inactive = 2
particles = 1000
#Fuel material settings:
"""
fuel_material = openmc.Material(11, name='Thorium fuel')
fuel_material.set_density('g/cm3', 10.062)
fuel_material.add_nuclide("Th232", 94)
fuel_material.add_nuclide("U235", 5)
fuel_material.add_nuclide("U238", 0.33)
fuel_material.add_nuclide("Xe135", 0.001)
fuel_material.add_nuclide("O16", 0.33)
fuel_material.depletable = True
"""
fuel_material = openmc.Material(11, name='Thorium fuel')
fuel_material.set_density('g/cm3', 10.062)
fuel_material.add_element("Th", 0.9)
fuel_material.add_nuclide('Pu239', 0.092) #weaponsgrade ref:
fuel_material.add_nuclide('Pu240', 0.008) #https://www.world-nuclear.org/information-library/nuclear-fuel-cycle/fuel-recycling/plutonium.aspx
fuel_material.add_element("O", 2.0)
fuel_material.depletable = True
"""
fuel_material = openmc.Material(11, name='Thorium fuel')
fuel_material.set_density('g/cm3', 10.062)
fuel_material.add_element("Th", 0.96)
fuel_material.add_element("O", 2.0)
fuel_material.add_nuclide('U233', 0.04)
fuel_material.depletable = True
"""
def build_default_materials_and_geometry(self):
# Define materials needed for 1D/1G slab problem
uo2_data = openmc.Macroscopic('uo2_iso', '71c')
uo2 = openmc.Material(name='UO2', material_id=1)
uo2.set_density('macro', 1.0)
uo2.add_macroscopic(uo2_data)
clad_data = openmc.Macroscopic('clad_iso', '71c')
clad = openmc.Material(name='Clad', material_id=2)
clad.set_density('macro', 1.0)
clad.add_macroscopic(clad_data)
water_data = openmc.Macroscopic('lwtr_iso', '71c')
water = openmc.Material(name='LWTR', material_id=3)
water.set_density('macro', 1.0)
water.add_macroscopic(water_data)
# Define the materials file.
self.materials.default_xs = '71c'
self.materials += (uo2, clad, water)
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0, surface_id=200,
boundary_type='reflective')
right = openmc.XPlane(x0=10.0, surface_id=201,
boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0, surface_id=300,
boundary_type='reflective')
top = openmc.YPlane(y0=10.0, surface_id=301,
boundary_type='reflective')
down = openmc.ZPlane(z0=0.0, surface_id=0,
boundary_type='reflective')
fuel_clad_intfc = openmc.ZPlane(z0=2.0, surface_id=1)
clad_lwtr_intfc = openmc.ZPlane(z0=2.4, surface_id=2)
up = openmc.ZPlane(z0=5.0, surface_id=3,
boundary_type='reflective')
# Define cells
c1 = openmc.Cell(cell_id=1)
c1.region = +left & -right & +bottom & -top & +down & -fuel_clad_intfc
c1.fill = uo2
c2 = openmc.Cell(cell_id=2)
c2.region = +left & -right & +bottom & -top & +fuel_clad_intfc & -clad_lwtr_intfc
c2.fill = clad
c3 = openmc.Cell(cell_id=3)
c3.region = +left & -right & +bottom & -top & +clad_lwtr_intfc & -up
c3.fill = water
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
root.add_cells((c1,c2,c3))
# Define the geometry file.
geometry = openmc.Geometry()
geometry.root_universe = root
self.geometry = geometry
import openmc
mats = openmc.Materials()
mat = openmc.Material(1)
mat.name = "Plutonium"
mat.set_density('sum')
mat.add_nuclide('Pu239', 4.4422e-02)
mat.add_nuclide('Pu240', 2.1326e-03)
mat.add_nuclide('Pu241', 9.2538e-05)
mat.add_element('C', 1.9515e-04)
mat.add_element('Fe', 8.1943e-05)
mats.append(mat)
mat = openmc.Material(2)
mat.name = "Berylium Reflector"
mat.set_density('sum')
mat.add_element('Be', 1.2099e-01)
mat.add_element('O', 1.0449e-03)
mat.add_s_alpha_beta('c_Be')
mats.append(mat)
mat = openmc.Material(3)
mat.name = "Steel Cover"
mat.set_density('sum')
mat.add_element('Fe', 5.1280e-02)
mat.add_element('C', 3.4757e-04)
mat.add_element('Si', 8.9185e-03)
mat.add_element('Ti', 6.1034e-04)
mat.add_element('Cr', 1.4452e-02)
mat.add_element('Mn', 1.5198e-03)
import openmc
import numpy as np
# set wall material
wall = openmc.Material(1, "wall")
wall.add_nuclide('Pb205', 1) # which pb to use
wall.set_density('g/cm3', 1) # ???
# set observation point
detector = openmc.Material(2, "detector")
detector.add_nuclide('Xe135', 1)
detector.set_density('g/cm3', 1) # ?
# set source material
source = openmc.Material(3, "source")
source.add_nuclide('Pt196', 1) # which pt to use?
source.set_density('atom/cm3', 10**4)
class Start_point:
def __init__(self, x_, y_, z_):
self.x = x_
self.y = y_
self.z = z_
def truncation_create(plane_list):
#list[][0] pos plane list[][1] neg plane
shape = (-plane_list[0][0]) & (+plane_list[0][1])
for oppo_planes in plane_list:
shape = shape & ((-oppo_planes[0]) & (+oppo_planes[1]))
#!/usr/bin/env python3
"""example_material_plot.py: plots cross sections for materials."""
import openmc
import plotly.graph_objects as go
natural_Li4SiO4 = openmc.Material()
natural_Li4SiO4.add_element('Li', 4.0, percent_type='ao')
natural_Li4SiO4.add_element('Si', 1.0, percent_type='ao')
natural_Li4SiO4.add_element('O', 4.0, percent_type='ao')
natural_Li4SiO4.set_density('g/cm3', 1.877)
#Li4SiO4 density 1.8770150075137564 g/cm3
enrichment_fraction = 0.6
enriched_Li4SiO4 = openmc.Material()
enriched_Li4SiO4.add_nuclide('Li6',
4.0 * enrichment_fraction,
percent_type='ao')
enriched_Li4SiO4.add_nuclide('Li7',
4.0 * (1 - enrichment_fraction),
percent_type='ao')
enriched_Li4SiO4.add_element('Si', 1.0, percent_type='ao')
enriched_Li4SiO4.add_element('O', 4.0, percent_type='ao')
enriched_Li4SiO4.set_density(
'g/cm3', 1.844) # this density is lower as there is more Li6 and less Li7
#Li4SiO4 density 1.8441466011318948 g/cm3 with 60% Li6
# Try adding another candidate breeder material (e.g. Li2SiO3, Li2ZrO3 or Li2TiO3) to the plot
#Li2SiO3 density 2.619497078021483 g/cm3
#Li2ZrO3 density 2.5288596326567134 g/cm3
#Li2TiO3 density 2.8994147653592983 g/cm3
import openmc
from plotly import __version__
from plotly.offline import download_plotlyjs, plot
from plotly.graph_objs import Scatter, Layout
def calculate_crystal_structure_density(material, atoms_per_unit_cell,
volume_of_unit_cell_cm3):
molar_mass = material.average_molar_mass * len(material.nuclides)
atomic_mass_unit_in_g = 1.660539040e-24
density_g_per_cm3 = molar_mass * atomic_mass_unit_in_g * atoms_per_unit_cell / volume_of_unit_cell_cm3
#print('density =',density_g_per_cm3)
return density_g_per_cm3
natural_Li4SiO4 = openmc.Material()
natural_Li4SiO4.add_element('Li', 4.0, percent_type='ao')
natural_Li4SiO4.add_element('Si', 1.0, percent_type='ao')
natural_Li4SiO4.add_element('O', 4.0, percent_type='ao')
natural_Li4SiO4.set_density(
'g/cm3',
calculate_crystal_structure_density(natural_Li4SiO4, 14, 1.1543e-21))
print('natural_Li4SiO4 density', natural_Li4SiO4.density,
natural_Li4SiO4.density_units)
enrichment_fraction = 0.6
enriched_Li4SiO4 = openmc.Material()
enriched_Li4SiO4.add_nuclide('Li6',
4.0 * enrichment_fraction,
percent_type='ao')
enriched_Li4SiO4.add_nuclide('Li7',
def _build_inputs(self):
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cc', 10.0)
uo2.add_nuclide('U238', 1.0)
uo2.add_nuclide('U235', 0.02)
uo2.add_nuclide('O16', 2.0)
borated_water = openmc.Material(name='Borated water')
borated_water.set_density('g/cm3', 1)
borated_water.add_nuclide('B10', 10e-5)
borated_water.add_nuclide('H1', 2.0)
borated_water.add_nuclide('O16', 1.0)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, borated_water])
materials_file.export_to_xml()
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(surface_id=1,
x0=0,
y0=0,
r=1,
name='Fuel OR')
left = openmc.XPlane(surface_id=2, x0=-2, name='left')
right = openmc.XPlane(surface_id=3, x0=2, name='right')
bottom = openmc.YPlane(y0=-2, name='bottom')
top = openmc.YPlane(y0=2, name='top')
left.boundary_type = 'vacuum'
right.boundary_type = 'reflective'
top.boundary_type = 'reflective'
bottom.boundary_type = 'reflective'
# Instantiate Cells
fuel = openmc.Cell(name='fuel')
water = openmc.Cell(name='water')
# Use surface half-spaces to define regions
fuel.region = -fuel_or
water.region = +fuel_or & -right & +bottom & -top
# Register Materials with Cells
fuel.fill = uo2
water.fill = borated_water
# Instantiate pin cell Universe
pin_cell = openmc.Universe(name='pin cell')
pin_cell.add_cells([fuel, water])
# Instantiate root Cell and Universe
root_cell = openmc.Cell(name='root cell')
root_cell.region = +left & -right & +bottom & -top
root_cell.fill = pin_cell
root_univ = openmc.Universe(universe_id=0, name='root universe')
root_univ.add_cell(root_cell)
# Instantiate a Geometry, register the root Universe
geometry = openmc.Geometry(root_univ)
geometry.export_to_xml()
# Instantiate a Settings object, set all runtime parameters
settings_file = openmc.Settings()
settings_file.batches = 10
settings_file.inactive = 0
settings_file.particles = 1000
#settings_file.output = {'tallies': True}
# Create an initial uniform spatial source distribution
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:],\
only_fissionable=True)
settings_file.source = openmc.source.Source(space=uniform_dist)
settings_file.export_to_xml()
# Tallies file
tallies_file = openmc.Tallies()
# Create partial current tallies from fuel to water
# Filters
two_groups = [0., 4e6, 20e6]
energy_filter = openmc.EnergyFilter(two_groups)
polar_filter = openmc.PolarFilter([0, np.pi / 4, np.pi])
azimuthal_filter = openmc.AzimuthalFilter([0, np.pi / 4, np.pi])
surface_filter = openmc.SurfaceFilter([1])
cell_from_filter = openmc.CellFromFilter(fuel)
cell_filter = openmc.CellFilter(water)
# Use Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Use a Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('fuel_to_water_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create partial current tallies from water to fuel
# Filters
cell_from_filter = openmc.CellFromFilter(water)
cell_filter = openmc.CellFilter(fuel)
# Cell to cell filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_1'))
cell_to_cell_tally.filters = [cell_from_filter, cell_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Cell from + surface filters for partial current
cell_to_cell_tally = openmc.Tally(name=str('water_to_fuel_2'))
cell_to_cell_tally.filters = [cell_from_filter, surface_filter, \
energy_filter, polar_filter, azimuthal_filter]
cell_to_cell_tally.scores = ['current']
tallies_file.append(cell_to_cell_tally)
# Create a net current tally on inner surface using a surface filter
surface_filter = openmc.SurfaceFilter([1])
surf_tally1 = openmc.Tally(name='net_cylinder')
surf_tally1.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally1.scores = ['current']
tallies_file.append(surf_tally1)
# Create a net current tally on left surface using a surface filter
# This surface has a vacuum boundary condition, so leakage is tallied
surface_filter = openmc.SurfaceFilter([2])
surf_tally2 = openmc.Tally(name='leakage_left')
surf_tally2.filters = [surface_filter, energy_filter, polar_filter, \
azimuthal_filter]
surf_tally2.scores = ['current']
tallies_file.append(surf_tally2)
# Create a net current tally on right surface using a surface filter
# This surface has a reflective boundary condition, so the net current
# should be zero.
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
surface_filter = openmc.SurfaceFilter([3])
surf_tally3 = openmc.Tally(name='net_right')
surf_tally3.filters = [surface_filter, energy_filter]
surf_tally3.scores = ['current']
tallies_file.append(surf_tally3)
tallies_file.export_to_xml()
#!/usr/bin/env python3
"""example_isotope_plot.py: plots few 2D views of a simple geometry ."""
__author__ = "Jonathan Shimwell"
import openmc
import matplotlib.pyplot as plt
import os
mats = openmc.Materials()
natural_lead = openmc.Material(1, "natural_lead")
natural_lead.add_element('Pb', 1, 'ao')
mats.append(natural_lead)
mats.export_to_xml()
#example surfaces
surface_sph1 = openmc.Sphere(r=500)
surface_sph2 = openmc.Sphere(r=600)
volume_sph1 = +surface_sph1 & -surface_sph2 # above (+) surface_sph and below (-) surface_sph2
#add surfaces here using https://openmc.readthedocs.io/en/stable/usersguide/geometry.html#surfaces-and-regions
#example cell
cell1 = openmc.Cell(region=volume_sph1)
cell1.fill = natural_lead
#add another cell here
universe = openmc.Universe(
def make_materials_geometry_tallies(enrichment_fraction):
#MATERIALS#
breeder_material = openmc.Material(
1, "breeder_material") #Pb84.2Li15.8 with natural enrichment of Li6
breeder_material.add_element('Pb', 84.2, 'ao')
breeder_material.add_nuclide('Li6', enrichment_fraction * 15.8, 'ao')
breeder_material.add_nuclide('Li7', (1.0 - enrichment_fraction) * 15.8,
'ao')
breeder_material.set_density('atom/b-cm', 3.2720171e-2) # around 11 g/cm3
copper = openmc.Material(name='copper')
copper.set_density('g/cm3', 8.5)
copper.add_element('Cu', 1.0)
eurofer = openmc.Material(name='eurofer')
eurofer.set_density('g/cm3', 7.75)
eurofer.add_element('Fe', 89.067, percent_type='wo')
eurofer.add_element('C', 0.11, percent_type='wo')
eurofer.add_element('Mn', 0.4, percent_type='wo')
eurofer.add_element('Cr', 9.0, percent_type='wo')
eurofer.add_element('Ta', 0.12, percent_type='wo')
eurofer.add_element('W', 1.1, percent_type='wo')
eurofer.add_element('N', 0.003, percent_type='wo')
eurofer.add_element('V', 0.2, percent_type='wo')
mats = openmc.Materials([breeder_material, eurofer, copper])
#GEOMETRY#
central_sol_surface = openmc.ZCylinder(r=100)
central_shield_outer_surface = openmc.ZCylinder(r=110)
vessel_inner = openmc.Sphere(r=500)
first_wall_outer_surface = openmc.Sphere(r=510)
breeder_blanket_outer_surface = openmc.Sphere(r=610,
boundary_type='vacuum')
central_sol_region = -central_sol_surface & -vessel_inner
central_sol_cell = openmc.Cell(region=central_sol_region)
central_sol_cell.fill = copper
central_shield_region = +central_sol_surface & -central_shield_outer_surface & -vessel_inner
central_shield_cell = openmc.Cell(region=central_shield_region)
central_shield_cell.fill = eurofer
inner_vessel_region = -vessel_inner & +central_shield_outer_surface
inner_vessel_cell = openmc.Cell(region=inner_vessel_region)
first_wall_region = -first_wall_outer_surface & +vessel_inner
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
breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
breeder_blanket_cell.fill = breeder_material
breeder_blanket_cell.name = 'breeder_blanket'
universe = openmc.Universe(cells=[
central_sol_cell, central_shield_cell, inner_vessel_cell,
first_wall_cell, breeder_blanket_cell
])
geom = openmc.Geometry(universe)
#SIMULATION SETTINGS#
sett = openmc.Settings()
batches = 3
sett.batches = batches
sett.inactive = 0
sett.particles = 5000
sett.run_mode = 'fixed source'
source = openmc.Source()
source.space = openmc.stats.Point((350, 0, 0))
source.angle = openmc.stats.Isotropic()
source.energy = openmc.stats.Discrete([14e6], [1])
sett.source = source
#TALLIES#
tallies = openmc.Tallies()
cell_filter = openmc.CellFilter(breeder_blanket_cell)
tbr_tally = openmc.Tally(2, name='TBR')
tbr_tally.filters = [cell_filter]
tbr_tally.scores = [
'205'
] # 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(tbr_tally)
#RUN OPENMC #
model = openmc.model.Model(geom, mats, sett, tallies)
model.run()
sp = openmc.StatePoint('statepoint.' + str(batches) + '.h5')
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()
return {
'enrichment_fraction': enrichment_fraction,
'tbr_tally_result': tbr_tally_result,
'tbr_tally_std_dev': tbr_tally_std_dev
}
#!/usr/bin/env python3
import openmc
import openmc.model
import os
import matplotlib.pyplot as plt
#MATERIALS#
mats = openmc.Materials()
copper = openmc.Material(name='Copper')
copper.set_density('g/cm3', 8.5)
copper.add_element('Cu', 1.0)
mats.append(copper)
eurofer = openmc.Material(name='EUROFER97')
eurofer.set_density('g/cm3', 7.75)
eurofer.add_element('Fe', 89.067, percent_type='wo')
eurofer.add_element('C', 0.11, percent_type='wo')
eurofer.add_element('Mn', 0.4, percent_type='wo')
eurofer.add_element('Cr', 9.0, percent_type='wo')
eurofer.add_element('Ta', 0.12, percent_type='wo')
eurofer.add_element('W', 1.1, percent_type='wo')
eurofer.add_element('N', 0.003, percent_type='wo')
eurofer.add_element('V', 0.2, percent_type='wo')
mats.append(eurofer)
#GEOMETRY#
import openmc
steel = openmc.Material(1, 'steel')
steel.add_element('Si', 1.3603E-03)
steel.add_element('Ti', 5.9844E-04)
steel.add_element('Ni', 8.1369E-03)
steel.add_element('Cr', 1.6532E-02)
steel.add_element('Fe', 5.9088E-02)
steel.add_nuclide('Mn55', 1.3039E-03)
steel.set_density('atom/b-cm', 8.701954E-02)
steel.temperature = 293
water = openmc.Material(2, 'water')
water.add_nuclide('H1', 6.6742E-02)
water.add_nuclide('O16', 3.3371E-02)
water.set_density('atom/b-cm', 1.001130E-01)
water.add_s_alpha_beta('c_H_in_H2O')
water.temperature = 293
clad = openmc.Material(4, 'cladding')
clad.add_nuclide('B10', 1.0844E-02)
clad.add_nuclide('B11', 4.3649E-02)
clad.add_element('C', 1.3623E-02)
clad.set_density('atom/b-cm', 6.811600E-02)
clad.temperature = 293
fuel = openmc.Material(3, 'fuel')
fuel.add_nuclide('H1', 5.6221E-02)
fuel.add_nuclide('O16', 3.8624E-02)
fuel.add_nuclide('N14', 2.9898E-03)
fuel.add_nuclide('U234', 3.0893E-07)
#!/usr/bin/env python3
"""example_isotope_plot.py: plots few 2D views of a simple tokamak geometry ."""
__author__ = "Jonathan Shimwell"
import openmc
import matplotlib.pyplot as plt
import os
mats = openmc.Materials()
copper = openmc.Material(name='Copper')
copper.set_density('g/cm3', 8.5)
copper.add_element('Cu', 1.0)
mats.append(copper)
eurofer = openmc.Material(name='EUROFER97')
eurofer.set_density('g/cm3', 7.75)
eurofer.add_element('Fe', 89.067, percent_type='wo')
eurofer.add_element('C', 0.11, percent_type='wo')
eurofer.add_element('Mn', 0.4, percent_type='wo')
eurofer.add_element('Cr', 9.0, percent_type='wo')
eurofer.add_element('Ta', 0.12, percent_type='wo')
eurofer.add_element('W', 1.1, percent_type='wo')
eurofer.add_element('N', 0.003, percent_type='wo')
eurofer.add_element('V', 0.2, percent_type='wo')
mats.append(eurofer)
breeder_material = openmc.Material(name='breeder_material')
breeder_material.set_density('g/cm3', 9.1)
breeder_material.add_element('Pb', 84.2, percent_type='ao')
#!/usr/bin/env python2
# -*- coding: utf-8 -*-
"""
Created on Wed Aug 9 09:43:36 2017
@author: stu
"""
import openmc
# define the model materials
# define fuel: 3% (atom percent) enriched UO2
uo2 = openmc.Material()
uo2.add_nuclide('U235', 0.03)
uo2.add_nuclide('U238', 0.97)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.0)
# define zirconium
zirconium = openmc.Material()
zirconium.add_element('Zr', 1.0)
zirconium.set_density('g/cm3', 6.6)
# define water
water = openmc.Material()
water.add_element('H', 2.0)
water.add_element('O', 1.0)
water.set_density('g/cm3', 0.7)
water.add_s_alpha_beta('c_H_in_H2O')
mf = openmc.Materials((uo2, zirconium, water))
def zr():
zr = openmc.Material()
zr.add_element('Zr', 1.0)
zr.set_density('g/cm3', 1.0)
return zr
import openmc
import numpy as np
fuel=openmc.Material(1,'fuel')
fuel.add_nuclide('O16',4.1202e-2)
fuel.add_nuclide('U234',2.8563e-6)
fuel.add_nuclide('U235',4.8785e-4)
fuel.add_nuclide('U236',3.5348e-6)
fuel.add_nuclide('U238',2.0009e-2)
fuel.set_density('g/cm3',9.8)
water=openmc.Material(2,'water')
water.add_nuclide('H1',6.6706e-2)
water.add_nuclide('O16',3.3353e-2)
water.add_nuclide('Gd152',7.9656e-11)
water.add_nuclide('Gd154',8.6825e-10)
water.add_nuclide('Gd155',5.8946e-9)
water.add_nuclide('Gd156',8.1528e-9)
water.add_nuclide('Gd157',6.2331e-9)
water.add_nuclide('Gd158',9.8933e-9)
water.add_nuclide('Gd160',8.7064e-9)
water.set_density('g/cm3',0.997766)
water.add_s_alpha_beta('c_H_in_H2O')
clad6061=openmc.Material(3,'clad6061')
clad6061.add_element('Mg',6.6651e-4)
clad6061.add_nuclide('Al27',5.8433e-2)
clad6061.add_element('Si',3.4607e-4)
clad6061.add_element('Ti',2.5375e-5)
from math import pi
import openmc
import openmc.deplete
import matplotlib.pyplot as plt
###############################################################################
# Define materials
###############################################################################
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
helium = openmc.Material(name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_element('He', 2.4044e-4)
zircaloy = openmc.Material(name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_element('Sn', 0.014, 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001, 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
borated_water = openmc.Material(name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
inactive = 5
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
# Instantiate some Materials and register the appropriate Nuclides
fuel1 = openmc.Material(material_id=1, name='fuel')
fuel1.set_density('g/cc', 4.5)
fuel1.add_nuclide(u235, 1.)
fuel2 = openmc.Material(material_id=2, name='depleted fuel')
fuel2.set_density('g/cc', 4.5)
fuel2.add_nuclide(u238, 1.)
moderator = openmc.Material(material_id=3, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel1, fuel2, moderator])
def test_unstructured_mesh(test_opts):
openmc.reset_auto_ids()
# skip the test if the library is not enabled
if test_opts['library'] == 'moab' and not openmc.lib._dagmc_enabled():
pytest.skip("DAGMC (and MOAB) mesh not enbaled in this build.")
if test_opts['library'] == 'libmesh' and not openmc.lib._libmesh_enabled():
pytest.skip("LibMesh is not enabled in this build.")
# skip the tracklength test for libmesh
if test_opts['library'] == 'libmesh' and \
test_opts['estimator'] == 'tracklength':
pytest.skip("Tracklength tallies are not supported using libmesh.")
### Materials ###
materials = openmc.Materials()
fuel_mat = openmc.Material(name="fuel")
fuel_mat.add_nuclide("U235", 1.0)
fuel_mat.set_density('g/cc', 4.5)
materials.append(fuel_mat)
zirc_mat = openmc.Material(name="zircaloy")
zirc_mat.add_element("Zr", 1.0)
zirc_mat.set_density("g/cc", 5.77)
materials.append(zirc_mat)
water_mat = openmc.Material(name="water")
water_mat.add_nuclide("H1", 2.0)
water_mat.add_nuclide("O16", 1.0)
water_mat.set_density("atom/b-cm", 0.07416)
materials.append(water_mat)
materials.export_to_xml()
### Geometry ###
fuel_min_x = openmc.XPlane(-5.0, name="minimum x")
fuel_max_x = openmc.XPlane(5.0, name="maximum x")
fuel_min_y = openmc.YPlane(-5.0, name="minimum y")
fuel_max_y = openmc.YPlane(5.0, name="maximum y")
fuel_min_z = openmc.ZPlane(-5.0, name="minimum z")
fuel_max_z = openmc.ZPlane(5.0, name="maximum z")
fuel_cell = openmc.Cell(name="fuel")
fuel_cell.region = +fuel_min_x & -fuel_max_x & \
+fuel_min_y & -fuel_max_y & \
+fuel_min_z & -fuel_max_z
fuel_cell.fill = fuel_mat
clad_min_x = openmc.XPlane(-6.0, name="minimum x")
clad_max_x = openmc.XPlane(6.0, name="maximum x")
clad_min_y = openmc.YPlane(-6.0, name="minimum y")
clad_max_y = openmc.YPlane(6.0, name="maximum y")
clad_min_z = openmc.ZPlane(-6.0, name="minimum z")
clad_max_z = openmc.ZPlane(6.0, name="maximum z")
clad_cell = openmc.Cell(name="clad")
clad_cell.region = (-fuel_min_x | +fuel_max_x |
-fuel_min_y | +fuel_max_y |
-fuel_min_z | +fuel_max_z) & \
(+clad_min_x & -clad_max_x &
+clad_min_y & -clad_max_y &
+clad_min_z & -clad_max_z)
clad_cell.fill = zirc_mat
if test_opts['external_geom']:
bounds = (15, 15, 15)
else:
bounds = (10, 10, 10)
water_min_x = openmc.XPlane(x0=-bounds[0],
name="minimum x",
boundary_type='vacuum')
water_max_x = openmc.XPlane(x0=bounds[0],
name="maximum x",
boundary_type='vacuum')
water_min_y = openmc.YPlane(y0=-bounds[1],
name="minimum y",
boundary_type='vacuum')
water_max_y = openmc.YPlane(y0=bounds[1],
name="maximum y",
boundary_type='vacuum')
water_min_z = openmc.ZPlane(z0=-bounds[2],
name="minimum z",
boundary_type='vacuum')
water_max_z = openmc.ZPlane(z0=bounds[2],
name="maximum z",
boundary_type='vacuum')
water_cell = openmc.Cell(name="water")
water_cell.region = (-clad_min_x | +clad_max_x |
-clad_min_y | +clad_max_y |
-clad_min_z | +clad_max_z) & \
(+water_min_x & -water_max_x &
+water_min_y & -water_max_y &
+water_min_z & -water_max_z)
water_cell.fill = water_mat
# create a containing universe
geometry = openmc.Geometry([fuel_cell, clad_cell, water_cell])
### Tallies ###
# create meshes and mesh filters
regular_mesh = openmc.RegularMesh()
regular_mesh.dimension = (10, 10, 10)
regular_mesh.lower_left = (-10.0, -10.0, -10.0)
regular_mesh.upper_right = (10.0, 10.0, 10.0)
regular_mesh_filter = openmc.MeshFilter(mesh=regular_mesh)
if test_opts['holes']:
mesh_filename = "test_mesh_tets_w_holes.e"
else:
mesh_filename = "test_mesh_tets.e"
uscd_mesh = openmc.UnstructuredMesh(mesh_filename, test_opts['library'])
uscd_filter = openmc.MeshFilter(mesh=uscd_mesh)
# create tallies
tallies = openmc.Tallies()
regular_mesh_tally = openmc.Tally(name="regular mesh tally")
regular_mesh_tally.filters = [regular_mesh_filter]
regular_mesh_tally.scores = ['flux']
regular_mesh_tally.estimator = test_opts['estimator']
tallies.append(regular_mesh_tally)
uscd_tally = openmc.Tally(name="unstructured mesh tally")
uscd_tally.filters = [uscd_filter]
uscd_tally.scores = ['flux']
uscd_tally.estimator = test_opts['estimator']
tallies.append(uscd_tally)
### Settings ###
settings = openmc.Settings()
settings.run_mode = 'fixed source'
settings.particles = 1000
settings.batches = 10
# source setup
r = openmc.stats.Uniform(a=0.0, b=0.0)
theta = openmc.stats.Discrete(x=[0.0], p=[1.0])
phi = openmc.stats.Discrete(x=[0.0], p=[1.0])
space = openmc.stats.SphericalIndependent(r, theta, phi)
energy = openmc.stats.Discrete(x=[15.e+06], p=[1.0])
source = openmc.Source(space=space, energy=energy)
settings.source = source
model = openmc.model.Model(geometry=geometry,
materials=materials,
tallies=tallies,
settings=settings)
harness = UnstructuredMeshTest('statepoint.10.h5', model,
test_opts['inputs_true'],
test_opts['holes'])
harness.main()
###############################################################################
# Simulation Input File Parameters
###############################################################################
# OpenMC simulation parameters
batches = 20
inactive = 10
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])
materials_file.export_to_xml()
###############################################################################
# Exporting to OpenMC geometry.xml file
import openmc
mats = openmc.Materials()
mat = openmc.Material(1)
mat.name = "Inner Core 1"
mat.set_density('sum')
mat.add_nuclide('Pu240', 1.17624E-04)
mat.add_nuclide('Pu241', 1.33836E-05)
mat.add_nuclide('U235', 1.26065E-05)
mat.add_nuclide('U238', 5.79290E-03)
mat.add_nuclide('Pu239', 8.86528E-04)
mat.add_nuclide('Pu238', 3.33696E-07)
mat.add_nuclide('Pu242', 1.40289E-06)
mat.add_nuclide('Am241', 2.95925E-06)
mat.add_element('Cr', 2.69344E-03)
mat.add_element('Ni', 1.19793E-03)
mat.add_element('Fe', 1.28611E-02)
mat.add_element('Al', 4.02385E-06)
mat.add_element('Na', 9.27752E-03)
mat.add_element('O', 1.37526E-02)
mat.add_element('C', 3.66446E-05)
mat.add_element('Mo', 2.36059E-04)
mat.add_element('Mn', 2.25651E-04)
mat.add_element('Cu', 2.46789E-05)
mat.add_element('Si', 1.62374E-04)
mat.add_element('Cl', 2.98410E-07)
mat.add_element('Co', 8.32631E-07)
mats.append(mat)
mat = openmc.Material(2)
# 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, # CASMO4 time step
# 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5,
# 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5, 12.5,
# 12.5, 25.0, 25.0, 25.0, 25.0, 25.0, 25.0, 25.0,
# 62.5, 62.5, 62.5, 62.5, 62.5, 62.5, 62.5, 62.5,
# 62.5, 62.5, 62.5, 62.5, 62.5, 62.5, 62.5, 62.5 ])*24*60*60
#chain_file = './chain_casl.xml'
power_den = 40.0 # W/cm, for 2D simulations only (use W for 3D) 40.0 W/gU
###############################################################################
# Define materials
###############################################################################
# Instantiate some Materials and register the appropriate Nuclides
fuel_21 = openmc.Material(material_id=1, name='UO2 fuel at 2.1% wt enrichment')
fuel_21.set_density('g/cm3', 10.257)
fuel_21.add_nuclide('U234', 4.02487E-06)
fuel_21.add_nuclide('U235', 4.86484E-04)
fuel_21.add_nuclide('U236', 2.23756E-06)
fuel_21.add_nuclide('U238', 2.23868E-02)
fuel_21.add_nuclide('O16', 4.57590E-02)
fuel_21.depletable = True
fuel_31 = openmc.Material(material_id=2, name='UO2 fuel at 3.1% wt enrichment')
fuel_31.set_density('g/cm3', 10.257)
fuel_31.add_nuclide('U234', 6.11864E-06)
fuel_31.add_nuclide('U235', 7.18132E-04)
fuel_31.add_nuclide('U236', 3.29861E-06)
fuel_31.add_nuclide('U238', 2.21546E-02)
fuel_31.add_nuclide('O16', 4.57642E-02)
def _build_inputs(self):
model = openmc.model.Model()
### MATERIALS ###
fuel = openmc.Material(name='no-void fuel')
fuel.set_density('g/cc', 10.29769)
fuel.add_nuclide('U234', 0.93120485)
fuel.add_nuclide('U235', 0.00055815)
fuel.add_nuclide('U238', 0.022408)
fuel.add_nuclide('O16', 0.045829)
cladding = openmc.Material(name='clad')
cladding.set_density('g/cc', 6.55)
cladding.add_nuclide('Zr90', 0.021827)
cladding.add_nuclide('Zr91', 0.00476)
cladding.add_nuclide('Zr92', 0.0072758)
cladding.add_nuclide('Zr94', 0.0073734)
cladding.add_nuclide('Zr96', 0.0011879)
water = openmc.Material(name='water')
water.set_density('g/cc', 0.740582)
water.add_nuclide('H1', 0.049457)
water.add_nuclide('O16', 0.024672)
water.add_nuclide('B10', 8.0042e-06)
water.add_nuclide('B11', 3.2218e-05)
water.add_s_alpha_beta('c_H_in_H2O')
model.materials = openmc.Materials([fuel, cladding, water])
### GEOMETRY ###
# create the DAGMC universe
pincell_univ = openmc.DAGMCUniverse(filename='dagmc.h5m',
auto_geom_ids=True)
# create a 2 x 2 lattice using the DAGMC pincell
pitch = np.asarray((24.0, 24.0))
lattice = openmc.RectLattice()
lattice.pitch = pitch
lattice.universes = [[pincell_univ] * 2] * 2
lattice.lower_left = -pitch
left = openmc.XPlane(x0=-pitch[0],
name='left',
boundary_type='reflective')
right = openmc.XPlane(x0=pitch[0],
name='right',
boundary_type='reflective')
front = openmc.YPlane(y0=-pitch[1],
name='front',
boundary_type='reflective')
back = openmc.YPlane(y0=pitch[1],
name='back',
boundary_type='reflective')
# clip the DAGMC geometry at +/- 10 cm w/ CSG planes
bottom = openmc.ZPlane(z0=-10.0,
name='bottom',
boundary_type='reflective')
top = openmc.ZPlane(z0=10.0, name='top', boundary_type='reflective')
bounding_region = +left & -right & +front & -back & +bottom & -top
bounding_cell = openmc.Cell(fill=lattice, region=bounding_region)
model.geometry = openmc.Geometry([bounding_cell])
# settings
model.settings.particles = 100
model.settings.batches = 10
model.settings.inactive = 2
model.settings.output = {'summary': False}
model.export_to_xml()
import openmc
mats = openmc.Materials()
mat = openmc.Material(1)
mat.name = "Plutonium nitrate solution"
mat.set_density('sum')
mat.add_nuclide('Pu239', 1.2164e-04)
mat.add_nuclide('Pu240', 3.9010e-06)
mat.add_nuclide('N14', 1.3452e-03)
mat.add_nuclide('H1', 6.3772e-02)
mat.add_nuclide('O16', 3.5500e-02)
mat.add_element('Fe', 2.0380e-06)
mats.append(mat)
mat = openmc.Material(2)
mat.name = "347 stainless steel"
mat.set_density('sum')
mat.add_element('Fe', 6.0386e-02)
mat.add_element('Cr', 1.6678e-02)
mat.add_element('Ni', 9.8504e-03)
mats.append(mat)
mat = openmc.Material(3)
mat.name = "Water at 27 C"
mat.set_density('sum')
mat.add_nuclide('H1', 6.6622e-02)
mat.add_nuclide('O16', 3.3311e-02)
mats.append(mat)
mats.export_to_xml()
import openmc
import numpy as np
fuel = openmc.Material(1, 'fuel')
fuel.add_nuclide('O16', 4.6753e-2)
fuel.add_nuclide('U234', 5.1835e-6)
fuel.add_nuclide('U235', 1.0102e-3)
fuel.add_nuclide('U236', 5.1395e-6)
fuel.add_nuclide('U238', 2.2157e-2)
fuel.set_density('g/cm3', 10.4)
water = openmc.Material(2, 'water')
water.add_nuclide('H1', 6.6675e-2)
water.add_nuclide('O16', 3.3338e-2)
water.set_density('g/cm3', .997297)
water.add_s_alpha_beta('c_H_in_H2O')
clad6061 = openmc.Material(3, 'clad6061')
clad6061.add_element('Mg', 6.6651e-4)
clad6061.add_nuclide('Al27', 5.8433e-2)
clad6061.add_element('Si', 3.4607e-4)
clad6061.add_element('Ti', 2.5375e-5)
clad6061.add_element('Cu', 6.3731e-5)
clad6061.add_element('Zn', 3.0967e-5)
clad6061.add_element('Ni', 8.3403e-3)
clad6061.add_element('Cr', 6.2310e-5)
clad6061.add_element('Fe', 1.0152e-4)
clad6061.add_nuclide('Mn55', 2.2115e-5)
clad6061.set_density('g/cm3', 2.69)
rubber = openmc.Material(4, 'rubber')
def build_model(ppm_Boron):
# Create the pin materials
fuel = openmc.Material(name='1.6% Fuel')
fuel.set_density('g/cm3', 10.31341)
fuel.add_element('U', 1., enrichment=1.6)
fuel.add_element('O', 2.)
zircaloy = openmc.Material(name='Zircaloy')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_element('Zr', 1.)
water = openmc.Material(name='Borated Water')
water.set_density('g/cm3', 0.741)
water.add_element('H', 2.)
water.add_element('O', 1.)
# Include the amount of boron in the water based on the ppm,
# neglecting the other constituents of boric acid
water.add_element('B', ppm_Boron * 1e-6)
# Instantiate a Materials object
materials = openmc.Materials([fuel, zircaloy, water])
# Create cylinders for the fuel and clad
fuel_outer_radius = openmc.ZCylinder(r=0.39218)
clad_outer_radius = openmc.ZCylinder(r=0.45720)
# Create boundary planes to surround the geometry
min_x = openmc.XPlane(x0=-0.63, boundary_type='reflective')
max_x = openmc.XPlane(x0=+0.63, boundary_type='reflective')
min_y = openmc.YPlane(y0=-0.63, boundary_type='reflective')
max_y = openmc.YPlane(y0=+0.63, boundary_type='reflective')
# Create fuel Cell
fuel_cell = openmc.Cell(name='1.6% Fuel')
fuel_cell.fill = fuel
fuel_cell.region = -fuel_outer_radius
# Create a clad Cell
clad_cell = openmc.Cell(name='1.6% Clad')
clad_cell.fill = zircaloy
clad_cell.region = +fuel_outer_radius & -clad_outer_radius
# Create a moderator Cell
moderator_cell = openmc.Cell(name='1.6% Moderator')
moderator_cell.fill = water
moderator_cell.region = +clad_outer_radius & (+min_x & -max_x & +min_y
& -max_y)
# Create root Universe
root_universe = openmc.Universe(name='root universe', universe_id=0)
root_universe.add_cells([fuel_cell, clad_cell, moderator_cell])
# Create Geometry and set root universe
geometry = openmc.Geometry(root_universe)
# Finish with the settings file
settings = openmc.Settings()
settings.batches = 300
settings.inactive = 200
settings.particles = 10000
settings.run_mode = 'eigenvalue'
# Create an initial uniform spatial source distribution over fissionable zones
bounds = [-0.63, -0.63, -10, 0.63, 0.63, 10.]
uniform_dist = openmc.stats.Box(bounds[:3],
bounds[3:],
only_fissionable=True)
settings.source = openmc.source.Source(space=uniform_dist)
# We dont need a tallies file so dont waste the disk input/output time
settings.output = {'tallies': False}
model = openmc.model.Model(geometry, materials, settings)
return model
def test_external_mesh(cpp_driver):
# Materials
materials = openmc.Materials()
fuel_mat = openmc.Material(name="fuel")
fuel_mat.add_nuclide("U235", 1.0)
fuel_mat.set_density('g/cc', 4.5)
materials.append(fuel_mat)
zirc_mat = openmc.Material(name="zircaloy")
zirc_mat.add_element("Zr", 1.0)
zirc_mat.set_density("g/cc", 5.77)
materials.append(zirc_mat)
water_mat = openmc.Material(name="water")
water_mat.add_nuclide("H1", 2.0)
water_mat.add_nuclide("O16", 1.0)
water_mat.set_density("atom/b-cm", 0.07416)
materials.append(water_mat)
materials.export_to_xml()
# Geometry
fuel_min_x = openmc.XPlane(-5.0, name="minimum x")
fuel_max_x = openmc.XPlane(5.0, name="maximum x")
fuel_min_y = openmc.YPlane(-5.0, name="minimum y")
fuel_max_y = openmc.YPlane(5.0, name="maximum y")
fuel_min_z = openmc.ZPlane(-5.0, name="minimum z")
fuel_max_z = openmc.ZPlane(5.0, name="maximum z")
fuel_cell = openmc.Cell(name="fuel")
fuel_cell.region = +fuel_min_x & -fuel_max_x & \
+fuel_min_y & -fuel_max_y & \
+fuel_min_z & -fuel_max_z
fuel_cell.fill = fuel_mat
clad_min_x = openmc.XPlane(-6.0, name="minimum x")
clad_max_x = openmc.XPlane(6.0, name="maximum x")
clad_min_y = openmc.YPlane(-6.0, name="minimum y")
clad_max_y = openmc.YPlane(6.0, name="maximum y")
clad_min_z = openmc.ZPlane(-6.0, name="minimum z")
clad_max_z = openmc.ZPlane(6.0, name="maximum z")
clad_cell = openmc.Cell(name="clad")
clad_cell.region = (-fuel_min_x | +fuel_max_x |
-fuel_min_y | +fuel_max_y |
-fuel_min_z | +fuel_max_z) & \
(+clad_min_x & -clad_max_x &
+clad_min_y & -clad_max_y &
+clad_min_z & -clad_max_z)
clad_cell.fill = zirc_mat
bounds = (10, 10, 10)
water_min_x = openmc.XPlane(x0=-bounds[0],
name="minimum x",
boundary_type='vacuum')
water_max_x = openmc.XPlane(x0=bounds[0],
name="maximum x",
boundary_type='vacuum')
water_min_y = openmc.YPlane(y0=-bounds[1],
name="minimum y",
boundary_type='vacuum')
water_max_y = openmc.YPlane(y0=bounds[1],
name="maximum y",
boundary_type='vacuum')
water_min_z = openmc.ZPlane(z0=-bounds[2],
name="minimum z",
boundary_type='vacuum')
water_max_z = openmc.ZPlane(z0=bounds[2],
name="maximum z",
boundary_type='vacuum')
water_cell = openmc.Cell(name="water")
water_cell.region = (-clad_min_x | +clad_max_x |
-clad_min_y | +clad_max_y |
-clad_min_z | +clad_max_z) & \
(+water_min_x & -water_max_x &
+water_min_y & -water_max_y &
+water_min_z & -water_max_z)
water_cell.fill = water_mat
# create a containing universe
geometry = openmc.Geometry([fuel_cell, clad_cell, water_cell])
# Meshes
mesh_filename = "test_mesh_tets.h5m"
# Create a normal unstructured mesh to compare to
uscd_mesh = openmc.UnstructuredMesh(mesh_filename, 'moab')
# Create filters
uscd_filter = openmc.MeshFilter(mesh=uscd_mesh)
# Tallies
tallies = openmc.Tallies()
uscd_tally = openmc.Tally(name="unstructured mesh tally")
uscd_tally.filters = [uscd_filter]
uscd_tally.scores = ['flux']
uscd_tally.estimator = 'tracklength'
tallies.append(uscd_tally)
# Settings
settings = openmc.Settings()
settings.run_mode = 'fixed source'
settings.particles = 100
settings.batches = 10
# Source setup
space = openmc.stats.Point()
angle = openmc.stats.Monodirectional((-1.0, 0.0, 0.0))
energy = openmc.stats.Discrete(x=[15.e+06], p=[1.0])
source = openmc.Source(space=space, energy=energy, angle=angle)
settings.source = source
model = openmc.model.Model(geometry=geometry,
materials=materials,
tallies=tallies,
settings=settings)
harness = ExternalMoabTest(cpp_driver, 'statepoint.10.h5', model)
# Run open MC and check results
harness.main()
import openmc
mats = openmc.Materials()
mat = openmc.Material(1)
mat.name = "HEU"
mat.set_density('sum')
mat.add_nuclide('U234', 5.2111e-04)
mat.add_nuclide('U235', 4.2064e-02)
mat.add_nuclide('U238', 4.3626e-03)
mat.add_element('C', 1.1074e-03)
mat.add_element('Fe', 1.9320e-04)
mat.add_element('W', 5.3798e-05)
mats.append(mat)
mats.export_to_xml()
import openmc
clad = openmc.Material(1, 'cladding')
clad.add_element('C', 4.3736e-5)
clad.add_element('Si', 1.0627e-3)
clad.add_element('S', 2.9782e-6)
clad.add_element('Ni', 8.3403e-3)
clad.add_element('Cr', 1.6775e-2)
clad.add_nuclide('P31', 4.3170e-5)
clad.add_element('Fe', 5.9421e-2)
clad.add_nuclide('Mn55', 1.1561e-3)
clad.set_density('atom/b-cm', 8.68449842E-02)
air = openmc.Material(2, 'water')
air.add_nuclide('N14', 3.9016e-5)
air.add_nuclide('O16', 1.0409e-5)
air.set_density('atom/b-cm', 4.94250000E-05)
fuel = openmc.Material(3, 'fuel')
fuel.add_nuclide('H1', 5.7176E-02)
fuel.add_nuclide('N14', 2.8156E-03)
fuel.add_nuclide('O16', 3.7836E-02)
fuel.add_nuclide('U234', 5.9840E-07)
fuel.add_nuclide('U235', 7.4257E-05)
fuel.add_nuclide('U236', 7.4165E-08)
fuel.add_nuclide('U238', 6.6142E-04)
fuel.set_density('atom/b-cm', 9.85636637E-02)
fuel.add_s_alpha_beta('c_H_in_H2O')
mats = openmc.Materials([clad, air, fuel])
mats.cross_sections = '/home/tang/nndc_hdf5/cross_sections.xml'
# OpenMC simulation parameters
batches = 15
inactive = 5
particles = 10000
###############################################################################
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
# Instantiate some Materials and register the appropriate Nuclides
moderator = openmc.Material(material_id=41, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])
materials_file.default_xs = '71c'
materials_file.export_to_xml()
###############################################################################
#!/usr/bin/env python3
"""example_isotope_plot.py: plots few 2D views of a simple tokamak geometry with neutron flux."""
__author__ = "Jonathan Shimwell"
import openmc
import matplotlib.pyplot as plt
import os
#MATERIALS#
breeder_material = openmc.Material(1, "PbLi") #Pb84.2Li15.8 with natural enrichment of Li6
enrichment_fraction = 0.97
breeder_material.add_element('Pb', 84.2,'ao')
breeder_material.add_nuclide('Li6', enrichment_fraction*15.8, 'ao')
breeder_material.add_nuclide('Li7', (1.0-enrichment_fraction)*15.8, 'ao')
breeder_material.set_density('atom/b-cm',3.2720171e-2) # around 11 g/cm3
copper = openmc.Material(name='Copper')
copper.set_density('g/cm3', 8.5)
copper.add_element('Cu', 1.0)
eurofer = openmc.Material(name='EUROFER97')
eurofer.set_density('g/cm3', 7.75)
eurofer.add_element('Fe', 89.067, percent_type='wo')
eurofer.add_element('C', 0.11, percent_type='wo')
eurofer.add_element('Mn', 0.4, percent_type='wo')
eurofer.add_element('Cr', 9.0, percent_type='wo')
eurofer.add_element('Ta', 0.12, percent_type='wo')
eurofer.add_element('W', 1.1, percent_type='wo')
