores = ['nu-fission', 'scatter']
t.filters = [filt_mats, filt_eout]
t.nuclides = ['U235']
t.derivative = derivs[i]
self._model.tallies.append(t)
# Create an initial uniform spatial source distribution over fissionable zones
bounds = [-4., -4., -4., 4., 4., 4.]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml file
###############################################################################
# Instantiate some tally Filters
cell_filter = openmc.CellFilter(cell2)
energy_filter = openmc.EnergyFilter([0., 20.e6])
energyout_filter = openmc.EnergyoutFilter([0., 20.e6])
# Instantiate the first Tally
first_tally = openmc.Tally(tally_id=1, name='first tally')
first_tally.filters = [cell_filter]
scores = ['total', 'scatter', 'nu-scatter',
'absorption', 'fission', 'nu-fission']
first_tally.scores = scores
# Instantiate the second Tally
second_tally = openmc.Tally(tally_id=2, name='second tally')
second_tally.filters = [cell_filter, energy_filter]
second_tally.scores = scores
# Instantiate the third Tally
third_tally = openmc.Tally(tally_id=3, name='third tally')
