forked from crp/ecp-benchmarks
Remove tallies, don't clone materials unless needed
This commit is contained in:
Paul Romano
parent
e53ffa6ece
commit
2b8e9e9f5d
2 changed files with 40 additions and 119 deletions
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@ -22,9 +22,7 @@ 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('-t', '--tallies', choices=('cell', 'mat'), default='mat',
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help='Whether to use distribmats or distribcells for tallies')
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parser.add_argument('-a', '--axial', type=int, default=92,
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parser.add_argument('-a', '--axial', type=int, default=100,
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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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@ -99,40 +97,30 @@ for halfspace in surfs['lat grid box inner']:
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# Define geometry with a single assembly
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geometry = openmc.Geometry(root_universe)
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def clone(material):
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"""Perform copy of material but share nuclide densities"""
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shared_mat = copy.copy(material)
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shared_mat.id = None
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return shared_mat
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#### "Differentiate" the geometry if using distribmats
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h = active_fuel_length / args.axial
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if args.tallies == 'mat':
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# Count the number of instances for each cell and material
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geometry.determine_paths(instances_only=True)
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for cell in tqdm(geometry.get_all_material_cells().values(),
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desc='Differentiating materials'):
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if cell.fill in materials:
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# Fill cell with list of "differentiated" materials
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cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
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fuel_mats = {}
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# Determine volume of each fuel material
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if 'UO2 Fuel' in cell.fill[0].name:
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upper_right = cell.region.bounding_box[1]
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if isclose(upper_right[0], rings[0]):
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ri, ro = 0.0, rings[0]
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elif isclose(upper_right[0], rings[1]):
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ri, ro = rings[0], rings[1]
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else:
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ri, ro = rings[1], pellet_OR
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for mat in cell.fill:
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mat.volume = pi * (ro*ro - ri*ri) * h
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for cell in tqdm(geometry.get_all_material_cells().values(),
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desc='Assigning volume'):
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if cell.fill in materials:
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# Determine volume of each fuel material
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if 'UO2 Fuel' in cell.fill.name:
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upper_right = cell.region.bounding_box[1]
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if isclose(upper_right[0], rings[0]):
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ri, ro = 0.0, rings[0]
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elif isclose(upper_right[0], rings[1]):
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ri, ro = rings[0], rings[1]
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else:
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for mat in cell.fill:
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mat.volume = 1.0
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ri, ro = rings[1], pellet_OR
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if 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[ri] = cell.fill
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else:
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cell.fill = fuel_mats[ri]
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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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print('Getting materials...')
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@ -173,34 +161,3 @@ if args.multipole:
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}
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settings.export_to_xml(str(directory / 'settings.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 = 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 args.tallies == 'cell':
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# Extract all cells filled by a fuel material
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fuel_cells = []
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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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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]))
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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 args.tallies == '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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tally.filters = [openmc.MaterialFilter(materials)]
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tallies.append(tally)
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tallies.export_to_xml(str(directory / 'tallies.xml'))
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@ -4,6 +4,7 @@ import os
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import shutil
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import copy
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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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@ -11,26 +12,18 @@ import numpy as np
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import openmc
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from smr.materials import materials
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from smr.plots import core_plots
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from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
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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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def clone(mat):
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"""Make a shallow copy of a material (sharing compositions)."""
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mat_copy = copy.copy(mat)
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mat_copy.id = None
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return mat_copy
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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('-t', '--tallies', choices=('cell', 'mat'), default='cell',
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help='Whether to use distribmats or distribcells for tallies')
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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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@ -57,18 +50,24 @@ 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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#### "Differentiate" the geometry if using distribmats
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if args.tallies == 'mat':
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# Count the number of instances for each cell and material
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geometry.determine_paths(instances_only=True)
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h = active_fuel_length / args.axial
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fuel_mats = {}
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# Extract all cells filled by a fuel material
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fuel_mats = {m for m in materials if 'UO2 Fuel' in m.name}
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fuel_volume = pi * pellet_OR**2 * h / args.rings
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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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if 'UO2 Fuel' in cell.fill.name:
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r_o = cell.region.bounding_box[1][0]
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if 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[r_o] = cell.fill
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else:
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cell.fill = fuel_mats[r_o]
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else:
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cell.fill.volume = 1.0
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for cell in geometry.get_all_cells().values():
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if cell.fill in fuel_mats:
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# Fill cell with list of "differentiated" materials
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cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
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#### Create OpenMC "materials.xml" file
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all_materials = geometry.get_all_materials()
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@ -107,38 +106,3 @@ if args.multipole:
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settings.export_to_xml(str(directory / 'settings.xml'))
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#### Create OpenMC "plots.xml" file
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plots = core_plots()
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plots.export_to_xml(str(directory / 'plots.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 = 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 args.tallies == 'cell':
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# Extract all cells filled by a fuel material
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fuel_cells = []
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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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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]))
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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 args.tallies == '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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tally.filters = [openmc.MaterialFilter(materials)]
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tallies.append(tally)
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tallies.export_to_xml(str(directory / 'tallies.xml'))
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