Lingfeng He

Profile Information
Name
Lingfeng He
Institution
Idaho National Laboratory
Position
Staff Scientist
Affiliation
Advanced Characterization Department
h-Index
ORCID
0000-0003-2763-1462
Expertise
Ceramics, Nuclear Fuel
Publications:
"A transmission electron microscopy study of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750" Lingfeng He, Journal of Nuclear Materials Vol. 528 2020 1-9 Link
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 precipitates found in several previous studies. In X-750, L12-structured precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni3(Ti,Al)) of the precipitates and the superlattice reflections confirmed the stability of the crystal structure of these precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.
"Bubble Character, Kr Distribution and Chemical Equilibrium in UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Hunter Henderson, Michele Manuel, Andrew Nelson, Janne Pakarinen, Billy Valderrama, Journal of Nuclear Materials Vol. 2015 Link
"Bubble evolution in Kr-irradiated UO2 during annealing" Lingfeng He, Xianming Bai, Janne Pakarinen, Brian Jaques, Jian Gan, Andrew Nelson, Anter EL-AZAB, Todd Allen, Journal of Nuclear Materials Vol. 496 2017 242-250 Link
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.
"Bubble formation and Kr distribution in Kr-irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Andrew Nelson, Janne Pakarinen, Billy Valderrama, Lingfeng He, Abdel-Rahman Hassan, Hunter Henderson, Marquis Kirk, Michele Manuel, Journal of Nuclear Materials Vol. 456 2015 125-132 Link
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.
"Bubble, stoichiometry, and chemical equilibrium of krypton-irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Michele Manuel, Janne Pakarinen, Billy Valderrama, Abdel-Rahman Hassan, Marquis Kirk, Andrew Nelson, Journal of Nuclear Materials Vol. 456 2015 125-132 Link
In situ and ex situ transmission electron microscopy observation of small Kr bubbles in bothsingle-crystal and polycrystalline UO2 were conducted to understand the inert gas bubblebehavior in oxide nuclear fuel. The bubble size and volume swelling are shown as a weakfunction of ion dose but strongly depend on the temperature. The Kr bubble formation at roomtemperature was observed for the first time. The depth profiles of implanted Kr determined byatom probe tomography are in good agreement with the calculated profiles by SRIM, but themeasured concentration of Kr is about 1/3 of calculated one. This difference is mainly due to lowsolubility of Kr in UO2 matrix, which has been confirmed by both density-functional theorycalculations and chemical equilibrium analysis.
"Determining Oxidation States of Transition Metals in Molten Salt Corrosion using Electron Energy Loss Spectroscopy" Lingfeng He, Kaustubh Bawane, Panayotis Manganaris, Yachun Wang, Scripta Materialia Vol. 197 [unknown] 113790 Link
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.
"Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91" Huan Yan, Xiang Liu, Lingfeng He, James Stubbins, Journal of Nuclear Materials Vol. 557 2021 Link
"Effect of dpa rate on the temperature regime of void swelling in ion-irradiated pure chromium" Adam Gabriel, Laura Hawkins, Aaron French, Yongchang Li, Zhihan Hu, Lingfeng He, Pengyuan Xiu, Michael Nastasi, Frank Garner, Lin Shao, JNM Vol. 561 2022 Link
"Effect of Grain Boundaries on Krypton Segregation Behavior in Irradiated Uranium Dioxide" Todd Allen, Darryl Butt, Jian Gan, Lingfeng He, Hunter Henderson, Brian Jaques, Michele Manuel, Janne Pakarinen, Billy Valderrama, Journal of Metals Vol. 66 2014 2562-2568 Link
Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregating toward grain boundaries and eventually leading to a lowering in thermal conductivity and fuel swelling. Recent computational studies have identi?ed that differences in grain boundary structure have a signi?cant effect on the segregation behavior of fission products. However, experimental work supporting these simulations is lacking. Atom probe tomography was used to measure the Kr distribution across grain boundaries in UO2. Polycrystalline depleted UO2 samples were irradiated with 0.7 MeV and 1.8 MeV Kr-ions and annealed to 1000C, 1300C, and 1600C for 1 h to produce a Kr-bubble dominated microstructure. The results of this work indicate a strong dependence of Kr concentration as a function of grain boundary structure. Temperature also influences grain boundary chemistry with greater Kr concentration evident at higher temperatures, resulting in a reduced Kr concentration in the bulk. Although Kr segregation takes place at elevated temperatures, no change in grain size or texture was observed in the irradiated UO2 samples.
"Effect of neutron irradiation on defect evolution in Ti3SiC2 and Ti2AlC" Michel Barsoum, Lingfeng He, Elizabeth Hoffman, Gordon Kohse, Darin Tallman, Brenda Garcia-Diaz, Robert Sindelar, Journal of Nuclear materials Vol. 468 2015 1-13 Link
Herein we report on the characterization of defects formed in polycrystalline Ti3SiC2 and Ti2AlC samples exposed to neutron irradiation e 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 ¼ 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 wasz1.5 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.
"Effect of proton pre-irradiation on corrosion of Zr-0.5Nb model alloys with different Nb distributions" Lingfeng He, Corrosion Science Volume 173 Vol. 173 2020 108790 Link
The effect of proton irradiation on corrosion rate of α-annealed and β-quenched Zr-0.5Nb alloys is investigated. The major focuses of this study are to understand i) if the nucleation of irradiation-induced platelets (IIPs)/nanoclusters requires dissolution of Nb-rich native precipitates, ii) if the irradiated native precipitates and interlaths are stable in the oxide, and iii) how much Nb content in the solid solution is suitable to lower the corrosion rate for Zr-Nb alloys. To answer these questions, the major characterization techniques used in this study are APT and (S)TEM/EDS to study the microstructure and microchemistry evolution following irradiation and oxidation.
"Effects of neutron irradiation of Ti3SiC2 and Ti3AlC2 in the 121-1085 C temperature range" Michel Barsoum, Jian Gan, Elizabeth Hoffman, Darin Tallman, Lingfeng He, El'ad Caspi, Journal of Nuclear Materials Vol. 484 2017 120-134 Link
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.
"Electron microscopy characterization of fast reactor MOX Joint Oxyde-Gaine (JOG)" Fabiola Cappia, Brandon Miller, Jeffery Aguiar, Lingfeng He, Daniel Murray, Brian Frickey, John Stanek, Jason Harp, Journal of Nuclear Materials Vol. 531 2020 Link
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.
"Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels" Andrew Hoffman, Maalavan Arivu, Haiming Wen, Li He, Kumar Sridharan, Xin Wang, Xiang Liu, Lingfeng He, Yaqiao Wu, Materialia Vol. 13 2020 Link
Irradiation induced phase transformation of γ-austenite to α-ferrite has been observed in austenitic steels for the past several decades. This transformation can be detrimental to structural materials in a nuclear reactor environment as the increased fraction of the ferritic phase can increase corrosion and embrittlement and lead to stress corrosion cracking. This transformation is caused by both strain induced martensite transformation as well as radiation induced segregation and precipitation. In this study, two radiation tolerant nanostructured 304L austenitic steels (one ultrafine grained and one nanocrystalline) were manufactured using severe plastic deformation. These nanostructured 304L steels were compared to conventional coarse-grained 304L, after self-ion irradiation at 500°C up to a peak damage of 50 displacements per atom. Phase fraction after irradiation was analyzed using grazing incidence x-ray diffraction, precession electron diffraction, and electron backscatter diffraction. Nanostructured 304L steels showed significant resistance to irradiation induced austenite to ferrite transformation. This resistance was shown to be due to a decrease in defect formation, as well as a reduction in radiation induced segregation and precipitation.
"Formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation" Cheng Sun, David Sprouster, Khalid Hattar, Lynne Ecker, Lingfeng He, Y. Gao, Yipeng Gao, Yongfeng Zhang, Jian Gan, Scripta Materialia Vol. 149 2018 26-30 Link
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.
"Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding" Xiang Liu, Luca Capriotti, Tiankai Yao, Jason Harp, Michael Benson, Yachun Wang, Fei Teng, Lingfeng He, https://www.sciencedirect.com/science/article/pii/S002231152031196X#ack0001 Vol. 544 2021 Link
"Fuel-cladding chemical interation of a prototype annular U-10Zr fuel with HT-9 cladding" Lingfeng He, Journal of Nuclear Materials Vol. 544 2021
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 Fdm). The (U,Zr)(Fe,Cr)2phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fmm). 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 U-rich 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.
"Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite" Lingfeng He, Carbon Vol. 166 2020 Pages 113-122
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.
"Hydrothermal synthesis of silicon oxide clad uranium oxide nanowires" Lingfeng He, Jason Harp, Adrian Wagner, Rita Hoggan, Kevin Tolman, Journal of the American Cermic Society Vol. 2018 1004-1008 Link
"Impact of krypton irradiation on a single crystal tungsten: Multi-modal X-ray imaging study" Lingfeng He, Scripta Materialia Vol. 188 2020 296-301
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.
"In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation" Michael Moorehead, Calvin Parkin, Mohamed Elbakhshwan, Jing Hu, Wei-Ying Chen, Meimei Li, Lingfeng He, Kumar Sridharan, Adrien Couet, Acta Materialia Vol. 198 2020 85-99 Link
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 mecha- nisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr- FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to vari- ous levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irra- diation 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 Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinet- ics 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.#171118
"In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation" Lingfeng He, Acta Materialia Vol. 198 2020 85-99
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.
"In Situ TEM Observation of Dislocation Evolution in Polycrystalline UO2" Todd Allen, Jian Gan, Mahima Gupta, Janne Pakarinen, Lingfeng He, Marquis Kirk, JOM Vol. 66 2014 2553-2561 Link
In situ transmission electron microscopy observation of polycrystalline UO2 (with average grain size of about 5 µm) irradiated with Kr ions at 600°C and 800°C was conducted to understand the radiation-induced dislocation evolution under the influence of grain boundaries. The dislocation evolution in the grain interior of polycrystalline UO2 was similar under Kr irradiation at different ion energies and temperatures. As expected, it was characterized by the nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation lines and tangles at high doses. For the first time, a dislocation-denuded zone was observed near a grain boundary in the 1-MeV Kr-irradiated UO2 sample at 800°C. The denuded zone in the vicinity of grain boundary was not found when the irradiation temperature was at 600°C. The suppression of dislocation loop formation near the boundary is likely due to the enhanced interstitial diffusion toward grain boundary at the high temperature.
"Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system" Yachun Wang, Xiang Liu, Daniel Murray, Fei Teng, Wen Jiang, Mukesh Bachhav, Laura Hawkins, Emmanuel Perez, Cheng Sun, Xianming Bai, Jie Lian, Colin Judge, John Jackson, Robert Carter, Lingfeng He, Materials Science & Engineering Vol. 849 2022
"Microstructural changes of proton irradiated Hastelloy-N and in situ micropillar compression testing of one single grain at different local damage levels" Miguel Pena, Andres Morell-Pacheco, Ching-Heng Shiau, Boopathy Kombaiah, Lingfeng He, Laura Hawkins, Adam Gabriel, Frank Garner, Lin Shao, JNM Vol. 2022 Link
"Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation" Anter EL-AZAB, Jian Gan, Marat Khafizov, Andrew Nelson, Janne Pakarinen, Chris Wetteland, Lingfeng He, David Hurley, Todd Allen, Journal of Nuclear Materials Vol. 454 2014 283-289 Link
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.
"Microstructure evolution in Xe-irradiated UO2 at room temperature" Todd Allen, Anter EL-AZAB, Jian Gan, Lingfeng He, Janne Pakarinen, Marquis Kirk, Andrew Nelson, Xianming Bai, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 330 2014 55-60 Link
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.
"On Spinodal-like Phase Decomposition in U-50 Zr Alloy" Lingfeng He, Materialia Vol. 9 2020 Link
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.
"Phase and Defect Evolution in Uranium-Nitrogen-Oxygen System under Irradiation" Lingfeng He, Acta Materialia Vol. 2021 Link
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.
"Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature" Bowen Gong, Jason Harp, Jie Lian, Tiankai Yao, Lingfeng He, Michael Tonks, Journal of Nuclear Materials Vol. 498 2018 169-175 Link
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.
"STEM-EDS/EELS and APT characterization of ZrN coatings on UMo fuel kernels" Lingfeng He, Mukesh Bachhav, Dennis Keiser, Emmanuel Perez, Brandon Miller, Jian Gan, Ann Leenaers, Sven Van den Berghe, Journal of Nuclear Materials Vol. 511 2018 174-182 Link
"Subsurface imaging of grain microstructure using picosecond ultrasonics" Darryl Butt, Hunter Henderson, David Hurley, Brian Jaques, Marat Khafizov, Andrew Nelson, Janne Pakarinen, Michele Manuel, Lingfeng He, Acta Materialia Vol. 112 2016 1476-1477 Link
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.
"The influence of lattice defects, recombination, and clustering on thermal transport in single crystal thorium dioxide" Lingfeng He, APL Materials Vol. 8 2020 Link
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
"Thermal stability of helium bubble superlattice in Mo under TEM in-situ heating" Jian Gan, Cheng Sun, Lingfeng He, Yongfeng Zhang, Chao Jiang, Yipeng Gao, Journal of Nuclear Materials Vol. 505 2018 207-211 Link
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.
Presentations:
"Advanced Characterization of Irradiated UO2 Fuel" Lingfeng He, Michael Moorehead, Brandon Miller, Jason Harp, Xianming Bai, TMS 2018 March 11-15, (2018)
"Comparison of Computationally Simulated Fission Product Distribution with Correlative Characterization Techniques in Surrogate Nuclear Fuel Materials" Todd Allen, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Billy Valderrama, 2013 SACNAS National Conference October 2-6, (2013)
"Electron microscopy characterization of fast reactor MOX joint-oxide-gaine (JOG)" Fabiola Cappia, Brandon Miller, Daniel Murray, Lingfeng He, Brian Frickey, John Stanek, Jason Harp, EMRS 2019 May 27-31, (2019)
"Fission Products in Nuclear Fuel: Comparison of Simulated Distribution with Correlative Characterization Techniques" Todd Allen, Anter EL-AZAB, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Billy Valderrama, Clarissa Yablinsky, Microscopy and Microanalysis August 4-8, (2012)
"Influence of Irradiation-induced Microstructural Defects on the Thermal Conductivity of Single Crystal Thorium Dioxide" Marat Khafizov, Amey Khanolkar, Zilong Hua, Cody Dennett, wangthink Wang, Tiankai Yao, Lingfeng He, Jian Gan, David Hurley, TMS 2020 February 23-27, (2020)
"Microstructural Characterization of High-entropy Alloy Ion Irradiated at Cryogenic Temperatures" Michael Moorehead, Calvin Parkin, Lingfeng He, Jing Hu, Meimei Li, Adrien Couet, Kumar Sridharan, TMS 2019 March 10-14, (2019)
"Microstructural Defects in Neutron Irradiated Ti3SiC2 and Ti2AlC" Michel Barsoum, Lingfeng He, Elizabeth Hoffman, Gordon Kohse, Darin Tallman, ICACC'15 January 25-29, (2015)
"Microstructural Investigation of Kr Irradiated UO2" Todd Allen, Jian Gan, Mahima Gupta, Lingfeng He, Clarissa Yablinsky, The Minerals, Materials, and Metals Society, 2013 Annual Meeting & Exhibition March 3-7, (2013)
"Microstructural Investigations of Kr and Xe Irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Hunter Henderson, Michele Manuel, Janne Pakarinen, Billy Valderrama, Energy Frontier Research Centers Principal Investigators Meeting July 18-19, (2013)
"Nano-scale Irradiation Induced Chemistry Changes in Oxide" Todd Allen, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Janne Pakarinen, Billy Valderrama, 2014 TMS Annual Meeting February 16-20, (2014)
"Neutron Irradiation of MAX Phases" Michel Barsoum, Lingfeng He, Elizabeth Hoffman, Gordon Kohse, Darin Tallman, Il Trovatore Meeting 2 July 23-25, (2015)
"Nuclear Scientific User Facility: Neutron Irradiation of MAX Phases" Michel Barsoum, Lingfeng He, Elizabeth Hoffman, Gordon Kohse, Darin Tallman, NSUF User's Week June 22-26, (2015)
"Poster - Examining microstructural differences in irradiated HT9 correlated with differences in processing prior to irradiation" Theresa Mary Green, Li He, Todd Allen, Brandon Miller, Lingfeng He, NUMAT 2018 October 15-18, (2018)
"Recent observations from the microstructural characterization of irradiated U-Mo fuels using advanced techniques" Dennis Keiser, Brandon Miller, Jian Gan, Lingfeng He, Daniel Jadernas, Mukesh Bachhav, NUMAT 2018 October 15-18, (2018)
NSUF Articles:
U.S. DOE Nuclear Science User Facilities Awards 30 Rapid Turnaround Experiment Research Proposals - Awards total nearly $1.2 million The U.S. Department of Energy (DOE) Nuclear Science User Facilities (NSUF) has selected 30 new Rapid Turnaround Experiment (RTE) projects, totaling up to approximately $1.2 million. These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Wednesday, April 26, 2017 - Calls and Awards
DOE Awards 33 Rapid Turnaround Experiment Research Proposals - Projects total approximately $1.5 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Monday, May 14, 2018 - Calls and Awards
DOE awards 39 RTE Projects - Projects total approximately $1.3 million Thursday, February 1, 2018 - Calls and Awards
NSUF Researcher Feature: Lingfeng He - An instrumentation scientist at INL, He works with users to characterize fuel and materials The TEM group leader for INL, He manages the two TEMs in INL's Irradiated Materials and Characterization Laboratory, and it continually making sure that his group's expertise is cutting edge, but also that users have access to that expertise via NSUF. Monday, October 7, 2019 - Newsletter, Researcher Highlight
NSUF awards 30 Rapid Turnaround Experiment proposals - Approximately $1.53M has been awarded. Tuesday, June 14, 2022 - Calls and Awards
NSUF Research Collaborations

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

APT and TEM study of redistribution of alloying elements in ZrNb alloys following proton irradiation: effects on in-reactor corrosion kinetics. - FY 2017 RTE 3rd Call, #1001

Atom probe and transmission electron microscopy studies on neutron irradiated FeCrMnNi Compositionally Complex Alloy - FY 2022 RTE 1st Call, #4459

Characterization of Neutron-Irradiated Zr-1Nb-O Using Scanning Transmission Electron Microscopy - FY 2018 RTE 2nd Call, #1374

ChemisSTEM Characterization of Bulk Heavy Ion Irradiated Complex Concentrated Alloys - FY 2020 RTE 2nd Call, #4095

Dislocation loop and bubble evolution in helium irradiated ThO2 and UO2 single crystals - FY 2022 RTE 1st Call, #4442

Electron microscopy characterization of fast reactor MOX joint oxyde-gaine (JOG) - FY 2018 RTE 3rd Call, #1538

Electron Tomography for Three-Dimensional Characterization of Intragranular Fission Product Transport in Neutron-Irradiated Silicon Carbide in TRISO Fuel - FY 2018 RTE 1st Call, #1151

EPMA and TEM Characterization of a UO2 fuel pellet and cladding interaction layer - FY 2019 RTE 2nd Call, #1796

Fission Gas Behavior and Fuel Swelling of Accident Tolerant U3Si2 Fuels by Ion Beam Irradiation - FY 2017 RTE 2nd Call, #957

Grain Size Effects on He and Xe behavior in UO2 and ThO2 - FY 2018 RTE 3rd Call, #1561

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

In-situ Micro-tensile Testing for Measuring Grain Boundary Strength of NiCr Alloys under Simultaneous Irradiation and Corrosion Environments - FY 2020 RTE 1st Call, #3028

In-situ separate effect studies of thermal and radiation effects on Xe diffusion in alpha-U and U-10Zr. - FY 2019 RTE 1st Call, #1621

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

Investigation of gas bubble behavior in metals using in-situ Ne, Ar and Kr ion irradiation - FY 2018 RTE 1st Call, #1213

Investigation of the mechanism behind irradiation-decelerated corrosion of Ni-20Cr in molten fluoride salt - FY 2019 RTE 2nd Call, #1742

Ion Irradiation and TEM Characterization of Polymer Derived C-SiC-SiOC Nanocomposites - FY 2020 RTE 1st Call, #3004

Ion irradiation of ThO2 and UO2 single crystals - FY 2019 RTE 1st Call, #1663

Irradiation and TEM Characterization of induced Defects in a-U and d-UZr2+x Crystals - FY 2019 RTE 2nd Call, #1784

Microstructural characterization of grain boundaries in Hastelloy N corroded in molten FLiBe salt under neutron irradiation - FY 2018 RTE 3rd Call, #1545

Microstructural characterizations of in-core molten salt irradiated TRISO particles - FY 2019 RTE 2nd Call, #1785

Microstructure characterization of 6Li(n,a)3H reaction damage sapphire claddings - FY 2020 RTE 1st Call, #3018

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

Radiation Response and Microstructure of Accident Tolerant U3Si2 Fuels by Ion Beam Irradiation - FY 2017 RTE 1st Call, #835

TEM and APT Characterization of Ion-Irradiated High-Entropy Alloys for Sodium-Cooled Fast Reactors - FY 2018 RTE 2nd Call, #1380

TEM/EDS study of Nb redistribution in ZrNb alloys following proton irradiation - FY 2018 RTE 2nd Call, #1392

The effects of stress on void superlattice formation during Cr+ self-ion-irradiation of chromium - FY 2021 RTE 1st Call, #4320

The influence of proton irradiation damage on the corrosion of Hastelloy N exposed to FliNaK molten salt - FY 2019 RTE 3rd Call, #2833

The window of gas-bubble superlattice formation in bcc metals - FY 2017 RTE 1st Call, #846

Thermal Driven Grain Growth and Fission Gas Bubble Coarsening in Nano-grain Sized U3Si2 - FY 2019 RTE 1st Call, #1691

UCl4/UCl3 speciation - FY 2020 RTE 2nd Call, #3084

Understand the atomic positions of the metallic fission product in UCO Fuel Kernels and Determine the exact stoichiometry of UC, UO phase of Irradiated TRISO Fuel Particles by Using Titan Themis 200 with EELS characterization Capability - FY 2019 RTE 2nd Call, #1779

Understand the Fission Products Behavior and Irradiation Effects in UCO Fuel Kernels of Irradiated AGR-1 and AGR-2 TRISO Fuel Particles Using Titan Themis 200 with ChemiSTEM Capability - FY 2018 RTE 1st Call, #1257

Understand the Fission Products Behavior in UCO Fuel Kernels of safety tested AGR2 TRISO Fuel Particles by Using Titan Themis 200 with ChemiSTEM Capability - FY 2019 RTE 3rd Call, #2893

Understanding the role of grain boundary character in segregation behavior of solute elements in neutron irradiated 304 SS using Atom Probe Tomography. - FY 2018 RTE 1st Call, #1253