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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

An Integrated Computational Materials Engineering (ICME) Approach to Design Nonlinear Transition Zones Between Dissimilar Metals

Current approaches to designing graded transition joints (GTJs) between dissimilar metals often rely on linear changes in both composition profiles and thickness of each sublayer. This increases fabrication cost and may not be optimal with respect to residual stress or the formation of undesirable phases. Here, in this study, GTJs between P91 ferritic/martensitic steel and 347H austenitic stainless steel were designed using Integrated Computational Materials Engineering (ICME) principles with nonlinear composition and length profiles. Guided by inputs from classical mechanics and CALPHAD predictions of carbon chemical potential, a novel transition zone consisting of five discrete compositions was proposed, with the thickness of each sublayer varying according to a brachistochrone-inspired distribution. In addition to carbon potential gradients, CALPHAD was used to predict coefficients of thermal expansion, which were incorporated into finite element models to evaluate stress evolution. The proposed nonlinear design resulted in a smoother carbon potential gradient, lower carbon depletion at the P91 interface, and a comparable residual stress under long-term thermal exposure, compared to a conventional linear design using ten sublayers with equal thickness. This work introduces a brachistochrone-inspired distribution for GTJ design, offering a general framework for optimizing graded interfaces between dissimilar metals.

Directed Energy Deposition↗
DOE OSTI2023

Preliminary HEA Solid Phase Processing Development Report

The development of high temperature fuel cladding materials to withstand a variety of extreme environments have received much attention. Several potential materials systems that have been identified for the fuel systems and core structural materials application in advanced reactor systems are ferritic/martensitic steel (e.g., HT9), austenitic stainless steels (e.g., 316 LN), oxide-dispersion strengthened steels (e.g., 12 YWT), Ni-based alloys and ceramic-based composites depending on the type of the reactors. Though these material systems have promising properties conducive for radiation-resistant performance, they suffer beyond the design-limit from one or more damage processes such as void swelling, radiation embrittlement, phase instability, corrosion, and limited creep life.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural development in DED stainless steels: applying welding models to elucidate the impact of processing and alloy composition

Austenitic stainless steel microstructures produced by directed energy deposition (DED)are analogous to those developed during welding, particularly high energy density welding. To better understand microstructural development during DED, theories of microstructural evolution,which have been established to contextualize weld microstructures, are applied in this study to microstructural development in DED austenitic stainless steels. Phenomenological welding models that describe the development of oxide inclusions, compositional microsegregation, ferrite,matrix austenite grains, and dislocation substructures are utilized to clarify microstructural evolution during deposition of austenitic stainless steels. Two different alloys, 304L and 316L, arecompared to demonstrate the broad applicability of this framework for understanding microstmctural development during the DED process. Despite differences in grain morphology and solidification mode for these two alloys (which can be attributed to compositional differences),similar tensile properties are achieved. It is the fine-scale compositional segregation and dislocation structures that ultimately determine the strength of these materials. The evolution of microsegregation and dislocation structures is shown to be dependent on the rapid solidification and thermomechanical history of the DED processing method and not the composition of the starting material.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Phase Transformation of Thermally Aged and Neutron Irradiated Duplex Stainless Steel Used in LWRs. Final report

The lifetime of reactor components made of duplex stainless steels can be limited by the embrittlement from thermal aging, neutron irradiation or a synergistic effect between thermal aging and neutron irradiation. There is still a large scientific knowledge gap in understanding the phase evolution in duplex stainless steels upon thermal aging with or without neutron irradiation, the synergistic effect between thermal aging and neutron irradiation, and the impacts of structural evolution on material mechanical responses. In this project, totally, five major tasks were completed through this project: a) the in-situ WAXS tensile tests of thermally aged cast duplex stainless steels, b) the EXAFS investigation of thermally aged cast austenitic stainless steel and welds of austenite sainless steel, c) the characterization of irradiation effect on neutron irradiated CASS using APT, d) characterize the ferrite hardening of neutron irradiated CASS using nanoindentation, and e) the FEM modelling to validate the in-situ WAXS tensile tests and to simulate the stress/strain distribution during deformation for thermally aged duplex stainless steel. In summary, the thermal aging without neutron irradiation can significantly increase the strength of ferrite phase through the spinodal decomposition and G-phase precipitations. The in-situ X-ray experiments show that the hardening effect in ferrite phase can significantly alter the load partition between the ferrite and austenite phases. Atom probe tomography characterizations on samples irradiated to 0.08, 3, 5, 10, 20, and 40 dpa shows that neutron irradiation has dominated the microstructural evolution after 3 dpa, and the spinodal decomposition has saturated yet even at 40 dpa based on the quantified wavelength and amplitude.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Irradiation-assisted cracking of SA508-304L weldments with 308L groove filler and 309L butter in hot water immersion constant rate extension tests

Low alloy SA508 bainitic steel is often welded to 300 series austenitic stainless steel in nuclear power plants. These dissimilar metal weldments use 309L butter applied to the low alloy steel and 308L groove filler material to join the 309L to austenite. High alloy steels (such as the 308L and 309L material used here) have been identified as having greater susceptibility to irradiation-assisted stress corrosion cracking (IASCC). Evidence is presented here of cracking under tensile loads in 309L weldment butter in LWR hydrothermal conditions that correlates irradiation induced changes at δ ferrite–γ austenite phase boundaries and γ–γ grain boundaries; specifically intergranular Cr 23 C 6 precipitates. Cracking is not observed in irradiated 308L groove filler under the same conditions. Both the 309L butter and 308L filler material have a duplex or mixed δ–γ phase microstructure due to solidification during the welding procedure. We have performed constant extension rate autoclave immersion tests in BWR normal water chemistry hot water (288 °C, 10 MPa, 2000 wt. ppb dissolved oxygen, <100 nS/cm conductivity, neutral pH) of proton irradiated (2 MeV, approximately 3 displacements per atom at a depth of 10 μm) tensile specimens. Proton irradiation to the dpa levels used here induced Cr 23 C 6 carbide precipitation at γ–γ grain boundaries and δ–γ phase boundaries in the 309L butter but not the 308L filler. Cracking was observed in the proton-irradiated volume of 309L at these interfaces in the vicinity of the Cr 23 C 6 precipitates. We hypothesis the SA508 is the source of carbon in the 309L butter. Cr depletion at γ–γ grain boundaries was also observed and may also be a root cause of cracking. Furthermore, the 309L butter is under significant tensile residual stress in the weldment that adds to applied tensile loads and this too may be a contributing factor.

309L weld butter↗
DOE OSTI2020

Limiting Retained Austenite Decomposition in Quenched and Tempered Steels: Influences of Rapid Tempering and Silicon

Tempering reactions are critical to microstructure and property control in martensitic steels. Here, retained austenite decomposition and cementite precipitation are monitored using Mössbauer spectroscopy in 4340 and 300-M steel under conventional and rapid tempering conditions. Tempering times are compared at a constant tempered hardness by increasing tempering temperatures associated with short time conditions to achieve equivalent matrix softening to that of longer tempering times. Time-temperature combinations that provide equivalent tempered hardness generated microstructures with similar dislocation densities and cementite precipitation fractions; these mechanisms are controlled by self-diffusion. However, systematic differences in retained austenite content were observed at a given degree of softening, where shorter tempering times exhibited higher levels of retained austenite compared to more conventional conditions. At low temperatures, the differences in retained austenite preservation between explored time-temperature conditions are attributed to corresponding differences in carbon diffusion distance (in austenite), the controlling diffusional process of retained austenite decomposition. At higher temperatures, retained austenite decomposition exhibits C-curve kinetic behavior in 4340. Thus, reduced thermodynamic driving force for cementite and ferrite formation at higher temperature is believed to play a role in restricting retained austenite decomposition within some short-time, high temperature tempering regimes. The addition of silicon pushes cementite precipitation and retained austenite decomposition to higher temperatures, although retained austenite decomposition is suppressed to a greater extent than cementite precipitation. Potential is illustrated for coupling rapid tempering with silicon alloying to produce appreciably tempered martensite (~490 HV) with relatively less retained austenite decomposition compared to conventional tempering conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development of Predictive Model for Accurate Rupture Time from Multi-Axial Creep in Alloy 709 with Physics-Based Simulations

A physics-based model is developed to predict multiaxial creep behavior in Alloy 709 (A709), an advanced austenitic stainless steel intended for high-temperature applications such as Sodium Fast Reactors (SFRs). Compared to conventional stainless steels like 316H, A709 offers superior high-temperature performance; however, comprehensive data on its multiaxial creep response remain limited. To address this gap, a crystal plasticity finite element (CPFE) framework is used to simulate the deformation and failure mechanisms of A709 under multiaxial loading conditions. The model incorporates an extended Hu-Cocks dislocation creep formulation that accounts for precipitation effects, along with the Sham–Needleman model to capture grain boundary cavitation-driven failure. These advanced constitutive models enable a detailed understanding of the interplay between microstructural evolution and macroscopic creep response. Furthermore, the study evaluates the predictive accuracy of various effective stress measures in estimating creep rupture life, leveraging simulated multiaxial creep data. The findings provide critical insights into the applicability of different stress measures for engineering design and life prediction of A709 components operating under complex loading conditions. This work contributes to improving the reliability of high-temperature structural components by advancing predictive modeling capabilities for advanced austenitic steels.

Alloy 709↗
DOE OSTI2021

Invited Review Paper in Commemoration of Over 50 Years of Oxidation of Metals: Addressing the Role of Water Vapor on Long-Term Stainless Steel Oxidation Behavior

In the past 20 years, it has become clear that the long-term exposure of chromia-forming austenitic stainless steels to air containing ~ 10 vol% water vapor at 650–800 °C significantly increases the Cr consumption rate compared to laboratory air. However, a similar acceleration does not occur in 100% steam because O 2 is not available to form volatile CrO 2 (OH) 2 . As the research has shown, this accelerated Cr loss is particularly important for thin-walled components like gas turbine recuperators. Typically, conventional austenitic stainless steels exhibit accelerated attack (i.e., rapid formation of Fe-rich oxide nodules), while higher alloyed steels resist this type of degradation until the alloy surface becomes Cr depleted. The development of strategies and solutions for this issue are reviewed. Model Fe–Cr–Ni alloys have been used to study composition effects on this behavior. For example, increasing the Mn content up to ~ 4 wt% was not beneficial, but increasing the Ni content improved oxidation resistance in this environment. However, at 650–700 °C even highly alloyed steels showed surface Cr depletion at alloy grain boundaries resulting in Fe-rich oxide nodule formation. The path forward for this issue will require more refined mechanistic understanding and increased used of modeling to develop better application-specific lifetime models to identify the most cost-effective alloy solutions.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Microstructure, Hardness, and Residual Stress of the Dissimilar Metal Weldments of SA508-309L/308L-304L

A dissimilar metal weldment consisting of SA508-309L-308L-304L is widely used in light-water nuclear reactors. These weldments demonstrate dissimilar susceptibility to stress corrosion cracking that are related to the microstructure, properties, and residual stress. In this work, microstructures, hardness, and the residual stress distribution of the dissimilar metal weldments were investigated, with the correlation of increased hardness in the heat-affected zone (HAZ) to the microstructure. 304L HAZ demonstrated similar grain morphology as the base material, and the increase in hardness was primarily attributed to the increased dislocation density. SA508 HAZ demonstrated a change of grain morphology resulting from the different peak temperatures and cooling rates. The increased hardness in the SA508 HAZ was attributed to the refined grain morphology, higher dislocation density, and higher number density of precipitates. A ~ 20–30-μm-wide martensitic zone formed at the fusion boundary of SA508-309L, where Cr-rich carbide precipitates were observed, with the average size and the number density of 44.1 ± 16.9 nm and 1.5 × 10 21 m ₋3 , respectively. In conclusion, residual stress results demonstrated the largest tensile stress at 309L butter, indicating its high cracking susceptibility.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Stacking Fault Energy Based Alloy Screening for Hydrogen Compatibility

The selection of austenitic stainless steels for hydrogen service is challenging since there are few intrinsic metrics that relate alloy composition to hydrogen degradation. One such metric, explored here, is intrinsic stacking fault energy. Stacking fault energy has an influence on the character and structure of dislocations and on the formation of secondary crystalline phases created during mechanical deformation in austenitic alloys. In this work, a data-driven model for the intrinsic stacking fault energy of common austenitic stainless steel alloys is applied to compare the relative degradation of tensile performance in the presence of hydrogen. A transition in the tensile reduction of area of both 300-series and manganese stabilized stainless steels is observed at a calculated stacking fault energy of approximately 43 mJ m -2 , below which pronounced hydrogen degradation on tensile ductility is observed. The model is also applied to suggest alloying strategies for low nickel austenitic stainless steels for hydrogen service. Lastly, through this investigation, we find that calculated intrinsic stacking fault energy is a high-throughput screening metric that enables the ranking of the performance of a diverse range of austenitic stainless steel compositions, as well as the identification of new alloys, with regard to hydrogen compatibility.

08 HYDROGEN↗
DOE OSTI2021

Combined modeling and experiments to predict corrosion and embrittlement in dual-phase stainless steels within the MARMOT framework

Extending service of Light Water Reactors (LWRs) to beyond 60 years will demand high integrity of materials and components in the reactors. The accurate evaluation and prediction of materials performance under anticipated operating conditions are of particular importance for ensuring the safety of nuclear power plants over their extended lifetime. Stainless steels are extensively used as structural alloys in light water reactor (LWR) systems because of their excellent combination of mechanical properties and corrosion resistance. Austenitic stainless steel welds and cast austenitic stainless steels (CASS) contain significant amounts of ferrite ~5-30% and are ubiquitous in LWR piping (elbows, pump casings, valves) and internal structural components. During their lifetime, these components are subjected to thermal aging at the temperature range of 288-327°C for pressurized water reactors (PWR). A phase transformation within the ferrite phase of these materials leads to embrittlement and degradation in corrosion resistance, which undermines the materials selection criteria.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A machine learning aided interpretable model for rupture strength prediction in Fe-based martensitic and austenitic alloys

Abstract The class of 9–12% Cr ferritic-martensitic alloys (FMA) and austenitic stainless steels have received considerable attention due to their numerous applications in high temperature power generation industries. To design high strength steels with prolonged service life requires a thorough understanding of the long-term properties, e.g., creep rupture strength, rupture life, etc., as a function of the chemical composition and processing parameters that govern the microstructural characteristics. In this article, the creep rupture strength of both 9–12% Cr FMA and austenitic stainless steel has been parameterized using curated experimental datasets with a gradient boosting machine. The trained model has been cross validated against unseen test data and achieved high predictive performance in terms of correlation coefficient ( $$R^{2} > 0.98 $$ R 2 > 0.98 for 9–12% Cr FMA and $$R^{2} > 0.95 $$ R 2 > 0.95 for austenitic stainless steel) thus bypassing the need for additional comprehensive tensile test campaigns or physical theoretical calculations. Furthermore, the feature importance has been computed using the Shapley value analysis to understand the complex interplay of different features.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Crack Growth Rate and Fracture Toughness of Stainless Steel Welds in Light Water Reactor Environment

Austenitic stainless steel welds play a crucial role in maintaining the structural integrity of light water reactors. Within the reactor core, stainless steel welds endure a harsh service environment including elevated temperature, corrosive coolant, and intense neutron irradiation. Long-term exposure to such a severe service environment could lead to performance degradation. To ensure the safe and economic operation of light water reactors, this study performed crack growth rate and fracture toughness tests on irradiated and thermally aged weld samples harvested from decommissioned reactor components. In a low-corrosion-potential environment, stainless steel welds showed resistance to stress corrosion cracking, and the measured crack propagation rates were comparable to those of wrought stainless steels under similar test conditions. However, a marked decline in fracture toughness was observed in weld samples, particularly in irradiated samples. Fracture instability was observed at ~0.8 dpa, significantly lower than those observed in wrought stainless steels. Further research at higher doses is needed to evaluate the dose dependence of irradiation-induced embrittlement, and to understand the interplay between thermal aging and neutron irradiation in stainless steel welds.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1991

Cryogenic piping material selection for the Component Test Facility (CTF)

The anticipated high cost of the 8500 psi cryogenic and 15,000 psi gas piping systems used in the CTF at NASA's John C. Stennis Space Center led to the consideration of high-strength materials for these piping systems. Based on years of satisfactory service using austenitic stainless steels in cryogenic applications, particularly for hydrogen service, consideration was limited to the austenitic stainless steels. Attention was focused on alternatives to the 304/304L grades of stainless steel traditionally used in these applications. This paper discusses the various considerations that resulted in the decision to continue using 304/304L for the cryogenic piping and the selection of the nitrogen-strengthened 21Cr-6Ni-9Mn alloy (UNS S21903) for the high-pressure gas systems at the CTF.

St. Cyr, William W.↗