Problem in validation of Multi-group cross-sections obtained using OpenMC and MCNP

Hi everyone!
First of all I am very happy to find this group so that I can share my problem here. Thanks to Group Administrator and Members.
I want to find Multi-group cross-sections using OpenMC. I use tallies “flux” and “total” and then divide “total” by “flux” to get total cross-sections of a pin-cell. But the problem is this that the results obtained in this way do not match with the results of same pin-cell obtained with MCNP in the resonance region. The files of both OpenMC and MCNP are attached here. Excel file for calculations is also attached. Can anybody please help me out. I am student of Nuclear Engineering and need to sort out this. I would really appreciate the effort.

materials.xml (941 Bytes)

settings.xml (571 Bytes)

tallies.out (2.7 KB)

tallies.xml~ (611 Bytes)

settings.xml~ (586 Bytes)

.~lock.tallies.out# (77 Bytes)

geometry.xml~ (1.03 KB)

materials.xml~ (1.99 KB)

results.txt (2.81 KB)

geometry.xml (1.01 KB)

tallies.xml (612 Bytes)

cross-sections calculation.xlsx (12.9 KB)

MCNP-output (317 KB)

MCNP-input (2.02 KB)

Some things to consider:

  • The OpenMC geometry has a gas gap whereas the MCNP geometry does not
  • Different cross sections were used in OpenMC vs MCNP
  • The MCNP tallies looks like they are set up to get reaction rates for individual nuclides whereas the OpenMC tally is for all nuclides in the fuel.
  • I could be wrong, but I don’t think your tally multipliers in MCNP include the atom density of the material. OpenMC automatically multiplies by the atomic density so that you effectively get macroscopic quantities.
    The energy dependence of the two results looks comparable, so it seems to just be a normalization issue.


Thanks Paul.
It worked…! You were correct. Atom density was not included in tally multipliers of MCNP. Now I have got good enough results.
Mirza Younis Baig