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Allow configurable fuel in ring radii

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This commit is contained in:
Paul Romano 2019-12-19 07:38:41 -06:00
parent 314331f275
commit 5481118862
5 changed files with 37 additions and 24 deletions
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5
smr/build-assembly.py
View file

@ -9,7 +9,7 @@ from tqdm import tqdm
import openmc
from smr.materials import materials
from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core
from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
from smr.assemblies import assembly_universes
from smr.plots import assembly_plots
from smr import inlet_temperature
@ -41,7 +41,8 @@ else:
directory.mkdir(exist_ok=True)
# Define geometry with a single assembly
assembly = assembly_universes(args.rings, args.axial, args.depleted)
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
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(

5
smr/build-core-fresh.py
View file

@ -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
@ -48,7 +48,8 @@ else:
directory = args.output_dir
directory.mkdir(exist_ok=True)
geometry = core_geometry(args.rings, args.axial, args.depleted)
ring_radii = np.sqrt(np.arange(1, args.rings)*pellet_OR**2 / args.rings)
geometry = core_geometry(ring_radii, args.axial, args.depleted)
#### "Differentiate" the geometry if using distribmats
if args.tallies == 'mat':

9
smr/smr/assemblies.py
View file

@ -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
@ -138,7 +139,7 @@ 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 = {}

9
smr/smr/core.py
View file

@ -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

33
smr/smr/pins.py
View file

@ -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
@ -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)]
@ -829,7 +829,7 @@ 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 2.4 {}'.format(fuel)])
else:
fuel_fill = mats['UO2 2.4 {}'.format(fuel)]
@ -875,11 +875,20 @@ 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']
# Stack all axial pieces of 3.1% enriched fuel pin cell
univs['Fuel pin (3.1%) stack'] = make_pin_stack(