def make_firstwall_material(firstwall_armour_material,
firstwall_armour_fraction,
firstwall_coolant_material,
firstwall_coolant_fraction,
firstwall_structural_material,
firstwall_structural_fraction):
if firstwall_coolant_material == 'He':
temperature_in_C = 400
pressure_in_Pa = 8e6
elif firstwall_coolant_material in ['H2O', 'D2O']:
temperature_in_C = 305
pressure_in_Pa = 15.5e6
firstwall_material = MultiMaterial(
material_name='firstwall_material',
materials=[
Material(material_name=firstwall_armour_material),
Material(material_name=firstwall_coolant_material,
temperature_in_C=temperature_in_C,
pressure_in_Pa=pressure_in_Pa),
Material(material_name=firstwall_structural_material)
],
fracs=[
firstwall_armour_fraction, firstwall_coolant_fraction,
firstwall_structural_fraction
]).neutronics_material
return firstwall_material
def make_other_aspects_of_neutronics_model():
"""
Makes and runs a simple OpenMC neutronics model with
the materials with the same tags as the DAGMC neutronics
geometry. The model also specifies the computational
intensity (particles and batches) and the tally to record
"""
universe = openmc.Universe()
geom = openmc.Geometry(universe)
mat1 = Material(
material_name="Li4SiO4", material_tag="blanket_material"
).openmc_material
mat2 = Material(
material_name="copper", material_tag="pf_coil_material"
).openmc_material
mats = openmc.Materials([mat1, mat2])
settings = openmc.Settings()
settings.batches = 10
settings.inactive = 0
settings.particles = 100
settings.run_mode = "fixed source"
settings.dagmc = True
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
settings.source = source
tallies = openmc.Tallies()
tbr_tally = openmc.Tally(name="TBR")
tbr_tally.scores = ["(n,Xt)"] # where X is a wild card
tallies.append(tbr_tally)
model = openmc.model.Model(geom, mats, settings, tallies)
output_filename = model.run()
return output_filename
def make_blanket_material(
blanket_structural_material, blanket_structural_fraction,
blanket_coolant_material, blanket_coolant_fraction,
blanket_multiplier_material, blanket_multiplier_fraction,
blanket_breeder_material, blanket_breeder_fraction,
blanket_breeder_li6_enrichment_fraction,
blanket_breeder_packing_fraction, blanket_multiplier_packing_fraction):
if blanket_coolant_material == 'He':
temperature_in_C = 400
pressure_in_Pa = 8e6
elif blanket_coolant_material in ['H2O', 'D2O']:
temperature_in_C = 305
pressure_in_Pa = 15.5e6
blanket_material = MultiMaterial(
material_name='blanket_material',
materials=[
Material(material_name=blanket_structural_material),
Material(material_name=blanket_coolant_material,
temperature_in_C=temperature_in_C,
pressure_in_Pa=pressure_in_Pa),
Material(material_name=blanket_multiplier_material,
packing_fraction=blanket_multiplier_packing_fraction),
Material(
material_name=blanket_breeder_material,
enrichment_fraction=blanket_breeder_li6_enrichment_fraction,
temperature_in_C=500,
packing_fraction=blanket_breeder_packing_fraction),
],
fracs=[
blanket_structural_fraction, blanket_coolant_fraction,
blanket_multiplier_fraction, blanket_breeder_fraction
],
percent_type='vo').neutronics_material
return blanket_material
#!/usr/bin/env python3
"""example_CAD_simulation.py: uses a dagmc.h5m file for the geometry."""
__author__ = "Jonathan Shimwell"
import openmc
import json
import os
from neutronics_material_maker import Material
from parametric_plasma_source import PlasmaSource, SOURCE_SAMPLING_PATH
# MATERIALS using the neutronics material maker
breeder_material = Material(material_name='Li4SiO4',
enrichment=90,
material_tag='blanket_material'
).openmc_material
copper = Material(material_name="copper",
material_tag="copper"
).openmc_material
eurofer = Material(material_name='eurofer',
material_tag="eurofer").openmc_material
mats = openmc.Materials([breeder_material, eurofer, copper])
# GEOMETRY using dagmc doesn't contain any CSG geometry
universe = openmc.Universe()
def make_materials_geometry_tallies(v):
enrichment_fraction, thickness = v
inner_radius = 500
breeder_material_name = 'Li'
temperature_in_C = 500
print('simulating enrichment,', enrichment_fraction, 'thickness ',
thickness)
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment_fraction=enrichment_fraction,
temperature_in_C=temperature_in_C).neutronics_material
eurofer = Material(material_name='eurofer').neutronics_material
mats = openmc.Materials([breeder_material, eurofer])
# GEOMETRY
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius)
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius + thickness)
vessel_inner_surface = openmc.Sphere(r=inner_radius + thickness + 10)
vessel_outer_surface = openmc.Sphere(r=inner_radius + thickness + 20,
boundary_type='vacuum')
breeder_blanket_region = -breeder_blanket_outer_surface & +breeder_blanket_inner_surface
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
breeder_blanket_cell.name = 'breeder_blanket'
inner_void_region = -breeder_blanket_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
vessel_region = +vessel_inner_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = 'vessel'
vessel_cell.fill = eurofer
blanket_vessel_gap_region = -vessel_inner_surface & +breeder_blanket_outer_surface
blanket_vessel_gap_cell = openmc.Cell(region=blanket_vessel_gap_region)
blanket_vessel_gap_cell.name = 'blanket_vessel_gap'
universe = openmc.Universe(cells=[
inner_void_cell, breeder_blanket_cell, blanket_vessel_gap_cell,
vessel_cell
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS
sett = openmc.Settings()
# batches = 3 # this is parsed as an argument
sett.batches = batches
sett.inactive = 0
sett.particles = 500
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(
e0=14080000.0, m_rat=5.0, kt=20000.0
) # neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
# TALLIES
tallies = openmc.Tallies()
# define filters
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
cell_filter_vessel = openmc.CellFilter(vessel_cell)
particle_filter = openmc.ParticleFilter([1]) # 1 is neutron, 2 is photon
surface_filter_rear_blanket = openmc.SurfaceFilter(
breeder_blanket_outer_surface)
surface_filter_rear_vessel = openmc.SurfaceFilter(vessel_outer_surface)
energy_bins = openmc.mgxs.GROUP_STRUCTURES['VITAMIN-J-175']
energy_filter = openmc.EnergyFilter(energy_bins)
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = [
'(n,Xt)'
] # MT 205 is the (n,Xt) reaction where X is a wildcard, if MT 105 or (n,t) then some tritium production will be missed, for example (n,nt) which happens in Li7 would be missed
tallies.append(tally)
# RUN OPENMC
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
json_output = {
'enrichment_fraction': enrichment_fraction,
'inner_radius': inner_radius,
'thickness': thickness,
'breeder_material_name': breeder_material_name,
'temperature_in_C': temperature_in_C
}
tally = sp.get_tally(name='TBR')
df = tally.get_pandas_dataframe()
json_output['TBR'] = df['mean'].sum()
json_output['TBR_std_dev'] = df['std. dev.'].sum()
return json_output
import openmc
import numpy as np
import plotly.graph_objs as go
from neutronics_material_maker import Material, MultiMaterial
# Homogeneous mixture of helium and Li4SiO4 using MultiMaterial class
# MultiMaterial class requires list of Material objects to be passed
helium = Material('He', temperature_in_C=500, pressure_in_Pa=100000)
Li4SiO4 = Material('Li4SiO4', enrichment=60)
mixed_helium_Li4SiO4 = MultiMaterial(
material_name='mixed_helium_Li4SiO4', # name of homogeneous material
materials=[helium,
Li4SiO4], # list of material objects (NOT neutronics materials)
fracs=[0.36,
0.64], # list of combination fractions for each material object
percent_type='vo') # combination fraction type
# print(mixed_helium_Li4SiO4.neutronics_material)
# Homogenous mixture of tungsten carbide and water using mix_materials function
mixed_water_WC = openmc.Material.mix_materials(
name='mixed_water_WC', # name of homogeneous material
materials=[ # list of neutronics materials
Material('WC').neutronics_material,
Material('H2O', temperature_in_C=25,
pressure_in_Pa=100000).neutronics_material
],
fracs=[0.8,
0.2], # list of combination fractions for each neutronics material
import numpy as np
import plotly.graph_objs as go
from plotly.subplots import make_subplots
from neutronics_material_maker import Material
# Because the neutronics_material_maker makes it easy to adjust material properties, we can easily perform parameter studies
# Water density as a function of temperature (at constant pressure)
# Based on WCLL, pressure = 15.5 MPa, inlet and outlet temperatures = 285 and 325 degrees C
# We will show density as a function of temperature over a larger range, but at correct pressure
temperatures = np.linspace(0.1, 600., 200)
water_densities = [Material('H2O', temperature_in_C=temperature, pressure_in_Pa=15500000).neutronics_material.density for temperature in temperatures]
fig = make_subplots(rows=2, cols=2,
subplot_titles=("Water density as a function of temperature (at constant pressure)", "Helium density as a function of temperature (at constant pressure)", "Helium density as a function of pressure (at constant temperature)", ""))
fig.add_trace(go.Scatter(x=temperatures,
y=water_densities,
mode='lines+markers',
showlegend=False),
row=1, col=1)
fig.update_xaxes({'title': 'Temperature in C'}, row=1, col=1)
fig.update_yaxes({'title': 'Density (g/cm3)'}, row=1, col=1)
# Helium density as a function of temperature (at constant pressure)
def make_other_aspects_of_neutronics_model(my_reactor):
"""
Makes and runs a simple OpenMC neutronics model with
the materials with the same tags as the DAGMC neutronics
geometry. The model also specifies the computational
intensity (particles and batches) and the tally to record
"""
# these materials are overly simplified to keep the example short
firstwall_mat = Material(material_name='eurofer',
material_tag='firstwall_mat').openmc_material
inboard_tf_coils_mat = Material(
material_name='WC',
material_tag='inboard_tf_coils_mat').openmc_material
center_column_mat = Material(
material_name='WC',
material_tag='center_column_shield_mat').openmc_material
divertor_mat = Material(material_name='eurofer',
material_tag='divertor_mat').openmc_material
blanket_mat = Material(material_name='Li4SiO4',
enrichment=60,
material_tag='blanket_mat').openmc_material
blanket_rear_wall_mat = Material(
material_name='eurofer',
material_tag='blanket_rear_wall_mat').openmc_material
mats = openmc.Materials([
firstwall_mat,
inboard_tf_coils_mat,
center_column_mat,
divertor_mat,
blanket_mat,
blanket_rear_wall_mat,
])
# this is the underlying geometry container that is filled with the faceteted CAD model
universe = openmc.Universe()
geom = openmc.Geometry(universe)
# settings for the number of neutrons to simulate
settings = openmc.Settings()
settings.batches = 10
settings.inactive = 0
settings.particles = 100
settings.run_mode = 'fixed source'
settings.dagmc = True
# details of the birth locations and energy of the neutronis
source = openmc.Source()
source.space = openmc.stats.Point((my_reactor.major_radius, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
settings.source = source
# details about what neutrons interactions to keep track of (called a tally)
tallies = openmc.Tallies()
material_filter = openmc.MaterialFilter(blanket_mat)
tbr_tally = openmc.Tally(name='TBR')
tbr_tally.filters = [material_filter]
tbr_tally.scores = ['(n,Xt)'] # where X is a wild card
tallies.append(tbr_tally)
# make the model from gemonetry, materials, settings and tallies
model = openmc.model.Model(geom, mats, settings, tallies)
# run the simulation
output_filename = model.run()
return output_filename
#!/usr/bin/env python3
"""example_CAD_simulation.py: uses a dagmc.h5m file for the geometry."""
__author__ = "Jonathan Shimwell"
import openmc
import json
import os
from neutronics_material_maker import Material
from parametric_plasma_source import Plasma
# MATERIALS using the neutronics material maker
breeder_material = Material(material_name='Li4SiO4',
enrichment=90).openmc_material
copper = Material(material_name="copper").openmc_material
eurofer = Material(material_name='eurofer').openmc_material
mats = openmc.Materials([breeder_material, eurofer, copper])
# GEOMETRY using dagmc doesn't contain any CSG geometry
universe = openmc.Universe()
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS
# Instantiate a Settings object
sett = openmc.Settings()
def find_tbr_from_graded_blanket(
number_of_layers, layer_thickness_fractions, layer_li6_enrichments,
layer_breeder_fractions, layer_multiplier_fractions,
blanket_structural_material, blanket_multiplier_material,
blanket_breeder_material, firstwall_coolant,
blanket_structural_fraction, inner_radius, thickness,
firstwall_thickness):
batches = 10
thickness_fraction_scaler = thickness / sum(layer_thickness_fractions)
print('thickness_fraction_scaler', thickness_fraction_scaler)
if firstwall_coolant == 'no firstwall':
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius)
inner_void_region = -breeder_blanket_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
surfaces = [breeder_blanket_inner_surface]
additional_thickness = 0
all_layers_materials = []
all_cells = [inner_void_cell]
for i, (layer_thickness, layer_enrichment, layer_breeder_fraction,
layer_multiplier_fraction) in enumerate(
zip(layer_thickness_fractions, layer_li6_enrichments,
layer_breeder_fractions, layer_multiplier_fractions)):
print(i, layer_thickness, layer_enrichment, layer_breeder_fraction)
layer_material = MultiMaterial(
material_name='layer_' + str(i) + '_material',
materials=[
Material(blanket_breeder_material,
enrichment_fraction=layer_enrichment),
Material(blanket_multiplier_material),
Material(blanket_structural_material)
],
fracs=[
layer_breeder_fraction, layer_multiplier_fraction,
blanket_structural_fraction
]).neutronics_material
all_layers_materials.append(layer_material)
additional_thickness = additional_thickness + (
layer_thickness * thickness_fraction_scaler)
print('additional_thickness', additional_thickness)
if i == number_of_layers - 1:
layer_outer_surface = openmc.Sphere(r=inner_radius +
additional_thickness,
boundary_type='vacuum')
else:
layer_outer_surface = openmc.Sphere(r=inner_radius +
additional_thickness)
layer_inner_surface = surfaces[-1]
layer_region = -layer_outer_surface & +layer_inner_surface
layer_cell = openmc.Cell(region=layer_region)
layer_cell.fill = layer_material
layer_cell.name = 'layer_' + str(i) + '_cell'
surfaces.append(layer_outer_surface)
all_cells.append(layer_cell)
else:
#firstwall is present
firstwall_material = MultiMaterial(
material_name='firstwall_material',
materials=[
Material('tungsten'),
Material(firstwall_coolant),
Material(blanket_structural_material)
],
fracs=[0.055262, 0.253962,
0.690776] #based on HCPB paper with 2mm W firstwall
).neutronics_material
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius +
firstwall_thickness)
firstwall_outer_surface = openmc.Sphere(r=inner_radius)
surfaces = [firstwall_outer_surface, breeder_blanket_inner_surface]
inner_void_region = -firstwall_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
firstwall_region = +firstwall_outer_surface & -breeder_blanket_inner_surface
firstwall_cell = openmc.Cell(region=firstwall_region)
firstwall_cell.fill = firstwall_material
firstwall_cell.name = 'firstwall'
additional_thickness = 0
all_layers_materials = [firstwall_material]
all_cells = [inner_void_cell, firstwall_cell]
for i, (layer_thickness, layer_enrichment, layer_breeder_fraction,
layer_multiplier_fraction) in enumerate(
zip(layer_thickness_fractions, layer_li6_enrichments,
layer_breeder_fractions, layer_multiplier_fractions)):
print(i, layer_thickness, layer_enrichment, layer_breeder_fraction)
layer_material = MultiMaterial(
'layer_' + str(i) + '_material',
materials=[
Material(blanket_breeder_material,
enrichment_fraction=layer_enrichment),
Material(blanket_multiplier_material),
Material(blanket_structural_material)
],
volume_fractions=[
layer_breeder_fraction, layer_multiplier_fraction,
blanket_structural_fraction
])
all_layers_materials.append(layer_material)
additional_thickness = additional_thickness + (
layer_thickness * thickness_fraction_scaler)
print('additional_thickness', additional_thickness)
if i == number_of_layers - 1:
layer_outer_surface = openmc.Sphere(r=inner_radius +
firstwall_thickness +
additional_thickness,
boundary_type='vacuum')
else:
layer_outer_surface = openmc.Sphere(r=inner_radius +
firstwall_thickness +
additional_thickness)
layer_inner_surface = surfaces[-1]
layer_region = -layer_outer_surface & +layer_inner_surface
layer_cell = openmc.Cell(region=layer_region)
layer_cell.fill = layer_material
layer_cell.name = 'layer_' + str(i) + '_cell'
surfaces.append(layer_outer_surface)
all_cells.append(layer_cell)
universe = openmc.Universe(cells=all_cells)
geom = openmc.Geometry(universe)
mats = openmc.Materials(all_layers_materials)
sett = openmc.Settings()
# batches = 3 # this is parsed as an argument
sett.batches = batches
sett.inactive = 10
sett.particles = 500
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((0, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(
e0=14080000.0, m_rat=5.0, kt=20000.0
) #neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
#TALLIES#
tallies = openmc.Tallies()
# define filters
# cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
particle_filter = openmc.ParticleFilter(['neutron'
]) #1 is neutron, 2 is photon
tally = openmc.Tally(name='TBR')
tally.filters = [particle_filter]
tally.scores = ['(n,Xt)'] #could be (n,Xt)
tallies.append(tally)
model = openmc.model.Model(geom, mats, sett, tallies)
model.export_to_xml()
# model.run()
openmc.lib.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
tally = sp.get_tally(name='TBR')
df = tally.get_pandas_dataframe()
return df['mean'].sum()
#!/usr/bin/env python3
"""example_CAD_simulation.py: uses a dagmc.h5m file for the geometry."""
__author__ = "Jonathan Shimwell"
import openmc
import json
import os
from neutronics_material_maker import Material
# MATERIALS using the neutronics material maker
breeder_material = Material(material_name='Li4SiO4',
enrichment=90).neutronics_material
copper = Material(material_name="copper").neutronics_material
eurofer = Material(material_name='eurofer').neutronics_material
mats = openmc.Materials([breeder_material, eurofer, copper])
# GEOMETRY using dagmc doesn't contain any CSG geometry
universe = openmc.Universe()
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS
# Instantiate a Settings object
sett = openmc.Settings()
batches = 10
def sphere_with_firstwall_model(
material_for_structure,
blanket_breeder_material,
blanket_coolant_material,
firstwall_coolant_material,
blanket_breeder_li6_enrichment, # this is a percentage
coolant_pressure, # this is in Pa
blanket_coolant_temperature_in_C,
firstwall_coolant_temperature_in_C,
blanket_breeder_fraction,
blanket_coolant_fraction,
blanket_structural_fraction,
blanket_breeder_temperature_in_C = None, #needed for liquid breeders like lithium lead
firstwall_thickness = 2.7, # this is in cm
blanket_thickness = 200, # this is in cm
inner_radius = 1000,
firstwall_armour_fraction = 0.106305, # equivilent to 3mm and based on https://doi.org/10.1016/j.fusengdes.2017.02.008
firstwall_coolant_fraction= 0.333507, # based on https://doi.org/10.1016/j.fusengdes.2017.02.008
firstwall_structural_fraction = 0.560188, # based on https://doi.org/10.1016/j.fusengdes.2017.02.008
blanket_multipler_material = None, #used for combined breeder multiplier options
blanket_multiplier_fraction = None, #used for combined breeder multiplier options
blanket_breeder_material_packing_fraction = None, #used for combined breeder multiplier options
blanket_multiplier_packing_fraction = None, #used for combined breeder multiplier options
blanket_multiplier_material = None #used for combined breeder multiplier options
):
breeder_percent_in_breeder_plus_multiplier_ratio = 100 * (blanket_breeder_fraction / (blanket_breeder_fraction + blanket_multiplier_fraction))
inputs = locals()
"""
This function builds materials for the homogenised blanket material, homogenised firstwall material
The creates a simple sphere geometry with a simple point source and TBR tally on the blanket
The function also carries out the simulation and writes the results to a JSON file
"""
# creates homogensied blanket material using a single breeder / multiplier material (e.g lithium lead)
if blanket_breeder_material == 'Pb842Li158':
blanket_material = MultiMaterial(material_tag = 'blanket_material',
materials = [
Material(material_name = material_for_structure),
Material(material_name = blanket_coolant_material,
temperature_in_C = blanket_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = blanket_breeder_material,
enrichment = blanket_breeder_li6_enrichment,
temperature_in_C = blanket_breeder_temperature_in_C),
],
fracs = [blanket_structural_fraction,
blanket_coolant_fraction,
blanket_breeder_fraction],
percent_type='vo'
).openmc_material
# creates homogensied blanket material using a combined breeder multiplier material (e.g lithium ceramic with be multiplier)
else:
blanket_material = MultiMaterial(
material_tag = 'blanket_material',
materials = [
Material(material_name = material_for_structure),
Material(material_name = blanket_coolant_material,
temperature_in_C = blanket_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = blanket_breeder_material,
enrichment = blanket_breeder_li6_enrichment,
packing_fraction = blanket_breeder_material_packing_fraction),
Material(material_name = blanket_multipler_material,
packing_fraction = blanket_multiplier_packing_fraction),
],
fracs = [blanket_structural_fraction,
blanket_coolant_fraction,
blanket_breeder_fraction,
blanket_multiplier_fraction],
percent_type='vo'
).openmc_material
# creates homogensied firstwall material with eurofer, tungsten and a coolant
firstwall_material = MultiMaterial(material_tag = 'firstwall_material',
materials = [
Material(material_name = 'tungsten'),
Material(material_name = firstwall_coolant_material,
temperature_in_C = firstwall_coolant_temperature_in_C,
pressure_in_Pa = coolant_pressure),
Material(material_name = 'eurofer')],
fracs = [firstwall_armour_fraction,
firstwall_coolant_fraction,
firstwall_structural_fraction]
).openmc_material
mats = openmc.Materials([blanket_material, firstwall_material])
# creates surfaces
breeder_blanket_inner_surface = openmc.Sphere(r=inner_radius+firstwall_thickness)
firstwall_outer_surface = openmc.Sphere(r=inner_radius)
inner_void_region = -firstwall_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = 'inner_void'
firstwall_region = +firstwall_outer_surface & -breeder_blanket_inner_surface
firstwall_cell = openmc.Cell(name='firstwall', region=firstwall_region)
firstwall_cell.fill = firstwall_material
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius+firstwall_thickness+blanket_thickness, boundary_type='vacuum')
breeder_blanket_region = -breeder_blanket_outer_surface & +breeder_blanket_inner_surface
breeder_blanket_cell = openmc.Cell(name = 'breeder_blanket', region=breeder_blanket_region)
breeder_blanket_cell.fill = blanket_material
universe = openmc.Universe(cells=[inner_void_cell,
firstwall_cell,
breeder_blanket_cell])
geom = openmc.Geometry(universe)
# assigns simulation settings
sett = openmc.Settings()
sett.batches = 200 # this is minimum number of batches that will be run
sett.trigger_active = True
sett.trigger_max_batches = 1500 # this is maximum number of batches that will be run
sett.particles = 300
sett.verbosity = 1
sett.run_mode = 'fixed source'
# sets a 14MeV (distributuion) point source
source = openmc.Source()
source.space = openmc.stats.Point((0,0,0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Muir(e0=14080000.0, m_rat=5.0, kt=20000.0) #neutron energy = 14.08MeV, AMU for D + T = 5, temperature is 20KeV
sett.source = source
# this is the tally set up
tallies = openmc.Tallies()
# define filters
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
particle_filter = openmc.ParticleFilter(['neutron'])
# creates the TBR tally using the filters and sets a completion trigger
tally = openmc.Tally(name='TBR')
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ['(n,Xt)'] # where X is a wildcard, if MT 105 or (n,t) then some tritium production will be missed, for example (n,nt) which happens in Li7 would be missed
tally.triggers = [openmc.Trigger(trigger_type='rel_err', threshold=0.0001)] # This stops the simulation if the threshold is meet
tallies.append(tally)
# collects all the model parts and runs the model
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run(output=False)
# opens the output file and retrieves the tally results
sp = openmc.StatePoint(sp_filename)
tally = sp.get_tally(name='TBR')
df = tally.get_pandas_dataframe()
tally_result = df['mean'].sum()
tally_std_dev = df['std. dev.'].sum()
# combines the tally results with the input data
inputs.update({'tbr': tally_result})
inputs.update({'tbr_error': tally_std_dev})
return inputs
def simulate_model(
enrichment,
thickness,
breeder_material_name="Li4SiO4",
temperature_in_C=500,
batches=10,
nps=1000,
inner_radius=500,
):
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment=enrichment,
temperature_in_C=temperature_in_C,
).neutronics_material
eurofer = Material(material_name="eurofer").neutronics_material
copper = Material(material_name="copper").neutronics_material
mats = openmc.Materials([breeder_material, eurofer, copper])
mats.export_to_xml("materials.xml")
# GEOMETRY#
central_sol_surface = openmc.ZCylinder(r=100)
central_shield_outer_surface = openmc.ZCylinder(r=110)
first_wall_inner_surface = openmc.Sphere(r=inner_radius)
first_wall_outer_surface = openmc.Sphere(r=inner_radius + 10)
breeder_blanket_outer_surface = openmc.Sphere(r=inner_radius + 10.0 +
thickness)
vessel_outer_surface = openmc.Sphere(r=inner_radius + 10.0 + thickness +
10.0,
boundary_type="vacuum")
central_sol_region = -central_sol_surface & -breeder_blanket_outer_surface
central_sol_cell = openmc.Cell(region=central_sol_region)
central_sol_cell.fill = copper
central_shield_region = (+central_sol_surface
& -central_shield_outer_surface
& -breeder_blanket_outer_surface)
central_shield_cell = openmc.Cell(region=central_shield_region)
central_shield_cell.fill = eurofer
inner_void_region = -first_wall_inner_surface & +central_shield_outer_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = "inner_void"
first_wall_region = (-first_wall_outer_surface
& +first_wall_inner_surface
& +central_shield_outer_surface)
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = eurofer
breeder_blanket_region = (+first_wall_outer_surface
& -breeder_blanket_outer_surface
& +central_shield_outer_surface)
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
vessel_region = +breeder_blanket_outer_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = "vessel"
vessel_cell.fill = eurofer
universe = openmc.Universe(cells=[
central_sol_cell,
central_shield_cell,
inner_void_cell,
first_wall_cell,
breeder_blanket_cell,
vessel_cell,
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS#
sett = openmc.Settings()
sett.batches = batches
sett.inactive = 0
sett.particles = nps
sett.run_mode = "fixed source"
source = openmc.Source()
source.space = openmc.stats.Point((150, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14.08e6], [1])
sett.source = source
# sett.export_to_xml("settings.xml")
# tally filters
particle_filter = openmc.ParticleFilter("neutron")
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
# TALLIES#
tallies = openmc.Tallies()
tally = openmc.Tally(name="TBR")
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ["(n,Xt)"]
tallies.append(tally)
# RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
model.run(output=False)
# RETRIEVING TALLY RESULTS
sp = openmc.StatePoint("statepoint." + str(batches) + ".h5")
json_output = {
"batches": batches,
"nps": nps,
"enrichment": enrichment,
"inner_radius": inner_radius,
"thickness": thickness,
"breeder_material_name": breeder_material_name,
"temperature_in_C": temperature_in_C,
}
tally = sp.get_tally(name="TBR")
df = tally.get_pandas_dataframe()
json_output["TBR"] = df["mean"].sum()
json_output["TBR_std_dev"] = df["std. dev."].sum()
return json_output
import numpy as np
import plotly.graph_objs as go
from plotly.subplots import make_subplots
from neutronics_material_maker import Material
# Because the neutronics_material_maker makes it easy to adjust material properties, we can easily perform parameter studies
# Water density as a function of temperature (at constant pressure)
# Based on WCLL, pressure = 15.5 MPa, inlet and outlet temperatures = 285 and 325 degrees C
# We will show density as a function of temperature over a larger range, but at correct pressure
temperatures = np.linspace(0.1, 600., 200)
water_densities = [
Material('H2O', temperature_in_C=temperature,
pressure_in_Pa=15500000).openmc_material.density
for temperature in temperatures
]
fig = make_subplots(
rows=2,
cols=2,
subplot_titles=(
"Water density as a function of temperature (at constant pressure)",
"Helium density as a function of temperature (at constant pressure)",
"Helium density as a function of pressure (at constant temperature)",
""))
fig.add_trace(go.Scatter(x=temperatures,
y=water_densities,
mode='lines+markers',
from neutronics_material_maker import Material
# Some materials require arguments to correctly calculate material properties
# Water requires 'temperature' and 'pressure' arguments to be passed
# the following command creates a Material() object using the neutronics_material_maker
water = Material('H2O', temperature_in_C=25,
pressure_in_Pa=100000) # atmospheric
print(type(water))
print(water)
# the following command converts Material() objects into neutronics materials which can be used in OpenMC
water_openmc_material_object = water.openmc_material
# this is equivalent to:
# water_openmc_material_object = Material('H2O', temperature_in_C=25, pressure_in_Pa=100000).openmc_material
print(type(water_openmc_material_object))
print(water_openmc_material_object)
# Some materials can also take arguments which adjust material properties
# Lithium Orthosilicate (Li4SiO4) can take arguments of 'enrichment', 'enrichment_target', 'enrichment_type' and 'packing_fraction'
# Note: for some lithium crystals, 'enrichment_target' and 'enrichment_type' are defined by default, but can be changed
default_Li4SiO4 = Material('Li4SiO4').openmc_material
print(default_Li4SiO4)
def simulate_model(enrichment,
blanket_thickness=200,
firstwall_thickness=5,
breeder_material_name="Li4SiO4",
temperature_in_C=500,
threshold=0.0005,
inner_radius=500):
# MATERIALS from library of materials in neutronics_material_maker package
breeder_material = Material(
material_name=breeder_material_name,
enrichment=enrichment,
temperature_in_C=temperature_in_C,
).neutronics_material
SS316 = Material(material_name="SS316").neutronics_material
copper = Material(material_name="copper").neutronics_material
mats = openmc.Materials([breeder_material, SS316, copper])
mats.export_to_xml("materials.xml")
# GEOMETRY#
first_wall_inner_surface = openmc.Sphere(r=inner_radius)
first_wall_outer_surface = openmc.Sphere(r=inner_radius +
firstwall_thickness)
breeder_blanket_outer_surface = openmc.Sphere(
r=inner_radius + firstwall_thickness + blanket_thickness)
vessel_outer_surface = openmc.Sphere(r=inner_radius + firstwall_thickness +
blanket_thickness + 10.,
boundary_type="vacuum")
inner_void_region = -first_wall_inner_surface
inner_void_cell = openmc.Cell(region=inner_void_region)
inner_void_cell.name = "inner_void"
first_wall_region = (-first_wall_outer_surface & +first_wall_inner_surface)
first_wall_cell = openmc.Cell(region=first_wall_region)
first_wall_cell.fill = SS316
breeder_blanket_region = (+first_wall_outer_surface
& -breeder_blanket_outer_surface)
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
vessel_region = +breeder_blanket_outer_surface & -vessel_outer_surface
vessel_cell = openmc.Cell(region=vessel_region)
vessel_cell.name = "vessel"
vessel_cell.fill = SS316
universe = openmc.Universe(cells=[
inner_void_cell, first_wall_cell, breeder_blanket_cell, vessel_cell
])
geom = openmc.Geometry(universe)
# SIMULATION SETTINGS#
sett = openmc.Settings()
sett.batches = 2 # this is minimum number of batches that will be run
sett.trigger_active = True
sett.trigger_max_batches = 2000 # this is maximum number of batches that will be run
sett.inactive = 0
sett.particles = 1000
sett.run_mode = "fixed source"
source = openmc.Source()
source.space = openmc.stats.Point((150, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14.08e6], [1])
sett.source = source
# sett.export_to_xml("settings.xml")
# tally filters
particle_filter = openmc.ParticleFilter("neutron")
cell_filter_breeder = openmc.CellFilter(breeder_blanket_cell)
# TALLIES#
tallies = openmc.Tallies()
tally = openmc.Tally(name="TBR")
tally.filters = [cell_filter_breeder, particle_filter]
tally.scores = ["(n,Xt)"]
tally.triggers = [
openmc.Trigger(trigger_type='rel_err', threshold=threshold)
] # This stops the simulation if the threshold is meet
tallies.append(tally)
# RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
sp_filename = model.run(output=False)
# RETRIEVING TALLY RESULTS
sp = openmc.StatePoint(sp_filename)
json_output = {
"enrichment": enrichment,
"inner_radius": inner_radius,
"firstwall_thickness": firstwall_thickness,
"blanket_thickness": blanket_thickness,
"breeder_material_name": breeder_material_name,
"temperature_in_C": temperature_in_C,
}
tally = sp.get_tally(name="TBR")
df = tally.get_pandas_dataframe()
json_output["TBR"] = df["mean"].sum()
json_output["TBR_std_dev"] = df["std. dev."].sum()
os.system('rm *.h5')
return json_output
def make_neutronics_model(
reactor,
firstwall_radial_thickness,
firstwall_armour_material,
firstwall_coolant_material,
firstwall_structural_material,
firstwall_armour_fraction,
firstwall_coolant_fraction,
firstwall_coolant_temperature_C,
firstwall_coolant_pressure_Pa,
firstwall_structural_fraction,
blanket_rear_wall_coolant_material,
blanket_rear_wall_structural_material,
blanket_rear_wall_coolant_fraction,
blanket_rear_wall_structural_fraction,
blanket_rear_wall_coolant_temperature_C,
blanket_rear_wall_coolant_pressure_Pa,
blanket_lithium6_enrichment_percent,
blanket_breeder_material,
blanket_coolant_material,
blanket_multiplier_material,
blanket_structural_material,
blanket_breeder_fraction,
blanket_coolant_fraction,
blanket_multiplier_fraction,
blanket_structural_fraction,
blanket_breeder_packing_fraction,
blanket_multiplier_packing_fraction,
blanket_coolant_temperature_C,
blanket_coolant_pressure_Pa,
blanket_breeder_temperature_C,
blanket_breeder_pressure_Pa,
divertor_coolant_fraction,
divertor_structural_fraction,
divertor_coolant_material,
divertor_structural_material,
divertor_coolant_temperature_C,
divertor_coolant_pressure_Pa,
center_column_shield_coolant_fraction,
center_column_shield_structural_fraction,
center_column_shield_coolant_material,
center_column_shield_structural_material,
center_column_shield_coolant_temperature_C,
center_column_shield_coolant_pressure_Pa,
inboard_tf_coils_conductor_fraction,
inboard_tf_coils_coolant_fraction,
inboard_tf_coils_structure_fraction,
inboard_tf_coils_conductor_material,
inboard_tf_coils_coolant_material,
inboard_tf_coils_structure_material,
inboard_tf_coils_coolant_temperature_C,
inboard_tf_coils_coolant_pressure_Pa,
):
"""
Makes and runs a simple OpenMC neutronics model with
the materials with the same tags as the DAGMC neutronics
geometry. The model also specifies the computational
intensity (particles and batches) and the tally to record
"""
input_parameters = locals()
# this is the underlying geometry container that is filled with the faceteted CAD model
universe = openmc.Universe()
geom = openmc.Geometry(universe)
center_column_shield_material = MultiMaterial(
material_tag='center_column_shield_mat',
materials=[
Material(
material_name=center_column_shield_coolant_material,
temperature_in_C=center_column_shield_coolant_temperature_C,
pressure_in_Pa=center_column_shield_coolant_pressure_Pa),
Material(material_name=center_column_shield_structural_material)
],
fracs=[
center_column_shield_coolant_fraction,
center_column_shield_structural_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
firstwall_material = MultiMaterial(
material_tag='firstwall_mat',
materials=[
Material(material_name=firstwall_coolant_material,
temperature_in_C=firstwall_coolant_temperature_C,
pressure_in_Pa=firstwall_coolant_pressure_Pa),
Material(material_name=firstwall_structural_material),
Material(material_name=firstwall_armour_material)
],
fracs=[
firstwall_coolant_fraction, firstwall_structural_fraction,
firstwall_armour_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
if blanket_multiplier_material == None and blanket_multiplier_fraction == None and blanket_multiplier_packing_fraction == None:
blanket_material = MultiMaterial(
material_tag='blanket_mat',
materials=[
Material(material_name=blanket_coolant_material,
temperature_in_C=blanket_coolant_temperature_C,
pressure_in_Pa=blanket_coolant_pressure_Pa),
Material(material_name=blanket_structural_material),
Material(material_name=blanket_breeder_material,
enrichment=blanket_lithium6_enrichment_percent,
packing_fraction=blanket_breeder_packing_fraction,
temperature_in_C=blanket_breeder_temperature_C,
pressure_in_Pa=blanket_breeder_pressure_Pa)
],
fracs=[
blanket_coolant_fraction, blanket_structural_fraction,
blanket_breeder_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
else:
blanket_material = MultiMaterial(
material_tag='blanket_mat',
materials=[
Material(material_name=blanket_coolant_material,
temperature_in_C=blanket_coolant_temperature_C,
pressure_in_Pa=blanket_coolant_pressure_Pa),
Material(material_name=blanket_structural_material),
Material(material_name=blanket_multiplier_material,
packing_fraction=blanket_multiplier_packing_fraction),
Material(material_name=blanket_breeder_material,
enrichment=blanket_lithium6_enrichment_percent,
packing_fraction=blanket_breeder_packing_fraction,
temperature_in_C=blanket_breeder_temperature_C,
pressure_in_Pa=blanket_breeder_pressure_Pa)
],
fracs=[
blanket_coolant_fraction, blanket_structural_fraction,
blanket_multiplier_fraction, blanket_breeder_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
divertor_material = MultiMaterial(
material_tag='divertor_mat',
materials=[
Material(material_name=divertor_coolant_material,
temperature_in_C=divertor_coolant_temperature_C,
pressure_in_Pa=divertor_coolant_pressure_Pa),
Material(material_name=divertor_structural_material)
],
fracs=[divertor_coolant_fraction, divertor_structural_fraction],
percent_type='vo',
packing_fraction=1.0).openmc_material
inboard_tf_coils_material = MultiMaterial(
material_tag='inboard_tf_coils_mat',
materials=[
Material(material_name=inboard_tf_coils_coolant_material,
temperature_in_C=inboard_tf_coils_coolant_temperature_C,
pressure_in_Pa=inboard_tf_coils_coolant_pressure_Pa),
Material(material_name=inboard_tf_coils_conductor_material),
Material(material_name=inboard_tf_coils_structure_material)
],
fracs=[
inboard_tf_coils_coolant_fraction,
inboard_tf_coils_conductor_fraction,
inboard_tf_coils_structure_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
blanket_rear_wall_material = MultiMaterial(
material_tag='blanket_rear_wall_mat',
materials=[
Material(material_name=blanket_rear_wall_coolant_material,
temperature_in_C=blanket_rear_wall_coolant_temperature_C,
pressure_in_Pa=blanket_rear_wall_coolant_pressure_Pa),
Material(material_name=blanket_rear_wall_structural_material)
],
fracs=[
blanket_rear_wall_coolant_fraction,
blanket_rear_wall_structural_fraction
],
percent_type='vo',
packing_fraction=1.0).openmc_material
mats = openmc.Materials([
center_column_shield_material, firstwall_material, blanket_material,
divertor_material, inboard_tf_coils_material,
blanket_rear_wall_material
])
# settings for the number of neutrons to simulate
settings = openmc.Settings()
settings.batches = 10
settings.inactive = 0
settings.particles = 1000
settings.run_mode = 'fixed source'
settings.dagmc = True
# details of the birth locations and energy of the neutronis
source = openmc.Source()
source.space = openmc.stats.Point((reactor['major_radius'], 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
settings.source = source
settings.photon_transport = True # This line is required to switch on photons tracking
# details about what neutrons interactions to keep track of (called a tally)
tallies = openmc.Tallies()
material_filter = openmc.MaterialFilter(blanket_material)
tbr_tally = openmc.Tally(name='TBR')
tbr_tally.filters = [material_filter]
tbr_tally.scores = ['(n,Xt)'] # where X is a wild card
tallies.append(tbr_tally)
material_filter = openmc.MaterialFilter(
[blanket_material, firstwall_material, blanket_rear_wall_material])
blanket_heating_tally = openmc.Tally(name='blanket_heating')
blanket_heating_tally.filters = [material_filter]
blanket_heating_tally.scores = ['heating']
tallies.append(blanket_heating_tally)
# make the model from gemonetry, materials, settings and tallies
model = openmc.model.Model(geom, mats, settings, tallies)
# run the simulation
output_filename = model.run()
"""
Reads the output file from the neutronics simulation
and prints the TBR tally result to screen
"""
# open the results file
sp = openmc.StatePoint(output_filename)
# access TBR tally
tbr_tally = sp.get_tally(name='TBR')
df = tbr_tally.get_pandas_dataframe()
tbr_tally_result = df['mean'].sum()
tbr_tally_std_dev = df['std. dev.'].sum()
# access heating tally
blanket_heating_tally = sp.get_tally(name='blanket_heating')
df = blanket_heating_tally.get_pandas_dataframe()
blanket_heating_tally_result = df['mean'].sum() / 1e6
blanket_heating_tally_std_dev = df['std. dev.'].sum() / 1e6
# returns all the inputs and some extra reactor attributes, merged into a single dictionary
return {
**input_parameters,
**{
'tbr': tbr_tally_result,
'tbr_std_dev': tbr_tally_std_dev,
'blanket_heating': blanket_heating_tally_result,
'blanket_heating_std_dev': blanket_heating_tally_std_dev,
}
}
firstwall_materials.append(fw_material)
if st.button('Simulate model'):
for material in breeder_materials:
for float_key in [
'enrichment_fraction', 'packing_fraction', 'volume_fraction',
'temperature_in_C'
]:
if float_key in material.keys():
material[float_key] = float(material[float_key])
if len(materials) == 1:
for material in breeder_materials:
breeder_material = Material(**material).neutronics_material
else:
multimaterials = []
volume_fractions = []
for material in breeder_materials:
multimaterials.append(Material(**material))
volume_fractions.append(material['volume_fraction'])
breeder_material = MultiMaterial(
material_name='breeder_material',
materials=multimaterials,
volume_fractions=volume_fractions).neutronics_material
if model == 'sphere with firstwall':
for material in firstwall_materials:
for float_key in [