Error when running depletion calculation

User Support
1

Hi,

I’m running into this error every time I try to run a depletion calculation with ‘diff_burnable_mats=True’:

ValueError: Number of material instances have not been determined. Call the Geometry.determine_paths() method.

How can I fix this?

I can’t attach the code, so I will paste it here, hope this is not a problem!

#!/usr/bin/env python3
# -*- coding: utf-8 -*-
"""
Created on Thu Oct 20 10:23:41 2022

@author: Marti
"""
#!/usr/bin/env python3
# -*- coding: utf-8 -*-
"""
Created on Tue Oct 18 12:58:18 2022

@author: Marti
"""
from math import pi, sin, cos, sqrt
import numpy as np
import matplotlib.pyplot as plt
import openmc
import openmc.deplete

 

# Materials definitions
openmc.Materials.cross_sections = "/Users/Marti/Downloads/endfb71_hdf5/cross_sections.xml"
inner_fuel = openmc.Material(name='Inner fuel')
inner_fuel.add_nuclide('U235', 0.19, 'wo')
inner_fuel.add_nuclide('U238', 75.09, 'wo')
inner_fuel.add_nuclide('Pu238', 0.46, 'wo')
inner_fuel.add_nuclide('Pu239', 6.12, 'wo')
inner_fuel.add_nuclide('Pu240', 3.83, 'wo')
inner_fuel.add_nuclide('Pu241', 1.06, 'wo')
inner_fuel.add_nuclide('Am241', 0.10, 'wo')
inner_fuel.add_nuclide('O16', 11.81, 'wo')
inner_fuel.set_density('kg/m3', 10661.4)

outer_fuel = openmc.Material(name='Outer fuel')
outer_fuel.add_nuclide('U235', 0.18, 'wo')
outer_fuel.add_nuclide('U238', 73.0, 'wo')
outer_fuel.add_nuclide('Pu238', 0.54, 'wo')
outer_fuel.add_nuclide('Pu239', 7.11, 'wo')
outer_fuel.add_nuclide('Pu240', 4.45, 'wo')
outer_fuel.add_nuclide('Pu241', 1.24, 'wo')
outer_fuel.add_nuclide('Am241', 0.17, 'wo')
outer_fuel.add_nuclide('O16', 11.75, 'wo')
outer_fuel.set_density('kg/m3', 10661.4)

clad = openmc.Material(name= 'Clad')
clad.add_nuclide('C0', 0.1, 'wo')
clad.add_element('Mn', 0.3, 'wo')
clad.add_element('P', 0.02, 'wo')
clad.add_element('Ni', 0.5, 'wo')
clad.add_element('Cr', 21.5, 'wo')
clad.add_element('Ti', 0.6, 'wo')
clad.add_element('Co', 0.3, 'wo')
clad.add_element('Cu', 0.15, 'wo')
clad.add_element('Al', 5.75, 'wo')
clad.add_nuclide('O16', 0.15, 'wo')
clad.add_element('Y', 0.28, 'wo')
clad.add_element('Fe', 70.08, 'wo')
clad.set_density('kg/m3', 7250.0)

sodium = openmc.Material(name = 'Sodium')
sodium.add_nuclide('Na23', 1.0)
sodium.set_density('kg/m3', 830.0)

boron_carbide = openmc.Material(name='Boron Carbide')
boron_carbide.add_nuclide('B10', 0.198, 'wo')
boron_carbide.add_nuclide('C0', 0.802, 'wo')
boron_carbide.set_density('kg/m3', 2450.0)

helium = openmc.Material(name = 'Helium')
helium.add_nuclide('He4', 1.0)
helium.set_density('kg/m3', 0.4282)

reflector_sodium = openmc.Material(name = 'Reflector Sodium')
reflector_sodium.add_nuclide('Na23', 1.0)
reflector_sodium.set_density('kg/m3', 833.0)

em10_steel = openmc.Material(name = 'Wrapper')
em10_steel.add_nuclide('C0', 0.1, 'wo')
em10_steel.add_element('Cr', 8.5, 'wo')
em10_steel.add_element('Mo', 1.0, 'wo')
em10_steel.add_element('Ni', 0.5, 'wo')
em10_steel.add_element('Mn', 0.5, 'wo')
em10_steel.add_element('Nb', 0.02, 'wo')
em10_steel.add_element('Ti', 0.02, 'wo')
em10_steel.add_element('W', 0.02, 'wo')
em10_steel.add_element('Si', 0.2, 'wo')
em10_steel.add_element('Fe', 89.0, 'wo')
em10_steel.set_density('kg/m3', 7753.0)

reflector_material = openmc.Material.mix_materials([sodium, em10_steel],[0.08, 0.92],'wo')

materials_file = openmc.Materials([inner_fuel, outer_fuel, sodium, em10_steel,
                                   reflector_material, reflector_sodium, helium,
                                   boron_carbide, clad])
materials_file.export_to_xml()

#Geometry definitions

sleeve_min_z = openmc.ZPlane(z0=-50, boundary_type = 'transmission')
sleeve_max_z = openmc.ZPlane(z0=+50, boundary_type = 'transmission')
axial_reflector_max_z = openmc.ZPlane(z0= +131, boundary_type = 'vacuum')
axial_reflector_min_z = openmc.ZPlane(z0= -80, boundary_type = 'transmission') 
gas_plenum_max_z = openmc.ZPlane(z0 = +61, boundary_type = 'transmission') 
gas_plenum_min_z = openmc.ZPlane (z0 = -171, boundary_type = 'vacuum')

fuel_radius_inner = openmc.ZCylinder(r=0.943/2)
clad_inner_radius_inner = openmc.ZCylinder(r=0.973/2)
clad_outer_radius_inner = openmc.ZCylinder(r=1.073/2)

fuel_region_inner = -fuel_radius_inner & -sleeve_max_z & +sleeve_min_z
gap_region_inner  = +fuel_radius_inner & -clad_inner_radius_inner  & -sleeve_max_z & +sleeve_min_z
clad_region_inner = +clad_inner_radius_inner & -clad_outer_radius_inner & -axial_reflector_max_z & +gas_plenum_min_z
axial_reflector_region_top_inner = -clad_inner_radius_inner & -axial_reflector_max_z & +gas_plenum_max_z
axial_reflector_region_bottom_inner = -clad_inner_radius_inner & +axial_reflector_min_z & -sleeve_min_z
gas_plenum_region_top_inner = -clad_inner_radius_inner & +sleeve_max_z & -gas_plenum_max_z
gas_plenum_region_bottom_inner = -clad_inner_radius_inner & +gas_plenum_min_z & -axial_reflector_min_z
surrounding_region_inner = +clad_outer_radius_inner & -axial_reflector_max_z & +gas_plenum_min_z

inner_fuel_cell = openmc.Cell(fill=inner_fuel, region=fuel_region_inner)
gap_cell_inner = openmc.Cell(fill=helium, region=gap_region_inner)
clad_cell_inner = openmc.Cell(fill=clad, region=clad_region_inner)
axial_reflector_top_cell_inner = openmc.Cell(fill=reflector_material,
                                             region=axial_reflector_region_top_inner)
axial_reflector_bottom_cell_inner = openmc.Cell(fill=reflector_material,
                                                region=axial_reflector_region_bottom_inner)
gas_plenum_cell_top_inner = openmc.Cell(fill = helium, region=gas_plenum_region_top_inner)
gas_plenum_cell_bottom_inner = openmc.Cell(fill = helium, 
                                           region=gas_plenum_region_bottom_inner) #
surrounding_cell_inner = openmc.Cell(fill=sodium, region=surrounding_region_inner)

inner_universe = openmc.Universe(cells=(inner_fuel_cell, gap_cell_inner, clad_cell_inner,
                                        axial_reflector_top_cell_inner, 
                                        axial_reflector_bottom_cell_inner, 
                                        gas_plenum_cell_top_inner,
                                        gas_plenum_cell_bottom_inner,
                                        surrounding_cell_inner))

fuel_radius_outer = openmc.ZCylinder(r=0.943/2)
clad_inner_radius_outer = openmc.ZCylinder(r=0.973/2)
clad_outer_radius_outer = openmc.ZCylinder(r=1.073/2)

fuel_region_outer = -fuel_radius_outer & -sleeve_max_z & +sleeve_min_z
gap_region_outer  = +fuel_radius_outer & -clad_inner_radius_outer  & -sleeve_max_z & +sleeve_min_z
clad_region_outer = +clad_inner_radius_outer & -clad_outer_radius_outer & -axial_reflector_max_z & +gas_plenum_min_z
axial_reflector_region_top_outer = -clad_inner_radius_outer & -axial_reflector_max_z & +gas_plenum_max_z
axial_reflector_region_bottom_outer = -clad_inner_radius_outer & +axial_reflector_min_z & -sleeve_min_z
gas_plenum_region_top_outer = -clad_inner_radius_outer & +sleeve_max_z & -gas_plenum_max_z
gas_plenum_region_bottom_outer = -clad_inner_radius_outer & +gas_plenum_min_z & -axial_reflector_min_z
surrounding_region_outer = +clad_outer_radius_outer & -axial_reflector_max_z & +gas_plenum_min_z

outer_fuel_cell = openmc.Cell(fill=outer_fuel, region=fuel_region_outer)
gap_cell_outer = openmc.Cell(fill=helium, region=gap_region_outer)
clad_cell_outer = openmc.Cell(fill=clad, region=clad_region_outer)
axial_reflector_top_cell_outer = openmc.Cell(fill=reflector_material,
                                             region=axial_reflector_region_top_outer)
axial_reflector_bottom_cell_outer = openmc.Cell(fill=reflector_material,
                                                region=axial_reflector_region_bottom_outer)
gas_plenum_cell_top_outer = openmc.Cell(fill = helium, region=gas_plenum_region_top_outer)
gas_plenum_cell_bottom_outer = openmc.Cell(fill = helium, 
                                           region=gas_plenum_region_bottom_outer) #
surrounding_cell_outer = openmc.Cell(fill=sodium, region=surrounding_region_outer)

outer_universe = openmc.Universe(cells=(outer_fuel_cell, gap_cell_outer, clad_cell_outer,
                                        axial_reflector_top_cell_outer, 
                                        axial_reflector_bottom_cell_outer, 
                                        gas_plenum_cell_top_outer,
                                        gas_plenum_cell_bottom_outer,
                                        surrounding_cell_outer))

sodium_fill_cell = openmc.Cell(fill=sodium)
sodium_universe = openmc.Universe(cells=(sodium_fill_cell,))

sodium_prism = openmc.hexagonal_prism(edge_length=11.91073605,
                                      orientation='x')
sodium_prism_region = sodium_prism & +gas_plenum_min_z & -axial_reflector_max_z
sodium_prism_out = ~sodium_prism & +gas_plenum_min_z & -axial_reflector_max_z
sodium_prism_cell = openmc.Cell(fill=sodium, region=sodium_prism_region)
sodium_prism_cell_out = openmc.Cell(fill=sodium, region=sodium_prism_out)
sodium_prism_universe = openmc.Universe(cells=(sodium_prism_cell, sodium_prism_cell_out))


reflector_radius = openmc.ZCylinder(r=1.9)
reflector_cladding_radius = openmc.ZCylinder(r=2.0062)

reflector_region = -reflector_radius & +gas_plenum_min_z & -axial_reflector_max_z
reflector_clad_region = (+reflector_radius & -reflector_cladding_radius & 
                         +gas_plenum_min_z & -axial_reflector_max_z)

reflector_cell = openmc.Cell(fill=reflector_material, region=reflector_region)
reflector_clad_cell = openmc.Cell(fill=em10_steel, region=reflector_clad_region)
reflector_out_cell = openmc.Cell(fill=sodium, region = +reflector_cladding_radius
                                 & +gas_plenum_min_z & -axial_reflector_max_z)
reflector_universe = openmc.Universe(cells = (reflector_cell, reflector_clad_cell,
                                              reflector_out_cell))

control_rod_radius = openmc.ZCylinder(r=0.915)
control_clad_inner_radius = openmc.ZCylinder(r=1.0415)
control_clad_outer_radius = openmc.ZCylinder(r=1.1412)

control_rod_region = (-control_rod_radius & +sleeve_max_z & -axial_reflector_max_z)
control_gap_region = (+control_rod_radius & -control_clad_inner_radius &
                      +sleeve_max_z & -axial_reflector_max_z)
control_clad_region = (+control_clad_inner_radius & -control_clad_outer_radius &
                       +gas_plenum_min_z & -axial_reflector_max_z)
follower_region = (-control_clad_outer_radius & +gas_plenum_min_z &
                   -sleeve_max_z)
control_outside_region = (+control_clad_outer_radius &
                          +gas_plenum_min_z & -axial_reflector_max_z)

control_rod_cell = openmc.Cell(fill=boron_carbide, region=control_rod_region)
control_gap_cell = openmc.Cell(fill=helium, region=control_gap_region)
control_clad_cell = openmc.Cell(fill=em10_steel, region=control_clad_region)
follower_cell = openmc.Cell(fill=reflector_material, region=follower_region)
control_surroundings_cell = openmc.Cell(fill=sodium, region=control_outside_region)

control_universe = openmc.Universe(cells=(control_rod_cell, control_gap_cell,
                                          control_clad_cell, follower_cell,
                                          control_surroundings_cell))

inner_lattice = openmc.HexLattice()
inner_lattice.center = (0., 0.)
inner_lattice.pitch = (21.08/17,)
inner_lattice.orientation = 'x'
inner_lattice.outer = sodium_universe

inner_ring_one = [inner_universe] * 48
inner_ring_two = [inner_universe] * 42
inner_ring_three = [inner_universe] * 36
inner_ring_four = [inner_universe] * 30
inner_ring_five = [inner_universe] * 24
inner_ring_six = [inner_universe] * 18
inner_ring_seven = [inner_universe] * 12
inner_ring_eight = [inner_universe] * 6
inner_ring_nine = [inner_universe]

inner_lattice.universes = [inner_ring_one, inner_ring_two, inner_ring_three,
                           inner_ring_four, inner_ring_five, inner_ring_six,
                           inner_ring_seven, inner_ring_eight, inner_ring_nine]

inner_prism = openmc.model.hexagonal_prism(edge_length=11.65092843, orientation='x')
wrapper_tube_inner = openmc.model.hexagonal_prism(edge_length=11.91073605,
                                                        orientation='x')
wrapper_tube_region_inner = (~inner_prism & wrapper_tube_inner & 
                             +gas_plenum_min_z & -axial_reflector_max_z)
outside_wrapper_tube_inner = (~wrapper_tube_inner & -axial_reflector_max_z & 
                              +gas_plenum_min_z)

main_in_assembly_cell = openmc.Cell(fill=inner_lattice,
                                    region=inner_prism & -axial_reflector_max_z & 
                                    +gas_plenum_min_z)

wrapper_tube_inner_cell = openmc.Cell(fill=em10_steel, region=wrapper_tube_region_inner)
wrapper_tube_inner_outside_cell = openmc.Cell(fill=sodium,
                                              region=outside_wrapper_tube_inner)
inner_assembly_universe = openmc.Universe(cells=[main_in_assembly_cell, 
                                                 wrapper_tube_inner_cell,
                                                 wrapper_tube_inner_outside_cell])

outer_lattice = openmc.HexLattice()
outer_lattice.center = (0., 0.)
outer_lattice.pitch = (21.08/17,)
outer_lattice.orientation = 'x'
outer_lattice.outer = sodium_universe

out_ring_zero = [outer_universe] * 54
outer_ring_one = [outer_universe] * 48
outer_ring_two = [outer_universe] * 42
outer_ring_three = [outer_universe] * 36
outer_ring_four = [outer_universe] * 30
outer_ring_five = [outer_universe] * 24
outer_ring_six = [outer_universe] * 18
outer_ring_seven = [outer_universe] * 12
outer_ring_eight = [outer_universe] * 6
outer_ring_nine = [outer_universe]

outer_lattice.universes = [outer_ring_one, outer_ring_two, outer_ring_three,
                           outer_ring_four, outer_ring_five, outer_ring_six,
                           outer_ring_seven, outer_ring_eight, outer_ring_nine]

outer_prism = openmc.model.hexagonal_prism(edge_length=11.65092843, orientation='x')
wrapper_tube_outer = openmc.model.hexagonal_prism(edge_length=11.91073605,
                                                        orientation='x')
wrapper_tube_region_outer = (~outer_prism & wrapper_tube_outer & 
                             +gas_plenum_min_z & -axial_reflector_max_z)
outside_wrapper_tube_outer = (~wrapper_tube_outer & -axial_reflector_max_z & 
                              +gas_plenum_min_z)

main_out_assembly_cell = openmc.Cell(fill=outer_lattice,
                                    region=outer_prism & -axial_reflector_max_z & 
                                    +gas_plenum_min_z)
wrapper_tube_outer_cell = openmc.Cell(fill=em10_steel, region=wrapper_tube_region_outer)
wrapper_tube_outer_outside_cell = openmc.Cell(fill=sodium,
                                              region=outside_wrapper_tube_outer)
outer_assembly_universe = openmc.Universe(cells=[main_out_assembly_cell,
                                                 wrapper_tube_outer_cell,
                                                 wrapper_tube_outer_outside_cell])

reflector_lattice = openmc.HexLattice()
reflector_lattice.center = (0., 0.)
reflector_lattice.pitch = (4.315,)
reflector_lattice.orientation = 'x'
reflector_lattice.outer = sodium_universe

reflector_inner_ring_one = [reflector_universe] * 12
reflector_inner_ring_two = [reflector_universe] * 6
reflector_inner_ring_three = [reflector_universe]

reflector_lattice.universes = [reflector_inner_ring_one, reflector_inner_ring_two, reflector_inner_ring_three]
reflector_prism = openmc.model.hexagonal_prism(edge_length=11.65092843, orientation='x')
reflector_wrapper =openmc.model.hexagonal_prism(edge_length=11.91073605, orientation='x')

wrapper_tube_region_ref = (~reflector_prism & reflector_wrapper & 
                             +gas_plenum_min_z & -axial_reflector_max_z)
outside_wrapper_tube_ref = (~reflector_wrapper & -axial_reflector_max_z &
                              +gas_plenum_min_z)

wrapper_tube_ref_cell = openmc.Cell(fill=em10_steel, region=wrapper_tube_region_ref)
wrapper_tube_outside_ref_cell = openmc.Cell(fill=sodium,
                                              region=outside_wrapper_tube_ref)
main_reflector_assembly_cell = openmc.Cell(fill=reflector_lattice,
                                    region=reflector_prism & -axial_reflector_max_z &
                                    +gas_plenum_min_z)
reflector_assembly_universe = openmc.Universe(cells=[main_reflector_assembly_cell,
                                                     wrapper_tube_ref_cell,
                                                     wrapper_tube_outside_ref_cell])

control_lattice = openmc.HexLattice()
control_lattice.center = (0., 0.)
control_lattice.pitch = (2.43,)
control_lattice.orientation = 'x'
control_lattice.outer = sodium_universe
control_ring_one = [control_universe] * 18
control_ring_two = [control_universe] * 12
control_ring_three = [control_universe] * 6
control_ring_four = [control_universe]

control_lattice.universes = [control_ring_one, control_ring_two, control_ring_three, control_ring_four]
control_wrapper_1 = openmc.hexagonal_prism(edge_length=8.775724092, orientation='x')
control_wrapper_2 = openmc.hexagonal_prism(edge_length=9.006664199, orientation='x')
control_wrapper_3 = openmc.hexagonal_prism(edge_length=11.65092843, orientation='x')
control_wrapper_4 = openmc.hexagonal_prism(edge_length=11.91073605, orientation='x')

control_assembly_cell = openmc.Cell(fill=control_lattice, region=control_wrapper_1
                                    & +gas_plenum_min_z & -axial_reflector_max_z)
control_wrapper_1_cell = openmc.Cell(fill=em10_steel, region = ~control_wrapper_1
                                     & control_wrapper_2 & +gas_plenum_min_z
                                     & -axial_reflector_max_z)
control_wrapper_gap_cell = openmc.Cell(fill=sodium, region=~control_wrapper_2
                                       & control_wrapper_3 & +gas_plenum_min_z
                                       & -axial_reflector_max_z)
control_wrapper_2_cell = openmc.Cell(fill=em10_steel, region = ~control_wrapper_3
                                     & control_wrapper_4 & +gas_plenum_min_z
                                     & -axial_reflector_max_z)
control_outside_cell = openmc.Cell(fill=sodium, region=~control_wrapper_4 & +gas_plenum_min_z
                                   & -axial_reflector_max_z)

control_assembly_universe = openmc.Universe(cells=[control_assembly_cell,
                                                   control_wrapper_1_cell,
                                                   control_wrapper_gap_cell,
                                                   control_wrapper_2_cell,
                                                   control_outside_cell])

main_lattice = openmc.HexLattice()
main_lattice.center = (0., 0.)
main_lattice.pitch = (21.08,)
main_lattice.orientation = 'y'
main_lattice.outer = sodium_universe

reflector_ring_one = [reflector_assembly_universe] * 96
reflector_ring_two = [reflector_assembly_universe] * 90
reflector_ring_three = [reflector_assembly_universe] * 84
reflector_ring_four = ([reflector_assembly_universe] * 5 + [outer_assembly_universe] * 4 + 
                       [reflector_assembly_universe] * 4) * 6
reflector_ring_five = ([reflector_assembly_universe] + [outer_assembly_universe] * 11) * 6
out_ring_one = [outer_assembly_universe] * 66
out_ring_two = ([outer_assembly_universe] * 5 + [control_assembly_universe] + 
                ([outer_assembly_universe] * 9 + [control_assembly_universe]) * 5 + [outer_assembly_universe] * 4)
out_ring_three = ([outer_assembly_universe] * 2 + ([control_assembly_universe] + [inner_assembly_universe] * 4 +
                  [control_assembly_universe] + [outer_assembly_universe] * 3) * 5 +
                  [control_assembly_universe] + [inner_assembly_universe] * 4  
                  + [control_assembly_universe] + [outer_assembly_universe])
in_ring_one = [inner_assembly_universe] * 48
in_ring_two = [inner_assembly_universe] * 42
in_ring_three = ([inner_assembly_universe]*3 + [sodium_prism_universe]) * 9
in_ring_four = [inner_assembly_universe] * 30
in_ring_five = [inner_assembly_universe] * 24
in_ring_six = ([control_assembly_universe] + [inner_assembly_universe] * 2) * 6
in_ring_seven = [inner_assembly_universe] * 12
in_ring_eight = [inner_assembly_universe] * 6
in_ring_nine = [reflector_assembly_universe] * 1

main_lattice.universes = [reflector_ring_one, reflector_ring_two, reflector_ring_three,
                          reflector_ring_four, reflector_ring_five,
                          out_ring_one,out_ring_two,out_ring_three,in_ring_one,in_ring_two,
                          in_ring_three,in_ring_four,in_ring_five,in_ring_six,
                          in_ring_seven,in_ring_eight,in_ring_nine]

core_surface = openmc.model.hexagonal_prism(edge_length=347.82, boundary_type='vacuum')
core = openmc.Cell(fill=main_lattice,
                   region = core_surface & -axial_reflector_max_z & 
                   +gas_plenum_min_z)
outside_core = openmc.Cell(fill=sodium,
                           region = ~core_surface & -axial_reflector_max_z & 
                           +gas_plenum_min_z)
core_universe = openmc.Universe(cells=[core, outside_core])
geom = openmc.Geometry(inner_assembly_universe)
geom.export_to_xml()





point = openmc.stats.Point((0, 0, 0))
source = openmc.Source(space=point)
settings = openmc.Settings()
settings.source = source
settings.batches = 100
settings.inactive = 10
settings.particles = 1000
settings.export_to_xml()
#openmc.run()

#depletion calculation

inner_fuel.volume = np.pi * ((0.943/2) ** 2) * 217 * 225
outer_fuel.volume = np.pi * ((0.943/2) ** 2) * 217 * 228

chain = openmc.deplete.Chain.from_xml("/home/mpf/support_files/chain_casl_sfr.xml")
model = openmc.Model(geometry=geom, settings=settings, materials=materials_file)
operator = openmc.deplete.CoupledOperator(model, "/home/mpf/support_files/chain_casl_sfr.xml", diff_burnable_mats=True)

power = 16000000

time_steps = [30] * 2
integrator = openmc.deplete.PredictorIntegrator(operator, time_steps,
                                                power, timestep_units = 'd')
integrator.integrate()
2

Hi @marti. All you should need to do is to call:

geom.determine_paths()

after your geom variable has been created.

3

Hi Paul, thanks for the quick answer. I added geom.determine_paths() after I define geom and it still returns the same error.

4

@marti It appears that the problem is as follows:

  • When it tries to differentiate burnable materials, it’s finding that outer_fuel doesn’t have a number of instances determined (which should usually happen with geom.determine_paths())
  • Your Geometry object has a root universe of inner_assembly_universe
  • However, nowhere in inner_assembly_universe is the outer_fuel material, so the call to geom.determine_paths() doesn’t actually count the number of instances for that material.
  • You’ve explicitly given a list of materials to the Model class that includes outer_fuel, and this ends up confusing the diff_burnable_mats option.

To fix this problem, simply omit the list of materials when you’re creating a Model:

model = openmc.Model(geometry=geom, settings=settings)

Also, with that in place, you don’t need to explicitly call geom.determine_paths() as that’s already done under the hood – sorry for misleading you before.

1 Like
5

Hi Paul, when I do that I get the error:
DataError: Cross sections were not specified in Model.materials and the OPENMC_CROSS_SECTIONS environment variable is not set.

This is probably because I haven’t actually set an OPENMC_CROSS_SECTIONS. I am using a MacBook and I’m not sure how to do that, so I downloaded the ENDFB71 cross sections file.

6

Hi again @marti. You can set environment variables either directly from a terminal (export OPENMC_CROSS_SECTIONS=...) or it may be more convenient to set them permanently in your terminal profile. See here for further instructions on that.

7

Hi,

I’m coming across a similar problem. I’m creating a model by using geometry and materials brought in using the ‘.from_xml’ functions, then passing the universe to a model where the universe is a Geometry object. When I include the materials in the model parameters, I get the same error in the comments above, even after adding the determine_paths() function in:

ValueError: Number of material instances have not been determined. Call the Geometry.determine_paths() method.

However, when I omit the materials from the model, I get a different error:

KeyError: 'No material exists with ID=9.'
[WARNING] yaksa: 1 leaked handle pool objects

In the materials list I am importing in, there is no material with ID of 9. I tried to fill in with some empty material instances, and also to manually change the IDs to be consecutive, and each time it seems to be seeking an ID that doesn’t exist.

Any help appreciated! Main bits of code below:

if (run_depletion) : 
    #### DEPLETION CHAIN ###
    diff_burnable_flag = True
    diff_volume_flag = 'divide equally'
    deplete.pool.USE_MULTIPROCESSING = False # Only set true if running on head node, not prod nodes
    chain = openmc.deplete.Chain.from_xml(chain_file)
    procs_per_job = 6
    
    #universe.determine_paths()
    model = openmc.model.Model(geometry = universe, settings= settings)

    operator = deplete.CoupledOperator(model, chain_file=chain_file, diff_burnable_mats=diff_burnable_flag, diff_volume_method = diff_volume_flag)
                                           
    operator.output_dir = os.getcwd()
    
    ts_days = [1, 1, 63, 300, 730, 1095, 1460]
    
    integrator = deplete.PredictorIntegrator(operator, 
                                            ts_days, 
                                            power=reactor.power, 
                                            timestep_units='d',)
    integrator.integrate(output=False)