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 neutron induced reaction cross-sections for target nuclei from ²⁷Al to ²⁰⁹Bi and incident neutron energies above 0.1 MeV have been calculated with the TALYS code and the modified ALICE code using different models for the description of the nuclear level density. Several thousand experimental points from EXFOR have been used for the comparison with these calculations. The obtained results give the possibility to find the best approaches for the cross-section calculation for nuclei in different mass ranges.