Reactor pressure vessel integrity assessment by probabilistic fracture mechanics – A plant specific analysis

Abstract

The probabilistic fracture mechanics method is widely adopted in nuclear power plant industry, especially for the structural integrity assessment of reactor pressure vessels. In this work, the plant specific analyses of Taiwan's boiling water reactors, BWR/4 and BWR/6, under the design transients were performed by the FAVOR (Fracture Analysis of Vessel – Oak Ridge) code. The difference of reactor pressure vessel geometry, alloying elements, neutron fluence, and loading conditions were taken into account in the probabilistic fracture mechanics analyses. The failure probabilities of the axial welds at 64 effective full-power year (EFPY) for the analyzed BWR/4 and BWR/6 are less than 2.8 × 10−10/year and 1.5 × 10−4/year, respectively. Furthermore, the circumferential welds present much smaller failure probabilities. The results of this work are useful for the subsequent aging management of the operating reactor pressure vessels.