Monte Carlo Code MCS Multi-cycle Depletion Analysis of Westinghouse 3-loop PWR

Abstract

A refueling capability has been implemented in the Monte Carlo code MCS under development at the Ulsan National Institute of Science and Technology (UNIST) since 2013. The purpose of the refueling capability is to model and simulate multi-cycle problems of pressurized water reactor (PWR). MCS is a validated continuous energy 3D neutron-transport code with capabilities for solving large scale criticality/eigenvalue problems, including thermal hydraulic (TH) feedback effects. In this work, we present the verification and validation (V&V) of MCS refueling capability. The analysis is conducted on the operating cycles 19 to 24 of a Westinghouse 3 loop (WH3L) PWR operated in South Korea. For the verification, MCS results, including the power profiles and the boron letdown curve, are compared to results from the deterministic code system STREAM/RAST-K-2.0 and the nuclear design report (NDR). STREAM is a lattice/transport code which can be used for cross-section generation, RAST-K-2.0 is a nodal code for reactor analysis and STREAM/RAST-K-2.0 is a two-step reactor analysis approach, all developed by UNIST. The results from the NDR were generated by the deterministic two-step method code system (PHOENIX-P/ANC). For the validation, MCS results are compared to measured data. The MCS analysis of WH3L is based on a 3D quarter core model developed for the reactor. The fuel depletion simulation is conducted with TH feedback, equilibrium xenon, critical boron concentration search and on-the-fly Doppler broadening. ENDF/B-VII.1 nuclear data is used in the calculations. The goal of this paper is to demonstrate the reliability of MCS for the whole core depletion analysis of multi-cycle problems of PWRs.