@article{yang_hawkins_shang_tsai_he_lu_song_wang_lou_2024, title={Dislocation channel broadening-A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel}, volume={266}, ISSN={["1873-2453"]}, DOI={10.1016/j.actamat.2024.119650}, abstractNote={Additively manufactured (AM) 316 L stainless steel (SS) after hot isostatic pressing (HIP) was found to exhibit superior resistance to irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water, as compared to wrought 316 L SS. The well-accepted IASCC factors of radiation-induced segregation (RIS) and radiation hardening are not accurate descriptions of IASCC susceptibility in this case. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was confirmed to suppress crack initiation. A unique distribution of irradiation-induced defects in HIP AM 316 L SS eased dislocation cross-slip compared to those in the wrought counterpart, thus increasing the channel width near the grain boundaries. For the first time, this study highlights the importance of dislocation channel broadening as a potential mechanism to further improve the IASCC resistance of 316 L SS and provides direct experimental evidence based on commercial-grade materials.}, journal={ACTA MATERIALIA}, author={Yang, Jingfan and Hawkins, Laura and Shang, Zhongxia and Tsai, Benson Kunhung and He, Lingfeng and Lu, Yu and Song, Miao and Wang, Haiyan and Lou, Xiaoyuan}, year={2024}, month={Mar} }
 @article{mazumder_bawane_mann_french_shao_he_el-azab_2023, title={Evolution of dislocation loops and voids in post-irradiation annealed ThO<sub>2</sub>: A combined <i>in-situ</i> TEM and cluster dynamics investigation}, volume={586}, ISSN={["1873-4820"]}, DOI={10.1016/j.jnucmat.2023.154686}, abstractNote={The effect of isochronal annealing on the evolution of dislocation loop and void population in proton irradiated ThO2 has been investigated. Post-irradiation annealing in other actinide oxides like UO2 shows significant loop coarsening. ThO2 samples were irradiated with 2 MeV protons up to a dose of 0.1 dpa at 600 °C. Post-irradiation isochronal annealing was performed at 600, 800, 1000 and 1100 °C for 1 h at each temperature using in-situ TEM. Only faulted 1/3<111> type dislocation loops were observed, and their sizes and distribution were characterized. The population of self-interstitial atom (SIA) dislocation loops did not show any significant growth and coarsening. Additionally, nanometric voids were observed at annealing temperatures of 1000 and 1100 °C. Using cluster dynamics (CD), we have studied the nucleation and growth of point defects and defect clusters, i.e., SIA prismatic dislocation loops and nanometric and sub-nanometric voids in proton irradiated ThO2. The CD model was further utilized to predict the growth and coarsening of loops and voids during isochronal annealing at the experimental and higher temperatures. The model did not predict significant SIA loop growth which closely corresponds to the TEM observations. CD predicted SIA loop coarsening is insignificant even at high annealing temperature of 1500 °C because the model only considers the growth of defect clusters by absorption of like point defects, i.e., SIA loops absorb interstitials and voids absorb vacancies, and cannot account for their migration and coalescence due to elastic interaction. The CD model also predicts the evolution of nanometric voids having mean size within the error bounds of TEM observations.}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Mazumder, Sanjoy Kumar and Bawane, Kaustubh and Mann, J. Matthew and French, Aaron and Shao, Lin and He, Lingfeng and El-Azab, Anter}, year={2023}, month={Dec} }
 @article{yang_hawkins_he_mahmood_song_schulze_lou_2023, title={Intragranular irradiation-assisted stress corrosion cracking (IASCC) of 316L stainless steel made by laser direct energy deposition additive manufacturing: Delta ferrite-dislocation channel interaction}, volume={577}, ISSN={["1873-4820"]}, DOI={10.1016/j.jnucmat.2023.154305}, abstractNote={In this study, irradiation-assisted stress corrosion cracking (IASCC) resistance of as-built AM 316L stainless steel (SS) made by direct energy deposition (DED) was evaluated in a simulated boiling water reactor (BWR) environment. Different from the intergranular cracking typically seen from 316L SS, unique intragranular cracking away from grain boundaries (GB) was observed. The presence of high-density micron-sized delta ferrite near the melt pool boundaries promoted the in-grain strain localization and crack nucleation. Delta ferrite particles can efficiently retard dislocations during deformation, altering the strain localization from the GB to its surrounding area and thus reducing the chance of GB cracking. The phenomenon was further confirmed by the higher nanoindentation hardness measured from the corresponding area, and by characterizing the evolution of microstructure, cracking morphology, and dislocation channels.}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Yang, Jingfan and Hawkins, Laura and He, Lingfeng and Mahmood, Samsul and Song, Miao and Schulze, Kyle and Lou, Xiaoyuan}, year={2023}, month={Apr} }
 @article{yu_bachhav_teng_he_dubey_couet_2023, title={STEM/EDS and APT study on the microstructure and microchemistry of neutron irradiated ZIRLOTM}, volume={573}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2022.154139}, DOI={10.1016/j.jnucmat.2022.154139}, abstractNote={A nanoscale characterization study was carried out on SRA ZIRLO™ at various cycles, aiming to highlight the effect of irradiation-induced alloying element (Sn, Nb, Fe) redistribution on the in-reactor corrosion kinetics. Using a combination of scanning transmission electron microscopy (STEM), energy dispersive X-ray spectroscopy (EDS), and atom probe tomography (APT) in both oxide and metal, the results confirmed the existence of i) Nb-rich native precipitates and irradiation-induced platelets (IIPs)/nanoclusters in both metal and oxide, and ii) solute segregation to the metal and oxide grain boundaries. One of the most important findings is that the Nb/Zr ratio in the metal solid solution at high burnup is only about 0.2 at.%. All three alloying elements (Sn, Fe, Nb) have segregated to oxide grain boundaries at high burnup. The effects of these microstructural and microchemistry changes in SRA ZIRLO™ on its in-reactor corrosion mechanism are discussed.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yu, Zefeng and Bachhav, Mukesh and Teng, Fei and He, Lingfeng and Dubey, Megha and Couet, Adrien}, year={2023}, month={Jan}, pages={154139} }
 @article{liu_bawane_zheng_ge_halstenberg_maltsev_ivanov_dai_xiao_lee_et al._2023, title={Temperature-Dependent Morphological Evolution during Corrosion of the Ni-20Cr Alloy in Molten Salt Revealed by Multiscale Imaging}, volume={15}, ISSN={["1944-8252"]}, DOI={10.1021/acsami.2c23207}, abstractNote={Understanding the mechanisms leading to the degradation of alloys in molten salts at elevated temperatures is significant for developing several key energy generation and storage technologies, including concentrated solar and next-generation nuclear power plants. Specifically, the fundamental mechanisms of different types of corrosion leading to various morphological evolution characteristics for changing reaction conditions between the molten salt and alloy remain unclear. In this work, the three-dimensional (3D) morphological evolution of Ni-20Cr in KCl-MgCl2 is studied at 600 °C by combining in situ synchrotron X-ray and electron microscopy techniques. By further comparing different morphology evolution characteristics in the temperature range of 500-800 °C, the relative rates between diffusion and reaction at the salt-metal interface lead to different morphological evolution pathways, including intergranular corrosion and percolation dealloying. In this work, the temperature-dependent mechanisms of the interactions between metals and molten salts are discussed, providing insights for predicting molten salt corrosion in real-world applications.}, number={10}, journal={ACS APPLIED MATERIALS & INTERFACES}, author={Liu, Xiaoyang and Bawane, Kaustubh and Zheng, Xiaoyin and Ge, Mingyuan and Halstenberg, Phillip and Maltsev, Dmitry S. and Ivanov, Alexander S. and Dai, Sheng and Xiao, Xianghui and Lee, Wah-Keat and et al.}, year={2023}, month={Mar}, pages={13772–13782} }
 @article{hawkins_yang_song_schwen_zhang_shao_lou_he_2023, title={The effect of secondary phases on microstructure and irradiation damage in an as-built additively manufactured 316 L stainless steel with a hafnium compositional gradient}, volume={587}, ISSN={["1873-4820"]}, DOI={10.1016/j.jnucmat.2023.154708}, abstractNote={Additive manufacturing (AM) or rapid prototyping has become a crucial tool for reducing both cost and time while increasing efficiency in qualifying structural materials for reactor use. In this study, directed energy deposition (DED) was used to develop an as-built 316 L stainless steel sample with three regions of increasing Hf-dopant to study the effects of Hf on the irradiation response of the material. Morphological and microstructural changes were analyzed before and after 2 MeV proton irradiation at 360 °C to a damage of 2.5 dpa at ∼ 10 µm below the surface. The addition of Hf effectively suppressed radiation-induced damage (dislocation loops, radiation-induced segregation) due to enhanced point defect recombination. The radiation damage seen in the as-built sample was further compared to a thermo-mechanically treated counterpart of the same fabrication and irradiation parameters which was found to behave superiorly. The increased radiation resistance of this material may be attributed to the as-built microstructure, which includes undissolved Hf particles, delta ferrite grains and cellular sub-grain boundaries that can hinder defect motion.}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Hawkins, Laura and Yang, Jingfan and Song, Miao and Schwen, Daniel and Zhang, Yongfeng and Shao, Lin and Lou, Xiaoyuan and He, Lingfeng}, year={2023}, month={Dec} }
 @article{thomas_liu_he_murray_teng_kombaiah_winston_okuniewski_2023, title={Transmission electron microscopy investigation of phase transformation and fuel constituent redistribution in neutron irradiated U-10wt.%Zr fuel}, volume={581}, ISSN={["1873-4820"]}, url={https://doi.org/10.1016/j.jnucmat.2023.154443}, DOI={10.1016/j.jnucmat.2023.154443}, abstractNote={Uranium-10 wt.% zirconium (U-10 wt.%Zr) is a primary candidate for fast reactor nuclear fuels. However, there is a lack of data characterizing the crystallographic phases and chemistry of the neutron irradiated fuel. In the current study, the microstructural evolution of a U-10 wt.%Zr fuel neutron irradiated to a burnup of 5.7 at.% was investigated using scanning transmission electron microscopy, energy dispersive X-ray spectroscopy, and selected area electron diffraction to determine the major phases, alterations in microstructure, and variations in local chemical composition at different localities of a fuel cross-section. The current study revealed that the irradiated U-10 wt.%Zr fuel was comprised of various major phases, including α-U, β-U, and δ-UZr2, as well as amorphous and crystalline solid fission product (FP) precipitates within different regions of the fuel cross-section. Regions A and A/B in the center of the fuel were comprised of U-rich, U-intermediate, and U-lean localities with α-U and δ-UZr2 composing the major phases. Regions B and C in the intermediate and peripheral fuel localities, respectively, were comprised of α-U grains, U-rich (β-U) grains with Zr-rich precipitates, U-intermediate grains (δ-UZr2), and solid FP precipitates. The major phases identified were associated with the nanoscopic chemical concentrations, the phase diagram, and the as-characterized specimen temperature, with the exception of the non-equilibrium β-U phase identified in Region B. The U-lean localities in Regions A, A/B, B, and C were Zr-enriched pathways along subgrain/grain boundaries, suggesting that Zr is susceptible to radiation-induced segregation in U-Zr fuels and indicating a new mechanism for constituent redistribution.}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Thomas, Jonova and Liu, Xiang and He, Lingfeng and Murray, Daniel and Teng, Fei and Kombaiah, Boopathy and Winston, Alex and Okuniewski, Maria A.}, year={2023}, month={Aug} }
 @article{song_yang_liu_hawkins_jiao_he_zhang_schwen_lou_2023, title={Void swelling in additively manufactured 316L stainless steel with hafnium composition gradient under self-ion irradiation}, volume={578}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2023.154351}, DOI={10.1016/j.jnucmat.2023.154351}, abstractNote={Compositionally-graded austenitic 316L stainless steel (SS) samples with five different Hafnium (Hf) concentrations (up to 1 wt.%) were additively manufactured by directed energy deposition and then were irradiated using 5 MeV Fe2+ ions to 50 displacements per atom (dpa) at 500, 550, and 600 °C, respectively. A rastering beam was used to ensure homogenous irradiations for the large areas of ∼1.40 cm2. Composition-dependent void evolution was evaluated. Void size, void number density, and void swelling all tend to decrease with increasing Hf concentration at all three temperatures. When the nominal Hf concentration increases to 1 wt.% in the doped additively manufactured (AM) 316L SS, the void swelling is over an order of magnitude lower than that in pure AM 316L. Atom probe tomography results showed that about 0.14 wt.% Hf dissolved in the matrix at the 1 wt.%. Hf nominal concentration. The suppression of void swelling by Hf addition shows qualitative agreement with the numerical calculation based on the vacancy trapping mechanism, indicating that the oversized Hf reduces the steady state vacancy concentration, and hence the incubation period of swelling is extended, resulting in a dramatic difference in void swelling. Other factors including grain boundaries, dislocation and dislocation cells, Hf-rich particles, and delta ferrite are secondary to this mechanism. This work shows that additive manufacturing enabled microalloying is promising for developing void swelling resistant materials.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Song, Miao and Yang, Jingfan and Liu, Xiang and Hawkins, Laura R. and Jiao, Zhijie and He, Lingfeng and Zhang, Yongfeng and Schwen, Daniel and Lou, Xiaoyuan}, year={2023}, month={May}, pages={154351} }
 @article{deskins_khanolkar_mazumder_dennett_bawane_hua_ferrigno_he_mann_khafizov_et al._2022, title={A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide}, volume={241}, ISSN={1359-6454}, url={http://dx.doi.org/10.1016/j.actamat.2022.118379}, DOI={10.1016/j.actamat.2022.118379}, abstractNote={During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.}, journal={Acta Materialia}, publisher={Elsevier BV}, author={Deskins, W. Ryan and Khanolkar, Amey and Mazumder, Sanjoy and Dennett, Cody A. and Bawane, Kaustubh and Hua, Zilong and Ferrigno, Joshua and He, Lingfeng and Mann, J. Matthew and Khafizov, Marat and et al.}, year={2022}, month={Dec}, pages={118379} }
 @article{copeland-johnson_murray_cao_he_2022, title={Assessing the interfacial corrosion mechanism of Inconel 617 in chloride molten salt corrosion using multi-modal advanced characterization techniques}, volume={1}, ISSN={2813-3412}, url={http://dx.doi.org/10.3389/fnuen.2022.1049693}, DOI={10.3389/fnuen.2022.1049693}, abstractNote={The United States Department of Energy (DOE) has committed to expanding the domestic clean energy portfolio in response to the rising challenges of energy security in the wake of climate change. Accordingly, the construction of a series of Generation IV reactor technologies are being demonstrated, including sodium-cooled, small modular, and molten chloride fast reactors (MCFRs). To date, there are no fully qualified structural materials for constructing MCFRs. A number of commercial structural alloys have been considered for the construction of MCFRs, including alloys from the Inconel and Hastelloy series. Informed qualification of structural materials for the construction of MCFRs in the future can only be ensured by expanding the current fundamental knowledgebase of information pertaining to material performance under environmental stressors relevant to operation of the reactor, including corrosion susceptibility. The purpose of this investigation is to illustrate how a correlative multi-modal electron microscopy characterization approach, including the novel application of focused-ion beam 3D reconstruction capabilities, can elucidate the corrosion mechanism of a candidate structural material Inconel 617 for MCFR in NaCl-MgCl2 eutectic salt at 700°C for 1,000 h. Evidence of intergranular corrosion, Ni and Fe dealloying, and Cr-O enrichment along the grain boundary, which most likely corresponds to Cr2O3, is a phenomenon that has been documented in other Ni-based superalloys exposed to chloride molten salt systems. Additional corrosion products, including the formation of insoluble MgAl2O4, within the porous network produced by the salt attack is a novel observation. In addition, Mo3Si5 and τ2 precipitates are detected in the alloy bulk and are dissolved by the salt. Furthermore, the lack of detection of design γ′ precipitates in Inconel 617 after 1,000 h could indicate that the molten salt corrosion mechanism has indirectly induced a phase transformation of Al2TiNi (τ2) and Ni3(Al,Ti) (γ’) phase. This investigation provides a comprehensive understanding of molten salt corrosion mechanisms in a complex material system such as a commercial structural alloy for applications in MCFRs.}, journal={Frontiers in Nuclear Engineering}, publisher={Frontiers Media SA}, author={Copeland-Johnson, Trishelle M. and Murray, Daniel J. and Cao, Guoping and He, Lingfeng}, year={2022}, month={Dec} }
 @article{ditter_smiles_lussier_altman_bachhav_he_mara_degueldre_minasian_shuh_2022, title={Chemical and elemental mapping of spent nuclear fuel sections by soft X-ray spectromicroscopy}, volume={29}, url={https://doi.org/10.1107/S1600577521012315}, DOI={10.1107/S1600577521012315}, abstractNote={Soft X-ray spectromicroscopy at the O K-edge, U N
                  4,5-edges and Ce M
                  4,5-edges has been performed on focused ion beam sections of spent nuclear fuel for the first time, yielding chemical information on the sub-micrometer scale. To analyze these data, a modification to non-negative matrix factorization (NMF) was developed, in which the data are no longer required to be non-negative, but the non-negativity of the spectral components and fit coefficients is largely preserved. The modified NMF method was utilized at the O K-edge to distinguish between two components, one present in the bulk of the sample similar to UO2 and one present at the interface of the sample which is a hyperstoichiometric UO2+x
                   species. The species maps are consistent with a model of a thin layer of UO2+x
                   over the entire sample, which is likely explained by oxidation after focused ion beam (FIB) sectioning. In addition to the uranium oxide bulk of the sample, Ce measurements were also performed to investigate the oxidation state of that fission product, which is the subject of considerable interest. Analysis of the Ce spectra shows that Ce is in a predominantly trivalent state, with a possible contribution from tetravalent Ce. Atom probe analysis was performed to provide confirmation of the presence and localization of Ce in the spent fuel.}, number={1}, journal={Journal of Synchrotron Radiation}, publisher={International Union of Crystallography (IUCr)}, author={Ditter, Alexander Scott and Smiles, Danil E. and Lussier, Daniel and Altman, Alison B. and Bachhav, Mukesh and He, Lingfeng and Mara, Michael W. and Degueldre, Claude and Minasian, Stefan G. and Shuh, David K.}, year={2022}, month={Jan}, pages={67–79} }
 @article{yang_hawkins_song_he_bachhav_pan_shao_schwen_lou_2022, title={Compositionally graded specimen made by laser additive manufacturing as a high-throughput method to study radiation damages and irradiation-assisted stress corrosion cracking}, volume={560}, url={http://dx.doi.org/10.1016/j.jnucmat.2021.153493}, DOI={10.1016/j.jnucmat.2021.153493}, abstractNote={This study demonstrates the feasibility of using compositionally gradient specimens, fabricated by laser additive manufacturing (AM) and post-AM thermo-mechanical treatment, to accelerate alloy synthesis, radiation experiment, and the assessment of irradiation properties in light water reactor environments. The effects of minor Hafnium (Hf) doping in austenitic 316L stainless steel (SS) was selected as the topic of interest. By comparing to the data in literature, we confirmed that the compositionally graded specimen produces the same trend of void swelling, dislocation loops, radiation-induced segregation (RIS), radiation hardening as the wrought specimen produced by cast/forging process. Hf suppressed most radiation damages through strong interaction with point defects. The work also demonstrates the use of compositionally gradient specimens to study the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of Hf-modified SS. While the suppression of radiation hardening and RIS are consistent with the IASCC mitigation by Hf, we emphasize Hf can alter the intrinsic deformation behavior of 316L SS, which reduces grain-boundary strain localization. The advantages and challenges of using compositionally gradient design for high-throughput nuclear alloy development and qualification are also discussed.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yang, Jingfan and Hawkins, Laura and Song, Miao and He, Lingfeng and Bachhav, Mukesh and Pan, Qingyu and Shao, Lin and Schwen, Daniel and Lou, Xiaoyuan}, year={2022}, month={Mar}, pages={153493} }
 @article{chauhan_ferrigno_adnan_pakarinen_he_hurley_khafizov_2022, title={Comprehensive characterization of irradiation induced defects in ceria: Impact of point defects on vibrational and optical properties}, url={https://doi.org/10.1063/5.0099189}, DOI={10.1063/5.0099189}, abstractNote={Validation of multiscale microstructure evolution models can be improved when standard microstructure characterization tools are coupled with methods sensitive to individual point defects. We demonstrate how electronic and vibrational properties of defects revealed by optical absorption and Raman spectroscopies can be used to compliment transmission electron microscopy (TEM) and x-ray diffraction (XRD) in the characterization of microstructure evolution in ceria under non-equilibrium conditions. Experimental manifestation of non-equilibrium conditions was realized by exposing cerium dioxide (CeO2) to energetic protons at elevated temperature. Two sintered polycrystalline CeO2 samples were bombarded with protons accelerated to a few MeVs. These irradiation conditions produced a microstructure with resolvable extended defects and a significant concentration of point defects. A rate theory (RT) model was parametrized using the results of TEM, XRD, and thermal conductivity measurements to infer point defect concentrations. An abundance of cerium sublattice defects suggested by the RT model is supported by Raman spectroscopy measurements, which show peak shift and broadening of the intrinsic T2g peak and emergence of new defect peaks. Additionally, spectroscopic ellipsometry measurements performed in lieu of optical absorption reveals the presence of Ce3+ ions associated with oxygen vacancies. This work lays the foundation for a coupled approach that considers a multimodal characterization of microstructures to guide and validate complex defect evolution models.}, journal={Journal of Applied Physics}, author={Chauhan, Vinay S. and Ferrigno, Joshua and Adnan, Saqeeb and Pakarinen, Janne and He, Lingfeng and Hurley, David H. and Khafizov, Marat}, year={2022}, month={Aug} }
 @article{he_yao_bawane_jin_jiang_liu_chen_mann_hurley_gan_et al._2022, title={Dislocation loop evolution in Kr‐irradiated ThO2}, url={https://doi.org/10.1111/jace.18478}, DOI={10.1111/jace.18478}, abstractNote={Abstract}, journal={Journal of the American Ceramic Society}, author={He, Lingfeng and Yao, Tiankai and Bawane, Kaustubh and Jin, Miaomiao and Jiang, Chao and Liu, Xiang and Chen, Wei‐Ying and Mann, J. Matthew and Hurley, David H. and Gan, Jian and et al.}, year={2022}, month={Aug} }
 @article{gabriel_hawkins_french_li_hu_he_xiu_nastasi_garner_shao_2022, title={Effect of dpa rate on the temperature regime of void swelling in ion-irradiated pure chromium}, volume={561}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2022.153519}, DOI={10.1016/j.jnucmat.2022.153519}, abstractNote={Pure chromium is a promising candidate for coating of Zircalloy fuel rods in light water-cooled power reactors to avoid or delay hydrogen generation in accident scenarios. Void swelling of chromium is one possible contributor to coating-interface failure and needs to be studied. The effect of dpa rate on void swelling of Cr was studied using 5 MeV Fe ion irradiation to 15 peak dpa at peak dpa rates of 3.5 × 10−5, 3.5 × 10−4, and 3.5 × 10−3 dpa/s, at six temperatures between 350 °C and 650 °C. The post-transient (steady-state) swelling rate of pure Cr is ∼0.05%/dpa. Swelling in Cr also appeared to start at a much higher swelling rate at very low doses. The observed dependence of peak swelling temperature on dpa rate agrees well with earlier theoretical models and clearly demonstrated the well-known “temperature shift” phenomenon. After determination of the peak swelling temperatures under ion irradiation at three dpa rates we extrapolated downwards to dpa rates characteristic of both sodium-cooled fast reactors and light water-cooled reactors.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Gabriel, Adam and Hawkins, Laura and French, Aaron and Li, Yongchang and Hu, Zhihan and He, Lingfeng and Xiu, Pengyuan and Nastasi, Michael and Garner, Frank.A. and Shao, Lin}, year={2022}, month={Apr}, pages={153519} }
 @article{sure_gill_wang_bawane_he_halstenberg_dai_mahurin_wishart_sasaki_2022, title={Electrochemical noise studies on localized corrosion of Ni and Ni-20Cr in molten ZnCl2}, volume={431}, ISSN={0013-4686}, url={http://dx.doi.org/10.1016/j.electacta.2022.141126}, DOI={10.1016/j.electacta.2022.141126}, abstractNote={The electrochemical noise (ECN) technique was employed to study corrosion of two model systems, i.e. pure Ni and a binary Ni-20 wt%Cr (Ni20Cr) alloy in molten ZnCl2 at 623 K. We measured ECN transients in current and open-circuit potential from two nominally identical Ni-Ni and Ni20Cr-Ni20Cr electrodes and one galvanic Ni-Ni20Cr electrode pair. The behavior of ECN is quite distinct among the three-electrode systems, and it is correlated with the various microscopic observations of micro- and nano-scale morphological features. Based on the ECN study coupled with the microstructural analysis, the origins of ECN in the molten salt environment as well as the mechanisms of localized corrosion in the three systems are discussed.}, journal={Electrochimica Acta}, publisher={Elsevier BV}, author={Sure, Jagadeesh and Gill, Simerjeet k and Wang, Yachun and Bawane, Kaustubh.K. and He, Lingfeng and Halstenberg, Phillip and Dai, Sheng and Mahurin, Shannon M. and Wishart, James F. and Sasaki, Kotaro}, year={2022}, month={Nov}, pages={141126} }
 @article{vijayan_bawane_giulia dilemma_he_fink_jinschek_2022, title={In situ</i> TEM Observations of Thermally Activated Phenomena in Materials Under Far-From-Equilibrium Conditions}, volume={28}, ISSN={1435-8115 1431-9276}, url={http://dx.doi.org/10.1017/S1431927622007255}, DOI={10.1017/S1431927622007255}, abstractNote={Structural materials, that have been fabricated using additive manufacturing (AM) or have been welded using metal joining processes, experience varying spatial and temporal thermal transients due to a very local but high energy heat source. Similarly, components experience varying thermal transients when ‘in service’, e.g. in next generation nuclear reactor cores, gas turbine engines, re-entry space vehicles and solder joints in micro-electronic packages. These varying thermal transients (extreme thermal gradients (10 4 - 10 6 K/m) and/or rapid thermal cycling (10 2 - 10 3 K/s)) cause microstructural changes due to solid-solid phase transformations under far-from-equilibrium conditions forming metastable phases with unknown properties.}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Vijayan, Sriram and Bawane, Kaustubh and Giulia Dilemma, Fidelma and He, Lingfeng and Fink, Carolin and Jinschek, Joerg R}, year={2022}, month={Aug}, pages={1844–1846} }
 @article{wang_liu_murray_teng_jiang_bachhav_hawkins_perez_sun_bai_et al._2022, title={Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system}, volume={849}, ISSN={0921-5093}, url={http://dx.doi.org/10.1016/j.msea.2022.143475}, DOI={10.1016/j.msea.2022.143475}, abstractNote={Grain boundaries (GBs), known as two-dimensional defects, are omnipresent in polycrystalline metallic alloys and thus influence a wide range of mechanical properties under different environmental conditions like irradiation and corrosion. Therefore, quantifying the strength of individual GBs is critical for understanding the degradation of mechanical properties of materials under different conditions. In this study we developed an efficient approach for the fabrication of micro-tensile specimens with a GB almost perpendicular to the tensile direction, which is expected to advance the development of individual GB tensile testing at micro or nanoscale in a wide scope of materials. An in-situ cantilever micro-tensile testing method was developed and used to quantify the strength of a ∑3 GB in Inconel X-750 with the combination of finite element modeling. The average ultimate tensile strength (UTS) of a non-irradiated ∑3 GB is estimated at around 1.4 GPa, comparable to that of a neutron-irradiated ∑3 GB with a dose of ∼1.5 dpa (1.3 GPa). Moreover, the in-situ push-to-pull micro-tensile testing technique developed in this work provides valuable insights into the high-angle GB deformation and fracture behavior. This method generates qualitatively similar ductility behavior before and after neutron irradiation as the bulk material testing. However, the ductility and UTS values obtained from this method are different from bulk measurements due to vastly different specimen dimensions.}, journal={Materials Science and Engineering: A}, publisher={Elsevier BV}, author={Wang, Yachun and Liu, Xiang and Murray, Daniel J. and Teng, Fei and Jiang, Wen and Bachhav, Mukesh and Hawkins, Laura and Perez, Emmanuel and Sun, Cheng and Bai, Xianming and et al.}, year={2022}, month={Aug}, pages={143475} }
 @article{pena_morell-pacheco_shiau_kombaiah_he_hawkins_gabriel_garner_shao_2022, title={Microstructural changes of proton irradiated Hastelloy-N and in situ micropillar compression testing of one single grain at different local damage levels}, url={https://doi.org/10.1016/j.jnucmat.2022.153939}, DOI={10.1016/j.jnucmat.2022.153939}, abstractNote={In situ micropillar compression was used to study the deformation of proton-irradiated Hastelloy-N at different damage levels. Multiple pillars were prepared from a single grain along the cross-section of 2.5 MeV proton-irradiated Hastelloy-N. Depending on the location of micropillars, the critical resolved shear stress was obtained as a function of local damage levels. Such an approach eliminates the variation of yield stress due to the difference in the Schmid factor . Microstructural characterization showed complicated defect structures , including (a) dislocation loops with many in corduroy-like alignments, (2) dislocations pile up, (3) element segregation, and (4) twin boundaries. Silicon atoms are found to segregate at dislocation lines , loops, and twin boundaries and form complicated patterns at nanometer scales. These complexities make it difficult to conclude which hardening mechanism contributes the most to the hardness changes. The critical resolved shear stress, τ c r s s , and hardening exponents were both extracted as a function of displacements per atom values up to 2.3. There was a 60% increase in τ c r s s at the highest damage level.}, journal={Journal of Nuclear Materials}, author={Pena, Miguel and Morell-Pacheco, Andres and Shiau, Ching-Heng and Kombaiah, Boopathy and He, Lingfeng and Hawkins, Laura and Gabriel, Adam and Garner, Frank A. and Shao, Lin}, year={2022}, month={Nov} }
 @article{parkin_moorehead_elbakhshwan_zhang_xiu_he_bachhav_sridharan_couet_2022, title={Phase stability, mechanical properties, and ion irradiation effects in face-centered cubic CrFeMnNi compositionally complex solid-solution alloys at high temperatures}, volume={565}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2022.153733}, DOI={10.1016/j.jnucmat.2022.153733}, abstractNote={Two CrFeMnNi face-centered cubic complex concentrated solid-solution alloys (CSA) have been evaluated for phase stability, mechanical properties, and radiation damage effects from heavy ions. Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35 were predicted by thermodynamic calculations to phase separate and maintain a single phase at 700 °C, respectively. Aging experiments at this temperature confirmed varying degrees of precipitation of a body-centered cubic phase in both Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35. The alloys showed promising strength in tensile deformation at room temperature, with yield strengths of 155 MPa and 151 MPa for Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35, respectively. At 500 °C, the yield strength of Cr18Fe27Mn27Ni28 fell to 93 MPa, and to 100 MPa in Cr15Fe35Mn15Ni35. Unlike Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35 gained some ductility at 500 °C compared to room temperature. The two CSAs were irradiated to 75 dpa at 500 °C in the plateau region of the displacement curve using 3.7 MeV Ni2+ ions, alongside model alloy 709 as a reference. Irradiation results produced similar densities and sizes of dislocations loops in the two CSAs compared to the reference. However, while large voids form in the plateau region of Cr18Fe27Mn27Ni28, small voids form just beyond the displacement peak of Cr15Fe35Mn15Ni35. Atom probe tomography and energy dispersive X-ray spectroscopy-equipped scanning transmission electron microscopes were used to characterize the alloys for changes in chemical distribution.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Parkin, Calvin and Moorehead, Michael and Elbakhshwan, Mohamed and Zhang, Xuan and Xiu, Pengyuan and He, Lingfeng and Bachhav, Mukesh and Sridharan, Kumar and Couet, Adrien}, year={2022}, month={Jul}, pages={153733} }
 @article{sun_wu_bao_li_wan_li_he_2022, title={Preparation and strengthening mechanism of prestressed ceramic tile components}, url={https://doi.org/10.1111/ijac.13757}, DOI={10.1111/ijac.13757}, abstractNote={Abstract}, journal={International Journal of Applied Ceramic Technology}, author={Sun, Yi and Wu, Tianye and Bao, Yiwang and Li, Yueming and Wan, Detian and Li, Kai and He, Lingfeng}, year={2022}, month={Jan} }
 @article{field_patki_sharaf_sun_hawkins_lynch_jacobs_morgan_he_field_2022, title={Real-time, On-Microscope Automated Quantification of Features in Microcopy Experiments Using Machine Learning and Edge Computing}, volume={28}, ISSN={1435-8115 1431-9276}, url={http://dx.doi.org/10.1017/S1431927622007929}, DOI={10.1017/S1431927622007929}, abstractNote={Machine learning (ML) techniques, including deep learning-based object detection models, are rapidly becoming common in data intensive microscopy workflows. The rise of ML over human-based, or even “classic” machine vision techniques (e.g., image thresholding, Hough transforms, etc.) is the result of ML techniques being significantly faster, less computationally intensive (on inference), and highly repeatable between various microscopists and research groups [1]. Furthermore, ML techniques are far easier to scale towards large (>16M pixels) image datasets, including time-series data collected via video-based formats. Scalable ML is achieved with ML-based methods becoming more light weight and computationally inexpensive, while commodity computing hardware, such as graphical processing units (GPUs) and tensor cores, are witnessing exponential growth rates in performance. However, adoption of ML-based methods in various microscopy workflows has been restricted to only a few “super-user” groups due to the limited toolsets for performing distributed ML computing and difficulties with sharing software stacks that can be easily downloaded and run by any microscopist.}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Field, Kevin G and Patki, Priyam and Sharaf, Nasir and Sun, Kai and Hawkins, Laura and Lynch, Matthew and Jacobs, Ryan and Morgan, Dane D and He, Lingfeng and Field, Christopher R}, year={2022}, month={Aug}, pages={2046–2048} }
 @article{yang_liu_song_he_bankson_hamilton_prorok_lou_2022, title={Sensitization, desensitization, and carbide evolution of Alloy 800H made by laser powder bed fusion}, volume={50}, url={http://dx.doi.org/10.1016/j.addma.2021.102547}, DOI={10.1016/j.addma.2021.102547}, abstractNote={Additively manufactured (AM) Alloy 800H made by laser powder bed fusion (L-PBF) exhibited a greater degree of sensitization than wrought 800H, leading to both intergranular and intercellular corrosion. Dislocation cellular boundaries showed mild Cr depletion as compared to high-angle grain boundaries (HAGBs). Boundary misorientation of < 5⁰ was found mitigating sensitization. Carbide growth in AM 800H was controlled by particle net growth at the early stage and Ostwald ripening at the later stage. Cellular structure was confirmed producing faster elemental diffusion than bulk diffusion and random dislocation structures by cold rolling, and leading more rapid carbide growth, chromium depletion, and desensitization on HAGBs. Sensitization along cellular boundaries is not the main cause of intercellular corrosion in the sensitized AM Alloy 800H.}, journal={Additive Manufacturing}, publisher={Elsevier BV}, author={Yang, Jingfan and Liu, Xiang and Song, Miao and He, Lingfeng and Bankson, Stephen and Hamilton, Michael and Prorok, Bart and Lou, Xiaoyuan}, year={2022}, month={Feb}, pages={102547} }
 @article{herrmann_zhao_bawane_he_tolman_pu_2022, title={Synthesis and characterization of uranium trichloride in alkali-metal chloride media}, volume={565}, url={https://doi.org/10.1016/j.jnucmat.2022.153728}, DOI={10.1016/j.jnucmat.2022.153728}, abstractNote={Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Herrmann, Steven D. and Zhao, Haiyan and Bawane, Kaustubh K. and He, Lingfeng and Tolman, Kevin R. and Pu, Xiaofei}, year={2022}, month={Jul}, pages={153728} }
 @article{hurley_el-azab_bryan_cooper_dennett_gofryk_he_khafizov_lander_manley_et al._2022, title={Thermal Energy Transport in Oxide Nuclear Fuel}, url={https://doi.org/10.1021/acs.chemrev.1c00262}, DOI={10.1021/acs.chemrev.1c00262}, abstractNote={To efficiently capture the energy of the nuclear bond, advanced nuclear reactor concepts seek solid fuels that must withstand unprecedented temperature and radiation extremes. In these advanced fuels, thermal energy transport under irradiation is directly related to reactor performance as well as reactor safety. The science of thermal transport in nuclear fuel is a grand challenge as a result of both computational and experimental complexities. Here we provide a comprehensive review of thermal transport research on two actinide oxides: one currently in use in commercial nuclear reactors, uranium dioxide (UO2), and one advanced fuel candidate material, thorium dioxide (ThO2). In both materials, heat is carried by lattice waves or phonons. Crystalline defects caused by fission events effectively scatter phonons and lead to a degradation in fuel performance over time. Bolstered by new computational and experimental tools, researchers are now developing the foundational work necessary to accurately model and ultimately control thermal transport in advanced nuclear fuels. We begin by reviewing research aimed at understanding thermal transport in perfect single crystals. The absence of defects enables studies that focus on the fundamental aspects of phonon transport. Next, we review research that targets defect generation and evolution. Here the focus is on ion irradiation studies used as surrogates for damage caused by fission products. We end this review with a discussion of modeling and experimental efforts directed at predicting and validating mesoscale thermal transport in the presence of irradiation defects. While efforts in these research areas have been robust, challenging work remains in developing holistic tools to capture and predict thermal energy transport across widely varying environmental conditions.}, journal={Chemical Reviews}, author={Hurley, David H. and El-Azab, Anter and Bryan, Matthew S. and Cooper, Michael W. D. and Dennett, Cody A. and Gofryk, Krzysztof and He, Lingfeng and Khafizov, Marat and Lander, Gerard H. and Manley, Michael E. and et al.}, year={2022}, month={Feb} }
 @article{jiang_he_dennett_khafizov_mann_hurley_2022, title={Unraveling small-scale defects in irradiated ThO2 using kinetic Monte Carlo simulations}, url={https://doi.org/10.1016/j.scriptamat.2022.114684}, DOI={10.1016/j.scriptamat.2022.114684}, abstractNote={Point defects and their clusters generated through irradiation can have significant impact on the physical and mechanical properties of materials. However, direct experimental visualization of these small-scale defects using high-resolution scanning transmission electron microscopy remains a challenging task. Here, using thorium dioxide (ThO2) with the fluorite structure as a model system, we demonstrate the use of ab initio basin-hopping simulations in synergy with object kinetic Monte Carlo simulations as a powerful tool for identifying small defect complexes in irradiated materials. In addition to providing quantitative insights into defect evolution in ThO2 under irradiation, our study reveals an unexpected role of bound anti-Schottky defect clusters in mediating defect transport. Remarkably, despite their poor thermal stability against dissociation at high temperatures, the transient formation of bound anti-Schottky defects under irradiation and their subsequent migration provide the dominant mechanism for the growth of large interstitial loops that have been experimentally observed in ThO2.}, journal={Scripta Materialia}, author={Jiang, Chao and He, Lingfeng and Dennett, Cody A. and Khafizov, Marat and Mann, J. Matthew and Hurley, David H.}, year={2022}, month={Jun} }
 @article{bawane_liu_gakhar_woods_ge_xiao_lee_halstenberg_dai_mahurin_et al._2022, title={Visualizing time-dependent microstructural and chemical evolution during molten salt corrosion of Ni-20Cr model alloy using correlative quasi in situ TEM and in situ synchrotron X-ray nano-tomography}, volume={195}, url={http://dx.doi.org/10.1016/j.corsci.2021.109962}, DOI={10.1016/j.corsci.2021.109962}, abstractNote={In situ monitoring of corrosion processes is important to fundamentally understand the kinetics and evolution of materials in harsh environments. A quasi in situ transmission electron microscopy technique was utilized to study microstructural and chemical evolution of a Ni-20Cr disc sample exposed to molten KCl-MgCl2 salt for 60 s in consecutive 20 s iterations. In situ synchrotron X-ray nano-tomography was performed to characterize the morphological evolution of a Ni-20Cr microwire exposed to molten KCl-MgCl2. Both techniques captured key corrosion events and revealed mechanisms at different time and length scales, potentially bringing greater insights and deeper understanding beyond conventional analysis.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Bawane, Kaustubh and Liu, Xiaoyang and Gakhar, Ruchi and Woods, Michael and Ge, Mingyuan and Xiao, Xianghui and Lee, Wah-Keat and Halstenberg, Philip and Dai, Sheng and Mahurin, Shannon and et al.}, year={2022}, month={Feb}, pages={109962} }
 @article{he_hawkins_yang_liu_song_lou_zhang_shao_schwen_2021, title={Advanced Characterization of Additively Manufactured 316L Stainless Steel for Nuclear Applications}, volume={27}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/S1431927621007789}, DOI={10.1017/S1431927621007789}, abstractNote={,}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={He, Lingfeng and Hawkins, Laura and Yang, Jingfan and Liu, Xiang and Song, Miao and Lou, Xiaoyuan and Zhang, Yongfeng and Shao, Lin and Schwen, Daniel}, year={2021}, month={Jul}, pages={2160–2161} }
 @article{dennett_deskins_khafizov_hua_khanolkar_bawane_fu_mann_marianetti_he_et al._2021, title={An integrated experimental and computational investigation of defect and microstructural effects on thermal transport in thorium dioxide}, volume={213}, url={https://doi.org/10.1016/j.actamat.2021.116934}, DOI={10.1016/j.actamat.2021.116934}, abstractNote={Advanced nuclear reactor concepts aim to use fuels that must withstand unprecedented temperature and radiation extremes. In these fuels, thermal energy transport under irradiation is directly related to fuel longevity, reactor safety, and is arguably one of the most important performance metrics. Here we provide a comprehensive, first-principles-informed treatment of phonon mediated thermal transport in a defect-bearing actinide oxide with direct comparison to experimental measurements. Pristine and proton irradiated thorium dioxide was chosen as a model system to treat the complexity of thermal transport in the presence of lattice defects. A thermal transport model is implemented using the linearized Boltzmann transport equation (LBTE) with input from first principles calculations and defect evolution models. Density functional theory is used to calculate phonon dispersion in thorium dioxide and used as an input to calculate both intrinsic and extrinsic, defect-induced relaxation times. In addition, a defect evolution model is benchmarked using microstructure characterization of as-irradiated thorium dioxide using a combination of electron microscopy and optical spectroscopy. The output of the LBTE is compared directly to mesoscopic measurements of thermal conductivity on length scales commensurate with defect accumulation. Parametric measurements of conductivity with irradiation dose and temperature suggest a saturation in the reduction of thermal conductivity with increasing defect generation, which is partially captured in our defect evolution model and LBTE framework. This comprehensive, atomistic- to meso-scale treatment provides the necessary basis to investigate thermal transport under irradiation in more complex systems that exhibit strong electron correlation.}, journal={Acta Materialia}, publisher={Elsevier BV}, author={Dennett, Cody A. and Deskins, W. Ryan and Khafizov, Marat and Hua, Zilong and Khanolkar, Amey and Bawane, Kaustubh and Fu, Lyuwen and Mann, J. Matthew and Marianetti, Chris A. and He, Lingfeng and et al.}, year={2021}, month={Jul}, pages={116934} }
 @article{bawane_manganaris_wang_sure_ronne_halstenberg_dai_gill_sasaki_chen-wiegart_et al._2021, title={Determining oxidation states of transition metals in molten salt corrosion using electron energy loss spectroscopy}, volume={197}, DOI={10.1016/j.scriptamat.2021.113790}, abstractNote={This work utilizes electron energy loss spectroscopy (EELS) to identify oxidation state of alloying elements in Ni-based alloys after exposure to molten chloride salt systems. Pure Ni and Ni-20Cr model alloy were corroded in molten ZnCl2 and KCl-MgCl2 under argon atmosphere at various temperatures. Oxidation states of Cr (Cr3+) and Ni (Ni2+) in the molten salt after corrosion were determined by monitoring changes in the L2,3 edges of corresponding EELS spectra. Oxidation state mapping technique using principal component analysis and multiple linear least squares fitting in HyperSpy Python package was developed.}, journal={Scripta Materialia}, publisher={Elsevier BV}, author={Bawane, Kaustubh and Manganaris, Panayotis and Wang, Yachun and Sure, Jagadeesh and Ronne, Arthur and Halstenberg, Phillip and Dai, Sheng and Gill, Simerjeet K. and Sasaki, Kotaro and Chen-Wiegart, Yu-chen Karen and et al.}, year={2021}, month={May}, pages={113790} }
 @article{xiu_jin_bawane_tyburska-püschel_jaques_field_giglio_he_2021, title={Dislocation Loops in Proton Irradiated Uranium-Nitrogen-Oxygen System}, volume={557}, url={http://dx.doi.org/10.1016/j.jnucmat.2021.153244}, DOI={10.1016/j.jnucmat.2021.153244}, abstractNote={In this study, we investigated the type of dislocation loops formed in the proton-irradiated uranium-nitrogen-oxygen (U-N-O) system, which involves uranium mononitride (UN), uranium sesquinitride (α-U2N3), and uranium dioxide (UO2) phases. The dislocation loop formation is examined using specimens irradiated at 400°C and 710°C. Based on the detailed transmission-based electron microscopy characterization with i) the morphology-based on-zone and ii) the invisibility-criterion based two-beam condition imaging techniques, only a single type of dislocation loop in each phase is found: a/2⟨110⟩, a/2⟨111⟩, or a/3⟨111⟩ dislocation loops in UN, α-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in the UN phase indicate a critical loop size of ∼ 6 nm, above which perfect loops are thermodynamically favorable. This could explain the absence of faulted loops in the experimental observation of the irradiated UN phase at two temperatures. This work will enhance the understanding of irradiation induced microstructural evolution for uranium mononitride as an advanced nuclear fuel for the next-generation nuclear reactors.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Xiu, Pengyuan and Jin, Miaomiao and Bawane, Kaustubh and Tyburska-Püschel, Beata and Jaques, Brian J. and Field, Kevin G. and Giglio, Jeffrey J. and He, Lingfeng}, year={2021}, month={Dec}, pages={153244} }
 @article{yan_liu_he_stubbins_2021, title={Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91}, volume={557}, url={http://dx.doi.org/10.1016/j.jnucmat.2021.153207}, DOI={10.1016/j.jnucmat.2021.153207}, abstractNote={A Fe-9Cr ferritic martensitic (F/M) steel T91 was neutron-irradiated in the Advanced Test Reactor up to 3.96 dpa in two temperature ranges, 466 °C to 534 °C and 571 °C to 632 °C. The microstructure evolution including dislocation loops, precipitation, segregation of elements and phase stability were studied using analytical scanning-transmission electron microscopy and atom probe tomography. The hardening induced by irradiation was measured by nanoindentation. Ni/Si/Mn clusters were identified in all conditions except the one irradiated around 600 °C to 3.23 dpa. The compositions of Ni/Si/Mn clusters were found to be converging to G phase stoichiometrically with increasing dose, with Mn partially substituted by Cu. Significant coarsening of this phase was observed in high temperature cases, with total dissolution of intragranular G phase after prolonged irradiation. A similar trend was also identified for dislocation loops. The results obtained in this experiment provide evidences that the absence of dislocation loops in some specimens irradiated to high dose level under high irradiation temperature range (typically above 500 °C) is not due to the suppression of nucleation and growth by high point defect recombination rate, but rather fast coalescence of dislocation loops. Hardness measurements show different dose dependences for two temperature ranges. For specimen irradiated around 600 °C significant hardening before 0.5 dpa followed by softening process was observed, while in lower temperature range (450 °C ~ 500 °C) the normal hardening pattern with increasing dose was observed.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yan, Huan and Liu, Xiang and He, Lingfeng and Stubbins, James}, year={2021}, month={Dec}, pages={153207} }
 @article{yang_song_hawkins_liu_he_lou_2021, title={Effects of heat treatment on corrosion fatigue and stress corrosion crack growth of additive-manufactured Alloy 800H in high-temperature water}, volume={191}, url={http://dx.doi.org/10.1016/j.corsci.2021.109739}, DOI={10.1016/j.corsci.2021.109739}, abstractNote={Crack growth behaviour of additively manufactured (AM) Alloy 800H were studied under cyclic and static load in high-temperature water, focusing on heat treatment effects. Under cyclic loading, heat treatment gently affected corrosion fatigue. The growth rate is comparable to wrought 800H. As the load cycle frequency reduced to below 0.001 Hz, AM 800H exhibited inconsistent stress corrosion cracking response as compared to wrought 800H. Depending on heat treatment, inhomogeneous crack propagation and localized pinning were observed in stress-relieved and recrystallized AM 800H, attributed to carbides, microstructural inhomogeneity, and twin boundaries. Grain boundary sensitization at ˜10 wt% Cr caused fast crack growth.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Yang, Jingfan and Song, Miao and Hawkins, Laura R. and Liu, Xiang and He, Lingfeng and Lou, Xiaoyuan}, year={2021}, month={Oct}, pages={109739} }
 @article{xiao_he_bai_2021, title={First principle studies of effects of solute segregation on grain boundary strength in Ni-based alloys}, volume={874}, DOI={10.1016/j.jallcom.2021.159795}, abstractNote={Thermal annealing or radiation induced segregation of solute and impurity elements to grain boundaries (GBs) in metallic alloys changes GB chemistry and thus can alter the GB cohesive strength. In this work, first principles based density functional theory calculations are conducted to study how the segregation of substitutional solute and impurity elements (Al, C, Cr, Cu, P, Si, Ti, Fe, which are present in Ni-based X-750 alloys) influences the cohesive strength of Σ3111,Σ3112,Σ5210 and Σ5310 GBs in Ni. It is found that C and P show strong embrittlement potencies while Cr and Ti can strengthen GBs in most cases. Other solute elements, including Si, have mixed but insignificant effects on GB strength. In terms of GB character effect, these solute and impurity elements affect the GB strength of the Σ5210 GB most and that of the Σ3111 least. Detailed analyses of solute-GB chemical interactions are conducted using electron localization function, charge density map, partial density of states, and Bader charge analysis. The results suggest that the bond type and charge transfer between solutes and Ni atoms at GBs may play important roles on affecting the GB strength. For non-metallic solute elements (C, P, Si), their interstitial forms are also studied but their effects on GB strength are weaker than their substitutional counterparts.}, journal={Journal of Alloys and Compounds}, publisher={Elsevier BV}, author={Xiao, Ziqi and He, Lingfeng and Bai, Xian-Ming}, year={2021}, month={Sep}, pages={159795} }
 @article{liu_capriotti_yao_harp_benson_wang_teng_he_2021, title={Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding}, volume={544}, DOI={10.1016/j.jnucmat.2020.152588}, abstractNote={As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits.The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features.Here, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom.Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel.Instead, most lanthanides were retained in the newly formed UZr2 phase in the fuel center region.The interdiffusion of iron and uranium resulted in tetragonal (U,Zr)6Fe phase (space group I4/mcm) and cubic (U,Zr)(Fe,Cr)2 phase (space group Fd 3 ̅ m).The (U,Zr)(Fe,Cr)2 phase contains a high density of voids and intergranular uranium monocarbides of NaCltype crystal structure (space group Fm3 ̅ m).At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed.Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P42/mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified.Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular Urich carbides formed as a result of the diffusion of uranium into the cladding.These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Liu, Xiang and Capriotti, Luca and Yao, Tiankai and Harp, Jason M. and Benson, Michael T. and Wang, Yachun and Teng, Fei and He, Lingfeng}, year={2021}, month={Feb}, pages={152588} }
 @article{khanolkar_yao_hua_dennett_reese_schley_he_kennedy_hurley_2021, title={In situ monitoring of microstructure evolution during thermal processing of uranium-zirconium alloys using laser-generated ultrasound}, volume={4}, DOI={10.1016/j.jnucmat.2021.153005}, abstractNote={Laser-generated ultrasound was used to monitor microstructure evolution during thermal processing of as-cast, polycrystalline binary uranium-zirconium metallic fuel alloys with compositions of U-20wt.%Zr (U-20Zr), U-50wt.%Zr (U-50Zr) and U-80wt.%Zr (U-80Zr). Ultrasonic waveforms were recorded during heating and cooling the samples from room temperature to >973 K and back. A phase transition temperature for all three compositions was estimated from the temperature at which an abrupt and rapid reduction in ultrasonic velocities was observed. Microstructural features on the length scale of tens of micrometers were inferred from the observation of scattering of ultrasonic waves by elastic heterogeneities above ~823 K in U-20Zr, while a hysteresis in the ultrasonic velocities of U-80Zr upon cooling was attributed to a partial retention of the high temperature phase following thermal annealing. The U-50Zr alloy exhibited a reversible viscoelastic response above 933 K, as evidenced by the observation of high frequency attenuation of the shear component of the waveforms at high temperature. Ultrasonic measurements were supplemented by in situ transmission electron microscopy (TEM). The TEM images revealed that the δ-U-Zr matrix in the three compositions underwent a spinodal decomposition above ~823 K into nanoscale regions. The ultrasonic measurements revealed larger, micron-scale structure evolution in the U-20Zr alloy at the same temperature. This large-scale structure is associated with heterogeneous regions having different Zr content. Our findings show the potential heating rate dependence of microstructural evolution in U-Zr alloys and highlight differences in the thermomechanical response and associated length scales during thermal annealing between single- and dual-phase compositions. These results demonstrate the utility of laser ultrasonics to rapidly and efficiently scan phase boundaries and monitor micrometer-scale structure evolution in metallic fuel alloys.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Khanolkar, A. and Yao, T. and Hua, Z. and Dennett, C.A. and Reese, S.J. and Schley, R.S. and He, L. and Kennedy, J.R. and Hurley, D.H.}, year={2021}, month={Apr}, pages={153005} }
 @article{chauhan_pakarinen_yao_he_hurley_khafizov_2021, title={Indirect characterization of point defects in proton irradiated ceria}, volume={15}, url={https://doi.org/10.1016/j.mtla.2021.101019}, DOI={10.1016/j.mtla.2021.101019}, abstractNote={A rate theory model informed by multimodal characterization is used to evaluate the concentration of point defects in irradiated materials. Cerium dioxide (CeO2) is used as a model ionic compound, whose cation and anion sublattice point defects evolve independent of each other, but extended defects in the form of dislocation loops retain stoichiometry of the compound. To demonstrate this, we performed extensive measurement of defect evolution in CeO2 exposed to energetic protons at elevated temperature. Two sintered polycrystalline CeO2 samples were irradiated with protons having energies up to 2.5 MeV. Both samples were irradiated at 600°C to a dose of 0.14 dpa, but with different dose rates. These irradiation conditions produced a rich microstructure with resolvable extended defects and a significant concentration of point defects. Dislocation loop density revealed by electron microscopy and lattice constant changes measured by X-ray diffraction (XRD), and mesoscale thermal conductivity measurements were used to parameterize the rate theory model. The model, which includes point defect generation, recombination, and clustering into stoichiometric interstitial loops, suggests a large concentration of cerium vacancies and interstitials is present under these irradiation conditions. This work lays the foundation for expanded multimodal characterization of microstructure, including more direct characterization of point defects using optical spectroscopies.}, journal={Materialia}, publisher={Elsevier BV}, author={Chauhan, Vinay S. and Pakarinen, Janne and Yao, Tiankai and He, Lingfeng and Hurley, David H. and Khafizov, Marat}, year={2021}, month={Mar}, pages={101019} }
 @article{gill_sure_wang_layne_he_mahurin_wishart_sasaki_2021, title={Investigating corrosion behavior of Ni and Ni-20Cr in molten ZnCl2}, volume={179}, DOI={10.1016/j.corsci.2020.109105}, abstractNote={Corrosion of Ni and Ni-20Cr in molten ZnCl2 under argon atmosphere at 593 K was investigated using cyclic polarization in combination with various electron microscopy techniques. Cyclic polarization measurements yielded significantly lower corrosion rates for pure Ni than for Ni-20Cr. Corrosion of the Ni-20Cr alloy is attributed to selective anodic dissolution of Cr, which forms a chromium chloride film at the metal-salt interface. The corrosion in Ni and Ni-20Cr is accelerated by grain boundary attack through intergranular corrosion, followed by the formation of pore-salt network regions. The related corrosion mechanisms and dissolution kinetics are discussed.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Gill, Simerjeet K. and Sure, Jagadeesh and Wang, Yachun and Layne, Bobby and He, Lingfeng and Mahurin, Shannon and Wishart, James F. and Sasaki, Kotaro}, year={2021}, month={Feb}, pages={109105} }
 @article{bachhav_kane_teng_cappia_he_2021, title={Isotopic Analysis of Irradiated Ceramic Fuel for Burnup and Microchemical Assessment Using Atom Probe Tomography.}, volume={27}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/S1431927621002002}, DOI={10.1017/S1431927621002002}, abstractNote={Ceramic nuclear fuels, such as UO 2 and mixed oxide (MOX), are the primary fuels used in the current commercial light water reactors and are also fuel candidates in advanced reactors. Oxide fuels are known for their thermo-mechanical and adequate physico-chemical stability under harsh operating condition of nuclear reactors and are hence the preferred fuel of choice. The energy supplied by uranium (U) based fuel is determined by consumption of U-235 isotope. This burnup of fuel leads to the large number of fission products (FP) generated during burnup which can impact the chemical and mechanical properties of the fuel. Precise knowledge on U isotope quantification is desired for burnup assessment and its correlation to distribution of different fission products and its chemical states. Atom probe microscopy is a unique and highly accurate single model peak approach data}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Bachhav, Mukesh and Kane, Joshua and Teng, Fei and Cappia, Fabiola and He, Lingfeng}, year={2021}, month={Jul}, pages={416–417} }
 @article{yuan_zhang_he_zuo_2021, title={Machine Learning Based Precision Orientation and Strain Mapping from 4D Diffraction Datasets}, volume={27}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/S1431927621004785}, DOI={10.1017/S1431927621004785}, abstractNote={Four-dimensional diffraction datasets (4D-DDs) collected using the scanning electron nanodiffraction (SEND) [1] or 4D scanning transmission electron microscopy (4D-STEM) [2] techniques have gained increasing popularity in the electron microscopy community for their versatility in both electron diffraction and imaging. The advantage of having a 4D-DD over the traditional 2D imaging techniques comes from the rich information captured in diffraction patterns, which can be related to the sample structure, electric and magnetic fields. Experimental acquisition of large 4D-DDs has been significantly improved with the development of fast electron pixel array detectors [3]. Now, the large data}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Yuan, Renliang and Zhang, Jiong and He, Lingfeng and Zuo, Jian-Min}, year={2021}, month={Jul}, pages={1276–1278} }
 @article{lemma_jensen_kane_chen_liu_capriotti_adkins_kombaiah_winston_he_et al._2021, title={Metallic Fast Reactor Separate Effect Studies for Fuel Safety}, volume={7}, DOI={10.1115/1.4049721}, abstractNote={Abstract}, number={4}, journal={Journal of Nuclear Engineering and Radiation Science}, publisher={ASME International}, author={Lemma, Fidelma G. Di and Jensen, Colby B. and Kane, Joshua J. and Chen, Wei-Ying and Liu, Xiang and Capriotti, Luca and Adkins, Cynthia A. and Kombaiah, Boopathy and Winston, Alexander J. and He, Lingfeng and et al.}, year={2021}, month={Apr} }
 @article{yu_bachhav_teng_he_couet_2021, title={Nanoscale redistribution of alloying elements in high-burnup AXIOM-2 (X2®) and their effects on in-reactor corrosion}, volume={190}, url={http://dx.doi.org/10.1016/j.corsci.2021.109652}, DOI={10.1016/j.corsci.2021.109652}, abstractNote={A comprehensive characterization study was carried out on 1 and 4 cycles X2®, aiming to reveal effect of irradiation-induced alloying element redistribution on the in-reactor corrosion kinetics. Using a combination of (scanning) transmission electron microscopy ((S)/TEM) and atom probe tomography (APT), the results strongly evidenced the existence of Nb-rich native precipitates and irradiation-induced platelets (IIPs)/nanoclusters in both metal and oxide. In addition, the suboxide may reject Nb content back into the metal matrix to achieve a thermodynamic equilibrium state. On the other hand, Fe-rich transition cracks were observed for the first time, along with Fe-rich oxide grain boundaries.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Yu, Zefeng and Bachhav, Mukesh and Teng, Fei and He, Lingfeng and Couet, Adrien}, year={2021}, month={Sep}, pages={109652} }
 @article{he_khafizov_jiang_tyburska-püschel_jaques_xiu_xu_meyer_sridharan_butt_et al._2021, title={Phase and defect evolution in uranium-nitrogen-oxygen system under irradiation}, url={https://doi.org/10.1016/j.actamat.2021.116778}, DOI={10.1016/j.actamat.2021.116778}, abstractNote={Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to study the phase and defect evolution under proton irradiation in nitride-oxide composite. Phase composition, crystallographic orientation relationships (ORs) and dislocation loops were characterized using X-ray diffraction, transmission electron microscopy, and energy dispersive X-ray spectroscopy techniques. Proton-irradiation at elevated temperatures promoted the transformation of UN into uranium sesquinitride (U2N3) and UO2 phases. U2N3 and UO2 formed a fully coherent structure with two ORs: {002}U2N3‖{002}UO2 and [001]U2N3‖[001]UO2; U2N3{101}‖UO2{101} and U2N3[101]‖UO2[101] due to low lattice misfit (2.3%) and low interfacial energy (127 mJ/m2). Observed oxidation of UN and coherent interface are consistent with density-functional theory calculations which suggest lower energy for oxidized configuration and low energy of the interface. The dislocation loops grew while their number density decreased with the temperature and dose. The loop size was over three times larger in two nitride phases than that in UO2, while the number density was one order of magnitude higher in UO2 than in nitride phases. Loop density and diameter were analyzed using a kinetic rate theory that considers stoichiometric loop evolution. This analysis led to the conclusion in all compounds loop growth is governed by mobility of uranium interstitials, and enabled measurement of diffusion coefficients of uranium interstitials and non-metal interstitials and vacancies. This analysis provided a comparative study of early stage of microstructure evolution under irradiation which has implications for use of this mixture as advanced fuel in nuclear energy systems.}, journal={Acta Materialia}, author={He, Lingfeng and Khafizov, Marat and Jiang, Chao and Tyburska-Püschel, Beata and Jaques, Brian J. and Xiu, Pengyuan and Xu, Peng and Meyer, Mitchell K. and Sridharan, Kumar and Butt, Darryl P. and et al.}, year={2021}, month={Apr} }
 @article{yan_liu_he_stubbins_2021, title={Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9}, volume={557}, url={http://dx.doi.org/10.1016/j.jnucmat.2021.153252}, DOI={10.1016/j.jnucmat.2021.153252}, abstractNote={Ferritic Martensitic (F/M) steel HT9 specimens were irradiated in the Advanced Test Reactor up to 4.16 dpa in three temperature ranges (roughly from 300 to 600 °C). The post-irradiation microstructure, including dislocation structure, precipitation and radiation-induced segregation (RIS) was characterized using analytical scanning / transmission electron microscopy (S/TEM), and atom probe tomography (APT). Irradiation hardening was measured using nanoindentation. The results reveal a distinctive pattern of dislocation and precipitate evolution at high temperature, around 600 °C, where various defects and precipitates formed in the low dose regime followed by a recovering process with increasing dose. Dislocation loops formed in all temperature ranges, and the growth of dislocation loops is unconstrained above certain critical temperature, contributing to the increasing dislocation density even prior to doses of 0.5 dpa at 600 °C. Ni/Mn/Si clusters were identified in all temperature ranges and the compositions of these clusters converged to G phase stoichiometrically. Significant coarsening of G phase particles was observed at 600 °C, accompanied by the formation of G phase on grain boundaries. α’ precipitates were only found in the medium and low temperature ranges (below 500 °C). The number density and volume fraction were higher in the low temperature specimens, while larger particles were observed in the medium temperature range. RIS of Cr, Ni, Mn, Si, P was identified at dislocation lines, grain boundaries and phase boundaries, and the temperature dependence is consistent with previous studies. The RIS of Cr to the existing VN particles was confirmed by APT and may accelerate the transition of VN to Cr-rich nitrides. The irradiation hardening contribution from dislocation loops, dislocation lines, G phase and α’ phase was parsed based on a linear dispersed barrier hardening model. The results suggest that most irradiation hardening at high temperature is due to increasing dislocation density with dose.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yan, Huan and Liu, Xiang and He, Lingfeng and Stubbins, James}, year={2021}, month={Dec}, pages={153252} }
 @article{pakarinen_he_gan_nelson_el-azab_khafizov_allen_2021, title={Proton irradiation-induced blistering in UO2}, volume={11}, url={http://dx.doi.org/10.1557/s43580-021-00149-3}, DOI={10.1557/s43580-021-00149-3}, abstractNote={Abstract}, journal={MRS Advances}, publisher={Springer Science and Business Media LLC}, author={Pakarinen, Janne and He, Lingfeng and Gan, Jian and Nelson, Andrew T. and El-Azab, Anter and Khafizov, Marat and Allen, Todd R.}, year={2021}, month={Nov} }
 @article{liu_cinbiz_kombaiah_he_teng_lacroix_2021, title={Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding}, volume={556}, url={http://dx.doi.org/10.1016/j.jnucmat.2021.153196}, DOI={10.1016/j.jnucmat.2021.153196}, abstractNote={Structure of the pellet/cladding-interaction (PCI) layer of a high-burnup nuclear fuel with niobium-bearing cladding was investigated using modern transmission electron microscopy (TEM) techniques. Morphology of the PCI consisted of uniform zirconia layer on the cladding side and finger-like features towards to the ceramic fuel. The PCI layer showed a complex microstructure that constitutes three distinct zirconia layers formed by two zirconia phases. From cladding to fuel, monoclinic, tetragonal, and another tetragonal phase were determined via TEM diffraction patterns and precession electron diffraction (PED) using ASTAR microscopy platform (AMP). Characterizations showed that monoclinic and tetragonal phases close to the cladding were crack-free, but the tetragonal layer adjacent to the fuel contained cracks and pores. Presence of the monoclinic layer suggested that its formation involved the gaseous diffusion of oxygen below threshold dose of fission product damage before the fuel/cladding contact occurred. Furthermore, high-resolution TEM revealed compound phase of (U,Zr)O2 at fringes of zirconia fingers on the fuel side.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Liu, Xiang and Cinbiz, Mahmut Nedim and Kombaiah, Boopathy and He, Lingfeng and Teng, Fei and Lacroix, Evrard}, year={2021}, month={Dec}, pages={153196} }
 @article{bawane_liu_yao_khafizov_french_mann_shao_gan_hurley_he_2021, title={TEM Characterization of Dislocation Loops in Proton Irradiated Single Crystal ThO2}, volume={4}, DOI={10.1016/j.jnucmat.2021.152998}, abstractNote={This work focuses on the full characterization of dislocation loops induced by proton irradiation in single-crystal ThO2. Irradiation was performed using 2 MeV H+ ions with sample temperature at 600oC and a dose of up to 0.47 displacements per atom (dpa). Transmission electron microscopy (TEM) characterization was performed on a large number of dislocation loops. Burgers vector (b→) analysis using standard g→.b→=0 invisibility criterion revealed different variants of 1/3〈111〉 type dislocation loops. TEM analysis of edge-on dislocation loops was used to determine habit planes as {111} type. The nature of dislocation loops was revealed using the inside-outside contrast method as interstitial type. Rel-rod dark field images were obtained by selecting streak at g→=1/2[311¯] in diffraction pattern to identify the Frank loops present in the microstructure. Subsequent analysis of a single dislocation loop using atomic resolution scanning transmission electron microscopy (STEM) confirmed the interstitial nature of the loop with {111} habit plane.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Bawane, Kaustubh and Liu, Xiang and Yao, Tiankai and Khafizov, Marat and French, Aaron and Mann, J. Matthew and Shao, Lin and Gan, Jian and Hurley, David H. and He, Lingfeng}, year={2021}, month={Apr}, pages={152998} }
 @article{deskins_hamed_kumagai_dennett_peng_khafizov_hurley_el-azab_2021, title={Thermal conductivity of ThO2: Effect of point defect disorder}, url={https://doi.org/10.1063/5.0038117}, DOI={10.1063/5.0038117}, abstractNote={Thoria (ThO2) has lately gained attention due to its potential for use as a nuclear fuel. From a physics standpoint, ThO2 is an actinide-bearing material with no 5f electrons and is thus ideally suited as a baseline material for future studies of the physical properties of actinide systems with correlated electrons. Current investigations of ThO2 as a nuclear fuel focus on the influence of radiation-induced lattice defects on its thermal properties, especially the conductivity. This work presents a first investigation of the impact of point defect disorder on phonon thermal conductivity of ThO2 by solving the Boltzmann transport equation within the single-mode relaxation time approximation. The relaxation times of intrinsic, three-phonon scattering are calculated by a rigorous sampling of k-points within the irreducible Brillouin zone of the face-centered cubic crystal structure. The effect of point defects on the thermal conductivity of ThO2 is predicted using the classic model by Klemens for phonon relaxation times that result from the change in mass and induced lattice strain associated with point defects. Within this model, the change in force constants and atomic radii are computed using input from an atomistic model of ThO2. The defects considered are uranium substitution at a thorium site, oxygen vacancies and interstitials, and thorium vacancies and interstitials. The results show that the conductivity of ThO2 is highly sensitive to intrinsic point defects and less sensitive to U substitution on the cation sublattice.}, journal={Journal of Applied Physics}, author={Deskins, W. Ryan and Hamed, Ahmed and Kumagai, Tomohisa and Dennett, Cody A. and Peng, Jie and Khafizov, Marat and Hurley, David and El-Azab, Anter}, year={2021}, month={Feb} }
 @article{yuan_zhang_he_zuo_2021, title={Training artificial neural networks for precision orientation and strain mapping using 4D electron diffraction datasets}, volume={3}, DOI={10.1016/j.ultramic.2021.113256}, abstractNote={Techniques for training artificial neural networks (ANNs) and convolutional neural networks (CNNs) using simulated dynamical electron diffraction patterns are described. The premise is based on the following facts. First, given a suitable crystal structure model and scattering potential, electron diffraction patterns can be simulated accurately using dynamical diffraction theory. Secondly, using simulated diffraction patterns as input, ANNs can be trained for the determination of crystal structural properties, such as crystal orientation and local strain. Further, by applying the trained ANNs to four-dimensional diffraction datasets (4D-DD) collected using the scanning electron nanodiffraction (SEND) or 4D scanning transmission electron microscopy (4D-STEM) techniques, the crystal structural properties can be mapped at high spatial resolution. Here, we demonstrate the ANN-enabled possibilities for the analysis of crystal orientation and strain at high precision and benchmark the performance of ANNs and CNNs by comparing with previous methods. A factor of thirty improvement in angular resolution at 0.009˚ (0.16 mrad) for orientation mapping, sensitivity at 0.04% or less for strain mapping, and improvements in computational performance are demonstrated.}, journal={Ultramicroscopy}, publisher={Elsevier BV}, author={Yuan, Renliang and Zhang, Jiong and He, Lingfeng and Zuo, Jian-Min}, year={2021}, month={Mar}, pages={113256} }
 @article{yao_sen_wagner_teng_bachhav_ei-azab_murray_gan_hurley_wharry_et al._2021, title={Understanding spinodal and binodal phase transformations in U-50Zr}, volume={16}, DOI={10.1016/j.mtla.2021.101092}, abstractNote={Engineered spinodal decomposition and spinodal precursors to precipitation hardening are effective approaches for tailoring material's properties and performance. These approaches leverage the spinodal and binodal nature of miscibility gaps. However, little is known about the chemical and crystallographic mechanisms controlling phase evolution across the spinodal curve and into the binodal regime. This study aims to reveal spinodal-binodal phase transformation in a binary U-50 wt.% Zr model alloy through coupled X-ray diffraction (XRD), in-situ transmission electron microscopy (TEM), and atom probe tomography (APT). The hexagonal ⍵-UZr2+x phase initially undergoes a spinodal decomposition into interconnected nanosized hexagonal Zr-rich and hexagonal U-rich domains at ~575 °C, near the hexagonal-bcc phase transformation. Between 600 and 800 °C, the microstructure evolution is dominated by the coarsening and chemical purification of decomposed bcc domains and a transition from spinodal to binodal nucleation. At 800 °C, the metastable Zr domains transfer to stable α-Zr phase and stay the same phase up to 1000 °C. On the contrary, metastable U domains transfer to orthorhombic α’-U firstly at 800 °C and then to tetragonal β-U at 1000 °C. Spinodal decomposed domain size is found to be linearly related to extrinsic specimen geometry. The effect of oxygen on spinodal decomposition is also discussed. In-situ TEM observations successfully captured the spinodal decomposition and the subsequent transition to binodal phase separation with temperature, providing critical evidence for an expanded miscibility gap, concerning both composition and temperature, on the U-Zr phase diagram.}, journal={Materialia}, publisher={Elsevier BV}, author={Yao, Tiankai and Sen, Amrita and Wagner, Adrian and Teng, Fei and Bachhav, Mukesh and EI-Azab, Anter and Murray, Daniel and Gan, Jian and Hurley, David H. and Wharry, Janelle P. and et al.}, year={2021}, month={May}, pages={101092} }
 @article{liu_he_yan_bachhav_stubbins_2020, title={A transmission electron microscopy study of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750}, volume={528}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2019.151851}, DOI={10.1016/j.jnucmat.2019.151851}, abstractNote={The microstructure of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750 was investigated. Both alloys were irradiated at low dose rates (∼2 × 10−8 dpa/s) to a neutron fluence of 6.9 × 1022 n/cm2 (E > 0.1 MeV) at 371–389 °C. Different types of defects, including Frank loops, cavities, and precipitates were characterized. The Frank loops in Type 304 stainless steel (SS) are larger in size (∼50 nm in diameter) and lower in number density (2.58 × 1021 m−3), compared to most previous higher dose rate neutron irradiation studies. The Frank loops in X-750 have an average size 26.0 nm of and a number density of 9.44 × 1021 m−3. In 304 SS and X-750, cavities are of ∼20 nm and ∼14 nm in diameter, respectively. The swelling of both alloys was found to be insignificant. In 304 SS, Ni and Si were found enriched at the cavity surfaces and Ni,Si-rich precipitates were also found. Multivariate statistical analysis using non-negative matrix factorization reveals that these Ni,Si-rich precipitates contain only ∼5.7 at.% Si, differing from the Ni3Si γ’ precipitates found in several previous studies. In X-750, L12-structured γ’ precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni3(Ti,Al)) of the γ’ precipitates and the superlattice reflections confirmed the stability of the crystal structure of these γ’ precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Liu, Xiang and He, Lingfeng and Yan, Huan and Bachhav, Mukesh and Stubbins, James F.}, year={2020}, month={Jan}, pages={151851} }
 @article{bachhav_he_kane_liu_gan_vurpiliot_2020, title={Atom Probe Tomography for Burnup and Fission Product Analysis for Nuclear Fuels}, volume={26}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/S1431927620023776}, DOI={10.1017/S1431927620023776}, abstractNote={The conventional and modern power reactors (such as PWR, BWR and AGR) use UO2 as a standard nuclear fuel since it provides necessary thermodynamic stability required in operating conditions [1-2]. However, the nuclear fuel during service is subjected to extreme condition of neutron irradiation, thermal, mechanical constraints which can influence its chemical, physico-chemical and microstructural properties of interest. Also, neutron irradiation leads to the formation of large number of fission products (FP) generated during burnup which can impact the chemical and mechanical properties of the fuel [3]. For instance, the release of gaseous fission products (xenon and krypton) can increase the pressure inside the fuel elements and contribute to the internal stresses on the cladding. Therefore, it is important to characterize the chemical and physical states of fission products in irradiated nuclear fuels which is often associated with amount of burnup and radial temperature gradient across fuel pin. The microstructure, composition, chemical and physical states of fission products, including metal and oxide precipitates and fission gas bubbles in both irradiated ceramic and metallic fuels have been characterized using Atom Probe Tomography (APT), Atomic-Resolution Scanning Transmission Electron Microscope (STEM) equipped with super-X energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS) systems at Idaho National Laboratory (INL). More often the irradiated microstructure of the fuel is heterogeneous, a reliable method to estimate the local burnup in nuclear fuels with sub-micron high spatial resolution becomes significant for evaluation of the local irradiated microstructure and its correlation to the general behavior of fuels [4]. Specimens}, number={S2}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Bachhav, Mukesh and He, Lingfeng and Kane, Joshua and Liu, Xiang and Gan, Jian and Vurpiliot, Francois}, year={2020}, month={Jul}, pages={3086–3088} }
 @article{khafizov_riyad_wang_pakarinen_he_yao_el-azab_hurley_2020, title={Combining mesoscale thermal transport and x-ray diffraction measurements to characterize early-stage evolution of irradiation-induced defects in ceramics}, volume={193}, url={https://doi.org/10.1016/j.actamat.2020.04.018}, DOI={10.1016/j.actamat.2020.04.018}, abstractNote={In situ characterization of defects, microstructure, and properties will provide new perspectives regarding the structure-property relationship of materials in extreme environments. In this communication, we investigate the utility of laser-based thermal transport measurements in combination with X-ray diffraction as a means to characterize the early-stage evolution of irradiation-induced defects in ceramics. Uranium dioxide is used as a model system to analyze the impact of irradiation-induced defects with 2.6 MeV H and 3.9 MeV He ions up to a dose of 0.1 displacement per atom (dpa) at low temperature. For these radiation regimes, the formation of extended defects such as loops and voids is limited as compared to point defects. Lattice expansion was determined from X-ray diffraction analysis. Modulated thermoreflectance was used to measure the thermal conductivity of the ion damaged region. Both H and He irradiation leads to an expansion of the crystal lattice and a reduction in thermal conductivity. For the same dpa, the lattice expansion and conductivity reduction were notably different for H and He irradiated samples. The results were analyzed using simple models for lattice expansion and thermal conductivity reduction, informed by atomistic simulation from the literature. The modeling results suggest that the difference in the defect kinetics between two conditions can be attributed to ionization induced enhanced defect mobility and the stability of Schottky defects. These results demonstrate the utility of thermal conductivity measurements as a tool for characterization of microstructure under irradiation.}, journal={Acta Materialia}, publisher={Elsevier BV}, author={Khafizov, Marat and Riyad, M Faisal and Wang, Yuzhou and Pakarinen, Janne and He, Lingfeng and Yao, Tiankai and El-Azab, Anter and Hurley, David}, year={2020}, month={Jul}, pages={61–70} }
 @article{wang_yao_xi_lei_guo_he_frankel_lian_2020, title={Degradation mechanism of lead-vanado-iodoapatite in NaCl solution}, volume={172}, ISSN={0010-938X}, url={http://dx.doi.org/10.1016/j.corsci.2020.108720}, DOI={10.1016/j.corsci.2020.108720}, abstractNote={Degradation mechanism of lead-vanado-iodoapatite (IAPT) in NaCl solution at 90 °C was investigated through systematic characterization of surface morphology, microstructure and microchemistry evolution of the alteration layer. Nano-scale characterization indicated that IAPT crystals degraded from grain boundaries toward inner grains, forming ultrafine Cl-bearing crystallites. A coupled interface dissolution-reprecipitation replacement mechanism and the ion exchange between Cl− and I− across the degradation zone may together play roles in determining the nano-scale degradation behavior of IAPT. This work highlights the degradation through multiple reactions and elemental transport across the liquid-solid and solid-solid interfaces at length scales from sub-millimeter to nanoscale.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Wang, Yachun and Yao, Tiankai and Xi, Fayuan and Lei, Penghui and Guo, Xiaolei and He, Lingfeng and Frankel, Gerald S. and Lian, Jie}, year={2020}, month={Aug}, pages={108720} }
 @article{cheniour_tonks_gong_yao_he_harp_beeler_zhang_lian_2020, title={Development of a grain growth model for U3Si2 using experimental data, phase field simulation and molecular dynamics}, volume={532}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2020.152069}, DOI={10.1016/j.jnucmat.2020.152069}, abstractNote={The purpose of this work is to develop a model for normal grain growth in U3Si2. The average grain boundary energy was determined from previously published molecular dynamics simulations. The grain growth kinetics were quantified at various temperatures by annealing nanocrystalline samples. The mobility was determined by comparing phase field grain growth simulations to the experimental data. From these various methods, we found that the average grain size D in U3Si2 can be estimated over time t using the equation D2−D02=2αMγt, where D0 is the initial average grain size, the geometry factor α=0.96, the average grain boundary mobility M=6.30×10−18e−0.33[eV]kbTm4/(Js) with the Boltzmann constant kb and temperature T, and the average grain boundary energy has been found as a function of temperature, e.g. γ¯=0.83 J/m2 at 673 K.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Cheniour, Amani and Tonks, Michael R. and Gong, Bowen and Yao, Tiankai and He, Lingfeng and Harp, Jason M. and Beeler, Benjamin and Zhang, Yongfeng and Lian, Jie}, year={2020}, month={Apr}, pages={152069} }
 @article{zhang_lin_he_murugesan_pawar_sivakumar_ding_ding_liaw_dufek_et al._2020, title={Dual Functional Ni<sub>3</sub>S<sub>2</sub>@Ni Core–Shell Nanoparticles Decorating Nanoporous Carbon as Cathode Scaffolds for Lithium–Sulfur Battery with Lean Electrolytes}, volume={3}, ISSN={2574-0962 2574-0962}, url={http://dx.doi.org/10.1021/acsaem.0c00568}, DOI={10.1021/acsaem.0c00568}, abstractNote={Lithium–sulfur batteries are very promising for next-generation energy storage. However, most studies use flooded electrolytes to achieve a high specific capacity at the expense of lowering the spe...}, number={5}, journal={ACS Applied Energy Materials}, publisher={American Chemical Society (ACS)}, author={Zhang, Yulun and Lin, Yuxiao and He, Lingfeng and Murugesan, Vijayakumar and Pawar, Gorakh and Sivakumar, Bhuvana M. and Ding, Hanping and Ding, Dong and Liaw, Boryann and Dufek, Eric J. and et al.}, year={2020}, month={May}, pages={4173–4179} }
 @article{yu_kim_bachhav_liu_he_couet_2020, title={Effect of proton pre-irradiation on corrosion of Zr-0.5Nb model alloys with different Nb distributions}, volume={173}, ISSN={0010-938X}, url={http://dx.doi.org/10.1016/j.corsci.2020.108790}, DOI={10.1016/j.corsci.2020.108790}, abstractNote={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.}, journal={Corrosion Science}, publisher={Elsevier BV}, author={Yu, Zefeng and Kim, Taeho and Bachhav, Mukesh and Liu, Xiang and He, Lingfeng and Couet, Adrien}, year={2020}, month={Aug}, pages={108790} }
 @article{cappia_miller_aguiar_he_murray_frickey_stanek_harp_2020, title={Electron microscopy characterization of fast reactor MOX Joint Oxyde-Gaine (JOG)}, volume={531}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2019.151964}, DOI={10.1016/j.jnucmat.2019.151964}, abstractNote={The composition and crystal structure of the “Joint Oxyde Gaine” (JOG) has been investigated by means of electron microscopy. Microstructural characterization reveals a highly heterogeneous porous structure with inclusions containing both fission products and cladding components. Major fission products detected, other than Cs and Mo, are Te, I, Zr and Ba. The layer is composed by sub-micrometric crystallites. The diffraction data refinement, together with chemical mapping, confirms the presence of Cs2MoO4, which is the major component of the JOG. However, combinatorial analyses reveal that other non-stoichiometric phases are possible, highlighting the complex nature of the crystalline structure of the JOG. Fe is found in metallic Pd-rich precipitates with structure compatible with the tetragonal structure of FePd alloy. Cr is found in different locations of the JOG, in oxide form, but no structural data could be obtained due to local beam sensitization of the sample in those areas.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Cappia, F. and Miller, B.D. and Aguiar, J.A. and He, L. and Murray, D.J. and Frickey, B.J. and Stanek, J.D. and Harp, J.M.}, year={2020}, month={Apr}, pages={151964} }
 @article{hoffman_arivu_wen_he_sridharan_wang_xiong_liu_he_wu_2020, title={Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels}, url={https://doi.org/10.1016/j.mtla.2020.100806}, DOI={10.1016/j.mtla.2020.100806}, abstractNote={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.}, journal={Materialia}, author={Hoffman, Andrew and Arivu, Maalavan and Wen, Haiming and He, Li and Sridharan, Kumar and Wang, Xin and Xiong, Wei and Liu, Xiang and He, Lingfeng and Wu, Yaqiao}, year={2020}, month={Sep} }
 @article{johns_he_kane_windes_ubic_karthik_2020, title={Experimental evidence for ‘buckle, ruck and tuck’ in neutron irradiated graphite}, volume={159}, url={https://doi.org/10.1016/j.carbon.2019.12.028}, DOI={10.1016/j.carbon.2019.12.028}, abstractNote={The current mainstream theory for radiation-induced dimensional change of nuclear graphite is based on the notion that interstitially displaced carbon atoms will coalesce into dislocation loops (i.e., additional basal planes). This standard atomic-displacement model has been challenged by theories based on first principles calculations. The so-called ‘ruck and tuck’ of basal planes has been proposed as an alternative mechanism to explain the observed c-axis expansion under irradiation; however, no such defects have been observed experimentally so far. In this study, the first experimental evidence for the presence of a ‘ruck and tuck’ defect in high-temperature neutron-irradiated nuclear graphite is presented.}, journal={Carbon}, publisher={Elsevier BV}, author={Johns, Steve and He, Lingfeng and Kane, Joshua J. and Windes, William E. and Ubic, Rick and Karthik, Chinnathambi}, year={2020}, month={Apr}, pages={119–121} }
 @article{johns_he_bustillo_windes_ubic_karthik_2020, title={Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite}, volume={166}, ISSN={0008-6223}, url={http://dx.doi.org/10.1016/j.carbon.2020.05.028}, DOI={10.1016/j.carbon.2020.05.028}, abstractNote={Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥800 °C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.}, journal={Carbon}, publisher={Elsevier BV}, author={Johns, Steve and He, Lingfeng and Bustillo, Karen and Windes, William E. and Ubic, Rick and Karthik, Chinnathambi}, year={2020}, month={Sep}, pages={113–122} }
 @article{gill_topsakal_jossou_huang_hattar_mausz_elbakhshwan_yan_chu_sun_et al._2020, title={Impact of krypton irradiation on a single crystal tungsten: Multi-modal X-ray imaging study}, volume={188}, ISSN={1359-6462}, url={http://dx.doi.org/10.1016/j.scriptamat.2020.07.024}, DOI={10.1016/j.scriptamat.2020.07.024}, abstractNote={Understanding microstructural and strain evolutions induced by noble gas production in the nuclear fuel matrix or plasma-facing materials is crucial for designing next generation nuclear reactors, as they are responsible for volumetric swelling and catastrophic failure. We describe a multimodal approach combining synchrotron-based nanoscale X-ray imaging techniques with atomic-scale electron microscopy techniques for mapping chemical composition, morphology and lattice distortion in a single crystal W induced by Kr irradiation. We report that Kr-irradiated single crystal W undergoes surface deformation, forming Kr containing cavities. Furthermore, positive strain fields are observed in Kr-irradiated regions, which lead to compression of underlying W matrix.}, journal={Scripta Materialia}, publisher={Elsevier BV}, author={Gill, Simerjeet K. and Topsakal, Mehmet and Jossou, Ericmoore and Huang, Xiaojing and Hattar, Khalid and Mausz, Julia and Elbakhshwan, Mohamed and Yan, Hanfei and Chu, Yong S. and Sun, Cheng and et al.}, year={2020}, month={Nov}, pages={296–301} }
 @article{parkin_moorehead_elbakhshwan_hu_chen_li_he_sridharan_couet_2020, title={In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation}, volume={198}, ISSN={1359-6454}, url={http://dx.doi.org/10.1016/j.actamat.2020.07.066}, DOI={10.1016/j.actamat.2020.07.066}, abstractNote={This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35, and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the CrFeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr18Fe27Mn27Ni28 and 10 nm in Cr15Fe35Mn15Ni35. This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.}, journal={Acta Materialia}, publisher={Elsevier BV}, author={Parkin, Calvin and Moorehead, Michael and Elbakhshwan, Mohamed and Hu, Jing and Chen, Wei-Ying and Li, Meimei and He, Lingfeng and Sridharan, Kumar and Couet, Adrien}, year={2020}, month={Oct}, pages={85–99} }
 @article{hua_yao_khanolkar_ding_gofryk_he_benson_hurley_2020, title={Intragranular thermal transport in U–50Zr}, volume={534}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2020.152145}, DOI={10.1016/j.jnucmat.2020.152145}, abstractNote={A thermoreflectance method was used to measure intragranular thermal diffusivity and conductivity of samples having a composition of U-50 wt%Zr (U–50Zr). Three phases at this composition were investigated: the high temperature γ phase, the low temperature δ phase, and the metastable ω phase. This approach uses a tightly focused laser to inject micron scale thermal waves and a second tightly focused laser to monitor the temperature distribution. The thermal properties are extracted by comparing experimental temperature profiles to an analytical heat diffusion model. The probe laser can monitor the temperature field in orthogonal directions along the surface of the sample and is well suited to measure thermal anisotropy. We show that the δ phase exhibits significant thermal anisotropy. The γ phase has the highest thermal conductivity. The higher conductivity of the γ phase is thought to be due to the presence of Zr precipitates that slightly change the stoichiometry of the γ matrix. The highly disordered ω phase appears to be thermally isotropic and has a lower conductivity than the δ phase. Both observations are likely due to the presence of γ domains that reside between ω domains. This supposition is supported by the presence of thermal heterogeneities that appear as noise in the measured signals.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Hua, Zilong and Yao, Tiankai and Khanolkar, Amey and Ding, Xiaxin and Gofryk, Krzysztof and He, Lingfeng and Benson, Michael and Hurley, David}, year={2020}, month={Jun}, pages={152145} }
 @article{yao_wagner_liu_ei-azab_harp_gan_hurley_benson_he_2020, title={On spinodal-like phase decomposition in U–50Zr alloy}, url={https://doi.org/10.1016/j.mtla.2020.100592}, DOI={10.1016/j.mtla.2020.100592}, abstractNote={Finely dispersed two phase microstructures resulting from a spinodal decomposition are of interest as they are associated with enhanced mechanical properties and excessive interfaces to mitigate defect related behavior. This study reports a spinodal-like phase decomposition in a U–50Zr alloy by thermal annealing at 620 °C and ion irradiation at 550 °C, with the latter temperature too low to initiate pure thermal phase transformation. The results hold broad impact for U–Zr alloy systems and its application as advanced nuclear fuel.}, journal={Materialia}, author={Yao, Tiankai and Wagner, Adrian R and Liu, Xiang and EI-Azab, Anter and Harp, Jason M and Gan, Jian and Hurley, David H and Benson, Michael T and He, Lingfeng}, year={2020}, month={Mar} }
 @article{ronne_he_dolzhnikov_xie_ge_halstenberg_wang_manard_xiao_lee_et al._2020, title={Revealing 3D Morphological and Chemical Evolution Mechanisms of Metals in Molten Salt by Multimodal Microscopy}, volume={12}, DOI={10.1021/acsami.9b19099}, abstractNote={Growing interest in molten salts as effective high-temperature heat-transfer fluids for sustainable energy systems drives a critical need to fundamentally understand the interactions between metals and molten salts. This work utilizes the multimodal microscopy methods of synchrotron X-ray nano-tomography and electron microscopy to investigate the 3D morphological and chemical evolution of two model systems, pure nickel metal and Ni-20Cr binary alloy, in a representative molten salt (KCl-MgCl2 50-50 mol. %, 800 °C). In both systems, unexpected shell-like structures formed due to the presence of more noble tungsten, suggesting a potential route of using Ni-W alloys for enhanced molten-salt corrosion resistance. The binary alloy Ni-20Cr developed a bicontinuous porous structure, reassembling functional porous metals manufactured by dealloying. This work elucidates better mechanistic understanding of corrosion in molten salts, which can contribute to the design of more reliable alloys for molten salt applications including next-generation nuclear and solar power plants and opens the possibility of using molten salts to fabricate functional porous materials.}, number={15}, journal={ACS Applied Materials & Interfaces}, publisher={American Chemical Society (ACS)}, author={Ronne, Arthur and He, Lingfeng and Dolzhnikov, Dmitriy and Xie, Yi and Ge, Mingyuan and Halstenberg, Phillip and Wang, Yachun and Manard, Benjamin T. and Xiao, Xianghui and Lee, Wah-Keat and et al.}, year={2020}, month={Mar}, pages={17321–17333} }
 @misc{couet_he_bachhav_yu_2020, title={TEM and APT Characterization of Neutron Irradiated AXIOM-2(X2)}, url={http://dx.doi.org/10.13182/t123-33175}, DOI={10.13182/t123-33175}, journal={Transactions of the American Nuclear Society - Volume 123}, publisher={AMNS}, author={Couet, A. and He, L. and Bachhav, M. and Yu, Z.}, year={2020} }
 @article{chen_he_cullison_hay_burns_wu_tan_2020, title={The correlation between microstructure and nanoindentation property of neutron-irradiated austenitic alloy D9}, volume={195}, url={https://doi.org/10.1016/j.actamat.2020.05.020}, DOI={10.1016/j.actamat.2020.05.020}, abstractNote={The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications.}, journal={Acta Materialia}, publisher={Elsevier BV}, author={Chen, Tianyi and He, Lingfeng and Cullison, Mack H. and Hay, Charles and Burns, Jatuporn and Wu, Yaqiao and Tan, Lizhen}, year={2020}, month={Aug}, pages={433–445} }
 @article{dennett_hua_khanolkar_yao_morgan_prusnick_poudel_french_gofryk_he_et al._2020, title={The influence of lattice defects, recombination, and clustering on thermal transport in single crystal thorium dioxide}, url={https://doi.org/10.1063/5.0025384}, DOI={10.1063/5.0025384}, abstractNote={Thermal transport is a key performance metric for thorium dioxide in many applications where defect-generating radiation fields are present. An understanding of the effect of nanoscale lattice defects on thermal transport in this material is currently unavailable due to the lack of a single crystal material from which unit processes may be investigated. In this work, a series of high-quality thorium dioxide single crystals are exposed to 2 MeV proton irradiation at room temperature and 600 °C to create microscale regions with varying densities and types of point and extended defects. Defected regions are investigated using spatial domain thermoreflectance to quantify the change in thermal conductivity as a function of ion fluence as well as transmission electron microscopy and Raman spectroscopy to interrogate the structure of the generated defects. Together, this combination of methods provides important initial insight into defect formation, recombination, and clustering in thorium dioxide and the effect of those defects on thermal transport. These methods also provide a promising pathway for the quantification of the smallest-scale defects that cannot be captured using traditional microscopy techniques and play an outsized role in degrading thermal performance.}, journal={APL Materials}, author={Dennett, Cody A. and Hua, Zilong and Khanolkar, Amey and Yao, Tiankai and Morgan, Phyllis K. and Prusnick, Timothy A. and Poudel, Narayan and French, Aaron and Gofryk, Krzysztof and He, Lingfeng and et al.}, year={2020}, month={Nov} }
 @article{yao_capriotti_harp_liu_wang_teng_murray_winston_gan_benson_et al._2020, title={α-U and ω-UZr2 in neutron irradiated U-10Zr annular metallic fuel}, volume={542}, DOI={10.1016/j.jnucmat.2020.152536}, abstractNote={To develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace the sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boosts the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissions per initial heavy metal atoms. The low burnup was planned for initial screening of this fuel design. After irradiation, the initial U-10Zr microstructure separated into an α-U annular region and an UZr2+x center region with a nanoscale spinodal decomposed microstructure. Because of the large amount of interface areas created in this microstructure, the fission gas atoms and vacancies generated in the UZr2+x phase are possibly pinned at the interface areas, leading to ~20 times smaller fission gas bubbles than those in the neighboring α-U. The large bubbles in α-U become connected and merged into large pores that provide fast paths for fission gas release into capsule plenum which prevents further swelling of fuel slug. The fuel slug center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yao, Tiankai and Capriotti, Luca and Harp, Jason M. and Liu, Xiang and Wang, Yachun and Teng, Fei and Murray, Daniel J. and Winston, Alex J. and Gan, Jian and Benson, Michael T. and et al.}, year={2020}, pages={152536} }
 @article{bao_kuang_sun_li_wan_shen_ma_he_2019, title={A simple way to make pre-stressed ceramics with high strength}, volume={5}, ISSN={2352-8478}, url={http://dx.doi.org/10.1016/j.jmat.2019.06.001}, DOI={10.1016/j.jmat.2019.06.001}, abstractNote={A pre-stressing design and a simple fabrication technology to substantially improve the strength of ceramic components are presented. Residual surface compressive stress is generated in ceramic components by pressureless sintering of a green bulk coated with a thin layer of low coefficient of thermal expansion (CTE). The stress level can be controlled by changing the cross-section area ratio, Young's modulus ratio and CTE ratio of the coating. Pre-stressed ZrO2 ceramics coated with Al2O3 can achieve a flexural strength of 1330 ± 52 MPa, 45% higher than their uncoated counterpart. Similarly, the flexural strength of building porcelain tiles is increased by 70%, from 67 ± 3 MPa to 114 ± 5 MPa. The damage tolerance of pre-stressed ZrO2 ceramics is excellent with a high residual strength of ∼1200 MPa in a thermal shock test at 325 °C. This simple technique can improve the mechanical performance of ceramic components with no limitation of size and shape.}, number={4}, journal={Journal of Materiomics}, publisher={Elsevier BV}, author={Bao, Yiwang and Kuang, Fenghua and Sun, Yi and Li, Yueming and Wan, Detian and Shen, Zongyang and Ma, Delong and He, Lingfeng}, year={2019}, month={Dec}, pages={657–662} }
 @article{benson_xie_he_king_2019, title={Characterization of U-Pu-Zr alloys with Pd as a minor additive}, volume={120}, journal={Transactions of the American Nuclear Society}, author={Benson, M.T. and Xie, Y. and He, L. and King, J.A.}, year={2019} }
 @article{xie_benson_he_king_mariani_murray_2019, title={Diffusion Behavior between Metallic Fuel Alloys with Pd Addition and Fe}, volume={525}, url={http://dx.doi.org/10.1016/j.jnucmat.2019.07.028}, DOI={10.1016/j.jnucmat.2019.07.028}, abstractNote={Fission product lanthanides in metallic fuels are known to cause adverse fuel-cladding chemical interaction (FCCI). Palladium (Pd) is being explored as a potential additive to reduce or mitigate FCCI by forming stable Pd-Ln compounds. The current study is an investigation of diffusion behaviors between UZrPd, UZrPdLn, UPuZrPd, and UPuZrPdLn alloys (Ln = 53Nd-25Ce-16Pr-6La wt. %) and iron (Fe). Diffusion couple tests were performed followed by microstructural analysis using scanning electron microscopy (SEM) and transmission electron microscopy (TEM). It is found that the diffusion of Fe into the fuel is reduced by PdZr2 precipitates randomly dispersed throughout the fuel matrix. Ln-Pd compounds form in UZrPdLn alloys, and (Ln,Pu)Pd forms in the UPuZrPdLn alloy. In both diffusion couples with Ln, small amounts of lanthanides are diffusing into the Fe. This could be due to small amounts of dissolved lanthanides in the precipitates, or could be due to the Ln-Pd compounds decomposing. In either case, diffusion caused by lanthanides is greatly reduced as compared to alloys without Pd.}, journal={Journal of Nuclear Materials}, author={Xie, Y. and Benson, M.T. and He, L. and King, J.A. and Mariani, R.D. and Murray, D.J.}, year={2019}, month={Nov}, pages={111–124} }
 @article{xie_benson_he_king_2019, title={Diffusion behaviors between Fe and Pd-containing metallic fuel}, volume={120}, number={1}, journal={Transactions of the American Nuclear Society}, author={Xie, Y. and Benson, M.T. and He, L. and King, J.A.}, year={2019}, month={Jun}, pages={412–413} }
 @article{khafizov_pakarinen_he_hurley_2019, title={Impact of irradiation induced dislocation loops on thermal conductivity in ceramics}, volume={102}, url={https://doi.org/10.1111/jace.16616}, DOI={10.1111/jace.16616}, abstractNote={Abstract}, number={12}, journal={Journal of the American Ceramic Society}, publisher={Wiley}, author={Khafizov, Marat and Pakarinen, Janne and He, Lingfeng and Hurley, David H.}, year={2019}, month={Dec}, pages={7533–7542} }
 @book{palmer_holbrook_jaoude_core_cao_marsden_he_frank_calderoni_2019, title={Irradiation Testing Strategy for Commercialization of a Molten Chloride Fast Reactor System}, author={Palmer, A.J. and Holbrook, M.R. and Jaoude, A.Abou and Core, G.M. and Cao, G. and Marsden, K.C. and He, L. and Frank, S.M. and Calderoni, P.}, year={2019} }
 @article{benson_xie_he_tolman_king_harp_mariani_hernandez_murray_miller_2019, title={Microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln}, volume={518}, url={https://doi.org/10.1016/j.jnucmat.2019.03.014}, DOI={10.1016/j.jnucmat.2019.03.014}, abstractNote={Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is the microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La, using scanning electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). In both alloys, the matrix is comprised of ζ-U0.4Pu0.6 and δ-(U,Pu)Zr2. Based on the matrix compositions, modifications to the room temperature extrapolated U-Pu-Zr ternary phase diagram are suggested. In U-20Pu-10Zr-3.86Pd, there is very little δ phase, due to formation of PdZr2. In U-20Pu-10Zr-3.86Pd-4.3Ln, a lanthanide-rich phase is present, although it does not have the crystal structure for 53Nd-25Ce-16Pr-6La. This phase cannot be identified based on known compounds. (Ln,Pu)Pd is the primary lanthanide phase, with Pu substituting into the crystal structure in place of Nd.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Benson, Michael T. and Xie, Yi and He, Lingfeng and Tolman, Kevin R. and King, James A. and Harp, Jason M. and Mariani, Robert D. and Hernandez, Brandon J. and Murray, Daniel J. and Miller, Brandon D.}, year={2019}, month={May}, pages={287–297} }
 @article{yu_zhang_voyles_he_liu_nygren_couet_2019, title={Microstructure and microchemistry study of irradiation-induced precipitates in proton irradiated ZrNb alloys}, url={https://doi.org/10.1016/j.actamat.2019.08.012}, DOI={10.1016/j.actamat.2019.08.012}, abstractNote={Proton irradiation induced Nb redistribution in Zr-xNb alloys (x = 0.4, 0.5, 1.0 wt%) has been investigated using scanning transmission electron microscopy/energy dispersive X-ray spectroscopy (STEM/EDS). Zr-xNb alloys are mainly composed of Zr matrix, native Zr–Nb–Fe phases, and β-Nb precipitates. After 2 MeV proton irradiation at 350 °C, a decrease of Nb content in native precipitates, as well as irradiation-induced precipitation of Nb-rich platelets (135 ± 69 nm long and 27 ± 12 nm wide) were found. Nb-rich platelets and Zr matrix form the Burgers orientation relationship, [11¯1]//[21¯1¯0] and (011)//(0002). The platelets were found to be mostly coherent with the matrix with a few dislocations near the ends of the precipitate. The coherent strain field has been measured in the matrix and platelets by the 4D-STEM technique. The growth of Nb-rich platelets is mainly driven by coherency and dislocation-induced strain fields. Irradiation may both enhance the diffusion and induce segregation of interstitial Nb to the ends of the irradiation induced platelets, further facilitating their growth.}, journal={Acta Materialia}, author={Yu, Zefeng and Zhang, Chenyu and Voyles, Paul M. and He, Lingfeng and Liu, Xiang and Nygren, Kelly and Couet, Adrien}, year={2019}, month={Oct} }
 @article{nguyen_nakayama_suematsu_iwasawa_suzuki_otsuka_he_takahashi_niihara_2019, title={Self-healing behavior and strength recovery of ytterbium disilicate ceramic reinforced with silicon carbide nanofillers}, volume={39}, ISSN={0955-2219}, url={http://dx.doi.org/10.1016/j.jeurceramsoc.2019.03.040}, DOI={10.1016/j.jeurceramsoc.2019.03.040}, abstractNote={Environmental barrier coatings (EBCs) are necessary to protect SiC/SiC ceramic components against oxidation and hot corrosion in high-temperature applications. The volatilization of SiO2 in SiC-reinforced materials is a major obstacle for the implementation of these self-crack-healing ceramics. The Yb2Si2O7-Yb2SiO5-SiC composite is known as a self-healing material that can help to avoid this SiC recession. In this research, the crack-healing behavior of this composite is investigated by using pre-cracking followed by annealing in an oxidizing environment. The crack-healing mechanism is explored and elucidated as a function of the filler morphology, crack size, annealing time, and annealing temperature. The two main crack-healing mechanisms are the filling of cracks with SiO2 glass and the volume expansion of Yb2Si2O7 induced by the reaction between SiO2 and Yb2SiO5. Full crack recovery is achieved with only 10 vol% SiC, with evidence from XRD and EDS analyses. SiC nanoparticulates are more efficient fillers than nanofibers and nanowhiskers.}, number={10}, journal={Journal of the European Ceramic Society}, publisher={Elsevier BV}, author={Nguyen, Son Thanh and Nakayama, Tadachika and Suematsu, Hisayuki and Iwasawa, Hirokazu and Suzuki, Tsuneo and Otsuka, Yuichi and He, Lingfeng and Takahashi, Tsuyoshi and Niihara, Koichi}, year={2019}, month={Aug}, pages={3139–3152} }
 @inproceedings{he_murray_liu_jiang_bachhav_bai_teysseyre_2019, title={Small Scale Tensile Testing of Grain Boundary Strength of X-750 Alloy}, booktitle={Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors}, author={He, L. and Murray, Daniel and Liu, Xiang and Jiang, Wen and Bachhav, Mukesh and Bai, Xianming and Teysseyre, Sebastien}, year={2019} }
 @article{bachhav_gan_he_miller_keiser_2018, title={Challenges and Opportunities on Elucidating Irradiated Fuels with Atom Probe Tomography}, volume={24}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/s1431927618011510}, DOI={10.1017/s1431927618011510}, abstractNote={Fuel materials used in nuclear reactors are subjected to extreme conditions during their operation and storage resulting in microstructural changes. It is therefore critical to understand microstructural evolution in nuclear materials and correlate with their performance. One of the main mechanisms in guiding the microstructural changes of fuels is damage associated with the formation and migration of fission products [1-2]. For instance, the formation and the movement of fission product leads to void formation and grain boundary segregation which can promote integrity loss of the fuel [3-4]. Also, during irradiation at aggressive reactor conditions, interaction can occur between the fuel particle and the matrix in a dispersion fuel that results in development of an interaction layer that is unstable under irradiation, which can contribute to fuel plate failure [3-4]. Thus, it is desirable to access the microstructural information in irradiated fuels in order to understand migration of fission product and their affect on performance of fuels.}, number={S1}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Bachhav, Mukesh and Gan, Jian and He, Lingfeng and Miller, Brandon and Keiser, Dennis}, year={2018}, month={Aug}, pages={2206–2207} }
 @article{hoggan_harp_he_2018, title={Evaluation of U3Si2 Fuel Pellets Sintered in an Argon vs. Vacuum Environment}, volume={10}, ISSN={2637-4390}, url={http://dx.doi.org/10.1002/9781119543299.ch3}, DOI={10.1002/9781119543299.ch3}, abstractNote={This chapter examines the microstructure, phase composition, and related density measurements of pellets sintered in each environment. A batch of pellets fabricated by the process reported were prepared in an experiment for sintering with three pellets randomly sampled for sintering in each environment, namely Argon and Vacuum. Three sample pellets were used in each environment as that was the standard sintering batch size according to the size of the tantalum crucible used for all sintering. The argon samples were sintered in a high temperature CM box furnace inside an argon atmosphere glove box with an approximate O2 impurity content of up to 40 ppm. An iron rich phase was identified with arrows indicating its location in the sample. The iron is thought to be a contaminant in the original uranium feedstock. It is often observed as a small precipitate connected to an oxide precipitate. This iron phase is likely present in both argon and vacuum samples, but was observed only in the vacuum sample.}, journal={Ceramic Transactions Series}, publisher={Wiley}, author={Hoggan, Rita and Harp, Jason and He, Lingfeng}, year={2018}, month={Oct}, pages={21–26} }
 @article{formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation_2018, DOI={https://doi.org/10.1016/j.scriptamat.2018.01.023}, abstractNote={We report the formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation at 573 K. The transmission electron microscopy study shows that the helium bubble lattice constant measured from the in-plane d-spacing is ~4.5 nm, while it is ~3.9 nm from the out-of-plane measurement. The results of synchrotron-based small-angle x-ray scattering agree well with the transmission electron microscopy results in terms of the measurement of bubble lattice constant and bubble size. The coupling of transmission electron microscopy and synchrotron high-energy X-ray scattering provides an effective approach to study defect superlattices in irradiated materials.}, journal={Scripta Materialia}, year={2018}, month={Jan} }
 @article{sun_sprouster_hattar_ecker_he_gao_zhang_gan_2018, title={Formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation}, volume={149}, ISSN={1359-6462}, url={http://dx.doi.org/10.1016/j.scriptamat.2018.01.023}, DOI={10.1016/j.scriptamat.2018.01.023}, abstractNote={We report the formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation at 573 K. The transmission electron microscopy study shows that the helium bubble lattice constant measured from the in-plane d-spacing is ~4.5 nm, while it is ~3.9 nm from the out-of-plane measurement. The results of synchrotron-based small-angle x-ray scattering agree well with the transmission electron microscopy results in terms of the measurement of bubble lattice constant and bubble size. The coupling of transmission electron microscopy and synchrotron high-energy X-ray scattering provides an effective approach to study defect superlattices in irradiated materials.}, journal={Scripta Materialia}, publisher={Elsevier BV}, author={Sun, C. and Sprouster, D.J. and Hattar, K. and Ecker, L.E. and He, L. and Gao, Y. and Zhang, Y. and Gan, J.}, year={2018}, month={May}, pages={26–30} }
 @article{he_harp_wagner_hoggan_tolman_2018, title={Hydrothermal synthesis of silicon oxide clad uranium oxide nanowires}, url={https://doi.org/10.1111/jace.15295}, DOI={10.1111/jace.15295}, abstractNote={Abstract}, journal={Journal of the American Ceramic Society}, author={He, Lingfeng and Harp, Jason M. and Wagner, Adrian R. and Hoggan, Rita E. and Tolman, Kevin R.}, year={2018}, month={Mar} }
 @article{yao_gong_he_miao_harp_tonks_lian_2018, title={In-situ TEM study of the ion irradiation behavior of U3Si2 and U3Si5}, volume={511}, url={https://doi.org/10.1016/j.jnucmat.2018.08.058}, DOI={10.1016/j.jnucmat.2018.08.058}, abstractNote={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.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yao, Tiankai and Gong, Bowen and He, Lingfeng and Miao, Yinbin and Harp, Jason M. and Tonks, Michael and Lian, Jie}, year={2018}, month={Dec}, pages={56–63} }
 @article{hoggan_he_harp_2018, title={Interdiffusion behavior of U3Si2 with FeCrAl via diffusion couple studies}, volume={502}, url={https://doi.org/10.1016/j.jnucmat.2017.10.057}, DOI={10.1016/j.jnucmat.2017.10.057}, abstractNote={Uranium silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as light water reactor (LWR) fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. A class of Fe, Cr, Al alloys collectively known as FeCrAl alloys that have superior mechanical and oxidation resistance are being considered as an alternative to the standard Zirconium based LWR cladding. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with U3Si2 pellets were placed in diffusion couples. Individual tests were ran at temperatures ranging from 500 °C to 1000 °C for 30 h and 100 h. The interdiffusion was analyzed with an optical microscope, scanning electron microscope, and transmission electron microscope. Uniform and planar interdiffusion layers along the material interface were illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy. Electron diffraction was used to validate phases present in the system, including distinct U2Fe3Si/UFe2 and UFeSi layers at the material interface. U and Fe diffused far into the FeCrAl and U3Si2 matrix, respectively, in the higher temperature tests. No interaction was observed at 500 °C for 30 h.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Hoggan, Rita E. and He, Lingfeng and Harp, Jason M.}, year={2018}, month={Apr}, pages={356–369} }
 @article{benson_he_king_mariani_2018, title={Investigating Pd as a Potential Fuel Additive in a Transmutation Fuel to Control FCCI}, volume={118}, journal={Transactions of the American Nuclear Society}, author={Benson, M.T. and He, L. and King, J.A. and Mariani, R.D.}, year={2018}, pages={1379–1380} }
 @article{benson_he_king_mariani_2018, title={Microstructural characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln: Investigating Pd as a metallic fuel additive}, url={https://doi.org/10.1016/j.jnucmat.2018.02.012}, DOI={10.1016/j.jnucmat.2018.02.012}, abstractNote={Palladium is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and, given enough time or burn-up, eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln, where Ln = 53Nd-25Ce-16Pr-6La. The present study shows that Pd preferentially binds the lanthanides over other fuel constituents, which may prevent lanthanide migration and interaction with the cladding during irradiation. The SEM analysis indicates the 1:1 Pd-Ln compound is being formed, while the TEM analysis, due to higher resolution, found the 1:1 compound, as well as Pd-rich compounds Pd2Ln and Pd3Ln2.}, journal={Journal of Nuclear Materials}, author={Benson, Michael T. and He, Lingfeng and King, James A. and Mariani, Robert D.}, year={2018}, month={Apr} }
 @article{benson_he_king_mariani_winston_madden_2018, title={Microstructural characterization of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln}, volume={508}, url={https://doi.org/10.1016/j.jnucmat.2018.05.062}, DOI={10.1016/j.jnucmat.2018.05.062}, abstractNote={Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is a scanning electron microscopy (SEM) and transmission electron microscopy (TEM) investigation of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La. In U-20Pu-10Zr-3.86Pd, PdZr2 is forming, along with a possible ternary phase between Pu, Zr, and Pd. Pu is also present in the Pd-Ln precipitates formed in U-20Pu-10Zr-3.86Pd-4.3Ln. In the LnPd phase, Pu appears to be substitutional, forming (Ln,Pu)Pd. The other prominent phase, which appears to be Ln7Pd3, has a fine, lamellar structure. The lanthanides remain essentially constant across this fine structure, but Pu and Pd alternate as to which has the higher concentration.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Benson, Michael T. and He, Lingfeng and King, James A. and Mariani, Robert D. and Winston, Alexander J. and Madden, James W.}, year={2018}, month={Sep}, pages={310–318} }
 @article{yao_gong_he_harp_tonks_lian_2018, title={Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature}, volume={498}, url={https://doi.org/10.1016/j.jnucmat.2017.10.027}, DOI={10.1016/j.jnucmat.2017.10.027}, abstractNote={U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Yao, Tiankai and Gong, Bowen and He, Lingfeng and Harp, Jason and Tonks, Michael and Lian, Jie}, year={2018}, month={Jan}, pages={169–175} }
 @article{he_bachhav_keiser_madden_perez_miller_gan_renterghem_leenaers_berghe_2018, title={STEM-EDS/EELS and APT characterization of ZrN coatings on UMo fuel kernels}, volume={511}, url={https://doi.org/10.1016/j.jnucmat.2018.09.004}, DOI={10.1016/j.jnucmat.2018.09.004}, abstractNote={In the framework of the SELENIUM project, ZrN coated U-Mo fuel kernels were irradiated in the Belgium Reactor 2 of SCK•CEN to a plate average burnup of 48% 235U and a local maximum burnup just below 70% 235U. The microstructure and chemical composition of ZrN coating before and after neutron irradiation have been analysed using scanning transmission electron microscopy (STEM) equipped with energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS), and atom probe tomography (APT). The atomic ratio of N/Zr and fission product distribution determined by three techniques were compared and the combination of three techniques shows the advantages in characterization of chemical information for nuclear materials.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={He, L. and Bachhav, M. and Keiser, D.D. and Madden, J.W. and Perez, E. and Miller, B.D. and Gan, J. and Renterghem, W. Van and Leenaers, A. and Berghe, S. Van}, year={2018}, month={Dec}, pages={174–182} }
 @article{gan_sun_he_zhang_jiang_gao_2018, title={Thermal stability of helium bubble superlattice in Mo under TEM in-situ heating}, volume={505}, url={https://doi.org/10.1016/j.jnucmat.2018.04.030}, DOI={10.1016/j.jnucmat.2018.04.030}, abstractNote={Although the temperature window of helium ion irradiation for gas bubble superlattice (GBS) formation was found to be in the range of approximately 0.15–0.35 melting point in literature, the thermal stability of He GBS has not been fully investigated. This work reports the experiment using an in-situ heating holder in a transmission electron microscope (TEM). A 3.0 mm TEM disc sample of Mo (99.95% pure) was irradiated with 40 keV He ions at 300 °C to a fluence of 1.0E+17 ions/cm2, corresponding to a peak He concentration of approximately 10 at.%, in order to introduce He GBS. In-situ heating was conducted with a ramp rate of ∼25 °C/min, hold time of ∼30 min, and temperature step of ∼100 °C up to 850 °C (0.39Tm homologous temperature). The result shows good thermal stability of He GBS in Mo with no noticeable change on GBS lattice constant and ordering. The implication of this unique and stable ordered microstructure on mechanistic understanding of GBS and its advanced application are discussed.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Gan, Jian and Sun, Cheng and He, Lingfeng and Zhang, Yongfeng and Jiang, Chao and Gao, Yipeng}, year={2018}, month={Jul}, pages={207–211} }
 @article{he_bai_pakarinen_jaques_gan_nelson_el-azab_allen_2017, title={Bubble evolution in Kr-irradiated UO2 during annealing}, volume={496}, url={https://doi.org/10.1016/j.jnucmat.2017.09.036}, DOI={10.1016/j.jnucmat.2017.09.036}, abstractNote={Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO2 annealed at high temperature was conducted in order to understand the inert gas behavior in oxide nuclear fuel. The average diameter of intragranular bubbles increased gradually from 0.8 nm in as-irradiated sample at room temperature to 2.6 nm at 1600 °C and the bubble size distribution changed from a uniform distribution to a bimodal distribution above 1300 °C. The size of intergranular bubbles increased more rapidly than intragranular ones and bubble denuded zones near grain boundaries formed in all the annealed samples. It was found that high-angle grain boundaries held bigger bubbles than low-angle grain boundaries. Complementary atomistic modeling was conducted to interpret the effects of grain boundary character on the Kr segregation. The area density of strong segregation sites in the high-angle grain boundaries is much higher than that in the low angle grain boundaries.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={He, L. and Bai, X.M. and Pakarinen, J. and Jaques, B.J. and Gan, J. and Nelson, A.T. and El-Azab, A. and Allen, T.R.}, year={2017}, month={Dec}, pages={242–250} }
 @article{effects of neutron irradiation of ti3sic2 and ti3alc2 in the 121–1085°c temperature range_2017, url={http://www.sciencedirect.com/science/article/pii/S0022311516311370}, DOI={dx.doi.org/10.1016/j.jnucmat.2016.11.016}, abstractNote={Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti3SiC2 and Ti3AlC2 samples exposed to total fluences of ≈6 × 1020 n/m2, 5 × 1021 n/m2 and 1.7 × 1022 n/m2 at irradiation temperatures of 121(12), 735(6) and 1085(68)°C. These fluences correspond to 0.14, 1.6 and 3.4 dpa, respectively. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed via transmission electron microscopy in both Ti3SiC2 and Ti3AlC2. After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti3SiC2, with loop diameters of 21(6) and 30(8) nm after 1.6 dpa and 3.4 dpa, respectively. In Ti3AlC2, larger dislocation loops, 75(34) nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Cavities were observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive cavity formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 μm are observed in Ti3SiC2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3–5 μm in diameter, are damage free after irradiation at 1085 °C at this dose. The results shown herein confirm once again that the presence of the A-layers in the MAX phases considerably enhance their irradiation tolerance. Based on these results, and up to 3.4 dpa, Ti3SiC2 remains a promising candidate for high temperature nuclear applications as long as the temperature remains >700 °C.}, journal={Journal of Nuclear Materials}, year={2017} }
 @inbook{hoggan_he_harp_2017, series={The Minerals, Metals & Materials Series}, title={Interdiffusion Behavior of FeCrAl with U3Si2}, ISBN={9783030046385 9783030046392}, ISSN={2367-1181 2367-1696}, url={http://dx.doi.org/10.1007/978-3-030-04639-2_92}, DOI={10.1007/978-3-030-04639-2_92}, abstractNote={Advanced steels, including FeCrAl are being considered as an alternative to the standard light water fuel (LWR) cladding, Zircalloy. FeCrAl has superior mechanical and thermal properties and oxidation resistance relative to the Zircalloy standard . Uranium Silicide (U3Si2) is a candidate to replace uranium oxide (UO2) as LWR fuel because of its higher thermal conductivity and higher fissile density relative to the current standard, UO2. The interdiffusion behavior between FeCrAl and U3Si2 is investigated in this study. Commercially available FeCrAl, along with pellets fabricated at the Idaho National Laboratory were placed in diffusion couples. Individual tests have been run at temperatures ranging from 500 ℃ to 1000 ℃ for 30 h and 100 h. The interdiffusion is analyzed with an optical microscope and scanning electron microscope (SEM). Uniform and planar diffusion regions along the material interface are illustrated with backscatter electron micrographs and energy-dispersive X-ray spectroscopy (EDS).}, booktitle={Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors}, publisher={Springer International Publishing}, author={Hoggan, Rita E. and He, Lingfeng and Harp, Jason M.}, editor={Jackson, J. and Paraventi, D. and Wright, M.Editors}, year={2017}, month={Oct}, pages={1391–1400}, collection={The Minerals, Metals & Materials Series} }
 @article{hoggan_harp_he_2017, title={Interdiffusion Behavior of U3Si2 and FeCrAl via Diffusion Couple Studies}, volume={116}, journal={Transactions of the American Nuclear Society}, author={Hoggan, R. and Harp, J. and He, L.}, year={2017}, pages={403–406} }
 @misc{aitkaliyeva_he_wen_miller_bai_allen_2017, title={Irradiation effects in Generation IV nuclear reactor materials}, url={http://dx.doi.org/10.1016/b978-0-08-100906-2.00007-0}, DOI={10.1016/b978-0-08-100906-2.00007-0}, abstractNote={Generation IV reactor structural materials will be exposed to high doses and temperatures during reactor operation that may lead to irradiation-induced degradation. This degradation will differ from that seen in light water reactors and therefore understanding mechanisms controlling material performance during irradiation is critical for evaluating the viability of Generation IV nuclear reactor concepts. This chapter discusses irradiation effects and microstructural changes that affect mechanical properties and dimensional stability of Generation IV reactor materials.}, journal={Structural Materials for Generation IV Nuclear Reactors}, publisher={Elsevier}, author={Aitkaliyeva, A. and He, L. and Wen, H. and Miller, B. and Bai, X.M. and Allen, T.}, year={2017}, pages={253–283} }
 @article{he_harp_hoggan_wagner_2017, title={Microstructure studies of interdiffusion behavior of U3Si2/Zircaloy-4 at 800 and 1000 °C}, volume={486}, url={https://doi.org/10.1016/j.jnucmat.2017.01.035}, DOI={10.1016/j.jnucmat.2017.01.035}, abstractNote={Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions when in contact with its cladding. As new fuel types are developed, it is crucial to understand the interaction behavior between fuel and its cladding. Diffusion experiments between U3Si2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000 °C up to 100 h. The microstructure of pristine U3Si2 and U3Si2/Zry-4 interdiffusion products were examined using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) equipped with an energy dispersive X-ray spectroscopy (EDS) system. The primary interdiffusion product observed at 800 °C is ZrSi2, with secondary phases of U-Zr in the Zry-4, and Fe-Cr-W-Zr-Si phases at Zry-4/ZrSi2 interface and Fe-Cr-U-Si phases at ZrSi2/U-Si interface. The primary interdiffusion products at 1000 °C were Zr2Si, U-Zr-Fe-Ni, U, U-Zr, and a low melting point phase U6Fe.}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={He, Lingfeng and Harp, Jason M. and Hoggan, Rita E. and Wagner, Adrian R.}, year={2017}, month={Apr}, pages={274–282} }
 @article{nguyen_nakayama_suematsu_suzuki_he_cho_niihara_2017, title={Strength improvement and purification of Yb<sub>2
Si<sub>2
O<sub>7
-SiC nanocomposites by surface oxidation treatment}, volume={100}, ISSN={0002-7820}, url={http://dx.doi.org/10.1111/jace.14831}, DOI={10.1111/jace.14831}, abstractNote={Abstract}, number={7}, journal={Journal of the American Ceramic Society}, publisher={Wiley}, author={Nguyen, Son T. and Nakayama, Tadachika and Suematsu, Hisayuki and Suzuki, Tsuneo and He, Lingfeng and Cho, Hong-Baek and Niihara, Koichi}, year={2017}, month={Apr}, pages={3122–3131} }
 @article{zheng_he_carpenter_sridharan_2016, title={Corrosion-induced microstructural developments in 316 stainless steel during exposure to molten Li2BeF4 (FLiBe) salt}, volume={482}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84992143288&partnerID=MN8TOARS}, DOI={10.1016/j.jnucmat.2016.10.023}, abstractNote={The microstructural developments in the near-surface regions of AISI 316 stainless steel during exposure to molten Li2BeF4 (FLiBe) salt have been investigated with the goal of using this material for the construction of the fluoride salt-cooled high-temperature reactor (FHR), a leading nuclear reactor concept for the next generation nuclear plants (NGNP). Tests were conducted in molten FLiBe salt (melting point: 459 °C) at 700 °C in graphite crucibles and 316 stainless steel crucibles for exposure duration of up to 3000 h. Corrosion-induced microstructural changes in the near-surface regions of the samples were characterized using scanning electron microscopy (SEM) in conjunction with energy dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD), and scanning transmission electron microscopy (STEM) with EDS capabilities. Intergranular corrosion attack in the near-surface regions was observed with associated Cr depletion along the grain boundaries. High-angle grain boundaries (15–180°) were particularly prone to intergranular attack and Cr depletion. The depth of attack extended to the depths of 22 μm after 3000-h exposure for the samples tested in graphite crucible, while similar exposure in 316 stainless steel crucible led to the attack depths of only about 11 μm. Testing in graphite crucibles led to the formation of nanometer-scale Mo2C, Cr7C3 and Al4C3 particle phases in the near-surface regions of the material. The copious depletion of Cr in the near-surface regions induced a γ-martensite to α-ferrite phase (FeNix) transformation. Based on the microstructural analysis, a thermal diffusion controlled corrosion model was developed and experimentally validated for predicting long-term corrosion attack depth.}, journal={Journal of Nuclear Materials}, author={Zheng, G. and He, L. and Carpenter, D. and Sridharan, K.}, year={2016}, pages={147–155} }
 @article{tallman_he_garcia-diaz_hoffman_kohse_sindelar_barsoum_2016, title={Effect of neutron irradiation on defect evolution in Ti3SiC2 and Ti2AlC}, volume={468}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84948445492&partnerID=MN8TOARS}, DOI={10.1016/j.jnucmat.2015.10.030}, abstractNote={Herein we report on the characterization of defects formed in polycrystalline Ti3SiC2 and Ti2AlC samples exposed to neutron irradiation – up to 0.1 displacements per atom (dpa) at 350 ± 40 °C or 695 ± 25 °C, and up to 0.4 dpa at 350 ± 40 °C. Black spots are observed in both Ti3SiC2 and Ti2AlC after irradiation to both 0.1 and 0.4 dpa at 350 °C. After irradiation to 0.1 dpa at 695 °C, small basal dislocation loops, with a Burgers vector of b = 1/2 [0001] are observed in both materials. At 9 ± 3 and 10 ± 5 nm, the loop diameters in the Ti3SiC2 and Ti2AlC samples, respectively, were comparable. At 1 × 1023 loops/m3, the dislocation loop density in Ti2AlC was ≈1.5 orders of magnitude greater than in Ti3SiC2, at 3 × 1021 loops/m3. After irradiation at 350 °C, extensive microcracking was observed in Ti2AlC, but not in Ti3SiC2. The room temperature electrical resistivities increased as a function of neutron dose for all samples tested, and appear to saturate in the case of Ti3SiC2. The MAX phases are unequivocally more neutron radiation tolerant than the impurity phases TiC and Al2O3. Based on these results, Ti3SiC2 appears to be a more promising MAX phase candidate for high temperature nuclear applications than Ti2AlC.}, journal={Journal of Nuclear Materials}, author={Tallman, D.J. and He, L. and Garcia-Diaz, B.L. and Hoffman, E.N. and Kohse, G. and Sindelar, R.L. and Barsoum, M.W.}, year={2016}, pages={194–206} }
 @inbook{irradiation effects in generation iv nuclear reactor materials_2016, url={http://www.sciencedirect.com/science/article/pii/B9780081009062000070}, DOI={http://dx.doi.org/10.1016/B978-0-08-100906-2.00007-0}, abstractNote={Generation IV reactor structural materials will be exposed to high doses and temperatures during reactor operation that may lead to irradiation-induced degradation. This degradation will differ from that seen in light water reactors and therefore understanding mechanisms controlling material performance during irradiation is critical for evaluating the viability of Generation IV nuclear reactor concepts. This chapter discusses irradiation effects and microstructural changes that affect mechanical properties and dimensional stability of Generation IV reactor materials.}, year={2016}, month={Sep} }
 @article{barsoum_tallman_he_gan_2016, title={MAX Phases for the Nuclear Industry: Possibilities and Pitfalls}, volume={114}, journal={Transactions of the American Nuclear Society}, author={Barsoum, M.W. and Tallman, D.J. and He, L. and Gan, J.}, year={2016}, pages={1242–1243} }
 @article{he_gan_allen_2016, title={Microstructure Evolution in Ion-Irradiated UO2}, volume={114}, journal={Transactions of the American Nuclear Society}, author={He, L. and Gan, J. and Allen, T.R.}, year={2016}, pages={1130–1131} }
 @article{khafizov_pakarinen_he_henderson_manuel_nelson_jaques_butt_hurley_2016, title={Subsurface imaging of grain microstructure using picosecond ultrasonics}, volume={112}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84964328395&partnerID=MN8TOARS}, DOI={10.1016/j.actamat.2016.04.003}, abstractNote={We report on imaging subsurface grain microstructure using picosecond ultrasonics. This approach relies on elastic anisotropy of crystalline materials where ultrasonic velocity depends on propagation direction relative to the crystal axes. Picosecond duration ultrasonic pulses are generated and detected using ultrashort light pulses. In materials that are transparent or semitransparent to the probe wavelength, the probe monitors gigahertz frequency Brillouin oscillations. The frequency of these oscillations is related to the ultrasonic velocity and the optical index of refraction. Ultrasonic waves propagating across a grain boundary experience a change in velocity due to a change in crystallographic orientation relative to the ultrasonic propagation direction. This change in velocity is manifested as a change in the Brillouin oscillation frequency. Using the ultrasonic propagation velocity, the depth of the interface can be determined from the location in time of the transition in oscillation frequency. A subsurface image of the grain boundary is obtained by scanning the beam along the surface. We demonstrate this subsurface imaging capability using a polycrystalline UO2 sample. Cross section liftout analysis of the grain boundary using electron microscopy was used to verify our imaging results.}, journal={Acta Materialia}, author={Khafizov, M. and Pakarinen, J. and He, L. and Henderson, H.B. and Manuel, M.V. and Nelson, A.T. and Jaques, B.J. and Butt, D.P. and Hurley, D.H.}, year={2016}, pages={209–215} }
 @article{pakarinen_he_hassan_wang_gupta_el-azab_allen_2015, title={Annealing-induced lattice recovery in room-temperature xenon irradiated CeO2: X-ray diffraction and electron energy loss spectroscopy experiments}, volume={30}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84929707243&partnerID=MN8TOARS}, DOI={10.1557/jmr.2015.13}, abstractNote={Abstract}, number={9}, journal={Journal of Materials Research}, author={Pakarinen, J. and He, L. and Hassan, A.-R. and Wang, Y. and Gupta, M. and El-Azab, A. and Allen, T.R.}, year={2015}, pages={1555–1562} }
 @article{he_valderrama_hassan_yu_gupta_pakarinen_henderson_gan_kirk_nelson_et al._2015, title={Bubble formation and Kr distribution in Kr-irradiated UO2}, volume={456}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84908031776&partnerID=MN8TOARS}, DOI={10.1016/j.jnucmat.2014.09.026}, abstractNote={In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO2 matrix and high release of Kr from sample surface under irradiation.}, journal={Journal of Nuclear Materials}, author={He, L.F. and Valderrama, B. and Hassan, A.-R. and Yu, J. and Gupta, M. and Pakarinen, J. and Henderson, H.B. and Gan, J. and Kirk, M.A. and Nelson, A.T. and et al.}, year={2015}, pages={125–132} }
 @article{zheng_kelleher_he_cao_anderson_allen_sridharan_2015, title={High-Temperature corrosion of UNS N10003 in Molten Li2BeF4 (FLiBe) Salt}, volume={71}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84942913529&partnerID=MN8TOARS}, DOI={10.5006/1657}, abstractNote={Corrosion testing of UNS N10003 in molten fluoride salt was performed in purified molten 27LiF-BeF2 (66–34 mol%) (FLiBe) salt at 700°C for 1,000 h, in pure nickel and graphite capsules. In the nickel capsule tests, the near-surface region of the alloy exhibited an approximately 200 nm porous structure, an approximately 3.5 μm chromium-depleted region, and MoSi2 precipitates. In the tests performed in graphite capsules, the alloy samples gained weight because of the formation of a variety of Cr3C2, Cr7C3, Mo2C, and Cr23C6 carbide phases on the surface and in the subsurface regions of the alloy. A Cr-depleted region was observed in the near-surface region where Mo thermally diffused toward either the surface or the grain boundary, which induced an approximately 1.4 μm Ni3Fe alloy layer in this region. The carbide-containing layer extended to approximately 7 μm underneath the Ni3Fe layer. The presence of graphite dramatically changes the mechanisms of corrosion attack in UNS N10003 in molten FLiBe salt. In t...}, number={10}, journal={Corrosion}, author={Zheng, G. and Kelleher, B. and He, L. and Cao, G. and Anderson, M. and Allen, T. and Sridharan, K.}, year={2015}, pages={1257–1266} }
 @article{inert gas measurement of single bubble in ceo2_2015, url={https://www.cambridge.org/core/journals/microscopy-and-microanalysis/article/div-classtitleinert-gas-measurement-of-single-bubble-in-ceospan-classsub2spandiv/0099BC0BDFB9315961DDF4E94E99FC1F}, DOI={doi:10.1017/S1431927615004559}, abstractNote={Journal Article Inert Gas Measurement of Single Bubble in CeO2 Get access Lingfeng He, Lingfeng He Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Janne Pakarinen, Janne Pakarinen University of Wisconsin, Madison, WI, USA Search for other works by this author on: Oxford Academic Google Scholar Xianming Bai, Xianming Bai Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Jian Gan, Jian Gan Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Yongqiang Wang, Yongqiang Wang Los Alamos National Laboratory, Los Alamos, NM, USA Search for other works by this author on: Oxford Academic Google Scholar Anter El-Azab, Anter El-Azab Purdue University, West Lafayette, IN, USA Search for other works by this author on: Oxford Academic Google Scholar Todd R Allen Todd R Allen Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 21, Issue S3, 1 August 2015, Pages 751–752, https://doi.org/10.1017/S1431927615004559 Published: 23 September 2015}, journal={Microsc. Microanal.}, year={2015}, month={Aug} }
 @article{he_pakarinen_bai_gan_wang_el-azab_allen_2015, title={Inert Gas Measurement of Single Bubble in CeO2}, volume={21}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/s1431927615004559}, DOI={10.1017/s1431927615004559}, abstractNote={Uranium dioxide (UO2), an oxide with a fluorite crystal structure, is the main fuel used in commercial light water reactors. Inert fission gases such as Xe and Kr significantly impact the performance of UO2 during reactor operation and in storage. These gases have a large yield of approximately 25% and have a low solubility in UO2, resulting in the formation of large density of fission gas bubbles [1]. Such bubbles cause the fuel to swell, which promotes clad outward creep that shortens the cladding lifetime. Characterization of the inert fission gas content in bubbles can help us understand fuel swelling and fuel pin pressurization. This is performed for Xe bubbles in cerium dioxide, CeO2, which is considered a surrogate for UO2 due to similar crystal structure and properties.}, number={S3}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={He, Lingfeng and Pakarinen, Janne and Bai, Xianming and Gan, Jian and Wang, Yongqiang and El-Azab, Anter and Allen, Todd R.}, year={2015}, month={Aug}, pages={751–752} }
 @article{pakarinen_he_gupta_gan_nelson_el-azab_allen_2014, title={2.6 MeV proton irradiation effects on the surface integrity of depleted UO2}, volume={319}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84890011923&partnerID=MN8TOARS}, DOI={10.1016/j.nimb.2013.11.014}, abstractNote={The effect of low temperature proton irradiation in depleted uranium dioxide was examined as a function of fluence. With 2.6 MeV protons, the fluence limit for preserving a good surface quality was found to be relatively low, about 1.4 and 7.0 × 1017 protons/cm2 for single and poly crystalline samples, respectively. Upon increasing the fluence above this threshold, severe surface flaking and disintegration of samples was observed. Based on scanning electron microscopy (SEM) and X-ray diffraction (XRD) observations the causes of surface failure were associated to high H atomic percent at the peak damage region due to low solubility of H in UO2. The resulting lattice stress is believed to exceed the fracture stress of the crystal at the observed fluencies. The oxygen point defects from the displacement damage may hinder the H diffusion and further increase the lattice stress, especially at the peak damage region.}, journal={Nuclear Instruments and Methods in Physics Research, Section B: Beam Interactions with Materials and Atoms}, author={Pakarinen, J. and He, L. and Gupta, M. and Gan, J. and Nelson, A. and El-Azab, A. and Allen, T.R.}, year={2014}, pages={100–106} }
 @article{he_roman_leng_sridharan_anderson_allen_2014, title={Corrosion behavior of an alumina forming austenitic steel exposed to supercritical carbon dioxide}, volume={82}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84894736865&partnerID=MN8TOARS}, DOI={10.1016/j.corsci.2013.12.023}, abstractNote={Microstructure characterization of corrosion behavior of an alumina forming austenitic (AFA) steel exposed to supercritical carbon dioxide was conducted at 450–650 °C and 20 MPa. At low temperature and short exposure times, the oxidation kinetics were parabolic and the oxide scales were mainly composed of protective and continuous Al2O3 and (Cr, Mn)-rich oxide layers. As the temperature and exposure time increased, the AFA steel gradually suffered breakaway oxidation and its oxide scales showed a multilayer structure mainly composed of Fe3O4, (Cr, Fe)3O4, NiFe/FeCr2O4/Cr2O3/Al2O3, FeCr2O4/Al2O3, and NiFe/Cr2O3/Al2O3, in sequence. The corrosion mechanism based on the microstructure evolution is discussed in detail.}, journal={Corrosion Science}, author={He, L.-F. and Roman, P. and Leng, B. and Sridharan, K. and Anderson, M. and Allen, T.R.}, year={2014}, pages={67–76} }
 @book{nguyen_nakayama_rozali_suematsu_suzuki_jiang_amarume_he_niihara_2014, title={Developing yttria-based ceramics having high liquid metal corrosion resistance}, volume={250}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84908214876&partnerID=MN8TOARS}, journal={Ceramic Transactions}, author={Nguyen, S.T. and Nakayama, T. and Rozali, S.B. and Suematsu, H. and Suzuki, T. and Jiang, W. and Amarume, S. and He, L. and Niihara, K.}, year={2014}, pages={53–63} }
 @article{valderrama_he_henderson_pakarinen_jaques_gan_butt_allen_manuel_2014, title={Effect of Grain Boundaries on Krypton Segregation Behavior in Irradiated Uranium Dioxide}, volume={66}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84911981510&partnerID=MN8TOARS}, DOI={10.1007/s11837-014-1182-x}, number={12}, journal={JOM}, author={Valderrama, B. and He, L. and Henderson, H.B. and Pakarinen, J. and Jaques, B. and Gan, J. and Butt, D.P. and Allen, T.R. and Manuel, M.V.}, year={2014}, pages={2562–2568} }
 @article{he_gupta_kirk_pakarinen_gan_allen_2014, title={In Situ TEM Observation of Dislocation Evolution in Polycrystalline UO2}, volume={66}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84911965785&partnerID=MN8TOARS}, DOI={10.1007/s11837-014-1186-6}, number={12}, journal={JOM}, author={He, L.F. and Gupta, M. and Kirk, M.A. and Pakarinen, J. and Gan, J. and Allen, T.R.}, year={2014}, pages={2553–2561} }
 @article{pakarinen_khafizov_he_wetteland_gan_nelson_hurley_el-azab_allen_2014, title={Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation}, volume={454}, ISSN={0022-3115}, url={http://dx.doi.org/10.1016/j.jnucmat.2014.07.053}, DOI={10.1016/j.jnucmat.2014.07.053}, abstractNote={The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in X-ray diffraction after ion irradiation up to 5 × 1016 He2+/cm2 at low-temperature (<200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 μm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 μm) in the sample subjected to 5 × 1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9 × 1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55% for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.}, number={1-3}, journal={Journal of Nuclear Materials}, publisher={Elsevier BV}, author={Pakarinen, Janne and Khafizov, Marat and He, Lingfeng and Wetteland, Chris and Gan, Jian and Nelson, Andrew T. and Hurley, David H. and El-Azab, Anter and Allen, Todd R.}, year={2014}, month={Nov}, pages={283–289} }
 @article{he_pakarinen_kirk_gan_nelson_bai_el-azab_allen_2014, title={Microstructure evolution in Xe-irradiated UO2 at room temperature}, volume={330}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84899120066&partnerID=MN8TOARS}, DOI={10.1016/j.nimb.2014.03.018}, abstractNote={In situ Transmission Electron Microscopy was conducted for single crystal UO2 to understand the microstructure evolution during 300 keV Xe irradiation at room temperature. The dislocation microstructure evolution was shown to occur as nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation segments and tangles at higher doses. Xe bubbles with dimensions of 1–2 nm were observed after room-temperature irradiation. Electron Energy Loss Spectroscopy indicated that UO2 remained stoichiometric under room temperature Xe irradiation.}, journal={Nuclear Instruments and Methods in Physics Research, Section B: Beam Interactions with Materials and Atoms}, author={He, L.F. and Pakarinen, J. and Kirk, M.A. and Gan, J. and Nelson, A.T. and Bai, X.-M. and El-Azab, A. and Allen, T.R.}, year={2014}, pages={55–60} }
 @article{he_shirahata_suematsu_nakayama_suzuki_jiang_niihara_2014, title={Synthesis of BN nanosheet/nanotube-Fe nanocomposites by pulsed wire discharge and high-temperature annealing}, volume={117}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84890959203&partnerID=MN8TOARS}, DOI={10.1016/j.matlet.2013.12.008}, abstractNote={Abstract This letter presents a simple method to synthesize boron nitride nanosheet (BNNS)/boron nitride nanotube (BNNT)–Fe nanocomposites via pulsed wire discharge (PWD) and high-temperature annealing. BNNS–Fe nanocomposites were synthesized by discharging BNNS-coated Fe wires in ambient nitrogen. The as-synthesized nanocomposites were annealed at 1200 °C to develop BNNTs. The Fe nanoparticles were found as the catalysts for growing BNNTs.}, journal={Materials Letters}, author={He, L.F. and Shirahata, J. and Suematsu, H. and Nakayama, T. and Suzuki, T. and Jiang, W. and Niihara, K.}, year={2014}, pages={120–123} }
 @article{khafizov_park_chernatynskiy_he_lin_moore_swank_lillo_phillpot_el-azab_et al._2014, title={Thermal conductivity in nanocrystalline ceria thin films}, volume={97}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84893977235&partnerID=MN8TOARS}, DOI={10.1111/jace.12673}, abstractNote={The thermal conductivity of nanocrystalline ceria films grown by unbalanced magnetron sputtering is determined as a function of temperature using laser‐based modulated thermoreflectance. The films exhibit significantly reduced conductivity compared with stoichiometric bulk CeO2. A variety of microstructure imaging techniques including X‐ray diffraction, scanning and transmission electron microscopy, X‐ray photoelectron analysis, and electron energy loss spectroscopy indicate that the thermal conductivity is influenced by grain boundaries, dislocations, and oxygen vacancies. The temperature dependence of the thermal conductivity is analyzed using an analytical solution of the Boltzmann transport equation. The conclusion of this study is that oxygen vacancies pose a smaller impediment to thermal transport when they segregate along grain boundaries.}, number={2}, journal={Journal of the American Ceramic Society}, author={Khafizov, M. and Park, I.-W. and Chernatynskiy, A. and He, L. and Lin, J. and Moore, J.J. and Swank, D. and Lillo, T. and Phillpot, S.R. and El-Azab, A. and et al.}, year={2014}, pages={562–569} }
 @inproceedings{he_cao_sridharan_allen_moss_was_lian_2013, title={Characterization of Oxide Films on Alloys 600 and 690 Exposed to Supercritical and Subcritical Water}, booktitle={16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors}, author={He, L. and Cao, G. and Sridharan, K. and Allen, T.R. and Moss, T. and Was, G.S. and Lian, T.}, year={2013} }
 @article{fission products in nuclear fuel: comparison of simulated distribution with correlative characterization techniques_2013, url={http://journals.cambridge.org/action/displayAbstract?fromPage=online&aid=9032448}, DOI={http://dx.doi.org/10.1017/S1431927613006831}, abstractNote={Extended abstract of a paper presented at Microscopy and Microanalysis 2013 in Indianapolis, Indiana, USA, August 4 – August 8, 2013.}, year={2013}, month={Oct} }
 @article{valderrama_henderson_he_yablinsky_gan_hassan_el-azab_allen_manuel_2013, title={Fission Products in Nuclear Fuel: Comparison of Simulated Distribution with Correlative Characterization Techniques}, volume={19}, ISSN={1431-9276 1435-8115}, url={http://dx.doi.org/10.1017/s1431927613006831}, DOI={10.1017/s1431927613006831}, abstractNote={Extended abstract of a paper presented at Microscopy and Microanalysis 2013 in Indianapolis, Indiana, USA, August 4 – August 8, 2013.}, number={S2}, journal={Microscopy and Microanalysis}, publisher={Oxford University Press (OUP)}, author={Valderrama, B. and Henderson, H.B. and He, L. and Yablinsky, C. and Gan, J. and Hassan, A.-R. and El-Azab, A. and Allen, T.R. and Manuel, M.V.}, year={2013}, month={Aug}, pages={968–969} }
 @article{he_gupta_yablinsky_gan_kirk_bai_pakarinen_allen_2013, title={In situ TEM observation of dislocation evolution in Kr-irradiated UO 2 single crystal}, volume={443}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84880894907&partnerID=MN8TOARS}, DOI={10.1016/j.jnucmat.2013.06.050}, abstractNote={In situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 were similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 °C and 1 MeV at 800 °C) used in this work. Interstitial-type dislocation loops with Burgers vector along 〈1 1 0〉 were observed in the Kr-irradiated UO2. The irradiated specimens were denuded of dislocation loops near the surface.}, number={1-3}, journal={Journal of Nuclear Materials}, author={He, L.-F. and Gupta, M. and Yablinsky, C.A. and Gan, J. and Kirk, M.A. and Bai, X.-M. and Pakarinen, J. and Allen, T.R.}, year={2013}, pages={71–77} }
 @article{zhou_he_lin_wang_2013, title={Synthesis and structure-property relationships of a new family of layered carbides in Zr-Al(Si)-C and Hf-Al(Si)-C systems}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84879063018&partnerID=MN8TOARS}, DOI={10.1016/j.jeurceramsoc.2013.05.020}, abstractNote={The layered ternary and quaternary carbides in Zr-Al(Si)-C and Hf-Al(Si)-C systems with general formulae of (TC)nAl3C2, (TC)nAl4C3 and (TC)n[Al(Si)]4C3 (where T = Zr or Hf, n = 1, 2, 3…) have attracted increasing attentions due to their fascinating properties such as high specific stiffness, high strength and fracture toughness, refractory, machinability by electrical discharge method, thermal shock resistance, as well as high-temperature and ultrahigh-temperature oxidation resistance. The combination of these properties makes them promising as structural components or coatings for high- and ultrahigh-temperature applications. In this review, the progresses on processing, and structure–property relationships of the novel layered carbides are comprehensively outlined. The crystal structure characteristics are introduced first. Then, methods for processing powders and bulk samples are summarized. The third section focuses on the multi-scale structure–property relationships. Finally, the potential applications and further trends in tailoring the properties and developing low cost processing methods are highlighted.}, journal={Journal of the European Ceramic Society}, author={Zhou, Y.-C. and He, L.-F. and Lin, Z.-J. and Wang, J.-Y.}, year={2013} }
 @article{he_yablinsky_gupta_gan_kirk_allen_2013, title={Transmission electron microscopy investigation of krypton bubbles in polycrystalline CeO2}, volume={182}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84879071544&partnerID=MN8TOARS}, number={2}, journal={Nuclear Technology}, author={He, L. and Yablinsky, C. and Gupta, M. and Gan, J. and Kirk, M.A. and Allen, T.R.}, year={2013}, pages={164–169} }
 @article{lu_xiang_he_sun_zhou_2011, title={Effect of Ti dopant on the mechanical properties and oxidation behavior of Zr 2[Al(Si)] 4C 5 ceramics}, volume={94}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-79958145065&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2010.04300.x}, abstractNote={Ti-doped Zr(2)[Al(Si)](4)C(5) solid solutions were prepared by an in situ hot-pressing method and its effect on the mechanical properties and high-temperature oxidation behavior were investigated. The solid solutions show comparable hardness, strength, and fracture toughness with Zr(2)[Al(Si)](4)C(5) except modulus, which decreases with Ti dopant content. The stiffness is maintained up to 1600 degrees C, which derives from the clean grain boundaries without glassy phases. The oxidation resistance of [Zr(1-x)(Ti)(x)](2)[Al(Si)](4)C(5) solid solutions at 1000 degrees-1300 degrees C is improved remarkably. The improved oxidation resistance is mostly due to the formation of a more protective oxide scale consisting of (Zr,Ti)O(2), Al(2)O(3), and mullite.}, number={6}, journal={Journal of the American Ceramic Society}, author={Lu, X. and Xiang, H. and He, L.-F. and Sun, L. and Zhou, Y.}, year={2011}, pages={1872–1877} }
 @article{zhong_wang_he_jiang_zhai_lin_chen_chang_2011, title={Fabrication and characterization of tricalcium silicate bioceramics with high mechanical properties by spark plasma sintering}, volume={8}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-79955659133&partnerID=MN8TOARS}, DOI={10.1111/j.1744-7402.2011.02613.x}, abstractNote={Bioactive tricalcium silicate ceramics (Ca3SiO5) were fabricated by spark plasma sintering (SPS), and their sinterability and mechanical properties were examined. The bioactivity and in vitro biocompatibility of Ca3SiO5 ceramics were evaluated. Ca3SiO5 ceramics show higher density and superior mechanical properties compared with those prepared by conventional pressureless sintering. In addition, hydroxyapatite was induced to form on the surface of Ca3SiO5 ceramics when soaked in simulated body fluid and bone marrow mesenchymal stem cells were attached and spread well on the ceramics. Ca3SiO5 ceramics fabricated by SPS possess excellent mechanical properties, bioactivity, and biocompatibility and are promising bone repaired materials.}, number={3}, journal={International Journal of Applied Ceramic Technology}, author={Zhong, H. and Wang, L. and He, L. and Jiang, W. and Zhai, W. and Lin, K. and Chen, L. and Chang, J.}, year={2011}, pages={501–510} }
 @article{zhong_wang_fan_he_lin_jiang_chang_chen_2011, title={Mechanical properties and bioactivity of β-Ca 2SiO 4 ceramics synthesized by spark plasma sintering}, volume={37}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-79960328263&partnerID=MN8TOARS}, DOI={10.1016/j.ceramint.2011.03.037}, abstractNote={Bioactive beta-dicalcium silicate ceramics (β-Ca2SiO4) were fabricated by spark plasma sintering (SPS). The relative density of as-prepared β-Ca2SiO4 ceramics reached 98.1% when sintered at 1150 °C, leading to great improvement in bending strength (293 MPa), almost 10 times higher than that of the specimen prepared by conventional pressureless sintering (PLS). High fracture toughness (3.0 MPa m1/2) and Vickers hardness (5.8 GPa) of β-Ca2SiO4 ceramics were also achieved by SPS at 1150 °C. The simulated body fluid (SBF) results showed that β-Ca2SiO4 ceramics had a good in vitro bioactivity to induce hydraxyapatite (HAp) formation on their surface, which suggests that β-Ca2SiO4 ceramics are promising candidates for load-bearing bone implant materials.}, number={7}, journal={Ceramics International}, author={Zhong, H. and Wang, L. and Fan, Y. and He, L. and Lin, K. and Jiang, W. and Chang, J. and Chen, L.}, year={2011}, pages={2459–2465} }
 @article{he_shirahata_nakayama_suzuki_suematsu_ihara_bao_komatsu_niihara_2011, title={Mechanical properties of Y2Ti2O7}, volume={64}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-78651300311&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2010.11.042}, abstractNote={The mechanical properties of Y2Ti2O7 ceramics have been systematically characterized. Y2Ti2O7 shows higher hardness (12.1 GPa) than Y2O3 and TiO2. The Young’s modulus of Y2Ti2O7 (262 GPa) is close to that of TiO2 and 50% higher than that of Y2O3. The fracture toughness (1.9 MPa m1/2) and bending strength (206 MPa) are comparable with those of Y2O3 but half those of TiO2. The indentation size effect on the hardness in nanoindentation tests is discussed based on the Nix–Gao model.}, number={6}, journal={Scripta Materialia}, author={He, L.F. and Shirahata, J. and Nakayama, T. and Suzuki, T. and Suematsu, H. and Ihara, I. and Bao, Y.W. and Komatsu, T. and Niihara, K.}, year={2011}, pages={548–551} }
 @article{wan_he_zheng_zhang_bao_zhou_2010, title={A new method to improve the high-temperature mechanical properties of Ti3SiC2 by substituting Ti with Zr, Hf, or Nb}, volume={93}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77953098758&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2010.03637.x}, abstractNote={Ti3SiC2shows a unique combination of the properties of both metals and ceramics. However, its stiffness and strength lose rapidly above 1050°C, which is the main obstacle for the high‐temperature application of this material. To improve the high‐temperature mechanical properties of Ti3SiC2, Zr, Hf, or Nb were used as dopants in Ti3(SiAl)C2. At room temperature, the Zr‐, Hf‐, or Nb‐doped Ti3(SiAl)C2ceramics have comparable stiffness, hardness, strength, and fracture toughness with those of Ti3(SiAl)C2. At high temperatures, however, a significant improvement in stiffness and strength has been achieved for (Ti1−xTx)3(SiAl)C2(T=Zr, Hf, or Nb). (Ti1−xTx)3(SiAl)C2can retain high degrees of stiffness and strength up to 1200°C, which is 150°C higher than those for Ti3(SiAl)C2.}, number={6}, journal={Journal of the American Ceramic Society}, author={Wan, D.-T. and He, L.-F. and Zheng, L.-L. and Zhang, J. and Bao, Y.-W. and Zhou, Y.-C.}, year={2010}, pages={1749–1753} }
 @article{nian_he_li_wang_zhou_2010, title={Crystal structure and theoretical elastic property of a new ternary ceramic HfAl4C4}, volume={93}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77950477327&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2009.03543.x}, abstractNote={HfAl4C4, a new ternary aluminum carbide, was discovered and its crystal structure was determined by a combination of X‐ray diffraction, transmission electron microscopy, and first‐principles calculations. The crystal structure is trigonal belonging to thespace group. The refined lattice constants area=0.3308 nm,c=2.190 nm. First‐principles method was used to calculate the theoretical second‐order elastic constants, bulk modulus, shear modulus, and the Young's modulus of HfAl4C4. It shows that HfAl4C4has relatively high elastic stiffness.}, number={4}, journal={Journal of the American Ceramic Society}, author={Nian, H. and He, L. and Li, F. and Wang, J. and Zhou, Y.}, year={2010}, pages={1164–1168} }
 @article{hu_he_li_wu_wang_li_bao_zhou_2010, title={In situ reaction synthesis and mechanical properties of TaC-TaSi 2 composites}, volume={7}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-78449266486&partnerID=MN8TOARS}, DOI={10.1111/j.1744-7402.2009.02392.x}, abstractNote={TaC–TaSi2 composites were fabricated at 1700°C by an in situ reaction/hot pressing method using Ta, Si, and graphite as initial materials. TaSi2 content was 0–100 vol%. The microstructure and mechanical properties of the composites were investigated. It was found that the relative densities of composites were above 97.5% when the volume content of TaSi2 was above 10%. The TaC/10 vol% TaSi2 composite presented the highest flexural strength of 376 MPa. When the TaSi2 content was 30–50 vol%, the composites showed the highest fracture toughness of about 4.3 MP·am1/2. In addition, the composites could retain high Young's modulus up to at least 1525°C.}, number={6}, journal={International Journal of Applied Ceramic Technology}, author={Hu, C. and He, L. and Li, F. and Wu, L. and Wang, J. and Li, M. and Bao, Y. and Zhou, Y.}, year={2010}, pages={697–703} }
 @article{he_nian_lu_bao_zhou_2010, title={Mechanical and thermal properties of a Hf2[Al(Si)]4C5 ceramic prepared by in situ reaction/hot-pressing}, volume={62}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-73549118598&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2009.12.020}, abstractNote={Predominantly single-phase Hf2[Al(Si)]4C5 ceramic has been fabricated by an in situ reaction/hot-pressing method using Hf, Al, Si and graphite as starting materials. Hf2[Al(Si)]4C5 shows comparable mechanical properties to Zr2[Al(Si)]4C5, and lower hardness and stiffness but higher strength and toughness than HfC. The stiffness decreases slowly with temperature and at 1600 °C it remains 83% of that at ambient temperature. Compared to Zr2[Al(Si)]4C5 and HfC, however, Hf2[Al(Si)]4C5 exhibits a relatively higher coefficient of thermal expansion, an intermediate specific heat capacity and a lower thermal conductivity.}, number={6}, journal={Scripta Materialia}, author={He, L.F. and Nian, H.Q. and Lu, X.P. and Bao, Y.W. and Zhou, Y.C.}, year={2010}, pages={427–430} }
 @article{zhang_presser_berthold_nickel_wang_raisch_chassé_he_zhou_2010, title={Mechanisms and kinetics of the hydrothermal oxidation of bulk titanium silicon carbide}, volume={93}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77950478718&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2009.03570.x}, abstractNote={Hydrothermal oxidation of bulk Ti3SiC2in continuous water flow was studied at 500°–700°C under a hydrostatic pressure of 35 MPa. The oxidation was weak at 500°–600°C and accelerated at 700°C due to the formation of cracks in oxides. The kinetics obeyed a linear time‐law. Due to the high solubility of silica in hydrothermal water, the resulting oxide layers only consisted of titanium oxides and carbon. Besides general oxidation, two special modes are very likely present in current experiments: (1) preferential hydrothermal oxidation of lattice planes perpendicular to thec‐axis inducing cleavage of grains and (2) uneven hydrothermal oxidation related to the occurrence of TiC and SiC impurity inclusions. Nonetheless the resistance against hydrothermal oxidation is remarkably high up to 700°C.}, number={4}, journal={Journal of the American Ceramic Society}, author={Zhang, H. and Presser, V. and Berthold, C. and Nickel, K.G. and Wang, X. and Raisch, C. and Chassé, T. and He, L. and Zhou, Y.}, year={2010}, pages={1148–1155} }
 @article{he_li_lu_bao_zhou_2010, title={Microstructure, mechanical, thermal, and oxidation properties of a Zr2[Al(Si)]4C5-SiC composite prepared by in situ reaction/hot-pressing}, volume={30}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77953541682&partnerID=MN8TOARS}, DOI={10.1016/j.jeurceramsoc.2010.02.005}, abstractNote={The microstructure, mechanical and thermal properties, as well as oxidation behavior, of in situ hot-pressed Zr2[Al(Si)]4C5–30 vol.% SiC composite have been characterized. The microstructure is composed of elongated Zr2[Al(Si)]4C5 grains and embedded SiC particles. The composite shows superior hardness (Vickers hardness of 16.4 GPa), stiffness (Young's modulus of 386 GPa), strength (bending strength of 353 MPa), and toughness (fracture toughness of 6.62 MPa m1/2) compared to a monolithic Zr2[Al(Si)]4C5 ceramic. Stiffness is maintained up to 1600 °C (323 GPa) due to clean grain boundaries with no glassy phase. The composite also exhibits higher specific heat capacity and thermal conductivity as well as better oxidation resistance compared to Zr2[Al(Si)]4C5.}, number={11}, journal={Journal of the European Ceramic Society}, author={He, L.-F. and Li, F.-Z. and Lu, X.-P. and Bao, Y.-W. and Zhou, Y.-C.}, year={2010}, pages={2147–2154} }
 @article{he_li_nian_wang_bao_li_wang_zhou_2010, title={Oxidation behavior of ternary carbide ceramics in Hf-Al-C system in air}, volume={93}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-78649552470&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2010.03891.x}, abstractNote={ The oxidation behavior of Hf–Al–C ceramics containing 37.5 wt% Hf3Al3C5, 30.5 wt% Hf2Al4C5, and 32.0 wt% Hf3Al4C6 has been investigated at 900°–1300°C in air. The oxidation kinetics approximately follows a linear law with the activation energy of 194±12 kJ/mol. The oxidation resistance of Hf–Al–C ceramics at high temperature is superior to HfC. The oxide scale is porous and composed of well‐mixed t‐HfO2 and Al2O3 as well as residual free carbon. The stabilization mechanisms of t‐HfO2 and free carbon have been discussed. The simultaneous oxidation of Hf and Al in Hf–Al–C ceramics can be attributed to their close oxygen affinity as well as the strong coupling between Hf–C blocks and Al–C units in the crystal structures. }, number={10}, journal={Journal of the American Ceramic Society}, author={He, L.-F. and Li, J.-J. and Nian, H.-Q. and Wang, X.-H. and Bao, Y.-W. and Li, M.-S. and Wang, J.-Y. and Zhou, Y.-C.}, year={2010}, pages={3427–3431} }
 @article{wu_he_chen_lu_zhou_2010, title={Reciprocating friction and wear behavior of Zr2[Al(Si)] 4C5 and Zr2[Al(Si)]4C 5-SiC composite against Si3N4 ball}, volume={93}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77955757370&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2010.03718.x}, abstractNote={ The reciprocating sliding friction and wear properties of two novel materials of Zr2[Al(Si)]4C5 ceramic and Zr2[Al(Si)]4C5–30 vol% SiC composite against Si3N4 ball were investigated. The sliding friction process of Zr2[Al(Si)]4C5 against Si3N4 experiences two different stages under constant normal load and involves friction and wear mechanism transition. The static coefficient of friction increases with an increasing normal load. The friction force mainly comes from the interfacial shear between Si3N4 ball and Zr2[Al(Si)]4C5, which changes with varied sliding distances and normal loads. In contrast, the friction process of the composite experiences one stage and the friction coefficient is not related to the test durations and normal loads. The friction force between Zr2[Al(Si)]4C5–30 vol% SiC composite and Si3N4 is mainly from the plough between SiC particles and Si3N4 ball, which appears not to be influenced significantly by different normal load and sliding distance. In addition, microfracture induced mechanical wear is the rate‐control wear mechanism in both Zr2[Al(Si)]4C5 and Zr2[Al(Si)]4C5–30 vol% SiC composite. Adding SiC improves the wear resistance of the single‐phase material, because the second phase bears normal load and slows down material removal. }, number={8}, journal={Journal of the American Ceramic Society}, author={Wu, L. and He, L.-F. and Chen, J.-X. and Lu, X.-P. and Zhou, Y.-C.}, year={2010}, pages={2369–2376} }
 @article{guo_li_zhang_he_zhou_2010, title={Surface strengthening of Ti3SiC2 through magnetron sputtering of Mo and Zr and subsequent annealing}, volume={30}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-77952950712&partnerID=MN8TOARS}, DOI={10.1016/j.jeurceramsoc.2009.04.042}, abstractNote={Magnetron sputtering deposition of Mo and Zr and subsequent annealing were conducted with the motivation to modify the surface hardness of Ti3SiC2. For Mo-coated Ti3SiC2, Si diffused outward into the Mo layer and reacted with Mo to form molybdenum silicides in the temperature range of 1000–1100 °C. The MoSi2 layer, however, cracked and easily spalled off. For Zr-coated Ti3SiC2, Si also diffused outward to form Zr–Si intermetallic compounds at 900–1100 °C. The Zr–Si compounds layer had good adhesion with Ti3SiC2 substrate, which resulted in the increased surface hardness.}, number={10}, journal={Journal of the European Ceramic Society}, author={Guo, H. and Li, A. and Zhang, J. and He, L. and Zhou, Y.}, year={2010}, pages={2123–2130} }
 @article{he_lu_bao_wang_zhou_2009, title={High-temperature internal friction, stiffness and strength of Zr-Al(Si)-C ceramics}, volume={61}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-64849105114&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2009.03.004}, abstractNote={The temperature dependence of internal friction, stiffness and strength of Zr–Al(Si)–C ceramics was studied. For glass-free Zr–Al(Si)–C ceramics, the internal friction curve followed an exponential-like background. For liquid-phase-sintered Zr3Al3C5, an internal peak present at 1590 K revealed the grain-boundary sliding due to wetting of glass phase at grain boundaries and triple junctions. Zr–Al(Si)–C ceramics exhibited high degrees of stiffness and strength at high temperatures up to 1600 and 1400 °C, respectively, which render them good high-temperature structural materials.}, number={1}, journal={Scripta Materialia}, author={He, L.F. and Lu, X.P. and Bao, Y.W. and Wang, J.Y. and Zhou, Y.C.}, year={2009}, pages={60–63} }
 @article{zhang_he_zhou_2009, title={Highly conductive and strengthened copper matrix composite reinforced by Zr2Al3C4 particulates}, volume={60}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-62849094266&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2009.02.026}, abstractNote={A new copper matrix composite reinforced with Zr2Al3C4 particulates was successfully fabricated by powder metallurgy method. In the whole reinforcing range, Cu/Zr2Al3C4 composite possesses comparable electrical conductivity and better mechanical properties than those of Cu/graphite composite, which was widely accepted as the electro-contacting material. The high conductivity was attributed to the formation of a continuous copper network in the composite, while the high strength and modulus of Zr2Al3C4 ceramic led to a notable improvement in the mechanical properties of the composite.}, number={11}, journal={Scripta Materialia}, author={Zhang, J. and He, L. and Zhou, Y.}, year={2009}, pages={976–979} }
 @article{he_bao_wang_li_zhou_2009, title={Mechanical and thermophysical properties of Zr-Al-Si-C ceramics}, volume={92}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-60849121069&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2008.02879.x}, abstractNote={ The mechanical and thermophysical properties of quaternary‐layered carbides, Zr2[Al(Si)]4C5 and Zr3[Al(Si)]4C6 have been investigated and compared with those of Zr2Al3C4 and Zr3Al3C5. These four carbides are generally alike in mechanical and thermophysical properties due to their similar crystal structures that consisting of alternatively stacked ZrC layers and Al3C2/[Al(Si)]4C3 slabs. The layer thickness of zirconium carbide and aluminum carbide has effects on their properties. Thicker layer of zirconium carbide and/or thinner layer of aluminum carbides are in favor of stiffness, hardness, thermal, and electrical conductivities, but go against density, specific stiffness, Debye temperature, and coefficient of thermal expansion. }, number={2}, journal={Journal of the American Ceramic Society}, author={He, L.-F. and Bao, Y.-W. and Wang, J.-Y. and Li, M.-S. and Zhou, Y.-C.}, year={2009}, pages={445–451} }
 @article{he_bao_wang_li_zhou_2009, title={Microstructure and mechanical and thermal properties of ternary carbides in Hf-Al-C system}, volume={57}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-64649091299&partnerID=MN8TOARS}, DOI={10.1016/j.actamat.2009.02.027}, abstractNote={A Hf–Al–C composite composed of Hf3Al3C5, Hf2Al4C5 and Hf3Al4C6 has been successfully synthesized by a hot pressing method; its microstructure and mechanical and thermal properties were systematically characterized. Hf–Al–C composite conserves the high hardness and stiffness similar to HfC. Interestingly, the composite exhibits much higher strength and fracture toughness than HfC due to its fine and anisotropic grains. Diffusion-accommodated grain-boundary sliding of Hf–Al–C ceramics at high temperature is inhibited by glass-free grain boundaries and tight interlocking of grains at grain-edge triple junctions, resulting in high remaining stiffness up to 1600 °C. Dislocations on the basal planes of Hf–Al–C ceramics with a Burgers vector of 13〈112¯0〉 can be activated at high temperature. Hf–Al–C composite shows higher coefficient of thermal expansion and specific heat capacity as well as lower thermal conductivity than HfC. The superior mechanical and thermal properties make Hf–Al–C compounds good high-temperature structural materials.}, number={9}, journal={Acta Materialia}, author={He, L.F. and Bao, Y.W. and Wang, J.Y. and Li, M.S. and Zhou, Y.C.}, year={2009}, pages={2765–2774} }
 @article{wu_he_bao_zhou_2009, title={Tribological properties of a Zr2Al3C4 ceramic at ambient temperature}, volume={92}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-58149346018&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2008.02805.x}, abstractNote={In this work, reciprocating ball‐on‐flat sliding friction and wear tests as well as two‐body abrasive wear tests were performed on Zr2Al3C4, a new type of ceramic material. In the sliding wear tests, a Si3N4ball and an AISI 52100 steel ball were used as counter materials. When Zr2Al3C4was slid against an AISI 52100 ball, the coefficient of friction (COF) was as low as 0.20–0.42, being independent of normal loads, and the wear rates of Zr2Al3C4and AISI 52100 steel were in the range of 10−4–10−5mm3/m. A tribofilm between the tribopair was presumed to be responsible for the low COF and wear rate. According to Raman and energy‐dispersive spectroscopy analysis, the tribofilm consists of a mixture of oxides of Zr, Al, and Fe as well as amorphous carbon. When Zr2Al3C4was slid against a Si3N4ball, a transition from mild wear to severe wear was observed as the normal load increased. The transition occurs under certain contact stresses after a damage‐accumulating period. In the two‐body abrasive wear test, surface chipping and fragmentation resulting from coalescence of surface and subsurface microcracks were the main material removal mechanisms of Zr2Al3C4.}, number={1}, journal={Journal of the American Ceramic Society}, author={Wu, L. and He, L.-F. and Bao, Y.-W. and Zhou, Y.-C.}, year={2009}, pages={141–146} }
 @article{he_zhong_xu_li_bao_wang_zhou_2009, title={Ultrahigh-temperature oxidation of Zr2Al3C4 via rapid induction heating}, volume={60}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-58849152137&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2008.12.002}, abstractNote={The oxidation behavior of Zr 2 Al 3 C 4 at 1600 and 1750 °C was studied by means of induction heating. The oxidation kinetics follows a linear law and the interfacial reaction between Zr 2 Al 3 C 4 and O 2 is the rate-limiting step for scale growth. The microstructure of oxide scales gradually changes from intragranular (small ZrO 2 particles embedded in large Al 2 O 3 grains) to an interpenetrating dual-phase form as the exposure time increases, which is ascribed to the temperature decay of the top scale surface with scale growth.}, number={7}, journal={Scripta Materialia}, author={He, L.F. and Zhong, H.B. and Xu, J.J. and Li, M.S. and Bao, Y.W. and Wang, J.Y. and Zhou, Y.C.}, year={2009}, pages={547–550} }
 @article{he_bao_zhou_2009, title={Zirconium aluminum carbides: New precursors for synthesizing ZrO 2-Al2O3 composites}, volume={92}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-70350493137&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2009.03281.x}, abstractNote={ ZrO2–Al2O3 nanocrystalline powders have been synthesized by oxidizing ternary Zr2Al3C4 powders. The simultaneous oxidation of Al and Zr in Zr2Al3C4 results in homogeneous mixture of ZrO2 and Al2O3 at nanoscale. Bulk nano‐ and submicro‐composites were prepared by hot‐pressing as‐oxidized powders at 1100°–1500°C. The composition and microstructure evolution during sintering was investigated by XRD, Raman spectroscopy, SEM, and TEM. The crystallite size of ZrO2 in the composites increased from 7.5 nm for as‐oxidized powders to about 0.5 μm at 1500°C, while the tetragonal polymorph gradually converted to monolithic one with increasing crystallite size. The Al2O3 in the composites transformed from an amorphous phase in as oxidized powders to θ phase at 1100°C and α phase at higher temperatures. The hardness of the composite increased from 2.0 GPa at 1100°C to 13.5 GPa at 1400°C due to the increase of density. }, number={11}, journal={Journal of the American Ceramic Society}, author={He, L.-F. and Bao, Y.-W. and Zhou, Y.-C.}, year={2009}, pages={2751–2758} }
 @article{lin_he_wang_li_bao_zhou_2008, title={Atomic-scale microstructure and elastic properties of quaternary Zr-Al-Si-C ceramics}, volume={56}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-42949091917&partnerID=MN8TOARS}, DOI={10.1016/j.actamat.2007.12.055}, abstractNote={Transmission electron microscopy characterizations and elastic properties of two quaternary carbides, i.e. Zr2(Al(Si))4C5 and Zr3(Al(Si))4C6 are reported. The space group and atomic-scale microstructures of both compounds were determined using a combination of selected area electron diffraction, convergent beam electron diffraction, high-resolution transmission electron microscopy and Z-contrast scanning transmission electron microscopy. In addition, the combined experimental and theoretical studies on elastic properties for Zr2(Al(Si))4C5 are presented. A full set of second-order elastic constants, bulk modulus, shear modulus, and Young’s modulus were calculated using first-principles calculations. Both experimental and theoretical works demonstrated that quaternary Zr–Al–Si–C ceramics possess close elastic properties to ZrC. Furthermore, Zr2(Al(Si))4C5 retained a high Young’s modulus up to about 1580 °C, which can be attributed to its comparable activation energy of lattice drag process to that of ZrC.}, number={9}, journal={Acta Materialia}, author={Lin, Z.J. and He, L.F. and Wang, J.Y. and Li, M.S. and Bao, Y.W. and Zhou, Y.C.}, year={2008}, pages={2022–2031} }
 @article{he_lin_wang_bao_zhou_2008, title={Crystal structure and theoretical elastic property of two new ternary ceramics Hf3Al4C6 and Hf2Al4C5}, volume={58}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-39149128249&partnerID=MN8TOARS}, DOI={10.1016/j.scriptamat.2007.12.002}, abstractNote={Two new ternary aluminum carbides, Hf3Al4C6 and Hf2Al4C5, were identified and their crystal structure was determined by a combination of X-ray diffraction, Z-contrast scanning transmission electron microscopy and density-function calculations. Theoretical second-order elastic constants, bulk modulus, shear modulus and Young’s modulus of Hf3Al4C6 and Hf2Al4C5, as well as Hf3Al3C5, Hf2Al3C4, HfC and Al4C3, were calculated and compared. These new carbides show promisingly high elastic stiffness.}, number={8}, journal={Scripta Materialia}, author={He, L.F. and Lin, Z.J. and Wang, J.Y. and Bao, Y.W. and Zhou, Y.C.}, year={2008}, pages={679–682} }
 @article{he_bao_li_wang_zhou_2008, title={Improving the high-temperature oxidation resistance of Zr2 Al3C4 by silicon pack cementation}, volume={23}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-50449089651&partnerID=MN8TOARS}, DOI={10.1557/jmr.2008.0285}, abstractNote={Silicon pack cementation has been applied to improve the oxidation resistance of Zr2Al3C4. The Si pack coating is mainly composed of an inner layer of ZrSi2 and SiC and an outer layer of Al2O3 at 1200 °C. The growth kinetics of silicide coating at 1000–1200 °C obey a parabolic law with an activation energy of 110.3 ± 16.7 kJ/mol, which is controlled by inward diffusion of Si and outward diffusion of Al. Compared with Zr2Al3C4, the oxidation resistance of siliconized Zr2Al3C4 is greatly improved due to the formation of protective oxidation products, aluminosilicate glass, mullite, and ZrSiO4.}, number={8}, journal={Journal of Materials Research}, author={He, L.F. and Bao, Y.W. and Li, M.S. and Wang, J.Y. and Zhou, Y.C.}, year={2008}, pages={2275–2282} }
 @article{hu_li_he_liu_zhang_wang_bao_wang_zhou_2008, title={In Situ}, volume={91}, ISSN={0002-7820 1551-2916}, url={http://dx.doi.org/10.1111/j.1551-2916.2008.02424.x}, DOI={10.1111/j.1551-2916.2008.02424.x}, abstractNote={ In this work, a bulk Nb4AlC3 ceramic was prepared by an in situ reaction/hot pressing method using Nb, Al, and C as the starting materials. The reaction path, microstructure, physical, and mechanical properties of Nb4AlC3 were systematically investigated. The thermal expansion coefficient was determined as 7.2 × 10−6 K−1 in the temperature range of 200°–1100°C. The thermal conductivity of Nb4AlC3 increased from 13.5 W·(m·K)−1 at room temperature to 21.2 W·(m·K)−1 at 1227°C, and the electrical conductivity decreased from 3.35 × 106 to 1.13 × 106Ω−1·m−1 in a temperature range of 5–300 K. Nb4AlC3 possessed a low hardness of 2.6 GPa, high flexural strength of 346 MPa, and high fracture toughness of 7.1 MPa·m1/2. Most significantly, Nb4AlC3 could retain high modulus and strength up to very high temperatures. The Young's modulus at 1580°C was 241 GPa (79% of that at room temperature), and the flexural strength could retain the ambient strength value without any degradation up to the maximum measured temperature of 1400°C. }, number={7}, journal={Journal of the American Ceramic Society}, publisher={Wiley}, author={Hu, Chunfeng and Li, Fangzhi and He, Lingfeng and Liu, Mingyue and Zhang, Jie and Wang, Jiemin and Bao, Yiwang and Wang, Jingyang and Zhou, Yanchun}, year={2008}, month={Jul}, pages={2258–2263} }
 @article{hu_he_liu_wang_wang_li_bao_zhou_2008, title={In situ reaction synthesis and mechanical properties of V2AlC}, volume={91}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-57649165082&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2008.02774.x}, abstractNote={ A dense V2AlC ceramic was synthesized by an in situ reaction/hot pressing method using V, Al, and C powders as starting materials. The reaction path was investigated. It was found that V2AlC was produced by a reaction between Al8V5 and C. Single ‐phase V2AlC could be prepared using the optimized molar ratio of V:Al:C=2:1.2:0.9. The grain sizes of as‐prepared V2AlC samples were temperature dependent. On increasing the temperature from 1400° to 1700°C, the mean grain size of V2AlC increased from 49 μm in length and 19 μm in width to 405 μm in length and 106 μm in width. The V2AlC sample sintered at 1500°C exhibited the highest flexural strength of 289 MPa and a fracture toughness of 5.7 MPa·m1/2, while the V2AlC sample prepared at 1400°C showed the highest compressive strength of 742 MPa. The sample sintered at 1700°C possessed the highest damage tolerance. Additionally, V2AlC could retain a high degree of Young's modulus up to 1200°C. Below 800°C, V2AlC showed excellent thermal shock resistance. }, number={12}, journal={Journal of the American Ceramic Society}, author={Hu, C. and He, L. and Liu, M. and Wang, X. and Wang, J. and Li, M. and Bao, Y. and Zhou, Y.}, year={2008}, pages={4029–4035} }
 @article{he_lin_bao_li_wang_zhou_2008, title={Isothermal oxidation of bulk Zr2Al3C4 at 500 to 1000 °C in air}, volume={23}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-39749124821&partnerID=MN8TOARS}, DOI={10.1557/jmr.2008.0042}, abstractNote={The isothermal oxidation behavior of Zr2Al3C4 in the temperature range of 500 to 1000 °C for 20 h in air has been investigated. The oxidation kinetics follow a parabolic law at 600 to 800 °C and a linear law at higher temperatures. The activation energy is determined to be 167.4 and 201.2 kJ/mol at parabolic and linear stages, respectively. The oxide scales have a monolayer structure, which is a mixture of ZrO2 and Al2O3. As indicated by x-ray diffraction and Raman spectra, the scales formed at 500 to 700 °C are amorphous, and at higher temperatures are α-Al2O3 and t-ZrO2 nanocrystallites. The nonselective oxidation of Zr2Al3C4 can be attributed to the strong coupling between Al3C2 units and ZrC blocks in its structure, and the close oxygen affinity of Zr and Al.}, number={2}, journal={Journal of Materials Research}, author={He, L.F. and Lin, Z.J. and Bao, Y.W. and Li, M.S. and Wang, J.Y. and Zhou, Y.C.}, year={2008}, pages={359–366} }
 @article{hu_he_zhang_bao_wang_li_zhou_2008, title={Microstructure and properties of bulk Ta2AlC ceramic synthesized by an in situ reaction/hot pressing method}, volume={28}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-41149160354&partnerID=MN8TOARS}, DOI={10.1016/j.jeurceramsoc.2007.10.006}, abstractNote={Dense bulk Ta2AlC ceramic was synthesized by an in situ reaction/hot pressing method using Ta, Al, and C as initial materials. The average grain size of Ta2AlC is 15 μm in length and 3 μm in width. The physical and mechanical properties were investigated. Ta2AlC is a good electrical and thermal conductor. The flexural strength and fracture toughness of Ta2AlC were measured to be 360 MPa and 7.7 MPa m1/2, respectively. The typical layered grains contribute to the damage tolerance of this ceramic. After indentation up to 200 N at the tensile surface of the beam specimens, no obvious decrease of the residual flexural strength was observed. Even at above 1200 °C, Ta2AlC still retains a high Young's modulus and shows excellent thermal shock resistance, which renders it a promising high-temperature structural material.}, number={8}, journal={Journal of the European Ceramic Society}, author={Hu, C. and He, L. and Zhang, J. and Bao, Y. and Wang, J. and Li, M. and Zhou, Y.}, year={2008}, pages={1679–1685} }
 @article{he_bao_li_wang_zhou_2008, title={Oxidation of Zr<sub>2</sub>[Al(Si)]<sub>4</sub>C<sub>5</sub> and Zr<sub>3</sub>[Al(Si)]<sub>4</sub>C<sub>6</sub> in air}, volume={23}, ISSN={0884-2914 2044-5326}, url={http://dx.doi.org/10.1557/jmr.2008.0411}, DOI={10.1557/jmr.2008.0411}, abstractNote={The oxidation behavior of Zr2[Al(Si)]4C5 and Zr3[Al(Si)]4C6 in air has been investigated. The oxidation kinetics of bulk Zr2[Al(Si)]4C5 and Zr3[Al(Si)]4C6 at 900–1300 °C generally follow a parabolic law at a very short initial stage and then a linear law for a long period with the activation energy of 237.9 and 226.8 kJ/mol, respectively. The oxide scales have a duplex structure, consisting of mainly an outer porous layer of ZrO2, Al2O3, and aluminosilicate/mullite, and a thin inner compact layer of these oxides plus remaining carbon. The oxidation resistance of Zr2[Al(Si)]4C5 and Zr3[Al(Si)]4C6 has been improved compared with Zr2Al3C4, and is much better than ZrC due to larger fraction of protective oxidation products, Al2O3 and aluminosilicate/mullite.}, number={12}, journal={Journal of Materials Research}, publisher={Springer Science and Business Media LLC}, author={He, L.F. and Bao, Y.W. and Li, M.S. and Wang, J.Y. and Zhou, Y.C.}, year={2008}, month={Dec}, pages={3339–3346} }
 @article{l.f. he_2008, title={Synthesis, Microstructure, and Mechanical Properties of Al3BC3}, url={http://onlinelibrary.wiley.com/doi/10.1111/j.1551-2916.2008.02437.x/abstract}, DOI={DOI: 10.1111/j.1551-2916.2008.02424.x}, abstractNote={ In this work, a bulk Nb4AlC3 ceramic was prepared by an in situ reaction/hot pressing method using Nb, Al, and C as the starting materials. The reaction path, microstructure, physical, and mechanical properties of Nb4AlC3 were systematically investigated. The thermal expansion coefficient was determined as 7.2 × 10−6 K−1 in the temperature range of 200°–1100°C. The thermal conductivity of Nb4AlC3 increased from 13.5 W·(m·K)−1 at room temperature to 21.2 W·(m·K)−1 at 1227°C, and the electrical conductivity decreased from 3.35 × 106 to 1.13 × 106Ω−1·m−1 in a temperature range of 5–300 K. Nb4AlC3 possessed a low hardness of 2.6 GPa, high flexural strength of 346 MPa, and high fracture toughness of 7.1 MPa·m1/2. Most significantly, Nb4AlC3 could retain high modulus and strength up to very high temperatures. The Young's modulus at 1580°C was 241 GPa (79% of that at room temperature), and the flexural strength could retain the ambient strength value without any degradation up to the maximum measured temperature of 1400°C. }, author={L.F. He, L. He}, year={2008} }
 @article{li_zhou_he_liu_wang_2008, title={Synthesis, Microstructure, and Mechanical Properties of Al<sub>3}, volume={91}, ISSN={0002-7820 1551-2916}, url={http://dx.doi.org/10.1111/j.1551-2916.2008.02437.x}, DOI={10.1111/j.1551-2916.2008.02437.x}, abstractNote={In situsynthesis of bulk Al3BC3was achieved via a reactive hot‐pressing method using Al, B4C, and graphite powders at 1800°C for 2 h. The reaction path for synthesizing Al3BC3was investigated. It was found that Al3BC3formed via the reaction of C, B4C, and Al4C3above 1180°C. Dense Al3BC3was prepared with a little B4C and graphite remained. Microstructure observations revealed the plate‐like morphology of Al3BC3grains. Moreover, the mechanical properties of Al3BC3were characterized (Vickers hardness of 11.1 GPa, bending strength of 185 MPa, fracture toughness of 2.3 MPa·m1/2, and Young's modulus of 163 GPa). Young's modulus decreased slowly with increasing temperature, and at 1600°C remained 79% of that at ambient temperature. These results show that Al3BC3is a promising lightweight high temperature structural material.}, number={7}, journal={Journal of the American Ceramic Society}, publisher={Wiley}, author={Li, Fangzhi and Zhou, Yanchun and He, Lingfeng and Liu, Bin and Wang, Jingyang}, year={2008}, month={Jul}, pages={2343–2348} }
 @article{he_wang_bao_zhou_2007, title={Elastic and thermal properties of Zr2 Al3 C4: Experimental investigations and ab initio calculations}, volume={102}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-34548424959&partnerID=MN8TOARS}, DOI={10.1063/1.2773679}, abstractNote={This article presents the results of combined experimental and theoretical studies of elastic and thermal properties of Zr2Al3C4 carbide. The full set of second order elastic constants, bulk modulus, shear modulus, and Young’s modulus of Zr2Al3C4 were calculated and compared with those of Zr3Al3C5 and ZrC. The experimentally measured Young’s modulus and shear modulus are in good agreement with theoretical ones. The calculated Debye temperature from elastic constants of Zr2Al3C4 is 830 K, which is slightly higher than that of Zr3Al3C5, and exhibits pronounced enhancement in comparison with that of ZrC. The highest Debye temperature of Zr2Al3C4 is related with its highest specific stiffness, i.e., the stiffness-to-weight ratio. The heat capacity and thermal conductivity of Zr2Al3C4 were measured by means of the flash method. The thermal conductivity of Zr2Al3C4 decreases with increasing temperature, for instance the values at room temperature and 1600 K are 15.5 and 10.1 W/m K, respectively. The investigations provide information on elastic and thermal properties of Zr2Al3C4 with promising high temperature applications.}, number={4}, journal={Journal of Applied Physics}, author={He, L.F. and Wang, J.Y. and Bao, Y.W. and Zhou, Y.C.}, year={2007} }
 @article{li_li_zhou_zhang_he_2007, title={In situ synthesis and properties of Ti 3AlC 2/TiB 2 composites}, volume={90}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-35948964516&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2007.01954.x}, abstractNote={In order to improve the mechanical properties of Ti3AlC2, near‐fully dense Ti3AlC2/TiB2composites were synthesized using Ti, Al, graphite, and B4C powders as the initial materials. Compared with monolithic Ti3AlC2, the composites exhibit a much higher strength (for the compressive strength, from initial 723 MPa to maximal 2205 MPa; for flexural strength, from initial 340 MPa to maximal 861 MPa), and the strengthening effect can be held at least up to 1100°C. Moreover, besides the enhancement of the elastic modulus and hardness of Ti3AlC2, the introduction of a TiB2phase makes a positive contribution to its electrical conductivity.}, number={11}, journal={Journal of the American Ceramic Society}, author={Li, C. and Li, M. and Zhou, Y. and Zhang, J. and He, L.}, year={2007}, pages={3615–3620} }
 @article{lin_he_li_wang_zhou_2007, title={Layered stacking characteristics of ternary zirconium aluminum carbides}, volume={22}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-36549089117&partnerID=MN8TOARS}, DOI={10.1557/jmr.2007.0409}, abstractNote={Layered stacking characteristics of ternary Zr–Al–C carbides were investigated using scanning transmission electron microscopy (STEM). Three previously unknown compounds, i.e., Zr4Al3C6, Zr5Al6C9, and Zr7Al6C11 were identified. The present study extends the structural information of ternary Zr–Al–C ceramics. The influence of the thickness of the NaCl-type Zr-C slab on the elastic properties of ternary Zr–Al–C ceramics is discussed based on first-principles calculations. In addition, direct atomic-resolution observations illustrate the process for forming the unique layered crystal structures of ternary Zr–Al–C ceramics. These results also provide insights into the formation mechanism of layered ternary Zr–Al–C carbides.}, number={11}, journal={Journal of Materials Research}, author={Lin, Z.J. and He, L.F. and Li, M.S. and Wang, J.Y. and Zhou, Y.C.}, year={2007}, pages={3058–3066} }
 @article{hu_lin_he_bao_wang_li_zhou_2007, title={Physical and mechanical properties of bulk Ta4AlC3 ceramic prepared by an in situ reaction synthesis/hot-pressing method}, volume={90}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-34547670744&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2007.01804.x}, abstractNote={ Bulk Ta4AlC3 ceramic was prepared by an in situ reaction synthesis/hot‐pressing method using Ta, Al, and C powders as the starting materials. The lattice parameter and a new set of X‐ray diffraction data were obtained. The physical and mechanical properties of Ta4AlC3 ceramic were investigated. Ta4AlC3 is a good electrical and thermal conductor. The flexural strength and fracture toughness are 372 MPa and 7.7 MPa·m1/2, respectively. Typically, plate‐like layered grains contribute to the damage tolerance of Ta4AlC3. After indentation up to a 200 N load, no obvious degradation of the residual flexural strength of Ta4AlC3 was observed, demonstrating the damage tolerance of this ceramic. Even at above 1200°C in air, Ta4AlC3 still retains a high strength and shows excellent thermal shock resistance, which renders it a promising high‐temperature structural material. }, number={8}, journal={Journal of the American Ceramic Society}, author={Hu, C. and Lin, Z. and He, L. and Bao, Y. and Wang, J. and Li, M. and Zhou, Y.}, year={2007}, pages={2542–2548} }
 @article{he_lin_wang_bao_li_zhou_2007, title={Synthesis and characterization of bulk Zr2Al3C 4 ceramic}, volume={90}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-35948961626&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2007.01964.x}, abstractNote={Polycrystalline Zr2Al3C4was fabricated by anin situreactive hot‐pressing process using zirconium (zirconium hydrides), aluminum, and graphite as starting materials. The investigation on reaction path revealed that the liquid Al played an important role in triggering the formation of ternary zirconium aluminum carbides. The mechanical properties of Zr2Al3C4at room temperature were measured (Vickers hardness of 10.1 GPa, Young's modulus of 362 GPa, flexural strength of 405 MPa, and fracture toughness of 4.2 MPa·m1/2). The electrical resistivity and thermal expansion coefficient were determined as 1.10 μΩ·m and 8.1 × 10−6K−1, respectively.}, number={11}, journal={Journal of the American Ceramic Society}, author={He, L. and Lin, Z. and Wang, J. and Bao, Y. and Li, M. and Zhou, Y.}, year={2007}, pages={3687–3689} }
 @article{he_zhou_bao_wang_li_2007, title={Synthesis and oxidation of Zr3Al3C5 powders}, volume={98}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33847074954&partnerID=MN8TOARS}, number={1}, journal={International Journal of Materials Research}, author={He, L.F. and Zhou, Y.C. and Bao, Y.W. and Wang, J.Y. and Li, M.S.}, year={2007}, pages={3–9} }
 @article{he_zhou_bao_wang_li_2007, title={Synthesis and oxidation of Zr<inf>3</inf>Al<inf>3</inf>C<inf>5</inf> powders}, volume={98}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33847074954&partnerID=MN8TOARS}, number={1}, journal={International Journal of Materials Research}, author={He, L.F. and Zhou, Y.C. and Bao, Y.W. and Wang, J.Y. and Li, M.S.}, year={2007}, pages={3–9} }
 @article{he_zhou_bao_lin_wang_2007, title={Synthesis, physical, and mechanical properties of bulk Zr 3Al3C5 ceramic}, volume={90}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-34247168553&partnerID=MN8TOARS}, DOI={10.1111/j.1551-2916.2007.01518.x}, abstractNote={Anin situreactive hot‐pressing process using zirconium (zirconium hydride), aluminum, and graphite as staring materials and Si and Y2O3as additives was used to synthesize bulk Zr3Al3C5ceramics. This method demonstrates the advantages of easy synthesis, lower sintering temperature, high purity and density, and improved mechanical properties of synthesized Zr3Al3C5. Its electrical and thermal properties were measured. Compared with ZrC, Zr3Al3C5has a relatively low hardness (Vickers hardness of 12.5 GPa), comparable stiffness (Young's modulus of 374 GPa), but superior strength (flexural strength of 488 GPa) and toughness (fracture toughness of 4.68 MPa·m1/2). In addition, the stiffness decreases slowly with increasing temperature and at 1600°C remains 78% of that at ambient temperature, indicating that Zr3Al3C5is a potential high‐temperature structural ceramic.}, number={4}, journal={Journal of the American Ceramic Society}, author={He, L. and Zhou, Y. and Bao, Y. and Lin, Z. and Wang, J.}, year={2007}, pages={1164–1170} }
 @article{lin_zhuo_he_zhou_li_wang_2006, title={Atomic-scale microstructures of Zr2Al3C4 and Zr3Al3C5 ceramics}, volume={54}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33745988948&partnerID=MN8TOARS}, DOI={10.1016/j.actamat.2006.02.052}, abstractNote={The microstructures of bulk Zr2Al3C4 and Zr3Al3C5 ceramics have been investigated using transmission electron microscopy and scanning transmission electron microscopy. These two carbides were determined to have a point group 6/mmm and a space group P63/mmc using selected-area electron diffraction and convergent beam electron diffraction. The atomic-scale microstructures of Zr2Al3C4 and Zr3Al3C5 were investigated through high-resolution imaging and Z-contrast imaging. Furthermore, intergrowth between Zr2Al3C4 and Zr3Al3C5 was identified. Stacking faults in Zr3Al3C5 were found to result from the insertion of an additional Zr–C layer. Cubic ZrC was occasionally identified to be incorporated in elongated Zr3Al3C5 grains. In addition, Al may induce a twinned ZrC structure and lead to the formation of ternary zirconium aluminum carbides.}, number={14}, journal={Acta Materialia}, author={Lin, Z.J. and Zhuo, M.J. and He, L.F. and Zhou, Y.C. and Li, M.S. and Wang, J.Y.}, year={2006}, pages={3843–3851} }
 @article{wang_zhou_lin_liao_he_2006, title={First-principles prediction of the mechanical properties and electronic structure of ternary aluminum carbide Zr3 Al3 C5}, volume={73}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33645763036&partnerID=MN8TOARS}, DOI={10.1103/PhysRevB.73.134107}, abstractNote={In this paper, we predicted the possible mechanical properties and presented the electronic structure of Zr(3)Al(3)C(5) by means of first-principles pseudopotential total energy method. The equation of state, elastic parameters (including the full set of second order elastic coefficients, bulk and shear moduli, Young's moduli, and Poisson's ratio), and ideal tensile and shear strengths are reported and compared with those of the binary compound ZrC. Furthermore, the bond relaxation and bond breaking under tensile and shear deformation from elasticity to structural instability are illustrated. Because shear induced bond breaking occurs inside the NaCl-type ZrC(x) slabs, the ternary carbide is expected to have high hardness and strength, which are related to structural instability under shear deformation, similar to the binary carbide. In addition, mechanical properties are interpreted by analyzing the electronic structure and chemical bonding characteristics accompanying deformation paths. Based on the present results, Zr(3)Al(3)C(5) is predicted to be useful as a hard ceramic for high temperature applications.}, number={13}, journal={Physical Review B - Condensed Matter and Materials Physics}, author={Wang, J. and Zhou, Y. and Lin, Z. and Liao, T. and He, L.F.}, year={2006} }
 @inproceedings{harp_he_hoggan_wagner, title={Corrosion and Interdiffusion Studies of U3Si2}, booktitle={Top Fuel 2016}, author={Harp, J.M. and He, L. and Hoggan, R.E. and Wagner, A.R.}, pages={1341–1346} }