抄録
In this paper, fatigue verification of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, and relevant issues that are often discussed in connection to the power uprate of several Swedish BWR reactors in recent years, are dealt with. Key parameters involved in the fatigue verification, i.e. the alternating stress intensity S_<alt>, the penalty factor K_e and the cumulative damage factor U, and relevant computational procedures applicable for the assessment of low-cycle fatigue failure using straincontrolled data, are particularly addressed. A so-calle simplified elastic-plastic discontinuity analysis for alternative verification when basic fatigue requirements found unsatisfactory, and the procedures provided in NB-3600 for evaluating the alternating stress intensity S_<alt>, are reviewed in detail. Our emphasis is placed on other procedures alternative to the simplified elastic-plastic discontinuity analysis. A more in-depth discussion is given to an alternative suggested earlier by the authors using nonlinear finite element analyses. This paper is a continuation of our work presented in ICONE16/17/18, which attempted to categorize design rules in the code into linear design rules and non-linear design rules and to clarify corresponding design requirements and finite element analyses, in particular, those non-linear ones.
