Some questions when using OpenMC

Dear Sir/Madam:

Hello ! I am writing this mail because I some simple questions, and I so sorry if the answers to my question are too obvious.
In tallies.xml I use “flux fission” to get total flux and and fission rate in certain cells. But I find the values are small such as the following.

Flux 6.16807E-02 +/- 3.96370E-04
Fission Rate 2.56586E-03 +/- 1.67589E-05

I do not know the units of both tallies. I want to calculate the fission cross section, so I need to get a flux in a unit of “neutrons/cm^2 * s” . Could you please tell me how can I normalize the flux and fission rate value to get the right units ? Thanks a lot.

Yours Sincerely

Hi Chong,

Please see the following discussion on the user’s list:
https://groups.google.com/forum/#!searchin/openmc-users/cm%5E2/openmc-users/r5JtjyH9BIM

Best regards,
Paul