SCALE and SERPENT solutions of the OECD VVER-1000 LEU and MOX burnup computational benchmark

Highlights: • New solutions for the VVER-1000 LEU and MOX burnup computational benchmark have been obtained using ENDF/B-VII and JEFF3.1 nuclear data libraries. • The SERPENT and SCALE codes have been used for the first time to solve the benchmark exercises. • The comparison of our results with the ones available in literature shows generally a good agreement over all the reactor states considered in terms of reactivity values, pin-by-pin fission rates distributions and nuclide concentrations. • The SERPENT models for the LEU and MOX assemblies have also been tested with JEF2.2 making of this work also a new Monte Carlo reference solution for the benchmark exercise with modern nuclear data libraries. - Abstract: The loading of hybrid cores with Mixed Uranium Plutonium Oxide (MOX) and Low Enriched Uranium (LEU) fuels in commercial nuclear reactors requires well validated computational methods and codes capable of providing reliable predictions of the neutronics characteristics of such fuels in terms of reactivity conditions (k_inf), nuclide inventory and pin power generation over the entire fuel cycle length. Within the framework of Joint United States/Russian Fissile Materials Disposition Program an important task is to verify and validate neutronics codes for the use of MOX fuel in VVER-1000 reactors. Benchmark analyses are being performed for both computational benchmarks and experimental benchmarks. In this paper new solutions for the (UO₂ + Gd) and (UO₂ + PuO₂ + Gd) fuel assemblies proposed within the "OECD VVER-1000 Burnup Computational Benchmark" are presented, these being representative of the designs which are expected to be used in the plutonium disposition mission. The objective is to test the SERPENT and SCALE codes against previously obtained solutions and to provide new reference solutions to the benchmark with modern nuclear data libraries