Depletion on Fixed source simulation

User Support
1

Dear OpenMC experts,

Due to my last problem has figured out and the fixed source simulation run out fine, then I couple the depletion into the script but the error occurred

unknown2
unknown21393×963 66 KB

I’m not sure if using the chain_endfb71_sfr_1.xml for running this problem, and maybe it will cause the error due to lack of heating data? What makes this error occurred?

Also, I have plot the flux of my geometry and it comes as in figure below

unknown
unknown1920×1080 259 KB

These fluxes represent the position, and it is not distributed all over the geometry as I expected. Are these transport fluxes are not going to create any fission product in the purple zone except in the shape of a circle right? please corrected me.

Here is my python script running on jupyter notebook and my external source file
ADS_Python.py (8.0 KB)
ads_data.h5 (7.9 KB)

1 Like
2

If you are running fixed source simulations, you need to inform the depletion operator/integrator that this is the case. Make sure you set:

operator = openmc.deplete.Operator(..., normalization_mode="source-rate")

and

integrator = openmc.deplete.PredictorIntegrator(operator, timesteps, source_rates=...)

where source_rates is a list of the source rates for your problem in neutrons/sec. The normal power normalization that OpenMC attempts to do (and relies on energy deposition tallies) only makes sense if you are running a k-eigenvalue calculation, which is not the case here. For a fixed source problem, it’s up to you to decide what the proper source rate is.