Assessment of Steam Generator Tube Flaw Size and Leak Rate Models

摘要

The Canadian Nuclear Safety Commission recently developed the CANT1A (CANDU~(TM) Tube Inspection Assessment) methodology for probabilistic assessment of inspection strategies for steam generator (SG) tubes as a direct effect on the early detection and prevention of tube failure and primary-to-secondary leak of reactor coolant. In an effort to improve CANTIA, an SG tube integrity assessment code, a relevant survey of the literature on the discharge of subcooled water from cracks and critical flow models, SG tube cracks, leakage, and probabilistic assessment methodologies was carried out. The original CANTIA and ANL/CANTIA code models for the flaw opening area and flow leakage rate were reviewed. The predictions from the crack opening area and the leakage flow rate models were compared with experimental measured data from cracked SG tubes.