forked from crp/ecp-benchmarks
140 lines
4.2 KiB
Python
140 lines
4.2 KiB
Python
#!/usr/bin/env python3
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"""Creates a 2D 2x2 assembly colorset with periodic BCs."""
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import os
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import shutil
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import copy
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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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#### Query the user for options
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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 = (multipole == 'y')
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# Query the user on whether to use distribmats or distribcells
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# If using distribmats, the geometry must be "differentiated" with unique
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# material instances for each instance of a fuel cell
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distrib = input('Use distribmat or distribcells? [mat/cell]: ').lower()
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if distrib not in ['cell', 'mat']:
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raise InputError('Distrib type "{}" is unsupported'.format(distrib))
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#### "Differentiate" the geometry if using distribmats
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if distrib == 'mat':
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Assign distribmats for each material
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for cell in fuel_cells:
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new_materials = []
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for i in range(cell.num_instances):
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new_materials.append(cell.fill.clone())
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# Fill cell with list of "differentiated" materials
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cell.fill = new_materials
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#### Create OpenMC "materials.xml" file
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all_materials = openmc_geometry.get_all_materials()
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materials = openmc.Materials(all_materials.values())
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materials.export_to_xml()
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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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# Construct uniform initial source distribution over fissionable zones
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lower_left = [-21.41728, -21.41728, +192.5]
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upper_right = [+21.41728, +21.41728, +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 = [21.41728*2, 21.41728*2]
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plot.origin = [0., 0., 195.]
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plot.color_by = 'material'
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plot.filename = '2x2-periodic'
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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 "tallies.xml" file
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tallies = openmc.Tallies()
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# Extract all fuel materials
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materials = openmc_geometry.get_materials_by_name(name='Fuel', matching=False)
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# If using distribcells, create distribcell tally needed for depletion
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if distrib == 'cell':
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True, matching=False)
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for cell in fuel_cells:
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tally = openmc.Tally(name='depletion tally')
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tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
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'fission', '(n,2n)', '(n,3n)', '(n,4n)']
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tally.nuclides = cell.fill.get_nuclides()
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tally.filters.append(openmc.DistribcellFilter([cell.id]))
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tallies.append(tally)
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# If using distribmats, create material tally needed for depletion
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elif distrib == 'mat':
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tally = openmc.Tally(name='depletion tally')
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tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
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'fission', '(n,2n)', '(n,3n)', '(n,4n)']
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tally.nuclides = materials[0].get_nuclides()
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material_ids = [material.id for material in materials]
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tally.filters.append(openmc.MaterialFilter(material_ids))
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tallies.append(tally)
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tallies.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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