forked from crp/ecp-benchmarks
Add script to build long assembly
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Paul Romano
parent
0e409cba87
commit
2a24952857
2 changed files with 205 additions and 4 deletions
203
smr/build-assembly-long.py
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203
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('-m', '--multipole', action='store_true',
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help='Whether to 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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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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@ -37,7 +37,7 @@ pellet_OR = 0.3195*INCHES/2 # ML17013A274, Table 4.1-2
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pellet_length = 0.4*INCHES # ML17013A274, Table 4.1-2
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clad_IR = 0.326*INCHES/2 # ML17013A274, Table 4.1-2
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clad_OR = 0.374*INCHES/2 # ML17013A274, Table 4.1-2
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#active_fuel_length = 78.74*INCHES # ML17013A274, Figure 4.2-10
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active_fuel_length = 78.74*INCHES # ML17013A274, Figure 4.2-10
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plenum_length = 5.311*INCHES # ML17013A274, Figure 4.2-10
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fuel_rod_length = 85.00*INCHES # ML17013A274, Table 4.1-2
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lower_end_cap_length = 0.575*INCHES # ML17007A001, Table 3-2
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@ -75,8 +75,6 @@ grid_strap_side = 21.47270
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top_nozzle_height = 3.551*INCHES # ML17013A274, Figure 4.2-2
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top_nozzle_width = 8.406*INCHES # ML17013A274, Figure 4.2-2
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active_fuel_length = 10.0*pin_pitch
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# core radial parameters
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core_barrel_IR = 74*INCHES/2 # ML17013A274, Table 4.1-2
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core_barrel_OR = 78*INCHES/2 # ML17013A274, Table 4.1-2
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@ -85,7 +83,7 @@ rpv_IR = 96.5*INCHES/2 # ML17013A274, Table 5.3-1
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rpv_OR = 105*INCHES/2 # ML17013A274, Table 5.3-1
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# axial parameters
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reference_z = -36.007
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reference_z = -36.6205
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lowest_extent = reference_z
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bottom_support_plate = lowest_extent + 20.000
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top_support_plate = bottom_support_plate + 5.000
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