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Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code)” provides fracture evaluation method and criterion, based on the linear elastic fracture mechanics, for irradiated stainless steels of boiling water reactor (BWR) internal components. The fracture toughness criterion, however, was developed with limited materials testing data and knowledge available at that time and it has not been revised since the code originally established. In this study, fracture toughness criteria for structural integrity evaluation were discussed and developed with the latest database on fracture toughness of irradiated austenitic stainless steels, including additional material testing data obtained in this study for the neutron fluence range of interest from 1 to 3 dpa. First, the fracture toughness data of austenitic stainless steels irradiated in BWR conditions were compiled to evaluate the correlation between fracture toughness and neutron fluence. Material characteristics potentially affecting fracture toughness, such as chemical composition and specimen orientation, were also considered and discussed in the development of the fracture toughness criteria. Based on the results, the fracture toughness criteria for irradiated austenitic stainless steels were proposed for fracture evaluation of the BWR internal components.

Cracking on metal surface including Stress Corrosion Cracking (SCC) is a concern for nuclear power plant reactor system piping and components. Several repair techniques have been developed over the years that help address the cracking issues. In the United States, some crack remaining repair technologies such as Weld Overlay (WOL) technique have already been developed and also some technologies such as Excavate and Weld Repair (EWR) are still being actively discussed. Fitness-for-Service (FFS) codes in Japan also define some repair methods which allow repairing the components without removing the cracks. This paper introduces three repair methods which are described in the Fitness-for Service code in Japan, the WOL, the cap repair technique for Bottom Mounted Instrumentation (BMI) nozzles and seal welding. WOL is a repair method that maintains structural integrity by depositing Intergranular Stress Corrosion Cracking (IGSCC) resistant weld to the outer surface of cracked circumferential welded joints in primary piping made of austenitic stainless steel. The cap repair technique for BMI nozzles is a method of repairing from the outer surface when leakage due to SCC occurs. The cap develops a new boundary outside the BMI nozzle to prevent the leakage by retaining the leakage inside the cap. The seal welding technique is also a repair method to cover the cracks by welding to prevent propagation of the cracks and to prevent leakage of reactor water. Provided the cracks are adequately managed and evaluated, these crack remaining repair methods are very beneficial.