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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

A microscopic and crystallographic study of proton irradiated alloy 718

Solution annealing and age hardening are important processes for achieving good engineering and chemical properties of alloy 718. The composition of alloy 718 also plays an important role as it can affect both mechanical (γ”- and γ’-phase formation) and irradiation behaviors. Therefore, in this study five different sets of alloy 718 samples with two different chemical compositions were fabricated using different processing conditions and irradiated up to 4 dpa. Based on the effects of irradiation on the microstructural data of γ”- and γ’-phases, irradiation-induced dislocation statistics, and crystallographic data, it can be inferred that the samples processed at high solution annealing temperature (1093°C) behave better under irradiation compared to the samples processed at low solution annealing temperatures (945 and 1065°C).

36 MATERIALS SCIENCE↗
DOE OSTI2021

Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr) 2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr) 2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High radiation tolerance of an ultrastrong nanostructured NiCoCr alloy with stable dispersed nanooxides and fine grain structure

The present paper reports the irradiation behavior of nanostructured NiCoCr medium entropy alloy enhanced by Y-Hf-O nanooxides and fine grains produced by powder metallurgy. Ion beam irradiation to a peak dose of 130 dpa at 580°C produced neither detectable void swelling nor irradiation-induced hardening, in contrast to significant void swelling and property degradation of single phase NiCoCr concentrated solid-solution alloy under the similar irradiation condition. The dispersed nanooxides and high density of grain boundaries act as defect sinks that effectively annihilate irradiation-induced interstitials and vacancies. Furthermore, high-density nanooxides maintain their crystalline structures and stabilize the grain boundaries during irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Orientation-selected micro-pillar compression of additively manufactured 316L stainless steels: Comparison of as-manufactured, annealed, and proton-irradiated variants

In this work, irradiation response and deformation mechanisms of additively manufactured (AM) 316L stainless steel were studied by atomic scale characterization and micro-pillar compression. The AM 316L stainless steels were fabricated by direct energy deposition, a laser-based additive manufacturing process. Irradiation with 2 MeV protons at 360°C was performed to create ~1.8 displacements-per-atom (dpa) damage in AM 316L. Deformation behaviors of the as-manufactured, annealed, and proton-irradiated variants were studied, focusing on the effects of manufacturing-induced pores, residual stress, and irradiation-introduced defects (dislocation loops and voids). Micro-pillars were prepared from grains of pre-selected orientation, avoiding contributions of grain boundaries and allowing determination of resolved shear stress on {111} glide planes. Transmission electron microscopy was used to characterize the pre- and post-deformation microstructure. It was found that in the as-manufactured alloy variant, moving dislocations were the major deformation carrier, with noticeable blocking by fabrication-induced pores, In the annealed variant, hardness was reduced, and deformation was also accomplished by dislocation gliding. In the proton-irradiated variant, significant twinning was observed. Comparing measured resolved shear stress and predicted critical stress for dislocation dissociation, we conclude that irradiation hardening became high enough to activate twinning. Therefore, the deformation mechanism changes from dislocation gliding to twinning. The study is important for both processing optimization and performance evaluation of AM alloys for reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Friction stir welding and self-ion irradiation effects on microstructure and mechanical properties changes within oxide dispersion strengthened steel $\mathrm{MA956}$

We report the joining process for oxide dispersion strengthened (ODS) alloys remains a key challenge facing the nuclear community. The microstructure and mechanical properties were characterized in the base material and friction stir welded ODS MA956 irradiated with 5 MeV Fe 2+ ions from 400 to 500°C up to 25 dpa. Nanoindentation was performed to assess changes in hardness and yield stress, and the dispersed barrier hardening (DBH) model was applied to described results. A combination of scanning transmission electron microscopy (STEM) and atom probe tomography (APT) were used to assess evolution of the microstructure including dispersoids, network dislocations and dislocation loops, nanoclusters, and solid solution concentrations. Overall, softening was observed as a result of increased dose, which was exacerbated at 500°C. The formation and coarsening of new dispersoids was noted while nanoclusters tended to dissolve in the base material, and were not observed in the stir zone. Solute nanocluster evolution was identified as a primary driver of the changes in mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Design considerations for high entropy alloys in advanced nuclear applications

The demanding operating environments of advanced nuclear reactors require the development of new nuclear materials that can withstand their increased physical, chemical, thermal, and radiation-related challenges. High-entropy alloys (HEAs) have shown often-impressive mechanical, thermomechanical, and corrosion-resistant properties, and offer a massive, unexplored compositional space which allows for the targeted development of application-specific materials. Furthermore, although still in a nascent stage, research has shown that HEAs may exhibit unique irradiation tolerance, including reduced defect production and resistance to irradiation-induced swelling and hardening. The mechanisms behind this increased tolerance are not yet well-understood, although the HEA-specific attributes of a complex energy landscape, reduced thermal conductivity, and shift in defect migration energies and pathways provide promising explanations. Here this work assesses the current and future challenges faced by structural nuclear materials, identifying the specific applications in which HEAs may provide a competitive advantage compared to industry-standard materials with the aid of Ashby material selection maps. Considerations are provided for the design of future nuclear HEAs, including calculations of nuclear-relevant properties to assist in the initial down-selection of elements depending on application requirements (e.g., low neutron capture for in-core applications), narrowing the existing compositional space of HEAs to a manageable scope.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Cluster dynamics modeling of Mn-Ni-Si precipitates coupled with radiation-induced segregation in low-Cu reactor pressure vessel steels

Formation of precipitates enhanced or induced by irradiation causes hardening and embrittlement of nuclear structural materials. Post-irradiation microstructure characterization of reactor pressure vessel (RPV) steels has shown that precipitation can be strongly associated with radiation-induced segregation (RIS) of solutes on dislocations. However, RIS and precipitation have not been coupled in previous precipitation modeling of RPV steels. Here in this study, a new hybrid and spatially-dependent precipitation model is developed that couples cluster dynamics with RIS, providing an unique way to account for concurrent evolution of heterogeneous cluster densities as well as solute and point defect concentration profiles. The model is applied to study the segregation of Mn, Ni, and Si on dislocations and heterogeneous nucleation of Mn-Ni-Si rich precipitates (MNSPs) in a low-Cu RPV steel. The result shows that the onset of MNSP nucleation on dislocations occurs at the fluence of 2 x 10 23 n.m -2 . The number density and mean radius can reach ~ 10 24 m -3 and 1-2 nm at high fluence, respectively. These observations suggest that the role of RIS and heterogeneous nucleation can be significant for RPV steels under a high fluence of irradiation. Results of simulations adopting various dose rates and dislocation densities show an increased fraction of MNSPs on dislocations at the condition of low dose rate, high dislocation density, and high fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Heavy ion irradiation effects on CrFeMnNi and AlCrFeMnNi high entropy alloys

Co-free but Al-included medium/high entropy alloys (M/HEAs) have gained increasing interests due to their lower cost and the potential to tune the multi-phase microstructure. The irradiation response of two Co-free HEAs, face-centered cubic (FCC) CrFeMnNi with limited Cr enriched α' phase and body-centered cubic (BCC) AlCrFeMnNi with B2s phase and nanoprecipitates were explored. Ion irradiations using 5 MeV Fe 2+ ions were performed at 500°C to a peak fluence of 50 and/or 100 displacements per atom (dpa). In dual-phase AlCrFeMnNi, there was no significant radiation induced segregation or chemical intermixing at the coherent matrix (FeCrMn-rich)/second phase (AlNi-rich) boundaries. In CrFeMnNi, limited voids were only detected at the peak damage location of ~ 50 dpa. On the other hand, voids were widely distributed in AlCrFeMnNi: under 50 and 100 dpa irradiation conditions, voids were found with larger dimension and denser distribution in the FeCrMn-rich matrix, smaller and slightly lower density in an AlNi-rich second phase. In addition, the diameter of the FeCMn-rich nanoprecipitates didn't reveal any tendency of dissolution or growth. This is correlated with their superior structural stability against irradiation. Significant radiation-induced hardening (increases from 3.8 ± 0.2 GPa to 4.7 ± 0.6 GPa) was measured in CrFeMnNi, but only ~ 4% hardness increase (from 7.4 ± 0.8 GPa to 7.7 ± 0.4 GPa) was noted in AlCrFeMnNi. Finally, in addition to the radiation-induced defects, such as voids, dislocation loops and point defects, other factors, such as chemical short-range ordering may play an important role.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Insights into radiation resistance of titanium alloys from displacement cascade simulations

Radiation damage in beam window materials limits the use of high power proton beams in high energy physics research. The alloy Ti-6Al-4V is presently used as a beam window material but a prospective alternative, Ti-15V-3Cr-3Sn-3Al has been proposed. Since both these alloys contain dual phases at room temperature, we compare the radiation damage in the α and ß-phases of these two materials via primary knock-on atom (PKA) cascade simulations in the 10-40keV energy range. At PKA energies 30 and 40keV, the number of Frenkel pairs in the ballistic stage is higher in the ß-phase of Ti-6Al-4V than that in the ß-phase of Ti-15V-3Cr-3Sn-3Al almost by a factor of 2. The α-phase, of both these alloys, by far outperforms the ß-phases of the two alloys, both in terms of damage during the ballistic stage and in terms of the surviving defects. The average displacement threshold energy (E d ) in the α-phase of both alloys was found to be 66eV while that in the ß-phase of Ti-15-3 was 55 eV and in the ß-phase of Ti-6-4 was 46. Further, while the number of surviving defects is almost equal in both alloys, the vacancy and interstitial clustering mechanisms differ notably, which can impact the degree of radiation hardening and loss of ductility. Our simulations show larger vacancy and interstitial clusters form in Ti-6Al-4V as compared to that in Ti-15-3-3-3 alloy. These results indicate that Ti-15-3-3-3 alloys may be a promising candidate for next generation beam window material with a higher radiation tolerance than the existing Ti-6-4 alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Deformation localisation in ion-irradiated Fe and Fe10Cr

Determining the mechanisms for irradiation-induced ductility loss is crucial for the design of reactor structural components. Here, the deformation characteristics around nanoindents in Fe and Fe10Cr irradiated with Fe ions to ~1 displacement-per-atom at 313 K are non-destructively studied. Slip steps surrounding the nanoindents indicate that deformation is localised in the irradiated materials. Lattice rotation and strain fields near the indent site show over 87% confinement of plasticity in the irradiated material. Cr has little effect on the irradiation-induced changes in pile-up topography and deformation fields, suggesting it has limited impact on retaining strain hardening capacity and reducing irradiation-induced embrittlement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Correlating microstructure and mechanical properties of harvested high dose Zorita light water reactor internals

Here, in this study, microstructural studies and micro-mechanical testing of an ex-plant material harvested from the decommissioned pressurized water reactor (PWR) is carried out. Irradiated 304 stainless steel (SS) components were harvested by the Electric Power Research Institute, U.S. Nuclear Regulatory Commission, and members of the José Cabrera Nuclear Power Station (Zorita) in Spain. The bi-crystalline micro-tensile specimens, atom probe tomography (APT) tips, and transmission electron microscopy (TEM) lamellae were fabricated from the baffle plate materials irradiated to 0.05, 15 and 50 dpa to elucidate microstructural, microchemical, and local mechanical properties changes as a function of dose. TEM and APT studies reveal radiation induced Ni, Cr, and Si rich precipitates in the 15 dpa sample that become smaller and partially redissolve into the matrix of the 50 dpa sample (decreasing size and number density). Dislocation loop and cavity number density and size were quantified as a function of dose as well as swelling. Intergranular Ni and Si enrichments with concomitant Fe and Cr depletion were observed in the 15 dpa sample and were more pronounced in the 50 dpa sample. Micro-tensile testing performed in the scanning electron microscope (SEM) at room temperature and 300 °C shows that the material's local yield strength and ultimate tensile strength decreases with increased dose and elevated test temperature and provides further insights via localized strain mapping. Current and previous mechanical data was compared with calculated values using the dispersed barrier hardening model with inputs of dislocation loops, precipitates, and cavities from microstructural characterization.

36 MATERIALS SCIENCE↗
DOE OSTI2025

The interaction of edge dislocations with hydrogen-helium bubbles in tungsten

In fusion reactors, plasmas-facing components undergo degradation due to displacement damage and injection of gas impurities. Significant amounts of hydrogen (H) and helium (He) impurities can be introduced into materials through plasma exposure and nuclear transmutation, and their synergistic effects lead to the formation of mixed He/H bubbles under irradiation, changing their hardening mechanisms. In this study, molecular dynamics (MD) models were developed to study the interaction between ½<111> edge dislocations with mixed He/H-bubbles in W. The critical resolved shear stress (CRSS) that is required for dislocation breakaway increases with He/H-to-vacancy ratio at 600 K, indicating strong pinning effects of mixed He/H bubbles. However, as the temperature increases up to 1400 K, the mixed He/H bubbles become unstable as the H atoms are increasingly emitted from the mixed He/H bubbles and migrate into W matrix, which lowers bubble pressure and CRSS. Overall, these results highlight the synergistic effects of He and H on dislocation behavior in irradiated W during deformation.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Multi-faceted framework for extrapolating early age flexural strength to facilitate rapid lifting/handling of high-volume fly ash precast members

Maintaining adequate early-age structural performance for precast concrete components has grown in importance as more sustainable mix designs become more widespread. Achieving high-early flexural strength is particularly crucial to facilitate rapid removal of hardened concrete components from formwork, often within 24 h after fresh concrete placement. Limited research has assessed the effectiveness of traditional design methods in correlating flexural strength with compressive strength for next-generation mix designs, or demonstrated extrapolation of such material performance to larger-scale structural tests. This paper presents a multi-faceted framework to reassess early-age flexural strength for concretes made with relatively high proportions of fly ash from both fresh and harvested sources. Here, the framework provides several pathways, from which the user can select based upon available resources and the specific application, to improve accuracy of early-age cracking moment calculations. Furthermore, the scope includes evaluation of strength performance under curing conditions emulative of those in a precast facility, recommending modulus of rupture equations which are more performance-driven than current design provisions, and experimental tests on prefabricated concrete beams to validate the proposed methodologies. Correlations of early-age strength with both concrete age and maturity measurements compare the effectiveness of utilizing in-situ data to further enhance the prediction methods. Ultimately, the proposed framework helped reduce errors when calculating cracking moment capacity at early ages by tailoring calculations to reflect mix-dependent behavior. Furthermore, most estimates of cracking moment were within 25 % of their corresponding experimental test results, thus promoting confidence for using these strategies with high-volume fly ash precast structures.

42 ENGINEERING↗
DOE OSTI2020

Strain rate sensitivity, microstructure variations, and stress-assisted β → α" phase transformation investigation on the mechanical behavior of dual-phase titanium alloys

In this study, the microstructures of three widely used dual-phase titanium alloys: Ti-6Al-4V, Ti-6Al-2Sn-4Zr-2Mo, and Ti-6Al-2Sn-4Zr-6Mo are characterized using backscatter electron imaging and the deformation behaviors in their individual α and β phases are investigated using in situ high-energy X-ray diffraction during uniaxial tension experiments. This is accompanied by calculation of the strain rate sensitivity (SRS) in these alloys, as a function of applied loading. The results of the calculated SRS and the mechanical response (in terms of lattice strains and hardening behaviors) are rationalized based on the local microstructure, composition, and texture of the alloy. A stress-induced β (BCC) to α" (orthorhombic) phase transformation is observed during the in-situ diffraction experiment, which permitted analysis of the lattice strain evolution during the phase transformation. Results from this study provide insights into the relationship between microstructure, composition, applied loading, and the deformation behavior, including effects on loading with time holds. Specifically, the results show that the microstructure, and not necessarily the chemical composition (Mo content), along with the applied loading have a dominant role in the observed SRS variation in titanium alloys. Further, the evolution of the α" (orthorhombic) phase is observed for the first time during deformation of an industrially relevant titanium alloy, Ti-6246.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Unearthing principal strengthening factors tuning the additive manufactured 15-5 PH stainless steel

Additive manufactured (AM) materials exhibit enhanced performance, while post-processing is essential to reinforce its properties. Controlling the tunable factors and the underlying strengthening mechanisms during the post-processing are a key to answering the calls from the industry and end users. We identified the principal strengthening factors that contribute to AM stainless steel under heat-treatment and weighed their importance. The as-built sample exhibited a unique microstructure with a heterogeneous grain size distribution, a small amount of retained austenite, high dislocation density, and dispersed oxide particles. The strengthening primarily by dislocation forest and additionally by heterogeneous grain size contributed to the as-built sample. Although aging slightly decreased dislocation strengthening, superior aging enhancement was achieved by Cu-precipitates. The solid-solutionizing not only increased the effective grain size but also reduced the dislocation density by eliminating the heterogeneous microstructure, which deteriorated the precipitation hardenability. Our findings shed light on the performance optimization of AM metals.

36 MATERIALS SCIENCE↗
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 OSTI2023

Rapid grain refinement and compositional homogenization in a cast binary Cu50Ni alloy achieved by friction stir processing

Friction stir processing (FSP) has been increasingly adopted for joining and processing materials in automotive, aerospace and industrial construction. During FSP a dynamic competition between high-speed shear deformation and deformation-induced heating brings about a complex competition between multiple dynamic microstructural evolution mechanisms making it difficult to predict the microstructural evolution pathway. Hence, improved understanding of microstructural evolution mechanisms during FSP can be beneficial for continued growth in the adoption of FSP for demanding applications of future. Towards this goal, this study uses a model binary Cu – 50 at.% Ni alloy to clarify the effect of single and double pass FSP on the microstructural evolution of a coarse grained and compositionally heterogeneous cast microstructure. High energy synchrotron X-ray diffraction, electron backscatter diffraction, and nanoindentation are used to clarify the microstructural evolution due to FSP. The process of compositional homogenization of as-cast segregations is studied by energy dispersive spectroscopy and atom probe tomography. Our results show that a single fast FSP pass at 30 mm.s -1 produces a 100 µm deep layer of submicrometric and hall-petch hardened CuNi grains. The initial cast compositional heterogeneities in a micrometric scale is rapidly transformed to nano-sized domains, mainly confined at grain boundaries. Double pass FSP increases the penetration depth of the processed layer and leads to a 2.9 times grain growth relative to single pass FSP. Grain fragmentation, discontinuous dynamic recrystallization, grain growth, and twinning mechanisms are discussed. In conclusion, these results highlight the value of FSP for ultrafast grain refinement and compositional homogenization of cast alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Strengthening mechanisms, hardening/softening behavior, and microstructure evolution in an LPSO magnesium alloy at elevated temperatures

Here, rapid degradation of the mechanical properties of conventional Mg alloys with temperature precludes their wide application in industry. The LPSO (long period stacking ordered) Mg alloys, on the other hand, show considerable potential to achieve excellent high-temperature mechanical performance. In this study, a Mg-5.1Y-2.4Zn (wt.%) extrusion alloy with the microstructure comprising of α-Mg matrix and LPSO phase was prepared. It is shown that the yield strength of the alloy at 200 °C retained ~83% of the yield strength of 198 MPa at room temperature (RT). Importantly, the ductility also improved, owing to the increased non-basal slip activities at elevated temperatures. In contrast to RT and 200 °C, the alloy exhibited softening tensile behavior with an apparent loss of yield strength at 300 °C. In-situ lattice strain analysis revealed that load transfer from α-Mg to LPSO remains effective at temperatures up to 200 °C, governing the high yield strength of the alloy. The post-mortem analysis by means of electron microscopy confirmed that the LPSO phase is structurally stable at 200 °C. In particular, the high strain hardenability is rationalized by the sandwiched LPSO structures together with stacking faults serving as effective obstacles to the motion of non-basal dislocations. The load transfer effect disappears at the temperature of 300 °C, resulting in the fast deterioration of strength of the alloy. The cracking of the LPSO phase along with the recrystallizing of α-Mg grains were found to be the main causes for the softening behavior of the alloy at 300 °C.

36 MATERIALS SCIENCE↗