Sodium-cooled fast reactors (SFRs) utilize liquid sodium as a coolant to extract heat from the core and transfer it to water via steam generators for steam production. Therefore, the steam generator serves as the interface between sodium and water. If a rupture occurs in the heat transfer tubes, a large amount of water/steam can enter the high-temperature sodium, generating a significant amount of hydrogen and heat. This causes a rapid increase in temperature and pressure near the rupture, creating pressure waves that propagate through the secondary cooling system, potentially damaging pipes and equipment and compromising reactor safety. This scenario is referred to as a large leak sodium-water reaction accident. This study uses a small-scale experimental setup to simulate the coolant system of a large sodium-cooled fast reactor. By injecting a controlled amount of high-temperature and high-pressure water/steam into the experimental setup, the accident progression of a large leak sodium-water reaction in the steam generator is simulated. First, the changes in thermal parameters such as temperature, pressure, flow rate, and liquid level within the system after the accident are analyzed to understand the progression of the large leak sodium-water reaction. Second, parameters such as the buffer tank liquid level, pressure, and flow rates at the evaporator inlet and outlet are analyzed to verify the accuracy and effectiveness of the alarm logic for large leak sodium-water accidents. Finally, the changes in thermal parameters such as temperature, pressure, and flow rate during the execution of the protection actions for large leak sodium-water accidents are analyzed to verify the rationality of the protection actions.