# Material specified in cell not not found

**URL:** <https://openmc.discourse.group/t/material-specified-in-cell-not-not-found/5117>\
**Category:** User Support\
**Created:** [January 9, 2025, 1:15pm UTC](https://openmc.discourse.group/t/material-specified-in-cell-not-not-found/5117 "2025-01-09T13:15:47Z")\
**Posts on this page:** 1\
**Page:** 1

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**Author:** ![AsukuA](https://avatars.discourse-cdn.com/v4/letter/a/49beb7/32.png) [@AsukuA](https://openmc.discourse.group/u/AsukuA)\
**Post date:** [January 9, 2025, 1:15pm UTC](https://openmc.discourse.group/t/material-specified-in-cell-not-not-found/5117/1 "2025-01-09T13:15:47Z")

</div>

Hello,  
I am relatively new to OpenMC. I created a relatively complex reactor model using OpenMc. I defined my materials and made sure those were the materials i specified in all cells. However, when i run openmc i get this error message.

" Reading settings XML file…  
Reading cross sections XML file…  
Reading materials XML file…  
Reading geometry XML file…  
ERROR: Could not find material 86 specified on cell 17818  
Abort(-1) on node 0 (rank 0 in comm 0): application called MPI\_Abort(MPI\_COMM\_WORLD, -1) - process 0

this is a snippet from the code

import openmc  
import numpy as np  
import matplotlib.pyplot as plt  
import matplotlib.image as mpimg  
openmc.config.cross\_sections = ‘/openmc/cross-sections/cross\_sections.xml’

# Define Materials

# CR meat made of cadmium

cd = openmc.Material(name=‘control cadmium’)  
cd.set\_density(‘g/cm3’, 8.65)  
cd.add\_element(‘Cd’, 1.0)

# ss304 CR clad, Reactivty adjuster rod clad

ss304 = openmc.Material(name=’ stainless steel ss304’)  
ss304.set\_density(‘g/cm3’, 7.8)  
ss304.add\_element(‘Fe’, 0.70845, ‘wo’)  
ss304.add\_element(‘Cr’, 0.1800, ‘wo’)  
ss304.add\_element(‘Ni’, 0.0800, ‘wo’)  
ss304.add\_element(‘Mn’, 0.0200, ‘wo’)  
ss304.add\_element(‘S’, 0.0100, ‘wo’)  
ss304.add\_element(‘C’, 0.0008, ‘wo’)  
ss304.add\_element(‘P’, 0.00045, ‘wo’)  
ss304.add\_element(‘Si’, 0.00030, ‘wo’)

# Active Fuel Material

uo2 = openmc.Material(name=‘active fuel uo2’)  
uo2.set\_density(‘g/cm3’, 10.66)  
uo2.add\_nuclide(‘U235’, 0.114313381, ‘wo’)  
uo2.add\_nuclide(‘U238’, 0.761063318, ‘wo’)  
uo2.add\_nuclide(‘U234’, 0.001758667, ‘wo’)  
uo2.add\_nuclide(‘U236’, 0.002198334, ‘wo’)  
uo2.add\_element(‘O’, 0.11839, ‘wo’)  
uo2.add\_element(‘Al’, 0.000250000000, ‘wo’)  
uo2.add\_element(‘C’, 0.000100000000, ‘wo’)  
uo2.add\_element(‘Ca’, 0.000100000000, ‘wo’)

# Instantiate a Materials collection and export to xml

materials\_file = openmc.Materials([cd, uo2, ss304])

# Export materials to XML

materials\_file.export\_to\_xml()

#Geometry definitions

# Define Cylindrical Surfaces with Boundary Conditions

fuel\_outer\_radius = openmc.ZCylinder(r=0.215, boundary\_type=‘reflective’)  
clad\_outer\_radius = openmc.ZCylinder(r=0.275, boundary\_type=‘reflective’)

#defining planes  
uo2\_meat\_max = openmc.ZPlane(z0=11.50, boundary\_type=‘reflective’)  
uo2\_meat\_min = openmc.ZPlane(z0=-11.50, boundary\_type=‘reflective’)  
zr4\_clad\_max = openmc.ZPlane(z0=12.40, boundary\_type=‘reflective’)  
zr4\_clad\_min = openmc.ZPlane(z0=-12.40, boundary\_type=‘reflective’)

# Create a Universe to encapsulate the fuel rod

fuel\_universe = openmc.Universe(name=‘uo2 fuel Universe’)

# Create uO2 cell

uo2\_cell = openmc.Cell(name=‘uo2’)  
uo2\_cell.fill = uo2  
uo2\_cell.region = -fuel\_outer\_radius & +uo2\_meat\_min & -uo2\_meat\_max  
fuel\_universe.add\_cell(uo2\_cell)

Please what could be the problem. Please note that this is not the complete code. It is quite long.