I have a code that runs openmc for the stochastic volume solver, it runs fine, then assigns those volumes to the appropriate materials. However, when I call integrator to initiate the depletion it gives me an error saying:
" Reading settings XML file…
Reading cross sections XML file…
Reading materials XML file…
ERROR: Could not find element C in cross_sections.xml."
This doesn’t happen for the volume run, it lists all the nuclides in materials.xml and prints the path to them so I’m not sure what’s going on. Is it something weird when you call on it twice in the same script? Or does it has to do with the way depletion requires specific cross section xml formats?
You require a specific cross section file for decay chains, you can either make them or download one from here (OpenMC provides these already ready for use),
I would suggest the CASL simplified one if your application doesn’t require specific isotopes that’s it’s missing, but overall the accuracy is still there for approximately 1/4 the computational time. I would also follow this example for depletion,
If you have already done this and are still getting the same error, please provide your code of your integrator and your cross section declaration.
I remeber some older cross section libraries such as ENDF 7.1 used C or C0 for Carbon instead of having nuclide break downs. I can’t say for sure but it reminds me of that error which can be avoided by downloading the newer ENDF 8 cross sections
and also and ENDF 8 based chain file.
I think I found the issue. It is because I turned on photon transport, because I wanted to simulate the decay photons from the transmutation. Now I need to figure out how to do that.