STRUCTURAL INTEGRITY ANALYSIS OF VVER 1200 REACTOR VESSEL WALL DUE TO PRESSURIZED THERMAL SHOCK

Abstract

Reactor Pressure Vessel (RPV) is the most crucial and irreplaceable part of any Nuclear Power Plant (NPP). So, maintaining the integrity and safety of RPV during any kind of transient or accidental conditions when it can be detrimental e.g., Pressurized Thermal Shock (PTS), need to address as highest priority. Analyzing the PTS event is of utmost importance to ensure the safety and integrity of the VVER-1200, a Gen-III+ reactor model of the Russian Federation, during hypothetical transient scenarios. To analyze the PTS event, Beaver Valley-105 case which is Main Steam Line Break (MSLB) type Loss of Coolant Accident (LOCA), is taken into consideration for this study. A computational fluid dynamics (CFD) analyses was done by ANSYS Fluent considering the injection of cold-water from emergency core cooling system (ECCS) to the hot primary coolant system through cold leg. The beltline region is considered for the modeling to analyze the temperature evolution in high pressure at both coolant and inner wall surface during the transient event and found 167◦C temperature coolant at 200 sec simulations after ECCS water mixing. Sudden increase of pressure in primary system which can lead the scenario to PTS at low temperature is analyzed by ANSYS Mechanical to evaluate the stress profile. The stress analysis for the transient event, which caused PTS due to an instantaneous elevation to a primary system pressure of 16.2 MPa, has been done. It is observed that maximum stress, e.g. Equivalent (Von mises) stress was found 663 MPa and was generated at the edge of nozzle area and the stress values exceed the ultimate yield strength at 16.2 MPa i pressure. Furthermore, a comprehensive postulated crack modelling has been done considering the temperature and pressure of the selected transient case. Three different semi-elliptic crack model are analyzed at selected transient event and for elevated primary pressure of 18, 20 and 22 MPa respectively. Finally, the results of stress intensity factor (SIF) at 100 ◦C obtained for these three postulated crack cases have found, were analyzed with respect to fracture toughness value, KIC of initial state and end of service life state to assess the service life. Here, fracture toughness, KIC is calculated as per ASME and IAEA TECDOC considering the neutron fluence and Nil Ductility temperature values. SIF results (87-118 Mpa.m^1/2) were found below the limiting fracture toughness values, KIC at case II for every pressure conditions. For case I & III, SIF values exceed the KIC values limit at 22 MPa and 20 & 22 MPa respectively. A further study was done to determine the maximum allowable critical temperature for crack for these three cases.