Abstract A detailed assessment of the JEFF 3.1 and ENDF/B VII.0 nuclear data libraries with the KANEXT code system has been performed on each level of the reactor simulation steps. This allowed for the sensitivity estimation of the lattice and core parameters to changes in nuclide and specific reaction cross section. The direct calculation of cross section data influence, in particular for specific reaction types, is possible due to the unique substitution feature of the group constant module GRUCAL. Further on, the different impact of several reaction types on the lattice versus core calculation points out the “engineering uncertainty” using the two major libraries for the identical problem. MCNP(X) and KANEXT cell models can be mutually adjusted such that differences between stochastic and deterministic data processing can be identified. The current analysis emphasizes, that the two major nuclear data libraries JEFF 3.1 and ENDF/B VII.0 exhibit to some extent compensation of deviations rather than reliable integral data calculation. On a special type of a fast reactor application we show that exchange of JEFF 3.1 and ENDF/B VII.0 cross sections introduce large discrepancies. A major part is attributed to the inelastic cross section. Further investigated discrepancies concerning the capture and the fission cross section in the epithermal energy range might be also a concern for LWR which was beyond the scope of the current study.