What is the unit of flux given in tallies.out?
Score unit in OpenMC is " per source particle". And the unit of flux in tallies.out is “neutrons-cm/source”. For further information , go to openmc manual in details.
Best regards
Pranto
https://groups.google.com/forum/#!msg/openmc-users/r5JtjyH9BIM/Cyg-djYABQAJ
You can also check this post.May I ask where you are from
Here’s a link to the relevant section of the user’s manual that outlines what units are used for each score.
https://docs.openmc.org/en/stable/usersguide/tallies.html#scores
Best,
Paul
When i draw flux vs z position curve using spyder, it shows unit n/src.What does it mean?
Also since neutrons/cm2.sec is the unit of flux, how could neutrons.cm/src be the unit?
OpenMC calculate flux using " Collision estimator Tally". In monte carlo, history of neutrons are weighted by " per source particle". OpenMC accumulates 1/Sigma_t which has units of cm, so units are like " neutrons.cm/source". Read theory and methodology part from openmc manual https://docs.openmc.org/en/stable/methods/index.html
Best regards
Pranto
Just to add on to the discussion here – OpenMC, like all other neutron transport codes, can not give you an absolute value of a flux (in units of neutrons/cm^2-sec as was suggested) without knowing what the actual source in the problem is. In a k-eigenvalue calculation where the source comes from fission, the actual source rate is really determined by the power level, which is something that the code doesn’t know about. Another way to think about this is from textbook reactor physics; if you solve the diffusion equation in a slab, everyone knows you get a cosine shape, but there is also a constant out front that determines the magnitude. The analogy here is that OpenMC will give you that cosine shape, but it can’t give you the magnitude.
There are other posts in this mailing list discussing how to go from the units that OpenMC gives you to “real” units (neutrons/cm^2-sec). Have a look here, for example.
Best regards,
Paul
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