adding back files to be reviewed

examples/python/basic/build-xml.py

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+import openmc
+
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 15
+inactive = 5
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+# Instantiate some Materials and register the appropriate Nuclides
+moderator = openmc.Material(material_id=41, 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')
+
+fuel = openmc.Material(material_id=40, name='fuel')
+fuel.set_density('g/cc', 4.5)
+fuel.add_nuclide('U235', 1.)
+
+# 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 ZCylinder surfaces
+surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=7, name='surf 1')
+surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=9, name='surf 2')
+surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=11, name='surf 3')
+surf3.boundary_type = 'vacuum'
+
+# Instantiate Cells
+cell1 = openmc.Cell(cell_id=1, name='cell 1')
+cell2 = openmc.Cell(cell_id=100, name='cell 2')
+cell3 = openmc.Cell(cell_id=101, name='cell 3')
+cell4 = openmc.Cell(cell_id=2, name='cell 4')
+
+# Use surface half-spaces to define regions
+cell1.region = -surf2
+cell2.region = -surf1
+cell3.region = +surf1
+cell4.region = +surf2 & -surf3
+
+# Register Materials with Cells
+cell2.fill = fuel
+cell3.fill = moderator
+cell4.fill = moderator
+
+# Instantiate Universes
+universe1 = openmc.Universe(universe_id=37)
+root = openmc.Universe(universe_id=0, name='root universe')
+cell1.fill = universe1
+
+# Register Cells with Universes
+universe1.add_cells([cell2, cell3])
+root.add_cells([cell1, cell4])
+
+# 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 = [-4., -4., -4., 4., 4., 4.]
+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 tallies.xml file
+###############################################################################
+
+# Instantiate some tally Filters
+cell_filter = openmc.CellFilter(cell2)
+energy_filter = openmc.EnergyFilter([0., 20.e6])
+energyout_filter = openmc.EnergyoutFilter([0., 20.e6])
+
+# Instantiate the first Tally
+first_tally = openmc.Tally(tally_id=1, name='first tally')
+first_tally.filters = [cell_filter]
+scores = ['total', 'scatter', 'nu-scatter',
+          'absorption', 'fission', 'nu-fission']
+first_tally.scores = scores
+
+# Instantiate the second Tally
+second_tally = openmc.Tally(tally_id=2, name='second tally')
+second_tally.filters = [cell_filter, energy_filter]
+second_tally.scores = scores
+
+# Instantiate the third Tally
+third_tally = openmc.Tally(tally_id=3, name='third tally')
+third_tally.filters = [cell_filter, energy_filter, energyout_filter]
+third_tally.scores = ['scatter', 'nu-scatter', 'nu-fission']
+
+# Instantiate a Tallies collection and export to XML
+tallies_file = openmc.Tallies((first_tally, second_tally, third_tally))
+tallies_file.export_to_xml()