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Now showing items 1 - 16 of 7238

  • The incorporation and migration of a single xenon atom in ceria

    Yinbin Miao   Wei-Ying Chen   Aaron Oaks   Kun Mo   James F. Stubbins  

    Abstract The behavior of xenon gas is crucial for the performance of nuclear fuel materials. We report molecular statics calculation results for the characteristics of a single xenon atom in cerium oxide, a non-radioactive surrogate of uranium dioxide. A variety of possible xenon incorporation sites, including the octahedral interstitial position, single-Ce-vacancy clusters, and double-Ce-vacancy clusters were considered. The binding energies and corresponding xenon incorporation energies were computed to reveal the preferred xenon positions in ceria. Different migration mechanisms of single xenon atoms were found to be involved with various incorporation sites. The energy barriers of all possible migration pathways were calculated. Only the mobility of single xenon atoms in the double-Ce-vacancy sites, which is due to the vacancy-assisted xenon migration, can account for the xenon diffusivity implied by bubble formation observed in experiments. The results also validated the role of ceria as a reliable surrogate of uranium dioxide in studies involving xenon gas.
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  • Biaxial thermal creep of Inconel 617 and Haynes 230 at 850 and 950 °C

    Hsiao-Ming Tung   Kun Mo   James F. Stubbins  

    Abstract The biaxial thermal creep behavior of Inconel 617 and Haynes 230 at 850 and 950 °C was investigated. Biaxial stresses were generated using the pressurized tube technique. The detailed creep deformation and fracture mechanism have been studied. Creep curves for both alloys showed that tertiary creep accounts for a greater portion of the materials’ life, while secondary creep only accounts for a small portion. Fractographic examinations of the two alloys indicated that nucleation, growth, and coalescence of creep voids are the dominant micro-mechanisms for creep fracture. At 850 °C, alloy 230 has better creep resistance than alloy 617. When subjected to the biaxial stress state, the creep rupture life of the two alloys was considerably reduced when compared to the results obtained by uniaxial tensile creep tests. The Monkman–Grant relation proves to be a promising method for estimating the long-term creep life for alloy 617, whereas alloy 230 does not follow the relation. This might be associated with the significant changes in the microstructure of alloy 230 at high temperatures.
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  • High Temperature Aging and Corrosion Study on Alloy 617 and Alloy 230

    Kun Mo   Gianfranco Lovicu   Hsiao Ming Tung   Xiang Chen   James F. Stubbins  

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  • Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

    Xiang Liu   Yinbin Miao   Meimei Li   Marquis A. Kirk   Stuart A. Maloy   James F. Stubbins  

    Abstract In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 × 10 15 ions/cm 2 at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black-dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 × 10 15 ions/cm 2 , about 51% of the loops were a 0 / 2 〈 111 〉 type for the 673 K irradiation, and the dominant loop type was a 0 〈 100 〉 for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100–150 MPa. Graphical abstract Image 1 • Microstructure evolution of T91 under Krypton ion irradiation was studied. • Distinct microstructure evolution was explained by 〈111>−<100〉 loop transition. • A hardening model was used to establish the microstructure-property relationship.
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  • Load-partitioning in an oxide dispersion-strengthened 310 steel at elevated temperatures

    Yinbin Miao   Kun Mo   Zhangjian Zhou   Xiang Liu   Kuan-Che Lan   Guangming Zhang   Jun-Sang Park   Jonathan Almer   James F. Stubbins  

    Abstract Here the high temperature tensile performance of an oxide dispersion-strengthened (ODS) 310 steel is reported upon. The microstructure of the steel was examined through both transmission electron microscopy (TEM) and synchrotron scattering. In situ synchrotron X-ray tensile investigation was performed at a variety of temperatures, from room temperature up to 800 °C. Pyrochlore structure yttrium titanate and sodium chloride structure titanium nitride phases were identified in the steel along with an austenite matrix and marginal residual α′ -martensite. The inclusion phases strengthen the steel by taking extra load through particle-dislocation interaction during plastic deformation or dislocation creep procedures. As temperature rises, lattice strain measurement implies that the load partitioning effect of conventional precipitate phases starts to diminish, whereas those ultra-fine oxygen-enriched nanoparticles continue to maintain a considerable amount of extra lattice strain. Introduction of oxygen-enriched nanoparticles in austenitic steel is shown to improve the high temperature performance, making austenitic ODS steels promising for advanced nuclear applications. Highlights • An austenitic oxide dispersion-strengthened stainless steel was developed and investigated for nuclear applications. • Size distribution and orientation relationship of the ultra-fine oxygen-enriched nanoparticles were measured. • In situ tensile tests were conducted at various temperatures to reveal the strengthening effect of precipitates. • Only the ultra-fine oxygen-enriched nanoparticles are capable of retaining their strengthening effect at high temperatures. • The austenitic oxygen dispersion-strengthened stainless steel showed advantageous mechanical strength. Graphical abstract
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  • The evolution mechanism of the dislocation loops in irradiated lanthanum doped cerium oxide

    Yinbin Miao   Dilpuneet Aidhy   Wei-Ying Chen   Kun Mo   Aaron Oaks   Dieter Wolf   James F. Stubbins  

    Abstract Cerium dioxide, a non-radioactive surrogate of uranium dioxide, is useful for simulating the radiation responses of uranium dioxide and mixed oxide fuel (MOX). Controlled additions of lanthanum can also be used to form various levels of lattice oxide or anion vacancies. In previous transmission electron microscopy (TEM) experimental studies, the growth rate of dislocation loops in irradiated lanthanum doped ceria was reported to vary with lanthanum concentration. This work reports findings of the evolution mechanisms of the dislocation loops in cerium oxide with and without lanthanum dopants based on a combination of molecular statics and molecular dynamics simulations. These dislocation loops are found to be b = 1 / 3 〈 1 1 1 〉 interstitial type Frank loops. Calculations of the defect energy profiles of the dislocation loops with different structural configurations and radii reveal the basis for preference of nucleation as well as the driving force of growth. Frenkel pair evolution simulations and displacement cascade overlaps simulations were conducted for a variety of lanthanum doping conditions. The nucleation and growth processes of the Frank loop were found to be controlled by the mobility of cation interstitials, which is significantly influenced by the lanthanum doping concentration. Competition mechanisms coupled with the mobility of cation point defects were discovered, and can be used to explain the lanthanum effects observed in experiments.
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  • Size-dependent characteristics of ultra-fine oxygen-enriched nanoparticles in austenitic steels

    Yinbin Miao   Kun Mo   Zhangjian Zhou   Xiang Liu   Kuan-Che Lan   Guangming Zhang   Michael K. Miller   Kathy A. Powers   James F. Stubbins  

    Abstract Here, a coordinated investigation of the elemental composition and morphology of ultra-fine-scale nanoparticles as a function of size within a variety of austenitic oxide dispersion-strengthened (ODS) steels is reported. Atom probe tomography was utilized to evaluate the elemental composition of these nanoparticles. Meanwhile, the crystal structures and orientation relationships were determined by high-resolution transmission electron microscopy. The nanoparticles with sufficient size (>4 nm) to maintain a Y 2 Ti 2− x O 7−2 x stoichiometry were found to have a pyrochlore structure, whereas smaller Y x Tiy O z nanoparticles lacked a well-defined structure. The size-dependent characteristics of the nanoparticles in austenitic ODS steels differ from those in ferritic/martensitic ODS steels. Highlights • The structural and chemical characteristics of nanoparticles are revealed. • Nanoparticles' crystal structure and elemental composition are size-dependent. • Characteristics of austenitic ODS steels are compared to that of an F/M ODS steel. • Hypothesis about the formation mechanism of nanoparticles is proposed accordingly.
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  • Atom probe study of irradiation-enhanced α′ precipitation in neutron-irradiated Fe–Cr model alloys

    Wei-Ying Chen   Yinbin Miao   Yaqiao Wu   Carolyn A. Tomchik   Kun Mo   Jian Gan   Maria A. Okuniewski   Stuart A. Maloy   James F. Stubbins  

    Abstract Atom probe tomography (APT) was performed to study the effects of Cr concentrations, irradiation doses and irradiation temperatures on α′ phase formation in Fe–Cr model alloys (10–16 at.%) irradiated at 300 and 450 °C to 0.01, 0.1 and 1 dpa. For 1 dpa specimens, α′ precipitates with an average radius of 1.0–1.3 nm were observed. The precipitate density varied significantly from 1.1 × 10 23 to 2.7 × 10 24 1/m 3 , depending on Cr concentrations and irradiation temperatures. The volume fraction of α′ phase in 1 dpa specimens qualitatively agreed with the phase diagram prediction. For 0.01 dpa and 0.1 dpa, frequency distribution analysis detected slight Cr segregation in high-Cr specimens, but not in Fe–10Cr specimens. Proximity histogram analysis showed that the radial Cr concentration was highest at the center of α′ precipitates. For most precipitates, the Cr contents were significantly lower than that predicted by the phase diagram. The Cr concentration at precipitate center increased with increasing precipitate size.
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  • Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation

    Sumit Bhattacharya   Xiang Liu   Yinbin Miao   Kun Mo   Zhi-Gang Mei   Laura Jamison   Walid Mohamed   Aaron Oaks   Ruqing Xu   Shaofei Zhu   James F. Stubbins   Abdellatif M. Yacout  

    Abstract Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86 × 10 17 ions/cm 2 , or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction ( μ XRD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, D n − D 0 n = K ϕ , with n  = 5.1. The Al-ZrN interaction products (Al 3 Zr and AlN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance. Graphical abstract Image 1
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  • Effect of creep and oxidation on reduced fatigue life of Ni-based alloy 617 at 850 °C

    Xiang Chen   Zhiqing Yang   Mikhail A. Sokolov   Donald L. Erdman   Kun Mo   James F. Stubbins  

    Abstract Low cycle fatigue (LCF) and creep–fatigue testing of Ni-based alloy 617 was carried out at 850 °C. Compared with its LCF life, the material’s creep–fatigue life decreases to different extents depending on test conditions. To elucidate the microstructure-fatigue property relationship for alloy 617 and the effect of creep and oxidation on its fatigue life, systematic microstructural investigations were carried out using scanning electron microscopy, energy-dispersive X-ray spectroscopy, and electron backscatter diffraction (EBSD). In LCF tests, as the total strain range increased, deformations concentrated near high angle grain boundaries (HAGBs). The strain hold period in the creep–fatigue tests introduced additional creep damage to the material, which revealed the detrimental effect of the strain hold time on the material fatigue life in two ways. First, the strain hold time enhanced the localized deformation near HAGBs, resulting in the promotion of intergranular cracking of alloy 617. Second, the strain hold time encouraged grain boundary sliding, which resulted in interior intergranular cracking of the material. Oxidation accelerated the initiation of intergranular cracking in alloy 617. In the crack propagation stage, if oxidation was promoted and the cyclic oxidation damage was greater than the fatigue damage, oxidation-assisted intergranular crack growth resulted in a significant reduction in the material’s fatigue life.
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  • The electrical impedance response of steel surface oxide layers in molten lead–bismuth eutectic

    James F. Stubbins   Alan Michael Bolind   Xiang Chen  

    Stainless steel 316L samples were preoxidized and then immersed in molten lead-bismuth eutectic (LBE) alloy at 200degC. The changes in their electrical impedance responses were observed over time. Negligible impedance magnitudes were observed at first, followed by a rapid increase to thousands of ohm-cm 2. The impedance response is sensitive to changes in the immersed sample area. Micro-indentations on samples caused their impedance magnitudes to decrease initially, but the magnitudes recovered within a few days. SEM analysis showed that the indentations were still present and visible even after the recovery of impedance response, demonstrating that the physical features of the oxide layers which govern the electrical response must be smaller than the micrometer length scale. [All rights reserved Elsevier].
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  • The electrical impedance response of steel surface oxide layers in molten lead–bismuth eutectic

    James F. Stubbins   Alan Michael Bolind   Xiang Chen  

    Stainless steel 316L samples were preoxidized and then immersed in molten lead–bismuth eutectic (LBE) alloy at 200 °C. The changes in their electrical impedance responses were observed over time. Negligible impedance magnitudes were observed at first, followed by a rapid increase to thousands of ohm-cm2. The impedance response is sensitive to changes in the immersed sample area. Micro-indentations on samples caused their impedance magnitudes to decrease initially, but the magnitudes recovered within a few days. SEM analysis showed that the indentations were still present and visible even after the recovery of impedance response, demonstrating that the physical features of the oxide layers which govern the electrical response must be smaller than the micrometer length scale.
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  • Bulk processing of materials with radiation

    James F. Stubbins  

    The potential for using highly penetrating radiation for processing bulk material is examined. It is found that gamma radiation, which is highly penetrating and nonactivating, is an acceptable radiation source but that only very low numbers of atomic interactions are possible even with intense sources. This limits the usefulness to certain classes of processing. The most promising of these are ones where undersized solute atoms are preferentially displaced and concentrate at interfacial sinks. The possibility for radiation-assisted nucleation is also examined, as are source type, strength, and handling limitations
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  • Mechanical properties of UO2 thin films under heavy ion irradiation using nanoindentation and finite element modeling

    Mohamed S. Elbakhshwan   Yinbin Miao   James F. Stubbins   Brent J. Heuser  

    Abstract The mechanical response of UO 2 to irradiation is becoming increasingly important due to the shift to higher burn-up rates in the next generation of nuclear reactors. In the current study, thin films of UO 2 were deposited on YSZ substrates using reactive-gas magnetron sputtering. Nanoindentation was used to measure the mechanical properties of the as-grown and irradiated films. Finite element modeling was used to account for the substrate effect on the measurements. In order to study the effect of displacement cascades accompanying gas bubbles, 5000 Å UO 2 films were irradiated with 600 keV Kr + ions at 25 °C and 600 °C. These irradiation conditions were used to confine radiation damage effects and implanted gas within the film. Results showed an increase in the film hardness and yield strength with dose, while elastic modulus initially decreased with irradiation and then kept increasing with dose. The change in hardness and elastic modulus is attributed to the introduction of gas bubbles and displacement cascade damage. Irradiation at 600 °C resulted in a decrease in the hardness and elastic modulus after irradiation using 600 keV Kr + at a dose of 1E14 ions/cm 2 . Both hardness and elastic modulus then increased with irradiation dose. This behavior is attributed to recrystallization during irradiation at 600 °C and the formation of nanocrystallite regions with diameter and density that increase with dose. The calculation of the critical resolved shear stress (CRSS) demonstrated that nanocrystals are the primary cause for film hardening based on the Orowan hardening mechanism.
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  • Void swelling and radiation-induced phase transformation in high purity Fe-Ni-Cr alloys

    James F. Stubbins  

    Three high purity Fe-Ni-Cr alloys were irradiated with high energy Ni ions and with high energy electrons over a temperature range of 450 to 650degC to doses up to 30 displacements per atom (dpa) for heavy ions and 60 dpa for electrons. The alloy compositions were Fe-10Cr-15Ni, Fe-16Cr-15Ni, and Fe-20Cr-15Ni with very low carbon levels. These alloy compositions were chosen to bracket a range of compositions where large differences in void swelling are seen, and where alloy phase instabilities are predicted at lower temperatures. The results of transmission electron microscopy (TEM) of irradiated specimens showed that none of the phase transformations predicted by computed phase diagrams occurred during irradiation. In one case, M 23C 6 was formed under irradiation. A more significant finding was a martensitic reversion reaction which occurred under irradiation of ferritic regions of the 10Cr alloy at 450 and 550degC. This reaction was found for both heavy ion and electron irradiations. The void swelling behavior of these alloys was examined in detail. Voids were found only after irradiation at 650degC for the heavy ion cases, but all irradiation temperatures with electrons. The amount of swelling was found to be dependent on composition, but the HVEM studies indicate that these differences may be due to differences in incubation doses between alloy compositions
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  • High voltage electron and helium irradiation of molybdenum

    James F. Stubbins  

    High purity molybdenum was irradiated with 1.1 MeV electrons, or 100 keV He +1 ions, or both simultaneously at temperatures of 800, 850 and 900degC. Voids were observed only at 900degC in low dose electron irradiations, while vacancy loops were formed at the lower irradiation temperatures. Simultaneous He and electron bombardment produced small defect clusters at 850degC. He implantation alone at 900degC produced a large number density of small loops, and no visible voids/bubbles
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