The Steam Generator Tube Rupture (SGTR) accident is a high complex process influenced by multiple factors, including steam-water critical flow, water-LBE interaction, and the frictional pressure drop in the discharge pipeline. While most simulations demonstrate good agreement with experimental data in terms of pressure trends, they often overlook the comparison of instantaneous mass flow and steam bubble evolution behavior, leading to insufficient verification. This study systematically validates the accuracy of the Fast Reactor Safety Analysis Code (FASAC) in simulating SGTR accidents based on the experimental facility established at the China Nuclear Power Technology Research Institute (CNPTRI). A data normalization approach was adopted to conduct a detailed comparative analysis between experimental and simulation results in terms of water injection mass flow, reactor vessel pressure changes, and bubble evolution behavior. Both the simulation results for water injection mass flow and pressure changes show a strong agreement with the experimental results, with errors of 5.4% for critical mass flow, 3.5% for pressure rise rate, and 6% for maximum pressure. The bubble evolution behavior is consistent with the visualization from the temperature distribution matrix. The behavior of steam bubbles can be classified into three stages: generation and expansion, migration and LBE blockage of the discharge pipeline, and the stable distribution of gas–liquid volume fraction. This study validates the accuracy and reliability of the FASAC in predicting pressure trends, mass flow, and steam bubble evolution, further demonstrating its applicability in the safety analysis for Lead-cooled Fast Reactor (LFR).