Accessing XS database

User Support
1

Hello everybody,

I’m trying to get the files 94239.71c, 94239.72c, etc. but I did not find them.

I tried using the ENDF database (https://www-nds.iaea.org/exfor/endf.htm) but I could not succeed finding them neither.

Where can I download the cross section database?

Regards,
Mutaito.

2

Those cross sections are distributed from the NNDC. Note that the OpenMC source distribution includes a script (data/get_nndc_data.py) that will download, unpack, and set-up the cross sections for you as described here:

http://openmc.readthedocs.io/en/latest/usersguide/install.html#using-endf-b-vii-1-cross-sections-from-nndc

Best,
Paul

3

Hi Paul,

Thanks for your quick reply.
First I downloaded the ENDF-B-VII.1-neutron-293.6K.tar.gz file and I added the data/cross_sections.xml file, pointing to this folder.

I used the “stubs” example from the openMC workshop to test it.

After executing it the error was the following:

Reading settings XML file…
Reading cross sections XML file…
Reading geometry XML file…
WARNING: The use of ‘surfaces’ is deprecated and will be disallowed in a future
release. Use ‘region’ instead. The openmc-update-inputs utility can
be used to automatically update geometry.xml files.
Reading materials XML file…
Reading tallies XML file…
WARNING: Mesh type ‘rectangular’ is deprecated. Please use ‘regular’ instead.
Building neighboring cells lists for each surface…
Loading ACE cross section table: 5010.71c
At line 331 of file /home/mutaito/Software/openmc/src/ace.F90
Fortran runtime error: Non-existing record number

I saw your previous answer (https://groups.google.com/forum/#!topic/openmc-users/7nMAe9ttWGM) but did not work for me. So, I used the python script to prepare the input data. After generating it I copied the cross sections database (in binary format) generated in the “nndc” folder as well as the cross_sections.xml.

Then I executed openmc

I’m getting the error:

Reading settings XML file…
Reading cross sections XML file…
Reading geometry XML file…
WARNING: The use of ‘surfaces’ is deprecated and will be disallowed in a future
release. Use ‘region’ instead. The openmc-update-inputs utility can
be used to automatically update geometry.xml files.
Reading materials XML file…
ERROR: Could not find S(a,b) table lwtr.15t in cross_sections.xml file!
ERROR STOP

Do you know what may be the problem?

Regards,
Mutaito.

4

The workshop examples that you see are quite outdated and as such some of the input options have changed. You might be better off looking at the examples/ directory from the OpenMC repository itself or the material from our more recent PHYSOR workshop which stressed use of the Python API.

To address your actual problem though – in that materials.xml file you should change 15t to 71t. If you look at the cross_sections.xml file, towards the bottom you’ll see that all the S(alpha,beta) tables have extension 71t.

Best,
Paul