Abstract: Steam generator tube rupture (SGTR) accident is one of the accident scenarios that must be considered during the design and safety analysis process of lead-cooled fast reactors (LFR). Focused on this topic, numerous experiments and theoretical research have been conducted in recent decades, which could be generally categorized into two aspects: on the one hand, separation effect experiments and theoretical analysis focusing on the specific phenomena of lead-water interaction process; on the other hand, engineering test facilities and numerical simulations focusing on the overall process of LFR SGTR accident, mainly concentrated on the evolution of macro parameters such as system pressure and lead pool temperature during SGTR accident. In this paper, firstly, the basic mathematical and physical models of ACENA code was introduced; then the code was validated against the LBE-N₂ two-phase flow experiment HESTIA-2, the KYILN-Ⅱ-S LBE-H₂O interaction experiment and the analytical solution of the point-kinetic neutronic equations; after that the Eulerian multi-phase analysis code ACENA was applied to simulate the postulated unprotected SGTR process of European pool-type lead-cooled fast reactor ALFRED. Attention was paid to the simulated migration process of steam bubbles in the lead pool as well as the evolution of safety parameters such as maximum cladding temperature, fuel centerline temperature, relative nuclear power and reactor vessel pressure. Furthermore, the sensitivity of ruptured tube quantities, lead coolant circulation paths and neutron dynamics parameters on calculation results were analyzed. This paper’s study demonstrates that the Ishii-Chawla-Suzuki interfacial drag coefficient model could accurately predict the diffusion and migration characteristics of rising bubbles in circular/annular LBE flow channels when combined with the interfacial area concentration transport model proposed by Ishii et al; The simulation of KYLIN-Ⅱ-S experiment shows that ACENA code could reasonably reproduce phenomena such as system pressure fluctuations and temperature transients of liquid lead during the interaction process between LBE and water; the calculated results of ALFRED reactor’s key thermal-hydraulic parameters under hot full power condition by ACENA code are generally consistent with results obtained by internationally recognized one-dimensional system code TRACE/FRED, proving the reliability of ACENA code when applied to full reactor system level; The calculation of the hypothetical SGTR accident of the ALFRED reactor verifies the ACENA code’s capability of simulating the complex multi-component multi-phase flow, and the calculation results reveal that designing the primary coolant flow path rationally and reducing the number of ruptured pipes are of great significance to mitigating the potential consequences of LFR SGTR accidents. This work can provide reference for the safety analysis work of China’s pool-type lead-cooled fast reactors.