"Investigation of thermal aging effects on the tensile properties of Alloy 617 by in-situ synchrotron wide-angle X-ray scattering"
Xiang Liu, Kun Mo, Yinbin Miao, Kuan-Che Lan, Guangming Zhang, Wei-Ying Chen, Carolyn Tomchik, Rachel Seibert, Jeff Terry, James Stubbins,
Materials Science and Engineering: A
Vol. 651
2016
55-62
Link
The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 °C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property?microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions. |
"Advanced Investigations on the Strengthening Mechanisms in Austenitic ODS Stainless Steels" Wei-Ying Chen, Bai Cui, Kuan-Che Lan, Yinbin Miao, Kun Mo, International Conference on Fusion Reactor Materials ICFRM-18 November 5-10, (2017) | |
"In-Situ Synchrotron X-Ray Scattering Study on the Tensile Properties of Neutron Irradiated Ferritic/ Martensitic Alloys" Xiang Liu, Kuan-Che Lan, Meimei Li, Xuan Zhang, Chi Xu, James Stubbins, ANS Annual Meeting 2018 June 11-22, (2018) |
The Nuclear Science User Facilities (NSUF) is the U.S. Department of Energy Office of Nuclear Energy's only designated nuclear energy user facility. Through peer-reviewed proposal processes, the NSUF provides researchers access to neutron, ion, and gamma irradiations, post-irradiation examination and beamline capabilities at Idaho National Laboratory and a diverse mix of university, national laboratory and industry partner institutions.
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