Impact of LOCA, SG Tube Rupture, and Power Loss on VVER-1200 Reactor Safety

This study analyzes the transient behavior of the VVER-1200 nuclear power plant under various failure scenarios using the PCTran VVER-1200 simulation tool. Thermal-hydraulic and safety characteristics assessed in under 600 seconds. The pressurizer, reactor building water levels, steam generators (SG-A and SG-B), reactor core, and other components are essential to the system. Systematic temperature measurements are taken at the maximum fuel, average RCS, peak clad, and reactor building levels. The talk covers fuel Doppler feedback, soluble boron's effect on reactivity, nuclear boiling ratio (DNBR) aberrations, and temperature changes. After 5 seconds, faults may develop, and system specifications affect basic circumstances. A 75% loss-of-coolant accident (LOCA) in the cold leg, a 60% power loss in the AC system, steam generator tube ruptures at 30% in SG-A and 45% in SG-B, and one auxiliary pump failure are simulated failures. The analysis found significant disparities between nuclear power station water levels and RCS leak mass flow rate. Core thermal power, nuclear flux, and turbine load changes indicate reactor performance. The results show the system's sensitivity to catastrophic events and how passive safety mechanisms stabilize the reactor. Specialists' rigorous investigation of historical accident processes improved VVER-1200 reactor safety.