MCNP vs Serpent pin-cell burnup discrepancy in keff evolution

[emilmammadzada]

TL;DR

MCNP vs Serpent pin-cell burnup discrepancy in keff evolution

Hello Dear Experts,
I am performing a simple pin-cell burnup (depletion) calculation using MCNP and Serpent and comparing the keff evolution as a function of time and burnup.


Despite modeling the same physical system in both codes (geometry, materials, initial isotopic composition, power normalization, burnup steps, boundary conditions, and thermal scattering treatment), I observe non-negligible differences in keff values and depletion behavior between MCNP and Serpent.


The general trend of keff decrease with burnup is consistent, but the absolute keff levels and burnup-dependent behavior do not match quantitatively.


I would appreciate insights on:

• Common physical or numerical sources of MCNP–Serpent discrepancies in depletion problems
• The impact of depletion solvers, normalization methods (power vs flux), or nuclear data (decay, fission yields)
• Key input parameters that must be carefully aligned to obtain consistent burnup results
• Recommended benchmarking practices for pin-cell depletion validation

I will provide both MCNP and Serpent input files for direct comparison.


Any suggestions or references would be greatly appreciated.