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 _build_inputs(self):
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('UO2')
# Instantiate some Materials and register the appropriate objects
mat = openmc.Material(material_id=1, name='UO2 fuel')
mat.set_density('macro', 1.0)
mat.add_macroscopic(uo2_data)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([mat])
materials_file.cross_sections = "./mgxs.h5"
materials_file.export_to_xml()
# Instantiate ZCylinder surfaces
left = openmc.XPlane(surface_id=4, x0=-5., name='left')
right = openmc.XPlane(surface_id=5, x0=5., name='right')
bottom = openmc.YPlane(surface_id=6, y0=-5., name='bottom')
top = openmc.YPlane(surface_id=7, y0=5., name='top')
left.boundary_type = 'reflective'
right.boundary_type = 'vacuum'
top.boundary_type = 'reflective'
bottom.boundary_type = 'reflective'
# Instantiate Cells
fuel = openmc.Cell(cell_id=1, name='cell 1')
# Use surface half-spaces to define regions
fuel.region = +left & -right & +bottom & -top
# Register Materials with Cells
fuel.fill = mat
# Instantiate Universe
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
root.add_cells([fuel])
# Instantiate a Geometry, register the root Universe, and export to XML
geometry = openmc.Geometry(root)
geometry.export_to_xml()
settings_file = openmc.Settings()
settings_file.energy_mode = "multi-group"
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
# Create an initial uniform spatial source distribution
bounds = [-5, -5, -5, 5, 5, 5]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.export_to_xml()
def make_materials(self):
materials_file = openmc.Materials()
mats = []
macros = []
for i in range(len(self.mat_names)):
macros.append(openmc.Macroscopic(self.mat_names[i]))
mats.append(openmc.Material(name=self.mat_names[i]))
mats[-1].set_density('macro', 1.0)
mats[-1].add_macroscopic(macros[-1])
materials_file.add_material(mats[-1])
materials_file.cross_sections = GROUP_FILES[self.groups]
return mats, materials_file
def add_macroscopic(self, macroscopic):
"""Add a macroscopic to the material. This will also set the
density of the material to 1.0, unless it has been otherwise set,
as a default for Macroscopic cross sections.
Parameters
----------
macroscopic : str or openmc.Macroscopic
Macroscopic to add
"""
# Ensure no nuclides, elements, or sab are added since these would be
# incompatible with macroscopics
if self._nuclides or self._elements or self._sab:
msg = 'Unable to add a Macroscopic data set to Material ID="{0}" ' \
'with a macroscopic value "{1}" as an incompatible data ' \
'member (i.e., nuclide, element, or S(a,b) table) ' \
'has already been added'.format(self._id, macroscopic)
raise ValueError(msg)
if not isinstance(macroscopic, (openmc.Macroscopic, basestring)):
msg = 'Unable to add a Macroscopic to Material ID="{0}" with a ' \
'non-Macroscopic value "{1}"'.format(self._id, macroscopic)
raise ValueError(msg)
if isinstance(macroscopic, openmc.Macroscopic):
# Copy this Macroscopic to separate it from the Macroscopic in
# other Materials
macroscopic = deepcopy(macroscopic)
else:
macroscopic = openmc.Macroscopic(macroscopic)
if self._macroscopic is None:
self._macroscopic = macroscopic
else:
msg = 'Unable to add a Macroscopic to Material ID="{0}". ' \
'Only one Macroscopic allowed per ' \
'Material.'.format(self._id)
raise ValueError(msg)
# Generally speaking, the density for a macroscopic object will
# be 1.0. Therefore, lets set density to 1.0 so that the user
# doesnt need to set it unless its needed.
# Of course, if the user has already set a value of density,
# then we will not override it.
if self._density is None:
self.set_density('macro', 1.0)
0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200
], [
0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000
]]])
scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
h2o_xsdata.set_scatter_matrix(scatter_matrix)
mg_cross_sections_file = openmc.MGXSLibrary(groups)
mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata])
mg_cross_sections_file.export_to_hdf5()
###############################################################################
# Create materials for the problem
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('UO2')
h2o_data = openmc.Macroscopic('LWTR')
# Instantiate some Materials and register the appropriate Macroscopic objects
uo2 = openmc.Material(name='UO2 fuel')
uo2.set_density('macro', 1.0)
uo2.add_macroscopic(uo2_data)
water = openmc.Material(name='Water')
water.set_density('macro', 1.0)
water.add_macroscopic(h2o_data)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, water])
materials_file.cross_sections = "mgxs.h5"
materials_file.export_to_xml()
control_rod_xsdata.nu_fission = np.zeros(7)
control_rod_xsdata.chi = np.zeros(7)
mg_cross_sections_file = openmc.MGXSLibrary(groups)
mg_cross_sections_file.add_xsdatas([uo2_xsdata, mox43_xsdata, mox7_xsdata, mox87_xsdata,
fiss_chamber_xsdata, guide_tube_xsdata, water_xsdata,
control_rod_xsdata])
mg_cross_sections_file.export_to_xml()
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('uo2', '300k')
mox43_data = openmc.Macroscopic('mox43', '300k')
mox7_data = openmc.Macroscopic('mox7', '300k')
mox87_data = openmc.Macroscopic('mox87', '300k')
fiss_chamber_data = openmc.Macroscopic('fiss_chamber', '300k')
guide_tube_data = openmc.Macroscopic('guide_tube', '300k')
water_data = openmc.Macroscopic('water', '300k')
control_rod_data = openmc.Macroscopic('control_rod', '300k')
# Instantiate Materials dictionary
materials = {}
# Instantiate some Materials and register the appropriate Nuclides
materials['UO2'] = openmc.Material(material_id=1, name='UO2')
materials['UO2'].set_density('macro', 1.0)
materials['UO2'].add_macroscopic(uo2_data)
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
[0.0000000, 0.0000000, 0.0000000, 0.0910284, 0.4155100, 0.0637320, 0.0121390],
[0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290],
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200],
[0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]]
mg_cross_sections_file = openmc.MGXSLibrary(groups)
mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata])
mg_cross_sections_file.export_to_xml()
###############################################################################
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Macroscopic Data
uo2_data = openmc.Macroscopic('UO2', '300K')
h2o_data = openmc.Macroscopic('LWTR', '300K')
# Instantiate some Materials and register the appropriate Macroscopic objects
uo2 = openmc.Material(material_id=1, name='UO2 fuel')
uo2.set_density('macro', 1.0)
uo2.add_macroscopic(uo2_data)
water = openmc.Material(material_id=2, name='Water')
water.set_density('macro', 1.0)
water.add_macroscopic(h2o_data)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, water])
materials_file.default_xs = '300K'
materials_file.export_to_xml()
def create_mg_mode(self, xsdata_names=None, xs_ids=None,
tabular_legendre=None, tabular_points=33):
"""Creates an openmc.MGXSLibrary object to contain the MGXS data for the
Multi-Group mode of OpenMC as well as the associated openmc.Materials
and openmc.Geometry objects. The created Geometry is the same as that
used to generate the MGXS data, with the only differences being
modifications to point to newly-created Materials which point to the
multi-group data. This method only creates a macroscopic
MGXS Library even if nuclidic tallies are specified in the Library.
Parameters
----------
xsdata_names : Iterable of str
List of names to apply to the "xsdata" entries in the
resultant mgxs data file. Defaults to 'set1', 'set2', ...
xs_ids : str or Iterable of str
Cross section set identifier (i.e., '71c') for all
data sets (if only str) or for each individual one
(if iterable of str). Defaults to '1m'.
tabular_legendre : None or bool
Flag to denote whether or not the Legendre expansion of the
scattering angular distribution is to be converted to a tabular
representation by OpenMC. A value of `True` means that it is to be
converted while a value of `False` means that it will not be.
Defaults to `None` which leaves the default behavior of OpenMC in
place (the distribution is converted to a tabular representation).
tabular_points : int
This parameter is not used unless the ``tabular_legendre``
parameter is set to `True`. In this case, this parameter sets the
number of equally-spaced points in the domain of [-1,1] to be used
in building the tabular distribution. Default is `33`.
Returns
-------
mgxs_file : openmc.MGXSLibrary
Multi-Group Cross Section File that is ready to be printed to the
file of choice by the user.
materials : openmc.Materials
Materials file ready to be printed with all the macroscopic data
present within this Library.
geometry : openmc.Geometry
Materials file ready to be printed with all the macroscopic data
present within this Library.
Raises
------
ValueError
When the Library object is initialized with insufficient types of
cross sections for the Library.
See also
--------
Library.create_mg_library()
Library.dump_to_file()
"""
# Check to ensure the Library contains the correct
# multi-group cross section types
self.check_library_for_openmc_mgxs()
if xsdata_names is not None:
cv.check_iterable_type('xsdata_names', xsdata_names, basestring)
if xs_ids is not None:
if isinstance(xs_ids, basestring):
# If we only have a string lets convert it now to a list
# of strings.
xs_ids = [xs_ids for i in range(len(self.domains))]
else:
cv.check_iterable_type('xs_ids', xs_ids, basestring)
else:
xs_ids = ['1m' for i in range(len(self.domains))]
xs_type = 'macro'
# Initialize MGXS File
mgxs_file = openmc.MGXSLibrary(self.energy_groups)
# Create a copy of the Geometry to differentiate for these Macroscopics
geometry = copy.deepcopy(self.openmc_geometry)
materials = openmc.Materials()
# Get all Cells from the Geometry for differentiation
all_cells = geometry.get_all_material_cells()
# Create the xsdata object and add it to the mgxs_file
for i, domain in enumerate(self.domains):
# Build & add metadata to XSdata object
if xsdata_names is None:
xsdata_name = 'set' + str(i + 1)
else:
xsdata_name = xsdata_names[i]
# Create XSdata and Macroscopic for this domain
xsdata = self.get_xsdata(domain, xsdata_name, nuclide='total',
xs_type=xs_type, xs_id=xs_ids[i],
tabular_legendre=tabular_legendre,
tabular_points=tabular_points)
mgxs_file.add_xsdata(xsdata)
macroscopic = openmc.Macroscopic(name=xsdata_name, xs=xs_ids[i])
# Create Material and add to collection
material = openmc.Material(name=xsdata_name + '.' + xs_ids[i])
material.add_macroscopic(macroscopic)
materials.append(material)
# Differentiate Geometry with new Material
if self.domain_type == 'material':
# Fill all appropriate Cells with new Material
for cell in all_cells:
if cell.fill.id == domain.id:
cell.fill = material
elif self.domain_type == 'cell':
for cell in all_cells:
if cell.id == domain.id:
cell.fill = material
return mgxs_file, materials, geometry
def slab_mg(reps=None, as_macro=True):
"""Create a one-group, 1D slab model.
Parameters
----------
reps : list, optional
List of angular representations. Each item corresponds to materials and
dictates the angular representation of the multi-group cross
sections---isotropic ('iso') or angle-dependent ('ang'), and if Legendre
scattering or tabular scattering ('mu') is used. Thus, items can be
'ang', 'ang_mu', 'iso', or 'iso_mu'.
as_macro : bool, optional
Whether :class:`openmc.Macroscopic` is used
Returns
-------
model : openmc.model.Model
One-group, 1D slab model
"""
model = openmc.model.Model()
# Define materials needed for 1D/1G slab problem
mat_names = ['uo2', 'clad', 'lwtr']
mgxs_reps = ['ang', 'ang_mu', 'iso', 'iso_mu']
if reps is None:
reps = mgxs_reps
xs = []
i = 0
for mat in mat_names:
for rep in reps:
i += 1
if as_macro:
xs.append(openmc.Macroscopic(mat + '_' + rep))
m = openmc.Material(name=str(i))
m.set_density('macro', 1.)
m.add_macroscopic(xs[-1])
else:
xs.append(openmc.Nuclide(mat + '_' + rep))
m = openmc.Material(name=str(i))
m.set_density('atom/b-cm', 1.)
m.add_nuclide(xs[-1].name, 1.0, 'ao')
model.materials.append(m)
# Define the materials file
model.xs_data = xs
model.materials.cross_sections = "../1d_mgxs.h5"
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0, boundary_type='reflective')
right = openmc.XPlane(x0=10.0, boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0, boundary_type='reflective')
top = openmc.YPlane(y0=10.0, boundary_type='reflective')
# for each material add a plane
planes = [openmc.ZPlane(z0=0.0, boundary_type='reflective')]
dz = round(5. / float(len(model.materials)), 4)
for i in range(len(model.materials) - 1):
planes.append(openmc.ZPlane(z0=dz * float(i + 1)))
planes.append(openmc.ZPlane(z0=5.0, boundary_type='reflective'))
# Define cells for each material
model.geometry.root_universe = openmc.Universe(name='root universe')
xy = +left & -right & +bottom & -top
for i, mat in enumerate(model.materials):
c = openmc.Cell(fill=mat, region=xy & +planes[i] & -planes[i + 1])
model.geometry.root_universe.add_cell(c)
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[0.0, 0.0, 0.0], [10.0, 10.0, 5.]))
model.settings.energy_mode = "multi-group"
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (5.0, 5.0, 2.5)
plot.width = (2.5, 2.5)
plot.basis = 'xz'
plot.pixels = (3000, 3000)
plot.color_by = 'material'
model.plots.append(plot)
return model
