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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
NASA NTRS2024

Analyses of the MISSE 9-15 Polymers and Composites Experiment 1-4 (PCE 1-4) Contamination Samples

Spacecraft in low Earth orbit (LEO) are subjected to harsh environmental conditions, including radiation (cosmic rays, ultraviolet (UV), x-ray and charged particle radiation), micrometeoroids and orbital debris, temperature extremes, thermal cycling, and atomic oxygen (AO). These environmental exposures can result in erosion, embrittlement and optical property degradation of susceptible materials threatening spacecraft performance and durability. To increase our understanding of environmental effects such as AO erosion and radiation induced embrittlement of spacecraft materials, NASA Glenn Research Center developed a series of experiments that were flown as part of the Materials International Space Station Experiment (MISSE) missions on the exterior of the International Space Station (ISS). Recently, four Glenn experiments with 365 flight samples were flown on ISS’s MISSE-Flight Facility (MISSE-FF). These experiments are the Polymers and Composites Experiment-1 (PCE-1) flown as part of the MISSE-9 mission, the PCE-2 flown as part of the MISSE-10 mission, the PCE-3 flown as part of the MISSE-12 and MISSE-15 missions, and the PCE-4 flown as part of the MISSE-13 mission. Each of these experiments included passive contamination witness samples in each flight direction for post-flight molecular contamination analyses. A total of 13 contamination flight samples were flown. The post-flight analyses of the PCE 1-4 contamination samples include X-ray Photoelectron Spectroscopy (XPS) analyses (surface and ion sputter depth analyses) and optical properties (total reflectance, total transmittance and solar absorptance). This paper provides results of post-flight analyses of the PCE 1-4 contamination samples and their corresponding control samples. Knowledge of on-orbit contamination is important for the PCE 1-4 flight data interpretation.

Atomic Oxygen↗
DOE OSTI2025

Direct Imaging of Hydrogen‐Driven Dislocation and Strain Field Evolution in a Stainless Steel Grain

Hydrogen embrittlement (HE) poses a significant challenge to the durability of materials used in hydrogen production and utilization. Disentangling the competing nanoscale mechanisms driving HE often relies on simulations and electron-transparent sample techniques, limiting experimental insights into hydrogen-induced dislocation behavior in bulk materials. This study employs in situ Bragg coherent X-ray diffraction imaging to track three-dimensional (3D) dislocation and strain field evolution during hydrogen charging in a bulk grain of austenitic 316 stainless steel. Tracking a single dislocation reveals hydrogen-enhanced mobility and relaxation, consistent with dislocation dynamics simulations. Subsequent observations reveal dislocation unpinning and climb processes, likely driven by osmotic forces. Additionally, nanoscale strain analysis around the dislocation core directly measures hydrogen-induced elastic shielding. These findings experimentally validate theoretical predictions and offer mechanistic insights into hydrogen-driven dislocation behavior. The quantified nanoscale phenomena serve as critical inputs for multiscale modeling frameworks to predict bulk material responses and accelerate the development of HE-resistant alloys.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

Doping Effects on the Ductility of a Lightweight Refractory High-Entropy Alloy: Grain Boundary and Bulk Lattice Aspects

Doping elements in small amounts often segregate to grain boundaries (GBs) in alloys and can significantly impact mechanical properties and performance. Refractory high-entropy alloys (RHEAs) are known for their poor ductility, especially at low temperatures. Promoting GB cohesion through segregation can be an effective approach to mitigate embrittlement. Here, in this study, first-principles density functional theory (DFT) calculations were performed to examine the effects of important interstitial dopants (O, B, C, and N) and substitutional dopants (Cr, Y, La, and Ce) on the Σ5(310) [001] tilt GB of a lightweight RHEA Nb 32.5 Ti 27.5 Mo 22.5 Ta 12.5 Hf 2.5 Zr 2.5 . The DFT calculations reveal that certain dopants, such as B, C, Cr, N, and Y, exhibit favorable GB strengthening effects by improving bonding interactions with the bulk alloy. The impact of doping on the ductility parameter of the bulk lattice, defined as the ratio of surface energy to unstable stacking fault energy for the {110} <111> slip system, was also studied; and the results show that doping reduces the intrinsic ductility of the alloy, decreasing the D-parameter from 2.90 to 2.55, depending on the specific dopant. The present findings provide a foundational understanding at atomic level of the effect of representative dopants on mechanical properties of RHEAs and can be used to guide future alloy design for improved mechanical properties.

deformation charge density↗
DOE OSTI2023

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as leading candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy’s unique balance of inherently high fracture toughness and low sputter yields; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. Further, to alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni + and He + ion irradiations to mimic the expected displacement damage and gas production expected after five years of service as a plasma facing material component. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and an a-W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation assisted and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material’s overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Hydrogen-enhanced compatibility constraint for intergranular failure in FCC FeNiCoCrMn high-entropy alloy

The fundamental mechanism of hydrogen embrittlement was investigated in the high-entropy alloy FeNiCoCrMn using slow strain rate tensile tests with and without internal hydrogen. Hydrogen induced intergranular failure and reduced the average grain elongation parallel to the tensile axis, but also increased the local plasticity within grains. Here, the influence of hydrogen on plasticity establishes a compatibility constraint across grain boundaries, which results in failure along the hydrogen-weakened grain boundaries. This study is the first to directly confirm the presence of the hydrogen-enhanced compatibility constraint in a high-entropy alloy and highlights the importance of developing a physical understanding of grain-scale interactions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effects of hydrogen partial pressure on crack initiation and growth rate in vintage X52 steel

There is an increasing interest in ensuring compatibility of existing natural gas infrastructure with conveyance of hydrogen, especially for vintage (pre-1970s) pipeline steels. While hydrogen is known to affect the fatigue and fracture resistance of these materials, there is a lack of data on its effects on crack initiation at low hydrogen partial pressure. To address this gap, the current work presents a series of fatigue tests on vintage API Grade X52 steel in pure nitrogen, air and gaseous hydrogen at partial pressures of 1 and 207 bar. Circumferential notch tension (CNT) specimens with direct current potential difference (DCPD) measurements are used to get S–N curves for crack initiation and failure, and compact tension (CT) tests to assess crack growth rates. Results indicate that air accelerates crack initiation, low hydrogen partial pressure affects the crack growth rate, and higher pressures have the largest impact on both. Surprisingly, crack initiation is faster in air than at 1 bar hydrogen partial pressure. The CNT tests also reveal that the orientation of the banded microstructure has no effect on the crack initiation site, but it influences its growth behavior. Moreover, fractography shows that hydrogen also promotes instances of quasi-cleavage and intergranular fracture, linked to embrittlement of the material. Here, these results demonstrate that, while hydrogen may impact the entirety of the fatigue process, its effects on the early stages of damage accumulation become relevant only at larger pressures. More comprehensive studies of crack initiation in hydrogen atmospheres will be critical to future codes and standards for hydrogen infrastructure.

Circumferential notch tension↗
DOE OSTI2025

Influence of stacking fault energy and hydrogen on deformation mechanisms in high Mn austenitic steels during in-situ tensile testing

High Mn austenitic steels are considered an economical alloy system for hydrogen storage and transport applications. This study used stacking fault energy (SFE) as a design parameter to achieve hydrogen embrittlement (HE)-resistant high Mn austenitic alloys. The role of hydrogen on the deformation mechanisms of low (29 mJ/m 2 ) and high SFE (49 mJ/m 2 ) alloys was evaluated through in-situ neutron diffraction during tensile loading. Hydrogen-precharging increased yield strength, partly due to hydrogen-induced lattice distortion (i.e., solute strengthening). Hydrogen accelerated the increase in defect density, including dislocations and stacking faults. The formation of planar deformation structures (twins and stacking faults), relative to dislocations, plays a critical role in promoting hydrogen-assisted fracture. The stacking fault frequency parameter obtained from neutron diffraction quantifies planar deformation tendencies, correlated with HE sensitivity. The higher SFE alloy exhibited greater resistance to HE, associated with the reduced propensity to form stacking faults and twins upon deformation in the hydrogen-precharged condition.

High Mn austenitic steel↗
DOE OSTI2025

Interplay between hydrogen, temperature, and character angle on dissociated dislocation energies in Fe–Ni–Cr austenitic stainless steels

Dislocation energy has an important role in the mechanical performance of structural metals. While dislocation energies cannot be fully obtained from continuum theories due to the contribution of the dislocation core, they have been calculated via atomistic simulations in elemental metals. However, constraints on the local atomic environments have prevented the use of such approaches in systems that incorporate alloying or interstitial solutes. In this work, we develop robust molecular dynamics methods to resolve these issues through a geometric construction of dislocation dipoles and the calculation of time-averaged energies. Furthermore, we apply these methods to calculate dislocation energies (including core energies) in an Fe 70 Ni 11 Cr 19 austenitic steel at a variety of character angles, hydrogen concentrations, temperatures, and dipole spacings. The resulting highly converged energies show an excellent agreement with continuum expressions. Overall, hydrogen concentrations up to 1.0 % do not have a significant effect on the elastic parameters and dislocation energy. The methods and insights derived in this work have the potential to facilitate the calculation of dislocation energies in a wide range of systems, and to guide our understanding of hydrogen embrittlement.

Alloyed systems↗
DOE OSTI2020

Effects of HFIR neutron irradiation on fracture toughness properties of standard and Ni-doped F82H

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. Here, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58 Ni, and F82H doped with 1.4% 60 Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe 2 Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Deformation Strengthening of Uranium and Dilute Uranium Alloys

The strength and ductility of uranium and dilute uranium alloys can be substantially improved by deformation strengthening. The yield strength of unalloyed uranium can be nearly doubled by rolling in the vicinity of 250°C. The embrittling effect of hydrogen is also overcome, eliminating the need for vacuum outgassing to obtain good ductility. Significant differences in longitudinal versus transverse tensile properties result, in part, from texture introduced by unidirectional rolling. Warm rolling of U-2.3%Nb results in both traditional strain hardening and a significant Bauschinger effect, resulting in differences between tensile and compressive yield strength. A similar combination of strain hardening and Bauschinger effect also occurs in deformation strengthened U-0.75%Ti. Warm rolling of U-0.75%Ti results in significantly better combinations of tensile yield strength and ductility than can be obtained by conventional age hardening. Warm rolling prior to aging also reduces the loss of ductility which typically accompanies conventional age hardening. Here, an approach is developed in which the effects of prior deformation on tensile and compressive yield strengths are analyzed in terms of texture, long-range dislocation, and Bauschinger effects. The Bauschinger effect is shown to vary in sign with the type of deformation, thus providing opportunities to tailor tensile and compressive properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A potential emerging issue concerning repair welding of out-of-core PWR components involving tritium exposure and 3 He retention

This paper identifies a potential but previously unrecognized risk of helium-induced embrittlement and cracking during repair welding of out-of-core pressurized water reactor (PWR) components exposed to tritium-contaminated coolant. While previous weldability concerns centered on 4 He accumulation in neutron-irradiated alloys located within the in-core or near-core regions, new measurements show that 3 He generated by tritium decay can accumulate in out-of-core components. Because hydrogen isotopes readily diffuse to grain boundaries and become trapped there, significant 3 He generation at grain boundaries may lead to cracking during weld repairs. Initial data from far-below-core PWR flux thimble tubes confirm the presence of 3 He levels above known cracking thresholds for repair welds. In conclusion, these findings indicate that out-of-core regions should be considered when defining safe weld repair windows in reactors operating for 60–100 years.

Extended PWR lifetimes↗
DOE OSTI2022

High energy X-ray diffraction and small-angle scattering measurements of hydrogen fatigue damage in AISI 4130 steel

Accurate lifetime predictions are critical for repurposing existing pipelines for hydrogen transmission as well as for developing novel steels which are minimally susceptible to lifetime degradation by hydrogen. Ultimately, lifetime prediction models assess the amount of damage a material undergoes during a typical service cycle and the cumulative damage a material can withstand prior to failure. However, not all damage processes are equal, and neither is the manner in which mechanical loading translates to damage the same when materials are in inert environments compared to in hydrogen environments. For example, in the three leading proposed mechanisms of hydrogen embrittlement (Hydrogen-Enhanced Decohesion (HEDE), the Hydrogen-Enhanced Localized Plasticity (HELP), and the Nano-Void Coalescence (NVC)), hydrogen is proposed to enhance the manifestation of grain separation, dislocation generation/movement, and void coalescence, respectively. A full understanding of the damage modes requires a measurement capable of probing all three mechanisms at once. Here we present simultaneous High Energy X-ray Diffraction (HEXRD) and Small-Angle X-ray Scattering (SAXS) during fatiguing of steel in hydrogen. HEXRD measurements probe strain and dislocation density; SAXS measurements probe nano-pore generation and coalescence. We will discuss the differences in damage modes between steels fatigued in air and in hydrogen and the role these difference play in lifetime predictions.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation effects in tungsten—From surface effects to bulk mechanical properties

Advanced materials such as tungsten fibre-reinforced composites allow to overcome severe weaknesses of the baseline materials for plasma-facing components — copper and tungsten. The effect of the fusion environment on the mechanical properties of these materials, e.g. the embrittlement by neutron irradiation, plays a key role for the development of future fusion reactors. To simulate this effect, high-energy ions are used as a substitute for the displacement damage by neutrons. We propose the use of very fine tungsten wire as a possibility of studying the influence of irradiation damage on the mechanical properties. This is possible as they allow full-depth irradiation of almost the entire volume despite the limited penetration depth of ions. Geometrical size effects are mitigated due to the nanoscale microstructure of the wire. In addition, similar wire is used in tungsten fibre-reinforced composites. Thus, the investigation of irradiated wire can directly be used for the prediction of the bulk composite properties. For the proof of this concept tungsten wire with a diameter of 16 μm was electrochemically thinned to 5 μm and irradiated with 20.5 MeV W 6+ ions. The mechanical properties were subsequently determined by macroscopic tensile testing. Irradiation to 0.3, 1 and 9 dpa did not lead to a change of the mechanical behaviour. Both strength and ductility, the latter indicated by the reduction of area, were similar to the as-fabricated state.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A critical review of irradiation-induced changes in reactor pressure vessel steels

A large body of work has been conducted to investigate the embrittlement and degradation of reactor pressure vessel (RPV) steels. This includes experiments on alloys with different compositions, performed in research and test reactors and ion accelerators that span various temperatures, fluxes, and fluences. In this paper, we perform a critical review of the published experimental data and compile experimentally reported values for dislocation loop size/density, precipitate size/density and yield stress into an easily downloadable format that can be used by both experimentalists and modelers. This thorough experimental review is complemented by a brief review of simulation efforts at atomistic and mesoscopic length scales. Finally, this paper highlights key aspects of the behavior of RPV steels under irradiation, identifies gaps or discrepancies in current understanding, and identifies priority future research directions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Comments on “Failure analysis of corroded hydrogen-blended natural gas pipelines based on finite element analysis and genetic algorithm-back propagation neural network” [262 (2025) 111174]

This is a brief commentary paper to highlight and discuss the determination of hydrogen concentration in pipeline steel, effect of hydrogen embrittlement (HE) on the mechanical properties of the material, burst strength of corroded pipelines using finite element analysis (FEA) simulations, and curve-fit models for assessing remaining strength of X80 corroded pipelines for transporting hydrogen blended natural gas. Recently, Xie et al. [1] proposed a methodology to quantify the impact of HE on material properties and numerically determined burst pressure of X80 corroded pipelines. However, their HE quantification overestimated the degradation of tensile strength for hydrogen blending ratios beyond the original data range, and their FEA results of burst pressure are nonconservative. This work thus recharacterized the hydrogen concentration in the steel pipeline and the effect of HE on tensile strength, and then redetermined burst pressures for a set of typical corrosion defect cases considered by Xie et al. [1] based on an experimentally validated FEA modelling method. With the new FEA results, two empirical corrosion models were proposed for X80 corroded pipelines for hydrogen service. At zero hydrogen blending ratio, the novel empirical models predict burst pressures to be consistent with the industry-accepted corrosion models. Furthermore, both the numerical simulation method and the novel corrosion models are significant contributions to the pipeline industry and the hydrogen community. Application of these results will enhance the safety, reliability, and integrity of natural gas pipelines when used to transport hydrogen.

Burst pressure prediction↗
DOE OSTI2025

Rapid tempering to enhance dynamic performance of high and ultra-high strength steels

This study employed rapid, short-time duration (1 s) tempering to improve the dynamic performance of high-strength steel (HSS) and ultra-high-strength steel (UHSS), compared to quenched and conventionally (1800 s) tempered microstructures. Rapid tempering significantly improved the Charpy toughness of both steels compared to the conventionally tempered condition at an equivalent tempering parameter (TP) or hardness level. For conventionally tempered conditions, tempered martensite embrittlement was observed within the tempering temperature range of 200–400 °C, whereas rapid tempering exhibited increased Charpy toughness with increasing tempering temperature across all tempering conditions. Dynamic compression experiments indicated that rapid tempering led to improved ductility compared to the conventionally tempered condition at an equivalent TP. This work shows that the same rapid tempering strategy used for HSS is directly transferable to a higher-carbon UHSS and provides the first systematic evidence of improved strength-ductility combinations and cracking resistance under dynamic compression, supported by Kolsky bar experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A reactive molecular dynamics model for uranium/hydrogen containing systems

Uranium-based materials are valuable assets in the energy, medical, and military industries. However, understanding their sensitivity to hydrogen embrittlement is particularly challenging due to the toxicity of uranium and the computationally expensive nature of quantum-based methods generally required to study such processes. In this regard, we have developed a Chebyshev Interaction Model for Efficient Simulation (ChIMES) that can be employed to compute energies and forces of U and UH3 bulk structures with vacancies and hydrogen interstitials with accuracy similar to that of Density Functional Theory (DFT) while yielding linear scaling and orders of magnitude improvement in computational efficiency. Here, we show that the bulk structural parameters, uranium and hydrogen vacancy formation energies, and diffusion barriers predicted by the ChIMES potential are in strong agreement with the reference DFT data. We then use ChIMES to conduct molecular dynamics simulations of the temperature-dependent diffusion of a hydrogen interstitial and determine the corresponding diffusion activation energy. Our model has particular significance in studies of actinides and other high-Z materials, where there is a strong need for computationally efficient methods to bridge length and time scales between experiments and quantum theory.

36 MATERIALS SCIENCE↗