add example for white boundary

examples/python/white/build-hemisphere.py

@@ -0,0 +1,73 @@
+import numpy as np
+import openmc
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 500
+inactive = 100
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+# Instantiate a Material and register the Nuclide
+fuel = openmc.Material(material_id=1, 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([fuel])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml file
+###############################################################################
+
+# Instantiate Surfaces
+surf1 = openmc.ZPlane(surface_id=1, z0=0.0, name='surf 1')
+surf2 = openmc.Sphere(surface_id=2, x0=0.0, y0=0.0, z0=0.0, r=0.39218, name='surf 2')
+
+surf1.boundary_type = 'vacuum'
+surf2.boundary_type = 'white'
+
+# Instantiate Cell
+cell = openmc.Cell(cell_id=1, name='cell 1')
+
+# Use surface half-spaces to define region
+cell.region = -surf1 & -surf2 
+
+# Register Material with Cell
+cell.fill = fuel
+
+# Instantiate Universes
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cell with Universe
+root.add_cell(cell)
+
+# 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
+uniform_dist = openmc.stats.Point(xyz=(0, 0, -0.1))
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+settings_file.export_to_xml()

examples/python/white/build-pin.py

@@ -0,0 +1,75 @@
+import numpy as np
+import openmc
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 500
+inactive = 100
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+# Instantiate a Material and register the Nuclide
+fuel = openmc.Material(material_id=1, 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([fuel])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml file
+###############################################################################
+
+# Instantiate Surfaces
+surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='surf 1')
+surf2 = openmc.ZPlane(surface_id=2, z0=-1, name='surf 2')
+surf3 = openmc.ZPlane(surface_id=3, z0=+1, name='surf 3')
+
+surf1.boundary_type = 'white'
+surf2.boundary_type = 'vacuum'
+surf3.boundary_type = 'vacuum'
+
+# Instantiate Cell
+cell = openmc.Cell(cell_id=1, name='cell 1')
+
+# Use surface half-spaces to define region
+cell.region = -surf1 & +surf2 & -surf3 
+
+# Register Material with Cell
+cell.fill = fuel
+
+# Instantiate Universes
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cell with Universe
+root.add_cell(cell)
+
+# 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
+uniform_dist = openmc.stats.Point(xyz=(0, 0, 0))
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+settings_file.export_to_xml()

examples/python/white/build-plane.py

@@ -0,0 +1,81 @@
+import numpy as np
+import openmc
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 500
+inactive = 100
+particles = 10000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+# Instantiate a Material and register the Nuclide
+fuel = openmc.Material(material_id=1, 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([fuel])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml file
+###############################################################################
+
+# Instantiate Surfaces
+surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
+surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
+surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
+surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
+surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
+surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
+
+surf1.boundary_type = 'vacuum'
+surf2.boundary_type = 'vacuum'
+surf3.boundary_type = 'white'
+surf4.boundary_type = 'white'
+surf5.boundary_type = 'white'
+surf6.boundary_type = 'white'
+
+# Instantiate Cell
+cell = openmc.Cell(cell_id=1, name='cell 1')
+
+# Use surface half-spaces to define region
+cell.region = +surf1 & -surf2 & +surf3 & -surf4 & +surf5 & -surf6
+
+# Register Material with Cell
+cell.fill = fuel
+
+# Instantiate Universes
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cell with Universe
+root.add_cell(cell)
+
+# 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
+uniform_dist = openmc.stats.Point(xyz=(0, 0, 0))
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+settings_file.export_to_xml()