def __init__(self, *args, **kwargs):
super().__init__(*args, **kwargs)
# Initialize a two-group structure
energy_groups = openmc.mgxs.EnergyGroups(group_edges=[0, 0.625, 20.e6])
# Initialize MGXS Library for a few cross section types
self.mgxs_lib = openmc.mgxs.Library(self._model.geometry)
self.mgxs_lib.by_nuclide = False
# Test all MGXS types
self.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES + \
openmc.mgxs.MDGXS_TYPES
self.mgxs_lib.energy_groups = energy_groups
self.mgxs_lib.num_delayed_groups = 6
self.mgxs_lib.legendre_order = 3
self.mgxs_lib.domain_type = 'mesh'
# Instantiate a tally mesh
mesh = openmc.RegularMesh(mesh_id=1)
mesh.dimension = [2, 2]
mesh.lower_left = [-100., -100.]
mesh.width = [100., 100.]
self.mgxs_lib.domains = [mesh]
self.mgxs_lib.build_library()
# Add tallies
self.mgxs_lib.add_to_tallies_file(self._model.tallies, merge=False)
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
# Initialize a one-group structure
energy_groups = openmc.mgxs.EnergyGroups([0, 20.e6])
# Initialize MGXS Library for a few cross section types
# for one material-filled cell in the geometry
model.mgxs_lib = openmc.mgxs.Library(model.geometry)
model.mgxs_lib.by_nuclide = False
# Test all MGXS types
model.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES + openmc.mgxs.MDGXS_TYPES
model.mgxs_lib.energy_groups = energy_groups
model.mgxs_lib.num_delayed_groups = 6
model.mgxs_lib.correction = None # Avoid warning about P0 correction
model.mgxs_lib.legendre_order = 3
model.mgxs_lib.domain_type = 'mesh'
# Instantiate a tally mesh
mesh = openmc.RegularMesh(mesh_id=1)
mesh.dimension = [2, 2]
mesh.lower_left = [-100., -100.]
mesh.width = [100., 100.]
model.mgxs_lib.domains = [mesh]
model.mgxs_lib.build_library()
# Add tallies
model.mgxs_lib.add_to_tallies_file(model.tallies, merge=False)
return model
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 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_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 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
settings.batches = 100 settings.inactive = 10 settings.particles = 1000 # Create an initial uniform spatial source distribution over fissionable zones lower_left = (-pitch / 2, -pitch / 2, -1) upper_right = (pitch / 2, pitch / 2, 1) uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True) settings.source = openmc.source.Source(space=uniform_dist) settings.export_to_xml() ############################################################################### # Define tallies # Create a mesh that will be used for tallying mesh = openmc.RegularMesh() mesh.dimension = (100, 100) mesh.lower_left = (-pitch / 2, -pitch / 2) mesh.upper_right = (pitch / 2, pitch / 2) # Create a mesh filter that can be used in a tally mesh_filter = openmc.MeshFilter(mesh) # Now use the mesh filter in a tally and indicate what scores are desired mesh_tally = openmc.Tally(name="Mesh tally") mesh_tally.filters = [mesh_filter] mesh_tally.scores = ['flux', 'fission', 'nu-fission'] # Let's also create a tally to get the flux energy spectrum. We start by # creating an energy filter e_min, e_max = 1e-5, 20.0e6
def build_homog_input_files(order, source_loc, bc, prob_size, num_neutrons,
num_batches, SE_dim, seed, dir):
os.system("mkdir -p {dir}".format(dir=dir))
os.chdir(dir)
xmin = -1. * prob_size[0] / 2.
xmax = prob_size[0] / 2.
ymin = -1. * prob_size[1] / 2.
ymax = prob_size[1] / 2.
zmin = -1. * prob_size[2] / 2.
zmax = prob_size[2] / 2.
# Instantiate some Materials and register the appropriate Nuclides
uranyl_sulf = openmc.Material(name='Uranyl Sulfate')
uranyl_sulf.set_density("atom/b-cm", 9.9035E-02)
uranyl_sulf.add_nuclide("U235", 9.6795E-05)
uranyl_sulf.add_nuclide("U234", 7.4257E-07)
uranyl_sulf.add_nuclide("U238", 5.5518E-04)
uranyl_sulf.add_nuclide("S32", 6.5272E-04)
uranyl_sulf.add_nuclide("O16", 3.5185E-02)
uranyl_sulf.add_nuclide("H1", 6.2538E-02)
uranyl_sulf.add_s_alpha_beta('c_H_in_H2O')
uranyl_sulf.depletable = False
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uranyl_sulf])
materials_file.export_to_xml()
# Instantiate planar surfaces
x1 = openmc.XPlane(x0=xmin)
x2 = openmc.XPlane(x0=xmax)
y1 = openmc.YPlane(y0=ymin)
y2 = openmc.YPlane(y0=ymax)
z1 = openmc.ZPlane(z0=zmin)
z2 = openmc.ZPlane(z0=zmax)
# Set boundary conditions
surface_list = [x1, x2, y1, y2, z1, z2]
for i in range(len(surface_list)):
surface = surface_list[i]
surface.boundary_type = bc[i]
# Define Cell, Region, and Fill
uran_sulf_sol = openmc.Cell(name='uranyl sulfate solution')
uran_sulf_sol.region = +x1 & -x2 & +y1 & -y2 & +z1 & -z2
uran_sulf_sol.fill = uranyl_sulf
# Instantiate root universe
root = openmc.Universe(name='root universe')
root.add_cells([uran_sulf_sol])
# Instantiate a Geometry, register the root Universe, and export to XML
geometry = openmc.Geometry(root)
geometry.export_to_xml()
# Define runtime settings
settings = openmc.Settings()
point_source = openmc.stats.Point(xyz=source_loc)
settings.source = openmc.Source(space=point_source)
settings.batches = num_batches
settings.inactive = num_batches - 1
settings.particles = num_neutrons
# Define entropy mesh
entropy_mesh = openmc.RegularMesh()
entropy_mesh.lower_left = [xmin, ymin, zmin]
entropy_mesh.upper_right = [xmax, ymax, zmax]
entropy_mesh.dimension = SE_dim
settings.entropy_mesh = entropy_mesh
settings.seed = seed
settings.export_to_xml()
# Create a nu-fission tally
nu_fiss_tally = openmc.Tally()
nu_fiss_tally.scores = ['nu-fission']
# Create a Legendre polynomial expansion filter and add to tally
expand_filter = openmc.SpatialLegendreFilter(order, 'z', zmin, zmax)
nu_fiss_tally.filters.append(expand_filter)
tallies = openmc.Tallies([nu_fiss_tally])
tallies.export_to_xml()
os.chdir("./..")
import openmc
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
tallies = {}
# Instantiate a tally mesh
mesh = openmc.RegularMesh(mesh_id=1)
mesh.dimension = [51, 51, 1]
mesh.lower_left = [-32.13, -32.13, -1.e50]
mesh.upper_right = [32.13, 32.13, 1.e50]
# Instantiate some tally Filters
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate the Tally
tallies['Mesh Rates'] = openmc.Tally(tally_id=1, name='tally 1')
tallies['Mesh Rates'].filters = [mesh_filter]
tallies['Mesh Rates'].scores = ['flux', 'fission', 'nu-fission']
tallies['Global Rates'] = openmc.Tally(tally_id=2, name='tally 2')
tallies['Global Rates'].scores = ['flux', 'fission', 'nu-fission']
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 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
def test_export_to_xml(run_in_tmpdir):
s = openmc.Settings()
s.run_mode = 'fixed source'
s.batches = 1000
s.generations_per_batch = 10
s.inactive = 100
s.particles = 1000000
s.max_lost_particles = 5
s.rel_max_lost_particles = 1e-4
s.keff_trigger = {'type': 'std_dev', 'threshold': 0.001}
s.energy_mode = 'continuous-energy'
s.max_order = 5
s.source = openmc.Source(space=openmc.stats.Point())
s.output = {'summary': True, 'tallies': False, 'path': 'here'}
s.verbosity = 7
s.sourcepoint = {
'batches': [50, 150, 500, 1000],
'separate': True,
'write': True,
'overwrite': True
}
s.statepoint = {'batches': [50, 150, 500, 1000]}
s.surf_source_read = {'path': 'surface_source_1.h5'}
s.surf_source_write = {'surface_ids': [2], 'max_particles': 200}
s.confidence_intervals = True
s.ptables = True
s.seed = 17
s.survival_biasing = True
s.cutoff = {
'weight': 0.25,
'weight_avg': 0.5,
'energy_neutron': 1.0e-5,
'energy_photon': 1000.0,
'energy_electron': 1.0e-5,
'energy_positron': 1.0e-5
}
mesh = openmc.RegularMesh()
mesh.lower_left = (-10., -10., -10.)
mesh.upper_right = (10., 10., 10.)
mesh.dimension = (5, 5, 5)
s.entropy_mesh = mesh
s.trigger_active = True
s.trigger_max_batches = 10000
s.trigger_batch_interval = 50
s.no_reduce = False
s.tabular_legendre = {'enable': True, 'num_points': 50}
s.temperature = {
'default': 293.6,
'method': 'interpolation',
'multipole': True,
'range': (200., 1000.)
}
s.trace = (10, 1, 20)
s.track = [1, 1, 1, 2, 1, 1]
s.ufs_mesh = mesh
s.resonance_scattering = {
'enable': True,
'method': 'rvs',
'energy_min': 1.0,
'energy_max': 1000.0,
'nuclides': ['U235', 'U238', 'Pu239']
}
s.volume_calculations = openmc.VolumeCalculation(domains=[openmc.Cell()],
samples=1000,
lower_left=(-10., -10.,
-10.),
upper_right=(10., 10.,
10.))
s.create_fission_neutrons = True
s.log_grid_bins = 2000
s.photon_transport = False
s.electron_treatment = 'led'
s.write_initial_source = True
# Make sure exporting XML works
s.export_to_xml()
# Generate settings from XML
s = openmc.Settings.from_xml()
assert s.run_mode == 'fixed source'
assert s.batches == 1000
assert s.generations_per_batch == 10
assert s.inactive == 100
assert s.particles == 1000000
assert s.max_lost_particles == 5
assert s.rel_max_lost_particles == 1e-4
assert s.keff_trigger == {'type': 'std_dev', 'threshold': 0.001}
assert s.energy_mode == 'continuous-energy'
assert s.max_order == 5
assert isinstance(s.source[0], openmc.Source)
assert isinstance(s.source[0].space, openmc.stats.Point)
assert s.output == {'summary': True, 'tallies': False, 'path': 'here'}
assert s.verbosity == 7
assert s.sourcepoint == {
'batches': [50, 150, 500, 1000],
'separate': True,
'write': True,
'overwrite': True
}
assert s.statepoint == {'batches': [50, 150, 500, 1000]}
assert s.surf_source_read == {'path': 'surface_source_1.h5'}
assert s.surf_source_write == {'surface_ids': [2], 'max_particles': 200}
assert s.confidence_intervals
assert s.ptables
assert s.seed == 17
assert s.survival_biasing
assert s.cutoff == {
'weight': 0.25,
'weight_avg': 0.5,
'energy_neutron': 1.0e-5,
'energy_photon': 1000.0,
'energy_electron': 1.0e-5,
'energy_positron': 1.0e-5
}
assert isinstance(s.entropy_mesh, openmc.RegularMesh)
assert s.entropy_mesh.lower_left == [-10., -10., -10.]
assert s.entropy_mesh.upper_right == [10., 10., 10.]
assert s.entropy_mesh.dimension == [5, 5, 5]
assert s.trigger_active
assert s.trigger_max_batches == 10000
assert s.trigger_batch_interval == 50
assert not s.no_reduce
assert s.tabular_legendre == {'enable': True, 'num_points': 50}
assert s.temperature == {
'default': 293.6,
'method': 'interpolation',
'multipole': True,
'range': [200., 1000.]
}
assert s.trace == [10, 1, 20]
assert s.track == [1, 1, 1, 2, 1, 1]
assert isinstance(s.ufs_mesh, openmc.RegularMesh)
assert s.ufs_mesh.lower_left == [-10., -10., -10.]
assert s.ufs_mesh.upper_right == [10., 10., 10.]
assert s.ufs_mesh.dimension == [5, 5, 5]
assert s.resonance_scattering == {
'enable': True,
'method': 'rvs',
'energy_min': 1.0,
'energy_max': 1000.0,
'nuclides': ['U235', 'U238', 'Pu239']
}
assert s.create_fission_neutrons
assert s.log_grid_bins == 2000
assert not s.photon_transport
assert s.electron_treatment == 'led'
assert s.write_initial_source == True
assert len(s.volume_calculations) == 1
vol = s.volume_calculations[0]
assert vol.domain_type == 'cell'
assert len(vol.ids) == 1
assert vol.samples == 1000
assert vol.lower_left == (-10., -10., -10.)
assert vol.upper_right == (10., 10., 10.)
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)
############################################################################### # Exporting to OpenMC settings.xml file ############################################################################### # Instantiate a Settings object, set all runtime parameters, and export to XML settings_file = openmc.Settings() settings_file.batches = batches settings_file.inactive = inactive settings_file.particles = particles # Create an initial uniform spatial source distribution over fissionable zones bounds = [-plane_value, -plane_value, -1, plane_value, plane_value, 1] uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True) settings_file.source = openmc.source.Source(space=uniform_dist) entropy_mesh = openmc.RegularMesh() entropy_mesh.lower_left = [-entropy_mesh_value, -entropy_mesh_value, -1.e50] entropy_mesh.upper_right = [entropy_mesh_value, entropy_mesh_value, 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 ############################################################################### tallies_file = openmc.Tallies() energy_filter = openmc.EnergyFilter([0., 0.625, 20.0e6]) # Instantiate flux Tally in moderator and fuel
def test_weightwindows(model):
ww_files = ('ww_n.txt', 'ww_p.txt')
cwd = Path(__file__).parent.absolute()
filepaths = [cwd / Path(f) for f in ww_files]
with cdtemp(filepaths):
# run once with variance reduction off
model.settings.weight_windows_on = False
analog_sp = model.run()
os.rename(analog_sp, 'statepoint.analog.h5')
# 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)
# energy bounds matching those of the
# generated weight windows
e_bnds = [0.0, 0.5, 2E7]
ww_n = openmc.WeightWindows(ww_mesh,
ww_n_lower_bnds,
None,
10.0,
e_bnds,
survival_ratio=1.01)
ww_p = openmc.WeightWindows(ww_mesh,
ww_p_lower_bnds,
None,
10.0,
e_bnds,
survival_ratio=1.01)
model.settings.weight_windows = [ww_n, ww_p]
# run again with variance reduction on
model.settings.weight_windows_on = True
ww_sp = model.run()
os.rename(ww_sp, 'statepoint.ww.h5')
# load both statepoints and examine results
asp = openmc.StatePoint('statepoint.analog.h5')
wsp = openmc.StatePoint('statepoint.ww.h5')
analog_tally = asp.tallies[1]
ww_tally = wsp.tallies[1]
def compare_results(particle, analog_tally, ww_tally):
# get values from each of the tallies
an_mean = analog_tally.get_values(filters=[openmc.ParticleFilter],
filter_bins=[(particle, )])
ww_mean = ww_tally.get_values(filters=[openmc.ParticleFilter],
filter_bins=[(particle, )])
# expect that more bins were scored with weight windows than
# the analog run
assert np.count_nonzero(an_mean) < np.count_nonzero(ww_mean)
an_rel_err = analog_tally.get_values(
filters=[openmc.ParticleFilter],
filter_bins=[(particle, )],
value='rel_err')
ww_rel_err = ww_tally.get_values(filters=[openmc.ParticleFilter],
filter_bins=[(particle, )],
value='rel_err')
an_rel_err[an_mean == 0.0] = 1.0
ww_rel_err[ww_mean == 0.0] = 1.0
an_avg_rel_err = np.mean(an_rel_err)
ww_avg_rel_err = np.mean(ww_rel_err)
# expect that the average relative error in the tally
# decreases
assert an_avg_rel_err > ww_avg_rel_err
# ensure that the value of the mesh bin containing the
# source is statistically similar in both runs
an_std_dev = analog_tally.get_values(
filters=[openmc.ParticleFilter],
filter_bins=[(particle, )],
value='std_dev')
ww_std_dev = ww_tally.get_values(filters=[openmc.ParticleFilter],
filter_bins=[(particle, )],
value='std_dev')
# index of the mesh bin containing the source for the higher
# energy group
source_bin_idx = (an_mean.shape[0] // 2, 0, 0)
an_source_bin = ufloat(an_mean[source_bin_idx],
an_std_dev[source_bin_idx])
ww_source_bin = ufloat(ww_mean[source_bin_idx],
ww_std_dev[source_bin_idx])
diff = an_source_bin - ww_source_bin
# check that values are within two combined standard deviations
assert abs(diff.nominal_value) / diff.std_dev < 2.0
compare_results('neutron', analog_tally, ww_tally)
compare_results('photon', analog_tally, ww_tally)
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.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
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 pin_model_attributes():
uo2 = openmc.Material(material_id=1, name='UO2')
uo2.set_density('g/cm3', 10.29769)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
uo2.depletable = True
zirc = openmc.Material(material_id=2, name='Zirc')
zirc.set_density('g/cm3', 6.55)
zirc.add_element('Zr', 1.)
zirc.depletable = False
borated_water = openmc.Material(material_id=3, name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
borated_water.add_element('O', 2.4e-2)
borated_water.add_s_alpha_beta('c_H_in_H2O')
borated_water.depletable = False
mats = openmc.Materials([uo2, zirc, borated_water])
pitch = 1.25984
fuel_or = openmc.ZCylinder(r=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(r=0.45720, name='Clad OR')
box = openmc.model.rectangular_prism(pitch,
pitch,
boundary_type='reflective')
# Define cells
fuel_inf_cell = openmc.Cell(cell_id=1, name='inf fuel', fill=uo2)
fuel_inf_univ = openmc.Universe(universe_id=1, cells=[fuel_inf_cell])
fuel = openmc.Cell(cell_id=2,
name='fuel',
fill=fuel_inf_univ,
region=-fuel_or)
clad = openmc.Cell(cell_id=3, fill=zirc, region=+fuel_or & -clad_or)
water = openmc.Cell(cell_id=4, fill=borated_water, region=+clad_or & box)
# Define overall geometry
geom = openmc.Geometry([fuel, clad, water])
uo2.volume = pi * fuel_or.r**2
settings = openmc.Settings()
settings.batches = 100
settings.inactive = 10
settings.particles = 1000
# Create a uniform spatial source distribution over fissionable zones
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
uniform_dist = openmc.stats.Box(bounds[:3],
bounds[3:],
only_fissionable=True)
settings.source = openmc.source.Source(space=uniform_dist)
entropy_mesh = openmc.RegularMesh()
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.entropy_mesh = entropy_mesh
tals = openmc.Tallies()
tal = openmc.Tally(tally_id=1, name='test')
tal.filters = [openmc.MaterialFilter(bins=[uo2])]
tal.scores = ['flux', 'fission']
tals.append(tal)
plot1 = openmc.Plot(plot_id=1)
plot1.origin = (0., 0., 0.)
plot1.width = (pitch, pitch)
plot1.pixels = (300, 300)
plot1.color_by = 'material'
plot1.filename = 'test'
plot2 = openmc.Plot(plot_id=2)
plot2.origin = (0., 0., 0.)
plot2.width = (pitch, pitch)
plot2.pixels = (300, 300)
plot2.color_by = 'cell'
plots = openmc.Plots((plot1, plot2))
chain = './test_chain.xml'
chain_file_xml = """<?xml version="1.0"?>
<depletion_chain>
<nuclide name="Xe136" decay_modes="0" reactions="0" />
<nuclide name="U235" decay_modes="0" reactions="1">
<reaction type="fission" Q="200000000."/>
<neutron_fission_yields>
<energies>2.53000e-02</energies>
<fission_yields energy="2.53000e-02">
<products>Xe136</products>
<data>1.0</data>
</fission_yields>
</neutron_fission_yields>
</nuclide>
</depletion_chain>
"""
operator_kwargs = {'chain_file': chain}
return (mats, geom, settings, tals, plots, operator_kwargs, chain_file_xml)
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]
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.temperature = {
'method': 'nearest',
'tolerance': 300.0,
'multipole': False
}
settings_file.resonance_scattering = {'enable': False, 'method': 'dbrc'}
# Create an initial uniform spatial source distribution over fissionable zones
bounds = [-0.6300, -0.6300, -1, 0.6300, 0.6300, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
entropy_mesh = openmc.RegularMesh()
entropy_mesh.lower_left = [-0.4096, -0.4096, -1.e50]
entropy_mesh.upper_right = [0.4096, 0.4096, 1.e50]
entropy_mesh.dimension = [10, 10, 1]
settings_file.entropy_mesh = entropy_mesh
settings_file.export_to_xml()
# plot setting
plot = openmc.Plot(plot_id=1)
plot.origin = [0, 0, 0]
plot.width = [1.26, 1.26]
plot.pixels = [500, 500]
plot.color_by = 'material'
# Instantiate a Plots object and export to XML
plot_file = openmc.Plots([plot])
# Compute cell areas
volume = {}
volume[fuel_material] = np.pi * (fuel_surf.coefficients['r'] ** 2) * 2*np.abs(core_surf_lowertank_wall_in.coefficients['z0']) * 10 * 10 * 15*15
#volume[fuel_cold] = np.pi * s1.coefficients['r'] ** 2
# 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