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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

4D evolution of Cr 23 C 6 precipitates in neutron-irradiated and annealed HT-UPS steel observed via synchrotron micro-computed tomography

High-temperature-ultrafine precipitate strengthened (HT-UPS) steel is a potential structural material for advanced nuclear reactors; however, its irradiation response is not well understood. This research provides insight into irradiation-induced effects, such as precipitate evolution mechanisms and four-dimensional morphological evolution, in HT-UPS steel using synchrotron micro-computed tomography. Identical specimens were characterized pre-irradiation and post-irradiation following neutron exposure up to 0.3 displacements per atom at 600 °C. Irradiation effects were also differentiated from the annealing response of precipitates. Following neutron irradiation, the average Cr 23 C 6 precipitate size reduced, affected by the synergy of nucleation and growth, ballistic dissolution, and inverse coarsening, which was observed at fluences an order of magnitude lower than previously observed. Annealing at 600 °C for 32 h increased the average Cr 23 C 6 precipitate size and decreased the phase fraction, attributed to precipitate coarsening. The precipitate morphology evolution and resultant mechanisms can be utilized to parameterize and validate microstructural models simulating radiation damage or annealing.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Hydrogen-based ore-to-part manufacturing of near-net-shape stainless steel

Decarbonizing iron and steelmaking, combined with global disruptions to raw material supply chains, necessitates novel approaches to iron and steel production. In this work, we demonstrate a direct ore-to-part manufacturing route using a mixture of ore-derived oxide powders of Fe 2 O 3 , Cr 2 O 3 , NiO, and MoO 3 as feedstock for additive manufacturing, combined with sintering under H 2 to produce a near-net-shape austenitic stainless-steel. Complete reduction of all constituent oxides, including MoO 3 and Cr 2 O 3 , is achieved in-situ at 1300 °C, resulting in dense, crack-free bulk alloy. The fabricated part retains geometric fidelity while undergoing substantial volumetric shrinkage inherent to redox and sintering. Thermodynamic calculations elucidate the co-reduction mechanisms and alloying pathways that enable complete metallization. This work is the first demonstration of net-shaping metal parts directly from ore derived oxides, and this ore-to-part approach can minimize the emissions and lead time for manufacturing associated with downstream processing such as rolling, forging, and machining.

Yang, Mingzhang [Univ. of Waterloo, ON (Canada); F↗
DOE OSTI2021

Diffusion-mediated chemical concentration variation and void evolution in ion-irradiated NiCoFeCr high-entropy alloy

High-entropy alloys (HEAs) are proposed as potential structural materials for advanced nuclear systems, but little is known about the response of matrix chemistry in HEAs upon irradiation. Here, we reveal a substantial change of matrix chemical concentration as a function of irradiation damage (depth) in equiatomic NiCoFeCr HEA irradiated by 3 MeV Ni ions. After ion irradiation, the matrix contains more Fe/Cr in depth shallower than ~900–1000 nm but more Ni/Co from ~900–1000 nm to the end of the ion-damaged region due to the preferential diffusion of vacancies through Fe/Cr. Preferential diffusion also facilitates migration of vacancies from high radiation damage region to low radiation damage region, leading to no void formation below ~900–1000 nm and void formation around the end of the ion-damaged region at a fluence of 5 × 10 16 cm –2 (~123 dpa, displacements per atom, peak dose under full cascade mode). As voids grow significantly at an increased fluence (8 × 10 16 cm –2 , 196 dpa), the matrix concentration does not change dramatically due to new voids formed below ~900–1000 nm.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Diffusion-mediated chemical concentration variation and void evolution in ion-irradiated NiCoFeCr high-entropy alloy

High-entropy alloys (HEAs) are proposed as potential structural materials for advanced nuclear systems, but little is known about the response of matrix chemistry in HEAs upon irradiation. Here, we reveal a substantial change of matrix chemical concentration as a function of irradiation damage (depth) in equiatomic NiCoFeCr HEA irradiated by 3 MeV Ni ions. After ion irradiation, the matrix contains more Fe/Cr in depth shallower than ~900–1000 nm but more Ni/Co from ~900–1000 nm to the end of the ion-damaged region due to the preferential diffusion of vacancies through Fe/Cr. Preferential diffusion also facilitates migration of vacancies from high radiation damage region to low radiation damage region, leading to no void formation below ~900–1000 nm and void formation around the end of the ion-damaged region at a fluence of 5 × 10 16 cm -2 (~123 dpa, displacements per atom, peak dose under full cascade mode). As voids grow significantly at an increased fluence (8 × 10 16 cm -2 , 196 dpa), the matrix concentration does not change dramatically due to new voids formed below ~900–1000 nm.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

A composite electrodynamic mechanism to reconcile spatiotemporally resolved exciton transport in quantum dot superlattices

Quantum dot (QD) solids are promising optoelectronic materials; further advancing their device functionality requires understanding their energy transport mechanisms. The commonly invoked near-field Förster resonance energy transfer (FRET) theory often underestimates the exciton hopping rate in QD solids, yet no consensus exists on the underlying cause. In response, we use time-resolved ultrafast stimulated emission depletion (STED) microscopy, an ultrafast transformation of STED to spatiotemporally resolve exciton diffusion in tellurium-doped cadmium selenide–core/cadmium sulfide–shell QD superlattices. We measure the concomitant time-resolved exciton energy decay due to excitons sampling a heterogeneous energetic landscape within the superlattice. The heterogeneity is quantified by single-particle emission spectroscopy. This powerful multimodal set of observables provides sufficient constraints on a kinetic Monte Carlo simulation of exciton transport to elucidate a composite transport mechanism that includes both near-field FRET and previously neglected far-field emission/reabsorption contributions. Uncovering this mechanism offers a much-needed unified framework in which to characterize transport in QD solids and additional principles for device design.

36 MATERIALS SCIENCE↗
DOE OSTI2026

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

The daily as-run thermal analysis for the Advanced Gas Reactor (AGR)-5/6/7 experiment was documented in ECAR-5633 (Hawkes 2021). A finite element model was created for the entire test train, which consisted of five capsules. The fuel compacts, graphite holders, stainless-steel capsule walls, and all other major components were individually modeled, including all thermocouples (TCs). Daily heat rates for each compact and component in the model were input from daily as-run physics analyses. Daily gas compositions for each capsule were also input. The thermal conductivity of the compacts and graphite holders varied with fast neutron fluence. Gas mixture thermal conductivity was implemented using experimentally attained values from the literature. Fluence and temperature-dependent thermal conductivity was used for the graphite components and the fuel compacts. Radiation heat transfer was implemented. The model was tuned in an attempt to match the TC readings during the first cycle by adjusting the Neolube (graphite lubricant) thickness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Impacts of point defects on shallow doping in cubic boron arsenide: A first principles study

Cubic boron arsenide (BAs) stands out as a promising material for advanced electronics, thanks to its exceptional thermal conductivity and ambipolar mobility. However, effective control of p- and n-type doping in BAs poses a significant challenge, mostly as a result of the influence of defects. In the present study, we employed density functional theory (DFT) to explore the impacts of the common point defects and impurities on p-type doping of Be B and Si As , and on n-type doping of Si B and Se As . We found that the most favorable point defects formed by C, O, and Si are C As , O B O As , Si As , C As Si B , and O B Si As , which have formation energies of less than 1.5 eV. While the O impurity detrimentally affects both p- and n-type dopings, C and Si impurities are harmful for n-type dopings, making n-type doping a potential challenge. Interestingly, the antisite defect pair A s B B As benefits both p- and n-type doping. Finally, the doping limitation analysis presented in this study can potentially pave the way for strategic development in the area of BAs-based electronics.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Nanoparticle localization within chiral liquid crystal defect lines and nanoparticle interactions

Self-assembly of colloidal particles into predefined structures is a promising way to design inexpensive manmade materials with advanced macroscopic properties. Doping of nematic liquid crystals (LCs) with nanoparticles has a series of advantages in addressing these grand scientific and engineering challenges. It also provides a very rich soft matter platform for the discovery of unique condensed matter phases. The LC host naturally allows the realization of diverse anisotropic interparticle interactions, enriched by the spontaneous alignment of anisotropic particles due to the boundary conditions of the LC director. Here we demonstrate theoretically and experimentally that the ability of LC media to host topological defect lines can be used as a tool to probe the behavior of individual nanoparticles as well as effective interactions between them. LC defect lines irreversibly trap nanoparticles enabling controlled particle movement along the defect line with the use of a laser tweezer. Minimization of Landau–de Gennes free energy reveals a sensitivity of the ensuing effective nanoparticle interaction to the shape of the particle, surface anchoring strength, and temperature, which determine not only the strength of the interaction but also its repulsive or attractive character. Theoretical results are supported qualitatively by experimental observations. Furthermore, this work may pave the way toward designing controlled linear assemblies as well as one-dimensional crystals of nanoparticles such as gold nanorods or quantum dots with tunable interparticle spacing.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Options for Achieving Cost Reduction in Advanced Reactors through Open Architecture

A key contributor to high capital costs and schedule overruns for new nuclear power plants is lack of standardization, driven by site-specific customization and construction of multiple designs by competing vendors rather than commitment to a single standardized program. While advanced reactor vendors typically individually target repeat construction of standardized units, the many competing designs could exacerbate the problem. “Open Architecture”, the open specification of requirements and interfaces for structures, systems and components (SSCs), has been proposed as a means of promoting standardization, by facilitating existing non-nuclear suppliers to enter the industry and/or allowing SSCs to be configured for more than one reactor within the same technology type. Contracting mechanisms that facilitate information sharing and alignment of incentives between stakeholders may complement such an approach. A preliminary scheme is presented for selection of SSCs for which such strategies could be adopted, based on a vendor make/buy decision model and stakeholder interviews. SSCs are categorized according to number of suppliers and their contribution to the reactor’s competitive edge. SSCs with many potential suppliers and a high contribution to competitive edge may be attractive for widening the supply chain via open specification of system requirements and interfaces, e.g., SSCs in the power island. SSCs with few suppliers and low contribution to competitive edge may be potential avenues for common system specification between vendors, e.g., some of the auxiliary SSCs. Potential cost reductions from such strategies will depend upon the size of the build program and the reactor type.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Ribbon Wrinkling Origin and Its Effect on Tape-Wound Core Shape

Improved understanding of advanced manufacturing techniques is needed to fabricate next- generation soft magnetic tape-wound cores. These utilize metal amorphous nano-composites (MANCs) in the form of long thin ribbons. MANCs exhibit attractive soft magnetic properties for use in high frequency and high power density magnetic components. Of particular interest here, the permeability and hysteresis of MANC ribbons can be tuned in a strain annealing process. An unexpected consequence is that wider ribbons develop longitudinal wrinkle patterns. This can substantially reduce the tape-wound core stacking factor. However, when winding a wrinkled ribbon into a toroidal core, the wrinkles can completely flatten, or crinkle. By this phenomenon, the stacking factor can be recovered. To understand the crinkling effect we have developed analytical and finite element analysis (FEA) models. Experiments with known boundary conditions demonstrate the transition from wrinkled to crinkled versus winding curvature. The models utilize an elastic modulus measured via ribbon tensile testing and a ribbon thickness value measurement made by profilometry. The analytical model provides insight into the crinkling effect but cannot explain the length of the transition region from wrinkling to crinkling. With FEA, we successfully predict the transition region and depending on wrinkling amplitude can determine when a ribbon will wind into a useable core. Such mechanics-based insight provides critical feedback for allowable processing parameters in the strain annealing process.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2025

Singly occupied 4⁢𝑓 antiferromagnetic insulators: CePO 4 and CeVO 4

Rare-earth containing wide band gap oxides, which provide spin-photon interface and narrow linewidth optical emission, are getting significant attention as the most promising candidate materials in advancing quantum transduction and memories. Here, in this work, from ab initio calculations, we identify antiferromagnetic ground states in structurally preferred monoclinic CePO 4 and tetragonal CeVO 4 exhibiting localized occupied and unoccupied Ce 4⁢𝑓 states with 4⁢𝑓−4⁢𝑓 transition characteristics. Interestingly, in CePO 4 , O 2⁢𝑝 and P 3⁢𝑝 states hybridize negligibly with Ce 4⁢𝑓 states, while in CeVO 4 , V 3⁢𝑑 and O 2⁢𝑝 states hybridize and appear as extended states in between the occupied and unoccupied Ce 4⁢𝑓 states. Here, phonon calculations and analysis identify and differentiate Raman active phonon modes along with the spin phonon coupling of Ce in both CePO 4 and CeVO 4 that ultimately lead to different 4⁢𝑓 ground-state crystal-field multiplets, which are critical to accurately describe electronic transitions for foundational quantum transduction and memories. Further, the identified 𝐶 1 site symmetry of Ce, lacking inversion symmetry in CePO 4 , is relevant for quantum memories and 𝐷 2⁢𝑑 site symmetry of Ce exhibiting inversion symmetry in CeVO 4 is relevant for quantum transduction.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

Roadmap for Photonics with 2D Materials

Triggered by advances in atomic-layer exfoliation and growth techniques, along with the identification of a wide range of extraordinary physical properties in self-standing films consisting of one or a few atomic layers, two-dimensional (2D) materials such as graphene, transition metal dichalcogenides (TMDs), and other van der Waals (vdW) crystals now constitute a broad research field expanding in multiple directions through the combination of layer stacking and twisting, nanofabrication, surface-science methods, and integration into nanostructured environments. Photonics encompasses a multidisciplinary subset of those directions, where 2D materials contribute remarkable nonlinearities, long-lived and ultraconfined polaritons, strong excitons, topological and chiral effects, susceptibility to external stimuli, accessibility, robustness, and a completely new range of photonic materials based on layer stacking, gating, and the formation of moiré patterns. These properties are being leveraged to develop applications in electro-optical modulation, light emission and detection, imaging and metasurfaces, integrated optics, sensing, and quantum physics across a broad spectral range extending from the far-infrared to the ultraviolet, as well as enabling hybridization with spin and momentum textures of electronic band structures and magnetic degrees of freedom. The rapid expansion of photonics with 2D materials as a dynamic research arena is yielding breakthroughs, which this Roadmap summarizes while identifying challenges and opportunities for future goals and how to meet them through a wide collection of topical sections prepared by leading practitioners.

2D materials↗
DOE OSTI2020

Fabrication and thermophysical properties of UO 2 -UB 2 and UO 2 -UB 4 composites sintered via spark plasma sintering

Uranium dioxide (UO 2 ) composites with uranium diboride (UB 2 ) and uranium tetraboride (UB 4 ) have been proposed as advanced fuel candidates due to their high thermal conductivity, high melting point, high fissile density and their ability to incorporate a built-in burnable poison by tailoring the targeted 10 B/ 11 B ratio. As such, it is important to assess the fabrication, and thermal and micromechanical properties of such composites. In this work, UO 2 -UB 2 and UO 2 -UB 4 samples with boride phase fractions of 5, 15 and 30 wt% were fabricated to high densities (above 95 % theoretical density) via spark plasma sintering (SPS). This enabled sintering at relatively low temperatures and short timescales. SPS also aided in maintaining the target phase fractions of the samples as reactions between the constituent phases were suppressed due to the short timescales and reducing environment during sintering. Here, thermal diffusivity measurements from 299 to 1273 K were conducted through laser flash analysis (LFA). The diffusivity increased as a function of boride weight fraction, and UB 2 additions increased the thermal diffusivity of the composites more than UB 4 additions. Assessment of the LFA results indicated in-situ reactions between the UO 2 and boride phases that suppress the thermal diffusivity occur above 800 K for all samples. Oxidation of the boride phase was proposed as the underlying reaction. This was supported by thermodynamic assessments from the literature, as well as microstructural, crystallographic, and nanoindentation characterization performed on these samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UN-SiC TRISO Post Irradiation Examination Developmental Work

As part of efforts to strengthen INL?s post irradiation analysis capabilities with non-Advanced Gas Reactor (AGR) tristructural isotropic (TRISO) fuels, two developmental activities were conducted. The first activity was to determine how best to analyze uranium nitride TRISO fuel kernels using electron probe microanalysis, while the second activity focused on developing a method to deconsolidate TRISO fuel particles that have been encased in a silicon carbide matrix. Because these two activities were unrelated, they have been presented separately in this report. Initial EPMA analyses showed nitrogen contents that far exceeded the concentration expected for UN--a line compound. Further examination showed that current literature values for the mass absorption coefficient (MAC) for the N ka X-ray absorbed by U ranged from approximately 1600 to 9500, with most values tending toward 9500. Measuring UN with five different progressively increasing accelerating voltages followed by using the modeling program xMAC suggests the actual MAC is approximately 2115. Additional MAC modifications were required to produce reasonable analytical results. Because of the inaccuracies of necessary MAC coefficients, UN analysis via scanning electron microscopy (SEM) is likely to produce inaccurate results. This is because SEM software does not typically allow the user to alter MACs. Tests have been performed to examine the feasibility of an electrochemical technique to liberate irradiated TRISO fuel from a SiC matrix without damaging the outer pyrolytic carbon layer of the fuel particle. The method is performed by electrochemically exposing the SiC to magnesium metal forming Mg2Si and C. Following exposure, the small SiC samples showed slight mass increases with no evidence of conversion to Mg2Si and C nor obvious degradation of the SiC samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The effect of Zr on precipitation in oxide dispersion strengthened FeCrAl alloys

Oxide dispersion strengthened (ODS) FeCrAl alloys are promising candidate materials for advanced nuclear reactor applications requiring high-temperature strength, corrosion resistance, and irradiation tolerance. As these alloys have increased in compositional complexity through attempts to use highly reactive elements such as Zr to refine particle sizes and optimize nanoprecipitate dispersion characteristics, much debate has ensued as to the effects of these alloying element additions on alloy properties. In an attempt to reconcile differences in nanoprecipitate distributions reported in the literature over the past decade, a detailed investigation of a recently developed ODS FeCrAl alloy with nominal composition Fe–10Cr-6.1Al-0.3Zr+0.3Y 2 O 3 is presented here using a combination of atom probe tomography (APT), scanning/transmission electron microscopy (S/TEM), and computational thermodynamics modeling. It is illustrated that based on the amount of Zr available in the lattice, Zr competes with Al and Cr to form carbides and nitrides as opposed to oxygen-rich precipitates. This alloy system has a high number density (>10 23 m -3 ) of ~2–4 nm diameter (Y,Al,O)-rich nanoprecipitates, but it is shown that due to the compositional spread and unknown partitioning of Al between the matrix and precipitates, significant challenges still exist for quantifying the exact compositions of these precipitates using APT. However, the noted compositional spread is supported by identified complex oxides yttrium aluminum monoclinic (YAM) and yttrium aluminum garnet (YAG) using S/TEM. As a result of these findings, researchers developing ODS FeCrAl with reactive element additions must pay careful attention to C and N impurities when optimizing reactive element additions.

electron microscopy↗
DOE OSTI2025

Temperature and dose effects on dislocation loops in self-ion irradiated high-purity iron

Body-centered cubic (BCC) Fe-based alloys are promising candidate materials for advanced nuclear reactors. However, a detailed understanding of irradiation induced dislocation loop microstructure development remains unresolved. It is a widespread belief that 〈001〉 loops become increasingly favorable over ½〈111〉 loops as irradiation temperature rises above ∼300 C. Unfortunately, the temperature effects on 〈001〉 loop have been primarily examined in in-situ irradiation on TEM thin foils but poorly explored on bulk Fe due to exceedingly limited experimental studies on bulk specimens, raising concerns about the potential influence of TEM thin foil artifacts on observed results. Here, in this study, we conducted experiments on ultra-high purity BCC Fe specimens irradiated with 6.7–8 MeV Fe ions over a wide temperature range on bulk samples. We investigated the effects of temperature (T irr = 250–500 °C), dose rate (10⁻⁵ to 10⁻³ dpa/s), and dose (0.35 to 3.5 dpa) on the formation and evolution of 〈001〉 and ½〈111〉 loops as well as cavity (void) formation. Post-irradiation Burgers vector analysis via g•b method on dislocation segments and loops revealed that 〈001〉 loop fraction does not show a monotonic positive correlation with irradiation temperature. Combined with previous and current theoretical as well as experimental findings, we explore the temperature effects on all existing models of 〈001〉 loop formation. We conclude that the prevailing reports regarding the dominance of 〈001〉 loops in Fe at elevated temperatures are mainly attributable to the loss of glissile ½〈111〉 clusters in TEM thin foil experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The mechanism behind the high radiation tolerance of Fe-Cr alloys

With the great demand for high radiation tolerant materials for advanced nuclear energy technologies, Fe-Cr alloys are at the forefront with long standing validated performance. Yet, the real mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here we applied depth resolved atomic scale probe of defects to uncover the real mechanism on how Cr improves radiation resistance and explain the controversial impact of increasing Cr percentage. By combining depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy we investigated the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or a’ phase with high Cr composition. The study reveals that the well-known resistance to radiation in Fe-Cr alloys arises from the stabilization of vacancy clusters around Cr atoms which act as sinks for radiation-induced defects. Thus, Cr atoms do not provide a direct sink for interstitials; rather defect complexes for that consist of Cr atoms and vacancies in turn act as sinks for irradiation-induced vacancies and interstitials. Most importantly, we find that lower amounts of Cr create smaller, uniformly distributed defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous in size and spatial distribution, resulting in less efficient damage recombination. No evidence of phase a’ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Report Summarizing the Mechanical Properties of a Large ODS Ferritic Alloy Ingot by Forging at High Temperatures for Future Mother Tube Production

Oxide dispersion strengthened (ODS) ferritic alloys are considered the benchmark fuel cladding and core structural material in advanced nuclear energy reactors that require high-temperature strength and creep properties and resistance to radiation damage. The ODS ferritic alloys including 14YWT are produced by mechanical alloying (MA), which is time-consuming and is associated with high manufacturing costs that are not beneficial to being used in advanced nuclear energy reactors. This report summarizes the accomplishments in FY24 in the INM program for producing an ODS ferritic alloy by high-deformation, high-temperature processing of reactive and ferritic alloy powders for advanced reactor fuel cladding applications. Following four hot forging experiments, it was concluded that several significant challenges were encountered that were difficult to overcome. These challenges were related to using high annealing temperatures for pressure assisted sintering processes to produce dense microstructures, but the high annealing temperatures result in long range diffusion of Ti through the bcc Fe lattice and react with the YIG particles that are distributed on the prior surfaces of the ferritic alloy powders. In addition, it was determined that the high deformations induced by forging at high temperatures were not very effective for fracturing the reactive YIG particles into smaller particles and distributing them into the interior of the ferritic alloy powders. Spark plasma sintering was attempted for shortening the time at high temperatures, which led to full densification but the microstructure characterization results still showed that Ti atoms diffused over long distances in the bcc Fe lattice and reacted with the YIG particles. Finally, a short 30 minute high intensity ball milling experiment was conducted on blended 14WT and YIG powders for inducing severe deformations that caused the initially spherical powders of 14WT decorated with the YIG particles on the surfaces to transform to flakes and fracturing of YIG particles into smaller particles that were incorporated into the bcc Fe lattice. Two forgings were performed with the high intensity milled powders: first annealing for 20 minutes at 850ºC followed by forging and second annealing for 20 minutes at 1,100ºC followed by forging. At 850ºC, Ti atoms are effectively immobile, thus allowing for densification albeit incomplete. The deformation by forging after annealing for 20 minutes at 1,100ºC favored dynamic recrystallization processes that led to nano-size grains with very high stored energy due to high dislocation density. The corresponding VH data showed significant hardening that correlated with estimated strengthening of ~1650 MPa. This hybrid processing approach combining high intensity ball milling of powder with annealing and forging showed the most promise for producing an oxide dispersing strengthened ferritic alloy.

36 MATERIALS SCIENCE↗