Fretting-wear of nuclear heat exchange equipment is addressed at the design stage to demonstrate that components will meet their design life. Atomic Energy of Canada Limited (AECL) has developed a methodology to predict the progression of fretting-wear damage using the combination of predicted work-rates from a finite element model and experimentally-derived wear coefficients. The predicted progression of fretting-wear damage in a recent steam generator design is compared to inspection results from steam generators in a mid-life plant. The predicted wear is similar to the maximum observed wear. Therefore, AECL’s methodology is shown to provide reasonable quantitative predictions of the progression of fretting-wear.
Comparison of Predicted and Observed Fretting-Wear Damage in Nuclear Steam Generators
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Fisher, NJ, Han, Y, Gue´rout, FM, & Janzen, VP. "Comparison of Predicted and Observed Fretting-Wear Damage in Nuclear Steam Generators." Proceedings of the ASME 2005 Pressure Vessels and Piping Conference. Volume 4: Fluid Structure Interaction. Denver, Colorado, USA. July 17–21, 2005. pp. 535-545. ASME. https://doi.org/10.1115/PVP2005-71390
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