ERROR: Nuclear data library does not contain cross sections

Omar1001 · November 25, 2021, 8:43am

Hello!

I am trying to run depletion calculations using materials in different values of temperatures. Those temperatures do not exist in the ENDF7.1 crosse section library. Therefore, I tried to use the Multipole method. However, I end up with the error below. It’s only working when I mix the Nearest or the Interpolation method with the Multipole. This makes me wonder, Is OpenMC using my set of temperatures

or the temperatures in the ENDF7.1?

In the case that OpenMC is using the temperatures in the ENDF7.1, then how can I make it use my materials temperatures or any other temperature? Please.

My second question is related to the transfer of library from ace to h5 format using “openmc.data.IncidentNeutron.from_ace()”. It seems that when I transfer an ace file that contains multiple temperatures data, only the data related to the first temperature is converted to the h5 format. Therefore, I tried to use the method “add_temperature_from_ace()”. But, this required that the data for different temperatures should be placed in separate files, for example:

for suffix in [711, 712, 713, 714, 715, 716]:
u235.add_temperature_from_ace(‘92235.{}nc’.format(suffix))

Then, would someone know how to convert all the ace data at once?
depletion_simple.py (2.5 KB)

AlexandreTrottier · November 26, 2021, 4:32pm

Hello,
Have you tried using both the multipole and the temperature interpolation methods? I have recently run test problems in this manner with the endfb71_hdf5 library. Both data sets are loaded (ACE type and WMP). The interpolation method is used when WMP data are not available.

Shimwell · November 26, 2021, 5:06pm

For an example of adding ace files with a temperature you could take a look at this script that does the processing of JEFF data at different temperatures. We use the add_temperature_from_ace() method that you are using, but perhaps in a slightly different way using tables.

github.com

openmc-dev/data/blob/bcc99743f5c0efa0d7f3821201e04a31f8fd1d0d/convert_jeff32.py#L210

    
      
              name, xs = table.name.split('.')
              table.name = '.'.join((name.strip(digits), xs))
              data = openmc.data.ThermalScattering.from_ace(table)
          
          
    # For each higher temperature, add cross sections to the existing table
              for filename in filenames[1:]:
                  print(f'Adding: {filename}')
                  table = openmc.data.ace.get_table(filename)
                  name, xs = table.name.split('.')
                  table.name = '.'.join((name.strip(digits), xs))
                  data.add_temperature_from_ace(table)
          
          
    # Export HDF5 file
              h5_file = args.destination / f'{data.name}.h5'
              print('Writing {}...'.format(h5_file))
              data.export_to_hdf5(h5_file, 'w', libver=args.libver)
          
          
    # Register with library
              library.register_file(h5_file)
          
          
# Write cross_sections.xml

The GitHub - openmc-dev/data: Collection of scripts for managing data for OpenMC has some nice scripts for processing ace files and endf files into h5 files, if you can’t see the particular one you need then perhaps it can be added.

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ERROR: Nuclear data library does not contain cross sections

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