Steam generator tube rupture (SGTR) accidents in lead-bismuth cooled fast reactors result in the generation of numerous steam bubbles owing to the interaction of the high-temperature liquid lead-bismuth eutectic (LBE) from the primary circuit and high-pressure subcooled water of the secondary circuit. These steam bubbles carried by the LBE may enter the core, causing local heat transfer deterioration and a power transient, seriously affecting reactor safety.

This study aims to elucidate the movement and dynamic behaviors of steam bubbles in liquid LBE and develop a drag coefficient model applicable to bubble migration in LBE for assessing the safety of the core during SGTR accidents.

Based on the coupled level-set and volume-of-fluid (CLSVOF) method, a three-dimensional numerical model for calculating the drag coefficient of steam bubbles in LBE was established to study their movement and dynamic behaviors. Firstly, the deformation, velocity, and trajectory characteristics of bubbles in LBE were analyzed, and their drag coefficients were estimated. Then, the simulated drag coefficient values were compared with those from existing models whilst the prediction performance of the model proposed by Tomiyama was superior to those of the others. Finally, Tomiyama's drag model was optimized by introducing the We number, and its applicability was analyzed.

Analysis results show that the errors of the existing drag coefficient models are large under the condition of bubble–LBE two-phase flow, and the calculation error of the optimized model for the bubble drag coefficient in LBE is within 15%.

The feasibility of drag calculation model optimized in this study is verified, demonstrating its suitability for calculating the drag coefficient of steam bubbles in LBE.