Creep and Creep-Fatigue of Alloy 617 Weldments

Technical Report · Fri Aug 01 04:00:00 EDT 2014

DOI:https://doi.org/10.2172/1168621· OSTI ID:1168621

Wright, Jill K. ^[1]; Carroll, Laura J. ^[1]; Wright, Richard N. ^[1]

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1. Idaho National Lab. (INL), Idaho Falls, ID (United States)

Alloy 617 is the primary candidate material for the heat exchanger of a very high temperature gas cooled reactor intended to operate up to 950°C. While this alloy is currently qualified in the ASME Boiler and Pressure Vessel Code for non-nuclear construction, it is not currently allowed for use in nuclear designs. A draft Code Case to qualify Alloy 617 for nuclear pressure boundary applications was submitted in 1992, but was withdrawn prior to approval. Prior to withdrawal of the draft, comments were received indicating that there was insufficient knowledge of the creep and creep-fatigue behavior of Alloy 617 welds. In this report the results of recent experiments and analysis of the creep-rupture behavior of Alloy 617 welds prepared using the gas tungsten arc process with Alloy 617 filler wire. Low cycle fatigue and creep-fatigue properties of weldments are also discussed. The experiments cover a range of temperatures from 750 to 1000°C to support development of a new Code Case to qualify the material for elevated temperature nuclear design. Properties of the welded material are compared to results of extensive characterization of solution annealed plate base metal.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1168621
• Report Number(s): INL/EXT--14-32966
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Alloy 617
Creep