[Show abstract][Hide abstract] ABSTRACT: U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratory (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen.
[Show abstract][Hide abstract] ABSTRACT: An in situ ion-irradiation study, simultaneously examined using transmission electron microscopy, was performed to investigate irradiation-induced disordering and amorphization of Al3Ti-based intermetallic compounds. Thin foil samples of two crystalline structures: D022-structured Al3Ti and L12-structured (Al,Cr)3Ti were irradiated using 1.0 MeV Kr ions at a temperature range from 40 K to 573 K to doses up to 4.06 × 1015 ions/cm2. The results showed that both the compounds underwent an order-disorder transformation under irradiation, where both Al3Ti and (Al,Cr)3Ti ordered structures were fully transformed to the disordered face-centered cubic (FCC) structure except at the highest irradiation temperature of 573 K. A slightly higher irradiation dose was required for order-disorder transformation in case of Al3Ti as compared to (Al,Cr)3Ti at a given temperature. However, their amorphization resistances were different: while the disordered FCC (Al,Cr)3Ti amorphized at the irradiation dose of 6.25 × 1014 ions/cm2 (0.92 dpa) at 40 K and 100 K, the Al3Ti compound with the same disordered FCC structure maintained crystallinity up to 4.06 × 1015 ions/cm2 (5.62 dpa) at 40 K. The critical temperature for amorphization of (Al,Cr)3Ti under Kr ion irradiation is likely between 100 K and room temperature and the critical temperature for disordering between room temperature and 573 K.
[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: Tungsten is a prime candidate for building divertor components in fusion reactors. During its service life, these components may undergo up to 30-40 dpa of displacement damage per year, originated from the collision cascades of fusion neutrons. In this work, we investigated the production and evolution of radiation damage in tungsten with in-situ observations and analysis of 150 keV W + ion irradiations, so as to mimic the effects of the average primary recoil energy of 14 MeV fusion neutrons . TEM foils of pure tungsten (typically > 99.996 wt%, Plansee) were annealed (1673K, 20 h) and prepared from jet-electropolishing in a 0.5 wt.% NaOH aqueous solution. The in-situ irradiations were performed on the IVEM-Tandem facility at Argonne National Laboratory, with a focused beam of W + ions directed to the specimen surface, at a high incident angle of ~75°. The irradiation conditions covered a wide temperature range from 30 K to 1073 K and a dose range from 10 16 to 10 18 W + /m 2 (0.01 ~ 1.0 dpa) at a constant rate of ~ 6.25×10 14 W + /m 2 s. We recorded the defect dynamics at 15 frames/s, and performed defect characterizations and analyses (population; size distribution; geometry; nature) following the methods described in Jenkins et al . This comprehensive study of damage production and evolution in tungsten has led to several new findings. We have discovered that at doses ≤ 0.01 dpa, the first observable defects nucleated in tungsten were vacancy loops, predominantly of b = ½ <111>, and were formed within individual collision cascades. With the increase of temperature from stage I (30 K) to stage IV (1073 K), loops with b = ½ <111> gained increasing predominance over those with b = <100> and the analysis of defect size versus the frequency of occurrence suggested the engagement of strong elastic interactions among all radiation-induced defects in the cascade (more details in Mason et al) . At doses beyond the overlap of cascades (> 0.01 dpa), radiation damage in tungsten evolved through the 1D migration of defect clusters, the elastic interactions and (typically) non-conservative reaction among these defect clusters, rendered by irradiation temperature and dose. Notably, a transition from random distribution to spatial ordering of loops has been observed in the z = <001> grains at doses > 0.4 dpa and T ≥ 773 K, as partially illustrated in Figure 1. We've also noticed that non z = <001> orientations may considerably lower the threshold of this ordering phenomenon. The highlights of damage analysis in tungsten are summarized in Figure 2. The rates of defect accumulation in Figure 2a indicated a rapid saturation for T < 773 K, whereas for T ≥ 773 K, the first signs of saturation did not occur until > 0.4 dpa. Post-irradiation analysis of defect size distributions (Figure 2b) and defect geometry (Figure 2c) at 1.0 dpa suggested that 773 K (stage III, migration of monovacancies) is a characteristic temperature in the radiation damage evolution of tungsten. Together with evidence found in defect dynamic behavior and the results of defect nature determination, we have found that temperature and dose tend to drive the damage microstructure in tungsten towards an increased proportion of interstitial ½ <111> loops, an increased degree of spatial ordering among them and facilitate their size increase through coalescence reactions.
Microscopy and Microanalysis 2015, Portand, USA; 08/2015
[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