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Respond to @wbinventor comments on #632
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15 changed files with 146 additions and 138 deletions
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@ -35,7 +35,7 @@ iron = openmc.Material(material_id=3, name='iron')
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iron.set_density('g/cc', 7.9)
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iron.add_nuclide(fe56, 1.)
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# Instantiate a Materials object, register all Materials, and export to XML
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# Instantiate a Materials collection, register all Materials, and export to XML
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materials_file = openmc.Materials()
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materials_file.default_xs = '71c'
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materials_file.add_materials([moderator, fuel, iron])
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@ -105,7 +105,7 @@ lattice.outer = univ2
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# Fill Cell with the Lattice
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cell1.fill = lattice
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# Instantiate a Geometry and register the root Universe, and export to XML
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# Instantiate a Geometry, register the root Universe, and export to XML
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geometry = openmc.Geometry()
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geometry.root_universe = root
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geometry.export_to_xml()
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@ -115,7 +115,7 @@ geometry.export_to_xml()
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# Exporting to OpenMC settings.xml file
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###############################################################################
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# Instantiate a SettingsFile, set all runtime parameters, and export to XML
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# Instantiate a Settings object, set all runtime parameters, and export to XML
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settings_file = openmc.Settings()
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settings_file.batches = batches
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settings_file.inactive = inactive
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@ -151,7 +151,7 @@ plot_yz.width = [8, 8]
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plot_yz.pixels = [400, 400]
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plot_yz.color = 'mat'
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# Instantiate a Plots object, add plots, and export to XML
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# Instantiate a Plots collection, add plots, and export to XML
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plot_file = openmc.Plots()
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plot_file.add_plot(plot_xy)
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plot_file.add_plot(plot_yz)
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@ -167,7 +167,7 @@ tally = openmc.Tally(tally_id=1)
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tally.filters = [openmc.Filter(type='distribcell', bins=[cell2.id])]
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tally.scores = ['total']
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# Instantiate a Tallies object, register Tally/Mesh, and export to XML
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# Instantiate a Tallies collection, register Tally/Mesh, and export to XML
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tallies_file = openmc.Tallies()
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tallies_file.add_tally(tally)
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tallies_file.export_to_xml()
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