Neutron flux and reaction rates discrepancies with MCNP

Dear experts,

I am a new user of OpenMC and I am currently working on benchmarking fluxes, reaction rates and heat loads in the Molten Salt Fast Reactor against MCNP and Serpent2.

As expected, using consistent nuclear data, the keff values I obtained using the codes are in very good agreement but when looking at fluxes and reaction rates very strange discrepancies are appearing as shown below:

Code MCNP6 OpenMC
keff ± σ (pcm) 0.98523 ± 13 0.98511 ± 19
Mean flux in fuel ( × 10^15n/cm2/s) 3.61 3.88
233U fission rate (mol/day) 12.9 14.05
233U capture rate (mol/day) 1.93 2.11
232Th fission rate (mol/day) 0.23 0.26
232Th capture rate (mol/day) 15.2 16.59

I talked about the problem with my colleagues, did several tests but up to now, unfortunately, we didn’t manage to figure out what was going wrong in the modelling which is a rather simple one. I am tempted to suspect a normalization problem as, from code to code, the process is different and can be tricky. For normalizing the OpenMC results, I followed the instructions given in 8. Specifying Tallies — OpenMC Documentation about tally normalization and read the related posts in the forum and for the MCNP case, I am normalizing the value related to the reactor power, nu-bar, keff, and energy released per fission as recommended in literature. Maybe I misunderstood something there or I am missing something in the modelling that I am not able to see.

New users are unfortunately not allow to include in their post any attachment but I will of course provide any material needed for further investigation.

Thank you in advance for your help in solving this issue.

Best regards,