Probabilistic fracture analysis for boiling water reactor vessels considering seismic loads during decommissioning transition period

Abstract During decommissioning transition period, the hazards from internal pressurization and thermal transients to a reactor vessel can be eliminated, but the effects of seismic loads on vessel integrity may be of concern. In this paper, we employed the Oak Ridge National Laboratory’s (ORNL) FAVOR code to perform probabilistic integrity assessment for the Taiwan domestic boiling water reactor (BWR) vessels of a decommissioning plant considering seismic loads. A manner to transform seismic loads as the equivalent internal pressure for FAVOR analysis was proposed. Two probabilistic fracture mechanics (PFM) models, the conservative model that U.S. NRC has used for BWR regulation, and the realistic model reflecting the entire vessel beltline configuration, were analyzed and compared. It is found that the circumferential welds dominate the fracture risk but still can be negligible. Present results could be the reference for weld inspection relief application of the RPV during the decommissioning transition period in Taiwan.