5-MWe - reactor simulation

This notebook shows the input file for modeling and simulating the 5-MWe reactor in North Korea with ONIX.

OpenMC input part

import openmc
import onix

---------------------------------------------------------------------------
ModuleNotFoundError                       Traceback (most recent call last)
<ipython-input-1-51377a570dec> in <module>
----> 1 import openmc
      2 import onix

ModuleNotFoundError: No module named 'openmc'

The 5-MWe is made of graphite blocks with fuel channels through which carbon dioxyde gas flows to cool the fuel rod. The fuel rod is made of natural uranium and is cladded with magnox alloy (an alloy of magnesium with small amount of aluminum and other metals).

natU_mat = openmc.Material()
natU_mat.temperature = 800
natU_mat.set_density('g/cc', density = 17.98)
natU_mat.add_nuclide('U234', d.NATURAL_ABUNDANCE['U234'])
natU_mat.add_nuclide('U235', d.NATURAL_ABUNDANCE['U235'])
natU_mat.add_nuclide('U238', d.NATURAL_ABUNDANCE['U238'])

clad_mat = openmc.Material()
clad_mat.temperature = 700
Mg_frac = 98.95
Al_frac = 1
Be_frac = 0.05
clad_mat.add_nuclide('Mg24',0.79*Mg_frac)
clad_mat.add_nuclide('Mg25',0.10*Mg_frac)
clad_mat.add_nuclide('Mg26',0.11*Mg_frac)
clad_mat.add_nuclide('Al27',1*Al_frac)
clad_mat.add_nuclide('Be9',1*Be_frac)
clad_mat.set_density('g/cc', density = 1.65)

mod_mat = openmc.Material()
mod_mat.temperature = 650
mod_mat.set_density('g/cc', density = 1.628)
mod_mat.add_nuclide('C12',0.989)
mod_mat.add_nuclide('C13',0.011)
mod_mat.add_s_alpha_beta('c_Graphite')

sample_mat = openmc.Material()
sample_mat.temperature = 650
sample_mat.set_density('g/cc', density = 1.628)
sample_mat.add_nuclide('C12',0.989)
sample_mat.add_nuclide('C13',0.011)
sample_mat.add_s_alpha_beta('c_Graphite')

cool_mat = openmc.Material()
cool_mat.temperature = 650
cool_mat.set_density('g/cc', density = 1.628)
C_frac = 33.33
O_frac = 66.66
cool_mat.add_nuclide('C12',C_frac*0.989)
cool_mat.add_nuclide('C13',C_frac*0.011)
cool_mat.add_nuclide('O16',O_frac*0.9976)
# Instantiate a Materials collection
materials_file = openmc.Materials([natU_mat, clad_mat, cool_mat, mod_mat, sample_mat])
#materials_file.cross_sections = '/tigress/jdtdl/openmc/mcnp_endfb70/cross_sections.xml'
materials_file.cross_sections = '/home/groups/rewing1/cross-sections/lib80x_hdf5/cross_sections.xml'

# Export to "materials.xml"
materials_file.export_to_xml()

The following commands define the geometry and the cells.

In order to measure neutronics parameters toward the edges of the fuel channels (against the wall of the graphite block), a sample cell is added. This cell allows to tally reaction rates and neutron flux for nuclear archaeology methods that would use a monitor tag located in this region.

# Create cylinders for the fuel and clad
fuel_outer_radius = openmc.ZCylinder(x0=0.0, y0=0.0, R=1.45)
clad_outer_radius = openmc.ZCylinder(x0=0.0, y0=0.0, R=1.50)
coolant_outer_radius = openmc.ZCylinder(x0=0.0, y0=0.0, R=3.25)

#Create cylinder of the sample
sample_cylinder = openmc.ZCylinder(x0=0.0, y0=3.35, R=0.1)

# Create boundary planes to surround the geometry
min_x = openmc.XPlane(x0=-10, boundary_type='reflective')
max_x = openmc.XPlane(x0=+10, boundary_type='reflective')
min_y = openmc.YPlane(y0=-10, boundary_type='reflective')
max_y = openmc.YPlane(y0=+10, boundary_type='reflective')
min_z = openmc.ZPlane(z0=-0.5, boundary_type='reflective')
max_z = openmc.ZPlane(z0=+0.5, boundary_type='reflective')

# Create a Universe to encapsulate the fuel pin
pin_cell_universe = openmc.Universe(name='Fuel Pin')

# Create fuel1 Cell
fuel_cell = openmc.Cell(name='fuel')
fuel_cell.fill = natU_mat
fuel_cell.region = -fuel_outer_radius
pin_cell_universe.add_cell(fuel_cell)

# Create a clad Cell for 1
clad_cell = openmc.Cell(name='Clad')
clad_cell.fill = clad_mat
clad_cell.region = +fuel_outer_radius & -clad_outer_radius
pin_cell_universe.add_cell(clad_cell)

# Create a coolant Cell for 1
cool_cell = openmc.Cell(name='Cool')
cool_cell.fill = cool_mat
cool_cell.region = +clad_outer_radius & - coolant_outer_radius
pin_cell_universe.add_cell(cool_cell)

# Create a moderator Cell for 1
mod_cell = openmc.Cell(name='Mod')
mod_cell.fill = mod_mat
mod_cell.region = +coolant_outer_radius & +sample_cylinder
pin_cell_universe.add_cell(mod_cell)

# Create sample cell
sample_cell = openmc.Cell(name='Sample')
sample_cell.fill = sample_mat
sample_cell.region = -sample_cylinder
pin_cell_universe.add_cell(sample_cell)

# Create root Cell
root_cell = openmc.Cell(name='root cell')
root_cell.fill = pin_cell_universe
root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z

# Create root Universe
root_universe = openmc.Universe(universe_id=0, name='root universe')
root_universe.add_cell(root_cell)

# Create Geometry and set root Universe
openmc_geometry = openmc.Geometry(root_universe)

# Export to "materials.xml"
materials_file.export_to_xml()
# Export to "geometry.xml"
openmc_geometry.export_to_xml()

The following commands are for the simulations settings of OpenMC.

There are no particular requirements from ONIX here so users can input these commands as they would for a normal OpenMC simulation.

# OpenMC simulation parameters
batches = 100
inactive = 10
particles = 10000

# Instantiate a Settings object
settings_file = openmc.Settings()
settings_file.batches = batches
settings_file.inactive = inactive
settings_file.particles = particles

# Create an initial uniform spatial source distribution over fissionable zones
#bounds = [-0.65635, -0.65635, -0.65635, 0.65635, 0.65635, 0.65635]
#uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
point_dist = openmc.stats.Point(xyz=(0.0, 0.0, 0.0))
settings_file.source = openmc.source.Source(space=point_dist)

# Export to "settings.xml"
settings_file.export_to_xml()

ONIX input part

This section describes the input commands for ONIX.

Defning a sequence

As we assume that the 5-MWe reactor has never been operated to a burnup higher than 0.7 MWd/kg, we define a burnup interval up to that value. With a total active volume of approximately 166 cubic meters and a nominal thermal power output of 25 MWth, the volumic power density is set to 0.15 kW/l (other power densities yield almost identical neutronics results and therefore this value can be used for other thermal power values).

macrostep_vector = [0.0087, 0.05, 0.1, 0.2, 0.3, 0.4, 0.5, 0.6, 0.7]
macrostep_unit = 'MWd/kg'

norma_vector = [0.15]*len(macrostep_vector)
norma_unit = 'power'

microstep_vector = [3]*len(macrostep_vector)

sequence1 = onix.Sequence(1)
sequence1.set_macrostep(macrostep_vector, macrostep_unit)
sequence1.set_norma(norma_vector, norma_unit)
sequence1.microstep_vector = microstep_vector

sequence1.flux_approximation ='iv'

couple = onix.couple.Couple_openmc()

couple.set_bounding_box([-10.16, -10.16, -0.5], [10.16, 10.16, 0.5])

couple.openmc_bin_path = '/home/julien/virtualenvs/py3/bin/openmc'

couple.select_bucells([fuel_cell, sample_cell])

couple.import_openmc(root_cell)

vol_dict = {'fuel': 6.6051, 'Sample':0.03141 ,'total volume':400}
couple.set_vol(vol_dict)

couple.set_sequence(sequence1)

We can finally ask ONIX to burn the system and launch the coupled simulation

couple.burn()

Output results are going to be located in two types of folder. Per step folders include densities, power, neutron flux, burnup, one-group cross sections and burnup matrices for each macrostep, separately. The output summary folder contains aggregated output results for the whole simulation and for every macrostep.