Criticality and sensitivity analysis of VVER-1000 mock-up with SCALE and MCNP5 code using ENDF/B-VII.1 nuclear data library

Abstract

Accurate analysis of reactor criticality is essential for reactor design and safety assessments. This paper conducts a criticality study of a VVER-1000 mock-up benchmark experiment, which was performed at the LR-0 research reactor operated by the Research Center Rez in the Czech Republic. Benchmark calculations are performed using two Monte Carlo codes – SCALE (KENO-VI) and MCNP5 – utilizing the ENDF/B-VII.1 continuous-energy nuclear data library for criticality calculations. The mock-up was examined under six different critical configurations by varying coolant levels and boric acid concentrations. This paper provides a comparative analysis of the results from SCALE (KENO-VI) and MCNP5 to assess the suitability of SCALE (KENO-VI) as a verification tool in the regulatory process, with MCNP5 as the reference code. Additionally, the research work also investigates the sensitivity of various reactor system parameters’ uncertainty, highlighting their significant impact on criticality result, which could potentially lead to overly conservative safety margin. The study focuses uncertainty on five key technological parameters: fuel assembly pitch, fuel cladding thickness, fuel density, fuel enrichment and boric acid concentration. A comprehensive analysis of these uncertainties, along with an assessment of their sensitivity to the criticality results, is provided in this study.