Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Abstract

Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the nuclear power plant (NPP) Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 libraries are considered. In the first part of this work the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few 100 pcm in multiplication factor for a fresh fuel due to differences in 235U and 238U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout, and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1%. Some differences in isothermal temperature coefficient were observed, however, they only marginally affect core power defect going from zero to full power.