Added hexagonal example for #331

examples/python/lattice/hexagonal/build-xml.py

@@ -0,0 +1,175 @@
+import openmc
+
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 20
+inactive = 10
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml File
+###############################################################################
+
+# Instantiate some Nuclides
+h1 = openmc.Nuclide('H-1')
+o16 = openmc.Nuclide('O-16')
+u235 = openmc.Nuclide('U-235')
+
+# Instantiate some Materials and register the appropriate Nuclides
+fuel = openmc.Material(material_id=1, name='fuel')
+fuel.set_density('g/cc', 4.5)
+fuel.add_nuclide(u235, 1.)
+
+moderator = openmc.Material(material_id=2, name='moderator')
+moderator.set_density('g/cc', 1.0)
+moderator.add_nuclide(h1, 2.)
+moderator.add_nuclide(o16, 1.)
+moderator.add_s_alpha_beta('HH2O', '71t')
+
+# Instantiate a MaterialsFile, register all Materials, and export to XML
+materials_file = openmc.MaterialsFile()
+materials_file.set_default_xs('71c')
+materials_file.add_materials([moderator, fuel])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml File
+###############################################################################
+
+# Instantiate Surfaces
+left = openmc.XPlane(surface_id=1, x0=-2, name='left')
+right = openmc.XPlane(surface_id=2, x0=2, name='right')
+bottom = openmc.YPlane(surface_id=3, y0=-2, name='bottom')
+top = openmc.YPlane(surface_id=4, y0=2, name='top')
+fuel1 = openmc.ZCylinder(surface_id=5, x0=0, y0=0, R=0.4)
+fuel2 = openmc.ZCylinder(surface_id=6, x0=0, y0=0, R=0.3)
+fuel3 = openmc.ZCylinder(surface_id=7, x0=0, y0=0, R=0.2)
+
+left.set_boundary_type('vacuum')
+right.set_boundary_type('vacuum')
+top.set_boundary_type('vacuum')
+bottom.set_boundary_type('vacuum')
+
+# Instantiate Cells
+cell1 = openmc.Cell(cell_id=1, name='Cell 1')
+cell2 = openmc.Cell(cell_id=101, name='cell 2')
+cell3 = openmc.Cell(cell_id=102, name='cell 3')
+cell4 = openmc.Cell(cell_id=201, name='cell 4')
+cell5 = openmc.Cell(cell_id=202, name='cell 5')
+cell6 = openmc.Cell(cell_id=301, name='cell 6')
+cell7 = openmc.Cell(cell_id=302, name='cell 7')
+
+# Register Surfaces with Cells
+cell1.add_surface(left, halfspace=+1)
+cell1.add_surface(right, halfspace=-1)
+cell1.add_surface(bottom, halfspace=+1)
+cell1.add_surface(top, halfspace=-1)
+cell2.add_surface(fuel1, halfspace=-1)
+cell3.add_surface(fuel1, halfspace=+1)
+cell4.add_surface(fuel2, halfspace=-1)
+cell5.add_surface(fuel2, halfspace=+1)
+cell6.add_surface(fuel3, halfspace=-1)
+cell7.add_surface(fuel3, halfspace=+1)
+
+# Register Materials with Cells
+cell2.set_fill(fuel)
+cell3.set_fill(moderator)
+cell4.set_fill(fuel)
+cell5.set_fill(moderator)
+cell6.set_fill(fuel)
+cell7.set_fill(moderator)
+
+# Instantiate Universe
+univ1 = openmc.Universe(universe_id=1)
+univ2 = openmc.Universe(universe_id=2)
+univ3 = openmc.Universe(universe_id=3)
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cells with Universe
+univ1.add_cells([cell2, cell3])
+univ2.add_cells([cell4, cell5])
+univ3.add_cells([cell6, cell7])
+root.add_cell(cell1)
+
+# Instantiate a Lattice
+lattice = openmc.HexLattice(lattice_id=5)
+lattice.set_num_rings(2)
+lattice.set_center([0., 0.])
+lattice.set_pitch([1.])
+lattice.set_universes([
+     [univ2, univ2, univ2, univ2, univ2, univ2],
+     [univ1]])
+
+# Fill Cell with the Lattice
+cell1.set_fill(lattice)
+
+# Instantiate a Geometry and register the root Universe
+geometry = openmc.Geometry()
+geometry.set_root_universe(root)
+
+# Instantiate a GeometryFile, register Geometry, and export to XML
+geometry_file = openmc.GeometryFile()
+geometry_file.set_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.set_batches(batches)
+settings_file.set_inactive(inactive)
+settings_file.set_particles(particles)
+settings_file.set_source_space('box', [-1, -1, -1, 1, 1, 1])
+settings_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC plots.xml File
+###############################################################################
+
+plot = openmc.Plot(plot_id=1)
+plot.set_origin([0, 0, 0])
+plot.set_width([4, 4])
+plot.set_pixels([400, 400])
+plot.set_color('mat')
+
+# Instantiate a PlotsFile, add Plot, and export to XML
+plot_file = openmc.PlotsFile()
+plot_file.add_plot(plot)
+plot_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml File
+###############################################################################
+
+# Instantiate a tally mesh
+mesh = openmc.Mesh(mesh_id=1)
+mesh.set_type('rectangular')
+mesh.set_dimension([4, 4])
+mesh.set_lower_left([-2, -2])
+mesh.set_width([1, 1])
+
+# Instantiate tally Filter
+mesh_filter = openmc.Filter()
+mesh_filter.set_mesh(mesh)
+
+# Instantiate the Tally
+tally = openmc.Tally(tally_id=1)
+tally.add_filter(mesh_filter)
+tally.add_score('total')
+
+# Instantiate a TalliesFile, register Tally/Mesh, and export to XML
+tallies_file = openmc.TalliesFile()
+tallies_file.add_mesh(mesh)
+tallies_file.add_tally(tally)
+tallies_file.export_to_xml()