[Show abstract][Hide abstract] ABSTRACT: Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 1019 ions/m2 (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite–austenite phase boundary and presence of M23C6 carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M23C6 carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M23C6 carbides at 350 °C and 400 °C.
[Show abstract][Hide abstract] ABSTRACT: ZnO nanowires (NWs) have been extensively studied for various device applications. Although these nanowires are often suspected to be impractical and highly unstable under hostile radiation environments, to date little is known on their radiation tolerance. Here, we show outstanding resilience of ZnO NWs by using in situ Kr ion irradiation at room temperature inside a transmission electron microscope. Our studies show that ZnO nanowires with certain diameters become nearly immune to radiation damage due to the existence of dislocation loop denuded zones. A remarkable size effect also holds: the smaller the nanowire diameter, the lower the defect density. Rate theory modeling suggests that the size effect arises from fast interstitial migration and a limit in size to which interstitial loops can grow. In situ studies also revealed a surprising phenomenon: the pristine prismatic loops can prevail over the strongest known defect sinks, free surfaces, to trap radiation-induced defect clusters. This study comprises the first critical step toward in-depth understanding of radiation response of functional oxide nanowires for electronic device applications in extreme environments.
[Show abstract][Hide abstract] ABSTRACT: We report here the microstructural changes occurring in the zirconium alloy Excel (Zr-3.5 wt% Sn-0.8Nb-0.8Mo-0.2Fe) during heavy ion irradiation. In situ irradiation experiments were conducted at reactor operating temperatures on two Zr Excel alloy microstructures with different states of alloying elements, with the states achieved by different solution heat treatments. In the first case, the alloying elements were mostly concentrated in the beta (β) phase, whereas, in the second case, large Zr3(Mo,Nb,Fe)4 secondary phase precipitates (SPPs) were grown in the alpha (α) phase by long term aging. The heavy ion induced damage and resultant compositional changes were examined using transmission electron microscopy (TEM) in combination with scanning transmission electron microscope (STEM)-energy dispersive x-ray spectroscopy (EDS) mapping. Significant differences were seen in microstructural evolution between the two different microstructures that were irradiated under similar conditions. Nucleation and growth of -component loops and their dependence on the alloying elements are a major focus of the current investigation. It was observed that the -component loops nucleate readily at 100, 300, and 400 °C after a threshold incubation dose (TID), which varies with irradiation temperature and the state of alloying elements. It was found that the TID for the formation of -component loops increases with decrease in irradiation temperature. Alloying elements that are present in the form of SPPs increase the TID compared to when they are in the β phase solid solution. Dose and temperature dependence of loop size and density are presented. Radiation induced redistribution and clustering of alloying elements (Sn, Mo, and Fe) have been observed and related to the formation of -component loops. It has been shown that at the higher temperature tests, irradiation induced dissolution of precipitates occurs whereas irradiation induced amorphization occurs at 100 °C. Furthermore, dose and temperature seem to be the main factors governing the dissolution of SPPs and redistribution of alloying elements, which in turn controls the nucleation and growth of -component loops. The correlation between the microstructural evolution and microchemistry has been found by EDS and is discussed in detail.
Journal of Materials Research 05/2015; 30. DOI:10.1557/jmr.2015.89 · 1.65 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Material performance in extreme radiation environments is central to the design of future nuclear reactors. Radiation induces significant damage in the form of dislocation loops and voids in irradiated materials, and continuous radiation often leads to void growth and subsequent void swelling in metals with low stacking fault energy. Here we show that by using in situ heavy ion irradiation in a transmission electron microscope, pre-introduced nanovoids in nanotwinned Cu efficiently absorb radiation-induced defects accompanied by gradual elimination of nanovoids, enhancing radiation tolerance of Cu. In situ studies and atomistic simulations reveal that such remarkable self-healing capability stems from high density of coherent and incoherent twin boundaries that rapidly capture and transport point defects and dislocation loops to nanovoids, which act as storage bins for interstitial loops. This study describes a counterintuitive yet significant concept: deliberate introduction of nanovoids in conjunction with nanotwins enables unprecedented damage tolerance in metallic materials.
[Show abstract][Hide abstract] ABSTRACT: Using in-situ transmission electron microscopy, we have directly observed
nano-scale defects formed in ultra-high purity tungsten by low-dose high energy
self-ion irradiation at 30K. At cryogenic temperature lattice defects have
reduced mobility, so these microscope observations offer a window on the
initial, primary damage caused by individual collision cascade events. Electron
microscope images provide direct evidence for a power-law size distribution of
nano-scale defects formed in high-energy cascades, with an upper size limit
independent of the incident ion energy, as predicted by Sand et al. [Eur. Phys.
Lett., 103:46003, (2013)]. Furthermore, the analysis of pair distribution
functions of defects observed in the micrographs shows significant
intra-cascade spatial correlations consistent with strong elastic interaction
between the defects.
[Show abstract][Hide abstract] ABSTRACT: We describe aspects of transmission electron microscopy (TEM) technique to image and quantify the defect state following neutron or ion irradiation with an emphasis on experimental considerations. After outlining various neutron and ion irradiation scenarios, including some sample preparation suggestions, we discuss methods to measure defect densities, size distributions, structures, and interstitial or vacancy nature. The importance of the image simulations of Zhou is suggested for guidance to the most accurate quantification of the defect state. It is hoped that the usefulness of the present paper will be greatest for those experiments that compare defect states in materials after different irradiation conditions, or especially those studies designed to benchmark advanced computer model simulations of defect production and evolution. The successful simulation of the defect state in bulk samples neutron irradiated to high dose at high temperature is a goal to which the suggestions in this paper can contribute.
Journal of Materials Research 02/2015; DOI:10.1557/jmr.2015.19 · 1.65 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: 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.
JOM: the journal of the Minerals, Metals & Materials Society 10/2014; 66(12). DOI:10.1007/s11837-014-1186-6 · 1.76 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Oxide dispersion strengthened ferritic alloys have superior radiation tolerance and thus become appealing candidates as fuel cladding materials for next generation nuclear reactors. In this study we constructed a model system, Fe/Y2O3 nanolayers with individual layer thicknesses of 10 and 50 nm, in order to understand their radiation response and corresponding damage mitigation mechanisms. These nanolayers were subjected to in situ Kr ion irradiation at room temperature up to similar to 8 displacements-per-atom. As-deposited Y2O3 layers had primarily amorphous structure. Radiation induced prominent nanocrystallization and grain growth in 50 nm thick Y2O3 layers. Conversely, little crystallization occurred in 10 nm thick Y2O3 layers implying size dependent enhancement of radiation tolerance. In situ video also captured grain growth in both Fe and Y2O3 and outstanding morphological stability of layer interfaces against Kr ion irradiation.
[Show abstract][Hide abstract] ABSTRACT: Enhanced irradiation tolerance in crystalline multilayers has received significant attention lately. However, little is known on the irradiation response of crystal/amorphous nanolayers. We report on in situ Kr ion irradiation studies of a bulk Fe96Zr4 nanocomposite alloy. Irradiation resulted in amorphization of Fe2Zr and formed crystal/amorphous nanolayers. α-Fe layers exhibited drastically lower defect density and size than those in large α-Fe grains. In situ video revealed that mobile dislocation loops in α-Fe layers were confined by the crystal/amorphous interfaces and kept migrating to annihilate other defects. This study provides new insights on the design of irradiation-tolerant crystal/amorphous nanocomposites.
Materials Research Letters 08/2014; 3(1):35-42. DOI:10.1080/21663831.2014.951494
[Show abstract][Hide abstract] ABSTRACT: 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.
Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms 07/2014; 330:55–60. DOI:10.1016/j.nimb.2014.03.018 · 1.12 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Oxide Dispersion Strengthened (ODS) reduced activation ferritic steels are promising candidate materials for structural components of both nuclear fission and fusion reactors. However, when irradiated with energetic particles, they may suffer changes on their microstructures that degrade their mechanical performance. In-situ transmission electron microscopy studies on ion-irradiated ODS steels can give remarkable insights into fundamental aspects of radiation damage allowing dynamic observations of defect formation, mobilities, and interactions during irradiation. In this investigation, a commercially available PM2000 ODS steel was in-situ irradiated with 150 KeV Fe+ at room temperature and 700°C. These experiments showed that the oxide nanoparticles in these steels remain stable up to the higher irradiation dose (~ 1.5 dpa), and that these particles seem to be effective sinks for irradiation induced defects.
Journal of Physics Conference Series 06/2014; 522(1):012032. DOI:10.1088/1742-6596/522/1/012032
[Show abstract][Hide abstract] ABSTRACT: Phase stability of Ni3(Al, Ti) precipitates in Inconel X-750 under cascade damage was studied using heavy ion irradiation with transmission electron microscope (TEM) in situ observations. From 333 K to 673 K (60 °C to 400 °C), ordered Ni3(Al, Ti) precipitates became completely disordered at low irradiation dose of 0.06 displacement per atom (dpa). At higher dose, a trend of precipitate dissolution occurring under disordered state was observed, which is due to the ballistic mixing effect by irradiation. However, at temperatures greater than 773 K (500 °C), the precipitates stayed ordered up to 5.4 dpa, supporting the view that irradiation-induced disordering/dissolution and thermal recovery reach a balance between 673 K and 773 K (400 °C and 500 °C). Effects of Ti/Al ratio and irradiation dose rate are also discussed.
Metallurgical and Materials Transactions A 04/2014; 45a(6). DOI:10.1007/s11661-014-2309-y · 1.73 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: In order to understand radiation damage in the nickel based superalloy Inconel X-750 in thermal reactors, where (n, α) transmutation reaction also occurred in addition to fast neutron induced atomic displacement, heavy ion (1 MeV Kr2+) irradiation with pre-injected helium was performed under in-situ observations of an intermediate voltage electron microscope at Argonne National Laboratory. By comparing to our previous studies using 1 MeV Kr2+ irradiation solely, the pre-injected helium was found to be essential in cavity nucleation. Cavities started to be visible after Kr2+ irradiation to 2.7 dpa at ≥200 °C in samples containing 200 appm, 1000 appm, and 5000 appm helium, respectively, but not at lower temperatures. The cavity growth was observed during the continuous irradiation. Cavity formation appeared along with a reduced number density of stacking fault tetrahedra, vacancy type defects. With higher pre-injected helium amount, a higher density of smaller cavities was observed. This is considered to be the result of local trapping effect of helium which disperses vacancies. The average cavity size increases with increasing irradiation temperatures; the density reduced; and the distribution of cavities became heterogeneous at elevated temperatures. In contrast to previous characterization of in-reactor neutron irradiated Inconel X-750, no obvious cavity sink to grain boundaries and phase boundaries was found even at high doses and elevated temperatures. MC-type carbides were observed as strong sources for agglomeration of cavities due to their enhanced trapping strength of helium and vacancies.
[Show abstract][Hide abstract] ABSTRACT: In the current investigation, TEM in-situ heavy ion (1MeV Kr2þ) irradiation with helium
pre-injected at elevated temperature (400 �C) was conducted to simulate in-reactor neutron
irradiation induced damage in CANDU spacer material Inconel X-750, in an effort to understand
the effects of helium on irradiation induced cavity microstructures. Three different quantities of
helium, 400 appm, 1000 appm, and 5000 appm, were pre-injected directly into TEM foils at
400 �C. The samples containing helium were then irradiated in-situ with 1MeV Kr2þ at 400 �C to
a final dose of 5.4 dpa (displacement per atom). Cavities were formed from the helium injection
solely and the cavity density and size increased with increasing helium dosage. In contrast to
previous heavy ion irradiations with cold pre-injected helium, heterogeneous nucleation of
cavities was observed. During the ensuing heavy ion irradiation, dynamical observation showed
noticeable size increase in cavities which nucleated close to the grain boundaries. A
“bubble-void” transformation was observed after Kr2þ irradiation to high dose (5.4 dpa) in
samples containing 1000 appm and 5000 appm helium. Cavity distribution was found to be
consistent with in-reactor neutron irradiation induced cavity microstructures. This implies that the
distribution of helium is greatly dependent on the injection temperature, and helium pre-injection
at high temperature is preferred for simulating the migration of the transmutation produced