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models.py
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models.py
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import numpy as np
import openmc
import openmc.mgxs
import openmc.model
import pandas as pd
import plotly
import plotly.graph_objs as go
from plotly.offline import init_notebook_mode
init_notebook_mode(connected=True)
def slab_mg(reps=None, as_macro=True):
"""Create a one-group, 1D slab model.
Parameters
----------
reps : list, optional
List of angular representations. Each item corresponds to materials and
dictates the angular representation of the multi-group cross
sections---isotropic ('iso') or angle-dependent ('ang'), and if Legendre
scattering or tabular scattering ('mu') is used. Thus, items can be
'ang', 'ang_mu', 'iso', or 'iso_mu'.
as_macro : bool, optional
Whether :class:`openmc.Macroscopic` is used
Returns
-------
model : openmc.model.Model
One-group, 1D slab model
"""
model = openmc.model.Model()
# Define materials needed for 1D/1G slab problem
mat_names = ['uo2', 'clad', 'lwtr']
mgxs_reps = ['ang', 'ang_mu', 'iso', 'iso_mu']
if reps is None:
reps = mgxs_reps
xs = []
i = 0
for mat in mat_names:
for rep in reps:
i += 1
name = mat + '_' + rep
xs.append(name)
if as_macro:
m = openmc.Material(name=str(i))
m.set_density('macro', 1.)
m.add_macroscopic(name)
else:
m = openmc.Material(name=str(i))
m.set_density('atom/b-cm', 1.)
m.add_nuclide(name, 1.0, 'ao')
model.materials.append(m)
# Define the materials file
model.xs_data = xs
model.materials.cross_sections = "../1d_mgxs.h5"
# Define surfaces.
# Assembly/Problem Boundary
left = openmc.XPlane(x0=0.0, boundary_type='reflective')
right = openmc.XPlane(x0=10.0, boundary_type='reflective')
bottom = openmc.YPlane(y0=0.0, boundary_type='reflective')
top = openmc.YPlane(y0=10.0, boundary_type='reflective')
# for each material add a plane
planes = [openmc.ZPlane(z0=0.0, boundary_type='reflective')]
dz = round(5. / float(len(model.materials)), 4)
for i in range(len(model.materials) - 1):
planes.append(openmc.ZPlane(z0=dz * float(i + 1)))
planes.append(openmc.ZPlane(z0=5.0, boundary_type='reflective'))
# Define cells for each material
model.geometry.root_universe = openmc.Universe(name='root universe')
xy = +left & -right & +bottom & -top
for i, mat in enumerate(model.materials):
c = openmc.Cell(fill=mat, region=xy & +planes[i] & -planes[i + 1])
model.geometry.root_universe.add_cell(c)
model.settings.batches = 20
model.settings.inactive = 5
model.settings.particles = 200
model.settings.source = openmc.Source(space=openmc.stats.Box(
[0.0, 0.0, 0.0], [10.0, 10.0, 5.]))
model.settings.energy_mode = "multi-group"
plot = openmc.Plot()
plot.filename = 'mat'
plot.origin = (5.0, 5.0, 2.5)
plot.width = (2.5, 2.5)
plot.basis = 'xz'
plot.pixels = (3000, 3000)
plot.color_by = 'material'
model.plots.append(plot)
return model
def pwr_pin_cell():
"""Create a PWR pin-cell model.
This model is a single fuel pin with 2.4 w/o enriched UO2 corresponding to a
beginning-of-cycle condition and borated water. The specifications are from
the `BEAVRS <http://crpg.mit.edu/research/beavrs>`_ benchmark. Note that the
number of particles/batches is initially set very low for testing purposes.
Returns
-------
model : openmc.model.Model
A PWR pin-cell model
"""
model = openmc.model.Model()
# Define materials.
fuel = openmc.Material(name='UO2 (2.4%)')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Zircaloy')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, hot_water)
# Instantiate ZCylinder surfaces
pitch = 1.26
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
left = openmc.XPlane(x0=-pitch / 2, name='left', boundary_type='reflective')
right = openmc.XPlane(x0=pitch / 2, name='right', boundary_type='reflective')
bottom = openmc.YPlane(y0=-pitch / 2, name='bottom', boundary_type='reflective')
top = openmc.YPlane(y0=pitch / 2, name='top', boundary_type='reflective')
# Instantiate Cells
fuel_pin = openmc.Cell(name='Fuel', fill=fuel)
cladding = openmc.Cell(name='Cladding', fill=clad)
water = openmc.Cell(name='Water', fill=hot_water)
# Use surface half-spaces to define regions
fuel_pin.region = -fuel_or
cladding.region = +fuel_or & -clad_or
water.region = +clad_or & +left & -right & +bottom & -top
# Create root universe
model.geometry.root_universe = openmc.Universe(0, name='root universe')
model.geometry.root_universe.add_cells([fuel_pin, cladding, water])
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-pitch / 2, -pitch / 2, -1], [pitch / 2, pitch / 2, 1], only_fissionable=True))
plot = openmc.Plot.from_geometry(model.geometry)
plot.pixels = (300, 300)
plot.color_by = 'material'
model.plots.append(plot)
return model
def pwr_assembly():
"""Create a PWR assembly model.
This model is a reflected 17x17 fuel assembly from the the `BEAVRS
<http://crpg.mit.edu/research/beavrs>`_ benchmark. The fuel is 2.4 w/o
enriched UO2 corresponding to a beginning-of-cycle condition. Note that the
number of particles/batches is initially set very low for testing purposes.
Returns
-------
model : openmc.model.Model
A PWR assembly model
"""
model = openmc.model.Model()
############################################################################################################
# MATERIALS
# Define materials.
fuel = openmc.Material(name='Fuel')
fuel.set_density('g/cm3', 10.29769)
fuel.add_nuclide("U234", 4.4843e-6)
fuel.add_nuclide("U235", 5.5815e-4)
fuel.add_nuclide("U238", 2.2408e-2)
fuel.add_nuclide("O16", 4.5829e-2)
clad = openmc.Material(name='Cladding')
clad.set_density('g/cm3', 6.55)
clad.add_nuclide("Zr90", 2.1827e-2)
clad.add_nuclide("Zr91", 4.7600e-3)
clad.add_nuclide("Zr92", 7.2758e-3)
clad.add_nuclide("Zr94", 7.3734e-3)
clad.add_nuclide("Zr96", 1.1879e-3)
hot_water = openmc.Material(name='Hot borated water')
hot_water.set_density('g/cm3', 0.740582)
hot_water.add_nuclide("H1", 4.9457e-2)
hot_water.add_nuclide("O16", 2.4672e-2)
hot_water.add_nuclide("B10", 8.0042e-6)
hot_water.add_nuclide("B11", 3.2218e-5)
hot_water.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, hot_water)
############################################################################################################
# GEOMETRY
# Instantiate ZCylinder surfaces
fuel_or = openmc.ZCylinder(x0=0, y0=0, R=0.39218, name='Fuel OR')
clad_or = openmc.ZCylinder(x0=0, y0=0, R=0.45720, name='Clad OR')
# Create boundary planes to surround the geometry
pitch = 21.42
height = 200
min_x = openmc.XPlane(x0=-pitch / 2, boundary_type='reflective')
max_x = openmc.XPlane(x0=+pitch / 2, boundary_type='reflective')
min_y = openmc.YPlane(y0=-pitch / 2, boundary_type='reflective')
max_y = openmc.YPlane(y0=+pitch / 2, boundary_type='reflective')
min_z = openmc.ZPlane(z0=-height / 2, boundary_type='reflective')
max_z = openmc.ZPlane(z0=+height / 2, boundary_type='reflective')
# Create a control rod guide tube universe
guide_tube_universe = openmc.Universe(name='Guide Tube')
gt_inner_cell = openmc.Cell(name='guide tube inner water', fill=hot_water,
region=-fuel_or)
gt_clad_cell = openmc.Cell(name='guide tube clad', fill=clad,
region=+fuel_or & -clad_or)
gt_outer_cell = openmc.Cell(name='guide tube outer water', fill=hot_water,
region=+clad_or)
guide_tube_universe.add_cells([gt_inner_cell, gt_clad_cell, gt_outer_cell])
# Create fuel assembly Lattice
assembly = openmc.RectLattice(name='Fuel Assembly')
assembly.pitch = (pitch / 17, pitch / 17)
assembly.lower_left = (-pitch / 2, -pitch / 2)
# Create array indices for guide tube locations in lattice
template_x = np.array([5, 8, 11, 3, 13, 2, 5, 8, 11, 14, 2, 5, 8,
11, 14, 2, 5, 8, 11, 14, 3, 13, 5, 8, 11])
template_y = np.array([2, 2, 2, 3, 3, 5, 5, 5, 5, 5, 8, 8, 8, 8,
8, 11, 11, 11, 11, 11, 13, 13, 14, 14, 14])
# Create 17x17 array of universes
fuel_pin_universe = openmc.Universe(name='Fuel Pin')
fuel_cell = openmc.Cell(name='fuel', fill=fuel, region=-fuel_or)
clad_cell = openmc.Cell(name='clad', fill=clad, region=+fuel_or & -clad_or)
hot_water_cell = openmc.Cell(name='hot water', fill=hot_water, region=+clad_or)
fuel_pin_universe.add_cells([fuel_cell, clad_cell, hot_water_cell])
assembly.universes = np.tile(fuel_pin_universe, (17, 17))
assembly.universes[template_x, template_y] = guide_tube_universe
# Create root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = assembly
root_cell.region = +min_x & -max_x & \
+min_y & -max_y & \
+min_z & -max_z
# Create root Universe
model.geometry.root_universe = openmc.Universe(name='root universe')
model.geometry.root_universe.add_cell(root_cell)
############################################################################################################
# MESH
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
dim_x, dim_y, dim_z = [100, 100, 10]
mesh.dimension = [dim_x, dim_y, dim_z]
mesh.lower_left = [*assembly.lower_left, -height / 2]
mesh.width = [*assembly.pitch, height / dim_z]
# Create a mesh filter
mesh_filter = openmc.MeshFilter(mesh)
############################################################################################################
# CROSS-SECTION LIBRARY
energy_groups = openmc.mgxs.EnergyGroups()
energy_groups.group_edges = np.logspace(-3, 7.3, 21)
# Instantiate a 1-group EnergyGroups object
one_group = openmc.mgxs.EnergyGroups()
one_group.group_edges = np.array([energy_groups.group_edges[0], energy_groups.group_edges[-1]])
# Initialize an 20-energy-group and 6-delayed-group MGXS Library
mgxs_lib = openmc.mgxs.Library(model.geometry)
mgxs_lib.energy_groups = one_group
mgxs_lib.num_delayed_groups = 6
# Specify multi-group cross section types to compute
mgxs_lib.mgxs_types = ['total', 'transport', 'nu-scatter matrix', 'kappa-fission', 'inverse-velocity', 'chi-prompt',
'prompt-nu-fission', 'chi-delayed', 'delayed-nu-fission', 'beta']
# Specify a "mesh" domain type for the cross section tally filters
mgxs_lib.domain_type = 'mesh'
# Specify the mesh domain over which to compute multi-group cross sections
mgxs_lib.domains = [mesh]
# Construct all tallies needed for the multi-group cross section library
mgxs_lib.build_library()
############################################################################################################
# TALLIES
# Create a "tallies.xml" file for the MGXS Library
model.tallies = openmc.Tallies()
mgxs_lib.add_to_tallies_file(model.tallies, merge=True)
# Instantiate a flux tally; Other valid options: 'current', 'fission', etc
flux_tally = openmc.Tally(name='Flux')
flux_tally.filters = [mesh_filter]
flux_tally.scores = ['flux']
# Add tallies to the tallies file
model.tallies.append(flux_tally)
############################################################################################################
# SETTINGS
model.settings.batches = 10
model.settings.inactive = 5
model.settings.particles = 100
model.settings.source = openmc.Source(space=openmc.stats.Box(
[-pitch / 2, -pitch / 2, -height / 2], [pitch / 2, pitch / 2, height / 2], only_fissionable=True))
############################################################################################################
# PLOT
plot = openmc.Plot()
plot.origin = (0.0, 0.0, 0)
plot.width = (21.42, 21.42)
plot.pixels = (300, 300)
plot.color_by = 'material'
model.plots.append(plot)
openmc.plot_geometry(output=False)
return model
def pwr_core():
"""Create a PWR full-core model.
This model is the OECD/NEA Monte Carlo Performance benchmark which is a
grossly simplified pressurized water reactor (PWR) with 241 fuel
assemblies. Note that the number of particles/batches is initially set very
low for testing purposes.
Returns
-------
model : openmc.model.Model
Full-core PWR model
"""
model = openmc.model.Model()
############################################################################################################
# MATERIALS
# Define materials.
# 1 Fuel
fuel = openmc.Material(1, name='UOX fuel')
fuel.depletable = True
fuel.set_density('g/cm3', 10.062)
fuel.add_nuclide("U234", 4.9476e-6)
fuel.add_nuclide("U235", 4.8218e-4)
fuel.add_nuclide("U238", 2.1504e-2)
fuel.add_nuclide("Xe135", 1.0801e-8)
fuel.add_nuclide("O16", 4.5737e-2)
clad = openmc.Material(2, name='Zircaloy')
clad.set_density('g/cm3', 5.77)
clad.add_nuclide("Zr90", 0.5145)
clad.add_nuclide("Zr91", 0.1122)
clad.add_nuclide("Zr92", 0.1715)
clad.add_nuclide("Zr94", 0.1738)
clad.add_nuclide("Zr96", 0.0280)
cold_water = openmc.Material(3, name='Cold borated water')
cold_water.set_density('atom/b-cm', 0.07416)
cold_water.add_nuclide("H1", 2.0)
cold_water.add_nuclide("O16", 1.0)
cold_water.add_nuclide("B10", 6.490e-4)
cold_water.add_nuclide("B11", 2.689e-3)
cold_water.add_s_alpha_beta('c_H_in_H2O')
hot_water = openmc.Material(4, name='Hot borated water')
hot_water.set_density('atom/b-cm', 0.06614)
hot_water.add_nuclide("H1", 2.0)
hot_water.add_nuclide("O16", 1.0)
hot_water.add_nuclide("B10", 6.490e-4)
hot_water.add_nuclide("B11", 2.689e-3)
hot_water.add_s_alpha_beta('c_H_in_H2O')
rpv_steel = openmc.Material(5, name='Reactor pressure vessel steel')
rpv_steel.set_density('g/cm3', 7.9)
rpv_steel.add_nuclide("Fe54", 0.05437098, 'wo')
rpv_steel.add_nuclide("Fe56", 0.88500663, 'wo')
rpv_steel.add_nuclide("Fe57", 0.0208008, 'wo')
rpv_steel.add_nuclide("Fe58", 0.00282159, 'wo')
rpv_steel.add_nuclide("Ni58", 0.0067198, 'wo')
rpv_steel.add_nuclide("Ni60", 0.0026776, 'wo')
rpv_steel.add_nuclide("Mn55", 0.01, 'wo')
rpv_steel.add_nuclide("Cr52", 0.002092475, 'wo')
rpv_steel.add_nuclide("C0", 0.0025, 'wo')
rpv_steel.add_nuclide("Cu63", 0.0013696, 'wo')
lower_rad_ref = openmc.Material(6, name='Lower radial reflector')
lower_rad_ref.set_density('g/cm3', 4.32)
lower_rad_ref.add_nuclide("H1", 0.0095661, 'wo')
lower_rad_ref.add_nuclide("O16", 0.0759107, 'wo')
lower_rad_ref.add_nuclide("B10", 3.08409e-5, 'wo')
lower_rad_ref.add_nuclide("B11", 1.40499e-4, 'wo')
lower_rad_ref.add_nuclide("Fe54", 0.035620772088, 'wo')
lower_rad_ref.add_nuclide("Fe56", 0.579805982228, 'wo')
lower_rad_ref.add_nuclide("Fe57", 0.01362750048, 'wo')
lower_rad_ref.add_nuclide("Fe58", 0.001848545204, 'wo')
lower_rad_ref.add_nuclide("Ni58", 0.055298376566, 'wo')
lower_rad_ref.add_nuclide("Mn55", 0.0182870, 'wo')
lower_rad_ref.add_nuclide("Cr52", 0.145407678031, 'wo')
lower_rad_ref.add_s_alpha_beta('c_H_in_H2O')
upper_rad_ref = openmc.Material(7, name='Upper radial reflector / Top plate region')
upper_rad_ref.set_density('g/cm3', 4.28)
upper_rad_ref.add_nuclide("H1", 0.0086117, 'wo')
upper_rad_ref.add_nuclide("O16", 0.0683369, 'wo')
upper_rad_ref.add_nuclide("B10", 2.77638e-5, 'wo')
upper_rad_ref.add_nuclide("B11", 1.26481e-4, 'wo')
upper_rad_ref.add_nuclide("Fe54", 0.035953677186, 'wo')
upper_rad_ref.add_nuclide("Fe56", 0.585224740891, 'wo')
upper_rad_ref.add_nuclide("Fe57", 0.01375486056, 'wo')
upper_rad_ref.add_nuclide("Fe58", 0.001865821363, 'wo')
upper_rad_ref.add_nuclide("Ni58", 0.055815129186, 'wo')
upper_rad_ref.add_nuclide("Mn55", 0.0184579, 'wo')
upper_rad_ref.add_nuclide("Cr52", 0.146766614995, 'wo')
upper_rad_ref.add_s_alpha_beta('c_H_in_H2O')
bot_plate = openmc.Material(8, name='Bottom plate region')
bot_plate.set_density('g/cm3', 7.184)
bot_plate.add_nuclide("H1", 0.0011505, 'wo')
bot_plate.add_nuclide("O16", 0.0091296, 'wo')
bot_plate.add_nuclide("B10", 3.70915e-6, 'wo')
bot_plate.add_nuclide("B11", 1.68974e-5, 'wo')
bot_plate.add_nuclide("Fe54", 0.03855611055, 'wo')
bot_plate.add_nuclide("Fe56", 0.627585036425, 'wo')
bot_plate.add_nuclide("Fe57", 0.014750478, 'wo')
bot_plate.add_nuclide("Fe58", 0.002000875025, 'wo')
bot_plate.add_nuclide("Ni58", 0.059855207342, 'wo')
bot_plate.add_nuclide("Mn55", 0.0197940, 'wo')
bot_plate.add_nuclide("Cr52", 0.157390026871, 'wo')
bot_plate.add_s_alpha_beta('c_H_in_H2O')
bot_nozzle = openmc.Material(9, name='Bottom nozzle region')
bot_nozzle.set_density('g/cm3', 2.53)
bot_nozzle.add_nuclide("H1", 0.0245014, 'wo')
bot_nozzle.add_nuclide("O16", 0.1944274, 'wo')
bot_nozzle.add_nuclide("B10", 7.89917e-5, 'wo')
bot_nozzle.add_nuclide("B11", 3.59854e-4, 'wo')
bot_nozzle.add_nuclide("Fe54", 0.030411411144, 'wo')
bot_nozzle.add_nuclide("Fe56", 0.495012237964, 'wo')
bot_nozzle.add_nuclide("Fe57", 0.01163454624, 'wo')
bot_nozzle.add_nuclide("Fe58", 0.001578204652, 'wo')
bot_nozzle.add_nuclide("Ni58", 0.047211231662, 'wo')
bot_nozzle.add_nuclide("Mn55", 0.0156126, 'wo')
bot_nozzle.add_nuclide("Cr52", 0.124142524198, 'wo')
bot_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_nozzle = openmc.Material(10, name='Top nozzle region')
top_nozzle.set_density('g/cm3', 1.746)
top_nozzle.add_nuclide("H1", 0.0358870, 'wo')
top_nozzle.add_nuclide("O16", 0.2847761, 'wo')
top_nozzle.add_nuclide("B10", 1.15699e-4, 'wo')
top_nozzle.add_nuclide("B11", 5.27075e-4, 'wo')
top_nozzle.add_nuclide("Fe54", 0.02644016154, 'wo')
top_nozzle.add_nuclide("Fe56", 0.43037146399, 'wo')
top_nozzle.add_nuclide("Fe57", 0.0101152584, 'wo')
top_nozzle.add_nuclide("Fe58", 0.00137211607, 'wo')
top_nozzle.add_nuclide("Ni58", 0.04104621835, 'wo')
top_nozzle.add_nuclide("Mn55", 0.0135739, 'wo')
top_nozzle.add_nuclide("Cr52", 0.107931450781, 'wo')
top_nozzle.add_s_alpha_beta('c_H_in_H2O')
top_fa = openmc.Material(11, name='Top of fuel assemblies')
top_fa.set_density('g/cm3', 3.044)
top_fa.add_nuclide("H1", 0.0162913, 'wo')
top_fa.add_nuclide("O16", 0.1292776, 'wo')
top_fa.add_nuclide("B10", 5.25228e-5, 'wo')
top_fa.add_nuclide("B11", 2.39272e-4, 'wo')
top_fa.add_nuclide("Zr90", 0.43313403903, 'wo')
top_fa.add_nuclide("Zr91", 0.09549277374, 'wo')
top_fa.add_nuclide("Zr92", 0.14759527104, 'wo')
top_fa.add_nuclide("Zr94", 0.15280552077, 'wo')
top_fa.add_nuclide("Zr96", 0.02511169542, 'wo')
top_fa.add_s_alpha_beta('c_H_in_H2O')
bot_fa = openmc.Material(12, name='Bottom of fuel assemblies')
bot_fa.set_density('g/cm3', 1.762)
bot_fa.add_nuclide("H1", 0.0292856, 'wo')
bot_fa.add_nuclide("O16", 0.2323919, 'wo')
bot_fa.add_nuclide("B10", 9.44159e-5, 'wo')
bot_fa.add_nuclide("B11", 4.30120e-4, 'wo')
bot_fa.add_nuclide("Zr90", 0.3741373658, 'wo')
bot_fa.add_nuclide("Zr91", 0.0824858164, 'wo')
bot_fa.add_nuclide("Zr92", 0.1274914944, 'wo')
bot_fa.add_nuclide("Zr94", 0.1319920622, 'wo')
bot_fa.add_nuclide("Zr96", 0.0216912612, 'wo')
bot_fa.add_s_alpha_beta('c_H_in_H2O')
# Define the materials file.
model.materials = (fuel, clad, cold_water, hot_water, rpv_steel,
lower_rad_ref, upper_rad_ref, bot_plate,
bot_nozzle, top_nozzle, top_fa, bot_fa)
############################################################################################################
# GEOMETRY
# Define surfaces.
s1 = openmc.ZCylinder(R=0.410, surface_id=1)
s2 = openmc.ZCylinder(R=0.475, surface_id=2)
s3 = openmc.ZCylinder(R=0.560, surface_id=3)
s4 = openmc.ZCylinder(R=0.620, surface_id=4)
s5 = openmc.ZCylinder(R=187.6, surface_id=5)
s6 = openmc.ZCylinder(R=209.0, surface_id=6)
s7 = openmc.ZCylinder(R=229.0, surface_id=7)
s8 = openmc.ZCylinder(R=249.0, surface_id=8, boundary_type='vacuum')
s31 = openmc.ZPlane(z0=-229.0, surface_id=31, boundary_type='vacuum')
s32 = openmc.ZPlane(z0=-199.0, surface_id=32)
s33 = openmc.ZPlane(z0=-193.0, surface_id=33)
s34 = openmc.ZPlane(z0=-183.0, surface_id=34)
s35 = openmc.ZPlane(z0=0.000, surface_id=35)
s36 = openmc.ZPlane(z0=183.0, surface_id=36)
s37 = openmc.ZPlane(z0=203.0, surface_id=37)
s38 = openmc.ZPlane(z0=215.0, surface_id=38)
s39 = openmc.ZPlane(z0=223.0, surface_id=39, boundary_type='vacuum')
# Define pin cells.
fuel_cold = openmc.Universe(name='Fuel pin, cladding, cold water', universe_id=1)
c21 = openmc.Cell(cell_id=21, fill=fuel, region=-s1)
c22 = openmc.Cell(cell_id=22, fill=clad, region=+s1 & -s2)
c23 = openmc.Cell(cell_id=23, fill=cold_water, region=+s2)
fuel_cold.add_cells((c21, c22, c23))
tube_cold = openmc.Universe(name='Instrumentation guide tube, '
'cold water', universe_id=2)
c24 = openmc.Cell(cell_id=24, fill=cold_water, region=-s3)
c25 = openmc.Cell(cell_id=25, fill=clad, region=+s3 & -s4)
c26 = openmc.Cell(cell_id=26, fill=cold_water, region=+s4)
tube_cold.add_cells((c24, c25, c26))
fuel_hot = openmc.Universe(name='Fuel pin, cladding, hot water', universe_id=3)
c27 = openmc.Cell(cell_id=27, fill=fuel, region=-s1)
c28 = openmc.Cell(cell_id=28, fill=clad, region=+s1 & -s2)
c29 = openmc.Cell(cell_id=29, fill=hot_water, region=+s2)
fuel_hot.add_cells((c27, c28, c29))
tube_hot = openmc.Universe(name='Instrumentation guide tube, hot water', universe_id=4)
c30 = openmc.Cell(cell_id=30, fill=hot_water, region=-s3)
c31 = openmc.Cell(cell_id=31, fill=clad, region=+s3 & -s4)
c32 = openmc.Cell(cell_id=32, fill=hot_water, region=+s4)
tube_hot.add_cells((c30, c31, c32))
# Set positions occupied by guide tubes
tube_x = np.array([5, 8, 11, 3, 13, 2, 5, 8, 11, 14, 2, 5, 8, 11, 14,
2, 5, 8, 11, 14, 3, 13, 5, 8, 11])
tube_y = np.array([2, 2, 2, 3, 3, 5, 5, 5, 5, 5, 8, 8, 8, 8, 8,
11, 11, 11, 11, 11, 13, 13, 14, 14, 14])
# Define fuel lattices.
NUM_PINS = 17
PITCH_PIN = 1.26
l100 = openmc.RectLattice(name='Fuel assembly (lower half)', lattice_id=100)
l100.lower_left = (-PITCH_PIN * NUM_PINS / 2, -PITCH_PIN * NUM_PINS / 2)
l100.pitch = (PITCH_PIN, PITCH_PIN)
l100.universes = np.tile(fuel_cold, (NUM_PINS, NUM_PINS))
l100.universes[tube_x, tube_y] = tube_cold
l101 = openmc.RectLattice(name='Fuel assembly (upper half)', lattice_id=101)
l101.lower_left = (-PITCH_PIN * NUM_PINS / 2, -PITCH_PIN * NUM_PINS / 2)
l101.pitch = (PITCH_PIN, PITCH_PIN)
l101.universes = np.tile(fuel_hot, (NUM_PINS, NUM_PINS))
l101.universes[tube_x, tube_y] = tube_hot
# Define assemblies.
fa_cw = openmc.Universe(name='Water assembly (cold)', universe_id=5)
c50 = openmc.Cell(cell_id=50, fill=cold_water, region=+s34 & -s35)
fa_cw.add_cell(c50)
fa_hw = openmc.Universe(name='Water assembly (hot)', universe_id=7)
c70 = openmc.Cell(cell_id=70, fill=hot_water, region=+s35 & -s36)
fa_hw.add_cell(c70)
fa_cold = openmc.Universe(name='Fuel assembly (cold)', universe_id=6)
c60 = openmc.Cell(cell_id=60, fill=l100, region=+s34 & -s35)
fa_cold.add_cell(c60)
fa_hot = openmc.Universe(name='Fuel assembly (hot)', universe_id=8)
c80 = openmc.Cell(cell_id=80, fill=l101, region=+s35 & -s36)
fa_hot.add_cell(c80)
# Define core lattices
PITCH_ASSEMBLY = 21.42
l200 = openmc.RectLattice(name='Core lattice (lower half)', lattice_id=200)
l200.lower_left = (-PITCH_ASSEMBLY * 21 / 2, -PITCH_ASSEMBLY * 21 / 2)
l200.pitch = (PITCH_ASSEMBLY, PITCH_ASSEMBLY)
l200.universes = [
[fa_cw] * 21,
[fa_cw] * 21,
[fa_cw] * 7 + [fa_cold] * 7 + [fa_cw] * 7,
[fa_cw] * 5 + [fa_cold] * 11 + [fa_cw] * 5,
[fa_cw] * 4 + [fa_cold] * 13 + [fa_cw] * 4,
[fa_cw] * 3 + [fa_cold] * 15 + [fa_cw] * 3,
[fa_cw] * 3 + [fa_cold] * 15 + [fa_cw] * 3,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 2 + [fa_cold] * 17 + [fa_cw] * 2,
[fa_cw] * 3 + [fa_cold] * 15 + [fa_cw] * 3,
[fa_cw] * 3 + [fa_cold] * 15 + [fa_cw] * 3,
[fa_cw] * 4 + [fa_cold] * 13 + [fa_cw] * 4,
[fa_cw] * 5 + [fa_cold] * 11 + [fa_cw] * 5,
[fa_cw] * 7 + [fa_cold] * 7 + [fa_cw] * 7,
[fa_cw] * 21,
[fa_cw] * 21]
l201 = openmc.RectLattice(name='Core lattice (upper half)', lattice_id=201)
l201.lower_left = (-PITCH_ASSEMBLY * 21 / 2, -PITCH_ASSEMBLY * 21 / 2)
l201.pitch = (PITCH_ASSEMBLY, PITCH_ASSEMBLY)
l201.universes = [
[fa_hw] * 21,
[fa_hw] * 21,
[fa_hw] * 7 + [fa_hot] * 7 + [fa_hw] * 7,
[fa_hw] * 5 + [fa_hot] * 11 + [fa_hw] * 5,
[fa_hw] * 4 + [fa_hot] * 13 + [fa_hw] * 4,
[fa_hw] * 3 + [fa_hot] * 15 + [fa_hw] * 3,
[fa_hw] * 3 + [fa_hot] * 15 + [fa_hw] * 3,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 2 + [fa_hot] * 17 + [fa_hw] * 2,
[fa_hw] * 3 + [fa_hot] * 15 + [fa_hw] * 3,
[fa_hw] * 3 + [fa_hot] * 15 + [fa_hw] * 3,
[fa_hw] * 4 + [fa_hot] * 13 + [fa_hw] * 4,
[fa_hw] * 5 + [fa_hot] * 11 + [fa_hw] * 5,
[fa_hw] * 7 + [fa_hot] * 7 + [fa_hw] * 7,
[fa_hw] * 21,
[fa_hw] * 21]
# Define root universe.
root = openmc.Universe(universe_id=0, name='root universe')
c1 = openmc.Cell(cell_id=1, fill=l200, region=-s6 & +s34 & -s35)
c2 = openmc.Cell(cell_id=2, fill=l201, region=-s6 & +s35 & -s36)
c3 = openmc.Cell(cell_id=3, fill=bot_plate, region=-s7 & +s31 & -s32)
c4 = openmc.Cell(cell_id=4, fill=bot_nozzle, region=-s5 & +s32 & -s33)
c5 = openmc.Cell(cell_id=5, fill=bot_fa, region=-s5 & +s33 & -s34)
c6 = openmc.Cell(cell_id=6, fill=top_fa, region=-s5 & +s36 & -s37)
c7 = openmc.Cell(cell_id=7, fill=top_nozzle, region=-s5 & +s37 & -s38)
c8 = openmc.Cell(cell_id=8, fill=upper_rad_ref, region=-s7 & +s38 & -s39)
c9 = openmc.Cell(cell_id=9, fill=bot_nozzle, region=+s6 & -s7 & +s32 & -s38)
c10 = openmc.Cell(cell_id=10, fill=rpv_steel, region=+s7 & -s8 & +s31 & -s39)
c11 = openmc.Cell(cell_id=11, fill=lower_rad_ref, region=+s5 & -s6 & +s32 & -s34)
c12 = openmc.Cell(cell_id=12, fill=upper_rad_ref, region=+s5 & -s6 & +s36 & -s38)
root.add_cells((c1, c2, c3, c4, c5, c6, c7, c8, c9, c10, c11, c12))
# Assign root universe to geometry
model.geometry.root_universe = root
############################################################################################################
# MESH
mesh = openmc.Mesh(mesh_id=1)
mesh.type = 'regular'
mesh.dimension = (1000, 1000, 1)
mesh.lower_left = (-229.0, -229.0, -229.0)
mesh.upper_right = (229.0, 229.0, 229.0)
# Create a mesh filter
mesh_filter = openmc.MeshFilter(mesh)
############################################################################################################
# CROSS-SECTION LIBRARY
energy_groups = openmc.mgxs.EnergyGroups()
energy_bounds = np.logspace(-3, 7.3, 21)
energy_groups.group_edges = energy_bounds
energy_filter = openmc.EnergyFilter(np.logspace(-3, 7.3, 1000))
# Instantiate a 1-group EnergyGroups object
one_group = openmc.mgxs.EnergyGroups()
one_group.group_edges = np.array([energy_groups.group_edges[0], energy_groups.group_edges[-1]])
# Initialize an 20-energy-group and 6-delayed-group MGXS Library
mgxs_lib = openmc.mgxs.Library(model.geometry)
mgxs_lib.energy_groups = energy_groups
mgxs_lib.num_delayed_groups = 6
# Specify multi-group cross section types to compute
mgxs_lib.mgxs_types = ['total', 'transport', 'nu-scatter matrix', 'kappa-fission', 'inverse-velocity', 'chi-prompt',
'prompt-nu-fission', 'chi-delayed', 'delayed-nu-fission', 'beta']
# Specify a "mesh" domain type for the cross section tally filters
mgxs_lib.domain_type = 'mesh'
# Specify the mesh domain over which to compute multi-group cross sections
mgxs_lib.domains = [mesh]
# Construct all tallies needed for the multi-group cross section library
mgxs_lib.build_library()
############################################################################################################
# TALLIES
# Create a "tallies.xml" file for the MGXS Library
model.tallies = openmc.Tallies()
mgxs_lib.add_to_tallies_file(model.tallies, merge=True)
# Instantiate a flux tally; Other valid options: 'current', 'fission', etc
flux_tally = openmc.Tally(name='Mesh')
flux_tally.filters = [mesh_filter]
flux_tally.scores = ['flux']
energy_tally = openmc.Tally(name='Energy')
energy_tally.filters = [energy_filter]
energy_tally.scores = ['flux']
# Add tallies to the tallies file
model.tallies.append(flux_tally)
model.tallies.append(energy_tally)
############################################################################################################
# SETTINGS
model.settings.batches = 45
model.settings.inactive = 5
model.settings.particles = 5000
uniform_dist = openmc.stats.Box([-160, -160, -183], [160, 160, 183], only_fissionable=False)
model.settings.source = openmc.Source(space=uniform_dist)
# to track particles 3 and 4 from batch 1 and generation 2:
# model.settings.track = (1, 2, 3, 1, 2, 4)
############################################################################################################
# PLOT
plot = openmc.Plot().from_geometry(model.geometry)
plot.filename = 'Full-Core'
plot.pixels = (3000, 3000)
plot.basis = 'xy'
plot.color_by = 'material'
model.plots.append(plot)
openmc.plot_geometry(output=True)
############################################################################################################
return model
##########################################
# plt.plot(sp.k_generation)
# plt.plot(sp.entropy)
def flux_spectrum(statepoint_file, desired_score=''):
sp = openmc.StatePoint(filename=statepoint_file)
flux = sp.get_tally(name='Energy').get_slice(scores=['flux'])
df = flux.get_pandas_dataframe()
fluxes = df['mean'].values
fluxes = np.delete(fluxes, 0)
# Extract the energy bins from the Tally's EnergyFilter
energy_filter = flux.find_filter(openmc.EnergyFilter)
energies = energy_filter.bins
energies = np.delete(energies, [0, 1])
print(len(energies))
print(energies)
print(len(fluxes))
print(fluxes)
return energies, fluxes
# https://openmc.readthedocs.io/en/stable/io_formats/statepoint.html#io-statepoint
def statepoint_evaluation(statepoint_file, desired_score=''):
sp = openmc.StatePoint(filename=statepoint_file)
# Initialize MGXS Library with OpenMC statepoint data
# xs_lib.load_from_statepoint(sp)
# Extract the current tally separately
# current_tally = sp.get_tally(name='current tally')
score_df = sp.get_tally(name='Mesh').get_slice(scores=['flux']).get_pandas_dataframe()
print(score_df.head(5).to_string())
axial_dfs = []
for i in range(1):
score_df = score_df[score_df['mesh 1']['z'] == i + 1]
print(score_df.head(1000).to_string())
x = score_df['mesh 1']['x']
y = score_df['mesh 1']['y']
mean = score_df['mean']
new_df = pd.DataFrame(list(zip(x, y, mean)), columns=['x', 'y', 'mean'])
new_df = new_df.pivot(index='y', columns='x', values='mean')
axial_dfs.append(np.array(new_df.values))
return axial_dfs
# pwr_core().export_to_xml()
# pwr_core().geometry.export_to_xml()
# pwr_core().materials.export_to_xml()
# pwr_core().settings.export_to_xml()
# pwr_core().plots.export_to_xml()
# pwr_core().tallies.export_to_xml()
#
# im = Image.open(".ppm")
# im.save(".png")
# pwr_assembly().export_to_xml()
pwr_core().export_to_xml()
openmc.run(output=True)
axial_dfs = statepoint_evaluation('statepoint.45.h5')
energy, spectrum = flux_spectrum('statepoint.45.h5')
plot = True
if plot:
##################################################################
data_1 = [go.Scatter(x=energy,
y=spectrum,
mode='lines',
marker=dict(line=dict(width=1))
)]
layout_1 = dict(title='Test',
width=1500,
height=1000,
xaxis=dict(type='log'),
yaxis=dict(type='log'),
)
figure_1 = dict(data=data_1, layout=layout_1)
plotly.offline.plot(figure_1, filename='file1.html')
##################################################################
# Instantiate Data
data = [go.Surface(
# z=np.zeros(np.shape(axial_dfs[0])),
z=axial_dfs[0],
surfacecolor=axial_dfs[0],
colorscale='RdBu',
)]
# Layout
layout = dict(title='Test',
hovermode='closest',
width=1500,
height=1000,
scene=dict(
# xaxis=dict(range=[-10, 100]),
# yaxis=dict(range=[-10, 100]),
zaxis=dict(range=[0, .005])),
)
figure = dict(data=data, layout=layout)
plotly.offline.plot(figure, filename='file.html')
##################################################################
# if plot:
#
# maxes = []
# for m in range(len(axial_dfs)):
# c_max = np.amax(np.array(axial_dfs[m]))
# maxes.append(c_max)
# c_max = np.max(np.array(maxes))
#
# # Instantiate Data
# data = [go.Surface(z=np.zeros(np.shape(axial_dfs[0])),
# surfacecolor=axial_dfs[0],
# cmax=c_max,
# cmin=0,
# colorscale='Viridis',
# )]
#
# ###############################################
#
# # Instantiate Frames
# frames = []
# steps = []
# for k in range(len(axial_dfs)):
# frame_data = go.Surface(z=np.full(np.shape(axial_dfs[0]), k),
# surfacecolor=axial_dfs[k]
# )
# frame = dict(data=[frame_data], name='Axial Step {}'.format(k))
# frames.append(frame)
#
# slider_step = dict(args=[
# ['Axial Step {}'.format(k)],
# dict(frame=dict(duration=0, redraw=False),
# mode='immediate',
# transition={'duration': 0})
# ],
# label='{}'.format(k),
# method='animate')
# steps.append(slider_step)
#
# ##################################################################
#
# # Slider Control
# sliders_dict = dict(active=0, # Starting Position
# yanchor='top',
# xanchor='left',
# currentvalue=dict(
# font={'size': 20},
# prefix='Axial Step:',
# visible=True,
# xanchor='right'
# ),
# # Transition for slider button
# transition=dict(duration=500,
# easing='cubic-in-out'),
# pad={'b': 10, 't': 50},
# len=.9,
# x=0.1,
# y=0,
# steps=steps
# )
#
# ##################################################################
#
# # Layout
# layout = dict(title='Test',
# hovermode='closest',
# width=1500,
# height=1000,
# scene=dict(
# zaxis=dict(range=[0, 10])),
# updatemenus=[dict(type='buttons',
#
# buttons=[dict(args=[None,
# dict(frame=dict(duration=500,
# redraw=False),
# fromcurrent=True,
# transition=dict(duration=100,
# easing='quadratic-in-out'))],
# label=u'Play',
# method=u'animate'
# ),
#
# # [] around "None" are important!
# dict(args=[[None], dict(frame=dict(duration=0,
# redraw=False),
# mode='immediate',
# transition=dict(duration=0))],
# label='Pause',
# method='animate'
# )
# ],
#