forked from crp/ecp-benchmarks
121 lines
4.2 KiB
Python
121 lines
4.2 KiB
Python
#!/usr/bin/env python3
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import argparse
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from math import pi
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from pathlib import Path
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import numpy as np
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import openmc
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from tqdm import tqdm
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from smr.materials import materials, clone
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from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, \
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pellet_OR, active_fuel_length
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from smr.core import core_geometry
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from smr import inlet_temperature
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# Define command-line options
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parser = argparse.ArgumentParser()
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parser.add_argument('--multipole', action='store_true',
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help='Use multipole cross sections')
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parser.add_argument('--no-multipole', dest='multipole', action='store_false',
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help='Do not use multipole cross sections')
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parser.add_argument('--clone', action='store_true',
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help='Clone materials for each cell instance')
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parser.add_argument('--no-clone', dest='clone', action='store_false',
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help='Do not clone materials for each cell instance')
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parser.add_argument('-r', '--rings', type=int, default=10,
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help='Number of annular regions in fuel')
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parser.add_argument('-a', '--axial', type=int, default=196,
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help='Number of axial subdivisions in fuel')
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parser.add_argument('-d', '--depleted', action='store_true',
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help='Whether UO2 compositions should represent depleted fuel')
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parser.add_argument('-o', '--output-dir', type=Path, default=None)
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parser.set_defaults(clone=False, multipole=True)
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args = parser.parse_args()
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# Make directory for inputs
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if args.output_dir is None:
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if args.depleted:
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directory = Path('core-depleted')
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else:
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directory = Path('core-fresh')
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else:
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directory = args.output_dir
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directory.mkdir(exist_ok=True)
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if args.rings > 1:
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ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
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else:
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ring_radii = None
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geometry = core_geometry(ring_radii, args.axial, args.depleted)
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h = active_fuel_length / args.axial
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fuel_mats = {}
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# Count the number of instances for each cell and material
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if args.clone:
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geometry.determine_paths(instances_only=True)
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fuel_volume = pi * pellet_OR**2 * h / args.rings
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for cell in tqdm(geometry.get_all_cells().values(),
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desc='Differentiating materials / assigning volume'):
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if cell.fill in materials:
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# Determine if this material is fuel
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name = cell.fill.name
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is_fuel = 'UO2 Fuel' in name
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# Determine volume of each fuel material
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if is_fuel:
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if args.clone:
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# Fill cell with list of "differentiated" materials if requested
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cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
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for mat in cell.fill:
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mat.volume = fuel_volume
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else:
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r_o = cell.region.bounding_box[1][0]
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if (name, r_o) not in fuel_mats:
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cell.fill = cell.fill.clone()
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cell.fill.volume = fuel_volume
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fuel_mats[name, r_o] = cell.fill
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else:
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cell.fill = fuel_mats[name, r_o]
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else:
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cell.fill.volume = 1.0
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#### Create OpenMC "materials.xml" file
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all_materials = geometry.get_all_materials()
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materials = openmc.Materials(all_materials.values())
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materials.export_to_xml(str(directory / 'materials.xml'))
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#### Create OpenMC "geometry.xml" file
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geometry.export_to_xml(str(directory / 'geometry.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 = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
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upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
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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 = openmc.Settings()
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settings.batches = 200
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settings.inactive = 100
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settings.particles = 10000
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settings.output = {'tallies': False, 'summary': False}
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settings.source = source
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settings.sourcepoint = {'write': False}
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settings.temperature = {
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'default': inlet_temperature,
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'method': 'interpolation',
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'range': (300.0, 1500.0),
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}
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if args.multipole:
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settings.temperature['multipole'] = True
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settings.temperature['tolerance'] = 1000
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settings.export_to_xml(str(directory / 'settings.xml'))
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