Xiang Liu

Profile Information

Name: Dr. Xiang Liu
Institution: Idaho National Laboratory
Position: Postdoctoral Research Associate
Affiliation: Idaho National Laboratory
h-Index: 8
ORCID: 0000-0002-2634-1888
Biography: Dr. Xiang Liu joined Idaho National Laboratory in June, 2018 as a postdoctoral research associate in the Advanced PIE department. Xiang received his Ph.D. in nuclear engineering in 2018 from the University of Illinois at Urbana-Champaign. He worked on radiation effects in advanced alloys for his Ph.D. study. Xiang got his M.S. from Peking University in 2012 and B.S. from Huazhong University of Science and Technology in 2009. He is interested in structural materials under extreme conditions.
Expertise: Advanced Alloys, APT, Austenitic, Dislocation Loops, EDS, Ferritic/Martensitic (F/M) Steels, Irradiated Microstructure, Nanoindentation, Oxide Dispersion-Strengthened Alloy (ODS), Post Irradiation Examination (PIE), Radiation Induced Segregation, TEM, X-Ray Diffraction

Publications:

		"Comparison of ion irradiation effects in PM-HIP and forged alloy 625" Caleb Clement, Yangyang Zhao, Patrick Warren, Xiang Liu, Sichuang Xue, David Gandy, Janelle Wharry, Journal of Nuclear Materials Vol. 558 2022 Link
		"Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91" Huan Yan, Xiang Liu, Lingfeng He, James Stubbins, Journal of Nuclear Materials Vol. 557 2021 Link
		"Effect of laser welding on deformation mechanisms in irradiated austenitic stainless steel" Janelle Wharry, Keyou Mao, Cheng Sun, Xiang Liu, Haozheng Qu, Aaron French, Paula Freyer, Frank Garner, Lin Shao, Journal of Nuclear Materials Vol. 528 2020 151878 Link Deformation mechanism of a laser weld on neutron irradiated AISI 304L stainless steel was studied by in-situ microcompression test at room temperature. The deformation-induced austenite-to-martensite phase transformation occurs in {101}-oriented grains in the irradiated base metal, while deformation twinning prevails in {101}-oriented grains in the weld heat affected zone (HAZ). A high number density of irradiation-induced voids in the base metal provides sufficient nucleation sites for the austenite-to-martensite phase transformation under compression at room temperature. A deformation map is established to predict critical twinning stress for face-centered-cubic (fcc) metals and alloys. Our results show that irradiation-induced voids can tailor the deformation mechanisms of austenitic stainless steel.
		"Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels" Andrew Hoffman, Maalavan Arivu, Haiming Wen, Li He, Kumar Sridharan, Xin Wang, Xiang Liu, Lingfeng He, Yaqiao Wu, Materialia Vol. 13 2020 Link Irradiation induced phase transformation of γ-austenite to α-ferrite has been observed in austenitic steels for the past several decades. This transformation can be detrimental to structural materials in a nuclear reactor environment as the increased fraction of the ferritic phase can increase corrosion and embrittlement and lead to stress corrosion cracking. This transformation is caused by both strain induced martensite transformation as well as radiation induced segregation and precipitation. In this study, two radiation tolerant nanostructured 304L austenitic steels (one ultrafine grained and one nanocrystalline) were manufactured using severe plastic deformation. These nanostructured 304L steels were compared to conventional coarse-grained 304L, after self-ion irradiation at 500°C up to a peak damage of 50 displacements per atom. Phase fraction after irradiation was analyzed using grazing incidence x-ray diffraction, precession electron diffraction, and electron backscatter diffraction. Nanostructured 304L steels showed significant resistance to irradiation induced austenite to ferrite transformation. This resistance was shown to be due to a decrease in defect formation, as well as a reduction in radiation induced segregation and precipitation.
		"Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding" Xiang Liu, Luca Capriotti, Tiankai Yao, Jason Harp, Michael Benson, Yachun Wang, Fei Teng, Lingfeng He, https://www.sciencedirect.com/science/article/pii/S002231152031196X#ack0001 Vol. 544 2021 Link
		"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.
		"Irradiation effects in high entropy alloys and 316H stainless steel at 300° C" Wei-Ying Chen, Xiang Liu, Jien-Wei Yeh, Krishnamurti Natesan, Journal of Nuclear Materials Vol. 512 2018 421-430 Link High entropy alloys (HEAs) have been considered for applications in nuclear reactors due to their promising mechanical properties, corrosion and radiation resistance. It has been suggested that sluggish diffusion kinetics and lattice distortion of HEAs can enhance the annihilation of irradiation-induced defects, giving rise to a higher degree of tolerance to irradiation damage. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages over conventional austenitic stainless steels (SS), we performed in-situ ion irradiation experiments with 1 MeV krypton at 300 °C on two HEAs and a 316H SS under an identical irradiation condition. The irradiation introduced a high density of dislocation loops in all materials, and the microstructural evolution as a function of dose was similar for HEAs and 316H SS. Nanoindentation tests showed that the degree of irradiation hardening was also comparable between them. The similar microstructural evolution and irradiation hardening behavior between the HEAs and 316H indicate that, at low temperatures (≤300 °C), the irradiation damage of fcc alloys is not sensitive to compositional variation and configurational entropy.
		"Irradiation effects on Al0. 3CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500° C" Wei-Ying Chen, Xiang Liu, Yiren Chen, Journal of Nuclear Materials Vol. 539 2020 To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al0.3CoCrFeNi and CoCrFeMnNi, and a conventional Type 316H stainless steel (SS) were irradiated with 1 MeV krypton ions at 500 °C up to 1 dpa, and examined in-situ with a transmission electron microscope (TEM). After irradiation, a high density of ordered L12 nanoparticles was observed in Al0.3CoCrFeNi. In contrast, no phase transformation was observed in CoCrFeMnNi and 316H SS. In the thin foil regions of TEM samples, stacking-fault tetrahedra were observed in the HEAs. In the thick foil regions, interstitial dislocation loops were observed for all alloys, and the areal loop density increased linearly with foil thickness. The Al0.3CoCrFeNi had the largest loop size (the lowest density), followed by the CoCrFeMnNi and then the 316H SS. The higher loop nucleation rate in the 316H SS was attributed to carbon content. The degree of irradiation hardening was slightly lower for the HEAs than for the 316H SS, which is a promising sign for the nuclear application of HEAs at high temperatures.
		"Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system" Yachun Wang, Xiang Liu, Daniel Murray, Fei Teng, Wen Jiang, Mukesh Bachhav, Laura Hawkins, Emmanuel Perez, Cheng Sun, Xianming Bai, Jie Lian, Colin Judge, John Jackson, Robert Carter, Lingfeng He, Materials Science & Engineering Vol. 849 2022
		"Microstructure and microchemistry of laser welds of irradiated austenitic steels" Keyou Mao, Aaron French, Xiang Liu, Lucille Giannuzzi, Cheng Sun, Megha Dubey, Paula Freyer, Jonathan Tatman, Frank Garner, Lin Shao, Janelle Wharry, Materials and Design Vol. 206 2021 Link
		"Microstructure of laser weld repairs of irradiated austenitic steels" Janelle Wharry, Keyou Mao, Aaron French, Xiang Liu, Yaqiao Wu, Cheng Sun, Paula Freyer, Jonathan Tatman, Lucille Giannuzzi, Frank Garner, Lin Shao, Materials & Design Vol. 206 2021 109764 Link
		"Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9" Huan Yan, Xiang Liu, James Stubbins, Journal of Nuclear Materials Vol. 557 2021 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:

	"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)

NSUF Articles:

DOE Awards 33 Rapid Turnaround Experiment Research Proposals - Projects total approximately $1.2 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, June 18, 2018 - Calls and Awards
RTE 1st Call Awards Announced - Projects total approximately $1.4 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-NE. Friday, February 8, 2019 - Calls and Awards
DOE Awards 31 RTE Proposals, Opens FY-20 1st Call - Projects total $1.1 million; Next proposals due 10/31 Awards will go to 22 principal investigators from universities, six from national laboratories, and three from foreign universities. Tuesday, September 17, 2019 - Calls and Awards, Announcement

Accomplishments

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Accomplished Author

Authored 10+ NSUF-supported publications

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Presented an NSUF-supported publication

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Submitted an RTE Proposal to NSUF

Accomplished Researcher

Awarded 3+ RTE Proposals

Top 5% Collaborator

Top 5% of all RTE Proposal collaborations

Accomplished Reviewer

Reviewed 10+ RTE Proposals

NSUF Supported Research

Atom probe tomography study of the fuel cladding chemical interaction (FCCI) layer in irradiated U-10Zr fuel with HT-9 cladding - FY 2019 RTE 1st Call, #1635

Investigation of fission gas bubble distribution, phase transformations, and bubble growth kinetics in a FFTF-irradiated U-10Zr fuel - FY 2019 RTE 3rd Call, #2899

Investigation of the synergistic effects of dual beam irradiation in an austenitic-ferritic duplex alloy - FY 2018 RTE 3rd Call, #1532

NSUF Research Collaborations

A comparative study of the radiation response of Fe–12Cr, Fe–14Cr, Fe–19Cr model alloys and a Fe–14Cr ODS alloy - FY 2018 RTE 2nd Call, #1433

A study of the tensile response of HT-9 alloys following ATR irradiation to doses between 0.01 and 10 dpa at 300, 450 and 550C - FY 2019 RTE 1st Call, #1670

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

Advanced microstructural characterization of irradiation-induced phase transformation in 304 steel - FY 2019 RTE 3rd Call, #2858

Atom Probe Tomography Study of Elemental Segregation and Precipitation in Ion-Irradiated Advanced Austenitic Alloy A709 - FY 2020 RTE 1st Call, #2963

High resolution (S)TEM/EDS characterization of neutron irradiated commercial Zr-Nb alloys - FY 2019 RTE 3rd Call, #2860

In-situ microstructural evolution and phase transition of irradiated and transient metallic fuel for M&S validation - FY 2020 RTE 2nd Call, #3088

Local Deformation Mechanism of Neutron-Irradiated NF709 Austenitic Stainless Steel - FY 2019 RTE 1st Call, #1652

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2019 RTE 2nd Call, #1802

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2018 RTE 1st Call, #1203

Nanoindentation of Phases in Irradiated and Control U-10Zr Fuels - FY 2019 RTE 1st Call, #1666

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

Towards Understanding Fuel Clad Chemical Interactions in FeCrAl Alloys - FY 2020 RTE 2nd Call, #4108