Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP under PTS

One of the most severe accident conditions of a reactor pressure vessel (RP V) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at a high level. Such an accident is call ed pressurized thermal shock (PTS) transient according to the 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w = 0.05 similar to 0.9 were used to probe what kind of flaw and what kind of trans ient are most detrimental for the RPV in the end of life (EOL). Both the li near elastic and elastic-plastic material models were used in the stress an alysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were in vestigated for different PTS transients. The fracture evaluation for the RP V in question under PTS transients, comparing with the material crack initi ation toughness K-IC, is performed in this paper. (C) 2000 Elsevier Science S.A. All rights reserved.