forked from crp/ecp-benchmarks
131 lines
4.5 KiB
Python
131 lines
4.5 KiB
Python
#!/usr/bin/env python3
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import argparse
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from math import pi, isclose
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from pathlib import Path
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import openmc
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from smr.materials import materials
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from smr.surfaces import bottom_fuel_stack, top_active_core, \
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pellet_OR, surfs, pin_pitch, clad_IR, clad_OR
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import smr.surfaces
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import smr.pins
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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', action='store_false',
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help='Do not use multipole cross sections')
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parser.add_argument('-a', '--axial', type=int, default=3,
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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(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-short-depleted')
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else:
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directory = Path('core-short-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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# Modify lattice pitch
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smr.surfaces.lattice_pitch = lattice_pitch = 17*smr.surfaces.pin_pitch
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# Modify fuel length
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length = 3. * pin_pitch
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smr.surfaces.active_fuel_length = length
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smr.pins.top_active_core = length
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surfs['top active core'].z0 = length
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# Change top and bottom of model to contain only fuel
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surfs['lower bound'].z0 = 0.0
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surfs['lower bound'].boundary_type = 'reflective'
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surfs['upper bound'].z0 = length
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surfs['upper bound'].boundary_type = 'reflective'
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ring_radii = [0.1*pin_pitch, 0.2*pin_pitch]
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geometry = core_geometry(ring_radii, args.axial, args.depleted)
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h = length / args.axial
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fuel_mats = {}
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for cell in geometry.get_all_cells().values():
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if cell.fill in materials:
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# Determine volume of each fuel material
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name = cell.fill.name
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if 'UO2 Fuel' in name:
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upper_right = cell.region.bounding_box[1][0]
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if isclose(upper_right, ring_radii[0]):
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ri, ro = 0.0, ring_radii[0]
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elif isclose(upper_right, ring_radii[1]):
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ri, ro = ring_radii[0], ring_radii[1]
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else:
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ri, ro = ring_radii[1], pellet_OR
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if (name, ri) not in fuel_mats:
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cell.fill = cell.fill.clone()
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cell.fill.volume = pi * (ro*ro - ri*ri) * h
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fuel_mats[name, ri] = cell.fill
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else:
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cell.fill = fuel_mats[name, ri]
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elif name == 'Helium':
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cell.fill.volume = pi * (clad_IR**2 - pellet_OR**2) * h
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elif name == 'M5':
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# Clad is not subdivided
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cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * length
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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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# Check assembly power distribution
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core_lattice = geometry.get_cells_by_fill_name('Main core')[0].fill
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mesh = openmc.RegularMesh.from_rect_lattice(core_lattice)
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assembly_power = openmc.Tally()
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assembly_power.filters = [openmc.MeshFilter(mesh)]
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assembly_power.scores = ['nu-fission']
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tallies = openmc.Tallies([assembly_power])
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tallies.export_to_xml(directory / 'tallies.xml')
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