ASME STP-NU-039 PDF

The GEN IV reactor concepts require structural components to operate at high temperatures in a regime where creep damage may occur and cracks may grow. The U.S. Nuclear Regulatory Commission (NRC) has identified the lack of a quantitative methodology for evaluating creep and creep crack growth as a shortcoming of the ASME Subsection NH (Class 1 Components in Elevated Temperature Service) standard [1]. The development of elastic-plastic fracture mechanics methods and the concepts of leak-before-break (LBB) were led by the needs of the nuclear industry. These crack assessment methods are now well established and used routinely in PWR and BWR plant extension applications and new designs. Quantitative creep and creep-fatigue crack growth assessment procedures are now needed for these GEN IV developments.