def tally_derivatives(self):
if not self._derivs_read:
# Populate the dictionary if any derivatives are present.
if 'derivatives' in self._f['tallies']:
# Read the derivative ids.
base = 'tallies/derivatives'
deriv_ids = [int(k.split(' ')[1]) for k in self._f[base]]
# Create each derivative object and add it to the dictionary.
for d_id in deriv_ids:
group = self._f['tallies/derivatives/derivative {}'.format(
d_id)]
deriv = openmc.TallyDerivative(derivative_id=d_id)
deriv.variable = group['independent variable'][()].decode()
if deriv.variable == 'density':
deriv.material = group['material'][()]
elif deriv.variable == 'nuclide_density':
deriv.material = group['material'][()]
deriv.nuclide = group['nuclide'][()].decode()
elif deriv.variable == 'temperature':
deriv.material = group['material'][()]
self._derivs[d_id] = deriv
self._derivs_read = True
return self._derivs
def test_xml_roundtrip(run_in_tmpdir):
# Create a tally with all possible gizmos
mesh = openmc.RegularMesh()
mesh.lower_left = (-10., -10., -10.)
mesh.upper_right = (
10.,
10.,
10.,
)
mesh.dimension = (5, 5, 5)
mesh_filter = openmc.MeshFilter(mesh)
tally = openmc.Tally()
tally.filters = [mesh_filter]
tally.nuclides = ['U235', 'I135', 'Li6']
tally.scores = ['total', 'fission', 'heating']
tally.derivative = openmc.TallyDerivative(variable='nuclide_density',
material=1,
nuclide='Li6')
tally.triggers = [openmc.Trigger('rel_err', 0.025)]
tally.triggers[0].scores = ['total', 'fission']
tallies = openmc.Tallies([tally])
# Roundtrip through XML and make sure we get what we started with
tallies.export_to_xml()
new_tallies = openmc.Tallies.from_xml()
assert len(new_tallies) == 1
new_tally = new_tallies[0]
assert new_tally.id == tally.id
assert len(new_tally.filters) == 1
assert isinstance(new_tally.filters[0], openmc.MeshFilter)
assert np.allclose(new_tally.filters[0].mesh.lower_left, mesh.lower_left)
assert new_tally.nuclides == tally.nuclides
assert new_tally.scores == tally.scores
assert new_tally.derivative.variable == tally.derivative.variable
assert new_tally.derivative.material == tally.derivative.material
assert new_tally.derivative.nuclide == tally.derivative.nuclide
assert len(new_tally.triggers) == 1
assert new_tally.triggers[0].trigger_type == tally.triggers[0].trigger_type
assert new_tally.triggers[0].threshold == tally.triggers[0].threshold
assert new_tally.triggers[0].scores == tally.triggers[0].scores
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)