Hi everyone!
I’ve been struggling to understand exactly the difference between the “flux” tally score and “current” tally score, primarily with respect to if flux keeps track of back scattering/forward scattering neutrons.
In my case, I’m just simulating a simple spherical reactor design where neutrons are being emitted from a point source and enter different reactor components. The OpenMC website says current is only compatible with a surface filter as it is supposed to measure the number of particles per source particle (or neutrons per source neutron in my case) crossing a certain surface in one direction. Flux, however, is defined in OpenMC as particle*cm/source particle, and from the other user support topics I’ve found, it looks like flux measures particles going in all directions.
When I try computing the flux of neutrons in a specific cell, I can get up to hundreds of neutron*cm/source neutron. Whereas if I measure the current at the surface just before or after this cell. It’s usually no more than 1.5 neutrons per source neutron. If I divide the flux by the thickness of the cell to just get units of neutrons per source neutron across some average surface in the cell, I’ll still get up to 10 neutrons per source neutron.
A few questions regarding this setup: Does it make sense to divide the flux in the way that I did to match the units of current? If so, why do I get so many more neutrons per source neutron for flux than I do for current? Is flux taking into account multiple scattering events going in and out of the cell?
To summarize it in one question: What exactly is flux measuring in the way OpenMC defines it?
Thanks,
Marcos