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 (RIS), X-Ray Diffraction (XRD)

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
		"Comparison of ion irradiation effects in PM-HIP and forged alloy 625" Caleb Clement, Patrick Warren, Yangyang Zhao, Xiang Liu, David Gandy, Janelle Wharry, Sichuang Xue, Journal of Nuclear Materials Vol. 558 [unknown] Link The nuclear industry has growing interest in replacing forgings with structural components fabricated by powder metallurgy with hot isostatic pressing (PM-HIP), owing to their chemical homogeneity, uniform grain structure, and near-net shape production. This study compares the ion irradiation response of PM-HIP and forged Alloy 625, over 50 and 100 dpa, 400 °C and 500 °C. Microstructure is characterized using down-zone bright-field scanning transmission electron microscopy (DZBFSTEM), and hardening is characterized using nanoindentation. PM-HIP Alloy 625 has a lower initial dislocation line density, resulting in a more rapid onset of dislocation loop growth and unfaulting than the forged material. But the total defect population (i.e. loop line length plus dislocation density) is insensitive to fabrication method. This finding shows promise for the eventual qualification of PM-HIP alloys for nuclear applications.
		"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.
		"Effect of proton pre-irradiation on corrosion of Zr-0.5Nb model alloys with different Nb distributions" Zefeng Yu, Taeho Kim, Mukesh Bachhav, Xiang Liu, Adrien Couet, Lingfeng He, Corrosion Science Volume 173 Vol. 173 2020 108790 Link The effect of proton irradiation on corrosion rate of α-annealed and β-quenched Zr-0.5Nb alloys is investigated. The major focuses of this study are to understand i) if the nucleation of irradiation-induced platelets (IIPs)/nanoclusters requires dissolution of Nb-rich native precipitates, ii) if the irradiated native precipitates and interlaths are stable in the oxide, and iii) how much Nb content in the solid solution is suitable to lower the corrosion rate for Zr-Nb alloys. To answer these questions, the major characterization techniques used in this study are APT and (S)TEM/EDS to study the microstructure and microchemistry evolution following irradiation and oxidation.
		"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 Link 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 Link
		"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

Additional Publications:

	"Hot compression bonding of a 9Cr oxide dispersion strengthened alloy and a 9Cr reduced-activation ferritic/martensitic alloy" Bin Xu, Mingyue Sun, Xiang Liu, Dianzhong Li, Jianqiang Wang, [2025] Journal of Materials Science & Technology · DOI: 10.1016/j.jmst.2024.02.087
	"Evaluation of elastic constants of M23C6 and M7C3 embedded in Fe-Cr-C alloys using in-situ XRD tensile test and self-consistent model" Xiang Liu, Kuan-Che Lan, Huan Yan, Xiao Pan, Xuan Zhang, Jun-sang Park, Meimei Li, Jonathan Almer, James Stubbins, Hoon Lee, [2024] Materialia · DOI: 10.1016/j.mtla.2024.102274
	"A Study on the Radiation Resistance Performance of an Al2O3 Composite Tritium Permeation Barrier and Zirconium-Based Tritium-Absorbing Materials" Rui Shu, Yinghong Li, Long Wang, Runjie Fang, Lihong Nie, Qisen Ren, Xiang Liu, Jing Hu, Shaohong Zhang, Changzheng Li, [2024] Materials · DOI: 10.3390/ma17225600 The permeation of tritium from secondary neutron source rods in nuclear power plants presents a significant and unavoidable safety concern both for internal equipment and the external environment. This study primarily explores two feasible strategies for tritium permeation barriers: coating stainless steel surfaces with tritium permeation barrier (TPB) materials and utilizing materials with excellent tritium absorption properties. Through external ion irradiation tests, a comparative analysis was conducted on the tritium permeation performance, morphology, and nanohardness changes in two tritium-resistant designs, specifically Cr2O3/Al2O3 composite coatings and a zirconium-based tritium-absorbing material under varying irradiation doses. The results indicate that both approaches exhibit exceptional radiation resistance, maintaining an effective tritium permeation reduction factor (PRF) even after irradiation.
	"A Study on the Radiation Resistance Performance of an Al₂O₃ Composite Tritium Permeation Barrier and Zirconium-Based Tritium-Absorbing Materials" Rui Shu, Yinghong Li, Long Wang, Runjie Fang, Lihong Nie, Qisen Ren, Xiang Liu, Jing Hu, Shaohong Zhang, Changzheng Li, [2024] Materials · DOI: 10.3390/ma17225600 The permeation of tritium from secondary neutron source rods in nuclear power plants presents a significant and unavoidable safety concern both for internal equipment and the external environment. This study primarily explores two feasible strategies for tritium permeation barriers: coating stainless steel surfaces with tritium permeation barrier (TPB) materials and utilizing materials with excellent tritium absorption properties. Through external ion irradiation tests, a comparative analysis was conducted on the tritium permeation performance, morphology, and nanohardness changes in two tritium-resistant designs, specifically Cr2O3/Al2O3 composite coatings and a zirconium-based tritium-absorbing material under varying irradiation doses. The results indicate that both approaches exhibit exceptional radiation resistance, maintaining an effective tritium permeation reduction factor (PRF) even after irradiation.
	"Study on the Radiation Resistance Performance of Al₂O₃ Composite Tritium-Permeation-Barrier and Zirconium-Based Tritium Absorbing Materials" Rui Shu, Yinghong Li, Long Wang, Runjie Fang, Lihong Nie, Qisen Ren, Xiang Liu, Jing Hu, Shaohong Zhang, Changzheng Li, [2024] · DOI: 10.20944/preprints202410.1160.v1 The permeation of tritium from secondary neutron source rods in nuclear power plants presents a significant and unavoidable safety concern for both internal equipment and the external environment. This study primarily explores two feasible strategies for tritium permeation barriers: coating stainless steel surfaces with tritium-permeation-barrier (TPB) materials and utilizing materials with excellent tritium absorption properties. Through external ion irradiation tests, a comparative analysis was conducted on the tritium permeation performance, microstructure, and nano-hardness changes of two tritium-resistant designs, specifically Cr2O3/Al2O3 composite coatings, and a zirconium-based tritium-absorbing material under varying irradiation doses. The results indicate that both approaches exhibit exceptional radiation resistance, maintaining an effective tritium permeation reduction factor (PRF) even after irradiation.
	"Void swelling in additively manufactured 316L stainless steel with hafnium composition gradient under self-ion irradiation" Jingfan Yang, Xiang Liu, Laura R. Hawkins, Zhijie Jiao, Lingfeng He, Yongfeng Zhang, Daniel Schwen, Xiaoyuan Lou, Miao Song, [2023] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2023.154351
	"Microstructural stability of a 9Cr oxide dispersion strengthened alloy under thermal aging at high temperatures" Sheng Liu, Bin Xu, Mingyue Sun, Xiang Liu, Dianzhong Li, Yiyi Li, Jianqiang Wang, [2023] Journal of Alloys and Compounds · DOI: 10.1016/j.jallcom.2022.167691 · ISSN: 0925-8388
	"A novel approach on designing ultrahigh burnup metallic TWR fuels: Upsetting the current technological limits" Yuwen Xu, Jie Qiu, Xiang Liu, Chunyang Wen, Zhengyu Qian, Wenbo Liu, Wei Yan, Yanfen Li, Zhaohao Wang, Shilun Zheng, Shaoqiang Guo, Tan Shi, Chenyang Lu, Junli Gou, Liangxing Li, Jianqiang Shan, James F. Stubbins, Long Gu, Di Yun, Linna Feng, [2022] MRS Bulletin · DOI: 10.1557/s43577-022-00420-4 · ISSN: 0883-7694 Abstract The grand challenge of “net-zero carbon” emission calls for technological breakthroughs in energy production. The traveling wave reactor (TWR) is designed to provide economical and safe nuclear power and solve imminent problems, including limited uranium resources and radiotoxicity burdens from back-end fuel reprocessing/disposal. However, qualification of fuels and materials for TWR remains challenging and it sets an “end of the road” mark on the route of R&D of this technology. In this article, a novel approach is proposed to maneuver reactor operations and utilize high-temperature transients to mitigate the challenges raised by envisioned TWR service environment. Annular U-50Zr fuel and oxidation dispersion strengthened (ODS) steels are proposed to be used instead of the current U-10Zr and HT-9 ferritic/martensitic steels. In addition, irradiation-accelerated transport of Mn and Cr to the cladding surface to form a protective oxide layer as a self-repairing mechanism was discovered and is believed capable of mitigating long-term corrosion. This work represents an attempt to disruptively overcome current technological limits in the TWR fuels. Impact statement After the Fukushima accident in 2011, the entire nuclear industry calls for a major technological breakthrough that addresses the following three fundamental issues: (1) Reducing spent nuclear fuel reprocessing demands, (2) reducing the probability of a severe accident, and (3) reducing the energy production cost per kilowatt-hour. An inherently safe and ultralong life fast neutron reactor fuel form can be such one stone that kills the three birds. In light of the recent development findings on U-50Zr fuels, we hereby propose a disruptive, conceptual metallic fuel design that can serve the following purposes at the same time: (1) Reaching ultrahigh burnup of above 40% FIMA, (2) possessing strong inherent safety features, and (3) extending current limits on fast neutron irradiation dose to be far beyond 200 dpa. We believe that this technology will be able to bring about revolutionary changes to the nuclear industry by significantly lowering the operational costs as well as improving the reactor system safety to a large extent. Graphical abstract
	"Transmission electron microscopy study of a high burnup U-10Zr metallic fuel" Xiang Liu, Daniel J. Murray, Kyle M. Paaren, Fei Xu, Tsvetoslav Pavlov, Michael T. Benson, Luca Capriotti, Tiankai Yao, Daniele Salvato, [2022] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2022.153963 · ISSN: 0022-3115
	"Transmission Electron Microscopy based Characterization of a U-20Pu-10Zr Fuel Irradiated in Experimental Breeder Reactor-II" Xiang Liu, Yachun Wang, Fei Teng, Daniel J. Murray, Mitchell Meyer, Michael T. Benson, Luca Capriotti, Tiankai Yao, [2022] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2022.153846 · ISSN: 0022-3115
	"Dislocation loop evolution in Kr‐irradiated ThO ₂" Tiankai Yao, Kaustubh Bawane, Miaomiao Jin, Chao Jiang, Xiang Liu, Wei‐Ying Chen, J. Matthew Mann, David H. Hurley, Jian Gan, Marat Khafizov, Lingfeng He, [2022] Journal of the American Ceramic Society · DOI: 10.1111/jace.18478 · ISSN: 0002-7820 Abstract The early stage of microstructural evolution of ThO₂, under krypton irradiation at 600, 800, and 1000°C, was investigated using in situ transmission electron microscopy (TEM). Dislocation loops grew faster, whereas their number density decreased with increasing irradiation temperature. Loop density was found to decrease with ion dose. Interstitial dislocation loops, including Frank loops with Burgers vector of a/3〈111〉 and perfect loops with Burgers vector of a/2〈110〉, were determined by traditional TEM and atomic resolution–scanning TEM techniques. Atomistic and mesoscale level modeling are performed to interpret experimental observations. The migration energy barriers of defects in ThO₂ were calculated using density‐functional theory. The energetics of different dislocation loop types were studied using molecular dynamics simulations. Loop density and diameter were analyzed using a kinetic rate theory model that considers stoichiometric loop evolution. This analysis reveals that loop growth is governed by the mobility of cation interstitials, whereas loop nucleation is determined by the mobility of anion defects. Lastly, a rate theory model was used to extract the diffusion coefficients of thorium interstitials, oxygen interstitials, and vacancies.
	"Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system" Xiang Liu, Daniel J. Murray, Fei Teng, Wen Jiang, Mukesh Bachhav, Laura Hawkins, Emmanuel Perez, Cheng Sun, Xianming Bai, Jie Lian, Colin D. Judge, John H. Jackson, Robert G. Carter, Lingfeng He, Yachun Wang, [2022] Materials Science and Engineering: A · DOI: 10.1016/j.msea.2022.143475 · ISSN: 0921-5093
	"A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300 °C to 400 °C" Chih-Pin Chuang, Hsiao-Ming Tung, Kun Mo, Yinbin Miao, Xiang Liu, Hoon Lee, Jun-Sang Park, Jonathan Almer, James F. Stubbins, Kuan-Che Lan, [2022] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2022.153688
	"Sensitization, desensitization, and carbide evolution of Alloy 800H made by laser powder bed fusion" Xiang Liu, Miao Song, Lingfeng He, Stephen Bankson, Michael Hamilton, Bart Prorok, Xiaoyuan Lou, Jingfan Yang, [2022] Additive Manufacturing · DOI: 10.1016/j.addma.2021.102547 · ISSN: 2214-8604
	"Comparison of ion irradiation effects in PM-HIP and forged alloy 625" Yangyang Zhao, Patrick Warren, Xiang Liu, Sichuang Xue, David W. Gandy, Janelle P. Wharry, Caleb Clement, [2022] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2021.153390 · ISSN: 0022-3115
	"Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91" Xiang Liu, Lingfeng He, James Stubbins, Huan Yan, [2021] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2021.153207 · ISSN: 0022-3115
	"Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9" Xiang Liu, Lingfeng He, James Stubbins, Huan Yan, [2021] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2021.153252 · ISSN: 0022-3115
	"Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding" Mahmut Nedim Cinbiz, Boopathy Kombaiah, Lingfeng He, Fei Teng, Evrard Lacroix, Xiang Liu, [2021] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2021.153196 · ISSN: 0022-3115
	"Effects of heat treatment on corrosion fatigue and stress corrosion crack growth of additive-manufactured Alloy 800H in high-temperature water" Miao Song, Laura R. Hawkins, Xiang Liu, Lingfeng He, Xiaoyuan Lou, Jingfan Yang, [2021] Corrosion Science · DOI: 10.1016/j.corsci.2021.109739 · ISSN: 0010-938X
	"Microstructure and microchemistry of laser welds of irradiated austenitic steels" Aaron J. French, Xiang Liu, Yaqiao Wu, Lucille A. Giannuzzi, Cheng Sun, Megha Dubey, Paula D. Freyer, Jonathan K. Tatman, Frank A. Garner, Lin Shao, Janelle P. Wharry, Keyou S. Mao, [2021] Materials & Design · DOI: 10.1016/j.matdes.2021.109764 · ISSN: 0264-1275
	"TEM characterization of dislocation loops in proton irradiated single crystal ThO2" Xiang Liu, Tiankai Yao, Marat Khafizov, Aaron French, J. Matthew Mann, Lin Shao, Jian Gan, David H. Hurley, Lingfeng He, Kaustubh Bawane, [2021] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2021.152998
	"Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding" Luca Capriotti, Tiankai Yao, Jason M. Harp, Michael T. Benson, Yachun Wang, Fei Teng, Lingfeng He, Xiang Liu, [2021] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2020.152588
	"Investigation of the microstructure evolution of alpha uranium after in pile transient" Xiang Liu, Tommy V. Holschuh, Charles P. Folsom, Daniel J. Murray, Fei Teng, Colby B. Jensen, Fidelma G. Di Lemma, [2020] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2020.152467
	"α-U and ω-UZr2 in neutron irradiated U-10Zr annular metallic fuel" Luca Capriotti, Jason M. Harp, Xiang Liu, Yachun Wang, Fei Teng, Daniel J. Murray, Alex J. Winston, Jian Gan, Michael T. Benson, Lingfeng He, Tiankai Yao, [2020] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2020.152536
	"On the biaxial thermal creep-fatigue behavior of Ni-base Alloy 617 at 950 °C" Xiang Liu, Kuan-Che Lan, Hoon Lee, D.K.L. Tsang, James F. Stubbins, Yang Zhong, [2020] International Journal of Fatigue · DOI: 10.1016/j.ijfatigue.2020.105787
	"Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels" Maalavan Arivu, Haiming Wen, Li He, Kumar Sridharan, Xin Wang, Wei Xiong, Xiang Liu, Lingfeng He, Yaqiao Wu, Andrew Hoffman, [2020] Materialia · DOI: 10.1016/j.mtla.2020.100806
	"On spinodal-like phase decomposition in U–50Zr alloy" Adrian R Wagner, Xiang Liu, Anter EI-Azab, Jason M Harp, Jian Gan, David H Hurley, Michael T Benson, Lingfeng He, Tiankai Yao, [2020] Materialia · DOI: 10.1016/j.mtla.2020.100592
	"Radiation response of a Fe–20Cr–25Ni austenitic stainless steel under Fe2+ irradiation at 500 °C" Jonathan G. Gigax, Jonathan D. Poplawsky, Wei Guo, Hyosim Kim, Lin Shao, Frank A. Garner, James F. Stubbins, Xiang Liu, [2020] Materialia · DOI: 10.1016/j.mtla.2019.100542
	"A transmission electron microscopy study of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750" Lingfeng He, Huan Yan, Mukesh Bachhav, James F. Stubbins, Xiang Liu, [2020] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2019.151851 · ISSN: 0022-3115
	"Atom probe tomography for burnup and fission product analysis for nuclear fuels" Lingfeng He, Joshua Kane, Xiang Liu, Jian Gan, Francois Vurpiliot, Mukesh Bachhav, [2020] Microscopy and Microanalysis · DOI: 10.1017/s1431927620023776 · EID: 2-s2.0-85094837752 · ISSN: 1435-8115
	"Effect of proton pre-irradiation on corrosion of Zr-0.5Nb model alloys with different Nb distributions" Taeho Kim, Mukesh Bachhav, Xiang Liu, Lingfeng He, Adrien Couet, Zefeng Yu, [2020] Corrosion Science · DOI: 10.1016/j.corsci.2020.108790 · ISSN: 0010-938X
	"Irradiation effects on Al0.3CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500 °C" Marquis A. Kirk, Naoyuki Hashimoto, Jien-Wei Yeh, Xiang Liu, Yiren Chen, Wei-Ying Chen, [2020] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2020.152324 · ISSN: 0022-3115
	"Effect of laser welding on deformation mechanisms in irradiated austenitic stainless steel" Cheng Sun, Xiang Liu, Haozheng J. Qu, Aaron J. French, Paula D. Freyer, Frank A. Garner, Lin Shao, Janelle P. Wharry, Keyou S. Mao, [2019] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2019.151878 · ISSN: 0022-3115
	"Microstructure and microchemistry study of irradiation-induced precipitates in proton irradiated ZrNb alloys" Chenyu Zhang, Paul M. Voyles, Lingfeng He, Xiang Liu, Kelly Nygren, Adrien Couet, Zefeng Yu, [2019] Acta Materialia · DOI: 10.1016/j.actamat.2019.08.012
	"Radial microstructural evolution in low burnup fast reactor MOX fuel" Xiang Liu, Alexander Winston, Jason M. Harp, Assel Aitkaliyeva, Riley J. Parrish, [2019] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2019.06.009
	"The effect of thermal-aging on the microstructure and mechanical properties of 9Cr ferritic/martensitic ODS alloy" Zhangjian Zhou, Kun Mo, Yinbin Miao, Shuai Xu, Haodong Jia, Xiang Liu, James F. Stubbins, Guangming Zhang, [2019] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2019.05.023
	"Dissolution of Intermetallic Second-Phase Particles in Zircaloy-2 in High-Temperature Steam" Xiang Liu, Peter A. Mouche, Jun-Li Lin, Donghee Park, Mohamed S. Elbakhshwan, Simerjeet K. Gill, Yang Ren, James F. Stubbins, Brent J. Heuser, Weicheng Zhong, [2019] Metallurgical and Materials Transactions A · DOI: 10.1007/s11661-018-5090-5 · ISSN: 1073-5623
	"Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation" Xiang Liu, Yinbin Miao, Kun Mo, Zhi-Gang Mei, Laura Jamison, Walid Mohamed, Aaron Oaks, Ruqing Xu, Shaofei Zhu, James F. Stubbins, Abdellatif M. Yacout, Sumit Bhattacharya, [2019] Acta Materialia · DOI: 10.1016/j.actamat.2018.10.031
	"Small scale tensile testing of grain boundary strength of X-750 alloy" [2019] 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, EnvDeg 2019 · EID: 2-s2.0-85080084898
	"Irradiation effects in high entropy alloys and 316H stainless steel at 300 °C" Xiang Liu, Yiren Chen, Jien-Wei Yeh, Ko-Kai Tseng, Krishnamurti Natesan, Wei-Ying Chen, [2018] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2018.08.031
	"A special coarsening mechanism for intergranular helium bubbles upon heating: A combined experimental and numerical study" Hefei Huang, Xiang Liu, Chengbin Wang, James F. Stubbins, Yan Li, Jie Gao, [2018] Scripta Materialia · DOI: 10.1016/j.scriptamat.2018.01.006 · EID: 2-s2.0-85040358350
	"Helium ion irradiation-induced microstructure evolution on the surfaces of thin nickel foils" Hefei Huang, Xiang Liu, Xin Ou, Wanxia Wang, Guo Yang, Yan Li, Jie Gao, [2018] Nuclear Instruments and Methods in Physics Research, Section B: Beam Interactions with Materials and Atoms · DOI: 10.1016/j.nimb.2018.05.007 · EID: 2-s2.0-85047094054
	"Radiation resistance of oxide dispersion strengthened alloys: Perspectives from in situ observations and rate theory calculations" Yinbin Miao, Meimei Li, Marquis A. Kirk, Guangming Zhang, Shigeharu Ukai, Stuart A. Maloy, James F. Stubbins, Xiang Liu, [2018] Scripta Materialia · DOI: 10.1016/j.scriptamat.2018.01.018 · EID: 2-s2.0-85041468817
	"Stability of nanoclusters in an oxide dispersion strengthened alloy under neutron irradiation" Yinbin Miao, Yaqiao Wu, Stuart A. Maloy, James F. Stubbins, Xiang Liu, [2017] Scripta Materialia · DOI: 10.1016/j.scriptamat.2017.05.023
	"Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91" Yinbin Miao, Meimei Li, Marquis A. Kirk, Stuart A. Maloy, James F. Stubbins, Xiang Liu, [2017] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2017.04.047
	"Neutron and ion irradiation studies on advanced steels using the nuclear science user facilities" [2017] Transactions of the American Nuclear Society · EID: 2-s2.0-85033450329
	"Temperature effect of elastic anisotropy and internal strain development in advanced nanostructured alloys: An in-situ synchrotron X-ray investigation" Kun Mo, Di Yun, David T. Hoelzer, Yinbin Miao, Xiang Liu, Kuan-Che Lan, Jun-Sang Park, Jonathan Almer, Tianyi Chen, Huijuan Zhao, Yingye Gan, [2017] Materials Science and Engineering A · DOI: 10.1016/j.msea.2017.03.068 · EID: 2-s2.0-85015619155
	"In situ synchrotron tensile investigations on 14YWT, MA957, and 9-Cr ODS alloys" Kun Mo, Di Yun, Yinbin Miao, Xiang Liu, Huijuan Zhao, David T. Hoelzer, Jun-Sang Park, Jonathan Almer, Guangming Zhang, Zhangjian Zhou, James F. Stubbins, Abdellatif M. Yacout, Jun-Li Lin, [2016] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2015.10.049
	"Investigation of High-Energy Ion-Irradiated MA957 Using Synchrotron Radiation under In-Situ Tension" Di Yun, Yinbin Miao, Xiang Liu, Michael Pellin, Jonathan Almer, Jun-Sang Park, James Stubbins, Shaofei Zhu, Abdellatif Yacout, Kun Mo, [2016] Materials · DOI: 10.3390/ma9010015 In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner.
	"Investigation of thermal aging effects on the tensile properties of Alloy 617 by in-situ synchrotron wide-angle X-ray scattering" Kun Mo, Yinbin Miao, Kuan-Che Lan, Guangming Zhang, Wei-Ying Chen, Carolyn Tomchik, Rachel Seibert, Jeff Terry, James F. Stubbins, Xiang Liu, [2016] Materials Science and Engineering a-Structural Materials Properties Microstructure and Processing · DOI: 10.1016/j.msea.2015.10.098
	"Load-partitioning in an oxide dispersion-strengthened 310 steel at elevated temperatures" Kun Mo, Zhangjian Zhou, Xiang Liu, Kuan-Che Lan, Guangming Zhang, Jun-Sang Park, Jonathan Almer, James F. Stubbins, Yinbin Miao, [2016] Materials & Design · DOI: 10.1016/j.matdes.2016.09.015
	"Size-dependent characteristics of ultra-fine oxygen-enriched nanoparticles in austenitic steels" Kun Mo, Zhangjian Zhou, Xiang Liu, Kuan-Che Lan, Guangming Zhang, Michael K. Miller, Kathy A. Powers, James F. Stubbins, Yinbin Miao, [2016] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2016.08.014
	"The comparison of microstructures and mechanical properties between 14Cr-Al and 14Cr-Ti ferritic ODS alloys" Zhangjian Zhou, Kun Mo, Yinbin Miao, Shaofu Li, Xiang Liu, Man Wang, Jun-Sang Park, Jonathan Almer, James F. Stubbins, Guangming Zhang, [2016] Materials & Design · DOI: 10.1016/j.matdes.2016.02.117
	"The evolution of internal stress and dislocation during tensile deformation in a 9Cr ferritic/martensitic (F/M) ODS steel investigated by high-energy X-rays" Zhangjian Zhou, Kun Mo, Yinbin Miao, Xiang Liu, Jonathan Almer, James F. Stubbins, Guangming Zhang, [2015] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2015.09.014
	"The evolution of internal stress and dislocation during tensile deformation in a 9Cr ferritic/martensitic (F/M) {ODS} steel investigated by high-energy X-rays" Zhangjian Zhou, Kun Mo, Yinbin Miao, Xiang Liu, Jonathan Almer, James F. Stubbins, Guangming Zhang, [2015] Journal of Nuclear Materials · DOI: 10.1016/j.jnucmat.2015.09.014
	"In situ synchrotron tensile investigations on the phase responses within an oxide dispersion-strengthened ({ODS}) 304 steel" Kun Mo, Zhangjian Zhou, Xiang Liu, Kuan-Che Lan, Guangming Zhang, Michael K. Miller, Kathy A. Powers, Jonathan Almer, James F. Stubbins, Yinbin Miao, [2015] Materials Science and Engineering: A · DOI: 10.1016/j.msea.2014.12.017
	"Differential and angle-integrated cross sections for the 40Ca(n, $\upalpha$)37Ar reaction from 4.0 to 6.5 {MeV}" Jiaming Liu, Xiang Liu, Xiao Fan, Zhimin Wang, Jinxiang Chen, Guohui Zhang, Yu. M. Gledenov, M. V. Sedysheva, L. Krupa, G. Khuukhenkhuu, P. J. Szalanski, Jinhua Han, [2015] The European Physical Journal A · DOI: 10.1140/epja/i2015-15012-5
	"Load partitioning between ferrite/martensite and dispersed nanoparticles of a 9Cr ferritic/martensitic (F/M) ODS steel at high temperatures" Kun Mo, Yinbin Miao, Xiang Liu, Jonathan Almer, Zhangjian Zhou, James F. Stubbins, Guangming Zhang, [2015] Materials Science and Engineering A · DOI: 10.1016/j.msea.2015.04.037 · EID: 2-s2.0-84928626214
	"On the microstructure and strengthening mechanism in oxide dispersion-strengthened 316 steel: A coordinated electron microscopy, atom probe tomography and in situ synchrotron tensile investigation" Kun Mo, Zhangjian Zhou, Xiang Liu, Kuan-Che Lan, Guangming Zhang, Michael K. Miller, Kathy A. Powers, Zhi-Gang Mei, Jun-Sang Park, Jonathan Almer, James F. Stubbins, Yinbin Miao, [2015] Materials Science and Engineering A · DOI: 10.1016/j.msea.2015.05.064 · EID: 2-s2.0-84930632580
	"The microstructure and mechanical properties of Al-containing 9Cr ODS ferritic alloy" Zhangjian Zhou, Kun Mo, Pinghuai Wang, Yinbin Miao, Shaofu Li, Man Wang, Xiang Liu, Mengqiang Gong, Jonathan Almer, James F. Stubbins, Guangming Zhang, [2015] Journal of Alloys and Compounds · DOI: 10.1016/j.jallcom.2015.06.214 · EID: 2-s2.0-84936141534
	"Cross sections of the 57 Fe ( n , $\upalpha$ ) 54 Cr and 63 Cu ( n , $\upalpha$ ) 60 Co reactions in the {MeV} region" M. V. Sedysheva, V. A. Stolupin, Guohui Zhang, Jinhua Han, Zhimin Wang, Xiao Fan, Xiang Liu, Jinxiang Chen, G. Khuukhenkhuu, P. J. Szalanski, Yu. M. Gledenov, [2014] Physical Review C · DOI: 10.1103/physrevc.89.064607
	"Nanoparticles loading behavior before and after matrix necking: An in-situ synchrotron radiation study in a 9Cr ODS alloy" [2014] Transactions of the American Nuclear Society · EID: 2-s2.0-84904622343
	"Response of oxide particles to externally applied stress in austenitic ods alloys" [2014] Transactions of the American Nuclear Society · EID: 2-s2.0-84904705863
	"Measurement of Cross Sections for the 10 B( n ,$\upalpha$) 7 Li Reaction at 4.0 and 5.0 {MeV} Using an Asymmetrical Twin Gridded Ionization Chamber" Xiang Liu, Jia-Ming Liu, Zhi-Hua Xue, Hao Wu, Jin-Xiang Chen, Guo-Hui Zhang, [2011] Chinese Physics Letters · DOI: 10.1088/0256-307x/28/8/082801
	"Sm 149 ( n , $\upalpha$ ) Nd 146 Cross Sections in the {MeV} Region" Yu. M. Gledenov, G. Khuukhenkhuu, M. V. Sedysheva, P. J. Szalanski, P. E. Koehler, Yu. N. Voronov, Jiaming Liu, Xiang Liu, Jinhua Han, Jinxiang Chen, Guohui Zhang, [2011] Phys. Rev. Lett. · DOI: 10.1103/physrevlett.107.252502
	"Cross sections of the Zn 67 ( n , $\upalpha$ ) 64 Ni reaction at 4.0, 5.0, and 6.0 {MeV}" Yu. M. Gledenov, G. Khuukhenkhuu, M. V. Sedysheva, P. J. Szalanski, Jiaming Liu, Hao Wu, Xiang Liu, Jinxiang Chen, V. A. Stolupin, Guohui Zhang, [2010] Physical Review C · DOI: 10.1103/physrevc.82.054619
	"Cross-section measurement and analysis for the Sm 149 ( n , $\upalpha$ ) Nd 146 reaction at 6.0~{MeV}" Guohui Zhang, G. Khuukhenkhuu, M. V. Sedysheva, P. J. Szalanski, P. E. Koehler, Jiaming Liu, Hao Wu, Xiang Liu, Jinxiang Chen, Yu. M. Gledenov, [2010] Physical Review C · DOI: 10.1103/physrevc.82.014601
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NSUF Supported Research

Investigation of fission gas bubble distribution, phase transformations, and bubble growth kinetics in a FFTF-irradiated U-10Zr fuel - FY 2019 RTE 3rd Call, #19-2899

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, #19-1635

Investigation of the synergistic effects of dual beam irradiation in an austenitic-ferritic duplex alloy - FY 2018 RTE 3rd Call, #18-1532

NSUF Research Collaborations

Towards Understanding Fuel Clad Chemical Interactions in FeCrAl Alloys - FY 2020 RTE 2nd Call, #20-4108

In-situ microstructural evolution and phase transition of irradiated and transient metallic fuel for M&S validation - FY 2020 RTE 2nd Call, #20-3088

Atom Probe Tomography Study of Elemental Segregation and Precipitation in Ion-Irradiated Advanced Austenitic Alloy A709 - FY 2020 RTE 1st Call, #20-2963

High resolution (S)TEM/EDS characterization of neutron irradiated commercial Zr-Nb alloys - FY 2019 RTE 3rd Call, #19-2860

Advanced microstructural characterization of irradiation-induced phase transformation in 304 steel - FY 2019 RTE 3rd Call, #19-2858

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2019 RTE 2nd Call, #19-1802

Local Deformation Mechanism of Neutron-Irradiated NF709 Austenitic Stainless Steel - FY 2019 RTE 1st Call, #19-1652

Nanoindentation of Phases in Irradiated and Control U-10Zr Fuels - FY 2019 RTE 1st Call, #19-1666

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, #19-1670

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, #18-1433

Microstructure characterization of neutron-irradiated Fe-Cr-C model alloys - FY 2018 RTE 1st Call, #18-1203

Advanced Investigations on the Low-dose Neutron-irradiated MA957 - FY 2015 RTE 1st Call, #15-538

Post-irradiation Examinations on a Ferritic ODS Steel Irradiated at 450C - FY 2015 RTE 1st Call, #15-539