Prediction of reactor internals cracking in BWRs

This paper describes an approach to provide guidance for plant life extension decisions related to irradiation assisted stress corrosion cracking of BWR internals. Experiments evaluated critical concepts of crack advance under irradiated conditions and quantified specific contribution to irradiation assisted stress corrosion cracking (IASCC) of water radiolysis and stainless steel radiation damage. From these results, a data base and life prediction methodology was developed and validated to provide quantitative prediction of: (1) the factors that control IGSCC (intergranular stress corrosion cracking)/IASCC of stainless steel under BWR irradiation conditions; and (2) the useful lifetimes of irradiated BWR internal components. This paper will discuss the background to the fundamental modeling under unirradiated conditions, and will focus on the hypotheses, critical experiments, and resulting radiation modifications that were developed in this program. The validity of this approach is addressed by comparing predictions with lab and field data, and its unique capabilities are compared to existing design and life evaluation approaches