Hi!
I have run a depletion simulation where I know the source rate at the end of depletion. I then put power to 0, and run again for 8 seconds using the openmc.deplete module. After the decay I use the new fuel composition in another fixed source simulation with a new geometry. Here I would like to look at neutron flux, but how do I find the new source rate in order to normalize the results?
Hi ebba, welcome to the openmc community.
Since you said, “After the decay, I use the new fuel composition in another fixed source simulation with a new geometry”, then the normalization will come from the source strength you used for the fixed source simulation. If your model was for a 10^10 n/sec neutron source, then it would become your fixed source strength, your normalizations factor.
Hi,
Thank you for your response, and sorry for being a bit unclear. The depletion calculation is an eigenvalue simulation. Since the source is left to decay I was assuming I couldn’t use the original source strength.
Hi ebba, just to make sure. When you do the eigenvalues problem with a specific power for your model, then the source strength will be correlated to the power. I think you have read the 8.3. Normalization of Tally Results at the documentation 8. Specifying Tallies — OpenMC Documentation
But when you use the material produced during some eigenvalues problem to be used in a fixed source, then you need to know the source strength you used in your fixed source, because it was fixed in your fixed source problem.
If you think that there will be some neutrons being emitted caused by the decay of some isotope, then you will need to convert it from the activity of the isotope (neutron emitter) to the number of neutrons emitted. But if you used another neutron source, i.e. accelerator-based (ADS), then you will need to know the source strength of this neutron source from ADS. The same thing happens if you want to do the photon transport, you will need to know the photon source strength that might come from the decay isotopes.
I haven’t done this type of calculation, but from what I know, all tally scores in openmc are normalized per source particle simulated, so in fixed source, the source strength will be an input that comes from our description, either it only comes from the decay products or it also came from an external source.
If you look at this examples of shutdown dose rate, you can see that the source strength was defined as all the photon from decay because this example want to focused on it, so the source unit is photon/sec. activated_material.get_decay_photon_energy