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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Long Term Creep Behavior of a CoCrFeNiMn High Entropy Alloy

The potential of high entropy alloys (HEAs) to meet or exceed austenitic stainless-steel performance with the additional benefit of improved hot corrosion/oxidation resistance makes FCC HEAs attractive for use in energy applications. While shorter term creep tests have been reported in the literature on HEAs, not all methodologies utilize repeatable techniques. This manuscript reports on over 23,500 accumulated hours of tensile creep testing with adherence to ASTM standards on a melt solidified ingot of CoCrFeNiMn HEA converted to wrought plate using conventional thermo-mechanical processing techniques. The typical standard creep analyses are reported, i.e., Larson-Miller parameter, Monkman-Grant relationship, activation energy for creep, and creep stress exponents were calculated and compared to previously reported short-term creep tests. Additionally, characteristics of creep fracture and microstructural evolution are reported with cursory dislocation mechanisms investigated.

Rozman, Kyle A.↗
DOE OSTI2022

Long-Term Creep Behavior of a CoCrFeNi Medium-Entropy Alloy

We report that current research shows that high-entropy alloys (HEAs) can offer superior oxidation resistance compared to more traditional austenitic stainless steels with equivalent tensile mechanical performance. While many reports have been published on creep of CoCrFeNiMn, few focus on the CoCrFeNi family of low to medium entropy alloys. In this manuscript, uniaxial tensile creep test results for more than 46,000 accumulated hours are presented on the equiatomic CoCrFeNi alloy. The alloy was manufactured at NETL using a vacuum induction melt solidification approach with a computationally designed homogenization cycle. After completing the homogenization cycle, the ingot was first hot forged and then hot rolled into 10 mm thick plate, a product form that is relevant to many industries. Creep tests were performed in the temperature range of 600-650 °C using applied stresses from 120 to 189 MPa. This study builds upon previous results on the CoCrFeNiMn family of high-entropy alloys and offers a modicum of clarification about performance compared to small button melts by using ASTM standard test procedures and mechanical test specimens. The apparent creep activation energy and creep stress exponent were similar to those for CoCrFe medium-entropy and CoCrFeNiMn high-entropy alloys. The post-test creep microstructure showed some Cr segregation, which has also been noted for other CoCrFe alloys. The CoCrFeNi alloy also presents a double minimum or two minimum creep rate phenomena.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A thermochemical database from high-throughput first-principles calculations and its application to analyzing phase evolution in AM-fabricated IN718

A comprehensive thermochemical database is constructed based on high–throughput first-principles phonon calculations of over 3000 atomic structures in limited concentrations in Ni, Fe, and Co alloys involving a total of 26 elements including Al, B, C, Cr, Cu, Hf, La, Mn, Mo, N, Nb, O, P, Re, Ru, S, Si, Ta, Ti, V, W, Y, and Zr, providing thermochemical data largely unavailable from existing experiments. Here, the database can be employed to predict the equilibrium phase compositions and fractions directly from first-principles by minimizing the chemical potential of a multicomponent system with a fixed overall chemical composition and a fixed temperature. It is applied to the additively manufactured nickel-based IN718 superalloy to analyze the phase evolution with temperature. IN718 is known for its great performance in tensile, fatigue, creep, and rupture strength, combined with easy fabrication and corrosion resistance. In particular, we successfully predicted the formation of L1 0 -FeNi, γ’-Ni 3 (Fe,Al), α-Cr, δ-Ni 3 (Nb,Mo), γ”-Ni 3 Nb, and η-Ni 3 Ti at low temperatures (below 680 K), γ’-Ni 3 Al, δ-Ni 3 Nb, γ”-Ni 3 Nb, α-Cr, and γ-Ni(Fe,Cr,Mo) at intermediate temperatures (between 680 and 1140 K), and δ-Ni 3 Nb and γ-Ni(Fe,Cr,Mo) at high temperatures (above 1140 K) in IN718. These predictions are validated by EDS mapping of compositional distributions and corresponding identifications of phase distributions. The database is expected to be a valuable source for future thermodynamic analysis and microstructure prediction of alloys involving the 26 elements.

36 MATERIALS SCIENCE↗
DOE OSTI2024

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Irradiation Damages of Structural Materials under Different Irradiation Environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study’s Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

radiation damage, structural alloys, neutron irrad↗
DOE OSTI2025

Computational Design of Improved Fast Reactor Cladding

HT9 ferritic-martensitic (FM) steel has served as a leading candidate for sodium-cooled fast reactor (SFR) cladding due to its favorable resistance to irradiation-induced swelling and good thermal and chemical properties. However, its limited creep strength at temperatures above 600 °C and susceptibility to α′ phase embrittlement under specific conditions could limit its application in next-generation SFRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Mechanical creep testing and characterization of Cr coated Zircaloy

Zirconium based cladding has been developed for use in light water reactor fuel applications due to the material’s desirable low neutron scattering cross section. To help improve corrosion resistance in accident conditions, such as a loss of coolant accident (LOCA), a chromium coating can be applied to the outer diameter of the Zircaloy tube. Cladding strength at reactor relevant temperatures is important for the structural performance. Stress relaxation creep testing was performed on Cr coated Zircaloy-4 to measure strain rate as a function of stress for 350 and 650°C. These were performed at both 5 ± 1 % and 7.5 ± 1 % strain. The lower temperature provides results near normal operating temperature, while the higher temperature provides results relevant to off-normal conditions where recrystallization will take place. Characterization was also performed to analyze the structure of the Cr coating and Zircaloy-4 as well as the interface between the two.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

In situ self-ion (Fe + ) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation

Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used in situ self-ion irradiation (150 keV Fe + ) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe + irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10 15 ions·cm -2 (2 dpa) and 3.4 × 10 15 ions·cm -2 (4 dpa), using three damage dose rates (5 × 10 –5 , 3.3 × 10 –4 , and 2 × 10 –3 dpa·s -1 ). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Complete Optimization of LPBF Ni-Based Alloys Down-Selected from FY23 Candidate Materials Including, Thermodynamic Modeling, Sample Fabrication and Microstructure Characterization

The goal of the Advanced Materials and Manufacturing Technologies (AMMT) program is to accelerate the incorporation of new materials and manufacturing technologies into advanced nuclear-related systems. Although 316H stainless steel fabricated by laser powder bed fusion (LPBF) has already been identified as an alloy that could have a significant effect on various reactor technologies, many other materials and manufacturing techniques are being evaluated. Nickel-based alloys typically offer higher-temperature capabilities compared with advanced stainless steels, and previous reports looked at three Ni-based alloy categories: low-Co alloys with a potential use close to the reactor core; high-temperature, high-strength alloys; and molten salt–compatible alloys. In the first category, alloy 718 was studied in 2023, and creep testing at 600°C and 650°C revealed that the alloy exhibited great creep strength after the appropriate annealing but had low ductility. Advanced characterization was recently conducted to highlight the presence of strengthening γ' and γ" precipitates after creep testing and to show that brittle phases at grain boundaries might explain the low ductility of LPBF 718 compared with wrought 718. For the high-temperature, high-strength alloys, previously purchased powders of alloys 617, 230, and 625 were used to assess the printability of these three solution-strengthened alloys. Hot cracking could not be suppressed for alloy 617 and 230, and it was shown that these cracks, which were elongated along the build direction (BD), had a drastic effect on the ductility of alloy 230 at room temperature when specimens were machined perpendicular to the BD. On the contrary, LPBF printing of crack-free alloy 625 was achieved using similar printing parameters, and the alloy looked like a promising candidate for various reactor technologies. The fabrication of alloy 282 by LPBF, a γ'-strengthened alloy with great creep strength up to 800°C, was performed in 2023, and x-ray computed tomography (XCT) scans of the alloy before and after creep testing at 750°C were carried out to assess the effect of flaws on the alloy’s creep behavior. Correlation between the flaws’ volume fraction, creep ductility, and creep lifetime could be established, and future work on LPBF 625 will take full advantage of in situ printing data and ex situ XCT scans to accelerate the alloy qualification. Finally, single track experiments were performed on the two alloys previously identified as good molten salt–resistant, Ni-based candidates: Hastelloy N and 244. Various laser parameters were considered, and cracking was not observed for either of the two alloys. Wrought 244 offers better creep strength and molten salt compatibility than alloy 625, and future work will aim to establish the alloy LPBF processing window.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Characterization of Precipitation-Strengthening Heat-Resistant Austenitic Stainless Steels for Life-Prediction Modeling

In this study, the role of minor alloying additions in 347H stainless steels (UNS34709, ASTM A240/240M) on creep-rupture properties at 650-750°C and microstructure evolution during isothermal exposure at 750°C has been investigated, aiming to provide the experimental dataset as boundary conditions of physics-based modeling for material/component life prediction. Four different 347H heats containing various amounts of boron and nitrogen additions were prepared and evaluated. The combined additions of B and N are found to stabilize the strengthening secondary M23C6 carbides and retarding the transition from M23C6 to sigma phase precipitates during thermal exposure. The observed kinetics of microstructure evolution reasonably explains the improvement of creep-rupture properties of 347H stainless steels with the B and N additions.

Yamamoto, Yukinori↗
DOE OSTI2021

Creep of MARBN-type 9Cr martensitic steel in gaseous CO 2 environment

Here, with an interest in moving from steam to supercritical CO 2 as the working fluid in advanced energy systems, it is important to study the mechanical response of structural alloys to CO 2 -containing environments. MARBN-type 9Cr martensitic steel was originally developed for application in boilers for supercritical and ultra-supercritical power plants where the steam temperature reaches 650 °C. In this research, a MARBN-type 9Cr martensitic steel designed and manufactured at NETL was creep tested in a gaseous CO 2 environment (i.e., 0.1 MPa at 650 °C), and the results are directly compared to creep tests conducted in air. It was found that environmentally-assisted cracking facilitated by carbide formation beneath the growing oxide accelerated failure in gaseous CO 2 . This work indicates that creep-oxidation interactions in advanced martensitic steels represent an important consideration for materials selection in supercritical CO 2 power cycles. Furthermore, this work confirms that more oxidation-resistant alloys will be required for the highest temperature portions of these systems.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Theory-guided Design of Refractory MPEAs for High-Temperature, Harsh-Service Conditions

Focusing on novel materials for harsh-service environments (oxidation, corrosion, and load), durability, higher operational temperature, and direct impact on power-generation technology, the goal was, using a quantitative theory-guided design approach integrated with high-throughput synthesis and characterization for validation, to accelerate by 50% the development of high temperature, refractory-based multi-principal element alloys (MPEAs) materials platforms (with co-designed oxidation-resistant, self-healing coatings) that demonstrate phase-stable operation with desired properties at 10-20% higher temperatures in oxidizing environments over state-ofthe- art systems (like Ni-stainless steels, i.e., Haynes 282 – used in critical gas turbine applications, and refractory TZM, i.e., molybdenum-rich Mo99.4-Ti0.5-Zr0.08-C0.02 in wt%). Notably, refractory MPEAs may achieve higher operational temperatures with superior creep strength and offer 50-100% larger thermal conductivity (140-170 W/m-K) than Ni-based systems, potentially eliminating active cooling, reducing system weight, complexity, and cost. Achieving this goal addresses DOE material challenges in energy generation and efficiency (higher-temperature operation), reduced lifecycle energy (waste-heat recovery), and accelerating materials development in support of AMO’s Strategic Plan and QTR goals. DOE’s Advanced Manufacturing Office (AMO) goal of reducing industrial energy intensity and GHG emissions is addressed by development of coated refractory MPEAs exhibiting 20%+ higher operational temperature (direct increase of Carnot efficiency), higher melting temperatures (better phase stability), superior creep strength (better lifetime with creep rate lower by factor of 10), larger (50-100%) thermal conductivity (improved cooling without active cooling), reduced complexity (single-phase alloys), self-healing coating (reliable oxidation-resistance) and cost.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Concrete shrinkage and creep under drying/wetting cycles

Concrete shrinkage and creep under variable hydric conditions are important factors for the safety and durability of concrete especially in nuclear reactor or nuclear waste storage background. A large (and long – more than 900 days) experimental campaign, conducted on two different concretes, has therefore been designed to study the strains of non loaded and loaded concretes submitted to drying and liquid water imbibition cycles. For the purposes of comparison, concrete strains and mass variations during drying only (50% RH) and/or following cycles of drying/rewetting were also recorded. This allowed the identification of desiccation shrinkage, basic creep and drying creep at 10 MPa of axial stress. Important results were found and have shown that the final mass and strain are not deeply modified by the introduction of a rewetting phase either for a loaded or a non loaded material.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time of flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Prediction of moisture transfer in cement-based materials: Use of a porous network model to access transfer parameters

Knowing the water transport phenomena is fundamental for understanding the durability of cement-based materials. The moisture content within cementitious materials is of primary importance because it has a direct influence on mechanical (shrinkage, creep) and chemical (penetration of aggressive agents) degradation. Many models based on the determination of the transfer properties of the material under study already exist but require fitting of the parameters to find the experimental kinetics. Transfer parameters are generally estimated from the empirical approaches developed for soils by Van Genuchten and the water vapour diffusion reduction factor is often calculated via Millington's semi-empirical model. In this paper, we make a proposal to improve the prediction of moisture transfer, through mechanisms such as drying, capillary imbibition and water vapour permeability in cement-based materials, by using a porous network model to access the transfer parameters.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effect of Transverse Compression on the Dielectric Breakdown of Power Cable with Ethylene Propylene Rubber Insulation

Understanding of the failure mechanisms of submarine dynamic power cables (SDPC) is critical for innovative design to meet the 2035 cost reduction target of U.S. DOE Floating Offshore Wind Shot. This is important because the current design suffers a significant failure rate in the field. This project explored a multiphysics approach and conducted systematic electro-thermo-mechanical (ETM) experimental study on the power cores extracted from a 15 kV power cable. The power cable had three cores of copper conductor and ethylene propylene rubber (EPR) insulation. An ETM testing system with transverse compression capability was developed. Mechanical loading induced dielectric breakdown. Increasing temperature from room temperature (RT, 22 degrees Celsius) to 90 degrees Celsius resulted in the 67% decrease in the failure mechanical load as defined by the dielectric breakdown. With the same temperature rise, the creep mechanical load was reduced by 70% at a given dielectric breakdown time. The post-test measurement revealed an impressive recovery of electrical resistance, although the dielectric breakdown occurred in the ETM test that would not be captured without the use of ETM testing.

Wang, Hong [ORNL] (ORCID:0000000201730545)↗
DOE OSTI2021

Plastic deformation mechanism of calcium-silicate hydrates determined by deviatoric-stress Raman spectroscopy

Highlights: • C-(A-)S-H samples under stress were characterized at the bond and grain level. • Small angle scattering showed that 720 MPa reduced grain thickness by 30%. • In-situ Raman showed that cross-linked C-A-S-H slips along its intralayer. • However, in-situ Raman showed that C-S-H slips along the interlayer. • In-situ Raman also showed a growth in CaCO{sub 3}, indicating fracture in both samples. Creep of the cement matrix affects the structural stability of concrete. In Portland cements, the creep is largely controlled by the binding phase calcium-(aluminum-)silicate-hydrate, or C-(A-)S-H. This phase has a lamellar structure and under deviatoric stress aligns its c-axis with the principal stress. However, the limiting resistance to this reorientation is unknown at the nanocrystalline level. Small-angle X-ray scattering shows that the lamellae thickness decreases under 100's MPa deviatoric stress. Deviatoric stress Raman spectroscopy shows that there are two ways that this break-up can occur. If the material's silicate chains are cross-linked, then strain in SiO bonds does not increase above certain stresses, indicating a relaxation adjacent to the SiO bond. If the chains are not cross-linked, then the silicate chains are broken up by rastering against each other, introducing defects. These results show that the plastic deformation of C-(A-)S-H is relevant for Portland cement creep.

36 MATERIALS SCIENCE↗