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At least 1,153 records · Page 64

Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC

Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.

36 MATERIALS SCIENCE↗

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗

Irradiation-induced chemical disordering in ceramics: The case of SiC

Chemical disordering is one feature of damage that occurs in irradiation environments and is common in ceramic compounds. This paper reviews irradiation-induced chemical disordering in SiC-based materials at elevated temperatures. Here the results were obtained using advanced analytical tools to characterize atomistic defects and to evaluate the chemistry of defect clusters and homonuclear bond structures. Finally, this paper demonstrates that chemical disordering is crucial to understanding microstructural stability and establishing a mechanistic modeling of microstructural evolution of irradiated SiC.

36 MATERIALS SCIENCE↗

Irradiation response of innovatively engineered metastable TRIP high entropy alloy

For this work, properties and radiation responses of a metastable high alloy (HEA) exhibiting the transformation induced plasticity (TRIP) effect were studied. Innovative engineering used to manufacture this HEA has shown superior mechanical and corrosion properties in 3.5% NaCl than most advanced stainless steels. The microstructural evolution and corresponding mechanical response after irradiation have been evaluated using detailed transmission electron microscopy, and nanoindentation. The study shows a change in metastability of the alloy with irradiation via a recovery mechanism, where the irradiation-induced transformation is reversed by the temperature-induced transformation, thereby introducing the concept of self-healing, made possible due to the TRIP behavior of HEA.

36 MATERIALS SCIENCE↗

Atomistic simulations of He bubbles in Beryllium

Formation of He bubbles can have a significant effect on the microstructural evolution and properties of irradiated materials. In this report, we use atomistic simulations based on machine learning potentials to investigate the fundamental behavior of He bubbles in Be, with a specific focus on the shape, stability, and diffusivity of bubbles. Stability of He bubbles is quantified in terms of formation energies, which are determined as a function of the ratio of He/V. We find that He bubbles become unstable with respect to plastic deformation through punch-out dislocations around the bubble when the He/V ratio is larger than ~1.25, and the punch-out process induces the change of the regular bubble shape. In general, the bubble shape of He in Be is found to be ellipsoid-like. It is also found that for a fixed He/V ratio, the bubble attracts vacancies to become larger in size. If the bubble size is constant, the bubble attracts additional He atoms until the punch-out reaction occurs. The dominant diffusion mechanism of He bubbles changes from surface diffusion to volume diffusion as the temperature is increased, with a crossover occurring at about 900 K.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Manufacturing porous U-10Zr metallic fuels with controllable microstructure by volume control spark plasma sintering

In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NH 4 HCO 3 are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting of highly porous and highly densified areas. Uneven powder stacking in the green body results in a non-uniform microstructure, and in the closed-packed area, Joule heating accelerates the neck formation and densification. The addition of ethanol reduces the friction between the powders, resulting in isolated pores formed by the stacking of powders during the sintering. By adding NH 4 HCO 3 , the pore size, and structure can be well controlled, and an interconnected pore structure can be obtained upon the decomposition of the NH 4 HCO 3 . Further, a uniform microstructure and pore distributions can be achieved through the U-10Zr fuel pellets by controlling current flow during the volume control SPS sintering. The microstructure and phase characterization of the sintered porous U-10Zr pellets show major phases of α-U and α-Zr for the sample with short dwelling. For the sample with long dwelling (30 min), the ω UZr 2 in the Zr-enriched area has been observed. The strategy of volume control SPS sintering with the assistance of pore formers could be used to fabricate porous U-10Zr metallic fuels to mimic the microstructure evolution of irradiated metallic fuels (including porosity) and could enable a possible solution for the design of new sodium-free metallic fuels for high burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance

Here, the development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.

BISON↗

Interpreting experimental measurements of helium bubbles using stochastic cluster dynamics models of heterogeneous nucleation and growth in irradiated ferritic alloys

Among a number of other advantageous features, ferritic/martensitic steels are being considered as fusion reactor structural materials due to their low intrinsic swelling under irradiation. However, under high-energy neutron irradiation, He produced through (n, α) reactions stabilizes vacancy clusters, which then act as seeds for bubble formation and growth, negating the intrinsic swelling resistance of these alloys. Standard models of irradiation damage accumulation and microstructural evolution consider homogeneous nucleation as the basis for bubble formation and growth. However, this generally does not explain the large bubble densities and sizes observed experimentally under a number of different conditions. Here, we propose a more realistic physical model of bubble nucleation, formation, and growth designed to capture recent experimental measurements of He-bubble formation and evolution during co-implantation of He and Fe ions in model ferritic alloys. We find that experimental results are explained only when the following three features are considered simultaneously: (i) heterogeneous nucleation of He-vacancy bubbles at defect sinks (e.g., dislocations, grain boundaries, and second-phase precipitates), (ii) nucleation and growth of bubbles via the ‘trap mutation’ mechanism (i.e., spontaneous production of Frenkel pairs due to absorption of He atoms), and (iii) transition from a growth-limited, He-stabilized bubble regime to a ‘runaway’ void-type growth scenario. The model is implemented into a stochastic cluster dynamics framework that takes advantage of cluster size grouping methods to accelerate the simulations, allowing us to reach 10 dpa of simulated irradiated dose, and to capture cluster sizes in excess of 20 nm. Further, a careful extrapolation exercise conducted assuming classical nucleation theory leads to excellent agreement with the experimental measurements at 50 dpa of irradiation.

36 MATERIALS SCIENCE↗

Back to functional hydrides: Effects of neutron-irradiated microstructure on hydrogen retention in yttrium hydride

Functional hydrides are promising candidates for advanced nuclear reactors, particularly in portable or transportable applications, due to their high hydrogen-retention capabilities, enabling efficient neutron moderation, and compact reactor design. However, hydrogen mobility in hydrides at elevated irradiation temperatures poses significant technological challenges, necessitating a comprehensive understanding of their irradiation behavior. Furthermore, this study investigated the microstructural and chemical stability of neutron-irradiated yttrium hydrides to assess their hydrogen-retention capacity. A targeted literature review was also conducted to contextualize neutron-irradiation effects on functional hydrides with regards to structural stability and hydrogen retention. Experimental characterizations revealed that, at high temperatures, irradiated hydrides retained their phase stability, which was likely enhanced by irradiation-induced microstructure evolution. Notably, an amorphous yttrium and oxygen -rich surface layer was present at the free surface of the hydride. Its thickness decreased while a continuous crystalline Y-O-rich layer was formed with increasing neutron damage. Additionally, the number density of dislocation loops and cavities generally increased as a function of neutron dose. First-principles calculations of hydrogen behavior within yttrium vacancy clusters in yttrium hydrides demonstrated vacancy-size-dependent hydrogen stability and configuration, highlighting the role of vacancy geometry in regulating hydrogen retention. Thus, the presence of irradiation-induced dislocation loops and cavities were hypothesized to improve hydrogen retention. Collectively, these findings advance the understanding of hydride behavior under neutron irradiation as well as their technological readiness for portable or transportable nuclear reactors.

Defect clusters↗

Identifying microstructures susceptible to pulverization in commercially irradiated high burnup UO2 under LOCA conditions

High burnup fuel fragmentation (HBFF) has been a concern in the nuclear industry for many years. When UO2 reaches and exceeds pellet average burnups around 55 GWd/tU, microstructural changes in the fuel pellet begin to occur that render the pellet susceptible to fine fragmentation under loss-of-coolant accident (LOCA) conditions. During a LOCA event, the cladding balloons and bursts, potentially releasing part of these fine fuel fragments into the reactor pressure vessel. This process is known as fuel fragmentation, relocation, and dispersal (FFRD). Presently, the US nuclear industry is developing a safety basis for FFRD with the goal of increasing burnups and pressurized water reactor cycle lengths. Increased cycle lengths would push the burnup limits of the fuel rods past the known threshold for HBFF susceptibility. In this work, the microstructural impact on HBFF was investigated by comparing the microstructures of as-irradiated rod segments to their post-LOCA tested counterparts. This investigation found that varied operational histories result in different microstructural evolutions across the pellet radius, which impacts the fragmentation behavior and radial location of the fragmentation. It is noted that regions with a high bubble density and a high density of grain boundaries fragment under LOCA-relevant conditions. In addition to the fragmentation that has been previously reported at the periphery of the fuel, fragmentation was also noted in the dark zone toward the fuel center. Analysis of the power histories of the fuel samples suggests that the dark zone forms in the central regions of the pellet between temperatures of approximately 740 and 960 °C.

McKinney, Casey [ORNL] (ORCID:0000000335383614)↗

Nano/ultrafine grained austenitic stainless steel through the formation and reversion of deformation-induced martensite: Mechanisms, microstructures, mechanical properties, and TRIP effect

A comprehensive study was carried out on the strain-induced martensitic transformation, its reversion to austenite, the resultant grain refinement, and the enhancement of strength and strain-hardening ability through the transformation-induced plasticity (TRIP) effect in a commercial austenitic 304L stainless steel with emphasis on the mechanisms and the microstructural evolution. A straightforward magnetic measurement device, which is based on the measurement of the saturation magnetization, for evaluating the amount of strain-induced martensite after cold rolling and reversion annealing in metastable austenitic stainless steels was used, which its results were in good consistency with those of the X-ray diffraction (XRD) method. A new parameter called the effective reduction in thickness was introduced, which corresponds to the reasonable upper bound on the obtainable martensite fraction based on the saturation in the martensitic transformation. By means of thermodynamics calculations, the reversion mechanisms were estimated and subsequently validated by experimental results. The signs of thermal martensitic transformation at cooling stage after reversion at 850 °C were found, which was attributed to the rise in the martensite start temperature due to the carbide precipitation. After the reversion treatment, the average grain sizes were around 500 nm and the nanometric grains of the size of ~ 65 nm were also detected. The intense grain refinement led to the enhanced mechanical properties and observation of the change in the work-hardening capacity and TRIP effect behavior. A practical map as a guidance for grain refining and characterizing the stability against grain growth was proposed, which shows the limitation of the reversion mechanism for refinement of grain size. - Graphical abstract: Display Omitted - Highlights: • Nano/ultrafine grained austenitic stainless steel through martensite treatment • A parameter descriptive of a reasonable upper bound on the obtainable martensite • Characterization of martensite by virtue of a new developed magnetic device • Investigation of mechanical properties and TRIP effect and the effect of grain size.

36 MATERIALS SCIENCE↗

Formation and mechanism of nanocrystalline AZ91 powders during HDDR processing

Grain sizes of AZ91 alloy powders were markedly refined to about 15 nm from 100 to 160 μm by an optimized hydrogenation-disproportionation-desorption-recombination (HDDR) process. The effect of temperature, hydrogen pressure and processing time on phase and microstructure evolution of AZ91 alloy powders during HDDR process was investigated systematically by X-ray diffraction, optical microscopy, scanning electron microscopy and transmission electron microscopy, respectively. The optimal HDDR process for preparing nanocrystalline Mg alloy powders is hydriding at temperature of 350 °C under 4 MPa hydrogen pressure for 12 h and dehydriding at 350 °C for 3 h in vacuum. A modified unreacted core model was introduced to describe the mechanism of grain refinement of during HDDR process. - Highlights: • Grain size of the AZ91 alloy powders was significantly refined from 100 μm to 15 nm. • The optimal HDDR technology for nano Mg alloy powders is obtained. • A modified unreacted core model of grain refinement mechanism was proposed.

36 MATERIALS SCIENCE↗

Pre-processing related recrystallization behavior in β annealing of a near-β Ti-5Al-5Mo-5V-3Cr-1Zr titanium alloy

Highlights: • Recrystallization behavior of Ti-55531 alloy in β annealing under four typical pre-processing conditions is studied. • Recrystallization nucleation mode is governed by the size and distribution of substructures. • The sequence of phase transformation and β recrystallization significantly affects recrystallized grain size. • The uniform substructures produce a higher fraction of low angle grain boundaries after full recrystallization. - Abstract: Recrystallization behavior of Ti-55531 alloy in β annealing is explored for grain refinement during primary hot working. Four typical pre-processing conditions were adopted to characterize the diverse processing routes of primary hot working: deformation and annealing in the two-phase region, deformation of globularized structure in two-phase region, deformation of lamellar structure in two-phase region and deformation in single β region. The microstructure evolution in β annealing was quantified by Optical Microscopy (OM), Electron Backscattered Diffraction (EBSD), X-ray diffraction (XRD) and quantitative metallography. The results indicate that the nucleation mode of static recrystallization is determined by the size and distribution of substructures prior to annealing which is related to pre-processing route. The uniform substructures in as-deformed globularized structure produce a higher fraction of low angle grain boundaries after full recrystallization. Fine β recrystallized structure can be obtained when β recrystallization is accompanied by α-to-β phase transformation.

36 MATERIALS SCIENCE↗

Microstructural investigation of oxide dispersion strengthened alloy 617 after creep rupture at high temperature

Highlights: • The creep related microstructural evolution of ODS-617 at 950 °C under 25 MPa was investigated. • Dispersion of nanoscale Y-Al-O particles contributed to the superior creep properties of the ODS-617. • The creep mechanism of the ODS-617 at 950 °C under 25 MPa was Coble creep. • Cracks preferentially developed around the large rod-like AlN particles, resulting in the brittle creep fracture mode. - Abstract: Oxide dispersion strengthened (ODS) alloys have been widely considered as candidate structural materials for advanced nuclear reactors because of their good high temperature properties and irradiation resistance. In this study, the creep properties of oxide dispersion strengthened Alloy 617 (ODS-617) was examined at 950 °C under 25 MPa in the air. Conventional Alloy 617 tested under the same conditions was included for comparison. The ODS-617 exhibited superior creep resistance, including a much longer rupture life of 8533 h and a steady-state creep rate of 8.9 × 10{sup −8} (1/s), which was two orders of magnitude lower than that of the conventional alloy. The lower steady-state creep rate was related to the dispersion of nano-sized Y-Al-O particles, which may impede grain boundary diffusion. It seems that cracks preferentially developed near some large rod-like AlN particles, resulting in the brittle creep fracture mode.

36 MATERIALS SCIENCE↗

Formation of omega phase induced by laser shock peening in Ti-17 alloy

Synchrotron-based high-energy X-ray diffraction and transmission electron microscopy were used to study the in-depth compressive residual strains and microstructure evolution in Ti-17 alloy subjected to two-sided laser shock peening (LSP). A noticeable difference was observed in the lattice strains and extents of peak broadening on the two sides of the sample. The compressive residual strain was smaller and the peak broadening was more pronounced on the bottom surface in comparison to the top surface. This is due to the β → ω phase transformation induced by LSP on the bottom surface. It was determined that the ω phase with fine particle morphology and incommensurate structure is located mainly in the 15 μm-thick layer under the bottom surface. The formation of the LSP-induced ω phase on the bottom surface can be attributed to the adiabatic increase in temperature due to the higher shock peak pressure on the bottom surface, and can result in the relaxation of the accumulated residual compressive strain in the original phases.

36 MATERIALS SCIENCE↗

Tracking nickel oxide reduction in solid oxide cells via ex-situ ptychographic nano-tomography

In solid oxide cell (SOC) electrodes, nickel oxide reduction is a fundamental step occurring during the initial stages of cell operation. The reduction procedure is considered to be crucial in producing the final fuel electrode microstructure. This work studies the reduction reaction occurring in a SOC fuel electrode by means of X-ray ptychographic nanotomography. The same microstructure is analyzed in its pristine state and after 1 h of reduction at 850 °C, in dry hydrogen. The microstructure evolution is discussed in terms of changes in volume fractions, surface areas, and connectivity of the different constituent materials. The pore network in the unreduced cermet is revealed to be fully isolated, implying that reduction proceeds inwards from the edges of the sample towards the center. For the given experimental conditions we report that reduction is complete after 1 h of treatment but the resulting nickel network is not yet percolated and detached from the yttria-stabilized zirconia.

25 ENERGY STORAGE↗

Effect of aging on failure mechanism of Alloy 709 at various temperatures

Alloy 709 is a novel austenitic stainless steel with high temperature creep strength, weldability, and corrosion resistance. These properties make the material suitable for applications in the structure of next-generation nuclear power plants. Enduring high temperatures for an extended period of time in the harsh environments of a nuclear power plant results in thermal aging of the material. Therefore, it is imperative to study the effect of thermal aging on the microstructure and mechanical properties of Alloy 709 before its application in the next generation power plants. In this study, hot-processed (forged + rolled), annealed and quenched ingots of Alloy 709 are aged at 650 °C for 2000 h in air and then tested at various temperatures up to 850 °C in an in-situ heating-loading Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectroscopy (EDS) and Electron Backscatter Diffraction (EBSD). The effect of testing temperature and aging on microstructural evolutions during tensile testing of as-received and aged samples are studied. Furthermore, both as-received and aged samples displayed serrations and drop in ductility at elevated temperatures, due to the effect of dynamic strain aging (DSA). The occurrence of DSA activity was found within temperatures range of 500 °C–750 °C in as-received and 550 °C - 650 °C in aged samples. The aged samples showed less elongation accompanied by shallower dimples on the fractured surface, indicating less ductile failure mechanism compared to as-received samples. Failure mechanism observations on the fracture surface using SEM fractography are correlated to the observations made on the sample surface using in-situ SEM to achieve a complementary set of information to better understand the failure mechanism of this novel alloy.

36 MATERIALS SCIENCE↗

Tracking Dendritic Growth in Hydrogen-Based Hematite Reduction via Computer Vision

The reduction of hematite to metallic iron using hydrogen (H2) as a reducing agent presents a promising pathway for decarbonizing steel production. In this study, we employ a combination of in situ confocal scanning laser microscopy (CSLM) and advanced computer vision techniques to quantitatively analyze dendritic growth of ferrite during H2-based reduction of iron oxide at high temperatures. A workflow integrating Watershed Image Segmentation (WIS) and Lucas-Kanade Optical Flow (LKOF) is developed to extract both global and local kinetic information from time-resolved micrograph sequences. H2 reduction experiments conducted at 1400 degrees C and 1500 degrees C demonstrate a clear correlation between temperature and reduction rate, as evidenced by accuracy of fitted Johnson-Mehl-Avrami-Kolmogorov (JMAK) parameters. Optical flow analysis further elucidates the anisotropic and branched nature of dendritic growth, providing spatially resolved velocity fields that correlate well with global transformation kinetics. The proposed methodology demonstrates strong agreement with experimental measurements and literature values, offering a robust framework for automated image-based analysis to study kinetics through microstructural evolution in the reduction of iron ore, and likely other reaction-diffusion phenomena.

08 HYDROGEN↗