forked from crp/ecp-benchmarks
Merge pull request #13 from mit-crpg/ecp-assembly-long
Add script for assembly model used in AD-SE-08-61 milestone
This commit is contained in:
Paul Romano
GitHub
commit
e53ffa6ece
9 changed files with 487 additions and 44 deletions
206
smr/build-assembly-long.py
Normal file
206
smr/build-assembly-long.py
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#!/usr/bin/env python3
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import argparse
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import copy
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from math import pi, isclose
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from pathlib import Path
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import numpy as np
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from tqdm import tqdm
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import openmc
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from smr.materials import materials, mats
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from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
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top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR, \
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active_fuel_length
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from smr.pins import pin_universes, make_stack
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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='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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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('assembly-long-depleted')
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else:
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directory = Path('assembly-long-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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rings = [0.1*pin_pitch, 0.2*pin_pitch]
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assembly_long_surfs = [
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surfs['bottom FR'],
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surfs['bot active core'],
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surfs['top active core'],
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surfs['top pin plenum'],
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surfs['top FR'],
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surfs['bot upper nozzle'],
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surfs['top upper nozzle']
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]
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univs = pin_universes(rings, args.axial, args.depleted)
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fuel_univ = make_stack(
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'Fuel (3.1%) stack no grid',
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surfaces=assembly_long_surfs,
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universes=[
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univs['water pin'],
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univs['end plug'],
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univs['Fuel pin (3.1%) no grid'],
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univs['pin plenum'],
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univs['end plug'],
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univs['water pin']
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]
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)
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# Define the NumPy array indices for assembly locations where there
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# may be CR guide tubes, instrument tubes and burnable absorbers
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nonfuel_y = np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
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nonfuel_x = np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
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universes = np.full((17,17), fuel_univ)
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universes[nonfuel_y, nonfuel_x] = univs['GT empty']
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# Instantiate the lattice
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lattice = openmc.RectLattice(name='Pin lattice')
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lattice.lower_left = (-17.*pin_pitch/2., -17.*pin_pitch/2.)
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lattice.pitch = (pin_pitch, pin_pitch)
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lattice.universes = universes
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# Add lattice to bounding cell
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root_universe = openmc.Universe(name='Root universe')
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cell = openmc.Cell(name='Lattice cell')
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cell.fill = lattice
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z_bounds = +surfs['bottom FR'] & -surfs['top FR']
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cell.region = surfs['lat grid box inner'] & z_bounds
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root_universe.add_cell(cell)
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# Apply reflective boundaries on sides and vacuum on bottom/top
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surfs['bottom FR'].boundary_type = 'vacuum'
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surfs['top FR'].boundary_type = 'vacuum'
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for halfspace in surfs['lat grid box inner']:
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halfspace.surface.boundary_type = 'reflective'
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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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# 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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else:
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for mat in cell.fill:
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mat.volume = 1.0
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#### Create OpenMC "materials.xml" file
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print('Getting materials...')
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all_materials = geometry.get_all_materials()
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print('Creating materials collection...')
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materials = openmc.Materials(all_materials.values())
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print('Exporting materials to XML...')
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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 = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
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upper_right = (lattice_pitch/2, 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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if args.multipole:
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settings.temperature = {
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'multipole': True,
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'tolerance': 1000,
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'default': 531.5,
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'method': 'interpolation',
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'range': (500.0, 1300.0)
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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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201
smr/build-assembly-short.py
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201
smr/build-assembly-short.py
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@ -0,0 +1,201 @@
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#!/usr/bin/env python3
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import argparse
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import copy
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from math import pi, isclose
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from pathlib import Path
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import numpy as np
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from tqdm import tqdm
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import openmc
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from smr.materials import materials, mats
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from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
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top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR
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from smr.pins import pin_universes
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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='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=10,
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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('assembly-short-depleted')
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else:
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directory = Path('assembly-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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rings = [0.1*pin_pitch, 0.2*pin_pitch]
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# Define the NumPy array indices for assembly locations where there
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# may be CR guide tubes, instrument tubes and burnable absorbers
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nonfuel_y = np.array([2,2,2,3,3,5,5,5,5,5,8,8,8,8,8,11,11,11,11,11,13,13,14,14,14])
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nonfuel_x = np.array([5,8,11,3,13,2,5,8,11,14,2,5,8,11,14,2,5,8,11,14,3,13,5,8,11])
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# NO BURNABLE ABSORBERS
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pins = pin_universes(rings, args.axial, args.depleted)
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gtu = pins['GT empty']
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#gti = pins['GT empty instr']
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universes = np.empty((17,17), dtype=openmc.Universe)
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universes[:,:] = pins['Fuel pin (3.1%) no grid']
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universes[nonfuel_y, nonfuel_x] = [ gtu, gtu, gtu,
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gtu, gtu,
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gtu, gtu, gtu, gtu, gtu,
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gtu, gtu, gtu, gtu, gtu,
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gtu, gtu, gtu, gtu, gtu,
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gtu, gtu,
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gtu, gtu, gtu ]
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# Instantiate the lattice
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lattice = openmc.RectLattice(name='Pin lattice')
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lattice.lower_left = (-17.*pin_pitch/2., -17.*pin_pitch/2.)
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lattice.pitch = (pin_pitch, pin_pitch)
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lattice.universes = universes
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# Add lattice to bounding cell
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root_universe = openmc.Universe(name='Root universe')
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cell = openmc.Cell(name='Lattice cell')
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cell.fill = lattice
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z_bounds = +surfs['bot active core'] & -surfs['top active core']
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cell.region = surfs['lat grid box inner'] & z_bounds
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root_universe.add_cell(cell)
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# Apply reflective boundaries
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surfs['bot active core'].boundary_type = 'reflective'
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surfs['top active core'].boundary_type = 'reflective'
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for halfspace in surfs['lat grid box inner']:
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halfspace.surface.boundary_type = 'reflective'
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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 = 10.0*pin_pitch / 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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# Determine volume of each fuel material
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if 'UO2 Fuel' in cell.fill[0].name:
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lower_left, _ = cell.region.bounding_box
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if isclose(lower_left[0], rings[0]):
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ri, ro = 0.0, rings[0]
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elif isclose(lower_left[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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elif cell.fill[0].name == 'Borated Water':
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for mat in cell.fill:
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mat.volume = pin_pitch**2 - pi*clad_OR**2 * h
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elif cell.fill[0].name == 'Helium':
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for mat in cell.fill:
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mat.volume = pi * (clad_IR**2 - pellet_OR**2) * h
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elif cell.fill[0].name == 'M5':
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for mat in cell.fill:
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mat.volume = pi * (clad_OR**2 - clad_IR**2) * h
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elif cell.fill[0].name == 'Zircaloy-4':
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for mat in cell.fill:
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mat.volume = pi * (guide_tube_OR**2 - guide_tube_IR**2) * h
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#### Create OpenMC "materials.xml" file
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print('Getting materials...')
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all_materials = geometry.get_all_materials()
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print('Creating materials collection...')
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materials = openmc.Materials(all_materials.values())
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print('Exporting materials to XML...')
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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 = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
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upper_right = (lattice_pitch/2, 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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if args.multipole:
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settings.temperature = {
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'multipole': True,
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'tolerance': 1000,
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'default': 531.5,
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'method': 'interpolation',
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'range': (500.0, 1300.0)
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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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@ -9,7 +9,7 @@ from tqdm import tqdm
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import openmc
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from smr.materials import materials
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from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core
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from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
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from smr.assemblies import assembly_universes
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from smr.plots import assembly_plots
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from smr import inlet_temperature
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@ -17,8 +17,10 @@ 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('-m', '--multipole', action='store_true',
|
||||
help='Whether to use multipole cross sections')
|
||||
parser.add_argument('--multipole', action='store_true',
|
||||
help='Use multipole cross sections')
|
||||
parser.add_argument('--no-multipole', action='store_false',
|
||||
help='Do not use multipole cross sections')
|
||||
parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='mat',
|
||||
help='Whether to use distribmats or distribcells for tallies')
|
||||
parser.add_argument('-r', '--rings', type=int, default=10,
|
||||
|
|
@ -28,6 +30,7 @@ parser.add_argument('-a', '--axial', type=int, default=196,
|
|||
parser.add_argument('-d', '--depleted', action='store_true',
|
||||
help='Whether UO2 compositions should represent depleted fuel')
|
||||
parser.add_argument('-o', '--output-dir', type=Path, default=None)
|
||||
parser.set_defaults(multipole=True)
|
||||
args = parser.parse_args()
|
||||
|
||||
# Make directory for inputs
|
||||
|
|
@ -41,7 +44,11 @@ else:
|
|||
directory.mkdir(exist_ok=True)
|
||||
|
||||
# Define geometry with a single assembly
|
||||
assembly = assembly_universes(args.rings, args.axial, args.depleted)
|
||||
if args.rings > 1:
|
||||
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
|
||||
else:
|
||||
ring_radii = None
|
||||
assembly = assembly_universes(ring_radii, args.axial, args.depleted)
|
||||
lattice_sides = openmc.model.get_rectangular_prism(lattice_pitch, lattice_pitch,
|
||||
boundary_type='reflective')
|
||||
main_cell = openmc.Cell(
|
||||
|
|
|
|||
|
|
@ -11,7 +11,7 @@ import numpy as np
|
|||
import openmc
|
||||
from smr.materials import materials
|
||||
from smr.plots import core_plots
|
||||
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
|
||||
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
|
||||
from smr.core import core_geometry
|
||||
from smr import inlet_temperature
|
||||
|
||||
|
|
@ -25,8 +25,10 @@ def clone(mat):
|
|||
|
||||
# Define command-line options
|
||||
parser = argparse.ArgumentParser()
|
||||
parser.add_argument('-m', '--multipole', action='store_true',
|
||||
help='Whether to use multipole cross sections')
|
||||
parser.add_argument('--multipole', action='store_true',
|
||||
help='Use multipole cross sections')
|
||||
parser.add_argument('--no-multipole', action='store_false',
|
||||
help='Do not use multipole cross sections')
|
||||
parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='cell',
|
||||
help='Whether to use distribmats or distribcells for tallies')
|
||||
parser.add_argument('-r', '--rings', type=int, default=10,
|
||||
|
|
@ -36,6 +38,7 @@ parser.add_argument('-a', '--axial', type=int, default=196,
|
|||
parser.add_argument('-d', '--depleted', action='store_true',
|
||||
help='Whether UO2 compositions should represent depleted fuel')
|
||||
parser.add_argument('-o', '--output-dir', type=Path, default=None)
|
||||
parser.set_defaults(multipole=True)
|
||||
args = parser.parse_args()
|
||||
|
||||
# Make directory for inputs
|
||||
|
|
@ -48,7 +51,11 @@ else:
|
|||
directory = args.output_dir
|
||||
directory.mkdir(exist_ok=True)
|
||||
|
||||
geometry = core_geometry(args.rings, args.axial, args.depleted)
|
||||
if args.rings > 1:
|
||||
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
|
||||
else:
|
||||
ring_radii = None
|
||||
geometry = core_geometry(ring_radii, args.axial, args.depleted)
|
||||
|
||||
#### "Differentiate" the geometry if using distribmats
|
||||
if args.tallies == 'mat':
|
||||
|
|
|
|||
|
|
@ -120,13 +120,14 @@ def make_assembly(name, universes):
|
|||
return universe
|
||||
|
||||
|
||||
def assembly_universes(num_rings, num_axial, depleted):
|
||||
def assembly_universes(ring_radii, num_axial, depleted):
|
||||
"""Generate universes for SMR fuel assemblies.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
num_rings : int
|
||||
Number of annual regions in fuel
|
||||
ring_radii : iterable of float
|
||||
Radii of rings in fuel (note that this doesn't need to include the
|
||||
full fuel pin radius)
|
||||
num_axial : int
|
||||
Number of axial subdivisions in fuel
|
||||
depleted : bool
|
||||
|
|
@ -138,13 +139,13 @@ def assembly_universes(num_rings, num_axial, depleted):
|
|||
Dictionary mapping a universe name to a openmc.Universe object
|
||||
|
||||
"""
|
||||
pins = pin_universes(num_rings, num_axial, depleted)
|
||||
pins = pin_universes(ring_radii, num_axial, depleted)
|
||||
|
||||
# Create dictionary to store assembly universes
|
||||
univs = {}
|
||||
|
||||
# commonly needed universes
|
||||
gtu = pins['GT empty']
|
||||
gtu = pins['GT empty stack']
|
||||
gti = pins['GT empty instr']
|
||||
bas = pins['BA stack']
|
||||
ins = pins['IT stack']
|
||||
|
|
|
|||
|
|
@ -10,13 +10,14 @@ from .reflector import reflector_universes
|
|||
from .assemblies import assembly_universes
|
||||
|
||||
|
||||
def core_geometry(num_rings, num_axial, depleted):
|
||||
def core_geometry(ring_radii, num_axial, depleted):
|
||||
"""Generate full core SMR geometry.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
num_rings : int
|
||||
Number of annual regions in fuel
|
||||
ring_radii : iterable of float
|
||||
Radii of rings in fuel (note that this doesn't need to include the
|
||||
full fuel pin radius)
|
||||
num_axial : int
|
||||
Number of axial subdivisions in fuel
|
||||
depleted : bool
|
||||
|
|
@ -28,7 +29,7 @@ def core_geometry(num_rings, num_axial, depleted):
|
|||
SMR full core geometry
|
||||
|
||||
"""
|
||||
assembly = assembly_universes(num_rings, num_axial, depleted)
|
||||
assembly = assembly_universes(ring_radii, num_axial, depleted)
|
||||
reflector = reflector_universes()
|
||||
|
||||
# Construct main core lattice
|
||||
|
|
|
|||
|
|
@ -192,7 +192,6 @@ aB_bsg = aB10_bsg + aB11_bsg
|
|||
|
||||
# Create borosilicate glass material
|
||||
mats['BSG'] = openmc.Material(name='Borosilicate Glass')
|
||||
mats['BSG'].temperature = 300
|
||||
mats['BSG'].set_density('g/cc', 2.26)
|
||||
mats['BSG'].add_element('O', aO_bsg, 'ao')
|
||||
mats['BSG'].add_element('Si', aSi_bsg, 'ao')
|
||||
|
|
@ -205,7 +204,6 @@ mats['BSG'].add_nuclide('B11', aB11_bsg, 'ao')
|
|||
|
||||
# Create 1.6% enriched UO2 fuel material
|
||||
mat = openmc.Material(name='1.6% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.31341)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=1.61006)
|
||||
|
|
@ -213,7 +211,6 @@ mats['UO2 1.6 fresh'] = mat
|
|||
|
||||
# Create 2.4% enriched UO2 fuel material
|
||||
mat = openmc.Material(name='2.4% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.29748)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=2.39993)
|
||||
|
|
@ -221,7 +218,6 @@ mats['UO2 2.4 fresh'] = mat
|
|||
|
||||
# Create 3.1% enriched UO2 fuel material
|
||||
mat = openmc.Material(name='3.1% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.30166)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=3.10221)
|
||||
|
|
@ -229,7 +225,6 @@ mats['UO2 3.1 fresh'] = mat
|
|||
|
||||
# Depleted versions of 1.6%, 2.4%, 3.1% fuel
|
||||
mat = openmc.Material(name='2.4% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.29748)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=2.39993)
|
||||
|
|
@ -238,7 +233,6 @@ for nuc in _DEPLETION_NUCLIDES:
|
|||
mats['UO2 2.4 depleted'] = mat
|
||||
|
||||
mat = openmc.Material(name='1.6% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.31341)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=1.61006)
|
||||
|
|
@ -247,7 +241,6 @@ for nuc in _DEPLETION_NUCLIDES:
|
|||
mats['UO2 1.6 depleted'] = mat
|
||||
|
||||
mat = openmc.Material(name='3.1% Enr. UO2 Fuel')
|
||||
mat.temperature = 300
|
||||
mat.set_density('g/cc', 10.30166)
|
||||
mat.add_element('O', 2., 'ao')
|
||||
mat.add_element('U', 1., 'ao', enrichment=3.10221)
|
||||
|
|
|
|||
|
|
@ -142,13 +142,14 @@ def make_pin_stack(name, zsurfaces, universes, boundary, fuel_fill):
|
|||
return universe
|
||||
|
||||
|
||||
def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
||||
def pin_universes(ring_radii=None, num_axial=196, depleted=False):
|
||||
"""Generate universes for SMR fuel pins.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
num_rings : int
|
||||
Number of annual regions in fuel
|
||||
ring_radii : iterable of float
|
||||
Radii of rings in fuel (note that this doesn't need to include the
|
||||
full fuel pin radius)
|
||||
num_axial : int
|
||||
Number of axial subdivisions in fuel
|
||||
depleted : bool
|
||||
|
|
@ -230,8 +231,8 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
surfs['top upper nozzle']
|
||||
]
|
||||
|
||||
univs['GT empty'] = make_stack(
|
||||
'GT empty', surfaces=stack_surfs,
|
||||
univs['GT empty stack'] = make_stack(
|
||||
'GT empty stack', surfaces=stack_surfs,
|
||||
universes=[univs['water pin'],
|
||||
univs['water pin'],
|
||||
univs['water pin'],
|
||||
|
|
@ -706,12 +707,11 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
axial_splits = np.linspace(bottom_fuel_stack, top_active_core, num_axial + 1)[1:-1]
|
||||
axial_surfs = [openmc.ZPlane(z0=z) for z in axial_splits]
|
||||
|
||||
if num_rings > 1:
|
||||
if ring_radii is not None:
|
||||
# Get z-cylinder surfaces for each ring
|
||||
rings = []
|
||||
for i in range(1, num_rings):
|
||||
R = sqrt(i*pellet_OR**2/num_rings)
|
||||
cyl = openmc.ZCylinder(r=R, name='fuel ring {}'.format(i))
|
||||
for i, r in enumerate(ring_radii):
|
||||
cyl = openmc.ZCylinder(r=r, name='fuel ring {}'.format(i))
|
||||
rings.append(cyl)
|
||||
|
||||
def subdivided_fuel(fill):
|
||||
|
|
@ -719,14 +719,14 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
uo2_cells = []
|
||||
if num_axial > 1:
|
||||
for axial_region in subdivide(axial_surfs):
|
||||
if num_rings > 1:
|
||||
if ring_radii is not None:
|
||||
for ring_region in subdivide(rings):
|
||||
cell = openmc.Cell(fill=fill, region=axial_region & ring_region)
|
||||
uo2_cells.append(cell)
|
||||
else:
|
||||
uo2_cells.append(openmc.Cell(fill=fill, region=axial_region))
|
||||
else:
|
||||
if num_rings > 1:
|
||||
if ring_radii is not None:
|
||||
for ring_region in subdivide(rings):
|
||||
cell = openmc.Cell(fill=fill, region=ring_region)
|
||||
uo2_cells.append(cell)
|
||||
|
|
@ -737,7 +737,7 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
|
||||
# If rings/axial segments are present, create a universe for the subdivided
|
||||
# fuel. Otherwise just use a plain material.
|
||||
if num_rings > 1 or num_axial > 1:
|
||||
if ring_radii is not None or num_axial > 1:
|
||||
fuel_fill = subdivided_fuel(mats['UO2 1.6 {}'.format(fuel)])
|
||||
else:
|
||||
fuel_fill = mats['UO2 1.6 {}'.format(fuel)]
|
||||
|
|
@ -762,6 +762,12 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
materials=outside_pin_mats,
|
||||
grid='intermediate')
|
||||
|
||||
univs['Fuel pin (1.6%) no grid'] = make_pin(
|
||||
'Pin no grid',
|
||||
surfaces=[surfs['pellet OR']] + outside_pin_surfaces,
|
||||
materials=[fuel_fill] + outside_pin_mats
|
||||
)
|
||||
|
||||
# Stack all axial pieces of 1.6% enriched fuel pin cell
|
||||
|
||||
within_fuel_surfs = [
|
||||
|
|
@ -829,11 +835,17 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
|
||||
# If rings/axial segments are present, create a universe for the subdivided
|
||||
# fuel. Otherwise just use a plain material.
|
||||
if num_rings > 1 or num_axial > 1:
|
||||
if ring_radii is not None or num_axial > 1:
|
||||
fuel_fill = subdivided_fuel(mats['UO2 2.4 {}'.format(fuel)])
|
||||
else:
|
||||
fuel_fill = mats['UO2 2.4 {}'.format(fuel)]
|
||||
|
||||
univs['Fuel pin (2.4%) no grid'] = make_pin(
|
||||
'Pin no grid',
|
||||
surfaces=[surfs['pellet OR']] + outside_pin_surfaces,
|
||||
materials=[fuel_fill] + outside_pin_mats
|
||||
)
|
||||
|
||||
# Stack all axial pieces of 2.4% enriched fuel pin cell
|
||||
|
||||
univs['Fuel pin (2.4%) stack'] = make_pin_stack(
|
||||
|
|
@ -875,11 +887,26 @@ def pin_universes(num_rings=10, num_axial=196, depleted=False):
|
|||
|
||||
# If rings/axial segments are present, create a universe for the subdivided
|
||||
# fuel. Otherwise just use a plain material.
|
||||
if num_rings > 1 or num_axial > 1:
|
||||
if ring_radii is not None or num_axial > 1:
|
||||
fuel_fill = subdivided_fuel(mats['UO2 3.1 {}'.format(fuel)])
|
||||
else:
|
||||
fuel_fill = mats['UO2 3.1 {}'.format(fuel)]
|
||||
|
||||
if num_axial > 1:
|
||||
water_cells = []
|
||||
for i, r in enumerate(subdivide(axial_surfs)):
|
||||
cell = openmc.Cell(fill=mats['H2O'], region=r, name=f'Water ({i})')
|
||||
water_cells.append(cell)
|
||||
water_fill = openmc.Universe(cells=water_cells)
|
||||
else:
|
||||
water_fill = mats['H2O']
|
||||
|
||||
univs['Fuel pin (3.1%) no grid'] = make_pin(
|
||||
'Pin no grid',
|
||||
surfaces=[surfs['pellet OR']] + outside_pin_surfaces,
|
||||
materials=[fuel_fill, mats['He'], mats['M5'], water_fill]
|
||||
)
|
||||
|
||||
# Stack all axial pieces of 3.1% enriched fuel pin cell
|
||||
|
||||
univs['Fuel pin (3.1%) stack'] = make_pin_stack(
|
||||
|
|
|
|||
|
|
@ -83,7 +83,7 @@ rpv_IR = 96.5*INCHES/2 # ML17013A274, Table 5.3-1
|
|||
rpv_OR = 105*INCHES/2 # ML17013A274, Table 5.3-1
|
||||
|
||||
# axial parameters
|
||||
reference_z = 0.0
|
||||
reference_z = -36.6205
|
||||
lowest_extent = reference_z
|
||||
bottom_support_plate = lowest_extent + 20.000
|
||||
top_support_plate = bottom_support_plate + 5.000
|
||||
|
|
@ -166,8 +166,8 @@ surfs['BA IR 7'] = openmc.ZCylinder(
|
|||
r=burn_abs_r7, name='BA IR 7')
|
||||
surfs['BA IR 8'] = openmc.ZCylinder(
|
||||
r=burn_abs_r8, name='BA IR 8')
|
||||
surfs['IT IR'] = copy.deepcopy(surfs['BA IR 5'])
|
||||
surfs['IT OR'] = copy.deepcopy(surfs['BA IR 6'])
|
||||
surfs['IT IR'] = surfs['BA IR 5']
|
||||
surfs['IT OR'] = surfs['BA IR 6']
|
||||
|
||||
# Rectangular prisms for grid spacers
|
||||
surfs['rod grid box'] = \
|
||||
|
|
@ -184,8 +184,8 @@ surfs['bot support plate'] = openmc.ZPlane(
|
|||
surfs['top support plate'] = openmc.ZPlane(
|
||||
z0=top_support_plate, name='top support plate')
|
||||
surfs['bottom FR'] = openmc.ZPlane(z0=bottom_fuel_rod, name='bottom FR')
|
||||
surfs['top lower nozzle'] = copy.deepcopy(surfs['bottom FR'])
|
||||
surfs['bot lower nozzle'] = copy.deepcopy(surfs['top support plate'])
|
||||
surfs['top lower nozzle'] = surfs['bottom FR']
|
||||
surfs['bot lower nozzle'] = surfs['top support plate']
|
||||
|
||||
# axial surfaces
|
||||
surfs['bot active core'] = openmc.ZPlane(
|
||||
|
|
@ -193,7 +193,7 @@ surfs['bot active core'] = openmc.ZPlane(
|
|||
surfs['top active core'] = openmc.ZPlane(
|
||||
z0=top_active_core, name='top active core')
|
||||
|
||||
surfs['top lower thimble'] = copy.deepcopy(surfs['bot active core'])
|
||||
surfs['top lower thimble'] = surfs['bot active core']
|
||||
surfs['BA bot'] = openmc.ZPlane(
|
||||
z0=bot_burn_abs, name='bottom of BA')
|
||||
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue