from math import log10 import numpy as np import openmc import openmc.mgxs ############################################################################### # Create multigroup data # Instantiate the energy group data groups = openmc.mgxs.EnergyGroups(group_edges=[ 1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6]) # Instantiate the 7-group (C5G7) cross section data uo2_xsdata = openmc.XSdata('UO2', groups) uo2_xsdata.order = 0 uo2_xsdata.set_total( [0.1779492, 0.3298048, 0.4803882, 0.5543674, 0.3118013, 0.3951678, 0.5644058]) uo2_xsdata.set_absorption([8.0248E-03, 3.7174E-03, 2.6769E-02, 9.6236E-02, 3.0020E-02, 1.1126E-01, 2.8278E-01]) scatter_matrix = np.array( [[[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]]]) scatter_matrix = np.rollaxis(scatter_matrix, 0, 3) uo2_xsdata.set_scatter_matrix(scatter_matrix) uo2_xsdata.set_fission([7.21206E-03, 8.19301E-04, 6.45320E-03, 1.85648E-02, 1.78084E-02, 8.30348E-02, 2.16004E-01]) uo2_xsdata.set_nu_fission([2.005998E-02, 2.027303E-03, 1.570599E-02, 4.518301E-02, 4.334208E-02, 2.020901E-01, 5.257105E-01]) uo2_xsdata.set_chi([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', groups) h2o_xsdata.order = 0 h2o_xsdata.set_total([0.15920605, 0.412969593, 0.59030986, 0.58435, 0.718, 1.2544497, 2.650379]) h2o_xsdata.set_absorption([6.0105E-04, 1.5793E-05, 3.3716E-04, 1.9406E-03, 5.7416E-03, 1.5001E-02, 3.7239E-02]) scatter_matrix = np.array( [[[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]]]) scatter_matrix = np.rollaxis(scatter_matrix, 0, 3) h2o_xsdata.set_scatter_matrix(scatter_matrix) mg_cross_sections_file = openmc.MGXSLibrary(groups) mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata]) mg_cross_sections_file.export_to_hdf5() ############################################################################### # Create materials for the problem # Instantiate some Macroscopic Data uo2_data = openmc.Macroscopic('UO2') h2o_data = openmc.Macroscopic('LWTR') # Instantiate some Materials and register the appropriate Macroscopic objects uo2 = openmc.Material(name='UO2 fuel') uo2.set_density('macro', 1.0) uo2.add_macroscopic(uo2_data) water = openmc.Material(name='Water') water.set_density('macro', 1.0) water.add_macroscopic(h2o_data) # Instantiate a Materials collection and export to XML materials_file = openmc.Materials([uo2, water]) materials_file.cross_sections = "mgxs.h5" materials_file.export_to_xml() ############################################################################### # Define problem geometry # Create a surface for the fuel outer radius fuel_or = openmc.ZCylinder(r=0.54, name='Fuel OR') # Create a region represented as the inside of a rectangular prism pitch = 1.26 box = openmc.model.RectangularPrism(pitch, pitch, boundary_type='reflective') # Instantiate Cells fuel = openmc.Cell(fill=uo2, region=-fuel_or, name='fuel') moderator = openmc.Cell(fill=water, region=+fuel_or & -box, name='moderator') # Create a geometry with the two cells and export to XML geometry = openmc.Geometry([fuel, moderator]) geometry.export_to_xml() ############################################################################### # Define problem settings # Instantiate a Settings object, set all runtime parameters, and export to XML settings = openmc.Settings() settings.energy_mode = "multi-group" settings.batches = 100 settings.inactive = 10 settings.particles = 1000 # Create an initial uniform spatial source distribution over fissionable zones lower_left = (-pitch/2, -pitch/2, -1) upper_right = (pitch/2, pitch/2, 1) uniform_dist = openmc.stats.Box(lower_left, upper_right) settings.source = openmc.IndependentSource( space=uniform_dist, constraints={'fissionable': True}) settings.export_to_xml() ############################################################################### # Define tallies # Create a mesh that will be used for tallying mesh = openmc.RegularMesh() mesh.dimension = (100, 100) mesh.lower_left = (-pitch/2, -pitch/2) mesh.upper_right = (pitch/2, pitch/2) # Create a mesh filter that can be used in a tally mesh_filter = openmc.MeshFilter(mesh) # Now use the mesh filter in a tally and indicate what scores are desired mesh_tally = openmc.Tally(name="Mesh tally") mesh_tally.filters = [mesh_filter] mesh_tally.scores = ['flux', 'fission', 'nu-fission'] # Let's also create a tally to get the flux energy spectrum. We start by # creating an energy filter e_min, e_max = 1e-5, 20.0e6 groups = 500 energies = np.logspace(log10(e_min), log10(e_max), groups + 1) energy_filter = openmc.EnergyFilter(energies) spectrum_tally = openmc.Tally(name="Flux spectrum") spectrum_tally.filters = [energy_filter] spectrum_tally.scores = ['flux'] # Instantiate a Tallies collection and export to XML tallies = openmc.Tallies([mesh_tally, spectrum_tally]) tallies.export_to_xml()