def test_from_expression(reset):
# Create surface dictionary
s1 = openmc.ZCylinder(surface_id=1)
s2 = openmc.ZPlane(-10., surface_id=2)
s3 = openmc.ZPlane(10., surface_id=3)
surfs = {1: s1, 2: s2, 3: s3}
r = openmc.Region.from_expression('-1 2 -3', surfs)
assert isinstance(r, openmc.Intersection)
assert r[:] == [-s1, +s2, -s3]
r = openmc.Region.from_expression('+1 | -2 | +3', surfs)
assert isinstance(r, openmc.Union)
assert r[:] == [+s1, -s2, +s3]
r = openmc.Region.from_expression('~(-1)', surfs)
assert r == +s1
# Since & has higher precendence than |, the resulting region should be an
# instance of Union
r = openmc.Region.from_expression('1 -2 | 3', surfs)
assert isinstance(r, openmc.Union)
assert isinstance(r[0], openmc.Intersection)
assert r[0][:] == [+s1, -s2]
# ...but not if we use parentheses
r = openmc.Region.from_expression('1 (-2 | 3)', surfs)
assert isinstance(r, openmc.Intersection)
assert isinstance(r[1], openmc.Union)
assert r[1][:] == [-s2, +s3]
def test_complement(reset):
zcyl = openmc.ZCylinder(r=1., surface_id=1)
z0 = openmc.ZPlane(-5., surface_id=2)
z1 = openmc.ZPlane(5., surface_id=3)
outside = +zcyl | -z0 | +z1
inside = ~outside
outside_equiv = ~(-zcyl & +z0 & -z1)
inside_equiv = ~outside_equiv
# Check bounding box
for region in (inside, inside_equiv):
ll, ur = region.bounding_box
assert ll == pytest.approx((-1., -1., -5.))
assert ur == pytest.approx((1., 1., 5.))
assert_unbounded(outside)
assert_unbounded(outside_equiv)
# string represention
assert str(inside) == '~(1 | -2 | 3)'
# evaluate method
assert (0, 0, 0) in inside
assert (0, 0, 0) not in outside
assert (0, 0, 6) not in inside
assert (0, 0, 6) in outside
# translate method
inside_t = inside.translate((1.0, 1.0, 1.0))
ll, ur = inside_t.bounding_box
assert ll == pytest.approx((0., 0., -4.))
assert ur == pytest.approx((2., 2., 6.))
def build_default_materials_and_geometry(self):
# Define materials needed for 1D/1G slab problem
uo2_data = openmc.Macroscopic('uo2_iso', '71c')
uo2 = openmc.Material(name='UO2', material_id=1)
uo2.set_density('macro', 1.0)
uo2.add_macroscopic(uo2_data)
clad_data = openmc.Macroscopic('clad_ang_mu', '71c')
clad = openmc.Material(name='Clad', material_id=2)
clad.set_density('macro', 1.0)
clad.add_macroscopic(clad_data)
water_data = openmc.Macroscopic('lwtr_iso_mu', '71c')
water = openmc.Material(name='LWTR', material_id=3)
water.set_density('macro', 1.0)
water.add_macroscopic(water_data)
# Define the materials file.
self.materials.default_xs = '71c'
self.materials += (uo2, clad, water)
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0,
surface_id=200,
boundary_type='reflective')
right = openmc.XPlane(x0=10.0,
surface_id=201,
boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0,
surface_id=300,
boundary_type='reflective')
top = openmc.YPlane(y0=10.0,
surface_id=301,
boundary_type='reflective')
down = openmc.ZPlane(z0=0.0, surface_id=0, boundary_type='reflective')
fuel_clad_intfc = openmc.ZPlane(z0=2.0, surface_id=1)
clad_lwtr_intfc = openmc.ZPlane(z0=2.4, surface_id=2)
up = openmc.ZPlane(z0=5.0, surface_id=3, boundary_type='reflective')
# Define cells
c1 = openmc.Cell(cell_id=1)
c1.region = +left & -right & +bottom & -top & +down & -fuel_clad_intfc
c1.fill = uo2
c2 = openmc.Cell(cell_id=2)
c2.region = +left & -right & +bottom & -top & +fuel_clad_intfc & -clad_lwtr_intfc
c2.fill = clad
c3 = openmc.Cell(cell_id=3)
c3.region = +left & -right & +bottom & -top & +clad_lwtr_intfc & -up
c3.fill = water
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
root.add_cells((c1, c2, c3))
# Assign root universe to geometry
self.geometry.root_universe = root
def _add_outer_rings(self):
""" Adds BEAVRS RPV, liner, and downcomer """
self.s_upperBound = openmc.ZPlane(name='Highest Extent',
z0=c.struct_HighestExtent,
boundary_type='vacuum')
self.s_lowerBound = openmc.ZPlane(name='Lowest Extent',
z0=c.struct_LowestExtent,
boundary_type='vacuum')
# RPV
self.s_RPVOR = openmc.ZCylinder(name='RPV OR',
R=c.rpvOR,
boundary_type='vacuum')
self.s_RPVIR = openmc.ZCylinder(name='RPV IR', R=c.rpvIR)
self.c_RPV = openmc.Cell(name="RPV", fill=self.mats['Carbon Steel'])
self.c_RPV.region = (-self.s_RPVOR & +self.s_RPVIR & -self.s_upperBound
& +self.s_lowerBound)
# RPV liner
self.s_linerIR = openmc.ZCylinder(name='RPV Liner IR', R=c.linerIR)
self.c_liner = openmc.Cell(name="RPV Liner", fill=self.mats['SS304'])
self.c_liner.region = (-self.s_RPVIR & +self.s_linerIR
& -self.s_upperBound & +self.s_lowerBound)
# Downcomer
self.s_neutronShieldOR = openmc.ZCylinder(name='Shield Panel OR',
R=c.neutronShieldOR)
self.c_downcomer = openmc.Cell(name="Downcomer",
fill=self.mats['Borated Water'])
self.c_downcomer.region = (-self.s_linerIR & +self.s_neutronShieldOR
& -self.s_upperBound & +self.s_lowerBound)
def _build_inputs(self):
# Define materials
water = openmc.Material(1)
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.add_nuclide('B10', 0.0001)
water.add_s_alpha_beta('c_H_in_H2O')
water.set_density('g/cc', 1.0)
fuel = openmc.Material(2)
fuel.add_nuclide('U235', 1.0)
fuel.add_nuclide('Mo99', 0.1)
fuel.set_density('g/cc', 4.5)
materials = openmc.Materials((water, fuel))
materials.export_to_xml()
cyl = openmc.ZCylinder(surface_id=1, r=1.0, boundary_type='vacuum')
top_sphere = openmc.Sphere(surface_id=2,
z0=5.,
r=1.,
boundary_type='vacuum')
top_plane = openmc.ZPlane(surface_id=3, z0=5.)
bottom_sphere = openmc.Sphere(surface_id=4,
z0=-5.,
r=1.,
boundary_type='vacuum')
bottom_plane = openmc.ZPlane(surface_id=5, z0=-5.)
# Define geometry
inside_cyl = openmc.Cell(1,
fill=fuel,
region=-cyl & -top_plane & +bottom_plane)
top_hemisphere = openmc.Cell(2,
fill=water,
region=-top_sphere & +top_plane)
bottom_hemisphere = openmc.Cell(3,
fill=water,
region=-bottom_sphere & -top_plane)
root = openmc.Universe(0,
cells=(inside_cyl, top_hemisphere,
bottom_hemisphere))
geometry = openmc.Geometry(root)
geometry.export_to_xml()
# Set up stochastic volume calculation
ll, ur = root.bounding_box
vol_calcs = [
openmc.VolumeCalculation(list(root.cells.values()), 100000),
openmc.VolumeCalculation([water, fuel], 100000, ll, ur),
openmc.VolumeCalculation([root], 100000, ll, ur)
]
# Define settings
settings = openmc.Settings()
settings.run_mode = 'volume'
settings.volume_calculations = vol_calcs
settings.export_to_xml()
def _make_openmc_input(self):
"""Generate the OpenMC input XML
"""
# Define material
mat = openmc.Material()
mat.add_nuclide(self.nuclide, 1.0)
if self.thermal is not None:
name, suffix = self.thermal.split('.')
thermal_name = openmc.data.thermal.get_thermal_name(name)
mat.add_s_alpha_beta(thermal_name)
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')
# Set up geometry
x1 = openmc.XPlane(x0=-1.e9, boundary_type='reflective')
x2 = openmc.XPlane(x0=+1.e9, boundary_type='reflective')
y1 = openmc.YPlane(y0=-1.e9, boundary_type='reflective')
y2 = openmc.YPlane(y0=+1.e9, boundary_type='reflective')
z1 = openmc.ZPlane(z0=-1.e9, boundary_type='reflective')
z2 = openmc.ZPlane(z0=+1.e9, boundary_type='reflective')
cell = openmc.Cell(fill=materials)
cell.region = +x1 & -x2 & +y1 & -y2 & +z1 & -z2
geometry = openmc.Geometry([cell])
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.Isotropic()
source.energy = openmc.stats.Discrete([self.energy], [1.])
# 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.create_fission_neutrons = False
settings.export_to_xml(self.openmc_dir / 'settings.xml')
# Define tallies
energy_bins = np.logspace(np.log10(self._min_energy),
np.log10(1.0001 * self.energy),
self._bins + 1)
energy_filter = openmc.EnergyFilter(energy_bins)
tally = openmc.Tally(name='tally')
tally.filters = [energy_filter]
tally.scores = ['flux']
tallies = openmc.Tallies([tally])
tallies.export_to_xml(self.openmc_dir / 'tallies.xml')
def centers_z_cylinder():
cylinder = openmc.ZCylinder(r=1, x0=1, y0=2)
min_z = openmc.ZPlane(0)
max_z = openmc.ZPlane(1)
region = +min_z & -max_z & -cylinder
return openmc.model.pack_spheres(radius=_RADIUS,
region=region,
pf=_PACKING_FRACTION,
initial_pf=0.2)
def _add_fuel_pincells(self):
""" Adds BEAVRS fuel pincells """
self.enrichments = ['1.6%', '2.4%', '3.1%', '3.2%', '3.4%']
# Fuel radial surfaces
self.s_fuel_pellet_OR = openmc.ZCylinder(name='Fuel pellet OR', R=c.pelletOR)
self.s_fuel_clad_IR = openmc.ZCylinder(name='Fuel clad IR', R=c.cladIR)
self.s_fuel_clad_OR = openmc.ZCylinder(name='Fuel clad OR', R=c.cladOR)
self.s_fuel_plenumSpring_OR = openmc.ZCylinder(name='Fuel rod plenum spring OR', R=c.plenumSpringOR)
# Fuel axial surfaces
self.s_fuel_rod_bot = openmc.ZPlane(name='Fuel rod bottom', z0=c.fuel_Rod_bot)
self.s_fuel_lowerFitting_bot = self.s_fuel_rod_bot
self.s_fuel_lowerFitting_top = openmc.ZPlane(name='Fuel lower fitting top', z0=c.fuel_LowerFitting_top)
self.s_fuel_activeFuel_bot = self.s_fuel_lowerFitting_top
self.s_fuel_activeFuel_top = openmc.ZPlane(name='Fuel active region top', z0=c.fuel_ActiveFuel_top)
self.s_fuel_plenum_bot = self.s_fuel_activeFuel_top
self.s_fuel_plenum_top = openmc.ZPlane(name='Fuel plenum top', z0=c.fuel_Plenum_top)
self.s_fuel_upperFitting_bot = self.s_fuel_plenum_top
self.s_fuel_upperFitting_top = openmc.ZPlane(name='Fuel upper fitting top', z0=c.fuel_UpperFitting_top)
# Fuel pincell universes
self.u_fuel_active_pin = {}
for enr in self.enrichments:
self.u_fuel_active_pin[enr] = InfinitePinCell(name='Fuel rod active region - {0} enr'.format(enr))
self.u_fuel_active_pin[enr].add_ring(self.mats['Fuel {0}'.format(enr)], self.s_fuel_pellet_OR)
self.u_fuel_active_pin[enr].add_ring(self.mats['Helium'], self.s_fuel_clad_IR)
self.u_fuel_active_pin[enr].add_last_ring(self.mats['Zircaloy 4'])
self.u_fuel_active_pin[enr].finalize()
self.u_fuel_plenum = InfinitePinCell(name='Fuel rod plenum')
self.u_fuel_plenum.add_ring(self.mats['Inconel 718'], self.s_fuel_plenumSpring_OR)
self.u_fuel_plenum.add_ring(self.mats['Helium'], self.s_fuel_clad_IR)
self.u_fuel_plenum.add_last_ring(self.mats['Zircaloy 4'])
self.u_fuel_plenum.finalize()
# Fuel axial stack universes
self.u_fuel_p = {}
for enr in self.enrichments:
self.u_fuel_p[enr] = AxialPinCell(name='Fuel rod - {0} enr'.format(enr))
self.u_fuel_p[enr].add_axial_section(self.s_struct_supportPlate_bot, self.mats['Borated Water'])
self.u_fuel_p[enr].add_axial_section(self.s_fuel_rod_bot, self.mats['SS304 SPN'])
self.u_fuel_p[enr].add_axial_section(self.s_fuel_lowerFitting_top, self.mats['Zircaloy 4'])
self.u_fuel_p[enr].add_axial_section(self.s_fuel_activeFuel_top, self.u_fuel_active_pin[enr])
self.u_fuel_p[enr].add_axial_section(self.s_fuel_plenum_top, self.u_fuel_plenum)
self.u_fuel_p[enr].add_axial_section(self.s_fuel_upperFitting_top, self.mats['Zircaloy 4'])
self.u_fuel_p[enr].add_axial_section(self.s_struct_upperNozzle_bot, self.mats['Borated Water'])
self.u_fuel_p[enr].add_axial_section(self.s_struct_upperNozzle_top, self.mats['SS304 SPN'])
self.u_fuel_p[enr].add_last_axial_section(self.mats['Borated Water'])
self.u_fuel_p[enr] = self.u_fuel_p[enr].add_wrapper(self.u_grids, self.s_fuel_clad_OR)
self.u_fuel_p[enr].finalize()
def _make_openmc_input(self):
"""Generate the OpenMC input XML
"""
# Define material
mat = openmc.Material()
for element, fraction in self.elements:
mat.add_element(element, 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')
# Set up geometry
x1 = openmc.XPlane(x0=-1.e9, boundary_type='reflective')
x2 = openmc.XPlane(x0=+1.e9, boundary_type='reflective')
y1 = openmc.YPlane(y0=-1.e9, boundary_type='reflective')
y2 = openmc.YPlane(y0=+1.e9, boundary_type='reflective')
z1 = openmc.ZPlane(z0=-1.e9, boundary_type='reflective')
z2 = openmc.ZPlane(z0=+1.e9, boundary_type='reflective')
cell = openmc.Cell(fill=materials)
cell.region = +x1 & -x2 & +y1 & -y2 & +z1 & -z2
geometry = openmc.Geometry([cell])
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.Isotropic()
source.energy = openmc.stats.Discrete([self.energy], [1.])
source.particle = 'photon'
# Settings
settings = openmc.Settings()
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 tallies
energy_bins = np.logspace(np.log10(self._cutoff_energy),
np.log10(1.0001*self.energy), self._bins+1)
energy_filter = openmc.EnergyFilter(energy_bins)
particle_filter = openmc.ParticleFilter('photon')
tally = openmc.Tally(name='tally')
tally.filters = [energy_filter, particle_filter]
tally.scores = ['flux']
tallies = openmc.Tallies([tally])
tallies.export_to_xml(self.openmc_dir / 'tallies.xml')
def _add_bpra_pincells(self):
""" Adds BEAVRS BPRA pincells """
# BP radial surfaces
self.s_bp_innerclad_IR = openmc.ZCylinder(name='BPRA rod radius 1', R=c.burnabs1)
self.s_bp_innerclad_OR = openmc.ZCylinder(name='BPRA rod radius 2', R=c.burnabs2)
self.s_bp_poison_IR = openmc.ZCylinder(name='BPRA rod radius 3', R=c.burnabs3)
self.s_bp_poison_OR = openmc.ZCylinder(name='BPRA rod radius 4', R=c.burnabs4)
self.s_bp_outerclad_IR = openmc.ZCylinder(name='BPRA rod radius 5', R=c.burnabs5)
self.s_bp_outerclad_OR = openmc.ZCylinder(name='BPRA rod radius 6', R=c.burnabs6)
# BP axial surfaces
self.s_bp_rod_bot = openmc.ZPlane(name='Bottom of BPRA rod', z0=c.bpra_Rod_bot)
self.s_bp_lowerFitting_bot = self.s_bp_rod_bot
self.s_bp_lowerFitting_top = openmc.ZPlane(name='Top of lower fitting in BPRA rod', z0=c.bpra_LowerFitting_top)
self.s_bp_active_bot = self.s_bp_lowerFitting_top
self.s_bp_active_top = openmc.ZPlane(name='Top of active poison in BPRA rod', z0=c.bpra_Active_top)
self.s_bp_plenum_bot = self.s_bp_active_top
self.s_bp_plenum_top = openmc.ZPlane(name='Top of plenum in BPRA rod', z0=c.bpra_Plenum_top)
self.s_bp_rod_top = openmc.ZPlane(name='Top of BPRA rod', z0=c.bpra_Rod_top)
# BP pincell universes
self.u_bp_activePoison = InfinitePinCell(name='BPRA rod active poison')
self.u_bp_activePoison.add_ring(self.mats['Air'], self.s_bp_innerclad_IR)
self.u_bp_activePoison.add_ring(self.mats['SS304'], self.s_bp_innerclad_OR)
self.u_bp_activePoison.add_ring(self.mats['Air'], self.s_bp_poison_IR)
self.u_bp_activePoison.add_ring(self.mats['Borosilicate Glass'], self.s_bp_poison_OR)
self.u_bp_activePoison.add_ring(self.mats['Air'], self.s_bp_outerclad_IR)
self.u_bp_activePoison.add_last_ring(self.mats['SS304'])
self.u_bp_activePoison.finalize()
self.u_bp_plenum = InfinitePinCell(name='BPRA rod plenum')
self.u_bp_plenum.add_ring(self.mats['Air'], self.s_bp_innerclad_IR)
self.u_bp_plenum.add_ring(self.mats['SS304'], self.s_bp_innerclad_OR)
self.u_bp_plenum.add_ring(self.mats['Air'], self.s_bp_outerclad_IR)
self.u_bp_plenum.add_last_ring(self.mats['SS304'])
self.u_bp_plenum.finalize()
# BP axial stack universe
self.u_bp = AxialPinCell(name='BPRA rod')
self.u_bp.add_axial_section(self.s_struct_supportPlate_bot, self.mats['Borated Water'])
self.u_bp.add_axial_section(self.s_struct_lowerNozzle_top, self.mats['Water SPN'])
self.u_bp.add_axial_section(self.s_bp_rod_bot, self.mats['Borated Water'])
self.u_bp.add_axial_section(self.s_bp_lowerFitting_top, self.mats['SS304'])
self.u_bp.add_axial_section(self.s_bp_active_top, self.u_bp_activePoison)
self.u_bp.add_axial_section(self.s_bp_plenum_top, self.u_bp_plenum)
self.u_bp.add_axial_section(self.s_bp_rod_top, self.mats['SS304'])
self.u_bp.add_last_axial_section(self.mats['Borated Water'])
self.u_bp = self.u_bp.add_wrapper(self.u_gt, self.s_bp_outerclad_OR)
self.u_bp.finalize()
def model():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.0)
fuel.add_nuclide('U235', 1.0)
zr = openmc.Material()
zr.set_density('g/cm3', 1.0)
zr.add_nuclide('Zr90', 1.0)
model.materials.extend([fuel, zr])
box1 = openmc.model.rectangular_prism(10.0, 10.0)
box2 = openmc.model.rectangular_prism(20.0,
20.0,
boundary_type='reflective')
top = openmc.ZPlane(z0=10.0, boundary_type='vacuum')
bottom = openmc.ZPlane(z0=-10.0, boundary_type='vacuum')
cell1 = openmc.Cell(fill=fuel, region=box1 & +bottom & -top)
cell2 = openmc.Cell(fill=zr, region=~box1 & box2 & +bottom & -top)
model.geometry = openmc.Geometry([cell1, cell2])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
# Initialize a one-group structure
energy_groups = openmc.mgxs.EnergyGroups([0, 20.e6])
# Initialize MGXS Library for a few cross section types
# for one material-filled cell in the geometry
model.mgxs_lib = openmc.mgxs.Library(model.geometry)
model.mgxs_lib.by_nuclide = False
# Test all MGXS types
model.mgxs_lib.mgxs_types = openmc.mgxs.MGXS_TYPES + openmc.mgxs.MDGXS_TYPES
model.mgxs_lib.energy_groups = energy_groups
model.mgxs_lib.num_delayed_groups = 6
model.mgxs_lib.correction = None # Avoid warning about P0 correction
model.mgxs_lib.legendre_order = 3
model.mgxs_lib.domain_type = 'mesh'
# Instantiate a tally mesh
mesh = openmc.RegularMesh(mesh_id=1)
mesh.dimension = [2, 2]
mesh.lower_left = [-100., -100.]
mesh.width = [100., 100.]
model.mgxs_lib.domains = [mesh]
model.mgxs_lib.build_library()
# Add tallies
model.mgxs_lib.add_to_tallies_file(model.tallies, merge=False)
return model
def centers_rectangular_prism():
min_x = openmc.XPlane(0)
max_x = openmc.XPlane(1)
min_y = openmc.YPlane(0)
max_y = openmc.YPlane(1)
min_z = openmc.ZPlane(0)
max_z = openmc.ZPlane(1)
region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
return openmc.model.pack_spheres(radius=_RADIUS,
region=region,
pf=_PACKING_FRACTION,
initial_pf=0.2)
def __init__(self, xmin, xmax, ymin, ymax, zmin, zmax, **kwargs):
if xmin >= xmax:
raise ValueError('xmin must be less than xmax')
if ymin >= ymax:
raise ValueError('ymin must be less than ymax')
if zmin >= zmax:
raise ValueError('zmin must be less than zmax')
self.xmin = openmc.XPlane(x0=xmin, **kwargs)
self.xmax = openmc.XPlane(x0=xmax, **kwargs)
self.ymin = openmc.YPlane(y0=ymin, **kwargs)
self.ymax = openmc.YPlane(y0=ymax, **kwargs)
self.zmin = openmc.ZPlane(z0=zmin, **kwargs)
self.zmax = openmc.ZPlane(z0=zmax, **kwargs)
def build(self):
x0 = openmc.XPlane(x0=0, boundary_type="reflective")
right_inner_wall = openmc.Plane(A=cos(pi / 3),
B=-cos(pi / 6),
D=(RAD_MAJ - STEEL_THICK) *
cos(pi / 6))
right_outer_wall = openmc.Plane(A=cos(pi / 3),
B=-cos(pi / 6),
D=RAD_MAJ * cos(pi / 6))
right_refl_edge = openmc.Plane(boundary_type="reflective",
A=cos(pi / 6),
B=cos(pi / 3),
D=-GAP / 2 * cos(pi / 6) * 0)
ymin = openmc.YPlane(y0=-RAD_MAJ, boundary_type="vacuum", name="YMIN")
zmin = openmc.ZPlane(z0=-10, boundary_type="periodic", name="ZMIN")
zmax = openmc.ZPlane(z0=+10, boundary_type="periodic", name="ZMAX")
ru = openmc.Universe(name="root universe")
radialu = openmc.Universe(name="radial universe")
lattice = self.get_lattice()
dist = RAD_MAJ - STEEL_THICK - sqrt(3) / 2 * self.pitch
right_inner_water = openmc.Plane(A=cos(pi / 3),
B=-cos(pi / 6),
D=dist * cos(pi / 6))
# Fueled area
inner = openmc.Cell()
inner.region = -right_inner_water
inner.fill = lattice
radialu.add_cell(inner)
# Water buffer
buffer = openmc.Cell()
buffer.region = +right_inner_water & -right_inner_wall
buffer.fill = all_materials["Mod"]
radialu.add_cell(buffer)
# Reactor pressure vessel
rpv = openmc.Cell()
rpv.region = +right_inner_wall & -right_outer_wall
rpv.fill = all_materials["SS316"]
radialu.add_cell(rpv)
outside = openmc.Cell()
outside.region = +right_outer_wall
outside.fill = all_materials["Air"]
radialu.add_cell(outside)
# Root Universe
root_cell = openmc.Cell(name="root cell")
root_cell.fill = radialu
root_cell.region = +x0 & +ymin & +zmin & -zmax & -right_refl_edge
ru.add_cell(root_cell)
self._geometry = openmc.Geometry()
self._geometry.root_universe = ru
def _add_guide_tube_pincells(self):
""" Adds BEAVRS guide tube pincells """
# GT radial surfaces
self.s_gt_IR = openmc.ZCylinder(name='Guide tube IR', R=c.guideTubeIR)
self.s_gt_OR = openmc.ZCylinder(name='Guide tube OR', R=c.guideTubeOR)
self.s_gt_dashpot_IR = openmc.ZCylinder(name='Guide tube IR below dashpot', R=c.guideTubeDashIR)
self.s_gt_dashpot_OR = openmc.ZCylinder(name='Guide tube OR below dashpot', R=c.guideTubeDashOR)
# GT axial surfaces
self.s_gt_rod_bot = openmc.ZPlane(name='Bottom of GT rod', z0=c.gt_Rod_bot)
self.s_gt_dashpot_bot = self.s_gt_rod_bot
self.s_gt_dashpot_top = openmc.ZPlane(name='GT Dashpot plane', z0=c.gt_Dashpot_top)
self.s_gt_rod_top = openmc.ZPlane(name='Top of GT rod', z0=c.gt_Rod_top)
# GT pincell universes
self.u_gt_dashpot = InfinitePinCell(name='Empty GT below the dashpot')
self.u_gt_dashpot.add_ring(self.mats['Borated Water'], self.s_gt_dashpot_IR)
self.u_gt_dashpot.add_ring(self.mats['Zircaloy 4'], self.s_gt_dashpot_OR)
self.u_gt_dashpot.add_last_ring(self.mats['Borated Water'])
self.u_gt_dashpot.finalize()
self.u_gt_nodashpot_p = InfinitePinCell(name='Empty GT above the dashpot')
self.u_gt_nodashpot_p.add_ring(self.mats['Borated Water'], self.s_gt_IR)
self.u_gt_nodashpot_p.add_last_ring(self.mats['Zircaloy 4'])
self.u_gt_nodashpot_p.finalize()
# GT axial stack
self.u_gt = AxialPinCell(name='Empty Guide Tube')
self.u_gt.add_axial_section(self.s_struct_supportPlate_bot, self.mats['Borated Water'])
self.u_gt.add_axial_section(self.s_gt_rod_bot, self.mats['Water SPN'])
self.u_gt.add_axial_section(self.s_gt_dashpot_top, self.u_gt_dashpot)
self.u_gt.add_axial_section(self.s_gt_rod_top, self.u_gt_nodashpot_p)
self.u_gt.add_axial_section(self.s_struct_upperNozzle_top, self.mats['Water SPN'])
self.u_gt.add_last_axial_section(self.mats['Borated Water'])
self.u_gt = self.u_gt.add_wrapper(self.u_grids, self.s_gt_OR)
self.u_gt.finalize()
self.u_gt_nodashpot = AxialPinCell(name='Empty Guide Tube in Center Position')
self.u_gt_nodashpot.add_axial_section(self.s_struct_supportPlate_bot, self.mats['Borated Water'])
self.u_gt_nodashpot.add_axial_section(self.s_gt_rod_bot, self.mats['Water SPN'])
self.u_gt_nodashpot.add_axial_section(self.s_gt_rod_top, self.u_gt_nodashpot_p)
self.u_gt_nodashpot.add_axial_section(self.s_struct_upperNozzle_top, self.mats['Water SPN'])
self.u_gt_nodashpot.add_last_axial_section(self.mats['Borated Water'])
self.u_gt_nodashpot = self.u_gt_nodashpot.add_wrapper(self.u_grids, self.s_gt_OR)
self.u_gt_nodashpot.finalize()
def make_surfaces(zcore, hpitch):
overall_size = sum(zcore)
Z = [zcore[0]]
for i, zz in enumerate(zcore[1:]):
Z.append(zz + Z[i])
surfaces = []
surfaces.append(openmc.ZPlane(z0=0.0 - overall_size/2, boundary_type = 'vacuum'))
for i,z in enumerate(Z[:-1]):
surfaces.append(openmc.ZPlane(z0=z - overall_size/2))
surfaces.append(openmc.ZPlane(z0=Z[-1] - overall_size/2, boundary_type = 'vacuum'))
edge_length = (1./np.sqrt(3.0)) * hpitch
hexsurface = openmc.model.get_hexagonal_prism(edge_length=edge_length,
origin=(0.0, 0.0))
return hexsurface, surfaces
def _build_inputs(self):
# Set energy cutoff
energy_cutoff = 4.0
# Material is composed of H-1
mat = openmc.Material(material_id=1, name='mat')
mat.set_density('atom/b-cm', 0.069335)
mat.add_nuclide('H1', 40.0)
materials_file = openmc.Materials([mat])
materials_file.export_to_xml()
# Cell is box with reflective boundary
x1 = openmc.XPlane(surface_id=1, x0=-1)
x2 = openmc.XPlane(surface_id=2, x0=1)
y1 = openmc.YPlane(surface_id=3, y0=-1)
y2 = openmc.YPlane(surface_id=4, y0=1)
z1 = openmc.ZPlane(surface_id=5, z0=-1)
z2 = openmc.ZPlane(surface_id=6, z0=1)
for surface in [x1, x2, y1, y2, z1, z2]:
surface.boundary_type = 'reflective'
box = openmc.Cell(cell_id=1, name='box')
box.region = +x1 & -x2 & +y1 & -y2 & +z1 & -z2
box.fill = mat
root = openmc.Universe(universe_id=0, name='root universe')
root.add_cell(box)
geometry = openmc.Geometry(root)
geometry.export_to_xml()
# Set the running parameters
settings_file = openmc.Settings()
settings_file.run_mode = 'fixed source'
settings_file.batches = 10
settings_file.particles = 100
settings_file.cutoff = {'energy_neutron': energy_cutoff}
bounds = [-1, -1, -1, 1, 1, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
watt_dist = openmc.stats.Watt()
settings_file.source = openmc.source.Source(space=uniform_dist,
energy=watt_dist)
settings_file.export_to_xml()
# Tally flux under energy cutoff
tallies = openmc.Tallies()
tally = openmc.Tally(1)
tally.scores = ['flux']
energy_filter = openmc.filter.EnergyFilter((0.0, energy_cutoff))
tally.filters = [energy_filter]
tallies.append(tally)
tallies.export_to_xml()
def test_zplane():
s = openmc.ZPlane(z0=3.)
assert s.z0 == 3.
# Check bounding box
ll, ur = (+s).bounding_box
assert ll == pytest.approx((-np.inf, -np.inf, 3.))
assert np.all(np.isinf(ur))
ll, ur = (-s).bounding_box
assert ur == pytest.approx((np.inf, np.inf, 3.))
assert np.all(np.isinf(ll))
# __contains__ on associated half-spaces
assert (0, 0, 5) in +s
assert (0, 0, 5) not in -s
assert (-2, 1, -10) in -s
assert (-2, 1, -10) not in +s
# evaluate method
assert s.evaluate((0., 0., 10.)) == pytest.approx(7.)
# translate method
st = s.translate((0.0, 0.0, 1.0))
assert st.z0 == s.z0 + 1
# rotate method
# make sure rotating around z axis does nothing to coefficients
sr = s.rotate((0., 0., 123), order='zxy')
assert s._get_base_coeffs() == pytest.approx(sr._get_base_coeffs())
# rotate around x by -90 deg and y by 90 deg should give negative x-plane
sr = s.rotate((-90, 0., 90.))
assert (-1., 0., 0., 3.) == pytest.approx(sr._get_base_coeffs())
# Make sure repr works
repr(s)
def test_zplane():
s = openmc.ZPlane(z0=3.)
assert s.z0 == 3.
# Check bounding box
ll, ur = (+s).bounding_box
assert ll == pytest.approx((-np.inf, -np.inf, 3.))
assert np.all(np.isinf(ur))
ll, ur = (-s).bounding_box
assert ur == pytest.approx((np.inf, np.inf, 3.))
assert np.all(np.isinf(ll))
# __contains__ on associated half-spaces
assert (0, 0, 5) in +s
assert (0, 0, 5) not in -s
assert (-2, 1, -10) in -s
assert (-2, 1, -10) not in +s
# evaluate method
assert s.evaluate((0., 0., 10.)) == pytest.approx(7.)
# translate method
st = s.translate((0.0, 0.0, 1.0))
assert st.z0 == s.z0 + 1
# Make sure repr works
repr(s)
def __init__(self, *args, **kwargs):
super().__init__(*args, **kwargs)
# Extract universes encapsulating fuel and water assemblies
geometry = self._model.geometry
water = geometry.get_universes_by_name('water assembly (hot)')[0]
fuel = geometry.get_universes_by_name('fuel assembly (hot)')[0]
# Construct a 3x3 lattice of fuel assemblies
core_lat = openmc.RectLattice(name='3x3 Core Lattice', lattice_id=202)
core_lat.lower_left = (-32.13, -32.13)
core_lat.pitch = (21.42, 21.42)
core_lat.universes = [[fuel, water, water], [fuel, fuel, fuel],
[water, water, water]]
# Create bounding surfaces
min_x = openmc.XPlane(-32.13, boundary_type='reflective')
max_x = openmc.XPlane(+32.13, boundary_type='reflective')
min_y = openmc.YPlane(-32.13, boundary_type='reflective')
max_y = openmc.YPlane(+32.13, boundary_type='reflective')
min_z = openmc.ZPlane(0, boundary_type='reflective')
max_z = openmc.ZPlane(+32.13, boundary_type='reflective')
# Define root universe
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_cell = openmc.Cell(cell_id=1)
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
root_cell.fill = core_lat
root_univ.add_cell(root_cell)
# Over-ride geometry in the input set with this 3x3 lattice
self._model.geometry.root_universe = root_univ
# Initialize a "distribcell" filter for the fuel pin cell
distrib_filter = openmc.DistribcellFilter(27)
# Initialize the tallies
tally = openmc.Tally(name='distribcell tally', tally_id=27)
tally.filters.append(distrib_filter)
tally.scores.append('nu-fission')
# Assign the tallies file to the input set
self._model.tallies.append(tally)
# Specify summary output and correct source sampling box
self._model.settings.source = openmc.Source(space=openmc.stats.Box(
[-32, -32, 0], [32, 32, 32], only_fissionable=True))
def _build_inputs(self):
# Define materials
water = openmc.Material(1)
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.add_s_alpha_beta('c_H_in_H2O')
water.set_density('g/cc', 1.0)
fuel = openmc.Material(2)
fuel.add_nuclide('U235', 1.0)
fuel.set_density('g/cc', 4.5)
materials = openmc.Materials((water, fuel))
materials.default_temperature = '294K'
materials.export_to_xml()
# Define geometry
x_min = openmc.XPlane(surface_id=1, x0=-5., boundary_type='periodic')
x_max = openmc.XPlane(surface_id=2, x0=5., boundary_type='periodic')
x_max.periodic_surface = x_min
y_min = openmc.YPlane(surface_id=3, y0=-5., boundary_type='periodic')
y_max = openmc.YPlane(surface_id=4, y0=5., boundary_type='periodic')
z_min = openmc.ZPlane(surface_id=5, z0=-5., boundary_type='reflective')
z_max = openmc.ZPlane(surface_id=6, z0=5., boundary_type='reflective')
z_cyl = openmc.ZCylinder(surface_id=7, x0=-2.5, y0=2.5, r=2.0)
outside_cyl = openmc.Cell(1,
fill=water,
region=(+x_min & -x_max & +y_min & -y_max
& +z_min & -z_max & +z_cyl))
inside_cyl = openmc.Cell(2, fill=fuel, region=+z_min & -z_max & -z_cyl)
root_universe = openmc.Universe(0, cells=(outside_cyl, inside_cyl))
geometry = openmc.Geometry()
geometry.root_universe = root_universe
geometry.export_to_xml()
# Define settings
settings = openmc.Settings()
settings.particles = 1000
settings.batches = 4
settings.inactive = 0
settings.source = openmc.Source(space=openmc.stats.Box(
*outside_cyl.region.bounding_box))
settings.export_to_xml()
def __init__(self, center_base, height, radius, axis='z', **kwargs):
cx, cy, cz = center_base
check_greater_than('cylinder height', height, 0.0)
check_greater_than('cylinder radius', radius, 0.0)
check_value('cylinder axis', axis, ('x', 'y', 'z'))
if axis == 'x':
self.cyl = openmc.XCylinder(y0=cy, z0=cz, r=radius, **kwargs)
self.bottom = openmc.XPlane(x0=cx, **kwargs)
self.top = openmc.XPlane(x0=cx + height, **kwargs)
elif axis == 'y':
self.cyl = openmc.YCylinder(x0=cx, z0=cz, r=radius, **kwargs)
self.bottom = openmc.YPlane(y0=cy, **kwargs)
self.top = openmc.YPlane(y0=cy + height, **kwargs)
elif axis == 'z':
self.cyl = openmc.ZCylinder(x0=cx, y0=cy, r=radius, **kwargs)
self.bottom = openmc.ZPlane(z0=cz, **kwargs)
self.top = openmc.ZPlane(z0=cz + height, **kwargs)
def test_get_all_surfaces(uo2):
planes = [openmc.ZPlane(z0=z) for z in np.linspace(-100., 100.)]
slabs = []
for region in openmc.model.subdivide(planes):
slabs.append(openmc.Cell(fill=uo2, region=region))
geom = openmc.Geometry(slabs)
surfs = set(geom.get_all_surfaces().values())
assert not surfs ^ set(planes)
def _add_outer_rings(self):
""" Adds BEAVRS RPV, liner, and downcomer """
# Change z-dimension based on whether 2D or 3D problem
if self.is_2d:
self.s_upperBound = openmc.ZPlane(name='Highest Extent',
z0=c.struct_HighestExtent_2d,
boundary_type='reflective')
self.s_lowerBound = openmc.ZPlane(name='Lowest Extent',
z0=c.struct_LowestExtent_2d,
boundary_type='reflective')
else:
self.s_upperBound = openmc.ZPlane(name='Highest Extent',
z0=c.struct_HighestExtent,
boundary_type='vacuum')
self.s_lowerBound = openmc.ZPlane(name='Lowest Extent',
z0=c.struct_LowestExtent,
boundary_type='vacuum')
# RPV
self.s_RPVOR = openmc.ZCylinder(name='RPV OR',
r=c.rpvOR,
boundary_type='vacuum')
self.s_RPVIR = openmc.ZCylinder(name='RPV IR', r=c.rpvIR)
self.c_RPV = openmc.Cell(name="RPV", fill=self.mats['Carbon Steel'])
self.c_RPV.region = (-self.s_RPVOR & +self.s_RPVIR & -self.s_upperBound
& +self.s_lowerBound)
# RPV liner
self.s_linerIR = openmc.ZCylinder(name='RPV Liner IR', r=c.linerIR)
self.c_liner = openmc.Cell(name="RPV Liner", fill=self.mats['SS304'])
self.c_liner.region = (-self.s_RPVIR & +self.s_linerIR
& -self.s_upperBound & +self.s_lowerBound)
# Downcomer
self.s_neutronShieldOR = openmc.ZCylinder(name='Shield Panel OR',
r=c.neutronShieldOR)
self.c_downcomer = openmc.Cell(name="Downcomer",
fill=self.mats['Borated Water'])
self.c_downcomer.region = (-self.s_linerIR & +self.s_neutronShieldOR
& -self.s_upperBound & +self.s_lowerBound)
def model():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cm3', 10.0)
fuel.add_nuclide('U235', 1.0)
zr = openmc.Material()
zr.set_density('g/cm3', 1.0)
zr.add_nuclide('Zr90', 1.0)
model.materials.extend([fuel, zr])
box1 = openmc.model.rectangular_prism(10.0, 10.0)
box2 = openmc.model.rectangular_prism(20.0,
20.0,
boundary_type='reflective')
top = openmc.ZPlane(z0=10.0, boundary_type='vacuum')
bottom = openmc.ZPlane(z0=-10.0, boundary_type='vacuum')
cell1 = openmc.Cell(fill=fuel, region=box1 & +bottom & -top)
cell2 = openmc.Cell(fill=zr, region=~box1 & box2 & +bottom & -top)
model.geometry = openmc.Geometry([cell1, cell2])
model.settings.batches = 5
model.settings.inactive = 0
model.settings.particles = 1000
mesh = openmc.Mesh()
mesh.lower_left = (-10.0, -10.0, -10.0)
mesh.upper_right = (10.0, 10.0, 10.0)
mesh.dimension = (3, 3, 3)
mesh_surface_filter = openmc.MeshSurfaceFilter(mesh)
energy_filter = openmc.EnergyFilter([0.0, 0.253, 20.0e6])
tally1 = openmc.Tally()
tally1.filters = [mesh_surface_filter]
tally1.scores = ['current']
tally2 = openmc.Tally()
tally2.filters = [mesh_surface_filter, energy_filter]
tally2.scores = ['current']
model.tallies.extend([tally1, tally2])
return model
def test_get_surfaces():
s1 = openmc.XPlane()
s2 = openmc.YPlane()
s3 = openmc.ZPlane()
region = (+s1 & -s2) | +s3
# Make sure get_surfaces() returns all surfaces
surfs = set(region.get_surfaces().values())
assert not (surfs ^ {s1, s2, s3})
inverse = ~region
surfs = set(inverse.get_surfaces().values())
assert not (surfs ^ {s1, s2, s3})
def model():
model = openmc.model.Model()
fuel = openmc.Material()
fuel.set_density('g/cc', 10.0)
fuel.add_nuclide('U235', 1.0)
h1 = openmc.Material()
h1.set_density('g/cc', 0.1)
h1.add_nuclide('H1', 0.1)
inner_sphere = openmc.Sphere(x0=1.0, r=5.0)
fuel_cell = openmc.Cell(fill=fuel, region=-inner_sphere)
hydrogen_cell = openmc.Cell(fill=h1, region=+inner_sphere)
univ = openmc.Universe(cells=[fuel_cell, hydrogen_cell])
# Create one cell on top of the other. Only one
# has a rotation
box = openmc.rectangular_prism(15., 15., 'z', boundary_type='vacuum')
lower_z = openmc.ZPlane(-7.5, boundary_type='vacuum')
upper_z = openmc.ZPlane(22.5, boundary_type='vacuum')
middle_z = openmc.ZPlane(7.5)
lower_cell = openmc.Cell(fill=univ, region=box & +lower_z & -middle_z)
lower_cell.rotation = (10, 20, 30)
upper_cell = openmc.Cell(fill=univ, region=box & +middle_z & -upper_z)
upper_cell.translation = (0, 0, 15)
model.geometry = openmc.Geometry(root=[lower_cell, upper_cell])
model.settings.particles = 10000
model.settings.inactive = 5
model.settings.batches = 10
source_box = openmc.stats.Box((-4., -4., -4.), (4., 4., 4.))
model.settings.source = openmc.Source(space=source_box)
return model
def _add_structural_axials(self):
""" Adds structure axial surfaces common for most pincells """
self.s_struct_supportPlate_bot = openmc.ZPlane(name='Support plate bottom', z0=c.struct_SupportPlate_bot)
self.s_struct_supportPlate_top = openmc.ZPlane(name='Support plate top', z0=c.struct_SupportPlate_top)
self.s_struct_lowerNozzle_bot = openmc.ZPlane(name='Lower nozzle bottom', z0=c.struct_LowerNozzle_bot)
self.s_struct_lowerNozzle_top = openmc.ZPlane(name='Lower nozzle top', z0=c.struct_LowerNozzle_top)
self.s_struct_upperNozzle_bot = openmc.ZPlane(name='Upper nozzle bottom', z0=c.struct_UpperNozzle_bot)
self.s_struct_upperNozzle_top = openmc.ZPlane(name='Upper nozzle top', z0=c.struct_UpperNozzle_top)
def create_geom(radius, fuel, air):
"""Create geometry for some radius sphere
"""
# Create geometry
sphere = openmc.Sphere(R=radius)
min_x = openmc.XPlane(x0=-radius, boundary_type='vacuum')
max_x = openmc.XPlane(x0=+radius, boundary_type='vacuum')
min_y = openmc.YPlane(y0=-radius, boundary_type='vacuum')
max_y = openmc.YPlane(y0=+radius, boundary_type='vacuum')
min_z = openmc.ZPlane(z0=-radius, boundary_type='vacuum')
max_z = openmc.ZPlane(z0=+radius, boundary_type='vacuum')
# Create Universe
universe = openmc.Universe(name='Universe')
fuel_cell = openmc.Cell(name='fuel')
fuel_cell.fill = fuel
fuel_cell.region = -sphere
universe.add_cell(fuel_cell)
air_cell = openmc.Cell(name='air')
air_cell.fill = None
air_cell.region = +sphere
universe.add_cell(air_cell)
# Create root cell
root_cell = openmc.Cell(name='root_cell')
root_cell.fill = universe
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
root_universe = openmc.Universe(universe_id=0, name='root universe')
root_universe.add_cell(root_cell)
geometry = openmc.Geometry()
geometry.root_universe = root_universe
geometry.export_to_xml()
def create_cube(x_length,
y_length,
z_length,
start_point,
boundary_type_set='reflective'):
x_pos = openmc.XPlane(x0=max(x_length + start_point.x, start_point.x),
boundary_type=boundary_type_set)
x_neg = openmc.XPlane(x0=min(x_length + start_point.x, start_point.x),
boundary_type=boundary_type_set)
y_pos = openmc.YPlane(y0=max(y_length + start_point.y, start_point.y),
boundary_type=boundary_type_set)
y_neg = openmc.YPlane(y0=min(y_length + start_point.y, start_point.y),
boundary_type=boundary_type_set)
z_pos = openmc.ZPlane(z0=max(z_length + start_point.z, start_point.z),
boundary_type=boundary_type_set)
z_neg = openmc.ZPlane(z0=min(z_length + start_point.z, start_point.z),
boundary_type=boundary_type_set)
plane_list = []
plane_list.append([x_pos, x_neg])
plane_list.append([y_pos, y_neg])
plane_list.append([z_pos, z_neg])
return truncation_create(plane_list)
