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Updated sample input scripts for Python API to conform to new API specs
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Will Boyd
parent
2767d3f53f
commit
3bb1f8faf7
8 changed files with 291 additions and 330 deletions
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@ -52,60 +52,60 @@ 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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uo2.set_density('g/cm3', 10.29769)
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uo2.add_nuclide(u234, 4.4843e-6)
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uo2.add_nuclide(u235, 5.5815e-4)
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uo2.add_nuclide(u238, 2.2408e-2)
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uo2.add_nuclide(o16, 4.5829e-2)
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uo2.add_nuclide(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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helium.set_density('g/cm3', 0.001598)
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helium.add_nuclide(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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zircaloy.set_density('g/cm3', 6.55)
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zircaloy.add_nuclide(o16, 3.0743e-4)
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zircaloy.add_nuclide(o17, 7.4887e-7)
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zircaloy.add_nuclide(cr50, 3.2962e-6)
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zircaloy.add_nuclide(cr52, 6.3564e-5)
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zircaloy.add_nuclide(cr53, 7.2076e-6)
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zircaloy.add_nuclide(cr54, 1.7941e-6)
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zircaloy.add_nuclide(fe54, 8.6699e-6)
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zircaloy.add_nuclide(fe56, 1.3610e-4)
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zircaloy.add_nuclide(fe57, 3.1431e-6)
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zircaloy.add_nuclide(fe58, 4.1829e-7)
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zircaloy.add_nuclide(zr90, 2.1827e-2)
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zircaloy.add_nuclide(zr91, 4.7600e-3)
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zircaloy.add_nuclide(zr92, 7.2758e-3)
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zircaloy.add_nuclide(zr94, 7.3734e-3)
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zircaloy.add_nuclide(zr96, 1.1879e-3)
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zircaloy.add_nuclide(sn112, 4.6735e-6)
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zircaloy.add_nuclide(sn114, 3.1799e-6)
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zircaloy.add_nuclide(sn115, 1.6381e-6)
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zircaloy.add_nuclide(sn116, 7.0055e-5)
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zircaloy.add_nuclide(sn117, 3.7003e-5)
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zircaloy.add_nuclide(sn118, 1.1669e-4)
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zircaloy.add_nuclide(sn119, 4.1387e-5)
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zircaloy.add_nuclide(sn120, 1.5697e-4)
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zircaloy.add_nuclide(sn122, 2.2308e-5)
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zircaloy.add_nuclide(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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borated_water.set_density('g/cm3', 0.740582)
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borated_water.add_nuclide(b10, 8.0042e-6)
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borated_water.add_nuclide(b11, 3.2218e-5)
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borated_water.add_nuclide(h1, 4.9457e-2)
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borated_water.add_nuclide(h2, 7.4196e-6)
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borated_water.add_nuclide(o16, 2.4672e-2)
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borated_water.add_nuclide(o17, 6.0099e-5)
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borated_water.add_s_alpha_beta('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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materials_file.set_default_xs('71c')
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materials_file.add_materials([uo2, helium, zircaloy, borated_water])
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materials_file.export_to_xml()
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###############################################################################
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@ -121,10 +121,10 @@ 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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left.set_boundary_type('reflective')
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right.set_boundary_type('reflective')
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top.set_boundary_type('reflective')
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bottom.set_boundary_type('reflective')
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# Instantiate Cells
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fuel = openmc.Cell(cell_id=1, name='cell 1')
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@ -133,37 +133,37 @@ 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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fuel.add_surface(fuel_or, halfspace=-1)
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gap.add_surface(fuel_or, halfspace=+1)
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gap.add_surface(clad_ir, halfspace=-1)
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clad.add_surface(clad_ir, halfspace=+1)
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clad.add_surface(clad_or, halfspace=-1)
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water.add_surface(clad_or, halfspace=+1)
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water.add_surface(left, halfspace=+1)
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water.add_surface(right, halfspace=-1)
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water.add_surface(bottom, halfspace=+1)
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water.add_surface(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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fuel.set_fill(uo2)
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gap.set_fill(helium)
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clad.set_fill(zircaloy)
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water.set_fill(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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root.add_cells([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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geometry.set_root_universe(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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geometry_file.set_geometry(geometry)
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geometry_file.export_to_xml()
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###############################################################################
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@ -172,15 +172,15 @@ geometry_file.exportToXML()
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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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settings_file.set_batches(batches)
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settings_file.set_inactive(inactive)
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settings_file.set_particles(particles)
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settings_file.set_source_space('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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settings_file.set_entropy_lower_left([-0.39218, -0.39218, -1.e50])
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settings_file.set_entropy_upper_right([0.39218, 0.39218, 1.e50])
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settings_file.set_entropy_dimension([10, 10, 1])
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settings_file.export_to_xml()
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###############################################################################
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@ -189,26 +189,26 @@ settings_file.exportToXML()
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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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mesh.set_type('rectangular')
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mesh.set_dimension([100, 100, 1])
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mesh.set_lower_left([-0.62992, -0.62992, -1.e50])
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mesh.set_upper_right([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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mesh_filter.set_mesh(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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tally.add_filter(energy_filter)
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tally.add_filter(mesh_filter)
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tally.add_score('flux')
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tally.add_score('fission')
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tally.add_score('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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tallies_file.add_mesh(mesh)
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tallies_file.add_tally(tally)
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tallies_file.export_to_xml()
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