Lingfeng He

Comparison of microstructural and micro-chemical evolutions in the irradiated fuel kernels of AGR-1 and AGR-2 TRISO fuel particles

PROGRESS IN NUCLEAR ENERGY, 176.

Compositionally complex carbide ceramics: A perspective on irradiation damage

JOURNAL OF APPLIED PHYSICS, 135(20).

Dislocation channel broadening-A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel

Yang, J., Hawkins, L., Shang, Z., Tsai, B. K., He, L., Lu, Y., … Lou, X. (2024, March 1). ACTA MATERIALIA, Vol. 266.

Elucidating the Transition of 3D Morphological Evolution of Binary Alloys in Molten Salts with Metal Ion Additives

ACS APPLIED MATERIALS & INTERFACES, 16(34), 45606–45618.

Evolution of extended defects in UO₂ during high temperature annealing

JOURNAL OF NUCLEAR MATERIALS, 593.

Selective laser sintering and spark plasma sintering of (Zr,Nb,Ta,Ti,W)C compositionally complex carbide ceramics

Trinh, L., Hua, Z., Bawane, K., He, L., Malakkal, L., Chen, X., … Cui, B. (2024, July 23). JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol. 7.

Evolution of dislocation loops and voids in post-irradiation annealed ThO₂: A combined <i>in-situ</i> TEM and cluster dynamics investigation

JOURNAL OF NUCLEAR MATERIALS, 586.

Intragranular irradiation-assisted stress corrosion cracking (IASCC) of 316L stainless steel made by laser direct energy deposition additive manufacturing: Delta ferrite-dislocation channel interaction

JOURNAL OF NUCLEAR MATERIALS, 577.

Temperature-Dependent Morphological Evolution during Corrosion of the Ni-20Cr Alloy in Molten Salt Revealed by Multiscale Imaging

ACS APPLIED MATERIALS & INTERFACES, 15(10), 13772–13782.

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

JOURNAL OF NUCLEAR MATERIALS, 587.

Transmission electron microscopy investigation of phase transformation and fuel constituent redistribution in neutron irradiated U-10wt.%Zr fuel

JOURNAL OF NUCLEAR MATERIALS, 581.

Void swelling in additively manufactured 316L stainless steel with hafnium composition gradient under self-ion irradiation

Journal of Nuclear Materials, 578, 154351.

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

Acta Materialia, 241, 118379.

Assessing the interfacial corrosion mechanism of Inconel 617 in chloride molten salt corrosion using multi-modal advanced characterization techniques

Frontiers in Nuclear Engineering, 1.

Comprehensive characterization of irradiation induced defects in ceria: Impact of point defects on vibrational and optical properties

Journal of Applied Physics.

Dislocation loop evolution in Kr‐irradiated ThO2

Journal of the American Ceramic Society.

Effect of dpa rate on the temperature regime of void swelling in ion-irradiated pure chromium

Journal of Nuclear Materials, 561, 153519.

Electrochemical noise studies on localized corrosion of Ni and Ni-20Cr in molten ZnCl2

Electrochimica Acta, 431, 141126.

In situ</i> TEM Observations of Thermally Activated Phenomena in Materials Under Far-From-Equilibrium Conditions

Microscopy and Microanalysis, 28(S1), 1844–1846.

Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system

Materials Science and Engineering: A, 849, 143475.

Microstructural changes of proton irradiated Hastelloy-N and in situ micropillar compression testing of one single grain at different local damage levels

Journal of Nuclear Materials.

Phase stability, mechanical properties, and ion irradiation effects in face-centered cubic CrFeMnNi compositionally complex solid-solution alloys at high temperatures

Journal of Nuclear Materials, 565, 153733.

Real-time, On-Microscope Automated Quantification of Features in Microcopy Experiments Using Machine Learning and Edge Computing

Microscopy and Microanalysis, 28(S1), 2046–2048.

STEM/EDS and APT study on the microstructure and microchemistry of neutron irradiated ZIRLOTM

Journal of Nuclear Materials, 573, 154139.

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Journal of Nuclear Materials, 565, 153728.

Unraveling small-scale defects in irradiated ThO2 using kinetic Monte Carlo simulations

Scripta Materialia.

Advanced Characterization of Additively Manufactured 316L Stainless Steel for Nuclear Applications

Microscopy and Microanalysis, 27(S1), 2160–2161.

An integrated experimental and computational investigation of defect and microstructural effects on thermal transport in thorium dioxide

Acta Materialia, 213, 116934.

Chemical and elemental mapping of spent nuclear fuel sections by soft X-ray spectromicroscopy

Journal of Synchrotron Radiation, 29(1), 67–79.

Compositionally graded specimen made by laser additive manufacturing as a high-throughput method to study radiation damages and irradiation-assisted stress corrosion cracking

Journal of Nuclear Materials, 560, 153493.

Determining oxidation states of transition metals in molten salt corrosion using electron energy loss spectroscopy

Scripta Materialia, 197, 113790.

Dislocation Loops in Proton Irradiated Uranium-Nitrogen-Oxygen System

Journal of Nuclear Materials, 557, 153244.

Early-stage microstructural evolution and phase stability in neutron-irradiated ferritic-martensitic steel T91

Journal of Nuclear Materials, 557, 153207.

Effects of heat treatment on corrosion fatigue and stress corrosion crack growth of additive-manufactured Alloy 800H in high-temperature water

Corrosion Science, 191, 109739.

First principle studies of effects of solute segregation on grain boundary strength in Ni-based alloys

Journal of Alloys and Compounds, 874, 159795.

In situ monitoring of microstructure evolution during thermal processing of uranium-zirconium alloys using laser-generated ultrasound

Journal of Nuclear Materials, 4, 153005.

Indirect characterization of point defects in proton irradiated ceria

Materialia, 15, 101019.

Isotopic Analysis of Irradiated Ceramic Fuel for Burnup and Microchemical Assessment Using Atom Probe Tomography.

Microscopy and Microanalysis, 27(S1), 416–417.

Machine Learning Based Precision Orientation and Strain Mapping from 4D Diffraction Datasets

Microscopy and Microanalysis, 27(S1), 1276–1278.

Metallic Fast Reactor Separate Effect Studies for Fuel Safety

Journal of Nuclear Engineering and Radiation Science, 7(4).

Nanoscale redistribution of alloying elements in high-burnup AXIOM-2 (X2®) and their effects on in-reactor corrosion

Corrosion Science, 190, 109652.

Phase and defect evolution in uranium-nitrogen-oxygen system under irradiation

Acta Materialia.

Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9

Journal of Nuclear Materials, 557, 153252.

Preparation and strengthening mechanism of prestressed ceramic tile components

International Journal of Applied Ceramic Technology.

Proton irradiation-induced blistering in UO2

MRS Advances, 11.

Sensitization, desensitization, and carbide evolution of Alloy 800H made by laser powder bed fusion

Additive Manufacturing, 50, 102547.

Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding

Journal of Nuclear Materials, 556, 153196.

TEM Characterization of Dislocation Loops in Proton Irradiated Single Crystal ThO2

Journal of Nuclear Materials, 4, 152998.

Thermal Energy Transport in Oxide Nuclear Fuel

Chemical Reviews.

Thermal conductivity of ThO2: Effect of point defect disorder

Journal of Applied Physics.

Training artificial neural networks for precision orientation and strain mapping using 4D electron diffraction datasets

Ultramicroscopy, 3, 113256.

Understanding spinodal and binodal phase transformations in U-50Zr

Materialia, 16, 101092.

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

Corrosion Science, 195, 109962.

Atom Probe Tomography for Burnup and Fission Product Analysis for Nuclear Fuels

Microscopy and Microanalysis, 26(S2), 3086–3088.

Combining mesoscale thermal transport and x-ray diffraction measurements to characterize early-stage evolution of irradiation-induced defects in ceramics

Acta Materialia, 193, 61–70.

Degradation mechanism of lead-vanado-iodoapatite in NaCl solution

Corrosion Science, 172, 108720.

Development of a grain growth model for U3Si2 using experimental data, phase field simulation and molecular dynamics

Journal of Nuclear Materials, 532, 152069.

Dual Functional Ni₃S₂@Ni Core–Shell Nanoparticles Decorating Nanoporous Carbon as Cathode Scaffolds for Lithium–Sulfur Battery with Lean Electrolytes

ACS Applied Energy Materials, 3(5), 4173–4179.

Effect of proton pre-irradiation on corrosion of Zr-0.5Nb model alloys with different Nb distributions

Corrosion Science, 173, 108790.

Electron microscopy characterization of fast reactor MOX Joint Oxyde-Gaine (JOG)

Journal of Nuclear Materials, 531, 151964.

Enhanced Resistance to Irradiation Induced Ferritic Transformation in Nanostructured Austenitic Steels

Materialia.

Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

Journal of Nuclear Materials, 544, 152588.

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Carbon, 166, 113–122.

Impact of krypton irradiation on a single crystal tungsten: Multi-modal X-ray imaging study

Scripta Materialia, 188, 296–301.

In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

Acta Materialia, 198, 85–99.

Intragranular thermal transport in U–50Zr

Journal of Nuclear Materials, 534, 152145.

Investigating corrosion behavior of Ni and Ni-20Cr in molten ZnCl2

Corrosion Science, 179, 109105.

On spinodal-like phase decomposition in U–50Zr alloy

Materialia.

Revealing 3D Morphological and Chemical Evolution Mechanisms of Metals in Molten Salt by Multimodal Microscopy

ACS Applied Materials & Interfaces, 12(15), 17321–17333.

TEM and APT Characterization of Neutron Irradiated AXIOM-2(X2)

Presented at the 2020 ANS Virtual Winter Meeting.

The correlation between microstructure and nanoindentation property of neutron-irradiated austenitic alloy D9

Acta Materialia, 195, 433–445.

The influence of lattice defects, recombination, and clustering on thermal transport in single crystal thorium dioxide

APL Materials.

α-U and ω-UZr2 in neutron irradiated U-10Zr annular metallic fuel

Journal of Nuclear Materials, 542, 152536.

A simple way to make pre-stressed ceramics with high strength

Journal of Materiomics, 5(4), 657–662.

A transmission electron microscopy study of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750

Journal of Nuclear Materials, 528, 151851.

Characterization of U-Pu-Zr alloys with Pd as a minor additive

Transactions of the American Nuclear Society, 120.

Diffusion Behavior between Metallic Fuel Alloys with Pd Addition and Fe

Journal of Nuclear Materials, 525, 111–124.

Diffusion behaviors between Fe and Pd-containing metallic fuel

Transactions of the American Nuclear Society, 120(1), 412–413.

Experimental evidence for ‘buckle, ruck and tuck’ in neutron irradiated graphite

Carbon, 159, 119–121.

Impact of irradiation induced dislocation loops on thermal conductivity in ceramics

Journal of the American Ceramic Society, 102(12), 7533–7542.

Irradiation Testing Strategy for Commercialization of a Molten Chloride Fast Reactor System

Microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln

Journal of Nuclear Materials, 518, 287–297.

Microstructure and microchemistry study of irradiation-induced precipitates in proton irradiated ZrNb alloys

Acta Materialia.

Self-healing behavior and strength recovery of ytterbium disilicate ceramic reinforced with silicon carbide nanofillers

Journal of the European Ceramic Society, 39(10), 3139–3152.

Small Scale Tensile Testing of Grain Boundary Strength of X-750 Alloy

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Presented at the 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Boston, MA.

Challenges and Opportunities on Elucidating Irradiated Fuels with Atom Probe Tomography

Microscopy and Microanalysis, 24(S1), 2206–2207.

Evaluation of U3Si2 Fuel Pellets Sintered in an Argon vs. Vacuum Environment

Ceramic Transactions Series, 10, 21–26.

Formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation

Scripta Materialia.

Formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion implantation

Scripta Materialia, 149, 26–30.

In-situ TEM study of the ion irradiation behavior of U3Si2 and U3Si5

Journal of Nuclear Materials, 511, 56–63.

Investigating Pd as a Potential Fuel Additive in a Transmutation Fuel to Control FCCI

Transactions of the American Nuclear Society, 118, 1379–1380.

Microstructural characterization of annealed U-12Zr-4Pd and U-12Zr-4Pd-5Ln: Investigating Pd as a metallic fuel additive

Journal of Nuclear Materials.

Microstructural characterization of as-cast U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln

Journal of Nuclear Materials, 508, 310–318.