I have recently installed the new OpenMC (ver 0.9.0). I have been able to correctly configure the NNDC files. I am able to run eigenvalue calculations from the /examples subdirectory (but not view their tallies).
I am not able to run the python script, ‘openmc-plot-mesh-tally statepoint.***.h5’.
When I run this command, a window opens with the information show in the the subject heading of this post:
Invalid StatePoiint File!
File does not contain mesh tallies!
Can you please tell me what I need to do to view tallies using the XML files? I do not use the Python API, instead editing XML files directly.
I have included a copy of the XML files I am trying to run. Also included are the three output files: statepoint.**.h5, tallies.h5, and tallies.out.
Respectfully,
Shawn Wachter
PS - Note that I recently also had trouble with the neutron temperatures. I placed a previous post asking why the XML files were not working at all, and it turned out that I had to change the neutrons from MeVs to eVs. The previous post can be considered closed.
geometry.xml (4.02 KB)
materials.xml (2.81 KB)
plots.xml (1.92 KB)
settings.xml (1.72 KB)
tallies.xml (1.85 KB)
statepoint.100.h5 (95.3 KB)
summary.h5 (71.7 KB)
tallies.out (166 Bytes)