摘要
Fatigue verification of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, which is often discussed in connection to power uprate and life-extension of aging reactors in recent years, is dealt with. Key parameters involved in the fatigue verification, e.g., the alternating stress intensity Salt, the penalty factor Ke and the cumulative damage factor U, and relevant computational procedures applicable for the assessment of low-cycle fatigue failure using strain-controlled data, are particularly addressed. A so-called simplified elastic-plastic discontinuity analysis for alternative verification when fatigue requirements found unsatisfactory, and the procedures provided in NB-3600 for evaluating the alternating stress intensity S,j,, are reviewed in detail. An in-depth discussion is given to alternative procedures suggested earlier by the authors using nonlinear finite element analyses, which uses a nonlinear finite element analysis for directly determining the alternating stress, thus eliminating uncertainties resulted from the use of the penalty factor Ke. Using this alternative, unavoidable plastic strains can be correctly taken into account in a computationally affordable way, and the reliability of the verification will not be affected by uncertainties introduced in the simplified elastic-plastic analysis.
