Re-Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RT_PTS values were determined by the tangent approach. In the case that the RT_NDT at or beyond the RPV design life may exceed the RT_PTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RT_PTS values are increased to varied extents for the different loading transients.