High temperature nuclear reactors operating in the creep regime are designed to withstand numerous cyclic events. Current ASME code rules provide two basic paths for evaluating creep fatigue and ratcheting under these conditions; one based on full inelastic analysis intended to provide a representative stress and strain history and the other based on elastic material models with adjustments of varying complexity to account for inelastic stress and strain redistribution. More recent developments have used elastic-perfectly plastic analysis to bound the effects of cyclic service. However, these methods still rely on the separate evaluation of fatigue and creep damage utilizing a damage interaction diagram. There is a procedure under current development that uses creep-fatigue data from key feature test articles directly without the use of the damage interaction diagram. However, it requires a reasonable representation of the strain range in a structure as an input. This work develops a simplified procedure based on elastic perfectly-plasticity analysis that can be used to represent the strain range in a structure in the steady state under cyclic loading conditions.
The Role of Material Modeling on Strain Range Estimation for Elevated Temperature Cyclic Life Evaluation
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Messner, MC, Jetter, RI, Sham, T, & Wang, Y. "The Role of Material Modeling on Strain Range Estimation for Elevated Temperature Cyclic Life Evaluation." Proceedings of the ASME 2018 Pressure Vessels and Piping Conference. Volume 1B: Codes and Standards. Prague, Czech Republic. July 15–20, 2018. V01BT01A011. ASME. https://doi.org/10.1115/PVP2018-84100
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