# OpenMC aborted unexpectedly

**URL:** <https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181>\
**Category:** User Support\
**Created:** [July 5, 2023, 7:35pm UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181 "2023-07-05T19:35:34Z")\
**Posts on this page:** 5\
**Page:** 1

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**Author:** ![franpat91](https://avatars.discourse-cdn.com/v4/letter/f/df705f/32.png) [@franpat91](https://openmc.discourse.group/u/franpat91)\
**Post date:** [July 5, 2023, 7:35pm UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181/1 "2023-07-05T19:35:34Z")

</div>

Hello everyone,  
I’m currently trying to run an OpenMC program using MGXS to generate multi-group efficient sections. It consists firstly of calculating multi-group cross-sections using a mgxs library, and then using these calculated cross-sections. These are determined for each zone of my system. This latter consists of a ring of material (tungsten) surrounded by a vacuum (divided into two zones (the inner part of the ring and the outer)). I should point out that the outer walls (those of the red part) are of the ‘vacuum’ type.

My goal is to tally the “scatter” reaction rate depending on the scattering angle.  
However, strangely, the program stops working (raise RuntimeError(error\_msg)) when the number of simulated particles (nb\_particles) becomes too large (for example nb\_particles=1000000 and nb\_batches=10 leads to the bug)… This problem only appears on the multigroup part (the continuous part works fine).

I ran the program in debug mode (as indicated here [How to turn on debug](https://openmc.discourse.group/t/how-to-turn-on-debug/800)). What I obtained is shown in the photo.

I thought the problem must be with the way I was allocating angles in multigroup. So I tested “histogram” and “Legendre” (at different orders) for mgxs\_lib.scatter\_format, but to no avail. This might also be a problem with the source as I have got a warning "/usr/local/lib/python3.10/dist-packages/openmc/source.py:388: FutureWarning: This class is deprecated in favor of ‘IndependentSource’ "

Do you have any idea why this might be?  
Thanks for your help  
Best  
PS:the program is just after the photos

 ![Capture d'écran 2023-07-04 164527](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/4/4b07f71c6f262fd2be247af9077a042851f3e5a5.png)

```auto
#les import importants 
import openmc 
import sys
import os
import matplotlib.pyplot as plt
import numpy as np
import math

def run_multigroupe(energy_groups, liste_SS316, rayon, bord, ordre_remplissage, nb_batches, nb_particles, nb_cutoff): 
    
    #beginning of the continuous where cross sections will be computed
    
    if (len(rayon)!=len(ordre_remplissage)): 
        print("Please make sure rayon and ordre_remplissage are of the same size")
        sys.exit()
        
    nom_des_fichiers=["geometry.xml", "settings.xml", "summary.h5", "tallies.out", "tallies.xml"]
    for loop in range (len(nom_des_fichiers)):
        if os.path.exists(nom_des_fichiers[loop]):
            os.remove(nom_des_fichiers[loop])

    sphere=[openmc.Sphere(r=rayon[j]) for j in range (len(rayon))]
    plan_z_up=openmc.ZPlane(z0=bord, boundary_type = 'vacuum')
    plan_z_down=openmc.ZPlane(z0=-bord, boundary_type = 'vacuum')
    plan_x_up=openmc.XPlane(x0=bord, boundary_type = 'vacuum')
    plan_x_down=openmc.XPlane(x0=-bord, boundary_type = 'vacuum')
    plan_y_up=openmc.YPlane(y0=bord, boundary_type = 'vacuum')
    plan_y_down=openmc.YPlane(y0=-bord, boundary_type = 'vacuum')

    cells=[]
    for j in range (len(rayon)):
        if (ordre_remplissage[j]!="void"):
            cell=openmc.Cell(fill=ordre_remplissage[j])
        else:
            cell=openmc.Cell()
        if (j==0):
            region=-sphere[j]
        else :
            region=-sphere[j] & +sphere[j-1]
        cell.region=region
        cells.append(cell)

    cell_fin=openmc.Cell() 
    cell_fin.region=+sphere[len(sphere)-1] & -plan_z_up & -plan_x_up & -plan_y_up & +plan_x_down & +plan_y_down & +plan_z_down
    
    cells.append(cell_fin)

    universe=openmc.Universe(cells=cells)

    plot=universe.plot (width=(900, 900), color_by="cell", colors={cells[0]:"green", cells[1]:"yellow", cells[2]:"red"})

    geom=openmc.Geometry(universe)
    geom.export_to_xml()

    source=openmc.Source() 
    source.space=openmc.stats.Point((0., 0., 0.))
    source.angle=openmc.stats.Isotropic() 
    source.energy = openmc.stats.Uniform(a=14.06e6, b=14.07e6)

    
    groups = openmc.mgxs.EnergyGroups(energy_groups)
    mgxs_lib = openmc.mgxs.Library(geom)
    mgxs_lib.energy_groups = groups
    mgxs_lib.mgxs_types = ['total', 'absorption', 'nu-fission', 'fission',
                       'nu-scatter matrix', 'multiplicity matrix', 'chi']
    mgxs_lib.domain_type = "cell"
    mgxs_lib.domains = geom.get_all_cells().values()
    mgxs_lib.by_nuclide = False
    mgxs_lib.scatter_format ="legendre"
    #mgxs_lib.histogram_bins=nombre_angle
    mgxs_lib.legendre_order = 1
    mgxs_lib.check_library_for_openmc_mgxs()
    mgxs_lib.build_library()
    tallies=openmc.Tallies()
    mgxs_lib.add_to_tallies_file(tallies, merge=True)
    
    tallies.export_to_xml() 

    settings = openmc.Settings()
    settings.run_mode = 'fixed source'
    settings.batches = nb_batches # Number of batches
    settings.particles = nb_particles # Number of particles
    settings.cutoff = {'energy_neutron': nb_cutoff} # Kill neutrons with energy < 0.001 eV
    settings.source = source

    settings.export_to_xml()

    openmc.run(output="False")
    
    #end of the continuous part#
    
    ce_spfile='calcul_sections_efficaces.h5'
    ce_sumfile='summary_sections_efficaces.h5'
    os.rename('statepoint.{}.h5'.format(nb_batches), ce_spfile)
    os.rename('summary.h5'.format(nb_batches), ce_sumfile)
    sp = openmc.StatePoint(ce_spfile, autolink=False)
    su = openmc.Summary(ce_sumfile)
    sp.link_with_summary(su)

    nom_des_fichiers=["geometry.xml", "settings.xml", "summary.h5", "tallies.out", "tallies.xml"]
    for loop in range (len(nom_des_fichiers)):
        if os.path.exists(nom_des_fichiers[loop]):
            os.remove(nom_des_fichiers[loop])    

    mgxs_lib.load_from_statepoint(sp)
    mgxs_file = mgxs_lib.create_mg_library(xs_type='macro', xsdata_names=["step{}".format(j) for j in range (len(cells))])
    
    mgxs_file.export_to_hdf5()
    
    materiaux_liste=[]
    for loop in range (len(cells)):
        materiel=openmc.Material(name=("materiau{}".format(loop)), material_id=10+loop)
        materiel.add_macroscopic("step{}".format(loop))
        materiaux_liste.append(materiel)
    materiaux=openmc.Materials((materiaux_liste[0], materiaux_liste[1], materiaux_liste[2]))
    materiaux.cross_sections="mgxs.h5"
    materiaux.export_to_xml()
    
    sphere=[openmc.Sphere(r=rayon[j]) for j in range (len(rayon))]
    plan_z_up=openmc.ZPlane(z0=bord, boundary_type = 'vacuum')
    plan_z_down=openmc.ZPlane(z0=-bord, boundary_type = 'vacuum')
    plan_x_up=openmc.XPlane(x0=bord, boundary_type = 'vacuum')
    plan_x_down=openmc.XPlane(x0=-bord, boundary_type = 'vacuum')
    plan_y_up=openmc.YPlane(y0=bord, boundary_type = 'vacuum')
    plan_y_down=openmc.YPlane(y0=-bord, boundary_type = 'vacuum')

    cells=[]
    for j in range (len(rayon)):
        cell=openmc.Cell(fill=materiaux_liste[j])
        if (j==0):
            region=-sphere[j]
        else :
            region=-sphere[j] & +sphere[j-1]
        cell.region=region
        cells.append(cell)

    cell_fin=openmc.Cell() 
    cell_fin.region=+sphere[len(sphere)-1] & -plan_z_up & -plan_x_up & -plan_y_up & +plan_x_down & +plan_y_down & +plan_z_down
    
    cells.append(cell_fin)

    universe=openmc.Universe(cells=cells)

    #plot=universe.plot (width=(900, 900), color_by="material", colors={FLIBE:"green", SS316:"yellow", tungsten:"red", water:'blue'})

    geom=openmc.Geometry(universe)
    geom.export_to_xml()
    
    angular_filter=openmc.MuFilter(nombre_angle)
    
    tallies=openmc.Tallies()
    tally=openmc.Tally()
    tally.name=str(loop)
    tally.filters=[angular_filter]
    tally.scores=["scatter"]
    tallies.append (tally)

    tallies.export_to_xml() 

    source=openmc.Source() 
    source.space=openmc.stats.Point((0., 0., 0.))
    source.angle=openmc.stats.Isotropic() 
    source.energy = openmc.stats.Uniform(a=energy_groups[len(energy_groups)-2], b=energy_groups[len(energy_groups)-1])

    settings = openmc.Settings()
    settings.run_mode = 'fixed source'
    settings.energy_mode = 'multi-group'
    settings.batches = nb_batches # Number of batches
    settings.particles = nb_particles # Number of particles
    settings.cutoff = {'energy_neutron': nb_cutoff} # Kill neutrons with energy < 0.001 eV
    settings.source = source

    settings.export_to_xml()

    openmc.run(output="False")
    os.rename('statepoint.{}.h5'.format(nb_batches), 'reference.h5')

liste_SS316=[0.07, 2.0, 1.0, 0.05, 0.02, 18.5, 2.5, 13.0, 0] # les fractions en % de la composition du SS316#
#mise en forme des données
liste_SS316=[liste_SS316[j]/100 for j in range (len(liste_SS316))]
liste_SS316.append(1-sum(liste_SS316))

#définition de tous les matériaux#
FLIBE=openmc.Material(name="FLIBE")
FLIBE.add_element("F", 4.)
FLIBE.add_element ("Li", 2.)
FLIBE.add_element ("Be", 1.)
FLIBE.set_density("g/cm3", 1.94)

tungsten=openmc.Material(name="tungsten")
tungsten.add_element ("W", 1.0)
tungsten.set_density("g/cm3", 19.25)

SS316=openmc.Material(name="SS316")
SS316.add_element("C", liste_SS316[0])
SS316.add_element("Mn", liste_SS316[1])
SS316.add_element("Si", liste_SS316[2])
SS316.add_element("P", liste_SS316[3])
SS316.add_element("S", liste_SS316[4])
SS316.add_element("Cr", liste_SS316[5])
SS316.add_element("Mo", liste_SS316[6])
SS316.add_element("Ni", liste_SS316[7])
SS316.add_element("Ti", liste_SS316[8])
SS316.add_element("Fe", liste_SS316[9])
SS316.set_density("g/cm3", 7.99)

water=openmc.Material(name="water")
water.add_element("H", 2.)
water.add_element("O", 1.)
water.set_density("g/cm3", 1.)
water.add_s_alpha_beta('c_H_in_H2O')

materiaux=openmc.Materials([SS316, FLIBE, tungsten, water])
materiaux.export_to_xml()

rayon=[115, 220]
bord=400
ordre_remplissage=["void", tungsten]

nb_batches=10
nb_particles=1000000
nb_cutoff=0.01

nombre_angle=100

energy_groups=[0., 10e6, 14.0e6, 14.07e6]
run_multigroupe(energy_groups, liste_SS316, rayon, bord, ordre_remplissage, nb_batches, nb_particles, nb_cutoff)

```

---

<div class="post-metadata">

**Author:** ![paulromano](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/paulromano/32/486_2.png) [@paulromano](https://openmc.discourse.group/u/paulromano)\
**Post date:** [July 13, 2023, 4:59am UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181/2 "2023-07-13T04:59:39Z")

</div>

> [@franpat91](#):
>
> This might also be a problem with the source as I have got a warning "/usr/local/lib/python3.10/dist-packages/openmc/source.py:388: FutureWarning: This class is deprecated in favor of ‘IndependentSource’ "

This is just due to an update in the development branch of OpenMC. If you change `openmc.Source` to `openmc.IndependentSource`, that warning should go away.

Otherwise, not much comes to mind as far as what could be going wrong.

---

<div class="post-metadata">

**Author:** ![franpat91](https://avatars.discourse-cdn.com/v4/letter/f/df705f/32.png) [@franpat91](https://openmc.discourse.group/u/franpat91)\
**Post date:** [July 14, 2023, 2:03pm UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181/3 "2023-07-14T14:03:45Z")

</div>

Ok, I will try that.  
Thank you very much for your help  
Best

---

<div class="post-metadata">

**Author:** ![Elie\_Mweze](https://avatars.discourse-cdn.com/v4/letter/e/8dc957/32.png) [@Elie\_Mweze](https://openmc.discourse.group/u/Elie_Mweze)\
**Post date:** [April 30, 2024, 1:07pm UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181/4 "2024-04-30T13:07:01Z")

</div>

Hello Mr @paulromano,  
I also have the same problem. How can I change openmc.Source to openmc.IndependentSource?

I need some help. Thanks a lot!

---

<div class="post-metadata">

**Author:** ![paulromano](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/paulromano/32/486_2.png) [@paulromano](https://openmc.discourse.group/u/paulromano)\
**Post date:** [April 30, 2024, 7:47pm UTC](https://openmc.discourse.group/t/openmc-aborted-unexpectedly/3181/5 "2024-04-30T19:47:43Z")

</div>

Just make sure your Python script has:

```Python
source = openmc.IndependentSource(...)

```