Analyzing Thermal-Hydraulic Parameters During Cold Leg Loss Of Coolant Accident Without Reactor Scram Using Pctran Model Of Vver-1200

The loss of coolant accident (LOCA) in the cold leg, including breaks of 0.000507 m2 (Case-1), a break of 0.008110 m2 (Case-2), a break of 0.129720 m2 (Case-3), a break of 0.164170 m2 (Case-4), 0.202680 m2 (Case-5), and a break of 0.240800 m2 (Case-6), with the concurrent incident of the anticipated transient without scram (ATWS) and the loss of AC power, have been simulated and analyzed by applying the personal computer transient analyzer (PCTRAN) simulator model of the VVER-1200 nuclear power plant (NPP). The larger break size results in a smaller pressure and temperature of the reactor coolant system (RCS) during the simulation period. The RCS pressure was lower than 110% of the initial operating pressure in all cases. The secondary side pressure of the steam generator was stabilized earlier for the larger break accident cases. There is no decrease in the liquid level of the core in Cases 1 and 2, but it becomes empty in Cases 3, 4, 5, and 6 just after the initiation of the accident. The pressurizer liquid level increases rapidly for Cases 1 and 2, due to the liquid surge towards the pressurizer after the initiation of the accident, but it becomes empty in Cases 3, 4, 5, and 6. The falling liquid level of the steam generator is higher in smaller break accident cases. The break mass flow rate decreases to a plateau value earlier for smaller break cases. The thermal power of the reactor, peak cladding temperature (PCT), and fuel temperature showed a rapid drop after the initiation of the accident, followed by an increase to a peak value. At the end of the simulation, there was a drastic drop in reactor thermal power, PCT, and fuel temperatures for Case-6 due to the supply of boric acid solution to the reactor. Among the accident cases, the maximum values of PCT and peak fuel temperature for Case-6 were recorded as 646.20 °C and 2084.88 °C, respectively, which is within the acceptable range for a design basis accident of a pressurized water reactor. Thus, the major thermal-hydraulic parameters were within acceptable values for the cold leg break LOCA with the simultaneous initiation of the ATWS and the loss of AC power of the VVER-1200 NPP model of the PCTRAN simulator.