Hi,
I have a question on how tally fission Q value, I understand I can tally sth like kappa-fission, however, the unit is in ev per source particle. For example, for a 2D pin cell case, I got a kappa-fission value of 1.08e8 ev per source particle, how could I get the value in MeV?
Thanks for your help.
Tian
Hello! OpenMC probably can’t give you the number you want because by definition, Monte Carlo can only ever give you a shape of the flux, not a magnitude. So it can give you a power distribution that you normalize to the total power you set.
You can define power with several different tallies. I’ve had good success using simply the “fission” tally, but you can also use “kappa-fission” or “fission-q-recoverable” depending how sophisticated the assumption you want. In my experience, I’ve used a known normalization factor of 66945.4 W for a fuel rod based on the power output of an entire reactor divided by the number of rods in the reactor.
For a cool example of changing temperatures and water densities in-memory in OpenMC, check out this notebook: Jupyter Notebook Viewer
Hi,
Thank you for your reply. I understand in the Serpent output file, the FISSE option represent the fission q value. As what you said, if I need sth like this in openMC the only way is to set myself, am I correct?
Thank you!
Hi again, you can get the fission q value in OpenMC using the tally ‘fission-q-recoverable’, however, no Monte Carlo code can provide the power scale, just the power shape. So you have to set the total power you want to scale the output. This is also true for Serpent. The tallies can only give you quantities per source particle, nothing absolute.