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.
