Engineering topics
Jiao, Zhijie
Publications and source records attributed to Jiao, Zhijie.
Data on Cu- and Ni-Si-Mn-rich solute clustering in a neutron irradiated austenitic stainless steel
The data presented in this article is supplementary to the research article “Phase instabilities in austenitic steels during particle bombardment at high and low dose rates” (Levine et al.). Needle-shaped samples were prepared with focused ion beam milling from a 304L stainless steel that was irradiated with fast neutrons (E 0.1 MeV) in the BOR-60 reactor at 318 °C to 47.5 dpa. Atom probe tomography (APT) experiments in voltage mode were then conducted on a Cameca LEAP 5000X HR. Atom position, range, and mass spectrum files after reconstruction with Cameca’s IVAS software are included. Cu- and Ni-Si-Mn-rich solute nanoclusters were identified and analyzed using the Open Source Characterization of APT Reconstructions (OSCAR) program. Python code for OSCAR, information on the program’s underlying algorithm, and sample output files are provided. A proximity histogram of a Ni-Si-Mn-rich cluster and a 1D density/solute concentration profile of a Cu-rich cluster are given to demonstrate OSCAR’s analytical functionalities. The provided APT dataset is valuable for benchmarking phase instabilities in neutron-irradiated austenitic stainless steels that occur at high doses. The OSCAR program can be reused to process other APT data sets where solute nanoclustering is of interest.
Synergies between H, He and radiation damage in dual and triple ion irradiation of candidate fusion blanket materials
Three ferritic/martensitic alloys were studied to understand the synergistic effect between single ion beam (Fe 2+ ), dual ion beam (Fe 2+ +He 2+ and Fe 2+ +H + ), and triple ion beam (Fe 2+ +He 2+ +H + ) irradiations on cavity evolution. A commercial alloy, F82H, a castable nanostructured alloy, CNA3, and a high purity model alloy, Fe8Cr2W, were irradiated at 400°C to 600°C to a damage level of 50 dpa at a damage rate of 1 × 10 –3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field transmission electron microscopy and high-angle annular dark-field scanning transmission electron microscopy was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium resulted in an increased cavity number density and maximum cavity size, producing an increase in swelling over that from helium injection alone. Swelling in F82H appears to peak between 450°C and 500°C. At 600°C, swelling was minimal and cavities of high density and small size were confined to grain boundaries and dislocations while at 400°C, swelling is also low with a nearly homogeneous, high density, distribution of very small cavities throughout. Swelling was least in the commercial alloy F82H due to the high sink strength. The CNA3 alloy underwent dissolution of precipitates that lowered the sink strength and resulted in higher swelling than F82H, but less than the model alloy. Electron energy loss spectroscopy (EELS) elemental mapping revealed hydrogen forming a halo-like structure about the periphery of the cavities and helium residing within the cavities themselves. Finally, this observation suggests that hydrogen reduces the surface energy of helium-filled cavities which results in both increased cavity number density and cavity size in triple beam irradiation over dual beam irradiation.
Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel
We report that dual-ion irradiations using 5.0 MeV defocused Fe 2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He 2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 × 10 -5 dpa/s to 3 × 10 -3 dpa/s at 445°C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation.
Fission Product Diffusion in Silicon Carbide: As Revealed by Computer Simulations and Experimental Measurements
The diffusion and release of fission products through silicon carbide in the tri-structural isotropic (TRISO) fuel particles remain unsolved for decades. The underlying mechanism is quite challenging to be determined. To help unveil the mysterious story, the current work applies molecular dynamics method to show the stability of silver, palladium, ruthenium and iodine as an interstitial and their atomic diffusion along coincident site lattice (CSL) boundary, especially ?3 grain boundary (GB). The major finding presents a much faster diffusion along GB than in bulk for all elements considered. The reasonably close estimate to experiments and simulations where available has confirmed the important role of grain boundary diffusion of Ag and Pd in SiC. However, the discrepancy addressed in Ag with measurements from fuel studies suggest a more complicated mechanism, which might be in correlation with high energy grain boundaries or the presence of crack. The subsequent characterization of Ru and I distribution in SiC-PyC-SiC diffusion couples, which have been ion irradiated at 900?C to 10 dpa and 20 dpa, has performed by secondary ion mass spectrometry (SIMS) analysis. The experiment measurements correlate well with the grain boundary diffusion by simulation, which provide further evidences that the grain boundary diffusion cannot be neglected once the fission products are accessible at the grain boundary.
Effects of post-irradiation annealing on the IASCC susceptibility of neutron-irradiated 304L stainless steel
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Emulation of fast reactor irradiated T91 using dual ion beam irradiation
In this work, dual ion irradiations using 5 MeV defocused Fe 2+ ions and co-injected He 2+ ions were conducted on a ferritic-martensitic steel alloy, T91, in the temperature range of 406 °C–570 °C over a damage range of 14.6–35 dpa followed by characterization of the microstructure using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). Dislocation loops were observed to increase in diameter and decrease in density with temperature until only network dislocations were observed at the highest temperatures of 520 °C and 570 °C. Swelling exhibited the expected bell-shaped trend with temperature following the number density of cavities, peaking at 460 °C and with a bimodal size distribution except at 520 °C and 570 °C. Nickel- and silicon-rich clusters formed under dual ion irradiations near the surface at all but the highest temperatures of 520 °C and 570 °C. Very little Cr and Si segregation was observed at lath boundaries while Ni enriched at all temperatures examined. Segregation of Cr and Ni appeared to saturate by 17 dpa, while Si enriched up to 35 dpa. The dislocation and cavity microstructures of dual ion irradiated T91 and T91 irradiated in the BOR-60 fast reactor matched extremely well using a temperature shift of +60–70 °C. However, segregation to grain boundaries and formation of nickel-silicon rich clusters were minimal in the dual ion irradiated T91 and less than that in T91 irradiated in the BOR-60 fast reactor.
Reveal the fast and charge-insensitive lattice diffusion of silver in cubic silicon carbide via first-principles calculations
The lattice diffusion of silver with various charge states in cubic silicon carbide has been investigated by means of high-fidelity spin-polarized density functional theory calculations. The migration energy barrier of the Ag interstitial diffusion is 1.09 eV and 1.11 eV for neutral and q = +1e charge state, respectively, close to the activation energy of Ag diffusion measured in the German HTR fuel program. A general trend is that the migration energy barrier reduces with respect to the increase of charge on Ag, which is much less than the increment in the oxidation energy, suggesting that the lattice diffusion of silver prefers constant neutral state without redox in transition state. Furthermore, our results indicate a scenario that once Ag is deposited to interstitials via the kickout mechanism, it will highly likely perform lattice diffusion across the SiC layer leading to fast release of Ag in Tristructural-Isotropic fuel induced by irradiation.
Ab initio study of the stability of intrinsic and extrinsic Ag point defects in 3C SiC
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The influence of carbon on cavity evolution in ion-irradiated ferritic-martensitic steels
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The influence of helium on cavity evolution in ion-irradiated T91
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Multiple ion beam irradiation for the study of radiation damage in materials
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Accelerated Irradiations for High Dose Microstructures in Fast Reactor Alloys
The objective of this project is to determine the extent to which high dose rate, self-ion irradiation can be used as an accelerated irradiation tool to understand microstructure evolution at high doses and temperatures relevant to advanced fast reactors. We will accomplish the goal by evaluating phase stability and swelling of F-M alloys relevant to SFR systems at very high dose by combining experiment and modeling in an effort to obtain a quantitative description of the processes at high and low damage rates.
Microstructural processes in irradiated materials
This is an editorial article (preface) for the publication of symposium papers in the Journal of Nuclear materials: These proceedings contain the papers presented at two symposia, the Microstructural Processes in Irradiated Materials (MPIM) and Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation, held in the TMS 2015, 144th Annual Meeting & Exhibition at Walt Disney World, Orlando, Florida, USA on March 15–19, 2015.
Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature
Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. Here, the studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Lastly, transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.
Superior Corrosion Resistance Properties of TiN-Based Coatings on Zircaloy Tubes in Supercritical Water
Thin films of TiN and Ti0.35Al0.65N nanocomposite were deposited on polished Zircaloy-4 tubes. After exposure to supercritical water for 48 h, the coated tubes are remarkably intact, while the bare uncoated tube shows severe oxidation and breakaway corrosion. X-ray diffraction patterns, secondary electron images, backscattered electron images, and energy dispersive X-ray spectroscopy data from the tube surfaces and cross-sections show that a protective oxide, formed on the film surface, effectively prevents further oxidation and corrosion to the Zircaloy-4 tubes. This result demonstrates the effectiveness of thin film ceramics as protective coatings under extreme environments.
Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments
The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of radiation on these important materials. The objective of this project is to conduct critical experiments to understand the evolution of microstructural and microchemical features (loops, voids, precipitates, and segregation) and mechanical properties (hardening and creep) under high temperature and full dose range radiation, including the effect of differences in the initial material composition and microstructure on the microstructural response, including key questions related to saturation of the microstructure at high doses and temperatures.
Microstructural evolution in proton irradiated NF616 at 773 K to 3 dpa
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