This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of Japan sodium cooled fast reactor (JSFR), there may be dissimilar welded joints between ferritic and austenitic steels especially in intermediate heat exchangers (IHX) and steam generators (SG). Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and type 304 austenitic stainless steel (304SS) was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the heat affected zone (HAZ) in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in base metal (BM) of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
- Nuclear Engineering Division
Thermal Fatigue Test on Dissimilar Welded Joint Between Gr.91 and 304SS
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Wakai, T, Kobayashi, S, Kato, S, Ando, M, & Takasho, H. "Thermal Fatigue Test on Dissimilar Welded Joint Between Gr.91 and 304SS." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 2: Plant Systems, Structures, Components and Materials. Shanghai, China. July 2–6, 2017. V002T03A008. ASME. https://doi.org/10.1115/ICONE25-66099
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