Draft Fatigue Design Curves and Creep-Fatigue Interaction Diagram for the Alloy 709 Code Case

Abstract Due to its exceptional high-temperature properties, Alloy 709 (UNS 31025), an advanced austenitic stainless steel, has been selected by the U.S. Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies Program for qualification as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 for high temperature reactor applications. This initiative encompasses a comprehensive staged qualification plan for code cases with progressively longer design lives of 100,000, 300,000, and 500,000 hr, as long-term test data become available. A significant portion of this effort focuses on creating a data package that is essential for establishing material-specific design parameters for Alloy 709, intended for use in sodium fast reactors, molten salt reactors, and gas-cooled reactors. Data generation for the initial 100,000-hr code case was completed at the end of 2024. A key component of this data package involves conducting fatigue and creep-fatigue testing at elevated temperatures to develop fatigue design curves and the damage envelope for the creep-fatigue interaction diagram (D-diagram), both of which are common to all three code cases. Strain-controlled fatigue and creep-fatigue tests have been conducted on three commercial heats of Alloy 709 plates at the Oak Ridge National Laboratory and the Idaho National Laboratory. The comprehensive code case testing matrix includes strain ranges from 0.25 to 3%, hold times of up to 10 hr, and testing temperatures of 21, 427, 649, 760, 816, and 954 °C (70, 800, 1200, 1400, 1500, and 1750 °F). This paper summarizes the results of cyclic testing and provides a comprehensive analysis of the test results. It outlines the methodology for developing temperature-dependent fatigue curves and presents draft fatigue design curves for inclusion in the code cases. Furthermore, creep-fatigue test data with various tension hold times are used to support the development of the creep-fatigue interaction diagram. Both sets of design parameters are crucial for the code case data package.

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Published
2025-07-20
DOI
https://doi.org/10.1115/pvp2025-155653
Primary Topic
Fatigue and fracture mechanics
Type
article
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Draft Fatigue Design Curves and Creep-Fatigue Interaction Diagram for the Alloy 709 Code Case

Xuesong Fan, Ryann Elizabeth Bass, Heramb Prakash Mahajan, Yanli Wang et al.
Fatigue and fracture mechanics
article

Draft Fatigue Design Curves and Creep-Fatigue Interaction Diagram for the Alloy 709 Code Case

Xuesong Fan, Ryann Elizabeth Bass, Heramb Prakash Mahajan, Yanli Wang, T.-L. Sham
article en

Abstract

Abstract Due to its exceptional high-temperature properties, Alloy 709 (UNS 31025), an advanced austenitic stainless steel, has been selected by the U.S. Department of Energy, Office of Nuclear Energy, Advanced Reactor Technologies Program for qualification as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 for high temperature reactor applications. This initiative encompasses a comprehensive staged qualification plan for code cases with progressively longer design lives of 100,000, 300,000, and 500,000 hr, as long-term test data become available. A significant portion of this effort focuses on creating a data package that is essential for establishing material-specific design parameters for Alloy 709, intended for use in sodium fast reactors, molten salt reactors, and gas-cooled reactors. Data generation for the initial 100,000-hr code case was completed at the end of 2024. A key component of this data package involves conducting fatigue and creep-fatigue testing at elevated temperatures to develop fatigue design curves and the damage envelope for the creep-fatigue interaction diagram (D-diagram), both of which are common to all three code cases. Strain-controlled fatigue and creep-fatigue tests have been conducted on three commercial heats of Alloy 709 plates at the Oak Ridge National Laboratory and the Idaho National Laboratory. The comprehensive code case testing matrix includes strain ranges from 0.25 to 3%, hold times of up to 10 hr, and testing temperatures of 21, 427, 649, 760, 816, and 954 °C (70, 800, 1200, 1400, 1500, and 1750 °F). This paper summarizes the results of cyclic testing and provides a comprehensive analysis of the test results. It outlines the methodology for developing temperature-dependent fatigue curves and presents draft fatigue design curves for inclusion in the code cases. Furthermore, creep-fatigue test data with various tension hold times are used to support the development of the creep-fatigue interaction diagram. Both sets of design parameters are crucial for the code case data package.

United States Nuclear Regulatory Commission (US), Oak Ridge National Laboratory (US), Federal Energy Regulatory Commission (US), Idaho National Laboratory (US)
Openalex Percentile: Top 68%
Fatigue and fracture mechanics
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