Creep-fatigue design of Gen IV high temperature reactor plants

The service conditions expected for the GEN IV reactors pose significant challenges to structural materials selection and qualification efforts. The components will undergo varied service conditions which include exposure to high temperatures and high neutron doses compared to the service conditions in today's commercial reactors. A considerable damage mechanism for materials used in GEN IV components is predicted to be creep-fatigue (CF) damage, which arises due to startup and shutdown or power transients during normal operation. Design codes, such as RCC-MRx and ASME III NH, for GEN IV nuclear reactors use interaction diagram based method for CF assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. When standard laboratory CF tests with relatively large strain ranges and short hold or creep periods are assessed using the design code CF assessment procedures, the design codes seem to provide sufficient level of conservatism for safe design. But does the sufficient level of conservatism still remain with parameters relevant to power plant conditions? Can this type of test results and assessment and modelling methods emerging from them be extrapolated to GEN IV relevant conditions, where the stress and strain levels are lower, hold periods are significantly longer and lifetimes of components are expected to be about 60 years or more? Are there enough data and information on the long-term microstructural evolution and its effect on the creep and cyclic properties of e.g. P91 steel?