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Reports - Fast Reactor Structural

	Title	Link	ReportCategory	Description	ReportDate
	74406_A Summary of the Microstructural Characterization Results for the A709 Commercial Heats	74406_A Summary of the Microstructural Characterization Results for the A709 Commercial Heats	Fast Reactor Structural	To improve high temperature strength for advanced nuclear reactors, it is necessary to qualify additional structural materials for introduction into the American Society of Mechanical Engineers (AMSE) Boiler and Pressure Vessel Code (BPVC). Specifically, Alloy 709 was identified as a material for Sodium-cooled Fast Reactors to enhance time-dependent mechanical properties compared to 316H stainless steel, a code qualified high temperature alloy. Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) performed initial microstructure and mechanical property testing on a third, commercial heat of Alloy 709 stainless steel. The Alloy 709 plate was produced by Allegheny Technology Inc. (ATI) and in the solution annealed condition. The chemical composition was within the limits for an American Society of Testing and Materials (ASTM) A213 specification for seamless tube for material UNS S31025. Grain size was measured as an average ASTM size of 6.5, which was larger than the ASTM 7 required according to ASTM A213 for seamless tube. The chemical composition was within the limits for an ASTM A213, and the room temperature tensile tests and hardness data met ASTM A213 specifications. Creep-fatigue results showed similar behavior to previous heats of Alloy 709.	8/31/2023
	Effects of Notches on the Intermediate Creep-Rupture Life of Alloy 617 Weldment	Effects of Notches on the Intermediate Creep-Rupture Life of Alloy 617 Weldment	Fast Reactor Structural	A Code Case has recently been completed that adds Alloy 617 to Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, which covers high temperature nuclear components. However, additional information is needed to address concerns raised by the Nuclear Regulatory Commission that are not covered by the Code Case.	9/16/2020
	Status of INL Aged A709 Mechanical Testing	Status of INL Aged A709 Mechanical Testing	Fast Reactor Structural	A709 was selected to be qualified in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Idaho National Laboratory (INL), Argonne National Laboratory and Oak Ridge National Laboratory are collaborating to develop and qualify A709 plate. Aging A709 to form beneficial precipitates prior to service is investigated. One tensile test, five creep-rupture tests, and eight cyclic tests have been completed on aged A709. Two creep-rupture tests are in progress.	8/24/2020
	Single-Bar SMT Testing on Alloy 617 using Software Controls	Single-Bar SMT Testing on Alloy 617 using Software Controls	Fast Reactor Structural	NL has performed simplified model testing using a variable in the software control program to compare with the testing performed at ORNL, which uses hardware to modify the control signal. The INL tests compared well with the ORNL tests, showing similar stress and strain values, as well as similarly shaped hysteresis loops. This demonstrates that possibility of performing simplified model testing without any hardware modifications.	7/25/2020