From 5e788cdffb7f5ba836da4a9c5c114c19d0c3b8f6 Mon Sep 17 00:00:00 2001 From: Adam Nelson Date: Mon, 16 Nov 2015 05:31:40 -0500 Subject: [PATCH] Whoops, forgot to add --- .../python/pincell_multigroup/build-xml.py | 185 ++++++++++++++++++ 1 file changed, 185 insertions(+) create mode 100644 examples/python/pincell_multigroup/build-xml.py diff --git a/examples/python/pincell_multigroup/build-xml.py b/examples/python/pincell_multigroup/build-xml.py new file mode 100644 index 000000000..300f36ce0 --- /dev/null +++ b/examples/python/pincell_multigroup/build-xml.py @@ -0,0 +1,185 @@ +import openmc +import openmc.mgxs +import numpy as np + +############################################################################### +# Simulation Input File Parameters +############################################################################### + +# OpenMC simulation parameters +batches = 100 +inactive = 10 +particles = 1000 + +############################################################################### +# Exporting to OpenMC mg_cross_sections.xml File +############################################################################### + +# Instantiate the energy group data +groups = openmc.mgxs.EnergyGroups(group_edges=[1E-11, 0.0635E-6, 10.0E-6, + 1.0E-4, 1.0E-3, 0.5, 1.0, 20.0]) + +# Instantiate the 7-group (C5G7) cross section data +uo2_xsdata = openmc.Xsdata('UO2.300k', groups) +uo2_xsdata.order = 0 +uo2_xsdata.total = np.array([0.1779492, 0.3298048, 0.4803882, 0.5543674, + 0.3118013, 0.3951678, 0.5644058]) +uo2_xsdata.absorption = np.array([8.0248E-03, 3.7174E-03, 2.6769E-02, 9.6236E-02, + 3.0020E-02, 1.1126E-01, 2.8278E-01]) +scatter = [[[0.1275370, 0.0423780, 0.0000094, 0.0000000, 0.0000000, 0.0000000, 0.0000000], + [0.0000000, 0.3244560, 0.0016314, 0.0000000, 0.0000000, 0.0000000, 0.0000000], + [0.0000000, 0.0000000, 0.4509400, 0.0026792, 0.0000000, 0.0000000, 0.0000000], + [0.0000000, 0.0000000, 0.0000000, 0.4525650, 0.0055664, 0.0000000, 0.0000000], + [0.0000000, 0.0000000, 0.0000000, 0.0001253, 0.2714010, 0.0102550, 0.0000000], + [0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0012968, 0.2658020, 0.0168090], + [0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0085458, 0.2730800]]] +uo2_xsdata.scatter = np.array(scatter[:][:]) +uo2_xsdata.fission = np.array([7.21206E-03, 8.19301E-04, 6.45320E-03, + 1.85648E-02, 1.78084E-02, 8.30348E-02, + 2.16004E-01]) +uo2_xsdata.nu_fission = np.array([2.005998E-02, 2.027303E-03, 1.570599E-02, + 4.518301E-02, 4.334208E-02, 2.020901E-01, + 5.257105E-01]) +uo2_xsdata.chi = np.array([5.8791E-01, 4.1176E-01, 3.3906E-04, 1.1761E-07, + 0.0000E+00, 0.0000E+00, 0.0000E+00]) + +h2o_xsdata = openmc.Xsdata('LWTR.300k', groups) +h2o_xsdata.order = 0 +h2o_xsdata.total = np.array([0.15920605, 0.412969593, 0.59030986, 0.58435, + 0.718, 1.2544497, 2.650379]) +h2o_xsdata.absorption = np.array([6.0105E-04, 1.5793E-05, 3.3716E-04, + 1.9406E-03, 5.7416E-03, 1.5001E-02, + 3.7239E-02]) +scatter = [[[0.0444777, 0.1134000, 0.0007235, 0.0000037, 0.0000001, 0.0000000, 0.0000000], + [0.0000000, 0.2823340, 0.1299400, 0.0006234, 0.0000480, 0.0000074, 0.0000010], + [0.0000000, 0.0000000, 0.3452560, 0.2245700, 0.0169990, 0.0026443, 0.0005034], + [0.0000000, 0.0000000, 0.0000000, 0.0910284, 0.4155100, 0.0637320, 0.0121390], + [0.0000000, 0.0000000, 0.0000000, 0.0000714, 0.1391380, 0.5118200, 0.0612290], + [0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0022157, 0.6999130, 0.5373200], + [0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.0000000, 0.1324400, 2.4807000]]] +h2o_xsdata.scatter = np.array(scatter[:][:]) + +mg_cross_sections_file = openmc.MGXSLibraryFile(groups) +mg_cross_sections_file.add_xsdatas([uo2_xsdata,h2o_xsdata]) +mg_cross_sections_file.export_to_xml() + + +############################################################################### +# Exporting to OpenMC materials.xml File +############################################################################### + +# Instantiate some Macroscopic Data +uo2_data = openmc.Macroscopic('UO2', '300k') +h2o_data = openmc.Macroscopic('LWTR', '300k') + +# Instantiate some Materials and register the appropriate Nuclides +uo2 = openmc.Material(material_id=1, name='UO2 fuel') +uo2.set_density('macro', 1.0) +uo2.add_macroscopic(uo2_data) + +water = openmc.Material(material_id=2, name='Water') +water.set_density('macro', 1.0) +water.add_macroscopic(h2o_data) + +# Instantiate a MaterialsFile, register all Materials, and export to XML +materials_file = openmc.MaterialsFile() +materials_file.default_xs = '300k' +materials_file.add_materials([uo2, 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=0.54, name='Fuel OR') +left = openmc.XPlane(surface_id=4, x0=-0.63, name='left') +right = openmc.XPlane(surface_id=5, x0=0.63, name='right') +bottom = openmc.YPlane(surface_id=6, y0=-0.63, name='bottom') +top = openmc.YPlane(surface_id=7, y0=0.63, 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') +moderator = openmc.Cell(cell_id=2, name='cell 2') + +# Use surface half-spaces to define regions +fuel.region = -fuel_or +moderator.region = +fuel_or & +left & -right & +bottom & -top + +# Register Materials with Cells +fuel.fill = uo2 +moderator.fill = water + +# Instantiate Universe +root = openmc.Universe(universe_id=0, name='root universe') + +# Register Cells with Universe +root.add_cells([fuel, moderator]) + +# Instantiate a Geometry and register the root Universe +geometry = openmc.Geometry() +geometry.root_universe = root + +# Instantiate a GeometryFile, register Geometry, and export to XML +geometry_file = openmc.GeometryFile() +geometry_file.geometry = geometry +geometry_file.export_to_xml() + + +############################################################################### +# Exporting to OpenMC settings.xml File +############################################################################### + +# Instantiate a SettingsFile, set all runtime parameters, and export to XML +settings_file = openmc.SettingsFile() +settings_file.energy_mode = "multi-group" +settings_file.cross_sections = "./mg_cross_sections.xml" +settings_file.batches = batches +settings_file.inactive = inactive +settings_file.particles = particles +settings_file.set_source_space('box', [-0.63, -0.63, -1, \ + 0.63, 0.63, 1]) +settings_file.entropy_lower_left = [-0.54, -0.54, -1.e50] +settings_file.entropy_upper_right = [0.54, 0.54, 1.e50] +settings_file.entropy_dimension = [10, 10, 1] +settings_file.export_to_xml() + + +############################################################################### +# Exporting to OpenMC tallies.xml File +############################################################################### + +# Instantiate a tally mesh +mesh = openmc.Mesh(mesh_id=1) +mesh.type = 'regular' +mesh.dimension = [100, 100, 1] +mesh.lower_left = [-0.63, -0.63, -1.e50] +mesh.upper_right = [0.63, 0.63, 1.e50] + +# Instantiate some tally Filters +# energy_filter = openmc.Filter(type='energy', +# bins=[1E-11, 0.0635E-6, 10.0E-6, 1.0E-4, 1.0E-3, +# 0.5, 1.0, 20.0]) +energy_filter = openmc.Filter(type='energy') +mesh_filter = openmc.Filter() +mesh_filter.mesh = mesh + +# Instantiate the Tally +tally = openmc.Tally(tally_id=1, name='tally 1') +tally.add_filter(energy_filter) +tally.add_filter(mesh_filter) +tally.add_score('flux') +tally.add_score('fission') +tally.add_score('nu-fission') + +# Instantiate a TalliesFile, register all Tallies, and export to XML +tallies_file = openmc.TalliesFile() +tallies_file.add_mesh(mesh) +tallies_file.add_tally(tally) +tallies_file.export_to_xml()