def __init__(self, *args, **kwargs):
super(FilterMeshTestHarness, self).__init__(*args, **kwargs)
# Initialize Meshes
mesh_1d = openmc.Mesh(mesh_id=1)
mesh_1d.type = 'regular'
mesh_1d.dimension = [17]
mesh_1d.lower_left = [-182.07]
mesh_1d.upper_right = [182.07]
mesh_2d = openmc.Mesh(mesh_id=2)
mesh_2d.type = 'regular'
mesh_2d.dimension = [17, 17]
mesh_2d.lower_left = [-182.07, -182.07]
mesh_2d.upper_right = [182.07, 182.07]
mesh_3d = openmc.Mesh(mesh_id=3)
mesh_3d.type = 'regular'
mesh_3d.dimension = [17, 17, 17]
mesh_3d.lower_left = [-182.07, -182.07, -183.00]
mesh_3d.upper_right = [182.07, 182.07, 183.00]
# Initialize the filters
mesh_1d_filter = openmc.MeshFilter(mesh_1d)
mesh_2d_filter = openmc.MeshFilter(mesh_2d)
mesh_3d_filter = openmc.MeshFilter(mesh_3d)
# Initialized the tallies
tally = openmc.Tally(name='tally 1')
tally.filters = [mesh_1d_filter]
tally.scores = ['total']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 2')
tally.filters = [mesh_1d_filter]
tally.scores = ['current']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 3')
tally.filters = [mesh_2d_filter]
tally.scores = ['total']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 4')
tally.filters = [mesh_2d_filter]
tally.scores = ['current']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 5')
tally.filters = [mesh_3d_filter]
tally.scores = ['total']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 6')
tally.filters = [mesh_3d_filter]
tally.scores = ['current']
self._model.tallies.append(tally)
def __init__(self, *args, **kwargs):
super(TallyArithmeticTestHarness, self).__init__(*args, **kwargs)
# Initialize Mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = [2, 2, 2]
mesh.lower_left = [-160.0, -160.0, -183.0]
mesh.upper_right = [160.0, 160.0, 183.0]
# Initialize the filters
energy_filter = openmc.EnergyFilter((0.0, 0.253e-6, 1.0e-3, 1.0, 20.0))
material_filter = openmc.MaterialFilter((1, 3))
distrib_filter = openmc.DistribcellFilter(60)
mesh_filter = openmc.MeshFilter(mesh)
# Initialized the tallies
tally = openmc.Tally(name='tally 1')
tally.filters = [material_filter, energy_filter, distrib_filter]
tally.scores = ['nu-fission', 'total']
tally.nuclides = ['U234', 'U235']
self._model.tallies.append(tally)
tally = openmc.Tally(name='tally 2')
tally.filters = [energy_filter, mesh_filter]
tally.scores = ['total', 'fission']
tally.nuclides = ['U238', 'U235']
self._model.tallies.append(tally)
def make_tallies(self, r=None):
if not self.tally:
return None
if self.geom == 'IN':
return None
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = self.mesh_dim
if self.geom == 'SL' or self.geom == 'ISLC':
mesh.lower_left = [-r, -INF, -INF]
mesh.upper_right = [r, INF, INF]
# Instantiate some tally Filters
energy_filter = openmc.EnergyFilter(
GROUP_STRUCT[self.groups].group_edges)
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')
tally.filters = [energy_filter, mesh_filter]
tally.scores = ['flux', 'fission', 'nu-fission']
# Instantiate a Tallies collection, register all Tallies
tallies_file = openmc.Tallies([tally])
return tallies_file
def createTallies(meshname, create=True, load=True):
# Filters
# Unstructured mesh to calculate tallies upon
umesh = openmc.UnstructuredMesh(meshname, create, load)
mesh_filter = openmc.MeshFilter(umesh)
# Tallies
tally = openmc.Tally()
tally.filters = [mesh_filter]
tally.scores = ['heating-local', 'flux']
tally.estimator = 'tracklength'
tallies = openmc.Tallies([tally])
tallies.export_to_xml()
print("Created tallies.xml")
def model():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.0)
fuel.add_nuclide('U234', 1.0)
fuel.add_nuclide('U235', 4.0)
fuel.add_nuclide('U238', 95.0)
water = openmc.Material(name='light water')
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.set_density('g/cm3', 1.0)
water.add_s_alpha_beta('c_H_in_H2O')
model.materials.extend([fuel, water])
cyl1 = openmc.ZCylinder(r=5.0)
cyl2 = openmc.ZCylinder(r=10.0, boundary_type='vacuum')
cell1 = openmc.Cell(fill=fuel, region=-cyl1)
cell2 = openmc.Cell(fill=water, region=+cyl1 & -cyl2)
model.geometry = openmc.Geometry([cell1, cell2])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
mesh = openmc.RegularMesh()
mesh.dimension = (2, 2)
mesh.lower_left = (-10.0, -10.0)
mesh.upper_right = (10.0, 10.0)
energy_filter = openmc.EnergyFilter((0.0, 10.0, 20.0e6))
material_filter = openmc.MaterialFilter((fuel, water))
mesh_filter = openmc.MeshFilter(mesh)
tally = openmc.Tally(name='tally 1')
tally.filters = [material_filter, energy_filter]
tally.scores = ['nu-fission', 'total']
tally.nuclides = ['U234', 'U235']
model.tallies.append(tally)
tally = openmc.Tally(name='tally 2')
tally.filters = [energy_filter, mesh_filter]
tally.scores = ['total', 'fission']
tally.nuclides = ['U238', 'U235']
model.tallies.append(tally)
return model
def create_tally(radius):
"""Create tally mesh???
"""
tallies_file = openmc.Tallies()
mesh = openmc.Mesh()
mesh.dimension = [100, 100]
mesh.lower_left = [-radius, -radius]
mesh.upper_right = [radius, radius]
mesh_filter = openmc.MeshFilter(mesh)
mesh_filter.mesh = mesh
tally = openmc.Tally(name='flux')
tally.filters = [mesh_filter]
tally.scores = ['flux', 'fission']
tallies_file.append(tally)
tallies_file.export_to_xml()
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 test_unstructured_mesh(test_opts):
### 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
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.h5m"
else:
mesh_filename = "test_mesh_tets.h5m"
uscd_mesh = openmc.UnstructuredMesh(mesh_filename)
uscd_mesh.mesh_lib = 'moab'
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 = 100
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)
angle = openmc.stats.Monodirectional((-1.0, 0.0, 0.0))
energy = openmc.stats.Discrete(x=[15.e+06], p=[1.0])
source = openmc.Source(space=space, energy=energy, angle=angle)
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():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.0)
fuel.add_nuclide('U235', 1.0)
zr = openmc.Material()
zr.set_density('g/cm3', 1.0)
zr.add_nuclide('Zr90', 1.0)
model.materials.extend([fuel, zr])
box1 = openmc.model.rectangular_prism(10.0, 10.0)
box2 = openmc.model.rectangular_prism(20.0, 20.0, boundary_type='reflective')
top = openmc.ZPlane(z0=10.0, boundary_type='vacuum')
bottom = openmc.ZPlane(z0=-10.0, boundary_type='vacuum')
cell1 = openmc.Cell(fill=fuel, region=box1 & +bottom & -top)
cell2 = openmc.Cell(fill=zr, region=~box1 & box2 & +bottom & -top)
model.geometry = openmc.Geometry([cell1, cell2])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
translation = np.array((10, -5, 0))
llc = np.array([-9, -9, -9])
urc = np.array([9, 9, 9])
mesh_dims = (3, 4, 5)
filters = []
# un-translated meshes
reg_mesh = openmc.RegularMesh()
reg_mesh.dimension = mesh_dims
reg_mesh.lower_left = llc
reg_mesh.upper_right = urc
filters.append(openmc.MeshFilter(reg_mesh))
recti_mesh = openmc.RectilinearMesh()
recti_mesh.x_grid = np.linspace(llc[0], urc[0], mesh_dims[0])
recti_mesh.y_grid = np.linspace(llc[1], urc[1], mesh_dims[1])
recti_mesh.z_grid = np.linspace(llc[2], urc[2], mesh_dims[2])
filters.append(openmc.MeshFilter(recti_mesh))
llc = np.array(llc - translation)
urc = np.array(urc - translation)
# translated meshes
translated_reg_mesh = openmc.RegularMesh()
translated_reg_mesh.dimension = mesh_dims
translated_reg_mesh.lower_left = llc
translated_reg_mesh.upper_right = urc
filters.append(openmc.MeshFilter(translated_reg_mesh))
filters[-1].translation = translation
translated_recti_mesh = openmc.RectilinearMesh()
translated_recti_mesh.x_grid = np.linspace(llc[0], urc[0], mesh_dims[0])
translated_recti_mesh.y_grid = np.linspace(llc[1], urc[1], mesh_dims[1])
translated_recti_mesh.z_grid = np.linspace(llc[2], urc[2], mesh_dims[2])
filters.append(openmc.MeshFilter(translated_recti_mesh))
filters[-1].translation = translation
# Create tallies
for f in filters:
tally = openmc.Tally()
tally.filters = [f]
tally.scores = ['total']
model.tallies.append(tally)
return model
def test_external_mesh(cpp_driver):
# 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
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])
# Meshes
mesh_filename = "test_mesh_tets.h5m"
# Create a normal unstructured mesh to compare to
uscd_mesh = openmc.UnstructuredMesh(mesh_filename, 'moab')
# Create filters
uscd_filter = openmc.MeshFilter(mesh=uscd_mesh)
# Tallies
tallies = openmc.Tallies()
uscd_tally = openmc.Tally(name="unstructured mesh tally")
uscd_tally.filters = [uscd_filter]
uscd_tally.scores = ['flux']
uscd_tally.estimator = 'tracklength'
tallies.append(uscd_tally)
# Settings
settings = openmc.Settings()
settings.run_mode = 'fixed source'
settings.particles = 100
settings.batches = 10
# Source setup
space = openmc.stats.Point()
angle = openmc.stats.Monodirectional((-1.0, 0.0, 0.0))
energy = openmc.stats.Discrete(x=[15.e+06], p=[1.0])
source = openmc.Source(space=space, energy=energy, angle=angle)
settings.source = source
model = openmc.model.Model(geometry=geometry,
materials=materials,
tallies=tallies,
settings=settings)
harness = ExternalMoabTest(cpp_driver, 'statepoint.10.h5', model)
# Run open MC and check results
harness.main()
def model():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.0)
fuel.add_nuclide('U235', 1.0)
zr = openmc.Material()
zr.set_density('g/cm3', 1.0)
zr.add_nuclide('Zr90', 1.0)
model.materials.extend([fuel, zr])
box1 = openmc.model.rectangular_prism(10.0, 10.0)
box2 = openmc.model.rectangular_prism(20.0,
20.0,
boundary_type='reflective')
top = openmc.ZPlane(z0=10.0, boundary_type='vacuum')
bottom = openmc.ZPlane(z0=-10.0, boundary_type='vacuum')
cell1 = openmc.Cell(fill=fuel, region=box1 & +bottom & -top)
cell2 = openmc.Cell(fill=zr, region=~box1 & box2 & +bottom & -top)
model.geometry = openmc.Geometry([cell1, cell2])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
# Create meshes
mesh_1d = openmc.RegularMesh()
mesh_1d.dimension = [5]
mesh_1d.lower_left = [-7.5]
mesh_1d.upper_right = [7.5]
mesh_2d = openmc.RegularMesh()
mesh_2d.dimension = [5, 5]
mesh_2d.lower_left = [-7.5, -7.5]
mesh_2d.upper_right = [7.5, 7.5]
mesh_3d = openmc.RegularMesh()
mesh_3d.dimension = [5, 5, 5]
mesh_3d.lower_left = [-7.5, -7.5, -7.5]
mesh_3d.upper_right = [7.5, 7.5, 7.5]
recti_mesh = openmc.RectilinearMesh()
recti_mesh.x_grid = np.linspace(-7.5, 7.5, 18)
recti_mesh.y_grid = np.linspace(-7.5, 7.5, 18)
recti_mesh.z_grid = np.logspace(0, np.log10(7.5), 11)
# Create filters
reg_filters = [
openmc.MeshFilter(mesh_1d),
openmc.MeshFilter(mesh_2d),
openmc.MeshFilter(mesh_3d),
openmc.MeshFilter(recti_mesh)
]
surf_filters = [
openmc.MeshSurfaceFilter(mesh_1d),
openmc.MeshSurfaceFilter(mesh_2d),
openmc.MeshSurfaceFilter(mesh_3d),
openmc.MeshSurfaceFilter(recti_mesh)
]
# Create tallies
for f1, f2 in zip(reg_filters, surf_filters):
tally = openmc.Tally()
tally.filters = [f1]
tally.scores = ['total']
model.tallies.append(tally)
tally = openmc.Tally()
tally.filters = [f2]
tally.scores = ['current']
model.tallies.append(tally)
return model
# Set materials volume for depletion. Set to an area for 2D simulations
fuel_material.volume = volume[fuel_material] #+ volume[fuel_hot];
print("Volume:", volume[fuel_material])
###############################################################################
tallies_file = openmc.Tallies()
# XY-MESH
#Creating mesh and mesh filter
mesh_xy = openmc.RegularMesh(mesh_id=1)
mesh_xy.dimension = [200,200,1]
mesh_xy.lower_left = [-100,-100,-1.e50]
mesh_xy.upper_right = [100,100,1.e50]
mesh_filter_xy = openmc.MeshFilter(mesh_xy)
particle_filter = openmc.ParticleFilter(['neutron','photon'], filter_id=2)
#Creating mesh tally
tally_xy = openmc.Tally(name='flux_xy',tally_id=1)
tally_xy.filters = [mesh_filter_xy,particle_filter]
tally_xy.scores = ['flux','fission','nu-fission']
tallies_file.append(tally_xy)
# XZ-MESH
#Creating mesh and mesh filter
mesh_xz = openmc.RegularMesh(mesh_id=2)
mesh_xz.dimension = [200,1,200]
mesh_xz.lower_left = [-100,-1.e50,-100]
mesh_xz.upper_right = [100,1.e50,100]
def __init__(self, *args, **kwargs):
super().__init__(*args, **kwargs)
# Define nuclides and scores to add to both tallies
self.nuclides = ['U235', 'U238']
self.scores = ['fission', 'nu-fission']
# Define filters for energy and spatial domain
low_energy = openmc.EnergyFilter([0., 0.625])
high_energy = openmc.EnergyFilter([0.625, 20.e6])
merged_energies = low_energy.merge(high_energy)
cell_21 = openmc.CellFilter(21)
cell_27 = openmc.CellFilter(27)
distribcell_filter = openmc.DistribcellFilter(21)
mesh = openmc.RegularMesh(name='mesh')
mesh.dimension = [2, 2]
mesh.lower_left = [-50., -50.]
mesh.upper_right = [+50., +50.]
mesh_filter = openmc.MeshFilter(mesh)
self.cell_filters = [cell_21, cell_27]
self.energy_filters = [low_energy, high_energy]
# Initialize cell tallies with filters, nuclides and scores
tallies = []
for energy_filter in self.energy_filters:
for cell_filter in self.cell_filters:
for nuclide in self.nuclides:
for score in self.scores:
tally = openmc.Tally()
tally.estimator = 'tracklength'
tally.scores.append(score)
tally.nuclides.append(nuclide)
tally.filters.append(cell_filter)
tally.filters.append(energy_filter)
tallies.append(tally)
# Merge all cell tallies together
while len(tallies) != 1:
halfway = len(tallies) // 2
zip_split = zip(tallies[:halfway], tallies[halfway:])
tallies = list(map(lambda xy: xy[0].merge(xy[1]), zip_split))
# Specify a name for the tally
tallies[0].name = 'cell tally'
# Initialize a distribcell tally
distribcell_tally = openmc.Tally(name='distribcell tally')
distribcell_tally.estimator = 'tracklength'
distribcell_tally.filters = [distribcell_filter, merged_energies]
for score in self.scores:
distribcell_tally.scores.append(score)
for nuclide in self.nuclides:
distribcell_tally.nuclides.append(nuclide)
mesh_tally = openmc.Tally(name='mesh tally')
mesh_tally.estimator = 'tracklength'
mesh_tally.filters = [mesh_filter, merged_energies]
mesh_tally.scores = self.scores
mesh_tally.nuclides = self.nuclides
# Add tallies to a Tallies object
self._model.tallies = [tallies[0], distribcell_tally, mesh_tally]
entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50] entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50] entropy_mesh.dimension = [10, 10, 1] settings_file.entropy_mesh = entropy_mesh settings_file.export_to_xml() ############################################################################### # Exporting to OpenMC tallies.xml file ############################################################################### # Instantiate a tally mesh mesh = openmc.Mesh(mesh_id=1) mesh.type = 'regular' mesh.dimension = [100, 100, 1] mesh.lower_left = [-0.62992, -0.62992, -1.e50] mesh.upper_right = [0.62992, 0.62992, 1.e50] # Instantiate some tally Filters energy_filter = openmc.EnergyFilter([0., 4., 20.e6]) mesh_filter = openmc.MeshFilter(mesh) # Instantiate the Tally tally = openmc.Tally(tally_id=1, name='tally 1') tally.filters = [energy_filter, mesh_filter] tally.scores = ['flux', 'fission', 'nu-fission'] # Instantiate a Tallies collection and export to XML tallies_file = openmc.Tallies([tally]) tallies_file.export_to_xml()
mats.export_to_xml() # Create a 5 cm x 5 cm box filled with iron box = openmc.model.rectangular_prism(10.0, 10.0, boundary_type='vacuum') cell = openmc.Cell(fill=iron, region=box) geometry = openmc.Geometry([cell]) geometry.export_to_xml() # Tell OpenMC we're going to use our custom source settings = openmc.Settings() settings.run_mode = 'fixed source' settings.batches = 10 settings.particles = 1000 source = openmc.Source() source.library = 'build/libparameterized_source.so' source.parameters = 'radius=3.0, energy=14.08e6' settings.source = source settings.export_to_xml() # Finally, define a mesh tally so that we can see the resulting flux mesh = openmc.RegularMesh() mesh.lower_left = (-5.0, -5.0) mesh.upper_right = (5.0, 5.0) mesh.dimension = (50, 50) tally = openmc.Tally() tally.filters = [openmc.MeshFilter(mesh)] tally.scores = ['flux'] tallies = openmc.Tallies([tally]) tallies.export_to_xml()
def create_neutronics_model(self, method: str = None):
"""Uses OpenMC python API to make a neutronics model, including tallies
(cell_tallies and mesh_tally_2d), simulation settings (batches,
particles per batch).
Arguments:
method: (str): The method to use when making the imprinted and
merged geometry. Options are "ppp", "trelis", "pymoab".
Defaults to None.
"""
self.create_materials()
self.create_neutronics_geometry(method=method)
# this is the underlying geometry container that is filled with the
# faceteted DGAMC CAD model
self.universe = openmc.Universe()
geom = openmc.Geometry(self.universe)
# settings for the number of neutrons to simulate
settings = openmc.Settings()
settings.batches = self.simulation_batches
settings.inactive = 0
settings.particles = self.simulation_particles_per_batch
settings.run_mode = "fixed source"
settings.dagmc = True
settings.photon_transport = True
settings.source = self.source
settings.max_lost_particles = self.max_lost_particles
# details about what neutrons interactions to keep track of (tally)
self.tallies = openmc.Tallies()
if self.mesh_tally_3d is not None:
mesh_xyz = openmc.RegularMesh(mesh_id=1, name='3d_mesh')
mesh_xyz.dimension = self.mesh_3D_resolution
mesh_xyz.lower_left = [
-self.geometry.largest_dimension,
-self.geometry.largest_dimension,
-self.geometry.largest_dimension
]
mesh_xyz.upper_right = [
self.geometry.largest_dimension,
self.geometry.largest_dimension,
self.geometry.largest_dimension
]
for standard_tally in self.mesh_tally_3d:
if standard_tally == 'tritium_production':
score = '(n,Xt)' # where X is a wild card
prefix = 'tritium_production'
else:
score = standard_tally
prefix = standard_tally
mesh_filter = openmc.MeshFilter(mesh_xyz)
tally = openmc.Tally(name=prefix + '_on_3D_mesh')
tally.filters = [mesh_filter]
tally.scores = [score]
self.tallies.append(tally)
if self.mesh_tally_2d is not None:
# Create mesh which will be used for tally
mesh_xz = openmc.RegularMesh(mesh_id=2, name='2d_mesh_xz')
mesh_xz.dimension = [
self.mesh_2D_resolution[1],
1,
self.mesh_2D_resolution[0]
]
mesh_xz.lower_left = [
-self.geometry.largest_dimension,
-1,
-self.geometry.largest_dimension
]
mesh_xz.upper_right = [
self.geometry.largest_dimension,
1,
self.geometry.largest_dimension
]
mesh_xy = openmc.RegularMesh(mesh_id=3, name='2d_mesh_xy')
mesh_xy.dimension = [
self.mesh_2D_resolution[1],
self.mesh_2D_resolution[0],
1
]
mesh_xy.lower_left = [
-self.geometry.largest_dimension,
-self.geometry.largest_dimension,
-1
]
mesh_xy.upper_right = [
self.geometry.largest_dimension,
self.geometry.largest_dimension,
1
]
mesh_yz = openmc.RegularMesh(mesh_id=4, name='2d_mesh_yz')
mesh_yz.dimension = [
1,
self.mesh_2D_resolution[1],
self.mesh_2D_resolution[0]
]
mesh_yz.lower_left = [
-1,
-self.geometry.largest_dimension,
-self.geometry.largest_dimension
]
mesh_yz.upper_right = [
1,
self.geometry.largest_dimension,
self.geometry.largest_dimension
]
for standard_tally in self.mesh_tally_2d:
if standard_tally == 'tritium_production':
score = '(n,Xt)' # where X is a wild card
prefix = 'tritium_production'
else:
score = standard_tally
prefix = standard_tally
for mesh_filter, plane in zip(
[mesh_xz, mesh_xy, mesh_yz], ['xz', 'xy', 'yz']):
mesh_filter = openmc.MeshFilter(mesh_filter)
tally = openmc.Tally(name=prefix + '_on_2D_mesh_' + plane)
tally.filters = [mesh_filter]
tally.scores = [score]
self.tallies.append(tally)
if self.cell_tallies is not None:
for standard_tally in self.cell_tallies:
if standard_tally == 'TBR':
score = '(n,Xt)' # where X is a wild card
sufix = 'TBR'
tally = openmc.Tally(name='TBR')
tally.scores = [score]
self.tallies.append(tally)
self._add_tally_for_every_material(sufix, score)
elif standard_tally == 'spectra':
neutron_particle_filter = openmc.ParticleFilter([
'neutron'])
photon_particle_filter = openmc.ParticleFilter(['photon'])
energy_bins = openmc.mgxs.GROUP_STRUCTURES['CCFE-709']
energy_filter = openmc.EnergyFilter(energy_bins)
self._add_tally_for_every_material(
'neutron_spectra',
'flux',
[neutron_particle_filter, energy_filter]
)
self._add_tally_for_every_material(
'photon_spectra',
'flux',
[photon_particle_filter, energy_filter]
)
else:
score = standard_tally
sufix = standard_tally
self._add_tally_for_every_material(sufix, score)
# make the model from gemonetry, materials, settings and tallies
self.model = openmc.model.Model(
geom, self.mats, settings, self.tallies)
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 model():
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 = 200
settings.batches = 2
settings.max_splits = 200
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 = (5, 10, 15)
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)
# weight windows
# load pre-generated weight windows
# (created using the same tally as above)
ww_n_lower_bnds = np.loadtxt('ww_n.txt')
ww_p_lower_bnds = np.loadtxt('ww_p.txt')
# create a mesh matching the one used
# to generate the weight windows
ww_mesh = openmc.RegularMesh()
ww_mesh.lower_left = (-240, -240, -240)
ww_mesh.upper_right = (240, 240, 240)
ww_mesh.dimension = (5, 6, 7)
ww_n = openmc.WeightWindows(ww_mesh,
ww_n_lower_bnds,
None,
10.0,
e_bnds,
max_lower_bound_ratio=1.5)
ww_p = openmc.WeightWindows(ww_mesh,
ww_p_lower_bnds,
None,
10.0,
e_bnds,
max_lower_bound_ratio=1.5)
model.settings.weight_windows = [ww_n, ww_p]
return model
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()
plots = openmc.Plots([p])
plots.export_to_xml()
openmc.plot_geometry()
get_ipython().system('convert coreplot.ppm core1.png')
tallies = openmc.Tallies()
# Create mesh which will be used for tally
mesh_core = openmc.Mesh()
mesh_core.dimension = [21, 1, 1]
mesh_core.lower_left = (-100, -100, 0)
mesh_core.upper_right = (100, 100, 160)
# Create filter for tally
mesh_filtercore = openmc.MeshFilter(mesh_core)
core_filter = openmc.UniverseFilter([a1, a2, a3, a4, ba])
energy_filter = openmc.EnergyFilter([0.0, 1.0, 1.0e5, 2.0e7])
# Create mesh tally to score flux and fission rate
tally_core = openmc.Tally(name='flux')
tally_core.filters = [mesh_filtercore]
tally_core.scores = ['fission-q-prompt']
tallies.append(tally_core)
tallies.export_to_xml()
coretest = openmc.Settings()
coretest.batches = 40
coretest.inactive = 10
coretest.particles = 10000
def tallies_generation(root):
""" Creates tallies.xml file
Parameters
----------
root: openmc.Universe with all the relevant cells for the geometry.
Returns
-------
This function generates the tallies.xml file.
"""
tallies_file = openmc.Tallies()
# phase1a-b
energy_filter_b = openmc.EnergyFilter([1e-6, 20.0e6])
mesh_b = openmc.RegularMesh(mesh_id=16)
mesh_b.dimension = [1, 1]
L = 27.02
mesh_b.lower_left = [-L, -L]
mesh_b.upper_right = [L, L]
mesh_filter_b = openmc.MeshFilter(mesh_b)
tally_b = openmc.Tally(name='mesh tally b')
tally_b.filters = [mesh_filter_b, energy_filter_b]
tally_b.scores = ['delayed-nu-fission', 'nu-fission']
tallies_file.append(tally_b)
# phase1a-c
mesh_no = 0
for t in range(6):
mesh_no += 1
for x in range(2):
x_trans = t * T['A1']['P']['x']
y_trans = t * T['A1']['P']['y']
if x == 1:
mesh_no += 1
x_trans += T['A1']['F']['x']
y_trans += T['A1']['F']['y']
mesh_c = openmc.RegularMesh(mesh_id=mesh_no)
mesh_c.dimension = [1, 5]
mesh_c.lower_left = [
V['A1']['F']['L']['x'] + x_trans,
V['A1']['F']['B']['y'] + y_trans
]
mesh_c.upper_right = [
V['A1']['F']['R']['x'] + x_trans,
V['A1']['F']['T']['y'] + y_trans
]
mesh_filter_c = openmc.MeshFilter(mesh_c)
tally_c = openmc.Tally(name='mesh tally c' + str(mesh_no))
tally_c.filters = [mesh_filter_c]
tally_c.scores = ['fission']
tallies_file.append(tally_c)
# phase 1a-d
energy_filter_d = openmc.EnergyFilter([1e-5, 3, 1.0e5, 20.0e6])
mesh_d = openmc.RegularMesh(mesh_id=13)
mesh_d.dimension = [1, 1]
L = 27.02
mesh_d.lower_left = [-L, -L]
mesh_d.upper_right = [L, L]
mesh_filter_d = openmc.MeshFilter(mesh_d)
tally_d = openmc.Tally(name='mesh tally d')
tally_d.filters = [mesh_filter_d, energy_filter_d]
tally_d.scores = ['flux', 'nu-fission', 'fission']
tallies_file.append(tally_d)
# phase 1a-e
energy_filter_e = openmc.EnergyFilter([1e-5, 3, 0.1e6, 20.0e6])
mesh_e = openmc.RegularMesh(mesh_id=14)
mesh_e.dimension = [100, 100]
L = 27.02
mesh_e.lower_left = [-L, -L]
mesh_e.upper_right = [L, L]
mesh_filter_e = openmc.MeshFilter(mesh_e)
tally_e = openmc.Tally(name='mesh tally e')
tally_e.filters = [mesh_filter_e, energy_filter_e]
tally_e.scores = ['flux', 'nu-fission', 'fission']
tallies_file.append(tally_e)
# phase 1a-f
energy_filter_f = openmc.EnergyFilter(engs)
mesh_f = openmc.RegularMesh(mesh_id=15)
mesh_f.dimension = [1, 1]
L = 27.02
mesh_f.lower_left = [-L, -L]
mesh_f.upper_right = [L, L]
mesh_filter_f = openmc.MeshFilter(mesh_f)
tally_f = openmc.Tally(name='mesh tally f')
tally_f.filters = [mesh_filter_f, energy_filter_f]
tally_f.scores = ['flux', 'nu-fission', 'fission']
tallies_file.append(tally_f)
tallies_file.export_to_xml()
return