Experimental study of boiling critical heat flux with low mass flux under motion condition in a wetted perimeter equivalent rod

The significance of floating nuclear power plant has recently surged as a viable solution for addressing ocean energy supply issues, while also meeting energy conservation and emission reduction goals. During the initial phase of nuclear reactor startup or in the event of a LOCA (Loss of Coolant Accident), the occurrence of critical heat flux (CHF) at exceptionally low mass flux in the core of floating nuclear power plant is a possibility. This underscores the importance of CHF safety characteristics of floating nuclear power plant. Despite this, there has been a noticeable absence of published experimental research on CHF under motion conditions with high operational parameters. To bridge this gap, CHF experiments were carried out on rod test sections under both static and motion conditions, employing the wetted perimeter equivalent design method. This study scrutinized the impact of system variables on CHF and evaluated the effectiveness of the design strategy based on the wetted perimeter equivalent method, using experimental data as a reference. Additionally, it elucidated the CHF mechanism and identified the penalty factor on CHF under conditions of inclination, rolling, and heaving, with the maximum penalty factor reaching up to 15%. The insights gained from this research are instrumental in guiding the CHF test scheme design under motion conditions and lay a foundational basis for the thermal design of floating nuclear power plant.