Applications of Modelling and Simulation (Mar 2022)
Analysis of Thermal-Hydraulics Parameters During Steam Generator Tube Rupture Event of VVER-1200 NPP Using PCTRAN Simulator
Abstract
The overall performance of steam generators plays a significant role in ensuring the safety of a nuclear power plant (NPP) operation. The analysis of thermal-hydraulic parameters during a steam generator tube rupture (SGTR) event of VVER-1200 NPP is conducted by applying the personal computer transient analyzer (PCTRAN) simulator. Four cases, namely, 25%, 50%, 75% and 100% of one tube rupture in two steam generators with the concurrent loss of AC power have been performed. Among the four cases, major variation in time was not observed for the occurrence of the reactor scram, reactor coolant pump trip, main feed-water pump trip, and turbine trip. The pressure and the temperature of the reactor coolant system (RCS) increase rapidly to a peak value due to event initiation, and drop promptly after the reactor scram. The stabilized pressure and temperature of the RCS are higher for the smaller break size of the SGTR. The secondary pressure of the steam generator is also increased to a peak value, followed by an increasing and decreasing trend, in turn, due to the repeated opening and closing of safety relief valves of the steam generators. The liquid level of the pressurizer is increased rapidly due to the liquid surge towards the pressurizer after the event and it is stabilized after the opening of the safety relief valve. The stabilized liquid level of the pressurizer and the steam generator is higher for the smaller break size of the SGTR. The earlier emergency core coolant injection to the reactor was required for the larger break size of the SGTR. There is no increase in the peak cladding temperature and the peak fuel temperature during the calculation period for all these cases. The results of this study provide a valuable understanding of SGTR events with the concurrent loss of AC power for the PCTRAN model of the VVER-1200 NPP.
