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At least 1,693 records · Page 94
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Effect of vanadium on phase composition and hard magnetic properties of as-solidified and heat-treated Sm–Fe–(Ti,V) alloys

Although the intrinsic magnetic properties of Ti-stabilized Sm(Fe,Co,Ti) 12 compounds exhibit potential of excellent rare-earth-lean permanent magnets, it has been much easier to realize large coercivities with the isostructural compounds stabilized by either V or by certain combinations of Ti and V. To elucidate the influence of V on the microstructure and magnetic properties, a series of Sm 8.1 Fe 78.4 (Ti 1-x V x ) 13.5 alloys was studied after arc-melting and annealing at 850–1000 °C. The alloys were found to fall into three groups. For x ≤ 0.2, solidification generates mostly the Sm(Fe,Ti,V) 12 phase, but annealing converts at least part of it into the non-magnetic Sm(Fe,Ti,V) 11 and the magnetically soft Sm 2 (Fe,Ti,V) 17 phases. For 0.2 < x < 0.6, the alloys solidify into a near-equilibrium mixture of the Sm(Fe,Ti,V) 12 , TiFe 2 and Sm-rich phases. For x ≥ 0.6, solidification generates large fractions of α-Fe solid solution and Sm-rich phases; an annealing step is necessary to complete the formation of Sm(Fe,Ti,V) 12 phase. Also, for x ≥ 0.6 the temperature below which the Sm(Fe,Ti,V) 12 phase is stable decreases with x, as does the fraction of this phase formed during solidification. Here, the differences between these three groups of alloys suggest different strategies for developing hard magnetic properties, with the likelihood of a success increasing with increasing x. For x ≥ 0.6, heat treatment alone is demonstrated to generate a microstructure of micron and submicron Sm(Fe,Ti 1-x V x ) 12 crystallites separated by a Sm-rich phase and exhibiting a coercivity with values up to 3.5 and 5.7 kOe for x = 0.8 and 1.0, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ investigation of strengthening and strain hardening mechanisms of Cu-added medium-Mn steels by synchrotron-based high-energy X-ray diffraction

A novel Cu-added medium-Mn steel with a chemical composition of Fe–0.27C–9.1Mn–1.86Al–3.3Cu (wt.%) was designed and subjected to intercritical annealing (IA) temperature range from 620 °C to 680 °C for 1 h. The ultimate tensile strength (UTS) increases and the yielding strength (YS) decreases with the IA temperature increasing. The YS of 824 MPa, UTS of 1222 MPa, total elongation (TE) of 55%, and product of strength and elongation (PSE) of 67.2 GPa·% are achieved after IA at 660 °C. Transmission electron microscopy confirmed that Cu-rich nanoparticles precipitate in the ferrite. The in-situ high-energy X-ray diffraction (HE-XRD) experiments show that at the beginning of plastic deformation, both austenite and ferrite bear the applied load. The load is mainly undertaken by martensite with effective transformation-induced plasticity (TRIP) effect triggered. The YS of ferrite is significantly higher than that of austenite. The individual contribution of solid solution strengthening, grain refinement strengthening, dislocation strengthening, and precipitation strengthening in ferrite and austenite is analyzed. The discrepancy between the YS of ferrite and austenite is mainly attributed to the precipitation strengthening due to the Cu-rich nanoparticles precipitation. The moderate mechanical stability and the collaboration of TRIP and twinning-induced plasticity (TWIP) effects of austenite contributed to the enhanced strain hardening capability and resulted in large ductility.

36 MATERIALS SCIENCE↗
DOE OSTI2026

The effect of varying powder feedstock chemistry and printing atmosphere on the microstructure of additively manufactured nickel-based ODS alloys: Role on stabilization of cellular structures vs. oxide dispersion formation

Nickel-based alloys have a wide variety of structural applications due to their high corrosion resistance and mechanical strength which depend on solid solution strengthening, or the formation of oxides and/or intermetallic precipitation for their properties. In this study, the microstructure of six Ni–Cr–Y–Ti–Al powder batches designed for the production of oxide dispersion strengthened nickel were compared. These batches varied in chemistry and atomization technique used which included Gas Atomization Reactive Synthesis (GARS). The batches of powder were then consolidated via Additive Manufacturing (AM) Powder Bed Fusion using Laser Beam (PBF-LB) and characterized via transmission electron microscopy to elucidate the influence of powder feedstock (i.e. synthesis methodology and chemistry) on the PBF-LB microstructure. The study investigates (i) how the amount of yttrium and titanium additions in the powder feedstock and the addition of oxygen during the processing (through GARS) affect the microstructure of the powder itself and the AM printed microstructure, and (ii) how the control of oxygen addition in the printing atmosphere during the PBF-LB printing process itself is another important parameter for achieving the formation of the wanted oxide dispersion versus the stabilization of the cellular structure (often observed in AM processed alloys). Microstructural characterization of both powder particles and additively manufactured nickel alloys in this study provide important insights into the movement of yttrium within the material upon solidification, particularly along cell boundaries, and how yttrium behaves depending on alloy chemistry. When a threshold of yttrium content is reached within the system, yttrium consistently reacts to form an intermetallic along cell boundaries instead of forming oxide nanoparticles.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Mass transport in a highly immiscible alloy on extended shear deformation

Forced mixing to a single phase or supersaturated solid solution (SSS) and its prerequisite microstructure evolution in immiscible systems has been a focus of research for both fundamental science and a variety of applications. Controlling the formation of SSS by shear deformation assisted processing could enable a material design beyond conventional equilibrium microstructure in immiscible systems. Here, a highly immiscible (mixing enthalpy of ~ 20 kJ·mol -1 ) Cu-50 at. % Cr binary alloy was employed to investigate the microstructure evolution and localized tendencies of SSS during severe shear deformation. The present results demonstrate distinctive microstructure refinement process in each phase and how they lead to localized SSS as a function of shear strain. Preferential dynamic recrystallization occurs in the softer Cu phase due to strain localization, leading to substantial grain refinement. The refinement of Cr phase in the top-most layer, however, is enabled by the progressive evolution of grain lamination, splitting, and fragmentation as a function of shear strain. The eventual SSS is found to be strongly dependent on the local environments that affect the dislocation activity, including the level of microstructure refinement, the interfacial orientation relationship, hardness difference, and supposed stability of oxidation. Ab initio simulations comfirm that it is more favorable to oxidize Cr than Cu at incoherent Cu/Cr interfaces which then limits the mass transport on an incoherent boundary. Our results shed light on the underpinning mechanism for non-equilibrium mass transport in immiscible systems upon severe deformation that can be applicable to a variety of processing techniques aimed at producing immiscible alloys with superior mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Phase Development in a U-7 wt.% Mo vs. Al-7 wt.% Ge Diffusion Couple

Fuel development for the Reduced Enrichment for Research and Test Reactors (RERTR) program has demonstrated that U-Mo alloys in contact with Al develop interaction regions with phases that have poor irradiation behavior. The addition of Si to the Al has been considered with positive results. Compositional modification to replace Si with Ge is now under evaluation to attempt to further improve irradiation behavior. In this study, the microstructural and phase development of a diffusion couple of U-7 wt.% Mo in contact with Al-7 wt.% Ge was examined by transmission electron microscopy, scanning electron microscopy and energy dispersive spectroscopy. The interdiffusion zone developed a microstructure that included the cubic-UGe3 phase and amorphous phases. The UGe3 phase was observed with and without Mo and Al solid solutioning developing a (U,Mo)(Al,Ge)3 phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Thermodynamic assessment of the U–La–O system

The CALPHAD methodology was used to develop a thermodynamic assessment of the U-La-O system. The solid solution and liquid phases are described with the compound energy formalism and the partially ionic two-sublattice liquid model respectively. A density functional theory (DFT) calculation for the lattice stability of the fictive lanthanum oxide fluorite structure compound is used to determine the Gibbs energies for the La containing end-members in the CEF model for U 1-y La y O 2+x . Experimental thermodynamic and phase equilibria data were then used in optimizations to develop representations of the phases in the system that can be extended to include other actinide and fission products to develop multi-component models. The models that comprise this assessment very well reproduce experimentally determined oxygen potentials and the observed phase relations for the U-La-O system.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Modelling the thermal conductivity of (U x Th 1-x )O 2 and (U x Pu 1-x )O 2

The degradation of thermal conductivity due to the non-uniform cation lattice of (U x Th 1-x )O 2 and (U x Pu 1-x )O 2 solid solutions has been investigated by molecular dynamics, using the non-equilibrium method, from 300 to 2000 K. Degradation of thermal conductivity is predicted in (U x Th 1-x )O 2 and (U x Pu 1-x )O 2 as compositions deviate from the pure end members: UO 2 , PuO 2 and ThO 2 . The reduction in thermal conductivity is most apparent at low temperatures where phonon-defect scattering dominates over phonon-phonon interactions. The effect is greater for (U x Th 1-x )O 2 than U x Pu 1-x )O 2 due to the greater mismatch in cation size. Parameters for an analytical expressions have been developed that describe the predicted thermal conductivities over the full temperature and compositional ranges. Finally, these expressions may be used in higher level fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Experimental oxygen potentials of U 1-y Pr y O 2± x and thermodynamic assessment of the U-Pr-O system

Thermogravimetric analysis (TGA) was used to determine the oxygen potentials of fluorite urania-praseodymia (U 1-y Pr y O 2± x ) solid solutions for y = 0.10 and 0.20 between 1000 and 1500 °C. A thermodynamic assessment of U-Pr-O system was performed using the CALPHAD (CALculation of PHAse Diagrams) method. Furthermore, the models well reproduce the TGA measurements and the computed phase relations are in good agreement with those proposed from an X-ray diffraction investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Interstitial migration behavior and defect evolution in ion irradiated pure nickel and Ni-xFe binary alloys

Transition from long-range one-dimensional to short-range three-dimensional migration modes of interstitial defect clusters greatly reduces the damage accumulation in single-phase concentrated solid solution alloys under ion irradiation. A synergetic investigation with experimental, computational and modeling approaches revealed that both the resistance to void swelling and the delay in dislocation evolution in Ni-Fe alloys increased with iron concentration. This was attributed to the gradually increased sluggishness of defect migration, which enhances interstitial and vacancy recombination. Transition from long-range one-dimensional defect motion in pure nickel to short-range three-dimensional motion in concentrated Ni-Fe alloys is continuum, not abrupt, and within an iron concentration range up to 20%. Lastly, the gradual transition process can be quantitatively characterized by the mean free path of the interstitial defect clusters.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of the structure and chemical nature of Pd fission product agglomerations in irradiated TRISO particle SiC

Tristructural-isotropic (TRISO)-coated fuel particles are used in high-temperature gas-cooled nuclear reactors. Although the polycrystalline 3C-SiC layer acts as the main barrier to fission product release, post-irradiation examinations have shown that certain metallic fission products are found outside of the TRISO coated particle, with no observable micro-cracks or other mechanical failure feature. In this study, an atomic resolution transmission electron microscopy investigation of a SiC layer of a neutron irradiated (19.38% fissions per initial metal atom average burnup; 1072°C time-averaged temperature) TRISO-coated particle from the first advanced gas reactor experiment at Idaho National Laboratory containing mainly Pd and U was conducted. The fission product Pd was found to be present at dislocation cores associated with twins, stacking faults, and their intersections. The Pd present at the dislocation cores were bonded to Si, suggesting that the diffusion of Pd along the SiC dislocation cores involves the formation of Pd-Si bonds. Evidence of single atom grain boundary migration was also found. Fission product agglomerations at grain boundaries most likely consist of a random solid solution containing Si-FP, Si-C, and C-C type bonding arrangements. No difference in the Si- or C-bonding environment is found for the unoccupied grain boundary adjacent to a fission product agglomeration compared to the bulk SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Plasma and X-ray radiation-induced damage mechanisms in a tungsten heavy alloy

Tungsten-based materials are among the main candidate plasma-facing materials for future fusion reactors where extreme conditions of thermomechanical stresses, radiation, and energetic particles must be withstood. Here a hybrid X-pinch platform combined with a pulsed power machine was used to study damage mechanisms in 90W–4Cu–6Ni tungsten heavy alloy as a result of exposure to high energy density plasma, X-ray radiation, and the resulting thermal shock. It was found that the plasma-facing surface of the material undergoes severe melting and ablation with the extent of ablation being higher in the Cu–Ni phase than the W particles. Rapid melting and solidification transform the W particles into fine dendrite morphologies embedded in the Cu–Ni solid solution. Cracking was also observed in the subsurface layers at the interfaces between the W particles and the Cu-rich phase of the Cu–Ni matrix. X-ray radiation, on the other hand, cause slight melting on the surface without significant ablation and no cracking.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tensile properties of powder-metallurgical-processed tungsten alloys after neutron irradiation near recrystallization temperatures

The tensile properties of powder-metallurgical-processed Pure W, K-doped W, W-3%Re, and K-doped W-3%Re were examined after neutron irradiation up to 0.7 dpa at 910–1020 °C with a thermal neutron shield in the High Flux Isotope Reactor (HFIR). After irradiation, recrystallized Pure W (R) exhibited a brittle fracture mode, while recrystallized K-doped W-3%Re (R) exhibited a ductile fracture mode at 500 °C. K-doped W-3%Re (R) has fine grains, and hence, contains a considerable number of grain boundaries that act as sinks for irradiation defects. Solid solute Re in the W matrix could improve not only the mechanical properties of W, but also its resistance to neutron irradiation. At 500 °C, the ductility of K-doped W-3%Re after irradiation was significantly higher than that of Pure W. The irradiation at ~1000 °C did not induce hardening of stress-relieved (SR) W materials, but SR W materials tended to exhibit a decrease in the ultimate tensile strength (UTS) and an increase in total elongation (TE). Here, the softening due to the recovery and recrystallization of SR W materials and the hardening due to the formation of irradiation defect clusters were balanced during irradiation at ~1000 °C, and ductility was exhibited without an increase in strength.

36 MATERIALS SCIENCE↗
DOE OSTI2020

(S)TEM/EDS study of native precipitates and irradiation induced Nb-rich platelets in high-burnup M5®

We have investigated microstructure and microchemistry of precipitates and dislocation loops in high-burnup M5® using (scanning) transmission electron microscopy ((S)TEM) equipped with energy dispersive X-ray spectroscopy (EDS). Two (S)TEM lamellae were made by cryo-FIB from the same cladding sample. The Nb-rich native precipitates were found in the metal, in the suboxide and in the oxide. Upon diffraction analysis, most of the Nb-rich native precipitates in the metal matrix remain as β-Nb phase, while no β-Nb native precipitates were found in the oxide. Nearby the oxide and metal (O/M) interface, the native precipitates in the oxide were already oxidized into t-NbO2 phase. At further distance away from the O/M interface, the Nb-rich native precipitates were gradually oxidized and became amorphous. Besides the native precipitates, Nb-rich irradiation induced precipitates (IIPs) were found in the metal matrix. Using g = <0002> vector for imaging, the length of the IIPs was aligned with dislocation loops. However, no Nb segregation to the dislocation loops themselves was observed. For the first time, we report that IIPs indeed exist in the oxide but only within about 1.5 µm away from O/M interface. However, the oxidation state of the IIPs in the oxide is still unclear. The presence of both native precipitates and IIPs in the oxide may indicate the Nb concentration in the oxide solid solution remain low nearby the O/M interface, which may explain the reduced corrosion kinetics of in-pile M5®. On the other hand, no IIPs were observed in the oxide at further distance and this may indicate that they have eventually dissolved back into the oxide. A few mechanisms related to IIPs stability in the oxide are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

An Atom Probe Tomography Study of the Through Wall Attenuation Effect on Cu-rich Precipitate Formation in a Reactor Pressure Vessel Steel

In this study, high-Cu weld material harvested from an ex service reactor pressure vessel (RPV) steel from Unit 1 of the decommissioned Zion Nuclear Generating Station has been characterized using atom probe tomography. Samples taken from 4 different positions through the thicknesses of the pressure vessel wall from the water-side to the air-side were characterized, along with an unirradiated baseline material. In the baseline material, no precipitates were found and the Cu was observed to be fully in solid solution; however, scanning transmission electron microscopy combined with energy dispersive spectroscopy (STEM-EDS) revealed the presence of ϵ -Cu that form during processing of the material and results in the concomitant decrease of matrix Cu. Following irradiation, a high number density of nano-scale Cu-rich precipitates (CRPs) were observed, uniformly distributed throughout the matrix. The Cu content within the CRPs was found to be ~30-35 at.% regardless of location in the wall. No statistically significant variation in the compositions, mean radius, number density, or volume fraction as a function of location within the wall was observed. The measured matrix Cu level excluding CRPs contribution was found to be ~90 appm higher than the solubility limit suggesting that further nucleation and growth of the CRPs under continued operations would have occurred. These results clearly demonstrate that the neutron energy attenuation has no significant effect on the precipitation kinetics of CRPs regardless of location in the wall in high-Cu RPV steels under irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High radiation tolerance of an ultrastrong nanostructured NiCoCr alloy with stable dispersed nanooxides and fine grain structure

The present paper reports the irradiation behavior of nanostructured NiCoCr medium entropy alloy enhanced by Y-Hf-O nanooxides and fine grains produced by powder metallurgy. Ion beam irradiation to a peak dose of 130 dpa at 580°C produced neither detectable void swelling nor irradiation-induced hardening, in contrast to significant void swelling and property degradation of single phase NiCoCr concentrated solid-solution alloy under the similar irradiation condition. The dispersed nanooxides and high density of grain boundaries act as defect sinks that effectively annihilate irradiation-induced interstitials and vacancies. Furthermore, high-density nanooxides maintain their crystalline structures and stabilize the grain boundaries during irradiation.

36 MATERIALS SCIENCE↗