wbinventor / OpenMC-MGXS

LaTeX source for "Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code" published in Nuclear Technology (2019)
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Nelson's comments #2

Closed nelsonag closed 6 years ago

nelsonag commented 6 years ago
wbinventor commented 6 years ago
  1. I think you're right, I just defined \Sigma{tr} as the term correcting the total xs rather than the corrected total xs. What do you think about renaming the correction term on the LHS of equation 8 as \Delta\Sigma{tr}, and the LHS of equation 9 modified to \Sigma_{tr}?

  2. We can add in the fission production matrix, since there it wasn't omitted for any substantive reason. Most of the math formulation for the tallies is taken from my thesis, which describes only those MGXS which were needed for my OpenMOC simulations. Since OpenMOC uses the vector chi rather than the matrix, it was left out here. But for completeness, I think it would be good to add it in so I will.

  3. Since you've been more involved with OpenMC over the last year than I, what would you expect to be different for v0.10? I doubt any of the results would be substantively different unless some major physics bugs were discovered. Has openmc.mgxs changed in any major way?

  4. I'll remove mention of universes in that paragraph, since I don't think it adds anything in that context.

nelsonag commented 6 years ago

Regarding #3, my gut was wrong. The filters vs scores distinction for expansion tallies was changed after v0.10.0 (i.e., its on develop and not yet released), therefore there might not be anything of interest between v0.9 and 0.10. So I am ok with leaving it at 0.9.

wbinventor commented 6 years ago

I changed to v0.10.0 since @paulromano brought independently remarked on this, so closing this issue now.