Miao Song

Profile Information
Dr. Miao Song
University of Michigan
Assistant Reserach Scientist
"Enhanced radiation tolerance in immiscible Cu/Fe multilayers with coherent and incoherent layer interfaces" Youxing Chen, Engang Fu, Kaiyuan Yu, Miao Song, Yue Liu, Yongqiang Wang, Haiyan Wang, Xinghang Zhang, Journal of Materials Research Vol. 30 2015 1300 Link
"Grain refinement mechanisms and strength-hardness correlation of ultra-fine grained grade 91 steel processed by equal channel angular extrusion" Miao Song, Cheng Sun, Youxing Chen, Zhongxia Shang, Jin Li, Zhe Fan, Karl Hartwig, Xinghang Zhang, International Journal of Pressure Vessels and Piping Vol. 172 [unknown] 212-219 Link
"In situ Observation of Defect Annihilation in Kr Ion-Irradiated Bulk Fe/Amorphous-Fe2 Zr Nanocomposite Alloy" Kaiyuan Yu, Zhe Fan, Youxing Chen, Miao Song, Yue Liu, Haiyan Wang, Mark Kirk, Meimei Li, Xinghang Zhang, Materials Research Letters Vol. 3 2014 35 Link
"In situ studies of Kr ion irradiation response of Fe/Y2O3 nanolayers" Youxing Chen, Liang Jiao, Cheng Sun, Miao Song, Kaiyuan Yu, Yue Liu, Meimei Li, Haiyan Wang, Xinghang Zhang, Journal of Nuclear Materials Vol. 452 2014 321 Link
"In situ Study of Defect Migration Kinetics and Self-Healing of Twin Boundaries in Heavy Ion Irradiated Nanotwinned Metals" Jin Li, Kaiyuan Yu, Youxing Chen, Miao Song, Haiyan Wang, Mark Kirk, Meimei Li, Xinghang Zhang, Nano Letters Vol. 15 2015 2922 Link
"Insights into the stress corrosion cracking of solution annealed alloy 690 in simulated pressurized water reactor primary water under dynamic straining" wenjun kuang, Miao Song, Gary Was, Acta Materialia Vol. 151 2018 321-333
The intergranular attack near both stagnant and active stress corrosion crack (SCC) tips of solution annealed alloy 690 were characterized after constant extension rate tensile tests in simulated pressurized water reactor (PWR) primary water containing 18 cc H2/kg H2O. In both cracks, an intergranular oxide, composed of NiO and Cr2O3, was observed beyond the crack tip with an adjacent Cr depleted grain boundary migration zone. The stagnant crack has a compact Cr2O3 covering the crack tip and the adjacent grain boundary migration zone is deep and free of oxidation, while the active crack has porous penetrative oxide extending into the migration zone. The ability to form a compact Cr2O3 film at the crack tip is important to the IGSCC resistance and may be dictated by the efficiency of Cr diffusion up the migrated grain boundary. When a compact Cr2O3 film does not form at the crack tip, the Cr depleted grain boundary migration zone is subject to penetrative oxidation via the inward diffusion of oxygen, making the zone susceptible to crack propagation. Thus, even a high Cr content cannot guarantee the immunity of a Ni base alloy to SCC during dynamic straining.
"Irradiation assisted stress corrosion cracking of commercial and advanced alloys for light water reactor core internals" mi wang, Miao Song, Calvin Lear, Gary Was, Journal of Nuclear Materials Vol. 515 2019 52-70 Link
"Radiation tolerance of commercial and advanced alloys for core internals: a comprehensive microstructural characterization" Miao Song, Calvin Lear, Chad Parish, Mi Wang, Gary Was, Journal of Nuclear Materials Vol. 510 2018 396-413 Link
Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.
"Tailoring the strength and ductility of T91 steel by partial tempering treatment" Zhongxia Shang, Jie Ding, Cuncai Fan, Miao Song, Jin Li, Qiang Li, Sichuang Xue, Karl Hartwig, Xinghang Zhang, Acta Materialia Vol. 169 [unknown] 209-224 Link
NSUF Articles:
RTE 2nd Call Awards Announced - Projects total approximately $1.6 million These project awards went to principal investigators from 26 U.S. universities, eight national laboratories, two British universities, and one Canadian laboratory. Tuesday, May 14, 2019 - Calls and Awards