def simulate(self):
"""this runs a simple tbr simulation using openmc and returns the
tritium breeding ratio"""
breeder_material = openmc.Material(1, "PbLi") # Pb84.2Li15.8
breeder_material.add_element('Pb', 84.2, percent_type='ao')
breeder_material.add_element('Li',
15.8,
percent_type='ao',
enrichment=50.0,
enrichment_target='Li6',
enrichment_type='ao') # 50% enriched
breeder_material.set_density('atom/b-cm',
3.2720171e-2) # around 11 g/cm3
mats = openmc.Materials([breeder_material])
# GEOMETRY
# surfaces
vessel_inner = openmc.Sphere(r=500)
breeder_blanket_outer_surface = openmc.Sphere(r=600,
boundary_type='vacuum')
# cells
inner_vessel_region = -vessel_inner
inner_vessel_cell = openmc.Cell(region=inner_vessel_region)
breeder_blanket_region = -breeder_blanket_outer_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
# universe
universe = openmc.Universe(
cells=[inner_vessel_cell, breeder_blanket_cell])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS
# Instantiate a Settings object
sett = openmc.Settings()
sett.batches = 2
sett.inactive = 0
sett.particles = 100
sett.run_mode = 'fixed source'
# Create a DT point source
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
sett.source = source
tallies = openmc.Tallies()
# added a cell tally for tritium production
cell_filter = openmc.CellFilter(breeder_blanket_cell)
tbr_tally = openmc.Tally(name='TBR')
tbr_tally.filters = [cell_filter]
tbr_tally.scores = [
'(n,Xt)'
] # MT 205 is the (n,Xt) reaction where X is a wildcard, if MT 105 or (n,t) then some tritium production will be missed, for example (n,nt) which happens in Li7 would be missed
tallies.append(tbr_tally)
# Run OpenMC!
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run()
# open the results file
sp = openmc.StatePoint(sp_filename)
# access the tally using pandas dataframes
tbr_tally = sp.get_tally(name='TBR')
df = tbr_tally.get_pandas_dataframe()
tbr_tally_result = df['mean'].sum()
return tbr_tally_result
def test_full(run_in_tmpdir, problem, multiproc):
"""Full system test suite.
Runs an entire OpenMC simulation with depletion coupling and verifies
that the outputs match a reference file. Sensitive to changes in
OpenMC.
This test runs a complete OpenMC simulation and tests the outputs.
It will take a while.
"""
geometry, lower_left, upper_right = problem
# OpenMC-specific settings
settings = openmc.Settings()
settings.particles = 100
settings.batches = 10
settings.inactive = 0
space = openmc.stats.Box(lower_left, upper_right)
settings.source = openmc.Source(space=space)
settings.seed = 1
settings.verbosity = 1
# Create operator
chain_file = Path(__file__).parents[2] / 'chain_simple.xml'
op = openmc.deplete.Operator(geometry, settings, chain_file)
op.round_number = True
# Power and timesteps
dt1 = 15. * 24 * 60 * 60 # 15 days
dt2 = 1.5 * 30 * 24 * 60 * 60 # 1.5 months
N = floor(dt2 / dt1)
dt = np.full(N, dt1)
power = 2.337e15 * 4 * JOULE_PER_EV * 1e6 # MeV/second cm from CASMO
# Perform simulation using the predictor algorithm
openmc.deplete.pool.USE_MULTIPROCESSING = multiproc
openmc.deplete.PredictorIntegrator(op, dt, power).integrate()
# Get path to test and reference results
path_test = op.output_dir / 'depletion_results.h5'
path_reference = Path(__file__).with_name('test_reference.h5')
# If updating results, do so and return
if config['update']:
shutil.copyfile(str(path_test), str(path_reference))
return
# Load the reference/test results
res_test = openmc.deplete.ResultsList.from_hdf5(path_test)
res_ref = openmc.deplete.ResultsList.from_hdf5(path_reference)
# Assert same mats
for mat in res_ref[0].mat_to_ind:
assert mat in res_test[0].mat_to_ind, \
"Material {} not in new results.".format(mat)
for nuc in res_ref[0].nuc_to_ind:
assert nuc in res_test[0].nuc_to_ind, \
"Nuclide {} not in new results.".format(nuc)
for mat in res_test[0].mat_to_ind:
assert mat in res_ref[0].mat_to_ind, \
"Material {} not in old results.".format(mat)
for nuc in res_test[0].nuc_to_ind:
assert nuc in res_ref[0].nuc_to_ind, \
"Nuclide {} not in old results.".format(nuc)
tol = 1.0e-6
for mat in res_test[0].mat_to_ind:
for nuc in res_test[0].nuc_to_ind:
_, y_test = res_test.get_atoms(mat, nuc)
_, y_old = res_ref.get_atoms(mat, nuc)
# Test each point
correct = True
for i, ref in enumerate(y_old):
if ref != y_test[i]:
if ref != 0.0:
correct = np.abs(y_test[i] - ref) / ref <= tol
else:
correct = False
assert correct, "Discrepancy in mat {} and nuc {}\n{}\n{}".format(
mat, nuc, y_old, y_test)
# Compare statepoint files with depletion results
t_test, k_test = res_test.get_eigenvalue()
t_ref, k_ref = res_ref.get_eigenvalue()
k_state = np.empty_like(k_ref)
n_tallies = np.empty(N + 1, dtype=int)
# Get statepoint files for all BOS points and EOL
for n in range(N + 1):
statepoint = openmc.StatePoint("openmc_simulation_n{}.h5".format(n))
k_n = statepoint.k_combined
k_state[n] = [k_n.nominal_value, k_n.std_dev]
n_tallies[n] = len(statepoint.tallies)
# Look for exact match pulling from statepoint and depletion_results
assert np.all(k_state == k_test)
assert np.allclose(k_test, k_ref)
# Check that no additional tallies are loaded from the files
assert np.all(n_tallies == 0)
breeder_blanket_region = +first_wall_outer_surface & -breeder_blanket_outer_surface & +central_shield_outer_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
universe = openmc.Universe(cells=[
central_sol_cell, central_shield_cell, first_wall_cell,
breeder_blanket_cell
])
geom = openmc.Geometry(universe)
geom.export_to_xml()
# A blank settings.xml is exported to allow the openmc plotter to work
sett = openmc.Settings()
sett.export_to_xml()
# makes the 3d "cube" style geometry
vox_plot = openmc.Plot()
vox_plot.type = 'voxel'
vox_plot.width = (1500., 1500., 1500.)
vox_plot.pixels = (200, 200, 200)
vox_plot.filename = 'plot_3d_tokamak'
vox_plot.color_by = 'material'
# vox_plot.colors = {copper: 'blue'} # materials can be coloured using this command
plots = openmc.Plots([vox_plot])
plots.export_to_xml()
openmc.plot_geometry()
def simulate_cylinder_cask_csg(self, material, source, height,
outer_radius, thickness, batches,
particles):
"""Makes a CSG cask geometry runs a simulation and returns the result"""
mats = openmc.Materials([material])
outer_cylinder = openmc.ZCylinder(r=outer_radius)
inner_cylinder = openmc.ZCylinder(r=outer_radius - thickness)
inner_top = openmc.ZPlane(z0=height * 0.5)
inner_bottom = openmc.ZPlane(z0=-height * 0.5)
outer_top = openmc.ZPlane(z0=(height * 0.5) + thickness)
outer_bottom = openmc.ZPlane(z0=(-height * 0.5) - thickness)
sphere_1 = openmc.Sphere(r=100, boundary_type='vacuum')
cylinder_region = -outer_cylinder & +inner_cylinder & -inner_top & +inner_bottom
cylinder_cell = openmc.Cell(region=cylinder_region)
cylinder_cell.fill = material
top_cap_region = -outer_top & +inner_top & -outer_cylinder
top_cap_cell = openmc.Cell(region=top_cap_region)
top_cap_cell.fill = material
bottom_cap_region = +outer_bottom & -inner_bottom & -outer_cylinder
bottom_cap_cell = openmc.Cell(region=bottom_cap_region)
bottom_cap_cell.fill = material
inner_void_region = -inner_cylinder & -inner_top & +inner_bottom
inner_void_cell = openmc.Cell(region=inner_void_region)
# sphere 1 region is below -sphere_1 and not (~) in the other regions
sphere_1_region = -sphere_1
sphere_1_cell = openmc.Cell(region=sphere_1_region
& ~bottom_cap_region
& ~top_cap_region
& ~cylinder_region
& ~inner_void_region)
universe = openmc.Universe(cells=[
inner_void_cell, cylinder_cell, top_cap_cell, bottom_cap_cell,
sphere_1_cell
])
geom = openmc.Geometry(universe)
# Instantiate a Settings object
sett = openmc.Settings()
sett.batches = batches
sett.particles = particles
sett.inactive = 0
sett.run_mode = 'fixed source'
sett.photon_transport = True
sett.source = source
cell_filter = openmc.CellFilter(
[cylinder_cell, top_cap_cell, bottom_cap_cell])
tally = openmc.Tally(name='csg_heating')
tally.filters = [cell_filter]
tally.scores = ['heating']
tallies = openmc.Tallies()
tallies.append(tally)
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run()
# open the results file
results = openmc.StatePoint(sp_filename)
# access the tally using pandas dataframes
tally = results.get_tally(name='csg_heating')
tally_df = tally.get_pandas_dataframe()
return tally_df['mean'].sum()
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 make_model_mg(nameoflib):
element = name_el(PATH_INOUT_DATA)
nuclide = xml_f(PATH_REACTION_XML)
print("Start")
print(nuclide, element)
diction_1, diction_2, diction_3, dzz, hss, rods, densna, type_fuel, type_assembly = run(PATH_INOUT_DATA)
print('First point', dzz, hss,rods)
nkas = len(rods)
diction_mg = translate_mg(nkas, rods, dzz, [sum(dzz)])
outuniv, outsurface = _make_outer_mg_universe(diction_mg[(1,1)])
#
assemblies = {}
load_dict = {}
hexsurf, rodfaces = make_surfaces([sum(dzz)], HPITCH)
hexsurf, zfaces = make_surfaces(dzz, HPITCH)
for i in range(NASS):
if (rods[i]):
assemblies[i+1] = make_abl_mg_cells(diction_mg, hexsurf, rodfaces, i+1)
load_dict[i+1] = [i+1]
else:
assemblies[i+1] = make_abl_mg_cells(diction_mg, hexsurf, zfaces, i+1)
load_dict[i+1] = [i+1]
print("Making load ..")
##################################################################################
load = make_load("BN-800", outuniv, assemblies, NRING, HPITCH, sum(dzz), load_dict)
root_cell = openmc.Cell(cell_id=0, region=-outsurface, fill=load)
root_univ = openmc.Universe(universe_id=0, cells = [root_cell])
# Instantiate a Geometry, register the root Universe, and export to XML
openmc_geometry = openmc.Geometry()
openmc_geometry.root_universe = root_univ
openmc_geometry.export_to_xml()
materials_list = openmc_geometry.get_all_materials()
materials_file = openmc.Materials([v for v in materials_list.values()])
materials_file.cross_sections = nameoflib
materials_file.export_to_xml()
# Instantiate a Settings object, set all runtime parameters, and export to XML
settings_file = openmc.Settings()
##########################SETTINGS
settings_file.batches = 120
settings_file.inactive = 20
settings_file.particles = 1000
###########################
settings_file.temperature = {'range':(250,2500)}
# Tell OpenMC this is a multi-group problem
settings_file.energy_mode = 'multi-group'
settings_file.source = openmc.Source(space=openmc.stats.Box(
[-160.04, -160.04, 0.0], [160.04, 160.04, 170], only_fissionable=True))
settings_file.export_to_xml()
return openmc
plot_1 = openmc.Plot(plot_id=1)
plot_1.filename = 'plot_1'
plot_1.origin = [0.0, 0.0, 8.]
plot_1.width = [520.26, 520.26]
plot_1.pixels = [2400, 2400]
plot_1.color = 'mat'
plot_1.basis = 'xy'
plot_2 = openmc.Plot(plot_id=2)
plot_2.filename = 'plot_2'
plot_2.origin = [0.0, 0.0, 0.0]
plot_2.width = [520.26, 350.2]
plot_2.pixels = [1200, 1200]
plot_2.color = 'mat'
plot_2.basis = 'xz'
# Instantiate a Plots collection and export to XML
plot_file = openmc.Plots([plot_1, plot_2])
def make_geometry_tallies(batches, nps, enrichment_fraction, inner_radius,
thickness, breeder_material_name, temperature_in_C):
#print('simulating ',batches,enrichment_fraction,inner_radius,thickness,breeder_material_name)
#MATERIALS#
breeder_material = make_breeder_materials(enrichment_fraction,
breeder_material_name,
temperature_in_C)
eurofer = make_eurofer()
copper = make_copper()
mats = openmc.Materials([breeder_material, eurofer, copper])
mats.export_to_xml('materials.xml')
#GEOMETRY#
central_sol_surface = openmc.ZCylinder(R=100)
central_shield_outer_surface = openmc.ZCylinder(R=110)
first_wall_inner_surface = openmc.Sphere(R=inner_radius)
first_wall_outer_surface = openmc.Sphere(R=inner_radius + 10)
breeder_blanket_outer_surface = openmc.Sphere(R=inner_radius + 10.0 +
thickness)
vessel_outer_surface = openmc.Sphere(R=inner_radius + 10.0 + thickness +
10.0,
boundary_type='vacuum')
central_sol_region = -central_sol_surface & -breeder_blanket_outer_surface
central_sol_cell = openmc.Cell(region=central_sol_region)
central_sol_cell.fill = copper
central_shield_region = +central_sol_surface & -central_shield_outer_surface & -breeder_blanket_outer_surface
central_shield_cell = openmc.Cell(region=central_shield_region)
central_shield_cell.fill = eurofer
inner_void_region = -first_wall_inner_surface & +central_shield_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
first_wall_region = -first_wall_outer_surface & +first_wall_inner_surface & +central_shield_outer_surface
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = eurofer
breeder_blanket_region = +first_wall_outer_surface & -breeder_blanket_outer_surface & +central_shield_outer_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
vessel_region = +breeder_blanket_outer_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = 'vessel'
vessel_cell.fill = eurofer
universe = openmc.Universe(cells=[
central_sol_cell, central_shield_cell, inner_void_cell,
first_wall_cell, breeder_blanket_cell, vessel_cell
])
#plt.show(universe.plot(width=(1500,1500),basis='xz'))
geom = openmc.Geometry(universe)
# geom.export_to_xml('geometry.xml')
#SIMULATION SETTINGS#
sett = openmc.Settings()
sett.batches = batches
sett.inactive = 1
sett.particles = nps
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((150, 0, 0))
source.angle = openmc.stats.Isotropic()
#source.energy = openmc.stats.Discrete([14.08e6], [1])
source.energy = openmc.stats.Muir(
e0=14080000.0, m_rat=5.0, kt=20000.0
) #neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
sett.export_to_xml('settings.xml')
#tally filters
particle_filter = openmc.ParticleFilter([1]) #1 is neutron, 2 is photon
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
cell_filter_vessel = openmc.CellFilter(vessel_cell)
surface_filter_front = openmc.SurfaceFilter(first_wall_inner_surface)
surface_filter_rear = openmc.SurfaceFilter(breeder_blanket_outer_surface)
energy_bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
energy_filter = openmc.EnergyFilter(energy_bins)
#TALLIES#
tallies = openmc.Tallies()
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ['205']
tallies.append(tally)
tally = openmc.Tally(name='blanket_leakage')
tally.filters = [surface_filter_rear, particle_filter]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='vessel_leakage')
tally.filters = [surface_filter_rear, particle_filter]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='rear_neutron_spectra')
tally.filters = [surface_filter_rear, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='front_neutron_spectra')
tally.filters = [surface_filter_front, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='breeder_blanket_spectra')
tally.filters = [cell_filter_breeder, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='vacuum_vessel_spectra')
tally.filters = [cell_filter_vessel, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='DPA')
tally.filters = [cell_filter_vessel, particle_filter]
tally.scores = ['444']
tallies.append(tally)
#RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
#RETRIEVING TALLY RESULTS
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
json_output = {
'enrichment_fraction': enrichment_fraction,
'inner_radius': inner_radius,
'thickness': thickness,
'breeder_material_name': breeder_material_name,
'temperature_in_C': temperature_in_C
}
tallies_to_retrieve = ['TBR', 'DPA', 'blanket_leakage', 'vessel_leakage']
for tally_name in tallies_to_retrieve:
tally = sp.get_tally(name=tally_name)
tally_result = tally.sum[0][0][
0] / batches #for some reason the tally sum is a nested list
tally_std_dev = tally.std_dev[0][0][
0] / batches #for some reason the tally std_dev is a nested list
json_output[tally_name] = {
'value': tally_result,
'std_dev': tally_std_dev
}
spectra_tallies_to_retrieve = [
'front_neutron_spectra', 'breeder_blanket_spectra',
'vacuum_vessel_spectra', 'rear_neutron_spectra'
]
for spectra_name in spectra_tallies_to_retrieve:
spectra_tally = sp.get_tally(name=spectra_name)
spectra_tally_result = [entry[0][0] for entry in spectra_tally.mean]
spectra_tally_std_dev = [
entry[0][0] for entry in spectra_tally.std_dev
]
json_output[spectra_name] = {
'value': spectra_tally_result,
'std_dev': spectra_tally_std_dev,
'energy_groups': list(energy_bins)
}
return json_output
def hexfunction(pitch, pack_frac):
settings = openmc.Settings()
# Set high tolerance to allow use of lower temperature xs
settings.temperature['tolerance'] = 1000
settings.temperature['method'] = 'nearest'
settings.temperature['multipole'] = True
settings.cutoff = {'energy': 1e-8} #energy cutoff in eV
#############################
### MATERIALS ###
#############################
mat_list = []
enrichment = 20.0
uo2 = openmc.Material(1, "uo2")
uo2.add_element('U', 1.0, enrichment=enrichment)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.97)
uo2.temperature = 900 #kelvin
mat_list.append(uo2)
graphite = openmc.Material(2, "graphite")
graphite.set_density('g/cm3', 1.1995)
graphite.add_element('C', 1.0)
graphite.add_s_alpha_beta('c_Graphite')
graphite.temperature = 900 #kelvin
mat_list.append(graphite)
# sodium = openmc.Material(3, "sodium")
sodium = openmc.Material()
sodium.set_density('g/cm3', 0.8017) # 900 K
sodium.add_element('Na', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
sodium.temperature = 900 #kelvin
mat_list.append(sodium)
# naoh = openmc.Material(6, "naoh")
naoh = openmc.Material()
naoh.set_density('g/cm3', 1.5) # 900 K
naoh.add_element('Na', 1.0)
naoh.add_element('O', 1.0)
naoh.add_element('H', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
naoh.temperature = 900 #kelvin
mat_list.append(naoh)
# TRISO Materials
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.5)
# fuel.add_nuclide('U235', 4.6716e-02)
fuel.add_nuclide('U235', 0.0667372)
# fuel.add_nuclide('U238', 2.8697e-01)
fuel.add_nuclide('U238', 0.2669488)
fuel.add_nuclide('O16', 5.0000e-01)
fuel.add_element('C', 1.6667e-01)
mat_list.append(fuel)
buff = openmc.Material(name='Buffer')
buff.set_density('g/cm3', 1.0)
buff.add_element('C', 1.0)
buff.add_s_alpha_beta('c_Graphite')
mat_list.append(buff)
PyC1 = openmc.Material(name='PyC1')
PyC1.set_density('g/cm3', 1.9)
PyC1.add_element('C', 1.0)
PyC1.add_s_alpha_beta('c_Graphite')
mat_list.append(PyC1)
PyC2 = openmc.Material(name='PyC2')
PyC2.set_density('g/cm3', 1.87)
PyC2.add_element('C', 1.0)
PyC2.add_s_alpha_beta('c_Graphite')
mat_list.append(PyC2)
SiC = openmc.Material(name='SiC')
SiC.set_density('g/cm3', 3.2)
SiC.add_element('C', 0.5)
SiC.add_element('Si', 0.5)
mat_list.append(SiC)
fuel_temp = 900
homogeneous_fuel = build_fuel_material(fuel_temp, pack_frac)
mat_list.append(homogeneous_fuel)
mats = openmc.Materials(mat_list)
mats.export_to_xml()
#############################
### GEOMETRY ###
#############################
pitch = 17.4
fuel_bottom = -pitch / 2
fuel_top = pitch / 2
coolant_r = 4
# fuel_r = (pitch/2 - coolant_r)/2
fuel_r = (pitch / (3**(1 / 2)) - coolant_r) / 2
hex_universe = openmc.Universe()
top = openmc.ZPlane(z0=fuel_top, boundary_type='reflective')
bottom = openmc.ZPlane(z0=fuel_bottom, boundary_type='reflective')
surf_fuel = openmc.ZCylinder(r=fuel_r)
# Make TRISOS to be filled in fuel cylinders by chopping up
# fuel cylinder into segments
n_cyls = 40
fuel_segment_heights = np.linspace(fuel_bottom, fuel_top, n_cyls)
segment_height = fuel_segment_heights[1] - fuel_segment_heights[0]
fuel_planes = [bottom]
fuel_cells = []
for i, height in enumerate(fuel_segment_heights[1:-1]):
this_plane = openmc.ZPlane(z0=height)
fuel_planes.append(this_plane)
this_cell = openmc.Cell()
this_cell.region = +fuel_planes[i] & -fuel_planes[i + 1] & -surf_fuel
fuel_cells.append(copy.deepcopy(this_cell))
# last cell
fuel_planes.append(top)
this_cell = openmc.Cell()
this_cell.region = +fuel_planes[-2] & -fuel_planes[-1] & -surf_fuel
fuel_cells.append(copy.deepcopy(this_cell))
# Make fuel cylinder
fuel_cyl_top = openmc.ZPlane(z0=segment_height / 2)
fuel_cyl_bottom = openmc.ZPlane(z0=-segment_height / 2)
fuel_triso_region = -surf_fuel & +fuel_cyl_bottom & -fuel_cyl_top
outer_radius = 425. * 1e-4
# openmc.model.triso._Cylinder.from_region(fuel_region, outer_radius)
spheres = [openmc.Sphere(r=r * 1e-4) for r in [215., 315., 350., 385.]]
cells = [
openmc.Cell(fill=fuel, region=-spheres[0]),
openmc.Cell(fill=buff, region=+spheres[0] & -spheres[1]),
openmc.Cell(fill=PyC1, region=+spheres[1] & -spheres[2]),
openmc.Cell(fill=SiC, region=+spheres[2] & -spheres[3]),
openmc.Cell(fill=PyC2, region=+spheres[3])
]
triso_univ = openmc.Universe(cells=cells)
outer_radius = 425. * 1e-4
centers = openmc.model.pack_spheres(radius=outer_radius,
region=fuel_triso_region,
pf=pack_frac)
trisos = [openmc.model.TRISO(outer_radius, triso_univ, c) for c in centers]
outside_trisos = openmc.Intersection(~t.region for t in trisos)
# background_region = outside_trisos & +fuel_cyl_bottom & \
# -fuel_cyl_top & -surf_fuel
background_region = outside_trisos
background_cell = openmc.Cell(fill=graphite, region=background_region)
fuel_triso_univ = openmc.Universe()
fuel_triso_univ.add_cell(background_cell)
for idx, triso in enumerate(trisos):
fuel_triso_univ.add_cell(triso)
# Fill in fuel cells with triso cells and translate to location
for i, cell in enumerate(fuel_cells):
cell_height = segment_height * (i + 1 / 2) + fuel_bottom
cell.translation = [0, 0, cell_height]
cell.fill = fuel_triso_univ
fuel_cell_univ = openmc.Universe(cells=fuel_cells)
# For testing solid fuel
# test_region = +bottom & -top & -surf_fuel
# fuel_cell = openmc.Cell(region=test_region, fill=fuel)
# fuel_cell_univ = openmc.Universe(cells=[fuel_cell])
coolant_cyl = openmc.ZCylinder(r=coolant_r)
coolant_region = -coolant_cyl
coolant_cell = openmc.Cell()
coolant_cell.fill = naoh
coolant_cell.region = coolant_region
hex_universe.add_cell(coolant_cell)
hex_prism = openmc.get_hexagonal_prism(edge_length=pitch / (3**1 / 2),
boundary_type='reflective')
graphite_region = hex_prism & +coolant_cyl & -top & +bottom
graphite_cell = openmc.Cell()
graphite_cell.fill = graphite
graphite_cell.region = graphite_region
hex_universe.add_cell(graphite_cell)
fuel_cells = []
root3 = 3**(1 / 2)
half_to_vertex = pitch / root3 / 2
half_to_edge = pitch / 4
# fuel_id = 100
offset_angle = 30
n_pins = 6
for i in range(n_pins):
theta = (offset_angle + i / n_pins * 360) * pi / 180
r = coolant_r + fuel_r + 0.01
x = r * np.cos(theta)
y = r * np.sin(theta)
fuel_cyl_bound = openmc.ZCylinder(x0=x, y0=y, r=fuel_r)
graphite_cell.region &= +fuel_cyl_bound
fuel_cell = openmc.Cell()
fuel_cell.fill = copy.deepcopy(fuel_cell_univ)
fuel_cell.translation = [x, y, 0]
fuel_cell.region = -fuel_cyl_bound & -top & +bottom
# fuel_cell.id = fuel_id
# fuel_id += 1
fuel_cells.append(fuel_cell)
hex_universe.add_cell(fuel_cell)
geom = openmc.Geometry(hex_universe)
# geom = openmc.Geometry(fuel_cell_univ)
geom.export_to_xml()
#####################################
### SOURCE/BATCHES ###
#####################################
point = openmc.stats.Point((0, 0, 0))
src = openmc.Source(space=point)
settings.source = src
settings.batches = 50
settings.inactive = 10
settings.particles = 200
settings.export_to_xml()
#############################
### TALLIES ###
#############################
# Instantiate an empty Tallies object
tallies_file = openmc.Tallies()
# K-Eigenvalue (infinity) tallies
fiss_rate = openmc.Tally(name='fiss. rate')
fiss_rate.scores = ['nu-fission']
tallies_file.append(fiss_rate)
abs_rate = openmc.Tally(name='abs. rate')
abs_rate.scores = ['absorption']
tallies_file.append(abs_rate)
# Resonance Escape Probability tallies
therm_abs_rate = openmc.Tally(name='therm. abs. rate')
therm_abs_rate.scores = ['absorption']
therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_abs_rate)
# Thermal Flux Utilization tallies
# fuel_therm_abs_rate = openmc.Tally(name='fuel therm. abs. rate')
# fuel_therm_abs_rate.scores = ['absorption']
# fuel_therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625]),
# openmc.CellFilter([fuel_cell])]
# tallies_file.append(fuel_therm_abs_rate)
# Fast Fission Factor tallies
therm_fiss_rate = openmc.Tally(name='therm. fiss. rate')
therm_fiss_rate.scores = ['nu-fission']
therm_fiss_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_fiss_rate)
tallies_file.export_to_xml()
#############################
### PLOTTING ###
#############################
zs = np.linspace(0, 1, 2)
plots = []
for z in zs:
p = openmc.Plot()
p.filename = 'pinplot' + str(z)
p.width = (1.4 * pitch, 1.4 * pitch)
p.pixels = (2000, 2000)
p.color_by = 'material'
p.origin = [0, 0, z]
# p.color_by = 'cell'
# p.colors = {homogeneous_fuel: 'yellow', naoh: 'grey', graphite: 'black'}
p.colors = {fuel: 'yellow', naoh: 'grey', graphite: 'black'}
plots.append(copy.deepcopy(p))
plots = openmc.Plots(plots)
plots.export_to_xml()
# openmc.plot_geometry(output = False)
openmc.plot_geometry()
# pngstring = 'pinplot{}.png'.format(str(pitch))
# subprocess.call(['convert','pinplot.ppm',pngstring])
# subprocess.call(['mv',pngstring,'figures/'+pngstring])
#############################
### EXECUTION ###
#############################
# openmc.run(output=False)
openmc.run()
sp = openmc.StatePoint('statepoint.{}.h5'.format(settings.batches))
# Collect all the tallies
fiss_rate = sp.get_tally(name='fiss. rate')
fiss_rate_df = fiss_rate.get_pandas_dataframe()
abs_rate = sp.get_tally(name='abs. rate')
abs_rate_df = abs_rate.get_pandas_dataframe()
therm_abs_rate = sp.get_tally(name='therm. abs. rate')
therm_abs_rate_df = therm_abs_rate.get_pandas_dataframe()
fuel_therm_abs_rate = sp.get_tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate_df = fuel_therm_abs_rate.get_pandas_dataframe()
therm_fiss_rate = sp.get_tally(name='therm. fiss. rate')
therm_fiss_rate_df = therm_fiss_rate.get_pandas_dataframe()
# Compute k-infinity
kinf = fiss_rate / abs_rate
kinf_df = kinf.get_pandas_dataframe()
# Compute resonance escape probability
res_esc = (therm_abs_rate) / (abs_rate)
res_esc_df = res_esc.get_pandas_dataframe()
# Compute fast fission factor
fast_fiss = fiss_rate / therm_fiss_rate
fast_fiss_df = fast_fiss.get_pandas_dataframe()
# Compute thermal flux utilization
therm_util = fuel_therm_abs_rate / therm_abs_rate
therm_util_df = therm_util.get_pandas_dataframe()
# Compute neutrons produced per absorption
eta = therm_fiss_rate / fuel_therm_abs_rate
eta_df = eta.get_pandas_dataframe()
columns = [
'pitch', 'enrichment', 'kinf mean', 'kinf sd', 'res_esc mean',
'res_esc sd', 'fast_fiss mean', 'fast_fiss sd', 'therm_util mean',
'therm_util sd', 'eta mean', 'eta sd'
]
data = [[
pitch, enrichment, kinf_df['mean'][0], kinf_df['std. dev.'][0],
res_esc_df['mean'][0], res_esc_df['std. dev.'][0],
fast_fiss_df['mean'][0], fast_fiss_df['std. dev.'][0],
therm_util_df['mean'][0], therm_util_df['std. dev.'][0],
eta_df['mean'][0], eta_df['std. dev.'][0]
]]
all_tallies = pd.DataFrame(data, columns=columns)
return all_tallies
def simulate_model(
enrichment,
thickness,
breeder_material_name="Li4SiO4",
temperature_in_C=500,
batches=2,
nps=500,
inner_radius=500,
):
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment=enrichment,
temperature_in_C=temperature_in_C,
).neutronics_material
SS316 = Material(material_name="SS316").neutronics_material
copper = Material(material_name="copper").neutronics_material
mats = openmc.Materials([breeder_material, SS316, copper])
mats.export_to_xml("materials.xml")
# GEOMETRY#
first_wall_inner_surface = openmc.Sphere(r=inner_radius)
first_wall_outer_surface = openmc.Sphere(r=inner_radius + 10.)
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius + 10. +
thickness)
vessel_outer_surface = openmc.Sphere(r=inner_radius + 10.0 + thickness +
10.,
boundary_type="vacuum")
inner_void_region = -first_wall_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = "inner_void"
first_wall_region = (-first_wall_outer_surface & +first_wall_inner_surface)
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = SS316
breeder_blanket_region = (+first_wall_outer_surface
& -breeder_blanket_outer_surface)
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
vessel_region = +breeder_blanket_outer_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = "vessel"
vessel_cell.fill = SS316
universe = openmc.Universe(cells=[
inner_void_cell, first_wall_cell, breeder_blanket_cell, vessel_cell
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS#
sett = openmc.Settings()
sett.batches = batches # this is the minimum number of batches which will be run
sett.trigger_active = True
sett.trigger_max_batches = 200 # this is the maximum number of batches which will be run
sett.inactive = 0
sett.particles = nps # as we are using a trigger, we specify a small number of particles per batch
sett.run_mode = "fixed source"
source = openmc.Source()
source.space = openmc.stats.Point((150, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14.08e6], [1])
sett.source = source
# sett.export_to_xml("settings.xml")
# tally filters
particle_filter = openmc.ParticleFilter("neutron")
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
# TALLIES#
tallies = openmc.Tallies()
tally = openmc.Tally(name="TBR")
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ["(n,Xt)"]
tally.triggers = [openmc.Trigger(trigger_type='std_dev', threshold=0.01)]
tallies.append(tally)
# RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run(output=False)
# RETRIEVING TALLY RESULTS
sp = openmc.StatePoint(sp_filename)
json_output = {
"batches": batches,
"nps": nps,
"enrichment": enrichment,
"inner_radius": inner_radius,
"thickness": thickness,
"breeder_material_name": breeder_material_name,
"temperature_in_C": temperature_in_C,
}
tally = sp.get_tally(name="TBR")
df = tally.get_pandas_dataframe()
json_output["TBR"] = df["mean"].sum()
json_output["TBR_std_dev"] = df["std. dev."].sum()
return json_output
def make_materials_geometry_tallies(enrichment_list,
batches=2,
inner_radius=500,
thickness=100,
breeder_material_name='Li',
temperature_in_C=500):
if isinstance(enrichment_list, list):
enrichment = enrichment_list[0]
else:
enrichment = enrichment_list
print('simulating ', batches, enrichment, inner_radius, thickness,
breeder_material_name)
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment=enrichment,
enrichment_target='Li6',
enrichment_type='ao',
temperature_in_C=temperature_in_C).neutronics_material
eurofer = Material(material_name='eurofer').neutronics_material
mats = openmc.Materials([breeder_material, eurofer])
# GEOMETRY
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius)
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius + thickness)
vessel_inner_surface = openmc.Sphere(r=inner_radius + thickness + 10)
vessel_outer_surface = openmc.Sphere(r=inner_radius + thickness + 20,
boundary_type='vacuum')
breeder_blanket_region = -breeder_blanket_outer_surface & +breeder_blanket_inner_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
breeder_blanket_cell.name = 'breeder_blanket'
inner_void_region = -breeder_blanket_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
vessel_region = +vessel_inner_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = 'vessel'
vessel_cell.fill = eurofer
blanket_vessel_gap_region = -vessel_inner_surface & +breeder_blanket_outer_surface
blanket_vessel_gap_cell = openmc.Cell(region=blanket_vessel_gap_region)
blanket_vessel_gap_cell.name = 'blanket_vessel_gap'
universe = openmc.Universe(cells=[
inner_void_cell, breeder_blanket_cell, blanket_vessel_gap_cell,
vessel_cell
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS
sett = openmc.Settings()
# batches = 3 # this is parsed as an argument
sett.batches = batches
sett.inactive = 20
sett.particles = 5000
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(
e0=14080000.0, m_rat=5.0, kt=20000.0
) # neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
# TALLIES
tallies = openmc.Tallies()
# define filters
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
cell_filter_vessel = openmc.CellFilter(vessel_cell)
particle_filter = openmc.ParticleFilter(
'neutron') # 1 is neutron, 2 is photon
surface_filter_rear_blanket = openmc.SurfaceFilter(
breeder_blanket_outer_surface)
surface_filter_rear_vessel = openmc.SurfaceFilter(vessel_outer_surface)
energy_bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
energy_filter = openmc.EnergyFilter(energy_bins)
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder]
tally.scores = [
'205'
] # MT 205 is the (n,Xt) reaction where X is a wildcard, if MT 105 or (n,t) then some tritium production will be missed, for example (n,nt) which happens in Li7 would be missed
tallies.append(tally)
tally = openmc.Tally(name='blanket_leakage')
tally.filters = [surface_filter_rear_blanket]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='vessel_leakage')
tally.filters = [surface_filter_rear_vessel]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='breeder_blanket_spectra')
tally.filters = [cell_filter_breeder, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='vacuum_vessel_spectra')
tally.filters = [cell_filter_vessel, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='DPA')
tally.filters = [cell_filter_vessel]
tally.scores = ['444']
tallies.append(tally)
# RUN OPENMC
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
json_output = {
'enrichment': enrichment,
'inner_radius': inner_radius,
'thickness': thickness,
'breeder_material_name': breeder_material_name,
'temperature_in_C': temperature_in_C
}
tallies_to_retrieve = ['TBR', 'DPA', 'blanket_leakage', 'vessel_leakage']
for tally_name in tallies_to_retrieve:
tally = sp.get_tally(name=tally_name)
df = tally.get_pandas_dataframe()
tally_result = df['mean'].sum()
tally_std_dev = df['std. dev.'].sum()
json_output[tally_name] = {
'value': tally_result,
'std_dev': tally_std_dev
}
spectra_tallies_to_retrieve = [
'breeder_blanket_spectra', 'vacuum_vessel_spectra'
]
for spectra_name in spectra_tallies_to_retrieve:
spectra_tally = sp.get_tally(name=spectra_name)
spectra_tally_result = [entry[0][0] for entry in spectra_tally.mean]
spectra_tally_std_dev = [
entry[0][0] for entry in spectra_tally.std_dev
]
json_output[spectra_name] = {
'value': spectra_tally_result,
'std_dev': spectra_tally_std_dev,
'energy_groups': list(energy_bins)
}
return json_output
def build_inf_model(xsnames, xslibname, temperature, tempmethod='nearest'):
""" Building an infinite medium for openmc multi-group testing
Parameters:
----------
xsnames : list of str()
list with xs names
xslibname:
name of hdf5 file with cross-section library
temperature : float
value of a current temperature in K
tempmethod : {'nearest', 'interpolation'}
by default 'nearest'
"""
inf_medium = openmc.Material(name='test material', material_id=1)
inf_medium.set_density("sum")
for xs in xsnames:
inf_medium.add_nuclide(xs, 1)
INF = 11.1
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([inf_medium])
materials_file.cross_sections = xslibname
materials_file.export_to_xml()
# Instantiate boundary Planes
min_x = openmc.XPlane(boundary_type='reflective', x0=-INF)
max_x = openmc.XPlane(boundary_type='reflective', x0=INF)
min_y = openmc.YPlane(boundary_type='reflective', y0=-INF)
max_y = openmc.YPlane(boundary_type='reflective', y0=INF)
# Instantiate a Cell
cell = openmc.Cell(cell_id=1, name='cell')
cell.temperature = temperature
# Register bounding Surfaces with the Cell
cell.region = +min_x & -max_x & +min_y & -max_y
# Fill the Cell with the Material
cell.fill = inf_medium
# Create root universe
root_universe = openmc.Universe(name='root universe', cells=[cell])
# Create Geometry and set root Universe
openmc_geometry = openmc.Geometry(root_universe)
# Export to "geometry.xml"
openmc_geometry.export_to_xml()
# OpenMC simulation parameters
batches = 200
inactive = 5
particles = 5000
# Instantiate a Settings object
settings_file = openmc.Settings()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.energy_mode = 'multi-group'
settings_file.output = {'summary': False}
# Create an initial uniform spatial source distribution over fissionable zones
bounds = [-INF, -INF, -INF, INF, INF, INF]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
settings_file.temperature = {'method': tempmethod}
settings_file.source = openmc.Source(space=uniform_dist)
settings_file.export_to_xml()
def make_neutronics_model(
reactor,
firstwall_radial_thickness,
firstwall_armour_material,
firstwall_coolant_material,
firstwall_structural_material,
firstwall_armour_fraction,
firstwall_coolant_fraction,
firstwall_coolant_temperature_C,
firstwall_coolant_pressure_Pa,
firstwall_structural_fraction,
blanket_rear_wall_coolant_material,
blanket_rear_wall_structural_material,
blanket_rear_wall_coolant_fraction,
blanket_rear_wall_structural_fraction,
blanket_rear_wall_coolant_temperature_C,
blanket_rear_wall_coolant_pressure_Pa,
blanket_lithium6_enrichment_percent,
blanket_breeder_material,
blanket_coolant_material,
blanket_multiplier_material,
blanket_structural_material,
blanket_breeder_fraction,
blanket_coolant_fraction,
blanket_multiplier_fraction,
blanket_structural_fraction,
blanket_breeder_packing_fraction,
blanket_multiplier_packing_fraction,
blanket_coolant_temperature_C,
blanket_coolant_pressure_Pa,
blanket_breeder_temperature_C,
blanket_breeder_pressure_Pa,
divertor_coolant_fraction,
divertor_structural_fraction,
divertor_coolant_material,
divertor_structural_material,
divertor_coolant_temperature_C,
divertor_coolant_pressure_Pa,
center_column_shield_coolant_fraction,
center_column_shield_structural_fraction,
center_column_shield_coolant_material,
center_column_shield_structural_material,
center_column_shield_coolant_temperature_C,
center_column_shield_coolant_pressure_Pa,
inboard_tf_coils_conductor_fraction,
inboard_tf_coils_coolant_fraction,
inboard_tf_coils_structure_fraction,
inboard_tf_coils_conductor_material,
inboard_tf_coils_coolant_material,
inboard_tf_coils_structure_material,
inboard_tf_coils_coolant_temperature_C,
inboard_tf_coils_coolant_pressure_Pa,
):
"""
Makes and runs a simple OpenMC neutronics model with
the materials with the same tags as the DAGMC neutronics
geometry. The model also specifies the computational
intensity (particles and batches) and the tally to record
"""
input_parameters = locals()
# this is the underlying geometry container that is filled with the
# faceteted CAD model
universe = openmc.Universe()
geom = openmc.Geometry(universe)
center_column_shield_material = MultiMaterial(
material_tag="center_column_shield_mat",
materials=[
Material(
material_name=center_column_shield_coolant_material,
temperature_in_C=center_column_shield_coolant_temperature_C,
pressure_in_Pa=center_column_shield_coolant_pressure_Pa,
),
Material(material_name=center_column_shield_structural_material),
],
fracs=[
center_column_shield_coolant_fraction,
center_column_shield_structural_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
firstwall_material = MultiMaterial(
material_tag="firstwall_mat",
materials=[
Material(
material_name=firstwall_coolant_material,
temperature_in_C=firstwall_coolant_temperature_C,
pressure_in_Pa=firstwall_coolant_pressure_Pa,
),
Material(material_name=firstwall_structural_material),
Material(material_name=firstwall_armour_material),
],
fracs=[
firstwall_coolant_fraction,
firstwall_structural_fraction,
firstwall_armour_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
if (
blanket_multiplier_material is None
and blanket_multiplier_fraction is None
and blanket_multiplier_packing_fraction is None
):
blanket_material = MultiMaterial(
material_tag="blanket_mat",
materials=[
Material(
material_name=blanket_coolant_material,
temperature_in_C=blanket_coolant_temperature_C,
pressure_in_Pa=blanket_coolant_pressure_Pa,
),
Material(material_name=blanket_structural_material),
Material(
material_name=blanket_breeder_material,
enrichment=blanket_lithium6_enrichment_percent,
packing_fraction=blanket_breeder_packing_fraction,
temperature_in_C=blanket_breeder_temperature_C,
pressure_in_Pa=blanket_breeder_pressure_Pa,
),
],
fracs=[
blanket_coolant_fraction,
blanket_structural_fraction,
blanket_breeder_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
else:
blanket_material = MultiMaterial(
material_tag="blanket_mat",
materials=[
Material(
material_name=blanket_coolant_material,
temperature_in_C=blanket_coolant_temperature_C,
pressure_in_Pa=blanket_coolant_pressure_Pa,
),
Material(material_name=blanket_structural_material),
Material(
material_name=blanket_multiplier_material,
packing_fraction=blanket_multiplier_packing_fraction,
),
Material(
material_name=blanket_breeder_material,
enrichment=blanket_lithium6_enrichment_percent,
packing_fraction=blanket_breeder_packing_fraction,
temperature_in_C=blanket_breeder_temperature_C,
pressure_in_Pa=blanket_breeder_pressure_Pa,
),
],
fracs=[
blanket_coolant_fraction,
blanket_structural_fraction,
blanket_multiplier_fraction,
blanket_breeder_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
divertor_material = MultiMaterial(
material_tag="divertor_mat",
materials=[
Material(
material_name=divertor_coolant_material,
temperature_in_C=divertor_coolant_temperature_C,
pressure_in_Pa=divertor_coolant_pressure_Pa,
),
Material(material_name=divertor_structural_material),
],
fracs=[divertor_coolant_fraction, divertor_structural_fraction],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
inboard_tf_coils_material = MultiMaterial(
material_tag="inboard_tf_coils_mat",
materials=[
Material(
material_name=inboard_tf_coils_coolant_material,
temperature_in_C=inboard_tf_coils_coolant_temperature_C,
pressure_in_Pa=inboard_tf_coils_coolant_pressure_Pa,
),
Material(material_name=inboard_tf_coils_conductor_material),
Material(material_name=inboard_tf_coils_structure_material),
],
fracs=[
inboard_tf_coils_coolant_fraction,
inboard_tf_coils_conductor_fraction,
inboard_tf_coils_structure_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
blanket_rear_wall_material = MultiMaterial(
material_tag="blanket_rear_wall_mat",
materials=[
Material(
material_name=blanket_rear_wall_coolant_material,
temperature_in_C=blanket_rear_wall_coolant_temperature_C,
pressure_in_Pa=blanket_rear_wall_coolant_pressure_Pa,
),
Material(material_name=blanket_rear_wall_structural_material),
],
fracs=[
blanket_rear_wall_coolant_fraction,
blanket_rear_wall_structural_fraction,
],
percent_type="vo",
packing_fraction=1.0,
).openmc_material
mats = openmc.Materials(
[
center_column_shield_material,
firstwall_material,
blanket_material,
divertor_material,
inboard_tf_coils_material,
blanket_rear_wall_material,
]
)
# settings for the number of neutrons to simulate
settings = openmc.Settings()
settings.batches = 10
settings.inactive = 0
settings.particles = 1000
settings.run_mode = "fixed source"
settings.dagmc = True
# details of the birth locations and energy of the neutronis
source = openmc.Source()
source.space = openmc.stats.Point((reactor["major_radius"], 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
settings.source = source
settings.photon_transport = (
True # This line is required to switch on photons tracking
)
# details about what neutrons interactions to keep track of (called a
# tally)
tallies = openmc.Tallies()
material_filter = openmc.MaterialFilter(blanket_material)
tbr_tally = openmc.Tally(name="TBR")
tbr_tally.filters = [material_filter]
tbr_tally.scores = ["(n,Xt)"] # where X is a wild card
tallies.append(tbr_tally)
material_filter = openmc.MaterialFilter(
[blanket_material, firstwall_material, blanket_rear_wall_material]
)
blanket_heating_tally = openmc.Tally(name="blanket_heating")
blanket_heating_tally.filters = [material_filter]
blanket_heating_tally.scores = ["heating"]
tallies.append(blanket_heating_tally)
# make the model from gemonetry, materials, settings and tallies
model = openmc.model.Model(geom, mats, settings, tallies)
# run the simulation
output_filename = model.run()
"""
Reads the output file from the neutronics simulation
and prints the TBR tally result to screen
"""
# open the results file
sp = openmc.StatePoint(output_filename)
# access TBR tally
tbr_tally = sp.get_tally(name="TBR")
df = tbr_tally.get_pandas_dataframe()
tbr_tally_result = df["mean"].sum()
tbr_tally_std_dev = df["std. dev."].sum()
# access heating tally
blanket_heating_tally = sp.get_tally(name="blanket_heating")
df = blanket_heating_tally.get_pandas_dataframe()
blanket_heating_tally_result = df["mean"].sum() / 1e6
blanket_heating_tally_std_dev = df["std. dev."].sum() / 1e6
# returns all the inputs and some extra reactor attributes, merged into a
# single dictionary
return {
**input_parameters,
**{
"tbr": tbr_tally_result,
"tbr_std_dev": tbr_tally_std_dev,
"blanket_heating": blanket_heating_tally_result,
"blanket_heating_std_dev": blanket_heating_tally_std_dev,
},
}
def _build_inputs(self):
mat = openmc.Material()
mat.set_density('g/cm3', 2.6989)
mat.add_nuclide('Al27', 1.0)
materials = openmc.Materials([mat])
materials.export_to_xml()
cyl = openmc.XCylinder(r=1.0, boundary_type='vacuum')
x_plane_left = openmc.XPlane(-1.0, 'vacuum')
x_plane_center = openmc.XPlane(1.0)
x_plane_right = openmc.XPlane(1.0e9, 'vacuum')
inner_cyl_left = openmc.Cell()
inner_cyl_right = openmc.Cell()
outer_cyl = openmc.Cell()
inner_cyl_left.region = -cyl & +x_plane_left & -x_plane_center
inner_cyl_right.region = -cyl & +x_plane_center & -x_plane_right
outer_cyl.region = ~(-cyl & +x_plane_left & -x_plane_right)
inner_cyl_right.fill = mat
geometry = openmc.Geometry(
[inner_cyl_left, inner_cyl_right, outer_cyl])
geometry.export_to_xml()
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Monodirectional()
source.energy = openmc.stats.Discrete([14.0e6], [1.0])
source.particle = 'neutron'
settings = openmc.Settings()
settings.particles = 10000
settings.run_mode = 'fixed source'
settings.batches = 1
settings.photon_transport = True
settings.electron_treatment = 'ttb'
settings.cutoff = {'energy_photon': 1000.0}
settings.source = source
settings.export_to_xml()
surface_filter = openmc.SurfaceFilter(cyl)
particle_filter = openmc.ParticleFilter('photon')
current_tally = openmc.Tally()
current_tally.filters = [surface_filter, particle_filter]
current_tally.scores = ['current']
tally_tracklength = openmc.Tally()
tally_tracklength.filters = [particle_filter]
tally_tracklength.scores = ['total', 'heating']
tally_tracklength.nuclides = ['Al27', 'total']
tally_tracklength.estimator = 'tracklength'
tally_collision = openmc.Tally()
tally_collision.filters = [particle_filter]
tally_collision.scores = ['total', 'heating']
tally_collision.nuclides = ['Al27', 'total']
tally_collision.estimator = 'collision'
tally_analog = openmc.Tally()
tally_analog.filters = [particle_filter]
tally_analog.scores = ['total', 'heating']
tally_analog.nuclides = ['Al27', 'total']
tally_analog.estimator = 'analog'
tallies = openmc.Tallies(
[current_tally, tally_tracklength, tally_collision, tally_analog])
tallies.export_to_xml()
def define_Geo_Mat_Set(cells_num_dic,parameters_dic,settings_dic,temp_phy_mat,controlRod_deep,empty_reflector_height):
# Check file
if os.path.exists('geometry.xml'):
os.remove('geometry.xml')
if os.path.exists('materials.xml'):
os.remove('materials.xml')
if os.path.exists('settings.xml'):
os.remove('settings.xml')
if os.path.exists('tallies.xml'):
os.remove('tallies.xml')
# defualt temperature: 1073.5K
if settings_dic['temperature_update']== True:
temp_defualt = temp_phy_mat.min()
else:
temp_defualt = 1073.5
# Structural Material HAYNES230
structure_HAY = openmc.Material(name='HAYNES230')
structure_HAY.set_density('g/cm3',8.97)
structure_HAY.add_element('Ni',0.57,'wo')
structure_HAY.add_element('Cr',0.22,'wo')
structure_HAY.add_element('W',0.14,'wo')
structure_HAY.add_element('Mo',0.02,'wo')
structure_HAY.add_element('Fe',0.01875,'wo')
structure_HAY.add_element('Co',0.03125,'wo')
# Structural Material SS316
structure_SS = openmc.Material(name='SS316')
structure_SS.set_density('g/cm3',7.99)
structure_SS.add_element('Ni',0.12,'wo')
structure_SS.add_element('Cr',0.17,'wo')
structure_SS.add_element('Mo',0.025,'wo')
structure_SS.add_element('Mn',0.02,'wo')
structure_SS.add_element('Fe',0.665,'wo')
#Control Rod Material B4C
ControlRod_B4C = openmc.Material(name='B4C')
ControlRod_B4C.set_density('g/cm3',2.52)
ControlRod_B4C.add_nuclide('B10',4,'ao')
ControlRod_B4C.add_element('C',1,'ao')
#Reflector Material BeO
Reflector_BeO = openmc.Material(name='BeO')
Reflector_BeO.set_density('g/cm3',3.025)
Reflector_BeO.add_element('Be',1,'ao')
Reflector_BeO.add_element('O',1,'ao')
#Coolant Na
Coolant_Na = openmc.Material(name='Na')
Coolant_Na.set_density('g/cm3',0.76)
Coolant_Na.add_element('Na',1,'ao')
# Instantiate a Materials collection
materials_file = openmc.Materials([structure_HAY, structure_SS, ControlRod_B4C, Reflector_BeO, Coolant_Na])
# Number of cells
n_r = cells_num_dic['n_r']
n_r_outer = cells_num_dic['n_r_outer']
n_h = cells_num_dic['n_h']
# Effect of Temperature on density of fuel
fuel_list = []
density_mat = calUMoDensity(temp_phy_mat)
# Fuel U-10Mo
for j in range(n_h):
for i in range(n_r):
fuel_name = str((j+1)*10000+(i + 1)*100)
fuel = openmc.Material(name='U-10Mo '+ fuel_name)
fuel.set_density('g/cm3',density_mat[j,i])
fuel.add_element('Mo',0.1,'wo')
# fuel.add_nuclide('U235',0.1773,'wo') # for LEU (19.7%)
# fuel.add_nuclide('U238',0.7227,'wo')
fuel.add_nuclide('U235',0.837,'wo') # for HEU (93.0%)
fuel.add_nuclide('U238',0.063,'wo')
fuel_list.append(fuel)
materials_file.append(fuel)
for i in range(n_r_outer):
fuel_name = str((j+1)*10000+(i + n_r + 1)*100)
fuel = openmc.Material(name='U-10Mo '+ fuel_name)
fuel.set_density('g/cm3',density_mat[j,i+n_r])
fuel.add_element('Mo',0.1,'wo')
# fuel.add_nuclide('U235',0.1773,'wo') # for LEU (19.7%)
# fuel.add_nuclide('U238',0.7227,'wo')
fuel.add_nuclide('U235',0.837,'wo') # for HEU (93.0%)
fuel.add_nuclide('U238',0.063,'wo')
fuel_list.append(fuel)
materials_file.append(fuel)
# Export to "materials.xml"
materials_file.export_to_xml()
num_heat_pipe = 8
# Parameters of reactor
# Unit:cm
fuel_r = parameters_dic['fuel_r']
fuel_h = parameters_dic['fuel_h']
controlRod_r = parameters_dic['controlRod_r']
controlRod_h_max = parameters_dic['controlRod_h_max']
controlRod_l = parameters_dic['controlRod_l']
reflector_r = parameters_dic['reflector_r']
reflector_h = parameters_dic['reflector_h']
heat_pipe_R = parameters_dic['heat_pipe_R']
heat_pipe_r = parameters_dic['heat_pipe_r']
top_distance = parameters_dic['top_distance']
bottom_distance = parameters_dic['bottom_distance']
# Create cylinders for the fuel control rod and reflector
fuel_OD = openmc.ZCylinder(x0=0.0, y0=0.0, r=fuel_r)
controlRod_OD = openmc.ZCylinder(x0=0.0, y0=0.0, r=controlRod_r)
reflector_OD = openmc.ZCylinder(x0=0.0, y0=0.0, r=reflector_r, boundary_type='vacuum')
# Create planes for fuel control rod and reflector
reflector_TOP = openmc.ZPlane(z0 = (top_distance+fuel_h/2),boundary_type='vacuum')
reflector_BOTTOM = openmc.ZPlane(z0 = -(bottom_distance+fuel_h/2),boundary_type='vacuum')
reflector_empty_TOP = openmc.ZPlane(z0 = -(bottom_distance+fuel_h/2-empty_reflector_height))
fuel_TOP = openmc.ZPlane(z0 = fuel_h/2)
fuel_BOTTOM = openmc.ZPlane(z0 = -fuel_h/2)
controlRodSpace_TOP = openmc.ZPlane(z0 = (controlRod_h_max-bottom_distance-fuel_h/2))
# Create cylinders for heat pipes
heat_pipe_OD = openmc.ZCylinder(x0=fuel_r, y0=0, r=heat_pipe_R)
n_ang = num_heat_pipe
ang_mesh = np.pi/n_ang
r_mesh = np.linspace(controlRod_r,(fuel_r-heat_pipe_R),n_r+1)
r_outer_mesh = np.linspace(fuel_r-heat_pipe_R,fuel_r,n_r_outer+1)
h_mesh = np.linspace(-fuel_h/2,fuel_h/2,n_h+1)
line_1 = openmc.Plane(a=np.tan(-ang_mesh),b=-1.0,c=0.0,d=0.0,boundary_type='reflective')
line_2 = openmc.Plane(a=np.tan(ang_mesh),b=-1.0,c=0.0,d=0.0,boundary_type='reflective')
# Create volume vector and matrix
volume_vec = np.zeros(n_r+n_r_outer)
d_h = fuel_h/n_h
for i in range(n_r+n_r_outer):
if i >= n_r:
d = heat_pipe_R*(i-n_r)/n_r_outer
x_i = np.sqrt(2*heat_pipe_R*d-d*d)
d = heat_pipe_R*(i-n_r+1)/n_r_outer
x_i1 = np.sqrt(2*heat_pipe_R*d-d*d)
s = (x_i+x_i1)*heat_pipe_R/n_r_outer
volume_vec[i] = d_h*np.pi*(r_outer_mesh[i+1-n_r]*r_outer_mesh[i+1-n_r]-r_outer_mesh[i-n_r]*r_outer_mesh[i-n_r])/8-s
else:
volume_vec[i] = d_h*np.pi*(r_mesh[i+1]*r_mesh[i+1]-r_mesh[i]*r_mesh[i])/8
volume_mat = np.zeros((n_h,(n_r+n_r_outer)))
for i in range(n_h):
volume_mat[i,:] = volume_vec
# Create heat_pipe universe
heat_pipe_Inner = openmc.ZCylinder(r=heat_pipe_r)
coolant_cell = openmc.Cell(fill=Coolant_Na, region=(-heat_pipe_Inner & -reflector_TOP & +reflector_BOTTOM))
pipe_cell = openmc.Cell(fill=structure_HAY, region=(+heat_pipe_Inner & -reflector_TOP & +reflector_BOTTOM))
heat_pipe_universe = openmc.Universe(cells=(coolant_cell, pipe_cell))
# Create a Universe to encapsulate a fuel pin
pin_cell_universe = openmc.Universe(name='U-10Mo Pin')
# Create fine-fuel-cell (num of cells: n_r + n_r_outer)
fuel_cell_list = []
fuel_cell_ID_list = []
k = 0
for j in range(n_h):
cir_top = openmc.ZPlane(z0 = h_mesh[j+1])
cir_bottom = openmc.ZPlane(z0 = h_mesh[j])
for i in range(n_r):
cir_in = openmc.ZCylinder(r=r_mesh[i])
cir_out = openmc.ZCylinder(r=r_mesh[i+1])
fuel_cell = openmc.Cell()
fuel_cell.fill = fuel_list[k]
k = k+1
fuel_cell.region = +cir_in & -cir_out & +cir_bottom &-cir_top
fuel_cell.temperature = temp_phy_mat[j,i]
fuel_cell.id = (j+1)*10000+(i + 1)*100
fuel_cell_ID_list.append((j+1)*10000+(i + 1)*100)
fuel_cell_list.append(fuel_cell)
pin_cell_universe.add_cell(fuel_cell)
for i in range(n_r_outer):
cir_in = openmc.ZCylinder(r=r_outer_mesh[i])
cir_out = openmc.ZCylinder(r=r_outer_mesh[i+1])
fuel_cell = openmc.Cell()
fuel_cell.fill = fuel_list[k]
k = k+1
fuel_cell.region = +cir_in & -cir_out & +heat_pipe_OD & +cir_bottom &-cir_top
fuel_cell.temperature = temp_phy_mat[j,i+n_r]
fuel_cell.id = (j+1)*10000+(n_r + i + 1)*100
fuel_cell_ID_list.append((j+1)*10000+(n_r + i + 1)*100)
fuel_cell_list.append(fuel_cell)
pin_cell_universe.add_cell(fuel_cell)
# Create control rod Cell
controlRod_TOP = openmc.ZPlane(z0 = (controlRod_deep-fuel_h/2-bottom_distance))
controlRod_TOP.name = 'controlRod_TOP'
if controlRod_deep < controlRod_h_max:
controlRod_empty_cell = openmc.Cell(name='Control Rod Empty')
controlRod_empty_cell.region = -controlRod_OD & +controlRod_TOP & -controlRodSpace_TOP
pin_cell_universe.add_cell(controlRod_empty_cell)
if controlRod_deep > 0:
controlRod_cell = openmc.Cell(name='Control Rod')
controlRod_cell.fill = ControlRod_B4C
controlRod_cell.region = -controlRod_OD & +reflector_BOTTOM & -controlRod_TOP
controlRod_cell.tempearture = temp_defualt
pin_cell_universe.add_cell(controlRod_cell)
# Create heat pipe Cell
heat_pipe_cell = openmc.Cell(name='Heat Pipe')
heat_pipe_cell.fill = heat_pipe_universe
heat_pipe_cell.region = -heat_pipe_OD & +reflector_BOTTOM & -reflector_TOP
heat_pipe_cell.temperature = temp_defualt
heat_pipe_cell.translation = (fuel_r,0,0)
pin_cell_universe.add_cell(heat_pipe_cell)
# To be edited
# Create reflector Cell
if empty_reflector_height >0:
reflector_radial_empty_cell = openmc.Cell(name='Radial Reflector Empty')
reflector_radial_empty_cell.region = +fuel_OD & +heat_pipe_OD & +reflector_BOTTOM & -reflector_empty_TOP
pin_cell_universe.add_cell(reflector_radial_empty_cell)
reflector_radial_cell = openmc.Cell(name='Radial Reflector')
reflector_radial_cell.fill = Reflector_BeO
reflector_radial_cell.region = +fuel_OD & +heat_pipe_OD & +reflector_empty_TOP & -reflector_TOP
reflector_radial_cell.temperature = temp_defualt
pin_cell_universe.add_cell(reflector_radial_cell)
else:
reflector_radial_cell = openmc.Cell(name='Radial Reflector')
reflector_radial_cell.fill = Reflector_BeO
reflector_radial_cell.region = +fuel_OD & +heat_pipe_OD & +reflector_BOTTOM & -reflector_TOP
reflector_radial_cell.temperature = temp_defualt
pin_cell_universe.add_cell(reflector_radial_cell)
reflector_bottom_cell = openmc.Cell(name='Bottom Reflector')
reflector_bottom_cell.fill = Reflector_BeO
reflector_bottom_cell.region = +controlRod_OD & +heat_pipe_OD & -fuel_OD & -fuel_BOTTOM & +reflector_BOTTOM
reflector_bottom_cell.temperature = temp_defualt
pin_cell_universe.add_cell(reflector_bottom_cell)
reflector_top_cell = openmc.Cell(name='Top Reflector')
reflector_top_cell.fill = Reflector_BeO
reflector_top_cell.region = +heat_pipe_OD & -fuel_OD & +controlRodSpace_TOP & -reflector_TOP
reflector_top_cell.region = reflector_top_cell.region | (+controlRod_OD & +heat_pipe_OD & -fuel_OD & +fuel_TOP & -controlRodSpace_TOP)
reflector_top_cell.temperature = temp_defualt
pin_cell_universe.add_cell(reflector_top_cell)
# Create root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = pin_cell_universe
# Add boundary planes
root_cell.region = -reflector_OD & +line_2 & -line_1 & +reflector_BOTTOM & -reflector_TOP
# Create root Universe
root_universe = openmc.Universe(universe_id=0, name='root universe')
root_universe.add_cell(root_cell)
# Create Geometry and set root Universe
geometry = openmc.Geometry(root_universe)
# Export to "geometry.xml"
geometry.export_to_xml()
# Instantiate a Settings object
settings_file = openmc.Settings()
settings_file.batches = settings_dic['batches']
settings_file.inactive = settings_dic['inactive']
settings_file.particles = settings_dic['particles']
settings_file.temperature['multipole']= True
settings_file.temperature['method']= 'interpolation'
settings_file.source = openmc.Source(space=openmc.stats.Point((15,0,0)))
# Export to "settings.xml"
settings_file.export_to_xml()
# Instantiate an empty Tallies object
tallies_file = openmc.Tallies()
for i in range(len(fuel_cell_ID_list)):
tally = openmc.Tally(name='cell tally '+str(fuel_cell_ID_list[i]))
tally.filters = [openmc.DistribcellFilter(fuel_cell_ID_list[i])]
tally.scores = ['heating','flux']
tallies_file.append(tally)
# Export to "tallies.xml"
tallies_file.export_to_xml()
return volume_mat,fuel_cell_ID_list
def _build_inputs(self):
####################
# Materials
####################
moderator = openmc.Material(material_id=1)
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide('H1', 2.0)
moderator.add_nuclide('O16', 1.0)
dense_fuel = openmc.Material(material_id=2)
dense_fuel.set_density('g/cc', 4.5)
dense_fuel.add_nuclide('U235', 1.0)
light_fuel = openmc.Material(material_id=3)
light_fuel.set_density('g/cc', 2.0)
light_fuel.add_nuclide('U235', 1.0)
mats_file = openmc.Materials([moderator, dense_fuel, light_fuel])
mats_file.export_to_xml()
####################
# Geometry
####################
c1 = openmc.Cell(cell_id=1, fill=moderator)
mod_univ = openmc.Universe(universe_id=1, cells=[c1])
r0 = openmc.ZCylinder(r=0.3)
c11 = openmc.Cell(cell_id=11, region=-r0)
c11.fill = [dense_fuel, None, light_fuel, dense_fuel]
c12 = openmc.Cell(cell_id=12, region=+r0, fill=moderator)
fuel_univ = openmc.Universe(universe_id=11, cells=[c11, c12])
lat = openmc.RectLattice(lattice_id=101)
lat.lower_left = [-2.0, -2.0]
lat.pitch = [2.0, 2.0]
lat.universes = [[fuel_univ]*2]*2
lat.outer = mod_univ
x0 = openmc.XPlane(x0=-3.0)
x1 = openmc.XPlane(x0=3.0)
y0 = openmc.YPlane(y0=-3.0)
y1 = openmc.YPlane(y0=3.0)
for s in [x0, x1, y0, y1]:
s.boundary_type = 'reflective'
c101 = openmc.Cell(cell_id=101, fill=lat)
c101.region = +x0 & -x1 & +y0 & -y1
root_univ = openmc.Universe(universe_id=0, cells=[c101])
geometry = openmc.Geometry(root_univ)
geometry.export_to_xml()
####################
# Settings
####################
sets_file = openmc.Settings()
sets_file.batches = 5
sets_file.inactive = 0
sets_file.particles = 1000
sets_file.source = openmc.Source(space=openmc.stats.Box(
[-1, -1, -1], [1, 1, 1]))
sets_file.export_to_xml()
####################
# Plots
####################
plot1 = openmc.Plot(plot_id=1)
plot1.basis = 'xy'
plot1.color_by = 'cell'
plot1.filename = 'cellplot'
plot1.origin = (0, 0, 0)
plot1.width = (7, 7)
plot1.pixels = (400, 400)
plot2 = openmc.Plot(plot_id=2)
plot2.basis = 'xy'
plot2.color_by = 'material'
plot2.filename = 'matplot'
plot2.origin = (0, 0, 0)
plot2.width = (7, 7)
plot2.pixels = (400, 400)
plots = openmc.Plots([plot1, plot2])
plots.export_to_xml()
def make_geometry_tallies(batches, nps, inner_radius, thickness):
first_wall_inner_surface = openmc.Sphere(r=inner_radius)
first_wall_outer_surface = openmc.Sphere(r=inner_radius + thickness,
boundary_type='vacuum')
first_wall = +first_wall_inner_surface & -first_wall_outer_surface
first_wall = openmc.Cell(region=first_wall)
first_wall.fill = eurofer
inner_vac_cell = -first_wall_inner_surface
inner_vac_cell = openmc.Cell(region=inner_vac_cell)
universe = openmc.Universe(cells=[first_wall, inner_vac_cell])
geom = openmc.Geometry(universe)
geom.export_to_xml('geometry')
vox_plot = openmc.Plot()
vox_plot.type = 'voxel'
vox_plot.width = (10, 10, 10)
vox_plot.pixels = (200, 200, 200)
vox_plot.filename = 'plot_3d'
vox_plot.color_by = 'material'
vox_plot.colors = {eurofer: 'blue'}
plots = openmc.Plots([vox_plot])
plots.export_to_xml()
openmc.plot_geometry()
os.system('openmc-voxel-to-vtk plot_3d.h5 -o plot_3d.vti')
os.system('paraview plot_3d.vti')
sett = openmc.Settings()
sett.batches = batches
sett.inactive = 0
sett.particles = nps
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14.08e6], [1])
#source.energy = openmc.stats.Muir(e0=14080000.0, m_rat=5.0, kt=20000.0)
sett.source = source
sett.export_to_xml('settings.xml')
#tallies
particle_filter = openmc.ParticleFilter([1])
surface_filter_front = openmc.SurfaceFilter(first_wall_inner_surface)
surface_filter_rear = openmc.SurfaceFilter(first_wall_outer_surface)
bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
#think will need to change this
energy_filter = openmc.EnergyFilter(bins)
tallies = openmc.Tallies()
tally = openmc.Tally(name='wall_leakage')
tally.filters = [surface_filter_rear, particle_filter]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='incident_neutron_spectrum')
tally.filters = [surface_filter_rear, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='leakage_neutron_spectrum')
tally.filters = [surface_filter_front, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
tallies_to_retrieve = ['wall_leakage']
#at the moment, we only have one tally to retrieve, but we will set it up in a list anyway so we know how to do it
for tally_name in tallies_to_retrieve:
tally = sp.get_tally(name=tally_name)
df = tally.get_pandas_dataframe()
#defining something that stands for dataframe, need to investigate this
#its basically something that we use to obtain the mean value and the std deviation value of the tally
tally_result = df['mean'].sum()
tally_std_dev = df['std. dev.'].sum()
#for now, we will just try to print these values
#get spectra
spectra_tallies_to_retrieve = [
'incident_neutron_spectrum', 'leakage_neutron_spectrum'
]
for spectra_name in spectra_tallies_to_retrieve:
spectra_tally = sp.get_tally(name=spectra_name)
spectra_tally_result = [entry[0][0] for entry in spectra_tally.mean]
spectra_tally_std_dev = [
entry[0][0] for entry in spectra_tally.std_dev
]
def slab_mg(num_regions=1, mat_names=None, mgxslib_name='2g.h5'):
"""Create a 1D slab model.
Parameters
----------
num_regions : int, optional
Number of regions in the problem, each with a unique MGXS dataset.
Defaults to 1.
mat_names : Iterable of str, optional
List of the material names to use; defaults to ['mat_1', 'mat_2',...].
mgxslib_name : str, optional
MGXS Library file to use; defaults to '2g.h5'.
Returns
-------
model : openmc.model.Model
One-group, 1D slab model
"""
openmc.check_type('num_regions', num_regions, Integral)
openmc.check_greater_than('num_regions', num_regions, 0)
if mat_names is not None:
openmc.check_length('mat_names', mat_names, num_regions)
openmc.check_iterable_type('mat_names', mat_names, str)
else:
mat_names = []
for i in range(num_regions):
mat_names.append('mat_' + str(i + 1))
# # Make Materials
materials_file = openmc.Materials()
macros = []
mats = []
for i in range(len(mat_names)):
macros.append(openmc.Macroscopic('mat_' + str(i + 1)))
mats.append(openmc.Material(name=mat_names[i]))
mats[-1].set_density('macro', 1.0)
mats[-1].add_macroscopic(macros[-1])
materials_file += mats
materials_file.cross_sections = mgxslib_name
# # Make Geometry
rad_outer = 929.45
# Set a cell boundary to exist for every material above (exclude the 0)
rads = np.linspace(0., rad_outer, len(mats) + 1, endpoint=True)[1:]
# Instantiate Universe
root = openmc.Universe(universe_id=0, name='root universe')
cells = []
surfs = []
surfs.append(openmc.XPlane(x0=0., boundary_type='reflective'))
for r, rad in enumerate(rads):
if r == len(rads) - 1:
surfs.append(openmc.XPlane(x0=rad, boundary_type='vacuum'))
else:
surfs.append(openmc.XPlane(x0=rad))
# Instantiate Cells
cells = []
for c in range(len(surfs) - 1):
cells.append(openmc.Cell())
cells[-1].region = (+surfs[c] & -surfs[c + 1])
cells[-1].fill = mats[c]
# Register Cells with Universe
root.add_cells(cells)
# Instantiate a Geometry, register the root Universe, and export to XML
geometry_file = openmc.Geometry(root)
# # Make Settings
# Instantiate a Settings object, set all runtime parameters
settings_file = openmc.Settings()
settings_file.energy_mode = 'multi-group'
settings_file.tabular_legendre = {'enable': False}
settings_file.batches = 10
settings_file.inactive = 5
settings_file.particles = 1000
# Build source distribution
INF = 1000.
bounds = [0., -INF, -INF, rads[0], INF, INF]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.output = {'summary': False}
model = openmc.model.Model()
model.geometry = geometry_file
model.materials = materials_file
model.settings = settings_file
model.xs_data = macros
return model
def _build_openmc(self):
"""Generate the OpenMC input XML
"""
# Directory from which openmc is run
os.makedirs('openmc', exist_ok=True)
# Define material
mat = openmc.Material()
for nuclide, fraction in self.nuclides:
mat.add_nuclide(nuclide, fraction)
mat.set_density('g/cm3', self.density)
materials = openmc.Materials([mat])
materials.export_to_xml(os.path.join('openmc', 'materials.xml'))
# Instantiate surfaces
cyl = openmc.XCylinder(boundary_type='vacuum', R=1.e-6)
px1 = openmc.XPlane(boundary_type='vacuum', x0=-1.)
px2 = openmc.XPlane(boundary_type='transmission', x0=1.)
px3 = openmc.XPlane(boundary_type='vacuum', x0=1.e9)
# Instantiate cells
inner_cyl_left = openmc.Cell()
inner_cyl_right = openmc.Cell()
outer_cyl = openmc.Cell()
# Set cells regions and materials
inner_cyl_left.region = -cyl & +px1 & -px2
inner_cyl_right.region = -cyl & +px2 & -px3
outer_cyl.region = ~(-cyl & +px1 & -px3)
inner_cyl_right.fill = mat
# Create root universe and export to XML
geometry = openmc.Geometry(
[inner_cyl_left, inner_cyl_right, outer_cyl])
geometry.export_to_xml(os.path.join('openmc', 'geometry.xml'))
# Define source
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Monodirectional()
source.energy = openmc.stats.Discrete([self.energy], [1.])
source.particle = 'neutron'
# Settings
settings = openmc.Settings()
settings.source = source
settings.particles = self.particles
settings.run_mode = 'fixed source'
settings.batches = 1
settings.photon_transport = True
settings.electron_treatment = self.electron_treatment
settings.cutoff = {'energy_photon': 1000.}
settings.export_to_xml(os.path.join('openmc', 'settings.xml'))
# Define filters
surface_filter = openmc.SurfaceFilter(cyl)
particle_filter = openmc.ParticleFilter('photon')
energy_bins = np.logspace(3, np.log10(2 * self.energy), 500)
energy_filter = openmc.EnergyFilter(energy_bins)
# Create tallies and export to XML
tally = openmc.Tally(name='photon current')
tally.filters = [surface_filter, energy_filter, particle_filter]
tally.scores = ['current']
tallies = openmc.Tallies([tally])
tallies.export_to_xml(os.path.join('openmc', 'tallies.xml'))
def make_model_mg_by_nuclide(nameoflib):
nuclist, conc, temp, rodnuclist, concrod, temprod, dzz, hss, rods, type_fuel, type_assembly = make_datas()
nkas = len(rods)
diction_mg = translate_mg(nkas, rods, dzz, [sum(dzz)], nameoflib,
nuclist, conc, rodnuclist, concrod)
outuniv, outsurface = _make_outer_mg_universe(diction_mg[(1,1)])
assemblies = {}
load_dict = {}
hexsurf, rodfaces = make_surfaces([sum(dzz)], HPITCH)
hexsurf, zfaces = make_surfaces(dzz, HPITCH)
for i in range(NASS):
if (rods[i]):
assemblies[i+1] = make_abl_mg_cells(diction_mg, hexsurf, rodfaces, i+1, temprod)
load_dict[i+1] = [i+1]
else:
assemblies[i+1] = make_abl_mg_cells(diction_mg, hexsurf, zfaces, i+1, temp)
load_dict[i+1] = [i+1]
print("Making load ..")
##################################################################################
load = make_load("BN-800", outuniv, assemblies, NRING, HPITCH, sum(dzz), load_dict)
root_cell = openmc.Cell(cell_id=0, region=-outsurface, fill=load)
root_univ = openmc.Universe(universe_id=0, cells = [root_cell])
# Instantiate a Geometry, register the root Universe, and export to XML
openmc_geometry = openmc.Geometry()
openmc_geometry.root_universe = root_univ
openmc_geometry.export_to_xml()
materials_list = openmc_geometry.get_all_materials()
materials_file = openmc.Materials([v for v in materials_list.values()])
materials_file.cross_sections = nameoflib
materials_file.export_to_xml()
# Instantiate a Settings object, set all runtime parameters, and export to XML
settings_file = openmc.Settings()
##########################SETTINGS
settings_file.batches = 120
settings_file.inactive = 20
settings_file.particles = 1000
###########################
settings_file.temperature = {'range':(250,2500)}
# Tell OpenMC this is a multi-group problem
settings_file.energy_mode = 'multi-group'
#settings_file.survival_biasing = True
#settings_file.cutoff = {'energy_neutron' : 1.0,'weight':1.e-10}
settings_file.source = openmc.Source(space=openmc.stats.Box(
[-160.04, -160.04, 0.0], [160.04, 160.04, 170], only_fissionable=True))
settings_file.export_to_xml()
plot_1 = openmc.Plot(plot_id=1)
plot_1.filename = 'plot_1'
plot_1.origin = [0.0, 0.0, 8.]
plot_1.width = [520.26, 520.26]
plot_1.pixels = [2400, 2400]
plot_1.color = 'mat'
plot_1.basis = 'xy'
plot_2 = openmc.Plot(plot_id=2)
plot_2.filename = 'plot_2'
plot_2.origin = [0.0, 0.0, 0.0]
plot_2.width = [520.26, 350.2]
plot_2.pixels = [1200, 1200]
plot_2.color = 'mat'
plot_2.basis = 'xz'
def initial(self, opt, Mesh): #, nthreads):
#####################################
# Create OpenMC geometry
#####################################
FuelOR = opt.FuelOR * 100 #[cm]
CladIR = opt.CladIR * 100 #[cm]
CladOR = opt.CladOR * 100 #[cm]
# ExtraCladOR = opt.ExtraCladOR*100 #[cm]
Pitch = opt.PinPitch * 100 #[cm]
# Construct uniform initial source distribution over fissionable zones
lower_left = [-Pitch / 2, -Pitch / 2, 0]
upper_right = [+Pitch / 2, +Pitch / 2, +365.76]
uniform_dist = openmc.stats.Box(lower_left,
upper_right,
only_fissionable=True)
# Settings file
self.settings_file = openmc.Settings()
self.settings_file.batches = 1100
self.settings_file.inactive = 100
self.settings_file.particles = 20000
self.settings_file.generations_per_batch = 5
self.settings_file.output = {'tallies': False}
self.settings_file.temperature = {'multipole': True, 'tolerance': 3000}
self.settings_file.source = openmc.source.Source(space=uniform_dist)
self.settings_file.seed = np.random.randint(1, 100) # for correlation
self.settings_file.export_to_xml() #Removed for correlation
# Materials file
uo2 = openmc.Material(material_id=1,
name='UO2 fuel at 2.4% wt enrichment')
uo2.temperature = 900. #opt.Tin
uo2.set_density('g/cm3', 10.29769)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
# helium = openmc.Material(material_id=2, name='Helium for gap')
# helium.temperature = opt.Tin
# helium.set_density('g/cm3', 0.001598)
# helium.add_element('He', 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.temperature = opt.Tin
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_element('Sn', 0.014, 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001, 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
# borated_water = openmc.Material()
# borated_water.temperature = opt.Tin
# borated_water.set_density('g/cm3', 0.7406)
# 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 = openmc.model.borated_water(boron_ppm=432.473,
temperature=opt.Tin,
pressure=15)
self.materials_file = openmc.Materials([uo2, zircaloy,
borated_water]) #helium
self.materials_file.export_to_xml() #Removed for correlation
# Geometry file
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=FuelOR)
# clad_ir = openmc.ZCylinder(x0=0, y0=0, R=CladIR)
clad_or = openmc.ZCylinder(x0=0, y0=0, R=CladOR)
# extra_clad_or = openmc.ZCylinder(x0=0, y0=0, R=ExtraCladOR)
left = openmc.XPlane(x0=-Pitch / 2)
right = openmc.XPlane(x0=Pitch / 2)
back = openmc.YPlane(y0=-Pitch / 2)
front = openmc.YPlane(y0=Pitch / 2)
top = openmc.ZPlane(z0=396.24)
bottom = openmc.ZPlane(z0=-30.48)
z_list = []
for i in range(0, len(Mesh)):
z_list.append(openmc.ZPlane(z0=Mesh[i]))
left.boundary_type = 'reflective'
right.boundary_type = 'reflective'
front.boundary_type = 'reflective'
back.boundary_type = 'reflective'
top.boundary_type = 'vacuum'
bottom.boundary_type = 'vacuum'
# z_list[-1].boundary_type = 'vacuum'
# z_list[0].boundary_type = 'vacuum'
self.reflectTOP = openmc.Cell()
self.reflectBOT = openmc.Cell()
self.fuel_list = []
# self.gap_list = []
self.clad_list = []
# self.extra_clad_list = []
self.water_list = []
for i in range(0, len(Mesh) - 1):
self.fuel_list.append(openmc.Cell())
# self.gap_list.append(openmc.Cell())
self.clad_list.append(openmc.Cell())
# self.extra_clad_list.append(openmc.Cell())
self.water_list.append(openmc.Cell())
self.reflectTOP.region = +left & -right & +back & -front & +z_list[
-1] & -top
self.reflectBOT.region = +left & -right & +back & -front & +bottom & -z_list[
0]
self.reflectTOP.fill = borated_water
self.reflectBOT.fill = borated_water
j = 0
for fuels in self.fuel_list:
fuels.region = -fuel_or & +z_list[j] & -z_list[j + 1]
fuels.fill = uo2
j = j + 1
# j = 0
# for gaps in self.gap_list:
# gaps.region = +fuel_or & -clad_ir & +z_list[j] & -z_list[j+1]
# gaps.fill = helium
# j = j+1
j = 0
for clads in self.clad_list:
clads.region = +fuel_or & -clad_or & +z_list[j] & -z_list[
j + 1] #clad_ir instead of fuel_or
clads.fill = zircaloy
j = j + 1
# j = 0
# for extra_clads in self.extra_clad_list:
# extra_clads.region = +clad_or & -extra_clad_or & +left & -right & +back & -front & +z_list[j] & -z_list[j+1]
# grid_flag = 0
# for i in range(0,len(opt.GridBot_z)):
# if z_list[j].z0 == opt.GridBot_z[i]:
# extra_clads.fill = zircaloy
# grid_flag = 1
# if grid_flag == 1:
# extra_clads.fill = zircaloy
# else:
# extra_clads.fill = borated_water
# extra_clads.fill = borated_water
# j = j+1
j = 0
for waters in self.water_list:
waters.region = +clad_or & +left & -right & +back & -front & +z_list[
j] & -z_list[j + 1]
waters.fill = borated_water
j = j + 1
self.root = openmc.Universe(universe_id=0, name='root universe')
self.root.add_cells(self.fuel_list)
# self.root.add_cells(self.gap_list)
self.root.add_cells(self.clad_list)
# self.root.add_cells(self.extra_clad_list)
self.root.add_cells(self.water_list)
self.root.add_cells([self.reflectTOP, self.reflectBOT])
self.geometry_file = openmc.Geometry(self.root)
self.geometry_file.export_to_xml() #Removed for correlation
# Tallies
# power distribution: fission q recoverable (Sterling's note: starts 0, might be data pb)
# openmc accounts for incoming neutron energy and isotope
cell_filter = openmc.CellFilter(self.fuel_list)
t = openmc.Tally(tally_id=1)
t.filters.append(cell_filter)
t.scores = ['fission-q-recoverable']
tallies = openmc.Tallies([t])
tallies.export_to_xml() #Removed for correlation
# Plots
plot = openmc.Plot()
plot.width = [Pitch + 45, Pitch + 45]
plot.origin = [0., 0., -20]
plot.color_by = 'material'
plot.filename = 'fuel-pin'
plot.pixels = [1000, 1000]
plot.basis = 'yz'
# openmc.plot_inline(plot)
# Move Files to PinGeo folder #Removed for correlation
shutil.move('settings.xml', 'PinGeo/settings.xml')
shutil.move('materials.xml', 'PinGeo/materials.xml')
shutil.move('geometry.xml', 'PinGeo/geometry.xml')
shutil.move('tallies.xml', 'PinGeo/tallies.xml')
# shutil.move('plots.xml', 'PinGeo/plots.xml')
openmc.run(
cwd='PinGeo') #, threads=nthreads, mpi_args=['mpiexec','-n','2'])
sp = openmc.StatePoint('PinGeo/statepoint.' +
str(self.settings_file.batches) + '.h5')
tally = sp.get_tally(scores=['fission-q-recoverable'])
self.Tally = np.ndarray.flatten(tally.sum)
Pfactor = 66945.4 / sum(np.ndarray.flatten(tally.sum))
# print("Pfactor: ", Pfactor)
self.Tally = np.ndarray.flatten(tally.sum) * Pfactor
self.Var = np.divide(np.ndarray.flatten(tally.std_dev),
np.ndarray.flatten(tally.mean))
def _build_inputs(self):
# Define TRISO matrials
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.5)
fuel.add_nuclide('U235', 0.14154)
fuel.add_nuclide('U238', 0.85846)
fuel.add_nuclide('C0', 0.5)
fuel.add_nuclide('O16', 1.5)
porous_carbon = openmc.Material()
porous_carbon.set_density('g/cm3', 1.0)
porous_carbon.add_nuclide('C0', 1.0)
porous_carbon.add_s_alpha_beta('c_Graphite')
ipyc = openmc.Material()
ipyc.set_density('g/cm3', 1.90)
ipyc.add_nuclide('C0', 1.0)
ipyc.add_s_alpha_beta('c_Graphite')
sic = openmc.Material()
sic.set_density('g/cm3', 3.20)
sic.add_element('Si', 1.0)
sic.add_nuclide('C0', 1.0)
opyc = openmc.Material()
opyc.set_density('g/cm3', 1.87)
opyc.add_nuclide('C0', 1.0)
opyc.add_s_alpha_beta('c_Graphite')
graphite = openmc.Material()
graphite.set_density('g/cm3', 1.1995)
graphite.add_nuclide('C0', 1.0)
graphite.add_s_alpha_beta('c_Graphite')
# Create TRISO particles
spheres = [
openmc.Sphere(R=r * 1e-4) for r in [212.5, 312.5, 347.5, 382.5]
]
c1 = openmc.Cell(fill=fuel, region=-spheres[0])
c2 = openmc.Cell(fill=porous_carbon, region=+spheres[0] & -spheres[1])
c3 = openmc.Cell(fill=ipyc, region=+spheres[1] & -spheres[2])
c4 = openmc.Cell(fill=sic, region=+spheres[2] & -spheres[3])
c5 = openmc.Cell(fill=opyc, region=+spheres[3])
inner_univ = openmc.Universe(cells=[c1, c2, c3, c4, c5])
outer_radius = 422.5 * 1e-4
trisos = openmc.model.pack_trisos(radius=outer_radius,
fill=inner_univ,
domain_shape='cube',
domain_length=1.,
domain_center=(0., 0., 0.),
n_particles=100)
# Define box to contain lattice
min_x = openmc.XPlane(x0=-0.5, boundary_type='reflective')
max_x = openmc.XPlane(x0=0.5, boundary_type='reflective')
min_y = openmc.YPlane(y0=-0.5, boundary_type='reflective')
max_y = openmc.YPlane(y0=0.5, boundary_type='reflective')
min_z = openmc.ZPlane(z0=-0.5, boundary_type='reflective')
max_z = openmc.ZPlane(z0=0.5, boundary_type='reflective')
box = openmc.Cell(region=+min_x & -max_x & +min_y & -max_y & +min_z
& -max_z)
# Create lattice
ll, ur = box.region.bounding_box
shape = (3, 3, 3)
pitch = (ur - ll) / shape
lattice = openmc.model.create_triso_lattice(trisos, ll, pitch, shape,
graphite)
box.fill = lattice
root = openmc.Universe(0, cells=[box])
geom = openmc.Geometry(root)
geom.export_to_xml()
settings = openmc.Settings()
settings.batches = 5
settings.inactive = 0
settings.particles = 100
settings.source = openmc.Source(space=openmc.stats.Point())
settings.export_to_xml()
mats = openmc.Materials(
[fuel, porous_carbon, ipyc, sic, opyc, graphite])
mats.export_to_xml()
def _make_openmc_input(self):
"""Generate the OpenMC input XML
"""
# Define material
mat = openmc.Material()
for nuclide, fraction in self.nuclides:
mat.add_nuclide(nuclide, fraction)
mat.set_density('g/cm3', self.density)
materials = openmc.Materials([mat])
if self.xsdir is not None:
xs_path = (self.openmc_dir / 'cross_sections.xml').resolve()
materials.cross_sections = str(xs_path)
materials.export_to_xml(self.openmc_dir / 'materials.xml')
# Instantiate surfaces
cyl = openmc.XCylinder(boundary_type='vacuum', r=1.e-6)
px1 = openmc.XPlane(boundary_type='vacuum', x0=-1.)
px2 = openmc.XPlane(boundary_type='transmission', x0=1.)
px3 = openmc.XPlane(boundary_type='vacuum', x0=1.e9)
# Instantiate cells
inner_cyl_left = openmc.Cell()
inner_cyl_right = openmc.Cell()
outer_cyl = openmc.Cell()
# Set cells regions and materials
inner_cyl_left.region = -cyl & +px1 & -px2
inner_cyl_right.region = -cyl & +px2 & -px3
outer_cyl.region = ~(-cyl & +px1 & -px3)
inner_cyl_right.fill = mat
# Create root universe and export to XML
geometry = openmc.Geometry(
[inner_cyl_left, inner_cyl_right, outer_cyl])
geometry.export_to_xml(self.openmc_dir / 'geometry.xml')
# Define source
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Monodirectional()
source.energy = openmc.stats.Discrete([self.energy], [1.])
source.particle = 'neutron'
# Settings
settings = openmc.Settings()
if self._temperature is not None:
settings.temperature = {'default': self._temperature}
settings.source = source
settings.particles = self.particles // self._batches
settings.run_mode = 'fixed source'
settings.batches = self._batches
settings.photon_transport = True
settings.electron_treatment = self.electron_treatment
settings.cutoff = {'energy_photon': self._cutoff_energy}
settings.export_to_xml(self.openmc_dir / 'settings.xml')
# Define filters
surface_filter = openmc.SurfaceFilter(cyl)
particle_filter = openmc.ParticleFilter('photon')
energy_bins = np.logspace(np.log10(self._cutoff_energy),
np.log10(self.max_energy), self._bins + 1)
energy_filter = openmc.EnergyFilter(energy_bins)
# Create tallies and export to XML
tally = openmc.Tally(name='tally')
tally.filters = [surface_filter, energy_filter, particle_filter]
tally.scores = ['current']
tallies = openmc.Tallies([tally])
tallies.export_to_xml(self.openmc_dir / 'tallies.xml')
def simulate_model(
enrichment,
thickness,
breeder_material_name="Li4SiO4",
temperature_in_C=500,
batches=10,
nps=1000,
inner_radius=500,
):
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment=enrichment,
temperature_in_C=temperature_in_C,
).neutronics_material
eurofer = Material(material_name="eurofer").neutronics_material
copper = Material(material_name="copper").neutronics_material
mats = openmc.Materials([breeder_material, eurofer, copper])
mats.export_to_xml("materials.xml")
# GEOMETRY#
central_sol_surface = openmc.ZCylinder(r=100)
central_shield_outer_surface = openmc.ZCylinder(r=110)
first_wall_inner_surface = openmc.Sphere(r=inner_radius)
first_wall_outer_surface = openmc.Sphere(r=inner_radius + 10)
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius + 10.0 +
thickness)
vessel_outer_surface = openmc.Sphere(r=inner_radius + 10.0 + thickness +
10.0,
boundary_type="vacuum")
central_sol_region = -central_sol_surface & -breeder_blanket_outer_surface
central_sol_cell = openmc.Cell(region=central_sol_region)
central_sol_cell.fill = copper
central_shield_region = (+central_sol_surface
& -central_shield_outer_surface
& -breeder_blanket_outer_surface)
central_shield_cell = openmc.Cell(region=central_shield_region)
central_shield_cell.fill = eurofer
inner_void_region = -first_wall_inner_surface & +central_shield_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = "inner_void"
first_wall_region = (-first_wall_outer_surface
& +first_wall_inner_surface
& +central_shield_outer_surface)
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = eurofer
breeder_blanket_region = (+first_wall_outer_surface
& -breeder_blanket_outer_surface
& +central_shield_outer_surface)
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
vessel_region = +breeder_blanket_outer_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = "vessel"
vessel_cell.fill = eurofer
universe = openmc.Universe(cells=[
central_sol_cell,
central_shield_cell,
inner_void_cell,
first_wall_cell,
breeder_blanket_cell,
vessel_cell,
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS#
sett = openmc.Settings()
sett.batches = batches
sett.inactive = 0
sett.particles = nps
sett.run_mode = "fixed source"
source = openmc.Source()
source.space = openmc.stats.Point((150, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14.08e6], [1])
sett.source = source
# sett.export_to_xml("settings.xml")
# tally filters
particle_filter = openmc.ParticleFilter("neutron")
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
# TALLIES#
tallies = openmc.Tallies()
tally = openmc.Tally(name="TBR")
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ["(n,Xt)"]
tallies.append(tally)
# RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
model.run(output=False)
# RETRIEVING TALLY RESULTS
sp = openmc.StatePoint("statepoint." + str(batches) + ".h5")
json_output = {
"batches": batches,
"nps": nps,
"enrichment": enrichment,
"inner_radius": inner_radius,
"thickness": thickness,
"breeder_material_name": breeder_material_name,
"temperature_in_C": temperature_in_C,
}
tally = sp.get_tally(name="TBR")
df = tally.get_pandas_dataframe()
json_output["TBR"] = df["mean"].sum()
json_output["TBR_std_dev"] = df["std. dev."].sum()
return json_output
def _pre_run(self, root_cell):
MC_input_path = self.MC_input_path
pre_run_path = os.getcwd() +'/pre_run'
try:
shutil.rmtree(pre_run_path)
except OSError:
pass
os.mkdir(pre_run_path)
# Prepare the volume calculation
#bounding_box = root_cell.bounding_box
ll = self.bounding_box[0]
ur = self.bounding_box[1]
cell_dict = root_cell.get_all_cells()
cell_list = utils.cell_dict_to_cell_list(cell_dict)
cell_list.append(root_cell) # Add root_cell so that the total volume is calculated
vol1 = openmc.VolumeCalculation(cell_list, 100000, lower_left = ll, upper_right = ur)
settings = openmc.Settings()
settings.volume_calculations = [vol1]
settings.temperature = {'method':'interpolation'}
settings.run_mode='volume'
settings.export_to_xml(path = pre_run_path + '/settings.xml')
# Copy the geometry and material file to the new dummy dir
shutil.copyfile(MC_input_path + '/geometry.xml', pre_run_path + '/geometry.xml')
shutil.copyfile(MC_input_path + '/materials.xml', pre_run_path + '/materials.xml')
# By default, the openm_exec is set to 'openmc'
# For some reasons, this does not work on the cluster (della)
# On della, we need to explicitly define the absolute path to the bin we want to use
# Right now a temporary path that depends on my installation is used
#openmc.calculate_volumes(cwd = pre_run_path, openmc_exec='/tigress/jdtdl/openmc/py3-mpi-190324/bin/openmc')
openmc.calculate_volumes(cwd = pre_run_path, openmc_exec=self.openmc_bin_path)
#openmc.run()
# Read and set initial nuclides dict
self.set_init_nucl_dict(root_cell)
# # Read each material object and add 1atm nuclides chosen by the user
# if self.mode == 'no_const_lib':
# self.add_zero_dens_nuclides(self.nucl_list_dict)
self._set_initial_summary(pre_run_path)
self._set_cross_sections_path(pre_run_path)
# Read cross sections xml files, create MC_XS_nucl_list
self.set_MC_XS_nucl_list()
self.set_root_universe()
root_cell_name = 'root cell' # need to be specified by the user at some point
self._set_root_cell(root_cell_name)
# Extract cells from summary, add 1 atm nuclides to their material
self._change_cell_materials()
# Read and distribute volumes to cells
self.set_vol_to_cell(vol1, pre_run_path)
# pdb.set_trace()
shutil.rmtree(pre_run_path)
def sphere_with_firstwall_model(
material_for_structure,
blanket_breeder_material,
blanket_coolant_material,
firstwall_coolant_material,
blanket_breeder_li6_enrichment, # this is a percentage
coolant_pressure, # this is in Pa
blanket_coolant_temperature_in_C,
firstwall_coolant_temperature_in_C,
blanket_breeder_fraction,
blanket_coolant_fraction,
blanket_structural_fraction,
blanket_breeder_temperature_in_C = None, #needed for liquid breeders like lithium lead
firstwall_thickness = 2.7, # this is in cm
blanket_thickness = 200, # this is in cm
inner_radius = 1000,
firstwall_armour_fraction = 0.106305, # equivilent to 3mm and based on https://doi.org/10.1016/j.fusengdes.2017.02.008
firstwall_coolant_fraction= 0.333507, # based on https://doi.org/10.1016/j.fusengdes.2017.02.008
firstwall_structural_fraction = 0.560188, # based on https://doi.org/10.1016/j.fusengdes.2017.02.008
blanket_multipler_material = None, #used for combined breeder multiplier options
blanket_multiplier_fraction = None, #used for combined breeder multiplier options
blanket_breeder_material_packing_fraction = None, #used for combined breeder multiplier options
blanket_multiplier_packing_fraction = None, #used for combined breeder multiplier options
blanket_multiplier_material = None #used for combined breeder multiplier options
):
breeder_percent_in_breeder_plus_multiplier_ratio = 100 * (blanket_breeder_fraction / (blanket_breeder_fraction + blanket_multiplier_fraction))
inputs = locals()
"""
This function builds materials for the homogenised blanket material, homogenised firstwall material
The creates a simple sphere geometry with a simple point source and TBR tally on the blanket
The function also carries out the simulation and writes the results to a JSON file
"""
# creates homogensied blanket material using a single breeder / multiplier material (e.g lithium lead)
if blanket_breeder_material == 'Pb842Li158':
blanket_material = MultiMaterial(material_tag = 'blanket_material',
materials = [
Material(material_name = material_for_structure),
Material(material_name = blanket_coolant_material,
temperature_in_C = blanket_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = blanket_breeder_material,
enrichment = blanket_breeder_li6_enrichment,
temperature_in_C = blanket_breeder_temperature_in_C),
],
fracs = [blanket_structural_fraction,
blanket_coolant_fraction,
blanket_breeder_fraction],
percent_type='vo'
).openmc_material
# creates homogensied blanket material using a combined breeder multiplier material (e.g lithium ceramic with be multiplier)
else:
blanket_material = MultiMaterial(
material_tag = 'blanket_material',
materials = [
Material(material_name = material_for_structure),
Material(material_name = blanket_coolant_material,
temperature_in_C = blanket_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = blanket_breeder_material,
enrichment = blanket_breeder_li6_enrichment,
packing_fraction = blanket_breeder_material_packing_fraction),
Material(material_name = blanket_multipler_material,
packing_fraction = blanket_multiplier_packing_fraction),
],
fracs = [blanket_structural_fraction,
blanket_coolant_fraction,
blanket_breeder_fraction,
blanket_multiplier_fraction],
percent_type='vo'
).openmc_material
# creates homogensied firstwall material with eurofer, tungsten and a coolant
firstwall_material = MultiMaterial(material_tag = 'firstwall_material',
materials = [
Material(material_name = 'tungsten'),
Material(material_name = firstwall_coolant_material,
temperature_in_C = firstwall_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = 'eurofer')],
fracs = [firstwall_armour_fraction,
firstwall_coolant_fraction,
firstwall_structural_fraction]
).openmc_material
mats = openmc.Materials([blanket_material, firstwall_material])
# creates surfaces
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius+firstwall_thickness)
firstwall_outer_surface = openmc.Sphere(r=inner_radius)
inner_void_region = -firstwall_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
firstwall_region = +firstwall_outer_surface & -breeder_blanket_inner_surface
firstwall_cell = openmc.Cell(name='firstwall', region=firstwall_region)
firstwall_cell.fill = firstwall_material
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius+firstwall_thickness+blanket_thickness, boundary_type='vacuum')
breeder_blanket_region = -breeder_blanket_outer_surface & +breeder_blanket_inner_surface
breeder_blanket_cell = openmc.Cell(name = 'breeder_blanket', region=breeder_blanket_region)
breeder_blanket_cell.fill = blanket_material
universe = openmc.Universe(cells=[inner_void_cell,
firstwall_cell,
breeder_blanket_cell])
geom = openmc.Geometry(universe)
# assigns simulation settings
sett = openmc.Settings()
sett.batches = 200 # this is minimum number of batches that will be run
sett.trigger_active = True
sett.trigger_max_batches = 1500 # this is maximum number of batches that will be run
sett.particles = 300
sett.verbosity = 1
sett.run_mode = 'fixed source'
# sets a 14MeV (distributuion) point source
source = openmc.Source()
source.space = openmc.stats.Point((0,0,0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(e0=14080000.0, m_rat=5.0, kt=20000.0) #neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
# this is the tally set up
tallies = openmc.Tallies()
# define filters
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
particle_filter = openmc.ParticleFilter(['neutron'])
# creates the TBR tally using the filters and sets a completion trigger
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ['(n,Xt)'] # where X is a wildcard, if MT 105 or (n,t) then some tritium production will be missed, for example (n,nt) which happens in Li7 would be missed
tally.triggers = [openmc.Trigger(trigger_type='rel_err', threshold=0.0001)] # This stops the simulation if the threshold is meet
tallies.append(tally)
# collects all the model parts and runs the model
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run(output=False)
# opens the output file and retrieves the tally results
sp = openmc.StatePoint(sp_filename)
tally = sp.get_tally(name='TBR')
df = tally.get_pandas_dataframe()
tally_result = df['mean'].sum()
tally_std_dev = df['std. dev.'].sum()
# combines the tally results with the input data
inputs.update({'tbr': tally_result})
inputs.update({'tbr_error': tally_std_dev})
return inputs
def _build_inputs(self):
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cc', 10.0)
uo2.add_nuclide('U238', 1.0)
uo2.add_nuclide('U235', 0.02)
uo2.add_nuclide('O16', 2.0)
borated_water = openmc.Material(name='Borated water')
borated_water.set_density('g/cm3', 1)
borated_water.add_nuclide('B10', 10e-5)
borated_water.add_nuclide('H1', 2.0)
borated_water.add_nuclide('O16', 1.0)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, borated_water])
materials_file.export_to_xml()
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=1, \
name='Fuel OR')
left = openmc.XPlane(surface_id=2, x0=-2, name='left')
right = openmc.XPlane(surface_id=3, x0=2, name='right')
bottom = openmc.YPlane(y0=-2, name='bottom')
top = openmc.YPlane(y0=2, name='top')
left.boundary_type = 'vacuum'
right.boundary_type = 'reflective'
top.boundary_type = 'reflective'
bottom.boundary_type = 'reflective'
# Instantiate Cells
fuel = openmc.Cell(name='fuel')
water = openmc.Cell(name='water')
# Use surface half-spaces to define regions
fuel.region = -fuel_or
water.region = +fuel_or & -right & +bottom & -top
# Register Materials with Cells
fuel.fill = uo2
water.fill = borated_water
# Instantiate pin cell Universe
pin_cell = openmc.Universe(name='pin cell')
pin_cell.add_cells([fuel, water])
# Instantiate root Cell and Universe
root_cell = openmc.Cell(name='root cell')
root_cell.region = +left & -right & +bottom & -top
root_cell.fill = pin_cell
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Instantiate a Geometry, register the root Universe
geometry = openmc.Geometry(root_univ)
geometry.export_to_xml()
# Instantiate a Settings object, set all runtime parameters
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 0
settings_file.particles = 1000
#settings_file.output = {'tallies': True}
# Create an initial uniform spatial source distribution
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:],\
only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.export_to_xml()
# Tallies file
tallies_file = openmc.Tallies()
# Create partial current tallies from fuel to water
# Filters
two_groups = [0., 4e6, 20e6]
energy_filter = openmc.EnergyFilter(two_groups)
polar_filter = openmc.PolarFilter([0, np.pi / 4, np.pi])
azimuthal_filter = openmc.AzimuthalFilter([0, np.pi / 4, np.pi])
surface_filter = openmc.SurfaceFilter([1])
cell_from_filter = openmc.CellFromFilter(fuel)
cell_filter = openmc.CellFilter(water)
# Use Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Use a Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create partial current tallies from water to fuel
# Filters
cell_from_filter = openmc.CellFromFilter(water)
cell_filter = openmc.CellFilter(fuel)
# Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create a net current tally on inner surface using a surface filter
surface_filter = openmc.SurfaceFilter([1])
surf_tally1 = openmc.Tally(name='net_cylinder')
surf_tally1.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally1.scores = ['current']
tallies_file.append(surf_tally1)
# Create a net current tally on left surface using a surface filter
# This surface has a vacuum boundary condition, so leakage is tallied
surface_filter = openmc.SurfaceFilter([2])
surf_tally2 = openmc.Tally(name='leakage_left')
surf_tally2.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally2.scores = ['current']
tallies_file.append(surf_tally2)
# Create a net current tally on right surface using a surface filter
# This surface has a reflective boundary condition, but the zero
# net current is not picked up because particles are only tallied once
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
tallies_file.export_to_xml()
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.dagmc = False
# 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 not s.dagmc
def make_materials_geometry_tallies(enrichment_fraction_list,
batches=2,
inner_radius=500,
thickness=100,
breeder_material_name='Li',
temperature_in_C=500):
if isinstance(enrichment_fraction_list, list):
enrichment_fraction = enrichment_fraction_list[0]
else:
enrichment_fraction = enrichment_fraction_list
print('simulating ', batches, enrichment_fraction, inner_radius, thickness,
breeder_material_name)
#MATERIALS#
breeder_material = make_breeder_material(enrichment_fraction,
breeder_material_name,
temperature_in_C)
eurofer = make_eurofer()
mats = openmc.Materials([breeder_material, eurofer])
#GEOMETRY#
breeder_blanket_inner_surface = openmc.Sphere(R=inner_radius)
breeder_blanket_outer_surface = openmc.Sphere(R=inner_radius + thickness)
vessel_inner_surface = openmc.Sphere(R=inner_radius + thickness + 10)
vessel_outer_surface = openmc.Sphere(R=inner_radius + thickness + 20,
boundary_type='vacuum')
breeder_blanket_region = -breeder_blanket_outer_surface & +breeder_blanket_inner_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
breeder_blanket_cell.name = 'breeder_blanket'
inner_void_region = -breeder_blanket_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
vessel_region = +vessel_inner_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = 'vessel'
vessel_cell.fill = eurofer
blanket_vessel_gap_region = -vessel_inner_surface & +breeder_blanket_outer_surface
blanket_vessel_gap_cell = openmc.Cell(region=blanket_vessel_gap_region)
blanket_vessel_gap_cell.name = 'blanket_vessel_gap'
universe = openmc.Universe(cells=[
inner_void_cell, breeder_blanket_cell, blanket_vessel_gap_cell,
vessel_cell
])
geom = openmc.Geometry(universe)
#SIMULATION SETTINGS#
sett = openmc.Settings()
# batches = 3 # this is parsed as an argument
sett.batches = batches
sett.inactive = 10
sett.particles = 500
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(
e0=14080000.0, m_rat=5.0, kt=20000.0
) #neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
#TALLIES#
tallies = openmc.Tallies()
# define filters
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
cell_filter_vessel = openmc.CellFilter(vessel_cell)
particle_filter = openmc.ParticleFilter([1]) #1 is neutron, 2 is photon
surface_filter_rear_blanket = openmc.SurfaceFilter(
breeder_blanket_outer_surface)
surface_filter_rear_vessel = openmc.SurfaceFilter(vessel_outer_surface)
energy_bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
energy_filter = openmc.EnergyFilter(energy_bins)
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ['205']
tallies.append(tally)
tally = openmc.Tally(name='blanket_leakage')
tally.filters = [surface_filter_rear_blanket, particle_filter]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='vessel_leakage')
tally.filters = [surface_filter_rear_vessel, particle_filter]
tally.scores = ['current']
tallies.append(tally)
tally = openmc.Tally(name='breeder_blanket_spectra')
tally.filters = [cell_filter_breeder, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='vacuum_vessel_spectra')
tally.filters = [cell_filter_vessel, particle_filter, energy_filter]
tally.scores = ['flux']
tallies.append(tally)
tally = openmc.Tally(name='DPA')
tally.filters = [cell_filter_vessel, particle_filter]
tally.scores = ['444']
tallies.append(tally)
#RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
json_output = {
'enrichment_fraction': enrichment_fraction,
'inner_radius': inner_radius,
'thickness': thickness,
'breeder_material_name': breeder_material_name,
'temperature_in_C': temperature_in_C
}
tallies_to_retrieve = ['TBR', 'DPA', 'blanket_leakage', 'vessel_leakage']
for tally_name in tallies_to_retrieve:
tally = sp.get_tally(name=tally_name)
# for some reason the tally sum is a nested list
tally_result = tally.sum[0][0][0] / batches
# for some reason the tally std_dev is a nested list
tally_std_dev = tally.std_dev[0][0][0] / batches
json_output[tally_name] = {
'value': tally_result,
'std_dev': tally_std_dev
}
spectra_tallies_to_retrieve = [
'breeder_blanket_spectra', 'vacuum_vessel_spectra'
]
for spectra_name in spectra_tallies_to_retrieve:
spectra_tally = sp.get_tally(name=spectra_name)
spectra_tally_result = [entry[0][0] for entry in spectra_tally.mean]
spectra_tally_std_dev = [
entry[0][0] for entry in spectra_tally.std_dev
]
json_output[spectra_name] = {
'value': spectra_tally_result,
'std_dev': spectra_tally_std_dev,
'energy_groups': list(energy_bins)
}
return json_output
u = universes['UO2 Unrodded Assembly']
m = universes['MOX Unrodded Assembly']
lattices['Core'].universes = [[u, m, w], [m, u, w], [w, w, w]]
cells['Core'].fill = lattices['Core']
# Instantiate a Geometry, register the root Universe, and export to XML
geometry = openmc.Geometry()
geometry.root_universe = universes['Root']
geometry.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a Settings, set all runtime parameters, and export to XML
settings_file = openmc.Settings()
settings_file.energy_mode = "multi-group"
settings_file.cross_sections = "./mgxs.xml"
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.output = {'tallies': True, 'summary': True}
settings_file.source = openmc.Source(space=openmc.stats.Box(
[-32.13, -10.71, -1.0], [10.71, 32.13, 1.0], only_fissionable=True))
settings_file.entropy_lower_left = [-32.13, -32.13, -1.E50]
settings_file.entropy_upper_right = [32.13, 32.13, 1.E50]
settings_file.entropy_dimension = [51, 51, 1]
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC plots.xml File
h_cell = openmc.Cell(fill=lattice, region=h_box)
domain = openmc.model.rectangular_prism(width=1000,
height=1000,
origin=(0., 0., 0.),
boundary_type='vacuum')
domain_box = domain & -ceil & +floor & ~h_box
domain_cell = openmc.Cell(region=domain_box, fill=water)
model = openmc.model.Model()
model.geometry = openmc.Geometry([h_cell, domain_cell])
point = openmc.stats.Point((0, 0, 0))
settings = openmc.Settings()
src = openmc.Source(space=point, particle="photon")
src.angle = openmc.stats.Isotropic()
src.energy = openmc.stats.Discrete([1e6], [1.])
settings = openmc.Settings()
settings.run_mode = "fixed source"
settings.batches = 10
settings.inactive = 2
settings.particles = 100000
settings.photon_transport = True
settings.source = [src]
model.settings = settings
