Refactored examples into xml and python subdirectories. Implemented basic using Python API

2026-07-26 13:15:39 -04:00 · 2014-10-13 09:11:28 -04:00 · 2014-10-13 09:11:28 -04:00 · b0e5c88b57
commit b0e5c88b57
parent 6f4131ded4
40 changed files with 541 additions and 2 deletions
--- a/examples/python/basic/build-xml.py
+++ b/examples/python/basic/build-xml.py
@ -0,0 +1,141 @@
+import openmc
+
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 15
+inactive = 5
+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
+moderator = openmc.Material(material_id=41, name='moderator')
+moderator.setDensity('g/cc', 1.0)
+moderator.addNuclide(h1, 2.)
+moderator.addNuclide(o16, 1.)
+moderator.addSAlphaBeta('HH2O', '71t')
+
+fuel = openmc.Material(material_id=40, name='fuel')
+fuel.setDensity('g/cc', 4.5)
+fuel.addNuclide(u235, 1.)
+
+# Instantiate a MaterialsFile, register all Materials, and export to XML
+materials_file = openmc.MaterialsFile()
+materials_file.setDefaultXS('71c')
+materials_file.addMaterials([moderator, fuel])
+materials_file.exportToXML()
+
+
+###############################################################################
+#                 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.setBoundaryType('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')
+
+# Register Surfaces with Cells
+cell1.addSurface(surface=surf2, halfspace=-1)
+cell2.addSurface(surface=surf1, halfspace=-1)
+cell3.addSurface(surface=surf1, halfspace=+1)
+cell4.addSurface(surface=surf2, halfspace=+1)
+cell4.addSurface(surface=surf3, halfspace=-1)
+
+# Register Materials with Cells
+cell2.setFill(fuel)
+cell3.setFill(moderator)
+cell4.setFill(moderator)
+
+# Instantiate Universes
+universe1 = openmc.Universe(universe_id=37)
+root = openmc.Universe(universe_id=0, name='root universe')
+cell1.setFill(universe1)
+
+# Register Cells with Universes
+universe1.addCells([cell2, cell3])
+root.addCells([cell1, cell4])
+
+# Instantiate a Geometry and register the root Universe
+geometry = openmc.Geometry()
+geometry.setRootUniverse(root)
+
+# Instantiate a GeometryFile, register Geometry, and export to XML
+geometry_file = openmc.GeometryFile()
+geometry_file.setGeometry(geometry)
+geometry_file.exportToXML()
+
+
+###############################################################################
+#                   Exporting to OpenMC settings.xml File
+###############################################################################
+
+# Instantiate a SettingsFile, set all runtime parameters, and export to XML
+settings_file = openmc.SettingsFile()
+settings_file.setBatches(batches)
+settings_file.setInactive(inactive)
+settings_file.setParticles(particles)
+settings_file.setSourceSpace('box', [-4, -4, -4, 4, 4, 4])
+settings_file.exportToXML()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml File
+###############################################################################
+
+# Instantiate some tally Filters
+cell_filter = openmc.Filter(type='cell', bins=100)
+energy_filter = openmc.Filter(type='energy', bins=[0., 20.])
+energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.])
+
+# Instantiate the first Tally
+first_tally = openmc.Tally(tally_id=1, label='first tally')
+first_tally.addFilter(cell_filter)
+scores = ['total', 'scatter', 'nu-scatter', \
+          'absorption', 'fission', 'nu-fission']
+for score in scores:
+  first_tally.addScore(score)
+
+# Instantiate the second Tally
+second_tally = openmc.Tally(tally_id=2, label='second tally')
+second_tally.addFilter(cell_filter)
+second_tally.addFilter(energy_filter)
+scores = ['total', 'scatter', 'nu-scatter', \
+          'absorption', 'fission', 'nu-fission']
+for score in scores:
+  second_tally.addScore(score)
+
+# Instantiate the third Tally
+third_tally = openmc.Tally(tally_id=3, label='third tally')
+third_tally.addFilter(cell_filter)
+third_tally.addFilter(energy_filter)
+third_tally.addFilter(energyout_filter)
+scores = ['scatter', 'nu-scatter', 'nu-fission']
+for score in scores:
+  third_tally.addScore(score)
+
+# Instantiate a TalliesFile, register all Tallies, and export to XML
+tallies_file = openmc.TalliesFile()
+tallies_file.addTally(first_tally)
+tallies_file.addTally(second_tally)
+tallies_file.addTally(third_tally)
+tallies_file.exportToXML()