adding back files to be reviewed

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

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+import openmc
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 20
+inactive = 10
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+# 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_element('H', 2.)
+moderator.add_element('O', 1.)
+moderator.add_s_alpha_beta('c_H_in_H2O')
+
+# Instantiate a Materials collection and export to XML
+materials_file = openmc.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.boundary_type = 'vacuum'
+right.boundary_type = 'vacuum'
+top.boundary_type = 'vacuum'
+bottom.boundary_type = 'vacuum'
+
+# Instantiate Cells
+cell1 = openmc.Cell(cell_id=1, name='Cell 1')
+cell2 = openmc.Cell(cell_id=2, name='Cell 2')
+cell3 = openmc.Cell(cell_id=101, name='cell 3')
+cell4 = openmc.Cell(cell_id=102, name='cell 4')
+cell5 = openmc.Cell(cell_id=201, name='cell 5')
+cell6 = openmc.Cell(cell_id=202, name='cell 6')
+cell7 = openmc.Cell(cell_id=301, name='cell 7')
+cell8 = openmc.Cell(cell_id=302, name='cell 8')
+
+# Use surface half-space to define regions
+cell1.region = +left & -right & +bottom & -top
+cell2.region = +left & -right & +bottom & -top
+cell3.region = -fuel1
+cell4.region = +fuel1
+cell5.region = -fuel2
+cell6.region = +fuel2
+cell7.region = -fuel3
+cell8.region = +fuel3
+
+# Register Materials with Cells
+cell3.fill = fuel
+cell4.fill = moderator
+cell5.fill = fuel
+cell6.fill = moderator
+cell7.fill = fuel
+cell8.fill = moderator
+
+# Instantiate Universe
+univ1 = openmc.Universe(universe_id=1)
+univ2 = openmc.Universe(universe_id=2)
+univ3 = openmc.Universe(universe_id=3)
+univ4 = openmc.Universe(universe_id=5)
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cells with Universe
+univ1.add_cells([cell3, cell4])
+univ2.add_cells([cell5, cell6])
+univ3.add_cells([cell7, cell8])
+root.add_cell(cell1)
+univ4.add_cell(cell2)
+
+# Instantiate nested Lattices
+lattice1 = openmc.RectLattice(lattice_id=4, name='4x4 assembly')
+lattice1.lower_left = [-1., -1.]
+lattice1.pitch = [1., 1.]
+lattice1.universes = [[univ1, univ2],
+                      [univ2, univ3]]
+
+lattice2 = openmc.RectLattice(lattice_id=6, name='4x4 core')
+lattice2.lower_left = [-2., -2.]
+lattice2.pitch = [2., 2.]
+lattice2.universes = [[univ4, univ4],
+                      [univ4, univ4]]
+
+# Fill Cell with the Lattice
+cell1.fill = lattice2
+cell2.fill = lattice1
+
+# Instantiate a Geometry, register the root Universe, and export to XML
+geometry = openmc.Geometry(root)
+geometry.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC settings.xml file
+###############################################################################
+
+# Instantiate a Settings object, set all runtime parameters, and export to XML
+settings_file = openmc.Settings()
+settings_file.batches = batches
+settings_file.inactive = inactive
+settings_file.particles = particles
+
+# Create an initial uniform spatial source distribution over fissionable zones
+bounds = [-1, -1, -1, 1, 1, 1]
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+settings_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC plots.xml file
+###############################################################################
+
+plot = openmc.Plot(plot_id=1)
+plot.origin = [0, 0, 0]
+plot.width = [4, 4]
+plot.pixels = [400, 400]
+plot.color_by = 'material'
+
+# Instantiate a Plots object and export to XML
+plot_file = openmc.Plots([plot])
+plot_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml file
+###############################################################################
+
+# Instantiate a tally mesh
+mesh = openmc.RegularMesh(mesh_id=1)
+mesh.dimension = [4, 4]
+mesh.lower_left = [-2, -2]
+mesh.width = [1, 1]
+
+# Instantiate tally Filter
+mesh_filter = openmc.MeshFilter(mesh)
+
+# Instantiate the Tally
+tally = openmc.Tally(tally_id=1)
+tally.filters = [mesh_filter]
+tally.scores = ['total']
+
+# Instantiate a Tallies collection, register Tally/RegularMesh, and export to
+# XML
+tallies_file = openmc.Tallies([tally])
+tallies_file.export_to_xml()