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Implemented lattice/nested example input file using Python API
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Will Boyd
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6 changed files with 189 additions and 108 deletions
189
examples/python/lattice/nested/build-xml.py
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189
examples/python/lattice/nested/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 = 20
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inactive = 10
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particles = 10000
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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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o16 = openmc.Nuclide('O-16')
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u235 = openmc.Nuclide('U-235')
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# Instantiate some Materials and register the appropriate Nuclides
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fuel = openmc.Material(material_id=1, name='fuel')
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fuel.setDensity('g/cc', 4.5)
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fuel.addNuclide(u235, 1.)
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moderator = openmc.Material(material_id=2, name='moderator')
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moderator.setDensity('g/cc', 1.0)
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moderator.addNuclide(h1, 2.)
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moderator.addNuclide(o16, 1.)
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moderator.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([moderator, fuel])
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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 Surfaces
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left = openmc.XPlane(surface_id=1, x0=-2, name='left')
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right = openmc.XPlane(surface_id=2, x0=2, name='right')
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bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
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top = openmc.YPlane(surface_id=4, y0=2, name='top')
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fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
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fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
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fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
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left.setBoundaryType('vacuum')
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right.setBoundaryType('vacuum')
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top.setBoundaryType('vacuum')
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bottom.setBoundaryType('vacuum')
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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cell2 = openmc.Cell(cell_id=2, name='Cell 2')
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cell3 = openmc.Cell(cell_id=101, name='cell 3')
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cell4 = openmc.Cell(cell_id=102, name='cell 4')
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cell5 = openmc.Cell(cell_id=201, name='cell 5')
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cell6 = openmc.Cell(cell_id=202, name='cell 6')
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cell7 = openmc.Cell(cell_id=301, name='cell 7')
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cell8 = openmc.Cell(cell_id=302, name='cell 8')
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# Register Surfaces with Cells
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cell1.addSurface(left, halfspace=+1)
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cell1.addSurface(right, halfspace=-1)
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cell1.addSurface(bottom, halfspace=+1)
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cell1.addSurface(top, halfspace=-1)
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cell2.addSurface(left, halfspace=+1)
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cell2.addSurface(right, halfspace=-1)
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cell2.addSurface(bottom, halfspace=+1)
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cell2.addSurface(top, halfspace=-1)
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cell3.addSurface(fuel1, halfspace=-1)
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cell4.addSurface(fuel1, halfspace=+1)
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cell5.addSurface(fuel2, halfspace=-1)
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cell6.addSurface(fuel2, halfspace=+1)
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cell7.addSurface(fuel3, halfspace=-1)
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cell8.addSurface(fuel3, halfspace=+1)
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# Register Materials with Cells
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cell3.setFill(fuel)
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cell4.setFill(moderator)
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cell5.setFill(fuel)
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cell6.setFill(moderator)
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cell7.setFill(fuel)
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cell8.setFill(moderator)
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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univ2 = openmc.Universe(universe_id=2)
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univ3 = openmc.Universe(universe_id=3)
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univ4 = openmc.Universe(universe_id=5)
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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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univ1.addCells([cell3, cell4])
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univ2.addCells([cell5, cell6])
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univ3.addCells([cell7, cell8])
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root.addCell(cell1)
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univ4.addCell(cell2)
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# Instantiate nested Lattices
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lattice1 = openmc.Lattice(lattice_id=4, name='4x4 assembly')
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lattice1.setDimension([2, 2])
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lattice1.setLowerLeft([-1., -1.])
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lattice1.setWidth([1., 1.])
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lattice1.setUniverses([[univ1, univ2],
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[univ2, univ3]])
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lattice2 = openmc.Lattice(lattice_id=6, name='4x4 core')
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lattice2.setDimension([2, 2])
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lattice2.setLowerLeft([-2., -2.])
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lattice2.setWidth([2., 2.])
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lattice2.setUniverses([[univ4, univ4],
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[univ4, univ4]])
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# Fill Cell with the Lattice
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cell1.setFill(lattice2)
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cell2.setFill(lattice1)
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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', [-1, -1, -1, 1, 1, 1])
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settings_file.exportToXML()
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###############################################################################
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# Exporting to OpenMC plots.xml File
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###############################################################################
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plot = openmc.Plot(plot_id=1)
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plot.setOrigin([0, 0, 0])
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plot.setWidth([4, 4])
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plot.setPixels([400, 400])
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plot.setColor('mat')
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# Instantiate a PlotsFile, add Plot, and export to XML
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plot_file = openmc.PlotsFile()
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plot_file.addPlot(plot)
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plot_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([4, 4])
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mesh.setLowerLeft([-2, -2])
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mesh.setWidth([1, 1])
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# Instantiate tally Filter
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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(mesh_filter)
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tally.addScore('total')
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# Instantiate a TalliesFile, register Tally/Mesh, 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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@ -1,45 +0,0 @@
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<?xml version="1.0"?>
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<geometry>
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<cell id="1" fill="6" surfaces="1 -2 3 -4" />
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<cell id="2" universe="5" fill="4" surfaces="1 -2 3 -4" />
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<cell id="101" universe="1" material="1" surfaces="-5" />
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<cell id="102" universe="1" material="2" surfaces="5" />
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<cell id="201" universe="2" material="1" surfaces="-6" />
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<cell id="202" universe="2" material="2" surfaces="6" />
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<cell id="301" universe="3" material="1" surfaces="-7" />
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<cell id="302" universe="3" material="2" surfaces="7" />
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<!-- 4 x 4 assembly -->
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<lattice id="4">
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<type>rectangular</type>
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<dimension>2 2</dimension>
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<lower_left>-1.0 -1.0</lower_left>
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<width>1.0 1.0</width>
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<universes>
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1 2
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2 3
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</universes>
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</lattice>
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<!-- 4 x 4 core -->
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<lattice id="6">
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<type>rectangular</type>
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<dimension>2 2</dimension>
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<lower_left>-2.0 -2.0</lower_left>
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<width>2.0 2.0</width>
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<universes>
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5 5
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5 5
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</universes>
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</lattice>
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<surface id="1" type="x-plane" coeffs="-2.0" boundary="vacuum" />
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<surface id="2" type="x-plane" coeffs="2.0" boundary="vacuum" />
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<surface id="3" type="y-plane" coeffs="-2.0" boundary="vacuum" />
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<surface id="4" type="y-plane" coeffs="2.0" boundary="vacuum" />
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<surface id="5" type="z-cylinder" coeffs="0.0 0.0 0.4" />
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<surface id="6" type="z-cylinder" coeffs="0.0 0.0 0.3" />
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<surface id="7" type="z-cylinder" coeffs="0.0 0.0 0.2" />
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</geometry>
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<?xml version="1.0"?>
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<materials>
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<default_xs>71c</default_xs>
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<!-- Definition of materials -->
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<material id="1">
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<density value="4.5" units="g/cc" />
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<nuclide name="U-235" ao="1.0" />
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</material>
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<material id="2">
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<density value="1.0" units="g/cc" />
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<nuclide name="H-1" ao="2.0" />
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<nuclide name="O-16" ao="1.0" />
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<sab name="HH2O" xs="71t" />
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</material>
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</materials>
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<?xml version="1.0"?>
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<plots>
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<plot id="1" color="mat">
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<origin>0. 0. 0.</origin>
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<width>4.0 4.0</width>
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<pixels>400 400</pixels>
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</plot>
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</plots>
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<?xml version="1.0"?>
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<settings>
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<!-- Parameters for k-eigenvalue calculation -->
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<eigenvalue>
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<batches>20</batches>
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<inactive>10</inactive>
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<particles>10000</particles>
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</eigenvalue>
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<!-- Starting source -->
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<source>
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<space type="box">
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<parameters>-1 -1 -1 1 1 1</parameters>
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</space>
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</source>
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</settings>
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<?xml version="1.0"?>
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<tallies>
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<mesh id="1">
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<type>rectangular</type>
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<dimension>4 4</dimension>
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<lower_left>-2.0 -2.0</lower_left>
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<width>1.0 1.0</width>
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</mesh>
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<tally id="1">
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<filter type="mesh" bins="1" />
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<scores>total</scores>
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</tally>
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</tallies>
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