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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
NASA NTRS1959

Halogen-Containing Gases as Boundary Lubricants for Corrosion-Resistant Alloys at 1200 F

The extreme temperatures anticipated for lubricated parts in advanced flight powerplants dictate the consideration of unconventional methods of lubrication such as solid lubricants and the reactive gases described in the present research. These halogen-containing "reactive" gases such as dichlorodifluoromethane, CF2Cl2, are among the most stable of organic molecules. The high "flash" temperatures generated at the contacting asperities as a result of frictional heat are sufficient to cause local decomposition of the halogen-containing gases. The active atoms thus released (e.g., chlorine) then react with the metal to be lubricated to form halides capable of effective lubrication. The presence of small amounts of a sulfur-containing gas (e.g., 1 percent sulfur hexafluoride, SF6) was found to catalyze the formation of metal halides. Friction and wear studies were made with a hemisphere (3/16-in. rad.) rider sliding in a circumferential path on the flat surface of a rotating disk (2 1/2-in. diam.). The specimens of corrosion-resistant 2 alloys were run in an atmosphere of the various gases with a load of 1200 grams, a sliding velocity of 120 feet per minute, and temperature from 75 to 1200 F. An effective lubricant for ferritic materials (M-1 tool steel) was CF2Cl2, but significant corrosion occurred above 600 F. Corrosion evaluation in CF2Cl2 suggested a number of nickel- and cobalt-base alloys for additional lubrication study. Several combinations of gases and these metals were found to lubricate to 1200 F without excessive corrosion. The gases were CF2Cl2 Plus 1 percent SF6, monobromotrifluoromethane CF3Br plus 1 percent SF6, dibromodifluoromethane CF2Br2, iodotrifluoromethane, CF3I, and I2. Careful selection of metals and gas are necessary for successful lubrication over specific temperature ranges. Optimum combinations give friction coefficients as low as 0.05 without

Buckley, Donald H.↗
DOE OSTI2022

Investigation of Nano-Scale Segregation in Nanostructured Ferritic Alloy 14YWT after Heavy Ion Irradiation

Oxide-dispersion-strengthened (ODS) steels, which contain nano-scale Y-Ti-O particles, are being considered for high-temperature radiation environments of nuclear reactors. It is important to accurately characterize the structure of grain boundaries and understand the behavior of segregation at grain boundaries in ODS steels during irradiation. The effect of heavy ion irradiation at 700 °C on Nanostructured Ferritic Alloy 14YWT was investigated using Atom Probe Tomography. Enrichment of Cr occurs at the grain boundaries as well as at nano oxide particle surfaces in the unirradiated sample. The enrichment of Ti and Y at a grain boundary corresponds with Y-Ti-O nano oxide particles with larger size compared to those in the grain, and the Cr enrichment is particularly accentuated at these larger nano oxide particles. The segregation of W occurs at the grain boundaries that are without nano oxide particles. O is segregated at grain boundaries without oxide particles after irradiation. The segregation behavior of Cr, W, Ti, and Y at the grain boundary in the irradiated samples is similar to that in the unirradiated sample. The nano oxide particles embedded in the grain boundary are a primary reason for the increase in Cr segregation at the grain boundary.

36 MATERIALS SCIENCE↗
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 OSTI2024

Microstructure Optimization and Novel Processing Development of ODS Steels for Fusion Environments (Final ARPA-E Report)

This project aimed to develop scalable, cost-effective fabrication of high-performance, oxide- dispersion-strengthened (ODS) steel using advanced manufacturing methods (AMMs) for fusion blanket-breeding applications. Gas atomization reaction synthesis (GARS) enables the synthesis of precursor ODS steel powders without prolonged mechanical alloying. This process creates a chromium (Cr)-enriched surface oxide with yttrium/titanium (Y/Ti)-enriched intermetallics in powder interiors. GARS powders were consolidated to >99% of the theoretical density using a first-of-a-kind shear assisted processing and extrusion (ShAPE) and laser-based powder bed fusion (L-PBF) AM processes. These processes led to ODS steels containing a high-density of nano-oxide dispersoids that enhance high-temperature mechanical properties. Such scalable, cost-effective fabrication of ODS steels can enable efficient power conversion cycles (=40%) at operating temperatures beyond 900 K in future fusion power plants.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Progress on Producing an ODS Ferritic Alloy by High Deformation Processing of Reactive Powders

Oxide dispersion-strengthened (ODS) ferritic alloys, such as 14YWT, are considered as benchmarks for fuel cladding and core structural materials in advanced nuclear energy reactor systems that require high temperature strength and creep properties as well as resistance to radiation damage. Unfortunately, the popular processing method of mechanical alloying (MA) is time-consuming, incurs high manufacturing costs, and typically produces anisotropy in the microstructure and mechanical properties of ODS ferritic alloys. These considerations hinder their acceptance in many high-performance, high-temperature materials for advanced nuclear energy reactors. This work package is pursuing an alternative solid-state processing to MA involving high deformation at high temperatures for producing an ODS ferritic alloy for fuel cladding. This report summarizes the concept for this processing method and the progress toward selecting the reactive and ferritic alloy powders, preparing 316L stainless steel cans for filling with blended powder followed by degassing, and choosing the heat treatments and forging parameters for performing the first two forging experiments. The preliminary results of optical microscopy performed on polished samples that were cut in three orthogonal directions revealed remnant porosity in both samples. The analysis of the porosity using ImageJ revealed a lesser amount of ~2.49% in the HiLoMe-1 sample that was processed with the three heat treatments and forging parameters compared to ~5.70% for the Hi-1 sample that was processed with only the first sintering heat treatment and forging parameter. These initial results will provide insights that can be leveraged to improve the densification of the powders in future forging experiments that will be continued into FY24 for determining the optimum heat treatments and forging parameters coupled with detailed microstructure characterization studies for producing the ODS ferritic alloy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A Review of Medium-Mn, Low-Density Steels for Transportation Applications

Low-density steels constitute a broad and complex alloy space (Fe–Mn–Al–C) suitable for a variety of applications. In particular, there has been growing interest in duplex (ferrite + austenite) or multiphase (+ martensite, carbides) low-density steels as a lightweight, advanced high-strength steel (AHSS) for vehicle applications, spurred by extensive decarbonization efforts. Medium-Mn (med-Mn) (3 to 12 wt pct) steels with 3.5 to 10 wt pct Al additions have decreased densities, presenting an interesting opportunity for high-specific strength, intrusion-resistant, and energy-absorbing sheet components with reduced alloying contents compared to high-Mn grades like austenitic Fe–Mn–Al–C or twinning-induced plasticity steels. Compared to leaner med-Mn steels, the physical metallurgy of med-Mn, low-density steels (MMLS) is complex and distinguished by increased δ-ferrite fractions and austenite stacking fault energies, decreased martensite start temperatures, and modified phase transformation windows. Mechanical properties of MMLS are comparable to 3rd generation AHSS, attributable to the unique, multiphase microstructures, and the array of strengthening mechanisms that can be accessed. Despite this, challenges and unknowns remain with respect to their industrial implementation, and new processing routes may need to be developed. Here, this review aims to highlight the composition effects, processing methods, microstructural evolution, deformation behavior, and application properties geared toward manufacturing and performance, altogether assessing the potential of MMLS for transportation applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Microchemical evolution of irradiated additive-manufactured HT9

The microstructural responses under 5 MeV Fe2 + single-ion-beam irradiation of three conditions of additive-manufactured (AM) HT9 steel using a powder-based directed energy deposition (DED) technique with and without postbuild heat treatments were investigated. Besides the observed dislocation loop formation and the absence of cavities at the irradiation condition of 50 dpa at 460 °C, Ni/Si/Mn-rich precipitates are found to form in all three conditions of AM-HT9, whereas Cu-rich clusters that arise from Cu uptake from the DED process are only observed in the heat-treated conditions, and not in the as-built (ASB) condition. Coprecipitation of the Cu- and Ni/Si/Mn-rich clusters occur near defect sinks such as line dislocations and grain boundaries in the heat-treated AM-HT9. The variation in microchemical evolution can be directly linked to the starting sink strength of the three AM-HT9 conditions, and the ASB condition with higher sink strength suppressed the responses observed in the postbuild heat-treated specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Small Scale Creep Testing of 14YWT via In-situ Transmission Electron Microscopy Irradiation and Nanoindentation

The next generation of nuclear materials must withstand harsh operating conditions such as high temperatures and irradiation doses. Nanostructured ferritic alloys like oxide dispersion strengthened steels, especially 14YWT, have shown promise as a structural material to withstand these harsh operating conditions. For application, understanding the irradiation enhanced creep of the structural components is fundamental to evaluating the service life in a reactor. Challenges with evaluating irradiation enhanced creep are related to the limited space in nuclear reactors and the expense of conducting post irradiation experiments on irradiated cladding. Ion irradiations are considered to expedite irradiation testing, but such experiments have restricted depth of penetration into the material, limiting the ability to characterize changes to material properties after irradiation. Small scale mechanical testing can be used with ion beam irradiations as a potential avenue to evaluate the irradiation enhanced creep of materials. In this study, in-situ transmission electron microscopy nanopillar creep studies on 14YWT were performed and simultaneously ion beam irradiated with 2.8 MeV Au4+ ions. It was observed that the ion beam irradiation did increase the measured strain rate of the materials. In addition, ex-situ nanoindentation creep studies were performed over a range of temperatures on control 14YWT, and it was observed that there was a change in the deformation mechanism between 873 K and 1073 K that agrees well with macro-scale mechanical testing. Furthermore, these results show promise for applying these techniques to nuclear materials in the future.

14 YWT↗
DOE OSTI2024

Distribution of Tritium in the Near Surface of Candidate Structural Materials for Fusion Reactors. Type 304L Stainless Steel, Inconel, Hastelloy, Eurofer-97, Oxide Dispersion-Strengthened Alloy 14YWT

Five different candidate structural materials for fusion have undergone pure tritium gas soaking at room temperature and at 310-mbar(a) pressure. The tritium uptake on the surface and in the bulk of the alloys has been analyzed using surface leaching, chemical etching, and thermal desorption. The nickel-based alloys: Inconel-X-750 and Hastelloy-X, absorbed the least amount of total tritium compared with austenitic stainless steel AISI 304L, reduced activation ferritic-martensitic (RAFM) Eurofer-97, and advanced nanoferritic alloy 14YWT. Microstructural analyses using electron backscatter diffraction (EBSD) indicate that the number of grains and mean grain size is not a dominant factor in near surface tritium uptake. The quantity of iron dissolved in the surface oxide appears to be the major factor in encouraging tritium absorption.

O'Callaghan, James↗
DOE OSTI2020

A Low-Cost Iron-Based Current Collector for Alkaline Battery Electrodes

The use of three-dimensional porous nickel foam as the current collector of the nickel hydroxide electrode adds significantly to the cost of the nickel-based alkaline rechargeable batteries. Although iron is considerably less expensive than nickel, iron corrodes at the operating potential of the nickel hydroxide electrode. We have found that a 70–100 nm thick thermal coating of cobalt ferrite spinel protects the iron from corrosion. Such a coated iron substrate was found to be stable against corrosion even when polarized anodically at 10 mA cm -2 in 30% potassium hydroxide electrolyte for 1000 h. While the thermal coating of cobalt ferrite protected iron against corrosion, incorporation of lithium ions into the coating was found to enhance the electrical conductivity of the coating. XPS and EXAFS studies confirmed that the enhanced conductivity resulted from an increase in the population of Co 3+ in the ferrite spinel lattice. An inexpensive iron (steel) substrate protected by such a coating when used as a nickel hydroxide battery electrode exhibited a specific capacity of 0.25 Ah g -1 at C/5 discharge rate, comparable to a nickel hydroxide electrode based on a relatively expensive nickel foam substrate. The steel-based electrode also exhibited no noticeable degradation over 150 cycles at C/2 rate. This demonstration of a robust and economical steel substrate presents a unique opportunity for reducing the cost of the nickel hydroxide battery electrode in alkaline batteries

25 ENERGY STORAGE↗
DOE OSTI2020

Degradation of impact toughness in cast stainless steels during long-term thermal aging

Cast austenitic stainless steels (CASSs) have been extensively used for the large components of nuclear reactor primary coolant systems. Since the cast steels inevitably contain degradable metastable phases and replacement of the large coolant system components is impractical, the thermal embrittlement of CASS components has been a serious concern in the extended-term operation of nuclear power plants. This study aimed to systematically measure and analyze the effect of long-term thermal aging on the Charpy impact toughness to provide a comprehensive understanding of thermal degradation behavior and a practical aging model to predict the degree of thermal degradation in the cast stainless steels. Additionally, the materials tested in the research include eight CASS alloys (two CF3s, one CF3M, three CF8s, and two CF8Ms) and two reference wrought materials (304L and 316L), in which the nominal δ-ferrite content ranges from ~2% to 33%. These stainless steels have been thermally aged at two light water reactor (LWR) temperatures (290 and 330 °C) and at two accelerated-aging temperatures (360 and 400 °C) for up to 30,000 h; these include both under-aged and over-aged conditions relative to the extended service lifetime (80 years). Charpy impact testing was performed for aged and non-aged specimens, and the impact (absorbed) energy parameters were correlated with a new aging parameter (A). Both the reduction of impact fracture toughness and the shift of ductile-brittle transition temperature were strongly dependent on the δ-ferrite content and degree of thermal aging. A linear relationship was found between the increasing rate of the index transition temperature T 41J and aging parameter A; base on which an empirical model was proposed for prediction of the transition temperature as a simple function of the aging parameter (A) and δ-ferrite content (F). Finally, the critical aging parameter for embrittlement (AC) was evaluated and compared with the existing δ-ferrite content criteria.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The compositional optimization and secondary phases evaluation regarding the creep resistance in Grade 91 steel through the CALPHAD approach

The critical secondary phases along with critical temperatures in regard to the creep resistance of Grade 91 steel alloy were evaluated in order to optimize the composition of the alloy to improve the creep resistance. The critical temperatures which are highlighted in this paper are Ac1 (the threshold temperature in which austinite begins to form), and Ac3 (the threshold temperature at which ferrite is fully transformed into austenite). Furthermore, optimization was presented through various Cr, V, Nb, and N concentrations in Gr.91 in relation to these critical temperatures and the mole fraction of M 23 C 6 , MX, and Z-Phase as critical secondary phases. Here, the results indicate that increasing V and Nb will increase MX stability, whereas decreasing Cr and N lead to a decrease in Z-Phase and M 23 C 6 stability, which may lead to an increased creep resistance of the material. Upon further analyzing the V, Nb, and N concentrations in Gr.91, results indicate that increasing Nb content in the steel can lead to an increased stability of Nb-rich MX2 carbide which can be used to replace Z-Phase in low N steels. Simulation results indicate that elimination of N with a dramatic increase in Nb concentrations result in the stability increase of only M 23 C 6 and MX2 phases at temperature regions between 600 °C and 1370 °C and a complete removal of Z-Phase precipitation. Overall, with the elimination of N, which destabilizes Z-Phase precipitation, and an increase in Nb for NbC carbide stability, a possible solution to increasing both long-term and short-term creep resistance for Gr.91 can be achieved, though further optimization and creep experiments need to be conducted for final conclusions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The Complexity and Challenges of Irradiation-Induced Phase Instability Phenomenon in Nuclear Power Plant Components

High energy irradiation by neutrons or ions of metastable austenitic 300-series steels appears to initiate formation of bcc-phases with different morphology and elemental composition compared to that of retained ferrite. Extensive formation of specific Fe-rich ferrite was observed on the grain boundaries of annealed Russian steel 12Cr18Ni10Ti, an analog of AISI 321, after irradiation in the center-core region of sodium-cooled BN-350 fast reactor (SFR) doses up to 57 dpa at ~1×10-6dpa/sec. The ferrite fraction determined from scanning electron microscopy (SEM) closely matched the magnetic fraction determined using a ferritscope confirmed by X-ray diffraction analysis and selected area diffraction patterns. Chemical composition of the secondary phases was determined by energy dispersive X-ray analysis using transmission electron microscope, showing Fe-rich ferrite domains. Fe-ion irradiation (2.3 MeV) was used to study the accumulation of ferrite and martensite phases at higher dpa rates. The very high etchability of these ferrite particles in alcohol-based solutions may signal a very high etching in hot water, leading to accelerated intergranular cracking, especially upon long-term exposure during extended plant lifetimes.

D a, Merezhko↗
DOE OSTI2025

Interpreting experimental measurements of helium bubbles using stochastic cluster dynamics models of heterogeneous nucleation and growth in irradiated ferritic alloys

Among a number of other advantageous features, ferritic/martensitic steels are being considered as fusion reactor structural materials due to their low intrinsic swelling under irradiation. However, under high-energy neutron irradiation, He produced through (n, α) reactions stabilizes vacancy clusters, which then act as seeds for bubble formation and growth, negating the intrinsic swelling resistance of these alloys. Standard models of irradiation damage accumulation and microstructural evolution consider homogeneous nucleation as the basis for bubble formation and growth. However, this generally does not explain the large bubble densities and sizes observed experimentally under a number of different conditions. Here, we propose a more realistic physical model of bubble nucleation, formation, and growth designed to capture recent experimental measurements of He-bubble formation and evolution during co-implantation of He and Fe ions in model ferritic alloys. We find that experimental results are explained only when the following three features are considered simultaneously: (i) heterogeneous nucleation of He-vacancy bubbles at defect sinks (e.g., dislocations, grain boundaries, and second-phase precipitates), (ii) nucleation and growth of bubbles via the ‘trap mutation’ mechanism (i.e., spontaneous production of Frenkel pairs due to absorption of He atoms), and (iii) transition from a growth-limited, He-stabilized bubble regime to a ‘runaway’ void-type growth scenario. The model is implemented into a stochastic cluster dynamics framework that takes advantage of cluster size grouping methods to accelerate the simulations, allowing us to reach 10 dpa of simulated irradiated dose, and to capture cluster sizes in excess of 20 nm. Further, a careful extrapolation exercise conducted assuming classical nucleation theory leads to excellent agreement with the experimental measurements at 50 dpa of irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Pre-Oxidation to Improve Liquid Metal Compatibility

Liquid metals represent an aggressive environment where selective degradation can attack structural alloys. Pre-oxidizing structural alloys, particularly those able to form alumina scales, have shown promise for Sn, Pb and Pb–Li concepts for fusion and fission applications. For Pb at 700 °C, a dramatic reduction in mass loss was observed for an alumina-forming austenitic (AFA) steel after pre-oxidation. Likewise, a ferritic FeCrAlMo alloy showed better compatibility in static Sn at 400–500 °C and flowing PbLi when pre-oxidized at 1000 °C to form α-Al2O3. Recent experiments with flowing PbLi at 675–700 °C has shown significant degradation suggesting that alumina-forming alloys are limited to ~ 650 °C in this environment. While the concept is promising, most experiments have been relatively short-term and long-term degradation mechanisms need to be evaluated.

Pint, Bruce↗
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

Thermomechanical Behavior of Advanced Manufactured Parts, Subcomponents, and Their Weldments for Gen3 CSP

Generation 3 (Gen3) concentrating solar power (CSP) plants may require the use of molten chloride salt storage systems, solar receivers, and supercritical-CO 2 primary heat exchangers (PHX). The temperatures that would be expected in these parts and subcomponents could approach 760°C for hot side and 500°C for cold side at peak operating conditions. With the design limitations, highly corrosion- and creep-resistant alloys are needed for maximizing component lives. This report presents the results for the project "Thermomechanical Behavior of Advanced Manufactured Parts, Subcomponents, and their Weldments for Gen3 CSP", award number DE-EE00036334. In this project, creep enhanced ferritic alloy Grade 91, austenitic stainless-steel (SS) 304H, Ni claddings Ni201 and C22, and nickel-based superalloys Inconel 740H and Haynes 282 and 230 were evaluated for potential applications in Gen3 CSP systems. Advanced manufacturing of these parts, subcomponents and their welds, was investigated and a full technoeconomic analysis was made in comparison to conventional manufacturing techniques. The manufacturing techniques explored are explosion clad welding and combustion synthesis/combustion reaction for transfer pipes, additive manufacturing (AM) including laser-powder bed fusion (L-PBF) and electron beam AM (EBAM) with wire feedstock for PHX and solar receivers, and conventional fusion welding for similar and dissimilar joining of these various parts and subcomponents.

14 SOLAR ENERGY↗
DOE OSTI2025

A highly wear resistant nanostructured bainitic steel with accelerated transformation kinetics

A coupled Calculation of Phase Diagrams (CALPHAD), machine learning, and data mining approach was used to design a new, highly wear-resistant nanostructured bainitic steel. Arc melting of the designed compositions, dilatometry, and advanced microscopy indicate that the designed steel had a nanoscale dual-phase structure of ferrite and austenite (approximately 50 nm) with kinetics 7x faster for the onset of bainite and 2x faster for complete transformation. Under dry sliding conditions using the current state-of-the-art AISI 52100 bearing steel as the counter sample, the designed steel little to no wear, indicating its potential for applications in high-wear service conditions.

36 MATERIALS SCIENCE↗