adding back files to be reviewed

examples/python/lattice/hexagonal/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')
+
+iron = openmc.Material(material_id=3, name='iron')
+iron.set_density('g/cc', 7.9)
+iron.add_element('Fe', 1.)
+
+# Instantiate a Materials collection and export to XML
+materials_file = openmc.Materials([moderator, fuel, iron])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml file
+###############################################################################
+
+# Instantiate Surfaces
+left = openmc.XPlane(surface_id=1, x0=-3, name='left')
+right = openmc.XPlane(surface_id=2, x0=3, name='right')
+bottom = openmc.YPlane(surface_id=3, y0=-4, name='bottom')
+top = openmc.YPlane(surface_id=4, y0=4, name='top')
+fuel_surf = openmc.ZCylinder(surface_id=5, x0=0, y0=0, r=0.4)
+
+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=101, name='cell 2')
+cell3 = openmc.Cell(cell_id=102, name='cell 3')
+cell4 = openmc.Cell(cell_id=500, name='cell 4')
+cell5 = openmc.Cell(cell_id=600, name='cell 5')
+cell6 = openmc.Cell(cell_id=601, name='cell 6')
+
+# Use surface half-spaces to define regions
+cell1.region = +left & -right & +bottom & -top
+cell2.region = -fuel_surf
+cell3.region = +fuel_surf
+cell5.region = -fuel_surf
+cell6.region = +fuel_surf
+
+# Register Materials with Cells
+cell2.fill = fuel
+cell3.fill = moderator
+cell4.fill = moderator
+cell5.fill = iron
+cell6.fill = moderator
+
+# Instantiate Universe
+univ1 = openmc.Universe(universe_id=1)
+univ2 = openmc.Universe(universe_id=3)
+univ3 = openmc.Universe(universe_id=4)
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cells with Universe
+univ1.add_cells([cell2, cell3])
+univ2.add_cells([cell4])
+univ3.add_cells([cell5, cell6])
+root.add_cell(cell1)
+
+# Instantiate a Lattice
+lattice = openmc.HexLattice(lattice_id=5)
+lattice.center = [0., 0., 0.]
+lattice.pitch = [1., 2.]
+lattice.universes = \
+    [ [ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ],
+      [ [univ2] + [univ1]*11, [univ2] + [univ1]*5, [univ1] ],
+      [ [univ2] + [univ3]*11, [univ2] + [univ3]*5, [univ3] ] ]
+lattice.outer = univ2
+
+# Fill Cell with the Lattice
+cell1.fill = lattice
+
+# 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.keff_trigger = {'type' : 'std_dev', 'threshold' : 5E-4}
+settings_file.trigger_active = True
+settings_file.trigger_max_batches = 100
+settings_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC plots.xml file
+###############################################################################
+
+plot_xy = openmc.Plot(plot_id=1)
+plot_xy.filename = 'plot_xy'
+plot_xy.origin = [0, 0, 0]
+plot_xy.width = [6, 6]
+plot_xy.pixels = [400, 400]
+plot_xy.color_by = 'material'
+
+plot_yz = openmc.Plot(plot_id=2)
+plot_yz.filename = 'plot_yz'
+plot_yz.basis = 'yz'
+plot_yz.origin = [0, 0, 0]
+plot_yz.width = [8, 8]
+plot_yz.pixels = [400, 400]
+plot_yz.color_by = 'material'
+
+# Instantiate a Plots collection, add plots, and export to XML
+plot_file = openmc.Plots((plot_xy, plot_yz))
+plot_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml File
+###############################################################################
+
+# Instantiate a distribcell Tally
+tally = openmc.Tally(tally_id=1)
+tally.filters = [openmc.DistribcellFilter(cell2)]
+tally.scores = ['total']
+
+# Instantiate a Tallies collection and export to XML
+tallies_file = openmc.Tallies([tally])
+tallies_file.export_to_xml()