Study on dryout and rewetting during accidents including ATWS for the BWR at JAEA

Abstract

Post-boiling transition (Post-BT) heat transfer is important to analyze the duration of surface dryout and the peak cladding temperature during abnormal transient and design basis accident (DBA) conditions in light water reactors. Furthermore, the Japanese new safety regulation revised after the Fukushima Daiichi accident requires the analysis of the beyond-DBA conditions including the anticipated transient without scram (ATWS), where experimental databases for the code validation are almost-non-existent especially for those with highly oscillatory power change caused by thermo-nuclear coupling. Motivated by the importance, Japan Atomic Energy Agency has conducted a series of experiments for BT and rewetting phenomena to develop mechanistic models based on physical understanding and to validate thermal hydraulic codes. So far, the experimental databases for the Post-BT heat transfer rate, deposition rates of liquid droplets, and the dryout and rewetting behavior were developed mainly for the conditions covering the anticipated operational occurrences (AOO) for BWRs. The significance of the precursory cooling just before the rewetting was identified from comparison with a theoretical model for the two-dimensional transient heat conduction in the cladding. Recently, experiments including the simulation of the ATWS were initiated using the single-tube test apparatus with simulated spacers, liquid film heat transfer visualization apparatus for visual observation, and the HIDRA facility with test sections of a full-length 4 × 4 bundle and a short-length 3 × 3 length bundle. Some promising results have already been obtained, which show effects of simulated ferrule and swirl spacers on the post-BT heat transfer and rewetting velocity, and mechanism of the rewetting by the propagation of the liquid film on the initially heated surface of the heater rod.