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 __init__(self, *args, **kwargs):
super(FilterEnergyFunHarness, self).__init__(*args, **kwargs)
# Add Am241 to the fuel.
self._model.materials[1].add_nuclide('Am241', 1e-7)
# Define Am242m / Am242 branching ratio from ENDF/B-VII.1 data.
x = [1e-5, 3.69e-1, 1e3, 1e5, 6e5, 1e6, 2e6, 4e6, 3e7]
y = [0.1, 0.1, 0.1333, 0.158, 0.18467, 0.25618, 0.4297, 0.48, 0.48]
# Make an EnergyFunctionFilter directly from the x and y lists.
filt1 = openmc.EnergyFunctionFilter(x, y)
# Also make a filter with the .from_tabulated1d constructor. Make sure
# the filters are identical.
tab1d = openmc.data.Tabulated1D(x, y)
filt2 = openmc.EnergyFunctionFilter.from_tabulated1d(tab1d)
assert filt1 == filt2, 'Error with the .from_tabulated1d constructor'
# Make tallies.
tallies = [openmc.Tally(1), openmc.Tally(2)]
for t in tallies:
t.scores = ['(n,gamma)']
t.nuclides = ['Am241']
tallies[1].filters = [filt1]
self._model.tallies = tallies
def model():
model = openmc.model.Model()
m = openmc.Material()
m.set_density('g/cm3', 10.0)
m.add_nuclide('Am241', 1.0)
model.materials.append(m)
s = openmc.Sphere(r=100.0, boundary_type='vacuum')
c = openmc.Cell(fill=m, region=-s)
model.geometry = openmc.Geometry([c])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
# Define Am242m / Am242 branching ratio from ENDF/B-VII.1 data.
x = [1e-5, 3.69e-1, 1e3, 1e5, 6e5, 1e6, 2e6, 4e6, 3e7]
y = [0.1, 0.1, 0.1333, 0.158, 0.18467, 0.25618, 0.4297, 0.48, 0.48]
# Make an EnergyFunctionFilter directly from the x and y lists.
filt1 = openmc.EnergyFunctionFilter(x, y)
# Also make a filter with the .from_tabulated1d constructor. Make sure
# the filters are identical.
tab1d = openmc.data.Tabulated1D(x, y)
filt2 = openmc.EnergyFunctionFilter.from_tabulated1d(tab1d)
assert filt1 == filt2, 'Error with the .from_tabulated1d constructor'
# Make tallies
tallies = [openmc.Tally(), openmc.Tally()]
for t in tallies:
t.scores = ['(n,gamma)']
t.nuclides = ['Am241']
tallies[1].filters = [filt1]
model.tallies.extend(tallies)
return model
