Parameter making and preliminary test of ACE format libraries based on different ENDF/B

Evaluated nuclear data libraries are the basis of reactor physics analysis. This study investigates major versions of evaluated nuclear data libraries in various stages, namely ENDF/B-Ⅳ, Ⅴ.2, Ⅵ.8, Ⅶ.0 and Ⅶ.1, which are then processed by the internationally renowned nuclear data processing code NJOY to obtain five groups of continuous energy point cross section libraries. Micro cross sections of certain nuclides are compared and Reactor Monte Carlo Code (RMC) is used for verification of criticality benchmark. The result shows that continuous energy neutron cross section libraries, based on ENDF/B-Ⅶ.1, has higher accuracy and reliability.