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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 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Exploiting ultra-large linear elasticity over a wide temperature range in nanocrystalline NiTi alloy

Many shape memory alloys can support large recoverable strains of a few percent by reversible stress-induced martensite transformation, yet they behave non-linear within a narrow operating temperature range. Developing the bulk metallic materials with ultra-large linear elasticity over a wide temperature range has proven to be difficult. In this work, a material design concept was proposed, that is true elastic deformation and reversible twinning-detwinning deformation run in parallel to overcome this challenge. By engineering the residual internal stress to realize the concurrency of true elastic deformation and twinning-detwinning deformation, a bulk nanocrystalline NiTi that possesses an ultra-large linear elastic strain up to 5.1 % and a high yield stress of 2.16 GPa over a wide temperature range of 270 °C was developed. This study offers a new avenue for developing the metallic materials with ultra-large linear elasticity over a wide temperature range of 270 °C (from 70 °C to -197 °C).

36 MATERIALS SCIENCE↗
DOE OSTI2022

Variant selection in laser powder bed fusion of non-spherical Ti-6Al-4V powder

The presence of α/α' on prior β/β grain boundaries directly impacts the final mechanical properties of the titanium alloys. The β/β grain boundary variant selection of titanium alloys has been assumed to be unlikely owing to the high cooling rates in laser powder bed fusion (L-PBF). However, we hypothesize that powder characteristics such as morphology (non-spherical) and particle size (50–120 µm) could affect the initial variant selection in L-PBF processed Ti-6Al-4V alloy by locally altering the cooling rates. Despite the high cooling rate found in L-PBF, the results showed the presence of β/β grain boundary α' lath growth inside two adjacent prior β grains. Electron backscatter diffraction micrographs confirmed the presence of β/β grain boundary variant selection, and synchrotron X-ray high-speed imaging observation revealed the role of the “shadowing effect” on the locally decreased cooling rate because of keyhole depth reduction and the consequent β/β grain boundary α' lath growth. The self-accommodation mechanism was the main variant selection driving force, and the most abundant α/α boundary variant was type 4 (63.26°//[ $\bar{10}$ 5 5 $\bar{3}$ ]). The dominance of Category II α lath clusters associated with the type 4 α/α boundary variant was validated using the phenomenological theory of martensite transformations and analytical calculations, from which the stress needed for the β→α' transformation was calculated.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of carbonitrides and carbides on microstructure and properties of castable nanostructured alloys

In this work, to develop advanced reduced-activation ferritic-martensitic (RAFM) steels for fusion reactor structural applications, both carbonitride- and carbide-strengthened castable nanostructured alloys (CNAs) were explored for higher densities of MX (M = Ti/Ta/V/etc. and X = C/N) nanoprecipitates. Additionally, systematic comparisons between the two types of CNAs indicated generally similar microstructures and comparable tensile properties and creep resistance. However, the carbide-CNAs did show some advantages over the carbonitride-CNAs in terms of the uniformly distributed higher density of MC nanoprecipitates, greater Charpy impact upper shelf energies, less deuterium retention and swelling, and potentially less transmutation-induced composition changes and consequently thermodynamically more stable carbides. The carbide-CNAs showed the best-balanced high performance in the examined properties, in contrast to the significantly varied performance of oxide-dispersion-strengthened alloys and the generally lower performance of current RAFM steels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. In this work, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal ( U ,Zr) 6 Fe phase (space group I4/mcm) and cubic ( U ,Zr)( Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The ( U ,Zr)( Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Deuterium transport and retention properties of representative fusion blanket structural materials

Reduced activation ferritic-martensitic (RAFM) steels have been developed for decades for use as fusion blanket structural materials, and have advantages in both mechanical properties and irradiation resistance following careful engineering of the microstructure. However, the hydrogen isotope behavior in these proposed fusion structural materials is not well understood, but is important to assess since it impacts the fusion reactor safety and self-sufficient tritium fuel cycle. Here, we investigated deuterium transport and retention in representative advanced RAFM steels, including castable nanostructured alloys (CNAs), and oxide-dispersion-strengthened (ODS) steels. A gas-driven permeation (GDP) system was used to measure the permeability, diffusivity and solubility of the studied materials, covering the temperature range from 623 K to 873 K, and the loading pressures from 1.8 x 10 4 to 1.0 x 10 5 Pa. The results indicated that the deuterium permeability has little material dependence. In contrast, the deuterium diffusivity of the studied materials showed significant variation. The deuterium diffusivity in ODS steels is one order of magnitude lower than that in RAFM steels and CNAs, and correspondingly, have an effective solubility that is 2–10 times larger than RAFM steels and CNAs. In addition, thermal desorption spectroscopy (TDS) measurements were performed to assess the deuterium retention and desorption of these materials following a static thermal deuterium charging at 723 Kfor 1 hour under the deuterium pressure of 1.0 x 10 5 Pa. It was found that ODS steels exhibit the highest deuterium retention and have broader desorption peaks. Microstructural features contributing to deuterium retention and impacting deuterium transport are discussed to rationalize the observed deuterium behavior in the studied RAFM steels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Neutron irradiation induced defects and clustering in NF616 and T91

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel

We report that dual-ion irradiations using 5.0 MeV defocused Fe 2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He 2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 × 10 -5 dpa/s to 3 × 10 -3 dpa/s at 445°C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation.

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 OSTI2022

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

We report Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ~10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Creep behavior of an additively manufactured 9Cr steel in the as-built condition

Limited studies have evaluated the creep behavior of additively manufactured (AM) ferritic-martensitic (FM) steels. This work investigated the creep behavior of a 9Cr FM steel fabricated by powder blown directed energy deposition (DED) technique. Here, the creep testing at 550–650 °C and 150 MPa for the specimens along the deposition direction in the as-built condition, together with corresponding microstructural characterization, revealed a threshold temperature between 600 and 625 °C, below which the steel has creep resistance comparable with Grade 91 cross-welds and noticeably greater than 9Cr-1Mo steel. The threshold temperature distinguishes the creep behavior in two regimes differentiated in creep activation energy, creep deformation, and failure mechanism. Unlike the creep rupture surface ~45° from the loading direction when tested above the threshold temperature, the creep rupture for testing below the threshold temperature resembles type IV failure in the cross-welds of ferritic steels. The DED-induced layer structure in the as-built steel played a significant role on the change of creep behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The Thermo-Elastic Properties and Damping of U-6wt%Nb

While thermal expansion data exists for quenched (as well as aged) U-6wt%Nb, there is wide variation in the reported room temperature elastic moduli. To better understand the room temperature elastic behavior and to address the complete absence of data on the temperature dependence of the elastic moduli, room temperature and in-situ elevated temperature resonant ultrasonic spectroscopy (RUS) was performed along with x-ray diffraction (XRD) and dilatometry using a thermomechanical analyzer (TMA). An in-situ small- and wide-angle x-ray scattering (SAXS/WAXS) experiment was performed on a companion sample to help interpret the results. The room temperature polycrystalline dynamic moduli were measured to be E = 95.4 ± 4.5 GPa and G = 35.2 ± 0.2 GPa, respectively. As the temperature is raised, the stiffness slowly decreases, consistent with the empirical rule proposed by Varshni, up to the point of precipitation of the equilibrium α-U phase, which is stiffer than the martensitic phases. Homogenization theory of composites can be used to rationalize the observed response at high temperatures and after high-temperature exposures. Further, aging of the material at low temperatures (≤200°C) does not affect the linear elastic stiffness, but does impact the damping behavior which can be measured through RUS. This change in damping provides another perspective on the microstructure changes induced by low-temperature aging which also result in significant strengthening.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Characteristics of oxide-dispersion strengthened alloys produced by high-temperature severe deformation

This study is to explore an economically attractive and technically feasible processing method for oxide-nanoparticle strengthened alloys for fusion reactor application. Despite many scientific merits of the advanced oxide-dispersion strengthened (ODS) alloys, such as the nanostructured ferritic alloy (NFA) 14YWT, the only viable production path for a high-quality NFA is the high-power mechanical alloying process. This process is often a multi-day high-speed ball milling of alloy powder with a small quantity of yttria (Y 2 O 3 ) powder, followed by the milled-powder consolidation using extrusion or other methods and additional thermomechanical processing (TMP) for property control. This complex production path, including the low-temperature mechanical alloying in particular, has limited technical advancement toward the cost-effective and scale-up production of ODS alloy components. To overcome such a practical limitation, we proposed to explore alternative low-cost processing routes using traditional thermomechanical processing (TMP) method only. Further, a series of continuous TMP cycles, which were designed to impose high-temperature severe plastic deformation (HT-SPD) conditions to the consolidated powder mixtures, were applied to achieve the effective distribution of oxide particles in nanograin structure and thus desirable mechanical properties. Since the reduced-activation ferritic-martensitic (RAFM) alloy powders (Fe-10Cr and Fe-14Cr alloys with various Y contents) are available in our inventory, we focused to utilize the new solid-state synthesis approach for controlling oxide (oxygen source) dissolution and nanoscale clustering in nanograin structure in those alloys. A combination of powder consolidation at 900 °C and continuous thermomechanical activation at 600 °C yielded two essential ODS alloy microstructure contents–nanograin structure and nanoparticle distribution–and thus demonstrated a good combination of strength and ductility.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Gas tungsten arc welding and post weld heat treatment effects on microstructure and mechanical property of castable nanostructured alloy steel

Here, this paper details the first study of castable nanostructured alloy (CNA) steel gas tungsten arc weldability and the post-weld heat treatment (PWHT) effects. Effects of welding heat input, thermal cycles, and PWHT on microstructures, microhardness distributions, room temperature tensile properties, and fracture characteristics are discussed. Results show that CNA steel exhibits excellent weldability (i.e., no indication of welding defects and reasonable tensile properties). The welded joint exhibited heterogeneous microstructures with δ-ferrite as well as large microhardness variation and fluctuation. The welded joint yield and ultimate tensile strengths were similar to those of the base metal, but the elongations decreased by 30 %. However, with normalization and tempering PWHT, the δ-ferrite was eliminated, microstructure was modified, hardness was unified, and joint ductility was restored. The study indicated that the CNA reduced-activation ferritic-martensitic steel owns excellent weldability, and PWHT is needed for the industrial application of welded structures.

36 MATERIALS SCIENCE↗
DOE OSTI2026

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

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

fusion materials↗
DOE OSTI2022

Additive friction stir deposition of SS316: Effect of process parameters on microstructure evolution

Solid state nature of additive friction stir deposition (AFSD) additive manufacturing process is very advantageous in terms of defect formation and microstructural refinement in the material. Current study presents the process optimization, microstructural evolution and kinetics of recrystallization for AFSD deposited low stacking fault energy material - SS316. As deposited microstructure shows equiaxed ultra fine grains with an average grain size of ~5.0 ± 0.5 μm. Shear deformation at high temperature during processing leads to the operation of restoration mechanisms. Observation of necklace type microstructure in the as deposited SS316 is attributed to discontinuous dynamic recrystallization during processing. Recrystallization kinetics of the AFSD SS316 is characterized using Johnson-Mehl-Avarami-Kolmogorov (JMAK) model. Deformation – thermal cycling during AFSD process resulted in inconsistent recrystallization kinetics. Variation in strain, strain rate and temperature during processing with processing parameters result in varying microstructure and tool wear. Further, high strength-ductility combination and sustained work hardening in as deposited SS316 appear to arise from transformation and twinning during deformation, leading to the formation of hierarchical twins and martensitic phase after deformation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Characterization of Fuel Cladding Chemical Interaction on a High Burnup U-10Zr Metallic Fuel via Electron Energy Loss Spectroscopy Enhanced by Machine Learning

Fuel cladding chemical interaction (FCCI) is one of the main performance limiting factors for metallic nuclear fuels. The interaction destabilizes the martensitic microstructure and deteriorates mechanical properties of HT-9 cladding. The detection of low atomic number elements (Z<10) and overlapping of elemental peaks can be problematic in interpreting energy dispersive X-ray spectroscopy (EDS) data. Electron energy loss spectroscopy (EELS) provides precise elemental edge energy values and can detect elements with a low atomic number. This work utilizes EELS to study the distribution of lanthanides and light elements at the interaction region. The sample was prepared from the FCCI region of a U-10Zr (wt.%) solid fuel with HT-9 cladding, irradiated to a burnup of 13.2 at.%. Processing the EELS data included three major steps: 1) enhance the signal to noise ratio by denoising the spectrum with principal component analysis (PCA) method, removing background and performing deconvolution; 2) identify chemical elements with core energy loss edges; 3) confirm different phases using a popular machine learning method, K-means. This work presents qualitative assessment of lanthanides and light elements like carbon (C) and oxygen (O) enhanced by the application of machine learning algorithms. By comparing with EDS elemental maps, EELS provides higher resolution chemical maps, reveals the distribution of carbon at the interaction region supporting the formation of zirconium carbide, a rind-like microstructure feature that was proposed to mitigate the chemical interaction. Furthermore, the plasmon peak map was also found to indicate an energy shift associated with the formation of phases/compounds. K-means clustering method was used on the processed electron energy loss (EEL) spectrum to automatically reveal different phases. The resulting clustered maps from K-means clustering align well with elemental maps confirming certain phases, especially Fe-Ce and Zr-C, in the FCCI region.

EELS↗
DOE OSTI2024

Active interlocking metasurfaces enabled by shape memory alloys

Interlocking metasurfaces (ILMs) are a newly developed joining technology that relies on arrays of interlocking features that transmit force and constrain motion between adjoining bodies in one or more directions. This study explores harnessing the shape memory effect (SME) in Nickel-Titanium shape memory alloys (NiTi SMAs) in structures fabricated using additive manufacturing (AM) to advance the development of active ILMs by creating unit cells that open or close at specific temperatures. The study encompasses designing and fabricating two distinct interlocking array configurations using near-equiatomic NiTi powder and the laser powder bed fusion (L-PBF) AM technique, following a previously developed AM process optimization framework to manufacture defect-free parts. To guide the design process, finite element analysis (FEA) was employed to predict strain values during engage-disengage cycles. The martensitic transformation characteristics of the ILMs were characterized. Thermomechanical testing revealed that the ILMs demonstrate high locking force once engaged, coupled with complete shape recovery and good cyclic stability. Digital image correlation (DIC) was also employed to validate the FEA predictions during the engage-disengage cycles. The results indicate that NiTi SMA-based ILMs can be designed and fabricated into complex shapes using L-PBF. By leveraging the SME, the functionality of an ILM can be improved upon. The combination of computational modeling, additive manufacturing, and thermomechanical and physical property characterization provides a framework for designing future ILMs out of active materials.

Additive manufacturing↗