[openmc-users] Not getting reasonable output for cell flux

User Support
1

Hi Jake,

Is the cell in your geometry repeated in a lattice and/or multiple universes?

Best,
Paul

2

Hi Jake,
Did you divide the tally output by cell volume to get the flux in unit of 1/cm2 ?

Tarek

3

I eventually figured this problem out, thanks for the replies! My units were getting a bit mixed up.

To those also running into this problem: there is no easy way to determine the number of neutrons that ever intercept a cell. You’ll have to settle for flux, which has units in the raw openmc output of neutrons (per source neutron)-cm/s (which is NOT necessarily less than unity). Dividing flux by the volume of the cell (cm^3) gives you neutrons (per source neutron)/cm^2-s, or the mean flux per source neutron in the aforementioned cell or mesh volume.

If anyone has the time,it would be helpful if units could be added to the tally descriptions in the user’s guide. It would make life a lot easier for people who are learning OpenMC without any background from MCNP or GEANT.

Thanks,

Jake

4

Thanks for the suggestion Jake. I’ve just added a description of units for each score type in the documentation. It will show up in the online documentation for the next release.

To be clear, there is no “per second” in any of the tallies. To get a tallied quantity per second, you need to know the source strength in source particles per second, something that OpenMC could not infer.

Best,
Paul