Safety Analysis of Loss of NPP Off-Site Power with Failure of Reactor SCRAM (ATWS) for VVER-1000

Abstract

Loss of off-site power in the Nuclear Power Plant (NPP) with failure of reactor SCRAM, i.e, inability to automatically drop control rods into the core after reactor trip signal during transient for PWR, is considered as a Design Extension Condition (DEC). Such an event is called ATWS, i.e., Anticipated Transient Without Scram. The event has been analyzed using thermal hydraulic computer code RELAP-5/MOD 3.2 for Kudankulam Nuclear Power Plant (KKNPP). RELAP-5/MOD 3.2 uses a one-dimensional, two fluids, non-equilibrium, six equation hydrodynamic model with a simplified capability to treat multi-dimensional flows. KKNPP has two operating VVER-1000 Reactors. VVER-1000 is a Pressurized Water Reactor having active and passive safety systems for such event mitigation. As a result of the initiating event, i.e., Loss of NPP station service power, trip of reactor coolant pump, closure of turbine governor valve, and loss of steam generator feedwater take place. This affects heat removal from the reactor core due to loss-of-coolant circulation and pressurization of the primary and secondary circuits due to closure of the turbine governor valve. This results in the generation of reactor scram signal. Failure of reactor SCRAM is considered in the event. Thus, the reactor power is not decreased even after the scram signal and ATWS condition is identified. This results in the actuation of both passive and active safety systems for boron addition to the core, designed for ATWS mitigation. The objective of this study is to evaluate the thermal hydraulic and neutronic behavior of the core and verify the capability of safety systems for event mitigation. The thermal hydraulic parameters are checked against the applicable acceptance criterion for the event.