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Implemented pincell example file using Python API
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Will Boyd
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368184371e
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5 changed files with 214 additions and 145 deletions
214
examples/python/pincell/build-xml.py
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examples/python/pincell/build-xml.py
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import openmc
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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# OpenMC simulation parameters
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batches = 100
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inactive = 10
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particles = 1000
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###############################################################################
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# Exporting to OpenMC materials.xml File
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###############################################################################
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# Instantiate some Nuclides
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h1 = openmc.Nuclide('H-1')
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h2 = openmc.Nuclide('H-2')
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he4 = openmc.Nuclide('He-4')
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b10 = openmc.Nuclide('B-10')
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b11 = openmc.Nuclide('B-11')
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o16 = openmc.Nuclide('O-16')
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o17 = openmc.Nuclide('O-17')
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cr50 = openmc.Nuclide('Cr-50')
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cr52 = openmc.Nuclide('Cr-52')
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cr53 = openmc.Nuclide('Cr-53')
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cr54 = openmc.Nuclide('Cr-54')
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fe54 = openmc.Nuclide('Fe-54')
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fe56 = openmc.Nuclide('Fe-56')
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fe57 = openmc.Nuclide('Fe-57')
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fe58 = openmc.Nuclide('Fe-58')
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zr90 = openmc.Nuclide('Zr-90')
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zr91 = openmc.Nuclide('Zr-91')
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zr92 = openmc.Nuclide('Zr-92')
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zr94 = openmc.Nuclide('Zr-94')
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zr96 = openmc.Nuclide('Zr-96')
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sn112 = openmc.Nuclide('Sn-112')
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sn114 = openmc.Nuclide('Sn-114')
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sn115 = openmc.Nuclide('Sn-115')
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sn116 = openmc.Nuclide('Sn-116')
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sn117 = openmc.Nuclide('Sn-117')
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sn118 = openmc.Nuclide('Sn-118')
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sn119 = openmc.Nuclide('Sn-119')
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sn120 = openmc.Nuclide('Sn-120')
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sn122 = openmc.Nuclide('Sn-122')
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sn124 = openmc.Nuclide('Sn-124')
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u234 = openmc.Nuclide('U-234')
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u235 = openmc.Nuclide('U-235')
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u238 = openmc.Nuclide('U-238')
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# Instantiate some Materials and register the appropriate Nuclides
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uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
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uo2.setDensity('g/cm3', 10.29769)
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uo2.addNuclide(u234, 4.4843e-6)
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uo2.addNuclide(u235, 5.5815e-4)
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uo2.addNuclide(u238, 2.2408e-2)
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uo2.addNuclide(o16, 4.5829e-2)
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uo2.addNuclide(o17, 1.1164e-4)
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helium = openmc.Material(material_id=2, name='Helium for gap')
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helium.setDensity('g/cm3', 0.001598)
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helium.addNuclide(he4, 2.4044e-4)
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zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
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zircaloy.setDensity('g/cm3', 6.55)
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zircaloy.addNuclide(o16, 3.0743e-4)
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zircaloy.addNuclide(o17, 7.4887e-7)
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zircaloy.addNuclide(cr50, 3.2962e-6)
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zircaloy.addNuclide(cr52, 6.3564e-5)
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zircaloy.addNuclide(cr53, 7.2076e-6)
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zircaloy.addNuclide(cr54, 1.7941e-6)
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zircaloy.addNuclide(fe54, 8.6699e-6)
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zircaloy.addNuclide(fe56, 1.3610e-4)
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zircaloy.addNuclide(fe57, 3.1431e-6)
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zircaloy.addNuclide(fe58, 4.1829e-7)
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zircaloy.addNuclide(zr90, 2.1827e-2)
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zircaloy.addNuclide(zr91, 4.7600e-3)
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zircaloy.addNuclide(zr92, 7.2758e-3)
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zircaloy.addNuclide(zr94, 7.3734e-3)
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zircaloy.addNuclide(zr96, 1.1879e-3)
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zircaloy.addNuclide(sn112, 4.6735e-6)
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zircaloy.addNuclide(sn114, 3.1799e-6)
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zircaloy.addNuclide(sn115, 1.6381e-6)
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zircaloy.addNuclide(sn116, 7.0055e-5)
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zircaloy.addNuclide(sn117, 3.7003e-5)
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zircaloy.addNuclide(sn118, 1.1669e-4)
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zircaloy.addNuclide(sn119, 4.1387e-5)
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zircaloy.addNuclide(sn120, 1.5697e-4)
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zircaloy.addNuclide(sn122, 2.2308e-5)
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zircaloy.addNuclide(sn124, 2.7897e-5)
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borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
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borated_water.setDensity('g/cm3', 0.740582)
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borated_water.addNuclide(b10, 8.0042e-6)
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borated_water.addNuclide(b11, 3.2218e-5)
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borated_water.addNuclide(h1, 4.9457e-2)
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borated_water.addNuclide(h2, 7.4196e-6)
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borated_water.addNuclide(o16, 2.4672e-2)
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borated_water.addNuclide(o17, 6.0099e-5)
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borated_water.addSAlphaBeta('HH2O', '71t')
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# Instantiate a MaterialsFile, register all Materials, and export to XML
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materials_file = openmc.MaterialsFile()
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materials_file.setDefaultXS('71c')
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materials_file.addMaterials([uo2, helium, zircaloy, borated_water])
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materials_file.exportToXML()
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###############################################################################
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# Exporting to OpenMC geometry.xml File
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###############################################################################
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# Instantiate ZCylinder surfaces
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fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, R=0.39218, name='Fuel OR')
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clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, R=0.40005, name='Clad IR')
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clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, R=0.45720, name='Clad OR')
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left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
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right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
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bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
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top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
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left.setBoundaryType('reflective')
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right.setBoundaryType('reflective')
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top.setBoundaryType('reflective')
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bottom.setBoundaryType('reflective')
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# Instantiate Cells
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fuel = openmc.Cell(cell_id=1, name='cell 1')
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gap = openmc.Cell(cell_id=2, name='cell 2')
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clad = openmc.Cell(cell_id=3, name='cell 3')
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water = openmc.Cell(cell_id=4, name='cell 4')
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# Register Surfaces with Cells
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fuel.addSurface(fuel_or, halfspace=-1)
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gap.addSurface(fuel_or, halfspace=+1)
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gap.addSurface(clad_ir, halfspace=-1)
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clad.addSurface(clad_ir, halfspace=+1)
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clad.addSurface(clad_or, halfspace=-1)
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water.addSurface(clad_or, halfspace=+1)
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water.addSurface(left, halfspace=+1)
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water.addSurface(right, halfspace=-1)
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water.addSurface(bottom, halfspace=+1)
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water.addSurface(top, halfspace=-1)
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# Register Materials with Cells
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fuel.setFill(uo2)
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gap.setFill(helium)
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clad.setFill(zircaloy)
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water.setFill(borated_water)
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# Instantiate Universe
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root = openmc.Universe(universe_id=0, name='root universe')
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# Register Cells with Universe
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root.addCells([fuel, gap, clad, water])
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# Instantiate a Geometry and register the root Universe
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geometry = openmc.Geometry()
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geometry.setRootUniverse(root)
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# Instantiate a GeometryFile, register Geometry, and export to XML
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geometry_file = openmc.GeometryFile()
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geometry_file.setGeometry(geometry)
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geometry_file.exportToXML()
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###############################################################################
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# Exporting to OpenMC settings.xml File
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###############################################################################
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# Instantiate a SettingsFile, set all runtime parameters, and export to XML
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settings_file = openmc.SettingsFile()
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settings_file.setBatches(batches)
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settings_file.setInactive(inactive)
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settings_file.setParticles(particles)
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settings_file.setSourceSpace('box', [-0.62992, -0.62992, -1, \
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0.62992, 0.62992, 1])
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settings_file.setEntropyLowerLeft([-0.39218, -0.39218, -1.e50])
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settings_file.setEntropyUpperRight([0.39218, 0.39218, 1.e50])
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settings_file.setEntropyDimension([10, 10, 1])
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settings_file.exportToXML()
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###############################################################################
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# Exporting to OpenMC tallies.xml File
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###############################################################################
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# Instantiate a tally mesh
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mesh = openmc.Mesh(mesh_id=1)
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mesh.setType('rectangular')
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mesh.setDimension([100, 100, 1])
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mesh.setLowerLeft([-0.62992, -0.62992, -1.e50])
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mesh.setUpperRight([0.62992, 0.62992, 1.e50])
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# Instantiate some tally Filters
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energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.])
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mesh_filter = openmc.Filter()
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mesh_filter.setMesh(mesh)
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# Instantiate the Tally
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tally = openmc.Tally(tally_id=1)
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tally.addFilter(energy_filter)
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tally.addFilter(mesh_filter)
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tally.addScore('flux')
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tally.addScore('fission')
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tally.addScore('nu-fission')
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# Instantiate a TalliesFile, register all Tallies, and export to XML
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tallies_file = openmc.TalliesFile()
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tallies_file.addMesh(mesh)
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tallies_file.addTally(tally)
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tallies_file.exportToXML()
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