"ERROR: No fission sites banked on MPI rank 0" using DAGMC geometry

User Support
1

Hi all,

I have made a model of a reactor in CAD and I am following cubit/dagmc method to generated the .h5m input file.
The developed model consists of a fuel element coated with cladding material, while it is surrounded by water. For the development of the geometry and .h5m file generation I have followed the available DAGMC and Cubit documentation and methodology. To import the DAGMC geometry

import openmc
import os 
import openmc.stats
os.system('rm *.xml *.h5')  # deletes previous files

# Materials Definition
materials = openmc.Materials()
#Cladding
cladding = openmc.Material(name='Cladding')
cladding.set_density('g/cm3', 6.55)
cladding.add_element('Sn', 0.014  , 'wo')
cladding.add_element('Fe', 0.00165, 'wo')
cladding.add_element('Cr', 0.001  , 'wo')
cladding.add_element('Zr', 0.98335, 'wo')
# Demineralized Water
water = openmc.Material(name='Water')
water.add_nuclide('H1', 2.0)
water.add_nuclide('O16', 1.0)
water.set_density('g/cm3', 0.99825)
water.add_s_alpha_beta('c_H_in_H2O')
# fuel
fuel = openmc.Material(name='fuel')
fuel.set_density('g/cm3', 10.29769)
fuel.add_element('U', 1., enrichment=2.4)
fuel.add_element('O', 2.)

materials_list = openmc.Materials([cladding, water,fuel])
materials_list.export_to_xml()

# Import Geometry
dag_universe = openmc.DAGMCUniverse('Test_model.h5m')
#Graveyard Surface
vac_surf = openmc.Sphere(r=300, surface_id=99999, boundary_type="vacuum")
region = -vac_surf
cont_cell = openmc.Cell(cell_id=9999, region=region, fill=dag_universe)
geometry = openmc.Geometry(root=[cont_cell])
geometry.export_to_xml()

sources = openmc.Source()
sources.space = openmc.stats.Point((-70, -16.8, 32))
sources.angle = openmc.stats.Isotropic()

model = openmc.Model()
model.geometry = geometry
model.materials = materials_list
model.settings.source = sources
model.settings.batches = 100
model.settings.inactive=10
model.settings.particles = 100000
model.export_to_model_xml()

Evaluating the imported geometry through Openmc-plotter, I can see that all components have been assign to individual cells and materials seem to be assigned correctly (see attached picture).

Test_model_capture
Test_model_capture419×806 7.94 KB

However, when I try to run a keff calculation, I get the following error:

Minimum neutron data temperature: 294 K
 Maximum neutron data temperature: 294 K
 Preparing distributed cell instances...
 Writing summary.h5 file...
 Maximum neutron transport energy: 20000000 eV for Sn112
 Initializing source particles...

 ====================>     K EIGENVALUE SIMULATION     <====================

  Bat./Gen.      k            Average k
  =========   ========   ====================
 ERROR: No fission sites banked on MPI rank 0
--------------------------------------------------------------------------
MPI_ABORT was invoked on rank 0 in communicator MPI_COMM_WORLD
  Proc: [[54643,0],0]
  Errorcode: -1

NOTE: invoking MPI_ABORT causes Open MPI to kill all MPI processes.
You may or may not see output from other processes, depending on
exactly when Open MPI kills them.
--------------------------------------------------------------------------

Despite the fact that openmc-plotter recognize the fuel cell filled with fuel material, OpenMC does not recognize any fissionable material.
I am using the OpenMC version 0.15.0 on a Windows 10 wsl environment.
Any feedback/thought on what might be the reason of this error would be much appreciated.

Thank you in advance.

2

Hi VTheos, welcome to openmc community,
Regarding your case, have you tried to move your source point directly into the fuel cell material of your model?
I see that you use (-70, -16.8, 32) in your input
openmc.stats.Point((-70, -16.8, 32))
which I think is in the water from your model figure.

3

No fission can be sustained for the inactive batches calculation if your source is located in the current position. It should be moved to the fuel zone,

Best
Juan