Lingfeng He - Academia.edu (original) (raw)
Papers by Lingfeng He
APL Materials, 2020
Thermal transport is a key performance metric for thorium dioxide in many applications where defe... more 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 of Nuclear Materials, 2021
In this study, we investigated the dislocation loop types formed in the proton-irradiated uranium... more In this study, we investigated the dislocation loop types formed in the proton-irradiated uraniumnitrogen-oxygen (U-NO) system, which involves uranium mononitride (UN), uranium sesquinitride (-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 transmissionbased 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,-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in UN phases indicate a critical loop size of ~ 6 nm, above which perfect loops are energetically 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 nitrides as advanced nuclear fuels for the next-generation nuclear reactors.
Journal of Nuclear Engineering and Radiation Science, 2021
Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commerci... more Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commercially applied for electricity production. The benefits of SFR are well-known and include: the possibility of a closed fuel cycle, proliferation resistance, nuclear waste minimization via actinides burning, and fissile breeding capabilities. Metallic fuel used in SFR has well demonstrated irradiation performance. However, more studies are necessary to optimize and extend operational and safety limits for their commercialization and licensing. This could be achieved through a better understanding of fuel behaviors during transient and of fuel failure thresholds. This paper describes the experimental Research and Development (R&D) program aimed at providing the necessary data to support the development of SFR-optimized safety limits. This program integrates separate effects testing and integral effects testing, combined with advanced Modeling and Simulation (M&S). This R&D program, finally, ...
Acta Materialia, 2019
Proton irradiation induced Nb redistribution in Zr-xNb alloys (x ¼ 0.4, 0.5, 1.0 wt%) has been in... more Proton irradiation induced Nb redistribution in Zr-xNb alloys (x ¼ 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 ZreNbeFe phases, and b-Nb precipitates. After 2 MeV proton irradiation at 350 C, a decrease of Nb content in native precipitates, as well as irradiationinduced 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, [111]//[2110] 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.
Scripta Materialia, 2018
We report the formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion... more 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 of Nuclear Materials, 2017
Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 annea... more Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 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 of Nuclear Materials, 2018
U 3 Si 2 , an advanced fuel form proposed for light water reactors (LWR), has excellent thermal c... more U 3 Si 2 , an advanced fuel form proposed for light water reactors (LWR), has excellent thermal conductivity and a higher fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U 3 Si 2 is available at LWR conditions. This study explores the irradiation behavior of U 3 Si 2 by 300 KeV Xe + ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U 3 Si 2 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 U 3 Si 2 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 an 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 U 3 Si 2 irradiated at 64 dpa considering fission gas only.
Journal of Nuclear Materials, 2017
Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions wh... more 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 U 3 Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000C up to 100 hours. The microstructure of pristine U 3 Si 2 and U 3 Si 2 /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 800C
Nuclear Technology, 2013
To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycryst... more To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycrystalline CeO 2 , composed of both nanograins and micrograins, as a surrogate material for UO 2. The CeO 2 was implanted with 150-keV Kr ions up to a dose of 1 ϫ 10 16 ions/cm 2 at 6008C. Transmission electron microscopy characterizations of small Kr bubbles in nanograin and micrograin regions were compared. The grain boundary acted as an efficient defect sink, as evidenced by smaller bubbles and a lower bubble density in the nanograin region as compared to the micrograin region.
Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2014
In situ Transmission Electron Microscopy was conducted for single crystal UO 2 to understand the ... more In situ Transmission Electron Microscopy was conducted for single crystal UO 2 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 UO 2 remained stoichiometric under room temperature Xe irradiation.
Journal of Nuclear Materials, 2015
In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both sing... more In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO 2 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 UO 2 matrix and high release of Kr from sample surface under irradiation.
JOM, 2014
Fission products, such as krypton (Kr), are known to be insoluble within UO 2 , segregating towar... more Fission products, such as krypton (Kr), are known to be insoluble within UO 2 , segregating toward grain boundaries and eventually leading to a lowering in thermal conductivity and fuel swelling. Recent computational studies have identified that differences in grain boundary structure have a significant 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 UO 2. Polycrystalline depleted UO 2 samples were irradiated with 0.7 MeV and 1.8 MeV Kr-ions and annealed to 1000°C, 1300°C, and 1600°C 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 UO 2 samples.
Journal of Nuclear Materials, 2014
The microstructural changes and associated effects on thermal conductivity were examined in UO 2 ... more The microstructural changes and associated effects on thermal conductivity were examined in UO 2 after irradiation using 3.9 MeV He 2+ ions. Lattice expansion of UO 2 was observed in x-ray diffraction after ion irradiation up to 5×10 16 He 2+ /cm 2 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 5×10 16 He 2+ /cm 2 , the highest fluence reached, while similar features were not detected at 9×10 15 He 2+ /cm 2. 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.
Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2014
The effect of low temperature proton irradiation in depleted uranium dioxide was examined as a fu... more 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 Â 10 17 protons/cm 2 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 UO 2. 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 of Nuclear Materials, 2013
In situ transmission electron microscopy (TEM) observation of UO 2 single crystal irradiated with... more In situ transmission electron microscopy (TEM) observation of UO 2 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 UO 2 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 UO 2 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 h1 1 0i were observed in the Kr-irradiated UO 2. The irradiated specimens were denuded of dislocation loops near the surface.
MRS Advances, 2021
Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was car... more Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO2 surfaces revealed that the implanted H+ formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract
Acta Materialia, 2021
Abstract Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to ... more Abstract 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.
Optics and Lasers in Engineering, 2012
Applications of the digital image correlation method (DIC) for the determination of the mixed-mod... more Applications of the digital image correlation method (DIC) for the determination of the mixed-mode stress intensity factors (SIF) is investigated in this paper. Experiments were performed on an edge fatigue cracked aluminum specimen using a special loading device, which is an appropriate apparatus for experimental mixed-mode fracture analysis. The full-field displacements around the crack-tip region of the test sample were calculated using DIC. And then the SIF associated with unavoidable rigidbody displacement motion were calculated simultaneously from the experimental data. The effect of the rigid body motion on the measured displacements was then eliminated using the computed rigid body translation and rotation. A coarse-fine searching method was developed to determine the cracktip location. For validation, the SIF thus determined is compared with theoretical results, confirming the effectiveness and accuracy of the proposed technique. Therefore it reveals that the DIC is a practical and effective tool for full-field deformation and SIF measurement.
JOM, 2014
ABSTRACT Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregatin... more ABSTRACT 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 identified that differences in grain boundary structure have a significant 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.
APL Materials, 2020
Thermal transport is a key performance metric for thorium dioxide in many applications where defe... more 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 of Nuclear Materials, 2021
In this study, we investigated the dislocation loop types formed in the proton-irradiated uranium... more In this study, we investigated the dislocation loop types formed in the proton-irradiated uraniumnitrogen-oxygen (U-NO) system, which involves uranium mononitride (UN), uranium sesquinitride (-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 transmissionbased 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,-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in UN phases indicate a critical loop size of ~ 6 nm, above which perfect loops are energetically 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 nitrides as advanced nuclear fuels for the next-generation nuclear reactors.
Journal of Nuclear Engineering and Radiation Science, 2021
Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commerci... more Sodium-cooled Fast Reactors (SFR) are one of the advanced nuclear reactor concepts to be commercially applied for electricity production. The benefits of SFR are well-known and include: the possibility of a closed fuel cycle, proliferation resistance, nuclear waste minimization via actinides burning, and fissile breeding capabilities. Metallic fuel used in SFR has well demonstrated irradiation performance. However, more studies are necessary to optimize and extend operational and safety limits for their commercialization and licensing. This could be achieved through a better understanding of fuel behaviors during transient and of fuel failure thresholds. This paper describes the experimental Research and Development (R&D) program aimed at providing the necessary data to support the development of SFR-optimized safety limits. This program integrates separate effects testing and integral effects testing, combined with advanced Modeling and Simulation (M&S). This R&D program, finally, ...
Acta Materialia, 2019
Proton irradiation induced Nb redistribution in Zr-xNb alloys (x ¼ 0.4, 0.5, 1.0 wt%) has been in... more Proton irradiation induced Nb redistribution in Zr-xNb alloys (x ¼ 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 ZreNbeFe phases, and b-Nb precipitates. After 2 MeV proton irradiation at 350 C, a decrease of Nb content in native precipitates, as well as irradiationinduced 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, [111]//[2110] 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.
Scripta Materialia, 2018
We report the formation of tetragonal gas bubble superlattice in bulk molybdenum under helium ion... more 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 of Nuclear Materials, 2017
Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 annea... more Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 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 of Nuclear Materials, 2018
U 3 Si 2 , an advanced fuel form proposed for light water reactors (LWR), has excellent thermal c... more U 3 Si 2 , an advanced fuel form proposed for light water reactors (LWR), has excellent thermal conductivity and a higher fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U 3 Si 2 is available at LWR conditions. This study explores the irradiation behavior of U 3 Si 2 by 300 KeV Xe + ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U 3 Si 2 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 U 3 Si 2 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 an 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 U 3 Si 2 irradiated at 64 dpa considering fission gas only.
Journal of Nuclear Materials, 2017
Fuel swelling during normal reactor operations could lead to unfavorable chemical interactions wh... more 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 U 3 Si 2 and Zricaloy-4 (Zry-4) were conducted at 800 and 1000C up to 100 hours. The microstructure of pristine U 3 Si 2 and U 3 Si 2 /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 800C
Nuclear Technology, 2013
To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycryst... more To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycrystalline CeO 2 , composed of both nanograins and micrograins, as a surrogate material for UO 2. The CeO 2 was implanted with 150-keV Kr ions up to a dose of 1 ϫ 10 16 ions/cm 2 at 6008C. Transmission electron microscopy characterizations of small Kr bubbles in nanograin and micrograin regions were compared. The grain boundary acted as an efficient defect sink, as evidenced by smaller bubbles and a lower bubble density in the nanograin region as compared to the micrograin region.
Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2014
In situ Transmission Electron Microscopy was conducted for single crystal UO 2 to understand the ... more In situ Transmission Electron Microscopy was conducted for single crystal UO 2 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 UO 2 remained stoichiometric under room temperature Xe irradiation.
Journal of Nuclear Materials, 2015
In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both sing... more In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO 2 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 UO 2 matrix and high release of Kr from sample surface under irradiation.
JOM, 2014
Fission products, such as krypton (Kr), are known to be insoluble within UO 2 , segregating towar... more Fission products, such as krypton (Kr), are known to be insoluble within UO 2 , segregating toward grain boundaries and eventually leading to a lowering in thermal conductivity and fuel swelling. Recent computational studies have identified that differences in grain boundary structure have a significant 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 UO 2. Polycrystalline depleted UO 2 samples were irradiated with 0.7 MeV and 1.8 MeV Kr-ions and annealed to 1000°C, 1300°C, and 1600°C 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 UO 2 samples.
Journal of Nuclear Materials, 2014
The microstructural changes and associated effects on thermal conductivity were examined in UO 2 ... more The microstructural changes and associated effects on thermal conductivity were examined in UO 2 after irradiation using 3.9 MeV He 2+ ions. Lattice expansion of UO 2 was observed in x-ray diffraction after ion irradiation up to 5×10 16 He 2+ /cm 2 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 5×10 16 He 2+ /cm 2 , the highest fluence reached, while similar features were not detected at 9×10 15 He 2+ /cm 2. 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.
Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms, 2014
The effect of low temperature proton irradiation in depleted uranium dioxide was examined as a fu... more 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 Â 10 17 protons/cm 2 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 UO 2. 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 of Nuclear Materials, 2013
In situ transmission electron microscopy (TEM) observation of UO 2 single crystal irradiated with... more In situ transmission electron microscopy (TEM) observation of UO 2 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 UO 2 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 UO 2 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 h1 1 0i were observed in the Kr-irradiated UO 2. The irradiated specimens were denuded of dislocation loops near the surface.
MRS Advances, 2021
Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was car... more Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO2 surfaces revealed that the implanted H+ formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract
Acta Materialia, 2021
Abstract Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to ... more Abstract 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.
Optics and Lasers in Engineering, 2012
Applications of the digital image correlation method (DIC) for the determination of the mixed-mod... more Applications of the digital image correlation method (DIC) for the determination of the mixed-mode stress intensity factors (SIF) is investigated in this paper. Experiments were performed on an edge fatigue cracked aluminum specimen using a special loading device, which is an appropriate apparatus for experimental mixed-mode fracture analysis. The full-field displacements around the crack-tip region of the test sample were calculated using DIC. And then the SIF associated with unavoidable rigidbody displacement motion were calculated simultaneously from the experimental data. The effect of the rigid body motion on the measured displacements was then eliminated using the computed rigid body translation and rotation. A coarse-fine searching method was developed to determine the cracktip location. For validation, the SIF thus determined is compared with theoretical results, confirming the effectiveness and accuracy of the proposed technique. Therefore it reveals that the DIC is a practical and effective tool for full-field deformation and SIF measurement.
JOM, 2014
ABSTRACT Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregatin... more ABSTRACT 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 identified that differences in grain boundary structure have a significant 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.