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Updated sample input scripts for Python API to conform to new API specs
This commit is contained in:
Will Boyd
parent
2767d3f53f
commit
3bb1f8faf7
8 changed files with 291 additions and 330 deletions
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@ -22,20 +22,20 @@ 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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fuel.set_density('g/cc', 4.5)
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fuel.add_nuclide(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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moderator.set_density('g/cc', 1.0)
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moderator.add_nuclide(h1, 2.)
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moderator.add_nuclide(o16, 1.)
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moderator.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([moderator, fuel])
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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([moderator, fuel])
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materials_file.export_to_xml()
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###############################################################################
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@ -51,10 +51,10 @@ 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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left.set_boundary_type('vacuum')
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right.set_boundary_type('vacuum')
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top.set_boundary_type('vacuum')
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bottom.set_boundary_type('vacuum')
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# Instantiate Cells
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cell1 = openmc.Cell(cell_id=1, name='Cell 1')
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@ -67,28 +67,28 @@ 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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cell1.add_surface(left, halfspace=+1)
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cell1.add_surface(right, halfspace=-1)
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cell1.add_surface(bottom, halfspace=+1)
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cell1.add_surface(top, halfspace=-1)
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cell2.add_surface(left, halfspace=+1)
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cell2.add_surface(right, halfspace=-1)
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cell2.add_surface(bottom, halfspace=+1)
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cell2.add_surface(top, halfspace=-1)
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cell3.add_surface(fuel1, halfspace=-1)
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cell4.add_surface(fuel1, halfspace=+1)
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cell5.add_surface(fuel2, halfspace=-1)
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cell6.add_surface(fuel2, halfspace=+1)
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cell7.add_surface(fuel3, halfspace=-1)
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cell8.add_surface(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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cell3.set_fill(fuel)
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cell4.set_fill(moderator)
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cell5.set_fill(fuel)
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cell6.set_fill(moderator)
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cell7.set_fill(fuel)
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cell8.set_fill(moderator)
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# Instantiate Universe
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univ1 = openmc.Universe(universe_id=1)
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@ -98,39 +98,39 @@ 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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univ1.add_cells([cell3, cell4])
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univ2.add_cells([cell5, cell6])
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univ3.add_cells([cell7, cell8])
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root.add_cell(cell1)
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univ4.add_cell(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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lattice1.set_dimension([2, 2])
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lattice1.set_lower_left([-1., -1.])
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lattice1.set_width([1., 1.])
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lattice1.set_universes([[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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lattice2.set_dimension([2, 2])
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lattice2.set_lower_left([-2., -2.])
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lattice2.set_width([2., 2.])
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lattice2.set_universes([[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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cell1.set_fill(lattice2)
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cell2.set_fill(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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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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@ -139,11 +139,11 @@ 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', [-1, -1, -1, 1, 1, 1])
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settings_file.exportToXML()
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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', [-1, -1, -1, 1, 1, 1])
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settings_file.export_to_xml()
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###############################################################################
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@ -151,15 +151,15 @@ settings_file.exportToXML()
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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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plot.set_origin([0, 0, 0])
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plot.set_width([4, 4])
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plot.set_pixels([400, 400])
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plot.set_color('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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plot_file.add_plot(plot)
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plot_file.export_to_xml()
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###############################################################################
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@ -168,22 +168,22 @@ plot_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([4, 4])
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mesh.setLowerLeft([-2, -2])
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mesh.setWidth([1, 1])
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mesh.set_type('rectangular')
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mesh.set_dimension([4, 4])
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mesh.set_lower_left([-2, -2])
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mesh.set_width([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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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(mesh_filter)
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tally.addScore('total')
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tally.add_filter(mesh_filter)
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tally.add_score('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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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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