Global Comparison of TALYS and ALICE Code Calculations and Evaluated Data from ENDF/B, JENDL, FENDL, and JEFF Files with Measured Neutron Induced Reaction Cross-sections at Energies above 0.1 MeV

Abstract

The goal of the work is the study of uncertainties in the calculation of activation and transmutation cross-sections for neutron induced reactions using nuclear models and codes having direct relation to the generation of nuclear data files. The calculations were performed with the TALYS code and the ALICE/ASH code, which have been extensively used for nuclear data evaluation during the last decade. The results of calculations are compared with experimental cross-sections from EXFOR for neutron induced reactions for nuclei from Al to Bi. The comparison is done for main reaction channels, for which experimental data are available, without the distinction. The obtained deviation factors allow to define appropriate models for nuclear reaction cross-section calculation at different mass ranges of target nuclei.