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OpenMC/examples/python/pincell/build-xml.py
Will Boyd c738c73486 Changed label to name
2015-04-27 08:20:30 -04:00

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Python
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import openmc
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 100
inactive = 10
particles = 1000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
h2 = openmc.Nuclide('H-2')
he4 = openmc.Nuclide('He-4')
b10 = openmc.Nuclide('B-10')
b11 = openmc.Nuclide('B-11')
o16 = openmc.Nuclide('O-16')
o17 = openmc.Nuclide('O-17')
cr50 = openmc.Nuclide('Cr-50')
cr52 = openmc.Nuclide('Cr-52')
cr53 = openmc.Nuclide('Cr-53')
cr54 = openmc.Nuclide('Cr-54')
fe54 = openmc.Nuclide('Fe-54')
fe56 = openmc.Nuclide('Fe-56')
fe57 = openmc.Nuclide('Fe-57')
fe58 = openmc.Nuclide('Fe-58')
zr90 = openmc.Nuclide('Zr-90')
zr91 = openmc.Nuclide('Zr-91')
zr92 = openmc.Nuclide('Zr-92')
zr94 = openmc.Nuclide('Zr-94')
zr96 = openmc.Nuclide('Zr-96')
sn112 = openmc.Nuclide('Sn-112')
sn114 = openmc.Nuclide('Sn-114')
sn115 = openmc.Nuclide('Sn-115')
sn116 = openmc.Nuclide('Sn-116')
sn117 = openmc.Nuclide('Sn-117')
sn118 = openmc.Nuclide('Sn-118')
sn119 = openmc.Nuclide('Sn-119')
sn120 = openmc.Nuclide('Sn-120')
sn122 = openmc.Nuclide('Sn-122')
sn124 = openmc.Nuclide('Sn-124')
u234 = openmc.Nuclide('U-234')
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_nuclide(u234, 4.4843e-6)
uo2.add_nuclide(u235, 5.5815e-4)
uo2.add_nuclide(u238, 2.2408e-2)
uo2.add_nuclide(o16, 4.5829e-2)
uo2.add_nuclide(o17, 1.1164e-4)
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_nuclide(he4, 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_nuclide(o16, 3.0743e-4)
zircaloy.add_nuclide(o17, 7.4887e-7)
zircaloy.add_nuclide(cr50, 3.2962e-6)
zircaloy.add_nuclide(cr52, 6.3564e-5)
zircaloy.add_nuclide(cr53, 7.2076e-6)
zircaloy.add_nuclide(cr54, 1.7941e-6)
zircaloy.add_nuclide(fe54, 8.6699e-6)
zircaloy.add_nuclide(fe56, 1.3610e-4)
zircaloy.add_nuclide(fe57, 3.1431e-6)
zircaloy.add_nuclide(fe58, 4.1829e-7)
zircaloy.add_nuclide(zr90, 2.1827e-2)
zircaloy.add_nuclide(zr91, 4.7600e-3)
zircaloy.add_nuclide(zr92, 7.2758e-3)
zircaloy.add_nuclide(zr94, 7.3734e-3)
zircaloy.add_nuclide(zr96, 1.1879e-3)
zircaloy.add_nuclide(sn112, 4.6735e-6)
zircaloy.add_nuclide(sn114, 3.1799e-6)
zircaloy.add_nuclide(sn115, 1.6381e-6)
zircaloy.add_nuclide(sn116, 7.0055e-5)
zircaloy.add_nuclide(sn117, 3.7003e-5)
zircaloy.add_nuclide(sn118, 1.1669e-4)
zircaloy.add_nuclide(sn119, 4.1387e-5)
zircaloy.add_nuclide(sn120, 1.5697e-4)
zircaloy.add_nuclide(sn122, 2.2308e-5)
zircaloy.add_nuclide(sn124, 2.7897e-5)
borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_nuclide(b10, 8.0042e-6)
borated_water.add_nuclide(b11, 3.2218e-5)
borated_water.add_nuclide(h1, 4.9457e-2)
borated_water.add_nuclide(h2, 7.4196e-6)
borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_s_alpha_beta('HH2O', '71t')
# Instantiate a MaterialsFile, register all Materials, and export to XML
materials_file = openmc.MaterialsFile()
materials_file.default_xs = '71c'
materials_file.add_materials([uo2, helium, zircaloy, borated_water])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml File
###############################################################################
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
left.boundary_type = 'reflective'
right.boundary_type = 'reflective'
top.boundary_type = 'reflective'
bottom.boundary_type = 'reflective'
# Instantiate Cells
fuel = openmc.Cell(cell_id=1, name='cell 1')
gap = openmc.Cell(cell_id=2, name='cell 2')
clad = openmc.Cell(cell_id=3, name='cell 3')
water = openmc.Cell(cell_id=4, name='cell 4')
# Register Surfaces with Cells
fuel.add_surface(fuel_or, halfspace=-1)
gap.add_surface(fuel_or, halfspace=+1)
gap.add_surface(clad_ir, halfspace=-1)
clad.add_surface(clad_ir, halfspace=+1)
clad.add_surface(clad_or, halfspace=-1)
water.add_surface(clad_or, halfspace=+1)
water.add_surface(left, halfspace=+1)
water.add_surface(right, halfspace=-1)
water.add_surface(bottom, halfspace=+1)
water.add_surface(top, halfspace=-1)
# Register Materials with Cells
fuel.fill = uo2
gap.fill = helium
clad.fill = zircaloy
water.fill = borated_water
# Instantiate Universe
root = openmc.Universe(universe_id=0, name='root universe')
# Register Cells with Universe
root.add_cells([fuel, gap, clad, water])
# Instantiate a Geometry and register the root Universe
geometry = openmc.Geometry()
geometry.root_universe = root
# Instantiate a GeometryFile, register Geometry, and export to XML
geometry_file = openmc.GeometryFile()
geometry_file.geometry = geometry
geometry_file.export_to_xml()
###############################################################################
# Exporting to OpenMC settings.xml File
###############################################################################
# Instantiate a SettingsFile, set all runtime parameters, and export to XML
settings_file = openmc.SettingsFile()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles
settings_file.set_source_space('box', [-0.62992, -0.62992, -1, \
0.62992, 0.62992, 1])
settings_file.entropy_lower_left = [-0.39218, -0.39218, -1.e50]
settings_file.entropy_upper_right = [0.39218, 0.39218, 1.e50]
settings_file.entropy_dimension = [10, 10, 1]
settings_file.export_to_xml()
###############################################################################
# Exporting to OpenMC tallies.xml File
###############################################################################
# Instantiate a tally mesh
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'rectangular'
mesh.dimension = [100, 100, 1]
mesh.lower_left = [-0.62992, -0.62992, -1.e50]
mesh.upper_right = [0.62992, 0.62992, 1.e50]
# Instantiate some tally Filters
energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.])
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')
tally.add_filter(energy_filter)
tally.add_filter(mesh_filter)
tally.add_score('flux')
tally.add_score('fission')
tally.add_score('nu-fission')
# Instantiate a TalliesFile, register all Tallies, and export to XML
tallies_file = openmc.TalliesFile()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
tallies_file.export_to_xml()
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