Yinbin Miao

Profile Information

Name: Dr. Yinbin Miao
Institution: Argonne National Laboratory
Position: Principal Materials Scientist
h-Index: 24
ORCID: 0000-0002-3128-4275
Expertise: Accident Tolerant Fuel, Advanced Fuels, APT, Ceramic Fuel, FIB, Oxide Dispersion-Strengthened Alloy (ODS), Synchrotron Scattering

Publications:

		"Atom probe study of irradiation-enhanced a' precipitation in neutron-irradiated Fe–Cr model alloys" Wei-Ying Chen, Jian Gan, Stuart Maloy, Kun Mo, Maria Okuniewski, James Stubbins, Yinbin Miao, Yaqiao Wu, Carolyn Tomchik, Journal of Nuclear Materials Vol. 462 2015 242-249 Link Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 1023 to 2.7 × 1024 1/m3, depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.
		"Bubble morphology in U3Si2 implanted by high-energy Xe ions at 300ºC" Yinbin Miao, Jason Harp, Kun Mo, Shaofei Zhu, Tiankai Yao, Jie Lian, Abdellatif Yacout, Journal of Nuclear Materials Vol. 495 2017 146-153 Link The microstructure modifications of a high-energy Xe implanted U3Si2, a promising accident tolerant fuel candidate, were characterized and are reported upon. The U3Si2 pellet was irradiated at Argonne Tandem Linac Accelerator System (ATLAS) by an 84 MeV Xe ion beam at 300 °C. The irradiated specimen was then investigated using a series of transmission electron microscopy (TEM) techniques. A dense distribution of bubbles were observed near the range of the 84 MeV Xe ions. Xe gas was also found to accumulate at multiple types of sinks, such as dislocations and grain boundaries. Bubbles aggregated at those sinks are slightly larger than intragranular bubbles in lattice. At 300 °C, the gaseous swelling strain is limited as all the bubbles are below 10 nm, implying the promising fission gas behavior of U3Si2 under normal operating conditions in light water reactors (LWRs).
		"Cross section TEM characterization of high-energy-Xe-irradiated U-Mo" Laura Jamison, Yinbin Miao, Bei Ye, Sumit Bhattacharya, Gerard Hofman, Abdellatif Yacout, Journal of Nuclear Materials Vol. 488 2017 134-142 Link U-Mo alloys irradiated with 84 MeV Xe ions to various doses were characterized with transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) techniques. The TEM thin foils were prepared perpendicular to the irradiated surface to allow a direct observation of the entire region modified by ions. Therefore, depth-selective microstructural information was revealed. Varied irradiation-induced phenomena such as gas bubble formation, phase reversal, and recrystallization were observed at different ion penetration depths in U-Mo.
		"In situ synchrotron investigation of grain growth behavior of nano-grained UO2" Jie Lian, Yinbin Miao, Kun Mo, Jun-Sang Park, Tiankai Yao, Jonathan Almer, Sumit Bhattacharya, Abdellatif Yacout, Scripta Materialia Vol. 131 2017 29-32 Link The study of grain growth kinetics in nano-grained UO2 samples is reported. Dense nano-grained UO2 samples with well-controlled stoichiometry and grain size were fabricated using the spark plasma sintering technique. To determine the grain growth kinetics at elevated temperatures, a synchrotron wide-angle X-ray scattering (WAXS) study was performed in situ to measure the real-time grain size evolution based on the modified Williamson-Hall analysis. The unique grain growth kinetics of nanocrystalline UO2 at 730 °C and 820 °C were observed and explained by the difference in mobility of various grain boundaries.
		"In situ TEM Ion Irradiation Investigations on U3Si2 at LWR Temperatures" Jason Harp, Yinbin Miao, Kun Mo, Sumit Bhattacharya, Peter Baldo, Abdellatif Yacout, Journal of Nuclear Materials Vol. 484 2017 168-173 Link The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.
		"In-situ TEM study of the ion irradiation behavior of U3Si2 and U3Si5" Tiankai Yao, Bowen Gong, Yinbin Miao, Jason Harp, Jie Lian, Journal of Nuclear Materials Vol. 511 2018 56-63 Link U3Si2 and U3Si5 are two important uranium silicide phases currently under extensive investigation as potential fuel forms or components for light water reactors (LWRs) to enhance accident tolerance. In this paper, their irradiation behaviors are studied by ion beam irradiations with various ion mass and energies, and their microstructure evolution is investigated by in-situ transmission electron microscopy (TEM). U3Si2 can easily be amorphized by ion beam irradiations (by 1 MeV Ar2+ or Kr2+) at room temperature with the critical amorphization dose less than 1 dpa. The critical amorphization temperatures of U3Si2 irradiated by 1 MeV Kr2+ and 1 MeV Ar2+ ion are determined as 580 ± 10 K and 540 ± 5 K, respectively. In contrast, U3Si5 remains crystalline up to 8 dpa at room temperature and is stable against ion irradiation-induced amorphization up to ∼50 dpa by either 1 MeV Kr2+ or 150 KeV Kr+ at 623 K. These results provide valuable experimental data to guide future irradiation experiments, support the relevant post irradiation examination, and serve as the experimental basis for the validation of advanced fuel performance models.
		"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.
		"Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Stuart Maloy, James Stubbins, Journal of Nuclear Materials Vol. 490 2017 305-316 Link In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 × 1015 ions/cm2 at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black-dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 × 1015 ions/cm2, about 51% of the loops were type for the 673 K irradiation, and the dominant loop type was for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100–150 MPa.
		"Microstructure investigations of U3Si2 implanted by high-energy Xe ions at 600°C" Yinbin Miao, Jason Harp, Kun Mo, Yeon Soo Kim, Shaofei Zhu, Abdellatif Yacout, Journal of Nuclear Materials Vol. 503 2018 314-322 Link The microstructure investigations on a high-energy Xe-implanted U3Si2 pellet were performed. The promising accident tolerant fuel (ATF) candidate, U3Si2, was irradiated by 84?MeV Xe ions at 600?°C at Argonne Tandem Linac Accelerator System (ATLAS). The characterizations of the Xe implanted sample were conducted using advanced transmission electron microscopy (TEM) techniques. An oxidation layer was observed on the sample surface after irradiation under the 10-5?Pa vacuum. The study on the oxidation layer not only unveils the readily oxidation behavior of U3Si2 under high-temperature irradiation conditions, but also develops an understanding of its oxidation mechanism. Intragranular Xe bubbles with bimodal size distribution were observed within the Xe deposition region of the sample induced by 84?MeV Xe ion implantation. At the irradiation temperature of 600?°C, the gaseous swelling strain contributed by intragranular bubbles was found to be insignificant, indicating an acceptable fission gas behavior of U3Si2 as a light water reactor (LWR) fuel operating at such a temperature.
		"Nano-crystallization induced by high-energy heavy ion irradiation in UO2" Yinbin Miao, Tiankai Yao, Jie Lian, Shaofei Zhu, Sumit Bhattacharya, Aaron Oaks, Adbdellatif Yacout, Scripta Materialia Vol. 155 2018 169-174 Link Advanced microstructure investigations of the high-burnup structure (HBS) in UO2 produced by high-dose 84 MeV Xe ion irradiation are reported. Spark plasma sintered micro-grained UO2 was irradiated to 1357 dpa at 350 °C. The characteristic nano-grains and micro-pores of the HBS were formed. The grain size and grain boundary misorientation distributions of the HBS were measured using transmission electron microscopy based orientation imaging microscopy. Grain polygonization due to accumulation of radiation-induced dislocations was found to be the mechanism of nano-crystallization. The morphology of Xe bubbles was quantitatively investigated. This study provides crucial references for advanced fuel performance modeling of high-burnup UO2.
		"Neutron irradiation effects in Fe and Fe-Cr at 300°C" Wei-Ying Chen, Yinbin Miao, Jian Gan, Maria Okuniewski, Stuart Maloy, James Stubbins, Acta Materialia Vol. 111 2016 407-416 Link Fe and Fe-Cr (Cr = 10-16 at.%) specimens were neutron-irradiated at 300 °C to 0.01, 0.1 and 1 dpa. The TEM observations indicated that the Cr significantly reduced the mobility of dislocation loops and suppressed vacancy clustering, leading to distinct damage microstructures between Fe and Fe-Cr. Irradiation-induced dislocation loops in Fe were heterogeneously observed in the vicinity of grown-in dislocations, whereas the loop distribution observed in Fe-Cr is much more uniform. Voids were observed in the irradiated Fe samples, but not in irradiated Fe-Cr samples. Increasing Cr content in Fe-Cr results in a higher density, and a smaller size of irradiation-induced dislocation loops. Orowan mechanism was used to correlate the observed microstructure and hardening, which showed that the hardening in Fe-Cr can be attributed to the formation of dislocation loops and α′ precipitates.
		"Phase decomposition and bubble evolution in Xe implanted U3Si2 at 450°C" Yinbin Miao, Journal of Nuclear Materials Vol. 518 [unknown] 108-116 Link
		"Radiation resistance of oxide dispersion strengthened alloys: Perspectives from in situ observations and rate theory calculations" Xiang Liu, Yinbin Miao, Meimei Li, Marquis Kirk, Guangming Zhang, Shigeharu Ukai, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 148 2018 33-36 Link Here, in situ ion irradiation and rate theory calculations were employed to directly compare the radiation resistance of an oxide dispersion strengthened alloy with that of a conventional ferritic/martensitic alloy. Compared to the rapid buildup of dislocation loops, loop growth, and formation of network dislocations in the conventional ferritic/martensitic alloy, the superior radiation resistance of the oxide dispersion strengthened alloy is manifested by its stable dislocation structure under the same irradiation conditions. The results are consistent with rate theory calculations, which show that high-density nanoparticles can significantly reduce freely migrating defects and suppress the buildup of clustered defects.
		"Stability of nanoclusters in an oxide dispersion strengthened alloy under neutron irradiation" Xiang Liu, Yinbin Miao, Yaqiao Wu, Stuart Maloy, James Stubbins, Scripta Materialia Vol. 138 2017 57-61 Link Here, we report atom probe tomography results of the nanoclusters in a neutron-irradiated oxide dispersion strengthened alloy. Following irradiation to 5 dpa at target temperatures of 300 °C and 450 °C, fewer large nanoclusters were found and the residual nanoclusters tend to reach an equilibrium Guinier radius of 1.8 nm. With increasing dose, evident decrease in peak oxygen and titanium (but not yttrium) concentrations in the nanoclusters was observed, which was explained by atomic weight, solubility, diffusivity, and chemical bonding arguments. The chemical modifications indicate the equilibrium size is indeed a balance of two competing processes: radiation enhanced diffusion and collisional dissolution.
		"The comparison of microstructures and mechanical properties between 14 cr-Al and 14Cr-Ti ferritic ODS alloys" Yinbin Miao, Kun Mo, James Stubbins, Guangming Zhang, Zhangjian Zhou, Shaofu Li, Xiang Liu, Man Wang, Jun-Sang Park, Jonathan Almer, Materials & Design Vol. 98 2016 61-67 Link In this study, two kinds of 14Cr ODS alloys (14Cr-Al and 14Cr-Ti) were investigated to reveal the different effects between Al and Ti on the microstructures and mechanical properties of 14Cr ferritic ODS alloys. The microstructure information such as grains, minor phases of these two alloys has been investigated by high-energy X-ray diffraction and transmission electron microscopy (TEM). The in situ synchrotron X-ray diffraction tensile test was applied to investigate the mechanical properties of these two alloys. The lattice strains of different phases through the entire tensile deformation process in these two alloys were analyzed to calculate their elastic stresses. From the comparison of elastic stress, the strengthening capability of Y2Ti2O7 is better than TiN in 14Cr-Ti, and the strengthening capability of YAH is much better than YAM and AlN in 14Cr-Al ODS. The dislocation densities of 14Cr-Ti and 14Cr-Al ODS alloys during tensile deformation were also examined by modified Williamson-Hall analyses of peak broadening, respectively. The different increasing speed of dislocation density with plastic deformation reveals the better strengthening effect of Y-Ti-O particles in 14Cr-Ti ODS than that of Y-Al-O particles in 14Cr-Al ODS alloy.

Presentations:

	" Microstructure Investigations of U3Si2 Irradiated by Heavy Ions at LWR Temperatures" Yinbin Miao, Kun Mo, 2017 ANS Annual Meeting [unknown]
	"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)
	"Atom probe analysis of a neutron irradiated Fe-14Cr model alloy" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, Yaqiao Wu, ICFRM 2013 January 1-9, (2013)
	"Ferritic/Martensitic Steels for Fast Reactor Applications" Yinbin Miao, James Stubbins, Huan Yan, International Workshop on Fast Reactor Metallic Fuels RD Program October 10-11, (2017)
	"Microstructure and Mechanical Property Studies on Neutron-Irradiated Ferritic FeCr Model Alloys" Jian Gan, Stuart Maloy, Yinbin Miao, Kun Mo, James Stubbins, Yaqiao Wu, TMS Annual Meeting February 16-20, (2014)
	"Neutron and Ion Irradiation Studies on Advanced Steels Using the Nuclear Science User Facilities" Wei-Ying Chen, Yinbin Miao, James Stubbins, Transactions of the American Nuclear Society June 11-15, (2017)
	"Neutron irradiation damage in ferritic ODS steel MA957" Yinbin Miao, James Stubbins, TMS 2017 146th Annual Meeting & Exhibition February 26-2, (2017)
	"Neutron Radiation Response of Ferritic ODS Steel MA957" Yinbin Miao, James Stubbins, International Conference on Fusion Reactor Materials ICFRM-18 November 5-10, (2017)

NSUF Articles:

DOE Awards 33 Rapid Turnaround Experiment Research Proposals - Projects total approximately $1.5 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Monday, May 14, 2018 - Calls and Awards
DOE awards 39 RTE Projects - Projects total approximately $1.3 million Thursday, February 1, 2018 - Calls and Awards
DOE Awards 37 RTE Proposals - Awarded projects total nearly $1.4M in access awards Tuesday, July 14, 2020 - News Release, Calls and Awards
NSUF awards 30 Rapid Turnaround Experiment proposals - Approximately $1.53M has been awarded. Tuesday, June 14, 2022 - Calls and Awards

Accomplishments

NSUF Profile

This NSUF Profile is 55% Complete

Accomplished Author

Authored 10+ NSUF-supported publications

NSUF Presenter

Presented an NSUF-supported publication

Top 5% Submitter

Top 5% of all RTE Proposal submissions

Top 5% Researcher

Top 5% of all RTE Proposals awarded

Accomplished Collaborator

Collaborated on 3+ RTE Proposals

Accomplished Reviewer

Reviewed 10+ RTE Proposals

NSUF Supported Research

Advanced Investigations on the Low-dose Neutron-irradiated MA957 - FY 2015 RTE 1st Call, #538

Fission Gas Bubble Characterizations of High-Energy Xe Implanted U3Si2 - FY 2017 RTE 1st Call, #766

In Situ TEM Investigation on Fission Gas Behavior in U-10Zr - FY 2018 RTE 1st Call, #1226

Microstructural Modifications in U-10Zr Irradiated by High-Energy Xe Ions - FY 2018 RTE 2nd Call, #1457

Nano-Indentation Investigations on Nano-grained UO2 with and without High-Energy Ion Irradiation - FY 2016 RTE 1st Call, #618

Post-irradiation Examinations on a Ferritic ODS Steel Irradiated at 450C - FY 2015 RTE 1st Call, #539

Radiation Tolerance of Advanced Multilayered Coating: an in-situ TEM Investigation - FY 2020 RTE 2nd Call, #4207

Radiation Tolerance of Advanced Multilayered Coatings under Transient Temperature Conditions - FY 2022 RTE 1st Call, #4465

Xe Bubble Evolution in U3Si2: an in situ TEM Investigation - FY 2017 RTE 1st Call, #767

NSUF Research Collaborations

In-Situ Observation of Radiation-Induced Phase Transformation in U-Mo - FY 2019 RTE 1st Call, #1700

Investigation of irradiation-induced recrystallization in U-Mo fuel - FY 2017 RTE 3rd Call, #1116

Irradiation Influence on Alloys Fabricated by Powder Metallurgy and Hot Isostatic Pressing for Nuclear Applications - FY 2015 CINR, #1709

Microstructure characterization of U-Mo/Al fuel irradiated with Xe - FY 2016 RTE 1st Call, #627