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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

SOFC interconnect barriers and methods of making same using masks

A novel method to produce thin films spatially disposed on desired areas of workpieces is disclosed. Examples of include the formation of a yttria stabilized zirconia (YSZ) film formed on a desired portion of a stainless steel interconnect for solid oxide fuel cells by Atomic Layer Deposition (ALD). A number of methods to produce the spatially disposed YSZ film structures are described including polymeric and silicone rubber masks. The thin film structures have utility for preventing the reaction of glasses with metals, in particular alkali-earth containing glasses with ferritic stainless steels, allowing high temperature bonding of these materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Exploring Fission–Fusion Synergies to Accelerate Compatibility Understanding

To address the significant commercial interest in fusion energy, it will be necessary to accelerate the compatibility research associated with liquid breeders including Li, eutectic Pb–Li and LiF-BeF 2 (FLiBe) molten salt. Particularly for FLiBe, compatibility understanding is limited especially for fusion relevant materials such as reduced activation ferritic-martensitic steels, SiC and V alloys. The historical knowledge associated with molten salt reactors (MSRs) and recent work to commercialize MSRs can benefit fusion research. Recent experimental and modeling work has improved understanding and this knowledge can be applied to fusion relevant materials. For liquid metals (LMs), the comparisons to Li and Pb–Li are less direct but nevertheless can help guide the pathway toward commercialization. For Pb–Li, Al-rich coatings have been shown to inhibit dissolution and potentially increase operating temperatures. For commercialization, the experience with sensors and on-line cleanup can help guide future developments. Thus, it is worth considering the potential for fission-related research with LMs and molten salts to help accelerate fusion research.

36 MATERIALS SCIENCE↗
DOE OSTI2020

An Alternative Casting Technique to Improve the Creep Resistance of Cast INCONEL Alloy 740H

The increasing performance requirements of power plant designs, such as advanced-ultra supercritical (A-USC), require the use of Ni-based superalloys to replace high-strength, ferritic-martensitic steels for components subjected to temperatures above 898 K (625 °C) and for austenitic stainless steels components at temperatures above 973 K (700 °C). To date, commercial Ni-based superalloy INCONEL 740H has been shown to be appropriate for use in A-USC power plants as boiler components in a wrought product. However, large complex components in boilers as well as other casings in the turbine and valve chest require castings of a thick-wall nature. Using the alloy in its cast form would be significantly valuable in terms of range of component size, geometry and complexity. Previous investigations revealed short creep lives from cast INCONEL alloy 740H. Here in this investigation, an alternative casting route that utilized a melt procedure resulting in a fine-grain casting, and in conjunction with a computationally optimized homogenization heat treatment, not only controlled the grain size and grain boundary structure but minimized chemistry variability and segregation. A primarily equiaxed and homogenous grain size distribution was obtained from this approach with better repartition of M 23 C 6 carbides along the grain boundaries. Furthermore, better than 38 pct increase was obtained for this material in comparison to the creep life obtained from the best performing conventionally cast material. More importantly, the fine-grain homogenized (FGH) casting route resulted in the Larson–Miller plot for this material that coincided with that of wrought alloy 740. At low creep stresses (with a test still in progress), the FGH casting is resulting in higher values of the LMP than the wrought alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Assessment of Outliers in Alloy Datasets Using Unsupervised Techniques

We report advancements in data analytics techniques have enabled complex, disparate datasets to be leveraged for alloy design. Identifying outliers in a dataset can reduce noise, identify erroneous and/or anomalous records, prevent overfitting, and improve model assessment and optimization. In this work, two alloy datasets (9-12% Cr ferritic martensitic steels, and austenitic stainless steels) have been assessed for outliers using unsupervised techniques and supplemented with domain knowledge. Principal component analysis and k-means clustering were applied to the data, and points were assessed as outliers based on their distance away from other points in the cluster and from other points in the dataset. The outlier characteristics were investigated to determine both cluster-specific and overall trends in the properties of the outlier points. The approach demonstrated here is extensible to other alloy datasets for outlier identification and evaluation to improve the reliability of machine learning and modeling predictions for advanced alloy design.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of heavy ion irradiation dose rate and temperature on α' precipitation in high purity Fe-18%Cr alloy

Cr-rich alpha prime precipitates (CrRP) induce hardening and embrittlement of FeCr alloys, but the kinetics of CrRP formation due to particle irradiation are not well understood. In this study, Fe18wt.%Cr alloy in solid solution state and pre-aged to produce relatively coarse CrRP was irradiated with 8 MeV Fe ions. The irradiation conditions involved two midrange doses of 0.37 and 3.7 displacements per atom (dpa), a wide range of dose rates (10 -5 –10 -3 dpa/s) and temperatures (300–450 °C). The distributions of CrRP after irradiation were studied with atom probe tomography (APT). The critical irradiation conditions to suppress CrRP formation were identified as 300 °C and 10-3 dpa/s; CrRP formation occurred readily at lower dose rates or higher temperatures. From 0.37 to 3.7 dpa, CrRP were observed to slightly grow at 350 °C and strongly coarsen at 450 °C. Specimens with pre-existing CrRP evolved into a similar precipitate distribution as detected after ion irradiation on solution annealed specimens at 300–350 °C to 0.37 dpa, indicating that the precipitate microstructure approaches a quasi-equilibrium for doses <1 dpa. Limited shrinking of pre-existing CrRP was observed after irradiation at 450 °C to 0.37 dpa, indicating a higher recovery rate at this temperature. Additionally, the evolution of CrRP is quantitatively explained by employing corrections to the historic Nelson-Hudson-Mazey precipitate stability model, and a radiation modified precipitation mechanism is proposed to account for the competition between radiation enhanced diffusion and ballistic dissolution which results in the modifications on both size and solute concentration of CrRP.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Direct utilization of gaseous fuels in metal supported solid oxide fuel cells

Direct utilization and internal reforming of gaseous fuels is investigated on symmetric-architecture metal supported solid oxide fuel cells (MS-SOFCs) with thin ceramic electrolyte and scaffold backbone layers, and low cost ferritic stainless steel supports on both sides. Infiltrated Pr-oxide and Ni/samarium-doped ceria catalysts are added to the cathode and anode electrodes, respectively. Initial performance and durability is evaluated for MS-SOFCs operating with natural gas, propane, ammonia, and dimethyl ether at 700 °C. Cells for natural gas and propane utilize a novel high entropy alloy (HEA) catalyst for internal reforming with performance and degradation rates similar to H 2 (0.5W cm -2 and ~12%/100 h). In conclusion, initial testing with sulfur shows reversible degradation for levels found in natural gas and irreversible degradation for higher levels found in commercial propane. Overall, MS-SOFCs show successful fuel flexibility.

08 HYDROGEN↗
DOE OSTI2020

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300°C 450°C

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy (SiC-C@NFA) composite for potential applications as a cladding and structural material for next generation nuclear reactors. The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 450 °C. Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory. The effects of ion irradiation on various phases such as α-ferrite matrix, (Fe,Cr) 7 C 3 , and (Ti,W)C precipitates were evaluated. The α-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops (or loop strings) at >5 dpa. The size of the dislocation loops at 450 °C was larger than that at 300 °C. The nucleation and growth of new (Ti,W)C precipitates in α-ferrite grains were enhanced with the ion dose at 450 °C. This study provides new insight into the irradiation resistance of the SiC-C@NFA system.

(FeCr)(7)C-3↗
DOE OSTI2021

Neutron irradiation induced defects and clustering in NF616 and T91

NF616 is a third-generation ferritic martensitic steel, developed to have better creep resistance than the prior generation T91. Unlike relatively numerous studies of T91, there is a lack of understanding of the irradiation effect on the microstructural evolution and mechanical response of NF616. This work evaluated the microstructures and radiation hardening of NF616 irradiated up to 8.2 displacement per atom (dpa) at 292 °C – 431 °C, compared with T91 from two heats. Dislocation loops were observed in all investigated samples. NF616 exhibited comparable loop size but slightly lower loop density than those in the general T91 heat at 430 °C. Cavities were only observed in NF616 at 431 °C but absent at lower irradiation temperatures (292 °C and 359 °C). Ni-rich clusters were also observed in NF616 at 431 °C, while only weak Ni-clustering were observed at lower irradiation temperatures. Compared to the general T91 heat, NF616 demonstrated better swelling resistance (e.g., one third of swelling in the general T91 heat at 430 °C), a slightly higher number density of Ni-rich clusters, and slightly lower radiation hardening. The low-carbon T91 showed the greatest hardening with the largest swelling and loop sizes, despite its lowest irradiation temperature and intermediate dose. The calculated hardening from loops, cavities and Ni-rich clusters using the classic dispersed barrier-hardening model had reasonable agreement with the experiment-derived results, with the primary hardening contribution attributed to dislocation loops.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Deformation localisation in ion-irradiated Fe and Fe10Cr

Determining the mechanisms for irradiation-induced ductility loss is crucial for the design of reactor structural components. Here, the deformation characteristics around nanoindents in Fe and Fe10Cr irradiated with Fe ions to ~1 displacement-per-atom at 313 K are non-destructively studied. Slip steps surrounding the nanoindents indicate that deformation is localised in the irradiated materials. Lattice rotation and strain fields near the indent site show over 87% confinement of plasticity in the irradiated material. Cr has little effect on the irradiation-induced changes in pile-up topography and deformation fields, suggesting it has limited impact on retaining strain hardening capacity and reducing irradiation-induced embrittlement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗
DOE OSTI2025

Bonding of vanadium- and Iron-based alloys as interlayers for plasma-facing and structural materials in fusion systems

Vanadium alloys and FeCrAl were investigated as interlayers between tungsten and reduced activation ferritic martensitic steel for fusion system components to avoid formation of intermetallic phase at operating temperatures between 550 and 1100 °C, while maintaining a body centered cubic phase throughout the interface. Physical and mechanical properties need to be graded between tungsten and steel, but recent results showed a significant hardness increase at the FeCrAl to vanadium alloy interface. Here, a sintered sample of these alloys was annealed for extended time, and the microstructure was investigated to provide a better understanding of the phenomena. A comparison with an additively manufactured interface of the same material is provided. An unexpected L2 1 intermetallic phase formation has been revealed using microscopy and synchrotron techniques and will inform future additive manufacturing approaches of the interface. A Cr layer interface as a preliminary solution was proposed between the Vanadium alloy and FeCrAl alloy interface.

Additive manufacturing↗
DOE OSTI2021

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Here, second phase intermetallic particles in an advanced accident-tolerant FeCrAl (Fe-13Cr-5Al-2Mo) alloy are formed in the α-Fe matrix during processing. These particles are prominently related to the added Y. Neutron irradiation to ~7 displacements per atom (dpa) with a dose rate of ~8.16 × 10 -7 dpa/s at 282 °C resulted in the amorphization of these precipitates which could degrade the mechanical properties of the FeCrAl alloys. Analytical electron microscopy and diffraction analysis combined with structural freedom analysis have been used to investigate the radiation resistance of the second phase particles. Radiation tolerance is closely linked to the particle Fe-Y content and can be tailored using the structure freedom value.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Irradiation-induced formation of G-phase precipitates and M 2 X carbides in self-ion irradiated HT-9

Ferritic-martensitic steels with high chromium content are a promising material group for advanced nuclear systems due to their high temperature strength and good irradiation tolerance. HT-9 is an optimized and often-studied alloy in this group, but additional studies are required on its radiation response under extreme conditions to be experienced in various types of nuclear reactors, especially with respect to phase stability under irradiation. Self-ion irradiation of HT-9 by 5 MeV Fe ions was used to simulate neutron-induced behavior reaching peak doses of 100 and 300 dpa at temperatures ranging from 450 to 550 °C. M 23 C 6 carbides that existed prior to irradiation were found to remain stable under all examined irradiation conditions. As irradiation progressed at 450 and 500 °C, however, formation of spherical-like G-phase precipitates and needle-like M 2 X carbides was observed. G-phase precipitates were found to be enriched in Ni, Si, and Mn, and show no interface segregation, whereas needle-like M 2 X carbides were rich in Cr and Mo and clearly displayed interface segregation of Ni and Si. M 2 X carbide formation is believed to be assisted by vacancies, while G-phase precipitation is thought to be assisted by interstitials. Finally, this difference in defect-mediated formation leads to a difference in distribution with depth. M 2 X carbides are distributed over shallower depths than that of G-phase precipitates, consistent with defect imbalance predictions that consider the influence of the injected interstitial effect.

atom probe tomography↗
DOE OSTI2025

Inversion of dislocation-impurity interactions in α -Fe under magnetic state changes

Here, in this work, we investigate the dislocation-impurity interaction energies and their profiles for various 3d elements —V, Cr, Mn, Cu, Ni, and Co —in and around 1/2 $\langle111\rangle$ screw dislocations in α -Fe using ab initio methods. We consider the ferromagnetic and paramagnetic states, with the latter being modeled through both the disordered local moment model and a spin-wave approach. Our findings reveal that (1) magnetic effects are large compared to size misfit effects of substitutional impurities, and (2) dislocation-impurity interactions are dependent on the magnetic state of the matrix and thermal lattice expansion. In particular, Cu changes from core-attractive in the ferromagnetic state to repulsive in the paramagnetic state.

36 MATERIALS SCIENCE↗