A Sodium Fast Reactor geometry¶
This notebook can be used as a template for modeling Sodium fast reactors.
SOURCES: A. Facchini, V. Giusti, R. Ciolini, K. Tucek, D. Thomas, E. D'Agata, "Detailed Neutornics Study of the Power Evolution for the European Sodium Fast Reactor During a Positive Insertion of Reactivity," Nuc. Eng. Design 313 1-9 (2017) A. Ponomarev, A. Bednarova, K. Mikityuk, "New Sodium Fast Reactor Neutronics Benchmark," PHYSOR 2018
In [1]:
Copied!
%matplotlib inline
import openmc
%matplotlib inline
import openmc
In [2]:
Copied!
# Materials definitions
u235 = openmc.Material(name='U235')
u235.add_nuclide('U235', 1.0)
u235.set_density('g/cm3', 10.0)
u238 = openmc.Material(name='U238')
u238.add_nuclide('U238', 1.0)
u238.set_density('g/cm3', 10.0)
pu238 = openmc.Material(name='Pu238')
pu238.add_nuclide('Pu238', 1.0)
pu238.set_density('g/cm3', 10.0)
pu239 = openmc.Material(name='U235')
pu239.add_nuclide('Pu239', 1.0)
pu239.set_density('g/cm3', 10.0)
pu240 = openmc.Material(name='Pu240')
pu240.add_nuclide('Pu240', 1.0)
pu240.set_density('g/cm3', 10.0)
pu241 = openmc.Material(name='Pu241')
pu241.add_nuclide('Pu241', 1.0)
pu241.set_density('g/cm3', 10.0)
pu242 = openmc.Material(name='Pu242')
pu242.add_nuclide('Pu242', 1.0)
pu242.set_density('g/cm3', 10.0)
am241 = openmc.Material(name='Am241')
am241.add_nuclide('Am241', 1.0)
am241.set_density('g/cm3', 10.0)
o16 = openmc.Material(name='O16')
o16.add_nuclide('O16', 1.0)
o16.set_density('g/cm3', 10.0)
sodium = openmc.Material(name='Na')
sodium.add_nuclide('Na23', 1.0)
sodium.set_density('g/cm3', 0.96)
cu63 = openmc.Material(name='Cu63')
cu63.set_density('g/cm3', 10.0)
cu63.add_nuclide('Cu63', 1.0)
Al2O3 = openmc.Material(name='Al2O3')
Al2O3.set_density('g/cm3', 10.0)
Al2O3.add_element('O', 3.0)
Al2O3.add_element('Al', 2.0)
# Materials definitions
u235 = openmc.Material(name='U235')
u235.add_nuclide('U235', 1.0)
u235.set_density('g/cm3', 10.0)
u238 = openmc.Material(name='U238')
u238.add_nuclide('U238', 1.0)
u238.set_density('g/cm3', 10.0)
pu238 = openmc.Material(name='Pu238')
pu238.add_nuclide('Pu238', 1.0)
pu238.set_density('g/cm3', 10.0)
pu239 = openmc.Material(name='U235')
pu239.add_nuclide('Pu239', 1.0)
pu239.set_density('g/cm3', 10.0)
pu240 = openmc.Material(name='Pu240')
pu240.add_nuclide('Pu240', 1.0)
pu240.set_density('g/cm3', 10.0)
pu241 = openmc.Material(name='Pu241')
pu241.add_nuclide('Pu241', 1.0)
pu241.set_density('g/cm3', 10.0)
pu242 = openmc.Material(name='Pu242')
pu242.add_nuclide('Pu242', 1.0)
pu242.set_density('g/cm3', 10.0)
am241 = openmc.Material(name='Am241')
am241.add_nuclide('Am241', 1.0)
am241.set_density('g/cm3', 10.0)
o16 = openmc.Material(name='O16')
o16.add_nuclide('O16', 1.0)
o16.set_density('g/cm3', 10.0)
sodium = openmc.Material(name='Na')
sodium.add_nuclide('Na23', 1.0)
sodium.set_density('g/cm3', 0.96)
cu63 = openmc.Material(name='Cu63')
cu63.set_density('g/cm3', 10.0)
cu63.add_nuclide('Cu63', 1.0)
Al2O3 = openmc.Material(name='Al2O3')
Al2O3.set_density('g/cm3', 10.0)
Al2O3.add_element('O', 3.0)
Al2O3.add_element('Al', 2.0)
In [3]:
Copied!
# Material mixtures
inner = openmc.Material.mix_materials(
[u235, u238, pu238, pu239, pu240, pu241, pu242, am241, o16],
[0.0019, 0.7509, 0.0046, 0.0612, 0.0383, 0.0106, 0.0134, 0.001, 0.1181],
'wo')
outer = openmc.Material.mix_materials(
[u235, u238, pu238, pu239, pu240, pu241, pu242, am241, o16],
[0.0018, 0.73, 0.0053, 0.0711, 0.0445, 0.0124, 0.0156, 0.0017, 0.1176],
'wo')
clad = openmc.Material.mix_materials(
[cu63,Al2O3], [0.997,0.003], 'wo')
# Material mixtures
inner = openmc.Material.mix_materials(
[u235, u238, pu238, pu239, pu240, pu241, pu242, am241, o16],
[0.0019, 0.7509, 0.0046, 0.0612, 0.0383, 0.0106, 0.0134, 0.001, 0.1181],
'wo')
outer = openmc.Material.mix_materials(
[u235, u238, pu238, pu239, pu240, pu241, pu242, am241, o16],
[0.0018, 0.73, 0.0053, 0.0711, 0.0445, 0.0124, 0.0156, 0.0017, 0.1176],
'wo')
clad = openmc.Material.mix_materials(
[cu63,Al2O3], [0.997,0.003], 'wo')
In [5]:
Copied!
# Instantiate a Materials collection and export to xml
materials_file = openmc.Materials([inner, outer, sodium, clad])
materials_file.export_to_xml()
# Instantiate a Materials collection and export to xml
materials_file = openmc.Materials([inner, outer, sodium, clad])
materials_file.export_to_xml()
In [6]:
Copied!
# Geometry definitions
fuel_or = openmc.ZCylinder(surface_id=1, r=0.943/2)
clad_ir = openmc.ZCylinder(surface_id=2, r=0.973/2)
clad_or = openmc.ZCylinder(surface_id=3, r=1.073/2)
top = openmc.ZPlane(surface_id=4, z0=+50, boundary_type='vacuum')
bottom = openmc.ZPlane(surface_id=5, z0=-50, boundary_type='vacuum')
fuel_region = -fuel_or & -top & +bottom
gap_region = +fuel_or & -clad_ir & -top & +bottom
clad_region = +clad_ir & -clad_or & -top & +bottom
moderator_region = +clad_or & -top & +bottom
gap_cell = openmc.Cell(cell_id=1, fill=inner, region=gap_region)
clad_cell = openmc.Cell(cell_id=2, fill=clad, region=clad_region)
sodium_cell = openmc.Cell(cell_id=3, fill=sodium, region=moderator_region)
# Geometry definitions
fuel_or = openmc.ZCylinder(surface_id=1, r=0.943/2)
clad_ir = openmc.ZCylinder(surface_id=2, r=0.973/2)
clad_or = openmc.ZCylinder(surface_id=3, r=1.073/2)
top = openmc.ZPlane(surface_id=4, z0=+50, boundary_type='vacuum')
bottom = openmc.ZPlane(surface_id=5, z0=-50, boundary_type='vacuum')
fuel_region = -fuel_or & -top & +bottom
gap_region = +fuel_or & -clad_ir & -top & +bottom
clad_region = +clad_ir & -clad_or & -top & +bottom
moderator_region = +clad_or & -top & +bottom
gap_cell = openmc.Cell(cell_id=1, fill=inner, region=gap_region)
clad_cell = openmc.Cell(cell_id=2, fill=clad, region=clad_region)
sodium_cell = openmc.Cell(cell_id=3, fill=sodium, region=moderator_region)
In [7]:
Copied!
inner_fuel_cell = openmc.Cell(cell_id=4, fill=inner, region=fuel_region)
inner_u = openmc.Universe(universe_id=1, cells=(inner_fuel_cell, gap_cell, clad_cell, sodium_cell))
inner_fuel_cell = openmc.Cell(cell_id=4, fill=inner, region=fuel_region)
inner_u = openmc.Universe(universe_id=1, cells=(inner_fuel_cell, gap_cell, clad_cell, sodium_cell))
In [9]:
Copied!
outer_fuel_cell = openmc.Cell(cell_id=5, fill=outer, region=fuel_region)
outer_u = openmc.Universe(universe_id=2, cells=(outer_fuel_cell, gap_cell, clad_cell, sodium_cell))
outer_fuel_cell = openmc.Cell(cell_id=5, fill=outer, region=fuel_region)
outer_u = openmc.Universe(universe_id=2, cells=(outer_fuel_cell, gap_cell, clad_cell, sodium_cell))
In [ ]:
Copied!
# Creating filling for emtpy space in the core
sodium_mod_cell = openmc.Cell(cell_id=6, fill=sodium)
sodium_mod_u = openmc.Universe(universe_id=3, cells=(sodium_mod_cell,))
# Creating filling for emtpy space in the core
sodium_mod_cell = openmc.Cell(cell_id=6, fill=sodium)
sodium_mod_u = openmc.Universe(universe_id=3, cells=(sodium_mod_cell,))
In [10]:
Copied!
# Define a lattice for inner assemblies
in_lat = openmc.HexLattice(lattice_id=1, name='inner assembly')
in_lat.center = (0., 0.)
in_lat.pitch = (21.08/17,)
in_lat.orientation = 'x'
in_lat.outer = sodium_mod_u
# Create rings of fuel universes that will fill the lattice
inone = [inner_u]*48
intwo = [inner_u]*42
inthree = [inner_u]*36
infour = [inner_u]*30
infive = [inner_u]*24
insix = [inner_u]*18
inseven = [inner_u]*12
ineight = [inner_u]*6
innine = [inner_u]*1
in_lat.universes = [inone,intwo,inthree,infour,infive,insix,inseven,ineight,innine]
# Create the prism that will contain the lattice
outer_in_surface = openmc.model.hexagonal_prism(edge_length=12.1705, orientation='x')
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
main_in_assembly = openmc.Cell(cell_id=7, fill=in_lat, region=outer_in_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_in_assembly = openmc.Cell(cell_id=8, fill=sodium, region=~outer_in_surface & -top & +bottom)
# Create a universe that contains both
main_in_u = openmc.Universe(universe_id=4, cells=[main_in_assembly, out_in_assembly])
# Define a lattice for inner assemblies
in_lat = openmc.HexLattice(lattice_id=1, name='inner assembly')
in_lat.center = (0., 0.)
in_lat.pitch = (21.08/17,)
in_lat.orientation = 'x'
in_lat.outer = sodium_mod_u
# Create rings of fuel universes that will fill the lattice
inone = [inner_u]*48
intwo = [inner_u]*42
inthree = [inner_u]*36
infour = [inner_u]*30
infive = [inner_u]*24
insix = [inner_u]*18
inseven = [inner_u]*12
ineight = [inner_u]*6
innine = [inner_u]*1
in_lat.universes = [inone,intwo,inthree,infour,infive,insix,inseven,ineight,innine]
# Create the prism that will contain the lattice
outer_in_surface = openmc.model.hexagonal_prism(edge_length=12.1705, orientation='x')
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
main_in_assembly = openmc.Cell(cell_id=7, fill=in_lat, region=outer_in_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_in_assembly = openmc.Cell(cell_id=8, fill=sodium, region=~outer_in_surface & -top & +bottom)
# Create a universe that contains both
main_in_u = openmc.Universe(universe_id=4, cells=[main_in_assembly, out_in_assembly])
/Users/miriamrathbun/codes/openmc-mak/openmc/mixin.py:68: IDWarning: Another Lattice instance already exists with id=1. warn(msg, IDWarning)
In [11]:
Copied!
main_in_u.plot(origin = (0,0,0), pixels=(500, 500), width = (30.,30.), color_by = 'material')
main_in_u.plot(origin = (0,0,0), pixels=(500, 500), width = (30.,30.), color_by = 'material')
Out[11]:
<matplotlib.image.AxesImage at 0x1273ff790>
In [12]:
Copied!
# Define a lattice for outer assemblies
out_lat = openmc.HexLattice(lattice_id=2, name='outer assembly')
out_lat.center = (0., 0.)
out_lat.pitch = (21.08/17,)
out_lat.orientation = 'x'
out_lat.outer = sodium_mod_u
# Create rings of fuel universes that will fill the lattice
outone = [outer_u]*48
outtwo = [outer_u]*42
outthree = [outer_u]*36
outfour = [outer_u]*30
outfive = [outer_u]*24
outsix = [outer_u]*18
outseven = [outer_u]*12
outeight = [outer_u]*6
outnine = [outer_u]*1
out_lat.universes = [outone,outtwo,outthree,outfour,outfive,outsix,outseven,outeight,outnine]
# Create the prism that will contain the lattice
outer_out_surface = openmc.model.hexagonal_prism(edge_length=12.1705)
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
main_out_assembly = openmc.Cell(cell_id=9, fill=out_lat, region=outer_out_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_out_assembly = openmc.Cell(cell_id=10, fill=sodium, region=~outer_out_surface & -top & +bottom)
# Create a universe that contains both
main_out_u = openmc.Universe(universe_id=5, cells=[main_out_assembly, out_out_assembly])
# Define a lattice for outer assemblies
out_lat = openmc.HexLattice(lattice_id=2, name='outer assembly')
out_lat.center = (0., 0.)
out_lat.pitch = (21.08/17,)
out_lat.orientation = 'x'
out_lat.outer = sodium_mod_u
# Create rings of fuel universes that will fill the lattice
outone = [outer_u]*48
outtwo = [outer_u]*42
outthree = [outer_u]*36
outfour = [outer_u]*30
outfive = [outer_u]*24
outsix = [outer_u]*18
outseven = [outer_u]*12
outeight = [outer_u]*6
outnine = [outer_u]*1
out_lat.universes = [outone,outtwo,outthree,outfour,outfive,outsix,outseven,outeight,outnine]
# Create the prism that will contain the lattice
outer_out_surface = openmc.model.hexagonal_prism(edge_length=12.1705)
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
main_out_assembly = openmc.Cell(cell_id=9, fill=out_lat, region=outer_out_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_out_assembly = openmc.Cell(cell_id=10, fill=sodium, region=~outer_out_surface & -top & +bottom)
# Create a universe that contains both
main_out_u = openmc.Universe(universe_id=5, cells=[main_out_assembly, out_out_assembly])
/Users/miriamrathbun/codes/openmc-mak/openmc/mixin.py:68: IDWarning: Another Lattice instance already exists with id=2. warn(msg, IDWarning)
In [13]:
Copied!
# Create a hexagonal water cell
reflector_assembly = openmc.model.hexagonal_prism(edge_length=12.1705, orientation='x')
ref_cell = openmc.Cell(cell_id=11, fill=sodium, region=reflector_assembly & -top & +bottom)
out_ref_cell = openmc.Cell(cell_id=12, fill=sodium, region=~reflector_assembly & -top & +bottom)
ref_u = openmc.Universe(universe_id=6, cells=[ref_cell, out_ref_cell])
# Create a hexagonal water cell
reflector_assembly = openmc.model.hexagonal_prism(edge_length=12.1705, orientation='x')
ref_cell = openmc.Cell(cell_id=11, fill=sodium, region=reflector_assembly & -top & +bottom)
out_ref_cell = openmc.Cell(cell_id=12, fill=sodium, region=~reflector_assembly & -top & +bottom)
ref_u = openmc.Universe(universe_id=6, cells=[ref_cell, out_ref_cell])
We have 3 types of assemblies created. We can now make the entire reactor core by creating a lattice that is filled with the assemblies.
In [14]:
Copied!
# Define the core lattice
core_lat = openmc.HexLattice(lattice_id=3, name='core')
core_lat.center = (0., 0.)
core_lat.pitch = (21.08,)
core_lat.outer = sodium_mod_u
# Define the core lattice
core_lat = openmc.HexLattice(lattice_id=3, name='core')
core_lat.center = (0., 0.)
core_lat.pitch = (21.08,)
core_lat.outer = sodium_mod_u
/Users/miriamrathbun/codes/openmc-mak/openmc/mixin.py:68: IDWarning: Another Lattice instance already exists with id=3. warn(msg, IDWarning)
In [15]:
Copied!
# Create rings of fuel universes that will fill the lattice
ref_one = [ref_u] * 96
ref_two = [ref_u] * 90
ref_three = [ref_u] * 84
ref_four = ([ref_u] * 5 + [main_out_u] * 4 + [ref_u] * 4) * 6
ref_five = ([ref_u] + [main_out_u] * 11) * 6
out_one = [main_out_u]*66
out_two = [main_out_u]*60
out_three = ([main_out_u]*2 + [main_in_u]*6 + [main_out_u] * 1)*6
in_one = [main_in_u]*48
in_two = [main_in_u]*42
in_three = [main_in_u]*36
in_four = [main_in_u]*30
in_five = [main_in_u]*24
in_six = [main_in_u]*18
in_seven = [main_in_u]*12
in_eight = [main_in_u]*6
in_nine = [main_in_u]*1
core_lat.universes = [ref_one,ref_two,ref_three,ref_four,ref_five,out_one,out_two,out_three,in_one,in_two,in_three,in_four,in_five,in_six,in_seven,in_eight,in_nine]
# Create rings of fuel universes that will fill the lattice
ref_one = [ref_u] * 96
ref_two = [ref_u] * 90
ref_three = [ref_u] * 84
ref_four = ([ref_u] * 5 + [main_out_u] * 4 + [ref_u] * 4) * 6
ref_five = ([ref_u] + [main_out_u] * 11) * 6
out_one = [main_out_u]*66
out_two = [main_out_u]*60
out_three = ([main_out_u]*2 + [main_in_u]*6 + [main_out_u] * 1)*6
in_one = [main_in_u]*48
in_two = [main_in_u]*42
in_three = [main_in_u]*36
in_four = [main_in_u]*30
in_five = [main_in_u]*24
in_six = [main_in_u]*18
in_seven = [main_in_u]*12
in_eight = [main_in_u]*6
in_nine = [main_in_u]*1
core_lat.universes = [ref_one,ref_two,ref_three,ref_four,ref_five,out_one,out_two,out_three,in_one,in_two,in_three,in_four,in_five,in_six,in_seven,in_eight,in_nine]
In [16]:
Copied!
# Create the prism that will contain the lattice
outer_core_surface = openmc.model.hexagonal_prism(edge_length=347.82, boundary_type='vacuum')
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
core = openmc.Cell(cell_id=13, fill=core_lat, region=outer_core_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_core = openmc.Cell(cell_id=14, fill=outer, region=~outer_core_surface & -top & +bottom)
# Create a universe that contains both
main_u = openmc.Universe(universe_id=7, cells=[core, out_core])
# Create the prism that will contain the lattice
outer_core_surface = openmc.model.hexagonal_prism(edge_length=347.82, boundary_type='vacuum')
# Fill a cell with the lattice. This cell is filled with the lattice and contained within the prism.
core = openmc.Cell(cell_id=13, fill=core_lat, region=outer_core_surface & -top & +bottom)
# Fill a cell with a material that will surround the lattice
out_core = openmc.Cell(cell_id=14, fill=outer, region=~outer_core_surface & -top & +bottom)
# Create a universe that contains both
main_u = openmc.Universe(universe_id=7, cells=[core, out_core])
In [20]:
Copied!
main_u.plot(origin = (0,0,0), pixels=(1000, 1000), width = (660.,660.), color_by = 'material')
main_u.plot(origin = (0,0,0), pixels=(1000, 1000), width = (660.,660.), color_by = 'material')
Out[20]:
<matplotlib.image.AxesImage at 0x125a17e50>
We now have an entire reactor core defined! We can export the geometry and run the code.
In [17]:
Copied!
geom = openmc.Geometry(main_u)
geom.export_to_xml()
geom = openmc.Geometry(main_u)
geom.export_to_xml()
In [18]:
Copied!
# OpenMC simulation parameters
lower_left = [-300, -300, -50]
upper_right = [300, 300, 50]
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
src = openmc.Source(space=uniform_dist)
settings = openmc.Settings()
settings.source = src
settings.batches = 100
settings.inactive = 10
settings.particles = 1000
settings.export_to_xml()
# OpenMC simulation parameters
lower_left = [-300, -300, -50]
upper_right = [300, 300, 50]
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
src = openmc.Source(space=uniform_dist)
settings = openmc.Settings()
settings.source = src
settings.batches = 100
settings.inactive = 10
settings.particles = 1000
settings.export_to_xml()
In [23]:
Copied!
openmc.run()
openmc.run()
%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%
%%%%%%%%%%%%%%%%%%%%%%%%
############### %%%%%%%%%%%%%%%%%%%%%%%%
################## %%%%%%%%%%%%%%%%%%%%%%%
################### %%%%%%%%%%%%%%%%%%%%%%%
#################### %%%%%%%%%%%%%%%%%%%%%%
##################### %%%%%%%%%%%%%%%%%%%%%
###################### %%%%%%%%%%%%%%%%%%%%
####################### %%%%%%%%%%%%%%%%%%
####################### %%%%%%%%%%%%%%%%%
###################### %%%%%%%%%%%%%%%%%
#################### %%%%%%%%%%%%%%%%%
################# %%%%%%%%%%%%%%%%%
############### %%%%%%%%%%%%%%%%
############ %%%%%%%%%%%%%%%
######## %%%%%%%%%%%%%%
%%%%%%%%%%%
| The OpenMC Monte Carlo Code
Copyright | 2011-2020 MIT and OpenMC contributors
License | https://docs.openmc.org/en/latest/license.html
Version | 0.12.1-dev
Git SHA1 | 0228ddba1d6399e654a295e51539590f076b2a3c
Date/Time | 2021-02-22 22:22:08
OpenMP Threads | 8
Reading settings XML file...
Reading cross sections XML file...
Reading materials XML file...
Reading geometry XML file...
Reading Na23 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Na23.h5
Reading U235 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/U235.h5
Reading U238 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/U238.h5
Reading Pu239 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Pu239.h5
Reading Pu240 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Pu240.h5
Reading Pu241 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Pu241.h5
Reading Pu242 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Pu242.h5
Reading Am241 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Am241.h5
Reading O16 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/O16.h5
Reading Cu63 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Cu63.h5
Reading O17 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/O17.h5
Reading Al27 from
/Users/miriamrathbun/codes/openmc/nucleardata/endfb71_hdf5/Al27.h5
Minimum neutron data temperature: 294.0 K
Maximum neutron data temperature: 294.0 K
Preparing distributed cell instances...
Writing summary.h5 file...
Maximum neutron transport energy: 20000000.0 eV for Na23
Initializing source particles...
====================> K EIGENVALUE SIMULATION <====================
Bat./Gen. k Average k
========= ======== ====================
1/1 1.02481
2/1 1.12641
3/1 1.19114
4/1 1.14944
5/1 1.24417
6/1 1.18112
7/1 1.22681
8/1 1.25324
9/1 1.26160
10/1 1.24391
11/1 1.26299
12/1 1.22630 1.24465 +/- 0.01834
13/1 1.22617 1.23849 +/- 0.01225
14/1 1.17487 1.22258 +/- 0.01811
15/1 1.23711 1.22549 +/- 0.01433
16/1 1.25744 1.23081 +/- 0.01285
17/1 1.19600 1.22584 +/- 0.01195
18/1 1.22848 1.22617 +/- 0.01035
19/1 1.24147 1.22787 +/- 0.00929
20/1 1.23635 1.22872 +/- 0.00835
21/1 1.22312 1.22821 +/- 0.00757
22/1 1.22163 1.22766 +/- 0.00693
23/1 1.21767 1.22689 +/- 0.00642
24/1 1.16644 1.22257 +/- 0.00735
25/1 1.19885 1.22099 +/- 0.00702
26/1 1.21883 1.22086 +/- 0.00657
27/1 1.24819 1.22246 +/- 0.00638
28/1 1.26476 1.22481 +/- 0.00645
29/1 1.21935 1.22453 +/- 0.00611
30/1 1.23808 1.22520 +/- 0.00584
31/1 1.16740 1.22245 +/- 0.00620
32/1 1.21038 1.22190 +/- 0.00593
33/1 1.20477 1.22116 +/- 0.00572
34/1 1.26182 1.22285 +/- 0.00573
35/1 1.22385 1.22289 +/- 0.00550
36/1 1.30378 1.22600 +/- 0.00613
37/1 1.26827 1.22757 +/- 0.00610
38/1 1.27147 1.22914 +/- 0.00609
39/1 1.23282 1.22926 +/- 0.00587
40/1 1.25327 1.23006 +/- 0.00573
41/1 1.19087 1.22880 +/- 0.00569
42/1 1.28910 1.23068 +/- 0.00582
43/1 1.23440 1.23080 +/- 0.00564
44/1 1.22528 1.23063 +/- 0.00547
45/1 1.25779 1.23141 +/- 0.00537
46/1 1.26413 1.23232 +/- 0.00530
47/1 1.27685 1.23352 +/- 0.00529
48/1 1.27599 1.23464 +/- 0.00527
49/1 1.28307 1.23588 +/- 0.00528
50/1 1.26215 1.23654 +/- 0.00519
51/1 1.22880 1.23635 +/- 0.00507
52/1 1.22389 1.23605 +/- 0.00495
53/1 1.23361 1.23600 +/- 0.00484
54/1 1.22966 1.23585 +/- 0.00473
55/1 1.22266 1.23556 +/- 0.00463
56/1 1.25986 1.23609 +/- 0.00456
57/1 1.31764 1.23782 +/- 0.00479
58/1 1.29656 1.23905 +/- 0.00484
59/1 1.25418 1.23936 +/- 0.00475
60/1 1.21933 1.23896 +/- 0.00467
61/1 1.23886 1.23895 +/- 0.00458
62/1 1.22153 1.23862 +/- 0.00451
63/1 1.25440 1.23892 +/- 0.00443
64/1 1.23758 1.23889 +/- 0.00435
65/1 1.29599 1.23993 +/- 0.00439
66/1 1.25912 1.24027 +/- 0.00433
67/1 1.25699 1.24057 +/- 0.00426
68/1 1.21444 1.24011 +/- 0.00421
69/1 1.25080 1.24030 +/- 0.00414
70/1 1.30284 1.24134 +/- 0.00420
71/1 1.26208 1.24168 +/- 0.00415
72/1 1.25869 1.24195 +/- 0.00409
73/1 1.24739 1.24204 +/- 0.00403
74/1 1.25263 1.24220 +/- 0.00397
75/1 1.22410 1.24193 +/- 0.00391
76/1 1.29268 1.24269 +/- 0.00393
77/1 1.22569 1.24244 +/- 0.00388
78/1 1.28431 1.24306 +/- 0.00387
79/1 1.22645 1.24282 +/- 0.00382
80/1 1.24248 1.24281 +/- 0.00377
81/1 1.22740 1.24259 +/- 0.00372
82/1 1.29676 1.24335 +/- 0.00374
83/1 1.26406 1.24363 +/- 0.00370
84/1 1.18714 1.24287 +/- 0.00373
85/1 1.24634 1.24291 +/- 0.00368
86/1 1.25176 1.24303 +/- 0.00364
87/1 1.30837 1.24388 +/- 0.00369
88/1 1.24285 1.24386 +/- 0.00364
89/1 1.22941 1.24368 +/- 0.00360
90/1 1.18066 1.24289 +/- 0.00364
91/1 1.28586 1.24342 +/- 0.00363
92/1 1.26556 1.24369 +/- 0.00360
93/1 1.22569 1.24348 +/- 0.00356
94/1 1.20319 1.24300 +/- 0.00355
95/1 1.27998 1.24343 +/- 0.00354
96/1 1.20130 1.24294 +/- 0.00353
97/1 1.24200 1.24293 +/- 0.00349
98/1 1.27352 1.24328 +/- 0.00347
99/1 1.30545 1.24398 +/- 0.00350
100/1 1.31981 1.24482 +/- 0.00356
Creating state point statepoint.100.h5...
=======================> TIMING STATISTICS <=======================
Total time for initialization = 2.3018e+00 seconds
Reading cross sections = 2.2651e+00 seconds
Total time in simulation = 1.0711e+00 seconds
Time in transport only = 1.0512e+00 seconds
Time in inactive batches = 9.2840e-02 seconds
Time in active batches = 9.7826e-01 seconds
Time synchronizing fission bank = 4.0280e-03 seconds
Sampling source sites = 3.3290e-03 seconds
SEND/RECV source sites = 5.3600e-04 seconds
Time accumulating tallies = 2.5000e-05 seconds
Time writing statepoints = 8.3760e-03 seconds
Total time for finalization = 7.0000e-06 seconds
Total time elapsed = 3.3798e+00 seconds
Calculation Rate (inactive) = 107712.0 particles/second
Calculation Rate (active) = 92000.4 particles/second
============================> RESULTS <============================
k-effective (Collision) = 1.24514 +/- 0.00362
k-effective (Track-length) = 1.24482 +/- 0.00356
k-effective (Absorption) = 1.24692 +/- 0.00415
Combined k-effective = 1.24588 +/- 0.00254
Leakage Fraction = 0.11453 +/- 0.00147