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,


It seems this one escaped the forum.
It is very probable that the energy release per fission is different in the two codes. The standard value of 188 MeV in mcnp does not include delayed (n and gamma) energy deposition and thus is not the correct value to relate to thermal power of the system. In other words it is the worng number to get the source multiplier.

Dear August,

Thank you for your reply, time and consideration. This is very good point to look at.

Best regards,