def buildTriangUni(modSetting,pitch,diameter,boron,modT,fuelT,DyFrac,enrich):
uo2=getU(fuelT,DyFrac,enrich)
mod=getMod(modSetting,boron,modT) #loads the moderator
mats = openmc.Materials([uo2, mod])
mats.export_to_xml()
####################build the surfaces################################
fuel_or1 = openmc.ZCylinder(R=diameter/2)
box = openmc.get_hexagonal_prism(edge_length=pitch/math.sqrt(3.0),boundary_type='reflective')
#################Build the Cells #######################################
fuel_region = -fuel_or1 #beautiful abuse of operator overloading
mod_region = box & +fuel_or1
# c
# c Fuel
# c
# c throwback to MCNP
fuel = openmc.Cell(1, 'fuel')
fuel.fill = uo2
fuel.region = fuel_region
#
# Moderator
#
moderator = openmc.Cell(4, 'moderator')
moderator.fill = mod
moderator.region = mod_region
#I am groot!
groot = openmc.Universe(cells=(fuel, moderator))
geom = openmc.Geometry(groot)
geom.export_to_xml()
cell_filter = openmc.CellFilter([fuel, moderator])
# tallies
buildTallies(cell_filter)
return (fuel,moderator)
import openmc
pitch =2.0
fuel_or = openmc.ZCylinder(R=0.5)
fuel_region = -fuel_or
fuel_cell = openmc.Cell(1, 'fuel')
fuel_cell.region = fuel_region
hexagon = openmc.get_hexagonal_prism(edge_length = 1/3**(1/2)*pitch,
orientation = 'x',
boundary_type = 'reflective')
bottom = openmc.ZPlane(z0=-pitch/2, boundary_type='reflective')
top = openmc.ZPlane(z0=pitch/2, boundary_type='reflective')
water_region = +bottom & -top & hexagon & +fuel_or
moderator = openmc.Cell(2, 'moderator')
moderator.region = water_region
root = openmc.Universe(cells=(fuel_cell, moderator))
geom = openmc.Geometry(root)
geom.export_to_xml()
print(water_region)
print('(0,0,0)')
print((0,0,0) in fuel_region,(0,0,0) in water_region)
print('(0,0.6,0)')
print((0,0.6,0) in fuel_region,(0,0.6,0) in water_region)
print('(-pitch,pitch,0)')
print((-pitch,pitch,0) in fuel_region,(-pitch,pitch,0) in water_region)
def hexfunction(pitch, pack_frac):
settings = openmc.Settings()
# Set high tolerance to allow use of lower temperature xs
settings.temperature['tolerance'] = 1000
settings.temperature['method'] = 'nearest'
settings.temperature['multipole'] = True
settings.cutoff = {'energy': 1e-8} #energy cutoff in eV
#############################
### MATERIALS ###
#############################
enrichment = 20.0
uo2 = openmc.Material(1, "uo2")
uo2.add_element('U', 1.0, enrichment=enrichment)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.97)
uo2.temperature = 900 #kelvin
graphite = openmc.Material(2, "graphite")
graphite.set_density('g/cm3', 1.1995)
graphite.add_element('C', 1.0)
graphite.add_s_alpha_beta('c_Graphite')
graphite.temperature = 900 #kelvin
sodium = openmc.Material(3, "sodium")
sodium.set_density('g/cm3', 0.8017) # 900 K
sodium.add_element('Na', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
sodium.temperature = 900 #kelvin
naoh = openmc.Material(6, "naoh")
naoh.set_density('g/cm3', 1.5) # 900 K
naoh.add_element('Na', 1.0)
naoh.add_element('O', 1.0)
naoh.add_element('H', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
naoh.temperature = 900 #kelvin
uo2 = openmc.Material(1, "uo2")
uo2.add_element('U', 1.0, enrichment=enrichment)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.97)
uo2.temperature = 900 #kelvin
fuel_temp = 900
homogeneous_fuel = build_fuel_material(4, fuel_temp, pack_frac)
mats = openmc.Materials([uo2, graphite, sodium, naoh, homogeneous_fuel])
mats.export_to_xml()
#############################
### GEOMETRY ###
#############################
universe = openmc.Universe()
coolant_cyl = openmc.ZCylinder(R=0.5)
coolant_region = -coolant_cyl
coolant_cell = openmc.Cell(1, 'coolant')
coolant_cell.fill = naoh
coolant_cell.region = coolant_region
hex_prism = openmc.get_hexagonal_prism(edge_length=pitch / (3**1 / 2),
boundary_type='reflective')
top = openmc.YPlane(y0=pitch)
bottom = openmc.YPlane(y0=-pitch)
fuel_region = hex_prism & -top & +bottom
fuel_cell = openmc.Cell(2, 'moderator')
fuel_cell.fill = homogeneous_fuel
fuel_cell.region = fuel_region
root = openmc.Universe(cells=(fuel_cell, coolant_cell))
geom = openmc.Geometry(root)
geom.export_to_xml()
#####################################
### SOURCE/BATCHES ###
#####################################
point = openmc.stats.Point((0, 0, 0))
src = openmc.Source(space=point)
settings.source = src
settings.batches = 50
settings.inactive = 10
settings.particles = 200
settings.export_to_xml()
#############################
### TALLIES ###
#############################
# Instantiate an empty Tallies object
tallies_file = openmc.Tallies()
# K-Eigenvalue (infinity) tallies
fiss_rate = openmc.Tally(name='fiss. rate')
fiss_rate.scores = ['nu-fission']
tallies_file.append(fiss_rate)
abs_rate = openmc.Tally(name='abs. rate')
abs_rate.scores = ['absorption']
tallies_file.append(abs_rate)
# Resonance Escape Probability tallies
therm_abs_rate = openmc.Tally(name='therm. abs. rate')
therm_abs_rate.scores = ['absorption']
therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_abs_rate)
# Thermal Flux Utilization tallies
fuel_therm_abs_rate = openmc.Tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate.scores = ['absorption']
fuel_therm_abs_rate.filters = [
openmc.EnergyFilter([0., 0.625]),
openmc.CellFilter([fuel_cell])
]
tallies_file.append(fuel_therm_abs_rate)
# Fast Fission Factor tallies
therm_fiss_rate = openmc.Tally(name='therm. fiss. rate')
therm_fiss_rate.scores = ['nu-fission']
therm_fiss_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_fiss_rate)
tallies_file.export_to_xml()
#############################
### PLOTTING ###
#############################
p = openmc.Plot()
p.filename = 'pinplot'
p.width = (2 * pitch, 2 * pitch)
p.pixels = (200, 200)
p.color_by = 'material'
p.colors = {homogeneous_fuel: 'yellow', sodium: 'grey'}
plots = openmc.Plots([p])
plots.export_to_xml()
# openmc.plot_geometry(output = False)
openmc.plot_geometry()
pngstring = 'pinplot{}.png'.format(str(pitch))
subprocess.call(['convert', 'pinplot.ppm', pngstring])
subprocess.call(['mv', pngstring, 'figures/' + pngstring])
#############################
### EXECUTION ###
#############################
# openmc.run(output=False)
openmc.run()
sp = openmc.StatePoint('statepoint.{}.h5'.format(settings.batches))
# Collect all the tallies
fiss_rate = sp.get_tally(name='fiss. rate')
fiss_rate_df = fiss_rate.get_pandas_dataframe()
abs_rate = sp.get_tally(name='abs. rate')
abs_rate_df = abs_rate.get_pandas_dataframe()
therm_abs_rate = sp.get_tally(name='therm. abs. rate')
therm_abs_rate_df = therm_abs_rate.get_pandas_dataframe()
fuel_therm_abs_rate = sp.get_tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate_df = fuel_therm_abs_rate.get_pandas_dataframe()
therm_fiss_rate = sp.get_tally(name='therm. fiss. rate')
therm_fiss_rate_df = therm_fiss_rate.get_pandas_dataframe()
# Compute k-infinity
kinf = fiss_rate / abs_rate
kinf_df = kinf.get_pandas_dataframe()
# Compute resonance escape probability
res_esc = (therm_abs_rate) / (abs_rate)
res_esc_df = res_esc.get_pandas_dataframe()
# Compute fast fission factor
fast_fiss = fiss_rate / therm_fiss_rate
fast_fiss_df = fast_fiss.get_pandas_dataframe()
# Compute thermal flux utilization
therm_util = fuel_therm_abs_rate / therm_abs_rate
therm_util_df = therm_util.get_pandas_dataframe()
# Compute neutrons produced per absorption
eta = therm_fiss_rate / fuel_therm_abs_rate
eta_df = eta.get_pandas_dataframe()
columns = [
'pitch', 'enrichment', 'kinf mean', 'kinf sd', 'res_esc mean',
'res_esc sd', 'fast_fiss mean', 'fast_fiss sd', 'therm_util mean',
'therm_util sd', 'eta mean', 'eta sd'
]
data = [[
pitch, enrichment, kinf_df['mean'][0], kinf_df['std. dev.'][0],
res_esc_df['mean'][0], res_esc_df['std. dev.'][0],
fast_fiss_df['mean'][0], fast_fiss_df['std. dev.'][0],
therm_util_df['mean'][0], therm_util_df['std. dev.'][0],
eta_df['mean'][0], eta_df['std. dev.'][0]
]]
all_tallies = pd.DataFrame(data, columns=columns)
return all_tallies
def pincellfunction(pitch, enrichment):
settings = openmc.Settings()
# Set high tolerance to allow use of lower temperature xs
settings.temperature['tolerance'] = 10000
settings.temperature['method'] = 'nearest'
settings.temperature['multipole'] = True
settings.cutoff = {'energy': 1e-8} #energy cutoff in eV
#############################
### MATERIALS ###
#############################
uo2 = openmc.Material(1, "uo2")
uo2.add_element('U', 1.0, enrichment=enrichment)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.97)
uo2.temperature = 900 #kelvin
water = openmc.Material(3, "h2o")
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
#Using P = 15.5 Mpa
water.set_density('g/cm3', 0.66)
water.add_s_alpha_beta('c_H_in_H2O')
water.temperature = 600 #kelvin
mats = openmc.Materials([uo2, water])
mats.export_to_xml()
#############################
### GEOMETRY ###
#############################
universe = openmc.Universe()
fuel_or = openmc.ZCylinder(R=0.5)
fuel_region = -fuel_or
fuel_cell = openmc.Cell(1, 'fuel')
fuel_cell.fill = uo2
fuel_cell.region = fuel_region
hexagon = openmc.get_hexagonal_prism(edge_length=1 / 3**(1 / 2) * pitch,
orientation='y',
boundary_type='reflective')
bottom = openmc.ZPlane(z0=-pitch / 2, boundary_type='reflective')
top = openmc.ZPlane(z0=pitch / 2, boundary_type='reflective')
water_region = hexagon & +fuel_or
moderator = openmc.Cell(2, 'moderator')
moderator.fill = water
moderator.region = water_region
root = openmc.Universe(cells=(fuel_cell, moderator))
geom = openmc.Geometry(root)
geom.export_to_xml()
#####################################
### SOURCE/BATCHES ###
#####################################
point = openmc.stats.Point((0, 0, 0))
src = openmc.Source(space=point)
settings.source = src
settings.batches = 100
settings.inactive = 10
settings.particles = 1000
settings.export_to_xml()
#############################
### TALLIES ###
#############################
# Instantiate an empty Tallies object
tallies_file = openmc.Tallies()
# K-Eigenvalue (infinity) tallies
fiss_rate = openmc.Tally(name='fiss. rate')
fiss_rate.scores = ['nu-fission']
tallies_file.append(fiss_rate)
abs_rate = openmc.Tally(name='abs. rate')
abs_rate.scores = ['absorption']
tallies_file.append(abs_rate)
# Resonance Escape Probability tallies
therm_abs_rate = openmc.Tally(name='therm. abs. rate')
therm_abs_rate.scores = ['absorption']
therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_abs_rate)
# Thermal Flux Utilization tallies
fuel_therm_abs_rate = openmc.Tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate.scores = ['absorption']
fuel_therm_abs_rate.filters = [
openmc.EnergyFilter([0., 0.625]),
openmc.CellFilter([fuel_cell])
]
tallies_file.append(fuel_therm_abs_rate)
# Fast Fission Factor tallies
therm_fiss_rate = openmc.Tally(name='therm. fiss. rate')
therm_fiss_rate.scores = ['nu-fission']
therm_fiss_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_fiss_rate)
tallies_file.export_to_xml()
#############################
### PLOTTING ###
#############################
p = openmc.Plot()
p.filename = 'pinplot'
p.width = (1.5 * pitch, 1.5 * pitch)
p.pixels = (200, 200)
p.color_by = 'material'
p.colors = {uo2: 'yellow', water: 'blue'}
plots = openmc.Plots([p])
plots.export_to_xml()
openmc.plot_geometry()
pngstring = 'pinplot{}.png'.format(str(pitch))
subprocess.call(['convert', 'pinplot.ppm', pngstring])
subprocess.call(['mv', pngstring, 'figures/' + pngstring])
#############################
### EXECUTION ###
#############################
openmc.run()
sp = openmc.StatePoint('statepoint.{}.h5'.format(settings.batches))
# Collect all the tallies
fiss_rate = sp.get_tally(name='fiss. rate')
fiss_rate_df = fiss_rate.get_pandas_dataframe()
abs_rate = sp.get_tally(name='abs. rate')
abs_rate_df = abs_rate.get_pandas_dataframe()
therm_abs_rate = sp.get_tally(name='therm. abs. rate')
therm_abs_rate_df = therm_abs_rate.get_pandas_dataframe()
fuel_therm_abs_rate = sp.get_tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate_df = fuel_therm_abs_rate.get_pandas_dataframe()
therm_fiss_rate = sp.get_tally(name='therm. fiss. rate')
therm_fiss_rate_df = therm_fiss_rate.get_pandas_dataframe()
# Compute k-infinity
kinf = fiss_rate / abs_rate
kinf_df = kinf.get_pandas_dataframe()
# Compute resonance escape probability
res_esc = (therm_abs_rate) / (abs_rate)
res_esc_df = res_esc.get_pandas_dataframe()
# Compute fast fission factor
fast_fiss = fiss_rate / therm_fiss_rate
fast_fiss_df = fast_fiss.get_pandas_dataframe()
# Compute thermal flux utilization
therm_util = fuel_therm_abs_rate / therm_abs_rate
therm_util_df = therm_util.get_pandas_dataframe()
# Compute neutrons produced per absorption
eta = therm_fiss_rate / fuel_therm_abs_rate
eta_df = eta.get_pandas_dataframe()
columns = [
'pitch', 'enrichment', 'kinf mean', 'kinf sd', 'res_esc mean',
'res_esc sd', 'fast_fiss mean', 'fast_fiss sd', 'therm_util mean',
'therm_util sd', 'eta mean', 'eta sd'
]
data = [[
pitch, enrichment, kinf_df['mean'][0], kinf_df['std. dev.'][0],
res_esc_df['mean'][0], res_esc_df['std. dev.'][0],
fast_fiss_df['mean'][0], fast_fiss_df['std. dev.'][0],
therm_util_df['mean'][0], therm_util_df['std. dev.'][0],
eta_df['mean'][0], eta_df['std. dev.'][0]
]]
all_tallies = pd.DataFrame(data, columns=columns)
return all_tallies
def hexfunction(pitch, pack_frac):
settings = openmc.Settings()
# Set high tolerance to allow use of lower temperature xs
settings.temperature['tolerance'] = 1000
settings.temperature['method'] = 'nearest'
settings.temperature['multipole'] = True
settings.cutoff = {'energy': 1e-8} #energy cutoff in eV
#############################
### MATERIALS ###
#############################
mat_list = []
enrichment = 20.0
uo2 = openmc.Material(1, "uo2")
uo2.add_element('U', 1.0, enrichment=enrichment)
uo2.add_element('O', 2.0)
uo2.set_density('g/cm3', 10.97)
uo2.temperature = 900 #kelvin
mat_list.append(uo2)
graphite = openmc.Material(2, "graphite")
graphite.set_density('g/cm3', 1.1995)
graphite.add_element('C', 1.0)
graphite.add_s_alpha_beta('c_Graphite')
graphite.temperature = 900 #kelvin
mat_list.append(graphite)
# sodium = openmc.Material(3, "sodium")
sodium = openmc.Material()
sodium.set_density('g/cm3', 0.8017) # 900 K
sodium.add_element('Na', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
sodium.temperature = 900 #kelvin
mat_list.append(sodium)
# naoh = openmc.Material(6, "naoh")
naoh = openmc.Material()
naoh.set_density('g/cm3', 1.5) # 900 K
naoh.add_element('Na', 1.0)
naoh.add_element('O', 1.0)
naoh.add_element('H', 1.0)
# sodium.add_s_alpha_beta('c_Graphite')
naoh.temperature = 900 #kelvin
mat_list.append(naoh)
# TRISO Materials
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.5)
# fuel.add_nuclide('U235', 4.6716e-02)
fuel.add_nuclide('U235', 0.0667372)
# fuel.add_nuclide('U238', 2.8697e-01)
fuel.add_nuclide('U238', 0.2669488)
fuel.add_nuclide('O16', 5.0000e-01)
fuel.add_element('C', 1.6667e-01)
mat_list.append(fuel)
buff = openmc.Material(name='Buffer')
buff.set_density('g/cm3', 1.0)
buff.add_element('C', 1.0)
buff.add_s_alpha_beta('c_Graphite')
mat_list.append(buff)
PyC1 = openmc.Material(name='PyC1')
PyC1.set_density('g/cm3', 1.9)
PyC1.add_element('C', 1.0)
PyC1.add_s_alpha_beta('c_Graphite')
mat_list.append(PyC1)
PyC2 = openmc.Material(name='PyC2')
PyC2.set_density('g/cm3', 1.87)
PyC2.add_element('C', 1.0)
PyC2.add_s_alpha_beta('c_Graphite')
mat_list.append(PyC2)
SiC = openmc.Material(name='SiC')
SiC.set_density('g/cm3', 3.2)
SiC.add_element('C', 0.5)
SiC.add_element('Si', 0.5)
mat_list.append(SiC)
fuel_temp = 900
homogeneous_fuel = build_fuel_material(fuel_temp, pack_frac)
mat_list.append(homogeneous_fuel)
mats = openmc.Materials(mat_list)
mats.export_to_xml()
#############################
### GEOMETRY ###
#############################
pitch = 17.4
fuel_bottom = -pitch / 2
fuel_top = pitch / 2
coolant_r = 4
# fuel_r = (pitch/2 - coolant_r)/2
fuel_r = (pitch / (3**(1 / 2)) - coolant_r) / 2
hex_universe = openmc.Universe()
top = openmc.ZPlane(z0=fuel_top, boundary_type='reflective')
bottom = openmc.ZPlane(z0=fuel_bottom, boundary_type='reflective')
surf_fuel = openmc.ZCylinder(r=fuel_r)
# Make TRISOS to be filled in fuel cylinders by chopping up
# fuel cylinder into segments
n_cyls = 40
fuel_segment_heights = np.linspace(fuel_bottom, fuel_top, n_cyls)
segment_height = fuel_segment_heights[1] - fuel_segment_heights[0]
fuel_planes = [bottom]
fuel_cells = []
for i, height in enumerate(fuel_segment_heights[1:-1]):
this_plane = openmc.ZPlane(z0=height)
fuel_planes.append(this_plane)
this_cell = openmc.Cell()
this_cell.region = +fuel_planes[i] & -fuel_planes[i + 1] & -surf_fuel
fuel_cells.append(copy.deepcopy(this_cell))
# last cell
fuel_planes.append(top)
this_cell = openmc.Cell()
this_cell.region = +fuel_planes[-2] & -fuel_planes[-1] & -surf_fuel
fuel_cells.append(copy.deepcopy(this_cell))
# Make fuel cylinder
fuel_cyl_top = openmc.ZPlane(z0=segment_height / 2)
fuel_cyl_bottom = openmc.ZPlane(z0=-segment_height / 2)
fuel_triso_region = -surf_fuel & +fuel_cyl_bottom & -fuel_cyl_top
outer_radius = 425. * 1e-4
# openmc.model.triso._Cylinder.from_region(fuel_region, outer_radius)
spheres = [openmc.Sphere(r=r * 1e-4) for r in [215., 315., 350., 385.]]
cells = [
openmc.Cell(fill=fuel, region=-spheres[0]),
openmc.Cell(fill=buff, region=+spheres[0] & -spheres[1]),
openmc.Cell(fill=PyC1, region=+spheres[1] & -spheres[2]),
openmc.Cell(fill=SiC, region=+spheres[2] & -spheres[3]),
openmc.Cell(fill=PyC2, region=+spheres[3])
]
triso_univ = openmc.Universe(cells=cells)
outer_radius = 425. * 1e-4
centers = openmc.model.pack_spheres(radius=outer_radius,
region=fuel_triso_region,
pf=pack_frac)
trisos = [openmc.model.TRISO(outer_radius, triso_univ, c) for c in centers]
outside_trisos = openmc.Intersection(~t.region for t in trisos)
# background_region = outside_trisos & +fuel_cyl_bottom & \
# -fuel_cyl_top & -surf_fuel
background_region = outside_trisos
background_cell = openmc.Cell(fill=graphite, region=background_region)
fuel_triso_univ = openmc.Universe()
fuel_triso_univ.add_cell(background_cell)
for idx, triso in enumerate(trisos):
fuel_triso_univ.add_cell(triso)
# Fill in fuel cells with triso cells and translate to location
for i, cell in enumerate(fuel_cells):
cell_height = segment_height * (i + 1 / 2) + fuel_bottom
cell.translation = [0, 0, cell_height]
cell.fill = fuel_triso_univ
fuel_cell_univ = openmc.Universe(cells=fuel_cells)
# For testing solid fuel
# test_region = +bottom & -top & -surf_fuel
# fuel_cell = openmc.Cell(region=test_region, fill=fuel)
# fuel_cell_univ = openmc.Universe(cells=[fuel_cell])
coolant_cyl = openmc.ZCylinder(r=coolant_r)
coolant_region = -coolant_cyl
coolant_cell = openmc.Cell()
coolant_cell.fill = naoh
coolant_cell.region = coolant_region
hex_universe.add_cell(coolant_cell)
hex_prism = openmc.get_hexagonal_prism(edge_length=pitch / (3**1 / 2),
boundary_type='reflective')
graphite_region = hex_prism & +coolant_cyl & -top & +bottom
graphite_cell = openmc.Cell()
graphite_cell.fill = graphite
graphite_cell.region = graphite_region
hex_universe.add_cell(graphite_cell)
fuel_cells = []
root3 = 3**(1 / 2)
half_to_vertex = pitch / root3 / 2
half_to_edge = pitch / 4
# fuel_id = 100
offset_angle = 30
n_pins = 6
for i in range(n_pins):
theta = (offset_angle + i / n_pins * 360) * pi / 180
r = coolant_r + fuel_r + 0.01
x = r * np.cos(theta)
y = r * np.sin(theta)
fuel_cyl_bound = openmc.ZCylinder(x0=x, y0=y, r=fuel_r)
graphite_cell.region &= +fuel_cyl_bound
fuel_cell = openmc.Cell()
fuel_cell.fill = copy.deepcopy(fuel_cell_univ)
fuel_cell.translation = [x, y, 0]
fuel_cell.region = -fuel_cyl_bound & -top & +bottom
# fuel_cell.id = fuel_id
# fuel_id += 1
fuel_cells.append(fuel_cell)
hex_universe.add_cell(fuel_cell)
geom = openmc.Geometry(hex_universe)
# geom = openmc.Geometry(fuel_cell_univ)
geom.export_to_xml()
#####################################
### SOURCE/BATCHES ###
#####################################
point = openmc.stats.Point((0, 0, 0))
src = openmc.Source(space=point)
settings.source = src
settings.batches = 50
settings.inactive = 10
settings.particles = 200
settings.export_to_xml()
#############################
### TALLIES ###
#############################
# Instantiate an empty Tallies object
tallies_file = openmc.Tallies()
# K-Eigenvalue (infinity) tallies
fiss_rate = openmc.Tally(name='fiss. rate')
fiss_rate.scores = ['nu-fission']
tallies_file.append(fiss_rate)
abs_rate = openmc.Tally(name='abs. rate')
abs_rate.scores = ['absorption']
tallies_file.append(abs_rate)
# Resonance Escape Probability tallies
therm_abs_rate = openmc.Tally(name='therm. abs. rate')
therm_abs_rate.scores = ['absorption']
therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_abs_rate)
# Thermal Flux Utilization tallies
# fuel_therm_abs_rate = openmc.Tally(name='fuel therm. abs. rate')
# fuel_therm_abs_rate.scores = ['absorption']
# fuel_therm_abs_rate.filters = [openmc.EnergyFilter([0., 0.625]),
# openmc.CellFilter([fuel_cell])]
# tallies_file.append(fuel_therm_abs_rate)
# Fast Fission Factor tallies
therm_fiss_rate = openmc.Tally(name='therm. fiss. rate')
therm_fiss_rate.scores = ['nu-fission']
therm_fiss_rate.filters = [openmc.EnergyFilter([0., 0.625])]
tallies_file.append(therm_fiss_rate)
tallies_file.export_to_xml()
#############################
### PLOTTING ###
#############################
zs = np.linspace(0, 1, 2)
plots = []
for z in zs:
p = openmc.Plot()
p.filename = 'pinplot' + str(z)
p.width = (1.4 * pitch, 1.4 * pitch)
p.pixels = (2000, 2000)
p.color_by = 'material'
p.origin = [0, 0, z]
# p.color_by = 'cell'
# p.colors = {homogeneous_fuel: 'yellow', naoh: 'grey', graphite: 'black'}
p.colors = {fuel: 'yellow', naoh: 'grey', graphite: 'black'}
plots.append(copy.deepcopy(p))
plots = openmc.Plots(plots)
plots.export_to_xml()
# openmc.plot_geometry(output = False)
openmc.plot_geometry()
# pngstring = 'pinplot{}.png'.format(str(pitch))
# subprocess.call(['convert','pinplot.ppm',pngstring])
# subprocess.call(['mv',pngstring,'figures/'+pngstring])
#############################
### EXECUTION ###
#############################
# openmc.run(output=False)
openmc.run()
sp = openmc.StatePoint('statepoint.{}.h5'.format(settings.batches))
# Collect all the tallies
fiss_rate = sp.get_tally(name='fiss. rate')
fiss_rate_df = fiss_rate.get_pandas_dataframe()
abs_rate = sp.get_tally(name='abs. rate')
abs_rate_df = abs_rate.get_pandas_dataframe()
therm_abs_rate = sp.get_tally(name='therm. abs. rate')
therm_abs_rate_df = therm_abs_rate.get_pandas_dataframe()
fuel_therm_abs_rate = sp.get_tally(name='fuel therm. abs. rate')
fuel_therm_abs_rate_df = fuel_therm_abs_rate.get_pandas_dataframe()
therm_fiss_rate = sp.get_tally(name='therm. fiss. rate')
therm_fiss_rate_df = therm_fiss_rate.get_pandas_dataframe()
# Compute k-infinity
kinf = fiss_rate / abs_rate
kinf_df = kinf.get_pandas_dataframe()
# Compute resonance escape probability
res_esc = (therm_abs_rate) / (abs_rate)
res_esc_df = res_esc.get_pandas_dataframe()
# Compute fast fission factor
fast_fiss = fiss_rate / therm_fiss_rate
fast_fiss_df = fast_fiss.get_pandas_dataframe()
# Compute thermal flux utilization
therm_util = fuel_therm_abs_rate / therm_abs_rate
therm_util_df = therm_util.get_pandas_dataframe()
# Compute neutrons produced per absorption
eta = therm_fiss_rate / fuel_therm_abs_rate
eta_df = eta.get_pandas_dataframe()
columns = [
'pitch', 'enrichment', 'kinf mean', 'kinf sd', 'res_esc mean',
'res_esc sd', 'fast_fiss mean', 'fast_fiss sd', 'therm_util mean',
'therm_util sd', 'eta mean', 'eta sd'
]
data = [[
pitch, enrichment, kinf_df['mean'][0], kinf_df['std. dev.'][0],
res_esc_df['mean'][0], res_esc_df['std. dev.'][0],
fast_fiss_df['mean'][0], fast_fiss_df['std. dev.'][0],
therm_util_df['mean'][0], therm_util_df['std. dev.'][0],
eta_df['mean'][0], eta_df['std. dev.'][0]
]]
all_tallies = pd.DataFrame(data, columns=columns)
return all_tallies