forked from crp/ecp-benchmarks
143 lines
4.5 KiB
Python
143 lines
4.5 KiB
Python
"""Creates a 2D fuel assembly with reflective BCs."""
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import os
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import shutil
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import numpy as np
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import openmc
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from geometry import beavrs, openmc_geometry
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#### Create OpenMC "materials.xml" file
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beavrs.write_openmc_materials()
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#### Create OpenMC "geometry.xml" file
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openmc_geometry.export_to_xml()
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#### Create OpenMC "settings.xml" file
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# Query the user on whether to use multipole cross sections
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multipole = input('Use multipole cross sections? (y/n): ').lower()
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multipole = True if multipole == 'y' else False
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# Construct uniform initial source distribution over fissionable zones
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lower_left = [-10.70864, -10.70864, +192.5]
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upper_right = [+10.70864, +10.70864, +197.5]
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source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
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source.space.only_fissionable = True
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settings_file = openmc.Settings()
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settings_file.batches = 10
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settings_file.inactive = 5
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settings_file.particles = 10000
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settings_file.output = {'tallies': False}
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settings_file.source = source
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settings_file.sourcepoint_write = False
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if multipole:
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settings_file.temperature = {'multipole': True, 'tolerance': 1000}
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settings_file.export_to_xml()
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#### Create OpenMC "plots.xml" file
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# Initialize the BEAVRS color mapping scheme
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beavrs.write_openmc_plots()
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# Create a plot colored by materials
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plot = openmc.Plot()
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plot.width = [10.70864*2, 10.70864*2]
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plot.origin = [0., 0., 195.]
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plot.color = 'mat'
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plot.filename = 'assembly'
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plot.col_spec = beavrs.plots.colspec_mat
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plot.pixels = [1000, 1000]
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plot_file = openmc.Plots([plot])
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plot_file.export_to_xml()
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#### Create OpenMC MGXS libraries
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# Get all cells filled with a "fuel" material
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mat_cells = openmc_geometry.get_all_material_cells()
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fuel_cells = []
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for cell in mat_cells:
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if 'fuel' in cell.fill.name.lower():
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fuel_cells.append(cell)
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# CASMO 70-group structure
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energy_groups = openmc.mgxs.EnergyGroups()
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energy_groups.group_edges = np.array([
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0, 0.005, 0.01, 0.015, 0.02, 0.025, 0.03, 0.035, 0.042, 0.05, 0.058, 0.067,
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0.08, 0.1, 0.14, 0.18, 0.22, 0.25, 0.28, 0.3, 0.32, 0.35, 0.4, 0.5, 0.625,
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0.78, 0.85, 0.91, 0.95, 0.972, 0.996, 1.02, 1.045, 1.071, 1.097, 1.123,
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1.15, 1.3, 1.5, 1.855, 2.1, 2.6, 3.3, 4., 9.877, 15.968, 27.7, 48.052,
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75.501, 148.73, 367.26001, 906.90002, 1.4251e3, 2.2395e3, 3.5191e3, 5.53e3,
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9.118e3, 15.03e3, 24.78e3, 40.85e3, 67.34e3, 111.e3, 183e3, 302.5e3, 500e3,
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821e3, 1.353e6, 2.231e6, 3.679e6, 6.0655e6, 2e7])
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# Initialize a 70-group "distribcell" MGXS library
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cell_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
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cell_mgxs_lib.energy_groups = energy_groups
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cell_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
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cell_mgxs_lib.domain_type = 'distribcell'
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cell_mgxs_lib.domains = fuel_cells
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cell_mgxs_lib.correction = None
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cell_mgxs_lib.build_library()
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# Initialize a 70-group "material" MGXS library
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mat_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
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mat_mgxs_lib.energy_groups = energy_groups
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mat_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
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mat_mgxs_lib.domain_type = 'material'
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mat_mgxs_lib.correction = None
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mat_mgxs_lib.build_library()
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#### Create mesh tallies for verification of pin-wise reaction rates
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# Instantiate a tally Mesh
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mesh = openmc.Mesh(name='assembly mesh')
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mesh.type = 'regular'
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mesh.dimension = [17, 17, 1]
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mesh.lower_left = lower_left
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mesh.width = (np.array(upper_right) - np.array(lower_left))
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mesh.width[:2] /= 17
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mesh_filter = openmc.MeshFilter(mesh)
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# Instantiate energy-integrated fission rate mesh Tally
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fission_rates = openmc.Tally(name='fission rates')
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fission_rates.filters = [mesh_filter]
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fission_rates.scores = ['fission']
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# Instantiate energy-wise U-238 capture rate mesh Tally
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capture_rates = openmc.Tally(name='u-238 capture')
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capture_rates.filters = [mesh_filter]
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capture_rates.nuclides = ['U238']
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capture_rates.scores = ['absorption', 'fission']
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#### Create OpenMC "tallies.xml" file
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# Create a "tallies.xml" file for the mesh tallies
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tallies_file = openmc.Tallies([fission_rates, capture_rates])
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cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
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mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
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tallies_file.export_to_xml()
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#### Move all XML files to 'fresh' directory
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if not os.path.exists('fresh'):
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os.makedirs('fresh')
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shutil.move('materials.xml', 'fresh/materials.xml')
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shutil.move('geometry.xml', 'fresh/geometry.xml')
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shutil.move('settings.xml', 'fresh/settings.xml')
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shutil.move('tallies.xml', 'fresh/tallies.xml')
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shutil.move('plots.xml', 'fresh/plots.xml')
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