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1

RuntimeError: Nuclear data library does not contain cross sections for c_Graphite at temperatures that bound 294.0 K.

I’ve been getting this error …
whenever i use the command:
graphite.add_s_alpha_beta(‘c_Graphite’)

i have changed my cross section library to Jeff 3.3 but it still shows the same error…how can i solve this problem?

2

Welcome to the forum @Mashaba

Thermal scattering data for Graphite doesn’t come with the official JEFF-3.3 library. You can change your library to ENDF-7.1 or ENDF-8 which you can download from official library.

1 Like
3

thank you so much! @Pranto

i have tried both…it still shows the same error

4

That’s strange! Ok, What is the output of

grep -rw $OPENMC_CROSS_SECTIONS -e c_Graphite.h5

command?

5

FYI, TSL for Graphite is available at

['1000K',
 '1200K',
 '1600K',
 '2000K',
 '296K',
 '400K',
 '500K',
 '600K',
 '700K',
 '800K']

temperature from both ENDF-7.1 and ENDF-8.

6

Hi @Mashaba, if your graphite temperature is 294.0 K, for that temperature nuclear data doesn’t exist. In that case, you can add settings.temperature = {‘method’: ‘interpolation’} to your Settings to interpolate. More details here.

1 Like
7

@Pranto this command shows syntax error

8

i have done that as well. It still shows the same error

9

If you’re using temperature interpolation, you need data to be available at two temperatures that surround the actual temperature. In this case, your material is at 294 K but the lowest available temperature in the dataset is 296 K. I think your options are either to:

  1. Change the temperature of the material to be slightly higher (296 or 297 K) which should have minimal impact on results, or
  2. Don’t use temperature interpolation at all. The ‘nearest’ temperature method uses a tolerance of 10 K by default, so if the material is at 294 K it should work fine with the 296 K data.
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10

Apparently that problem is solved. I used the nearest option. Thank you so much!

But i am now experiencing a runtime error given below…Is it because of the version i am using?(0.12.1)

RuntimeError Traceback (most recent call last)
in
----> 1 openmc.run()

~/miniconda3/lib/python3.9/site-packages/openmc/executor.py in run(particles, threads, geometry_debug, restart_file, tracks, output, cwd, openmc_exec, mpi_args, event_based)
225 args = mpi_args + args
226
→ 227 _run(args, output, cwd)

~/miniconda3/lib/python3.9/site-packages/openmc/executor.py in _run(args, output, cwd)
36 error_msg = ’ '.join(error_msg.split())
37
—> 38 raise RuntimeError(error_msg)
39
40

RuntimeError: OpenMC aborted unexpectedly.

11

@Mashaba What did you see when you run openmc from the command line?

1 Like
12

i saw this after the openmc.run() command line

              %%%%%%%%%%%%%%%
                       %%%%%%%%%%%%%%%%%%%%%%%%
                    %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                  %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
               %%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%
                                %%%%%%%%%%%%%%%%%%%%%%%%
                                 %%%%%%%%%%%%%%%%%%%%%%%%
             ###############      %%%%%%%%%%%%%%%%%%%%%%%%
            ##################     %%%%%%%%%%%%%%%%%%%%%%%
            ###################     %%%%%%%%%%%%%%%%%%%%%%%
            ####################     %%%%%%%%%%%%%%%%%%%%%%
            #####################     %%%%%%%%%%%%%%%%%%%%%
            ######################     %%%%%%%%%%%%%%%%%%%%
            #######################     %%%%%%%%%%%%%%%%%%
             #######################     %%%%%%%%%%%%%%%%%
             ######################     %%%%%%%%%%%%%%%%%
              ####################     %%%%%%%%%%%%%%%%%
                #################     %%%%%%%%%%%%%%%%%
                 ###############     %%%%%%%%%%%%%%%%
                   ############     %%%%%%%%%%%%%%%
                      ########     %%%%%%%%%%%%%%
                                  %%%%%%%%%%%

               | The OpenMC Monte Carlo Code
     Copyright | 2011-2021 MIT and OpenMC contributors
       License | https://docs.openmc.org/en/latest/license.html
       Version | 0.12.1
      Git SHA1 | d49e6ccb837685e399efdef8b29e46cdc693ccec
     Date/Time | 2021-06-10 08:41:28
OpenMP Threads | 6

Reading settings XML file…
Reading cross sections XML file…
Reading materials XML file…
Reading geometry XML file…
Reading U235 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/U235.h5
Reading U238 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/U238.h5
Reading C12 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/C12.h5
Reading O16 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/O16.h5
Reading C13 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/C13.h5
Reading Si28 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/Si28.h5
Reading Si29 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/Si29.h5
Reading Si30 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/Si30.h5
Reading He3 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/He3.h5
Reading He4 from /home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/He4.h5
Reading c_Graphite from
/home/mashaba/Downloads/endfb80.tar(2)/endfb80_hdf5/c_Graphite.h5
Minimum neutron data temperature: 294.0 K
Maximum neutron data temperature: 294.0 K
Preparing distributed cell instances…
Writing summary.h5 file…
Maximum neutron transport energy: 20000000.0 eV for C13
Initializing source particles…

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

13

After showing ‘initializing source particles’ it says that openmc aborted unexpectedly @Pranto

15

@Mashaba Have you checked overlapping cell definitions? You can do this by running openmc.run(geometry_debug=True)

1 Like
16

If you’re using temperature nearest method, you can set tolerance

settings_file.temperature = {
		 'method'    :'nearest',
		 'tolerance' : 300.0
}
1 Like
17 
<===================

Cell ID No. Overlap Checks
1 244530
2 2357696
3 3145491
4 3596587
5 7385881

There were 0 overlapping cells
woaaahh! it is suddenly giving the outputs with INTERPOLATION COMMAND too!!!.. THANKS A TON!
@Pranto

18

Great @Mashaba

The output here means cell id and number of checks have been performed for each cell. You can adjust initial source particle to get a good coverage.

1 Like
19

and how should i do that…? im not sure if i understand that properly can u provide me a reference?
@Pranto

20

If you run openmc in geometry-debugging mode with fewer particles in the case of complex geometry where two surfaces are close to each other, you’ll see the number of checks quite low. In that case, you can adjust the source particle to get good coverage of each area of geometry.

Don’t worry, you’re fine now, you have simple geometry.

1 Like