mit-crpg/openmc — pull request #133 by nelsonag (2013-01-06 14:29:05+00:00)
Title: Kappa Fission Tallying Capability

Added a Kappa-Fission response type to OpenMC.
This can be accessed with the response type of *k-fission* or *kappa-fission* in a tallies.xml tally object.

The tallying was implemented by adding a k_fission attribute to the micro and macro cross-section classes which is populated with the relevant MT=18 Q-value multiplied by the relevant fission cross-section.

The MT=18 Q-value specifically was used because the ENDF format manual states that the Pseudo-Q-value for fission is contained there. I did not see a corresponding statement in the ACE format though.

I also checked a variety of fissile/fissionable nuclides in the MCNP5 ACE files; those with multiple fission reaction channels had the same Q-value for all fission channels. So only using MT=18's Q is fine as far as I can tell, and the most respecting of the ENDF manual.

This was tested for a few different problem types, the results all look exactly as expected. I also checked all different analog/tracklength/micro/macro means of tallying this and the methods all provide the same answer within statistics.