Benchmarking Study of Asyst Code Against Cora-18 Bwr Experiment with Integrated Uncertainty Analysis

Abstract

Verifying and validating thermohydraulic codes for severe accidents in nuclear reactors is an ongoing process that requires a combination of experimental data, benchmarking, uncertainty analysis, and real-world application. The goal is to ensure that these codes can accurately and precisely predict nuclear reactor behavior under severe accident scenarios, thereby supporting risk and safety analyses. ASYST code provides advanced modeling capabilities for nuclear reactor thermal hydraulics in light and heavy water, molten salts, and liquid metals reactors. In this paper, the CORA-18 experiment, part of a series conducted at the German CORA facility focusing on Boiling Water Reactor (BWR) scenarios, is simulated using ASYST. Unlike other tests in the series, CORA-18 uniquely examined the impact of heating and melting transients on a larger bundle scale. The CORA-18 experiment serves as a benchmark for evaluating the robustness and accuracy of such codes. The present study provides a detailed benchmarking of the ASYST code against the experimental results. Additionally, a probabilistic approach was employed by propagating uncertainties in both the input parameters and physical models’ parameters to quantify uncertainties in the model predictions. Although some variations were observed, particularly in predicting hydrogen generation, the findings validate ASYST's capability to simulate core damage processes under complex conditions. Additionally, the incorporation of integrated uncertainty analysis enhances the credibility of the code by providing a quantified measure of confidence in its predictions. This research enables further exploration across different scenarios, reactor types, and computational codes, contributing to ongoing advancements in nuclear safety analysis.