Zeros for reaction rates

User Support
1

Hi Folks,

This might be a bug, I’ve modified the boxes.xml example in order to put a fixed source, and some reaction rate tallies. The difference is that I have added a material, that contains only Au197, with the intention to use a tally filter on that nuclide to get the (n,gamma) reaction rate. The material with au197 is not a used by any cell.

I have tallied across all the cells, the (n,gamma), (n,2n) and total reaction rate and get all zeros.

Im hoping this is a gap in my understanding and not a bug.

Please find attached my inputs.

Thanks

Andy

geometry.xml (1.47 KB)

materials.xml (656 Bytes)

plots.xml (194 Bytes)

settings.xml (330 Bytes)

tallies.xml (313 Bytes)

2

Hi Folks

Ping! Any ideas?

Thanks

Andy

3

Hi Andy!

Thanks for the reminder and sorry it took a while to get back on this. I was on vacation for the last week and a half. Currently in OpenMC, if you are tallying a nuclide-wise reaction rate, the nuclide must appear in the material that you are tallying. If you want to get the micro xs for Au197 under the flux in cells 1-3, the solution in your case would be to put a trace amount of Au197 in each cell so that the tally is non zero and then divide out the trace number density.

Best,
Paul

4

Hi All,

I am new to openmc and I have a question related to this thread.

I’m trying to tally the total inelastic reaction for In115. I have marked the reaction for tallying by it’s MT number (5).
I have put a trace of In115 in the cells I wish to tally, but I’m getting zeros. However, if I do the inelastic scattering reactions individually I get non-zero answers.

Any suggestions? Do I have to tally all the individual inelastic level reactions and sum possibly?

I have modified Andys inputs to show my problem.

Thanks in advance,
Ander

geometry.xml (1.47 KB)

materials.xml (671 Bytes)

plots.xml (194 Bytes)

settings.xml (330 Bytes)

tallies.xml (292 Bytes)

5

Hi Ander,

MT=5 is actually a “miscellaneous” reaction. What you are probably looking for is MT=4, which is defined by the ENDF-6 formats manual to be the sum of MT=50-91 (level inelastic scattering reactions). That being said, most “summed” reactions cannot be tallied correctly in OpenMC presently. The only summed reactions that are guaranteed to work are total, fission, and absorption. We’ve run into this recently in another use case (e.g., if you ask for (n,p) and it is broken up into MT=600-649, the resulting rate will zero). I’m aiming to get this fixed by the next release. In the meantime, your best bet is to tally the individual level inelastic reactions and summed them manually.

Best regards,
Paul