Hello everyone
I have a question regarding OpenMC tallies. When computing the total neutron flux in a reflector or anything else region composed of multiple cells, what is the correct definition of “total flux”?
Should it be interpreted as a volume-integrated flux over the entire reflector, or is it acceptable to sum the cell-wise flux values directly?
In published studies, is this quantity typically obtained using volume-weighted averaging of cell tallies or using a mesh-based integration?
Thanks
I would usually understand total flux to be energy-integrated flux in a given volume.
Over an entire reflector, the total neutron flux will vary significantly; as you get further away from the neutron source or core, it will fall off.
It depends on what quantity or what you are looking at whether you’d be interested in the flux close to the core, far from the core or the reflector volume-integrated flux.
hi
I have a question regarding OpenMC tallies. In my reactor model, the moderator/reflector region is divided into many cells, and I also have internal and external irradiation sites.
If I want to report a single neutron flux value for the moderator, what is the correct approach?
Should the flux values from all moderator cells be summed directly, or should the moderator flux be computed as a volume-weighted (or volume-integrated) quantity over the entire moderator region?
Additionally, for the internal and external irradiation sites, is it more appropriate to report the local neutron flux at each irradiation position instead of using a global moderator flux?
If you tally the moderator cells together or volume-weight the distinct cells them, it should be the same result.
I would only tally in the active core length (extent of the fuel meat).
Yes, the neutron flux of the irradiation sites should be tallied and reported locally.
This sounds like a research reactor, they (and others) have significantly varying flux across the core, moderator, reflector.
Flux in one position could be 1e15, in another 1e12
Thank you, this clarifies my confusion.
So, for the moderator/reflector region, I can either tally all moderator cells together or compute a volume-weighted flux, and both approaches should give approximately the same result.
For the internal and external irradiation sites, I should instead report the local neutron flux at each specific position, since the flux varies significantly throughout the reactor.