adding back files to be reviewed

2026-07-27 05:35:49 -04:00 · 2019-10-28 11:55:45 -05:00 · 2019-10-28 11:55:45 -05:00 · bc09d1ef55
commit bc09d1ef55
parent ae28233110
1244 changed files with 301904 additions and 0 deletions
--- a/examples/python/pincell/build-xml.py
+++ b/examples/python/pincell/build-xml.py
@ -0,0 +1,138 @@
+import openmc
+
+###############################################################################
+#                      Simulation Input File Parameters
+###############################################################################
+
+# OpenMC simulation parameters
+batches = 100
+inactive = 10
+particles = 1000
+
+
+###############################################################################
+#                 Exporting to OpenMC materials.xml file
+###############################################################################
+
+
+# Instantiate some Materials and register the appropriate Nuclides
+uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
+uo2.set_density('g/cm3', 10.29769)
+uo2.add_element('U', 1., enrichment=2.4)
+uo2.add_element('O', 2.)
+
+helium = openmc.Material(material_id=2, name='Helium for gap')
+helium.set_density('g/cm3', 0.001598)
+helium.add_element('He', 2.4044e-4)
+
+zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
+zircaloy.set_density('g/cm3', 6.55)
+zircaloy.add_element('Sn', 0.014  , 'wo')
+zircaloy.add_element('Fe', 0.00165, 'wo')
+zircaloy.add_element('Cr', 0.001  , 'wo')
+zircaloy.add_element('Zr', 0.98335, 'wo')
+
+borated_water = openmc.Material(material_id=4, name='Borated water')
+borated_water.set_density('g/cm3', 0.740582)
+borated_water.add_element('B', 4.0e-5)
+borated_water.add_element('H', 5.0e-2)
+borated_water.add_element('O', 2.4e-2)
+borated_water.add_s_alpha_beta('c_H_in_H2O')
+
+# Instantiate a Materials collection and export to XML
+materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
+materials_file.export_to_xml()
+
+
+###############################################################################
+#                 Exporting to OpenMC geometry.xml file
+###############################################################################
+
+# Instantiate ZCylinder surfaces
+fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='Fuel OR')
+clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.40005, name='Clad IR')
+clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=0.45720, name='Clad OR')
+left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
+right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
+bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
+top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
+
+left.boundary_type = 'reflective'
+right.boundary_type = 'reflective'
+top.boundary_type = 'reflective'
+bottom.boundary_type = 'reflective'
+
+# Instantiate Cells
+fuel = openmc.Cell(cell_id=1, name='cell 1')
+gap = openmc.Cell(cell_id=2, name='cell 2')
+clad = openmc.Cell(cell_id=3, name='cell 3')
+water = openmc.Cell(cell_id=4, name='cell 4')
+
+# Use surface half-spaces to define regions
+fuel.region = -fuel_or
+gap.region = +fuel_or & -clad_ir
+clad.region = +clad_ir & -clad_or
+water.region = +clad_or & +left & -right & +bottom & -top
+
+# Register Materials with Cells
+fuel.fill = uo2
+gap.fill = helium
+clad.fill = zircaloy
+water.fill = borated_water
+
+# Instantiate Universe
+root = openmc.Universe(universe_id=0, name='root universe')
+
+# Register Cells with Universe
+root.add_cells([fuel, gap, clad, water])
+
+# 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 = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
+uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
+settings_file.source = openmc.source.Source(space=uniform_dist)
+
+entropy_mesh = openmc.RegularMesh()
+entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
+entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
+entropy_mesh.dimension = [10, 10, 1]
+settings_file.entropy_mesh = entropy_mesh
+settings_file.export_to_xml()
+
+
+###############################################################################
+#                   Exporting to OpenMC tallies.xml file
+###############################################################################
+
+# Instantiate a tally mesh
+mesh = openmc.RegularMesh()
+mesh.dimension = [100, 100, 1]
+mesh.lower_left = [-0.62992, -0.62992, -1.e50]
+mesh.upper_right = [0.62992, 0.62992, 1.e50]
+
+# Instantiate some tally Filters
+energy_filter = openmc.EnergyFilter([0., 4., 20.e6])
+mesh_filter = openmc.MeshFilter(mesh)
+
+# Instantiate the Tally
+tally = openmc.Tally(tally_id=1, name='tally 1')
+tally.filters = [energy_filter, mesh_filter]
+tally.scores = ['flux', 'fission', 'nu-fission']
+
+# Instantiate a Tallies collection and export to XML
+tallies_file = openmc.Tallies([tally])
+tallies_file.export_to_xml()