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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Comparative characterization of mixed spectra and thermal neutron shielded irradiated tungsten

The effect of mixed spectra and thermal neutron shielded irradiation on tungsten was evaluated with plasma exposure in the tritium plasma experiment followed by thermal desorption spectroscopy, X-ray photoelectron spectroscopy, and transmission electron microscopy. The two different irradiation campaigns were performed at the High Flux Isotope Reactor to 0.39–0.74 displacement per atom (dpa) in the 894–1379 K temperature range. A neutron spectrum influence on the void size and void number density was not observed. However, a strong correlation was found between void size and void number density with temperature, but not with dpa in the limited dpa range of this study. Thermal neutron shielding significantly reduced the transmutation to Re+Os. Higher irradiation temperature will lead to larger voids with lower number density, which reduces deuterium retention. In conclusion, grain growth was also observed for high-temperature irradiation of over ~1300 K within the limited grains visible in the transmission electron microscopy specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Radiation damage study of POCO ZXF-5Q graphite for neutrino production targets using 4.5 MeV helium ions

To address the challenges of increased beam power and target survivability associated with next-generation particle production beam lines, high dose, high-energy proton beam conditions are simulated using irradiation from low-energy ion beams. A low-energy ion irradiation study of POCO ZXF-5Q graphite under conditions similar to those of the NuMI NT-02 neutrino production target at the Fermi National Accelerator Laboratory is reported. Helium ion irradiation was performed at 100 ∘ C to a maximum damage level of 0.9 displacements per atom (DPA). Irradiation induced hardening, swelling of the irradiated region, inter-plane lattice expansion, and intraplane lattice contraction with increasing ion fluence was observed using micromechanical (nanoindentation, atomic force microscopy) and electron microscopy (high-resolution imaging, selected area diffraction) characterization. Similar changes were also observed in post irradiation examination of the NT-02 target indicating that ion irradiation can be a valuable tool for estimating radiation damage in proton beam targets. Caution must be exercised though, because the hardening, lattice alteration, and swelling occur to different magnitudes for a given damage level. The observed hardening and embrittlement were greater for ion irradiated graphite. For He ion irradiated samples the lattice spacing changes were smaller at low damage levels (78% less expansion and 71% less contraction at 0.1 DPA) and larger at high damage levels (38% more expansion and 5% more contraction at 0.9 DPA) relative to that observed in the NT-02 target. The magnitude of swelling was 8.5× greater under ion irradiation which is influenced by the differing damage gradients and inclusion of implanted He ions in the region of interest.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Fracture behavior of irradiation induced nanocrystalline UO 2 studied by in-situ mechanical testing in transmission electron microscopy

Uranium Dioxide (UO 2 ) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO 2 fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO 2 and UO 2 irradiated with 84 MeV Xe 26+ ions to a peak dose of 1357 displacements per atom (dpa). Further, the irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO 2 high burnup structure on local fracture properties. Fresh UO 2 fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO 2 structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO 2 as a result of Xe-irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of post-irradiation annealing on mechanical performance of irradiated 718 alloy

Here, the effect of post-irradiation annealing on solution-annealed 718 alloy was investigated using advanced mechanical testing, fractography, scanning electron microscopy, and transmission electron microscopy. Specimens were extracted from a proton beam window operated at the Spallation Neutron Source, irradiated with 940 MeV protons to a maximum dose of approximately 9.7 displacements per atom (dpa) at a calculated temperature not exceeding 110 °C while in service. Helium and hydrogen concentrations reached about 1700 and 6900 atomic parts per million (appm), respectively. Despite irradiation and high tensile strength (yield stress over 1 GPa), the material exhibited significant ductility. Annealing at 500 °C, 700 °C, and 900 °C for 30 min resulted in an appreciable decrease in yield strength and an increase in ductility for annealing treatments at 500 °C and 900 °C relative to the strength and ductility of the as-irradiated material. The presence of helium and hydrogen led to cavity formation and cleavage-like brittle features on fractured surfaces; however, high-magnification imaging revealed the presence of small-scale ductile dimples, indicating that the fracture mechanism remained mixed. The annealed specimens retained total elongation levels of 14–33 %, and there was no sudden drop in ductility after heat treatments. The ductility level in the irradiated and annealed material is notable despite the presence of helium and hydrogen.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Microstructural evolution and hardness changes in ion irradiated nickel-based Haynes 282 superalloy

This study investigated the crucial aspects of thermal and irradiation stability in precipitation hardened Haynes 282 Ni-based superalloy. The Haynes 282 Ni-based alloy was irradiated by 8 MeV Ni 3+ ions to assess its resistance to phase instabilities and mechanical property alterations. The mid-range doses (at a depth of ∼1 µm) were 1 and 10 displacements per atom (dpa) at temperatures of 600°C and 750°C. Nanoindentation tests provided insights into bulk equivalent hardness of the irradiated and pristine regions, while scanning transmission electron microscopy (STEM) and energy dispersive X-ray spectroscopy (EDS) were used to examine the microstructural evolution of irradiation-induced defects and γ′ -Ni 3 (Al, Ti) precipitates and defect structure under irradiation. These precipitates, with an average diameter of 29 nm and a number density of 5 x 10 21 /m 3 , acted as robust dispersion strengthening agents with high radiation point defect sink strength. Remarkably, irradiation did not significantly alter the size or number density of the γ′ precipitates, indicating exceptional thermal stability and radiation resistance of these precipitates at the examined conditions. Radiation-induced dislocation loops were observed at 600°C, albeit without a substantial impact on mechanical properties due to the dominance of γ′ precipitates on the overall alloy strength. The superior stability of γ′ precipitates observed in this study contrasts with several previous research findings on Ni-based alloys that reported poor precipitate stability. Plausible reasons for this difference are discussed. Moreover, this work explicitly outlines a physically grounded approach to ensure accurate microstructure-hardness correlations and clarifies the hardening model by addressing common misapplications of superposition in prior studies.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanical properties, strain hardening, and fracture behavior of ultrasonic additively manufactured Zircaloy-4 after low-temperature neutron irradiation

Ultrasonic additive manufacturing (UAM) is a solid-state, layer-by-layer advanced manufacturing process that has the potential to create custom spatially controlled composites with embedded wires and sensors for nuclear component manufacture. For this work, to assess the feasibility of using UAM for nuclear-relevant materials research, the technique was used to produce a 3.5-mm-thick Zircaloy-4 plate for irradiation testing. The UAM Zircaloy-4 specimens were irradiated in the High Flux Isotope Reactor at a target irradiation temperature of 117 °C to 2.9 displacements per atom (dpa) to assess differences in irradiation-hardening behavior as a function of alloy processing path. The UAM and reference baseplate (BP) materials increased in yield strength by 372±27 MPa and 346±21 MPa, respectively, and both suffered significant reductions in uniform and total elongation attributed to irradiation hardening at low-temperature. Although the materials had similar nanoscale defect structures, including nanoscale black dot/loop features and strain-induced dislocation channels, the UAM material’s processing-related defects resulted in accelerated strain localization and failure as demonstrated by lower post-irradiation uniform elongation of UAM specimens (0.5 %) compared to BP (1.5 %) material. The UAM material also showed considerable anisotropy in mechanical response due to crack propagation along weld boundaries, resulting in differences in strength & ductility when tested parallel and perpendicular to the prior UAM build orientation. Therefore, although the fundamental irradiation response of UAM-processed Zircaloy-4 was phenomenologically comparable to that of BP reference material, additional optimization of the UAM processing is needed to produce irradiation-resistant and nuclear-relevant materials.

Digital image correlation↗
DOE OSTI2026

In situ study on radiation response of tungsten manufactured by laser powder bed fusion

Tungsten (W) produced by laser powder bed fusion (LPBF) was examined by in situ Krypton (Kr) ion irradiation at 400 °C up to 2.52 displacements per atom (dpa) to investigate its radiation response. Dislocation loops with identical Burgers vectors form aligned raft structures, inducing significant grain misorientation accumulation. Defect saturation was observed beyond 0.36 dpa, marked by constant loop density and raft spacing. WO 3 nanoparticles are found in the as-printed matrix and served as efficient defect sinks. Dislocation loops were absorbed at the W/WO 3 interface, facilitating defect annihilation and suppressing defect accumulation. In conclusion, these findings highlight the role of LPBF microstructure and oxide interfaces in mediating radiation-induced defect evolution, offering insights for designing radiation tolerant W-based materials.

Defect sink↗
DOE OSTI2021

Comparison of void swelling of ferritic-martensitic and ferritic HT9 alloys after high-dose self-ion irradiation

In this work, ferritic-martensitic (F/M) HT9 along with fully ferritic HT9 specimens were irradiated using self-ions to 600 peak displacement-per-atom (dpa) at 450 °C. To investigate and minimize the carbon contamination effect on void swelling, two identical F/M HT9 specimens were prepared: one without a coating and the other one with a tungsten and silver coating on the surface prior to irradiation. The specimen without a coating showed carbide formation in the matrix, resulting from carbon contamination during irradiation, with no voids observed. The coated sample did not form matrix carbides, and voids were observed within the irradiated region. The same coating was applied to the ferritic HT9, and void swelling results of both the F/M HT9 and the ferritic HT9 alloys were compared after irradiation. The F/M HT9 exhibited a maximum of 2.6% swelling, while the ferritic HT9 showed a maximum of 0.8% swelling at a 400–500 nm depth, equivalent to 334 local dpa. The results were correlated with the instability of carbides in the ferritic HT9 under irradiation. Carbides in the F/M HT9 and the ferritic HT9 are both M23C6, but the matrix carbides in the ferritic HT9 were less stable under the irradiation, compared to the ones on the grain boundaries in the F/M HT9. The study shows the impact of chemistry on the void swelling which overpowers the microstructural factors such as grain size and phase.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The role of Cr concentration and temperature on cavity swelling with co-injected helium in dual-ion irradiated Fe and Fe-Cr alloys

The level of chromium plays an essential role in irradiation tolerance of Fe-Cr ferritic alloys. However, conflicting results have been reported regarding the dependence of cavity swelling under irradiation on Cr level and temperature. Here, we have performed a comprehensive set of simultaneous dual-ion (Ni + He) irradiations to high dose (~30 displacements per atom, dpa) at 400–550 °C on a series of ultra-high purity Fe and Fe-Cr binary alloys (3–14 wt.%Cr). Helium co-implantation rates of 0.1 and 10 appm He/dpa were selected to examine He synergistic effects relevant for fission and fusion reactor conditions, respectively. Cavities were observed in all irradiated samples by transmission electron microscopy. The results show that higher He implantation rate causes a shift in the swelling peak to higher temperatures in both Fe and Fe-Cr alloys. When assuming smaller cavities as biased sinks, the non-monotonic nature of the cavity swelling behavior is related to the ratio of biased to unbiased point defect sink strengths. Cr-enriched precipitates were observed in Fe-14Cr irradiated at 400 °C by atom probe tomography. Our analysis suggests the formation of Cr-enriched precipitates could suppress cavity swelling for Fe-Cr alloys with Cr content above 10 wt%.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Comparing structure-property evolution for PM-HIP and forged alloy 625 irradiated with neutrons to 1 dpa

The nuclear power industry has growing interest in qualifying powder metallurgy with hot isostatic pressing (PM-HIP) to replace traditional alloy fabrication methods for reactor structural components. But there is little known about the response of PM-HIP alloys to reactor conditions. This study directly compares the response of PM-HIP to forged Ni-base Alloy 625 under neutron irradiation doses ~0.5–1 displacements per atom (dpa) at temperatures ranging ~321–385 °C. Post-irradiation examination involves microstructure characterization, ASTM E8 uniaxial tensile testing, and fractography. Up through 1 dpa, PM-HIP Alloy 625 appears more resistant to irradiation-induced cavity nucleation than its forged counterpart, and consequently experiences significantly less hardening. This observed difference in performance can be explained by the higher initial dislocation density of the forged material, which represents an interstitial-biased sink that leaves a vacancy supersaturation to nucleate cavities. These findings show promise for qualification of PM-HIP Alloy 625 for nuclear applications, although higher dose studies are needed to assess the steady-state irradiated microstructure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

From suppressed void growth to significant void swelling in NiCoFeCr complex concentrated solid-solution alloy

Void swelling can result in dimensional instability and undermine the safe operation of nuclear reactors. Current strategies to inhibit void swelling mainly focus on enhancing defect absorption and recombination by introducing high-density defect sinks. Complex concentrated solid-solution alloys (CSAs), including high-entropy alloys, can withstand severe radiation damage due to their inherent chemical complexity without interfaces. However, the underlying mechanisms for void suppression in CSAs are far from clear. In this research, we studied the void evolution with respect to irradiation depths, doses, and temperatures in equiatomic NiCoFeCr under 3 MeV Ni ion irradiations. At relatively low doses (16 and 54 displacements per atom, dpa), voids form mainly outside of the ion-damaged region, and void formation in the peak damage region is suppressed, leading to negligible swelling. However, with further increase of dose (86 up to 250 dpa), significant void growth occurs in the peak damage region and extended dislocation lines dominate instead of short dislocation lines and loops formed at lower doses. From 500 to 700 °C, the dislocation density decreases while dislocations grow. Although the overall void swelling increases dramatically from 500 to 580 °C at 54 dpa, void growth in the peak damage region is still suppressed. The transition from suppressed void growth to significant void swelling is attributed to dislocation evolution and local chemical inhomogeneity (enrichment of Fe/Cr in the matrix) at higher doses. Our study shows that controlling element diffusion and defect evolution through tuning chemical complexity can further enhance the swelling resistance of CSAs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructural stability of tantalum-alloyed ferritic-martensitic steel with neutron irradiation to 7.4 dpa at ~490 °C

Specimens of a reduced-activation ferritic-martensitic steel, called CNA1, were irradiated to 7.4 displacements per atom at ~490 °C in the High Flux Isotope Reactor. Vickers hardness and tensile properties revealed consistent softening that is attributed to some recovery during irradiation, with reduced laths and dislocation density. Detailed microstructural characterization by transmission electron microscopy and energy dispersive x-ray spectroscopy showed radiation-induced dislocation loops, few cavities, and slight dissolution of M 23 C 6 and MX precipitates. Partial amorphization was only observed in M 23 C 6 precipitated in matrix. Transmutation somewhat destabilized M 23 C 6 and MX, with a stronger effect on the later one by noticeable depletion of Ta with enriched W, which is consistent with the transmutation calculations. The microstructural evolution and associated strength changes were discussed with the assistance of computational thermodynamics.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation-induced amorphization of Fe-Y-based second phase particles in accident-tolerant FeCrAl alloys

Here, second phase intermetallic particles in an advanced accident-tolerant FeCrAl (Fe-13Cr-5Al-2Mo) alloy are formed in the α-Fe matrix during processing. These particles are prominently related to the added Y. Neutron irradiation to ~7 displacements per atom (dpa) with a dose rate of ~8.16 × 10 -7 dpa/s at 282 °C resulted in the amorphization of these precipitates which could degrade the mechanical properties of the FeCrAl alloys. Analytical electron microscopy and diffraction analysis combined with structural freedom analysis have been used to investigate the radiation resistance of the second phase particles. Radiation tolerance is closely linked to the particle Fe-Y content and can be tailored using the structure freedom value.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Multi–length scale characterization of point defects in thermally oxidized, proton irradiated iron oxides

A key for the success of safe nuclear power generation system is to consider structural materials that are economical, meet mechanical property needs, possess good corrosion resistance, and are radiation tolerant. Nevertheless, fundamental insights that elucidate the details of radiation damage on materials corrosion performance are lacking. This includes the behavior of surface oxides which often regulate corrosion. For example, it is unclear how non-equilibrium point defects, oxide structure, mass transport in oxides, and subsequent oxidation behavior are altered by the radiation. Here, in this work, some of the effects of proton irradiation on the attributes of point defects, iron oxide microstructures, and the physical nature of the oxidation product were correlated with corrosion behavior. Iron oxides, fabricated by thermal oxidation in air at 400°C and 800°C for 1 h, were subjected to 200 keV, 0.03 dpa (displacements per atom) of proton irradiation, and subjected to corrosion reactivity assessment using AC and DC electrochemical methods. Experimental methods that target materials information at different length scales, such as positron annihilation spectroscopy (atomistic), transmission electron microscopy (mesoscopic), and electrochemical methods (macroscopic), were coupled to shed light on the impact of radiation-induced defect modifications and structural changes in oxides on corrosion reactivity which ultimately affects durability in harsh environments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of TiN coating on suppressing Ce-Fe interaction under irradiation

Advanced cladding is critical for fast reactors with the adequate thermal conductivity, mechanical stability and radiation tolerance of the cladding base material, corrosion resistance and high temperature coolant compatibility of the cladding surface, and chemical stability of the cladding inner wall against fuel cladding chemical interaction (FCCI). The preliminary results of recent ion irradiation studies of two diffusion-couple samples of cerium (Ce)/oxide-dispersion strengthened steel (ODS) and Ce/TiN/ODS, irradiated with 80 MeV xenon (Xe) ions to 100 displacements per atom (dpa) at 500°C, are summarized. Significant Ce-Fe interaction occurred in the Ce/ODS sample, and no noticeable Ce-Fe interaction was found in the Ce/TiN/ODS sample. It shows the effectiveness of 1-µm TiN diffusion barrier coated by the pulsed laser deposition on suppressing Ce-Fe interaction, a major contributor to FCCI in cladding. Here, density function theory (DFT) calculations of the impurity diffusivities of Ce and Fe within the Ti sublattice of TiN were performed to assist a mechanistic understanding of the experimental results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Damage buildup in multilayer hexagonal boron nitride films under Ar ion bombardment

Ion irradiation can be used to modify properties of two-dimensional hexagonal boron nitride (2D-hBN). However, the behavior of multilayer 2D-hBN under ion bombardment remains essentially unexplored. Here, in this study, we systematically study the damage buildup in 40-nm-thick multilayer 2D-hBN films irradiated at room temperature with 500 keV Ar ions with doses up to 0.64 displacements per atom. Transmission electron microscopy, soft X-ray absorption spectroscopy, and Raman spectroscopy are employed to study effects of ion irradiation. Results reveal a monotonic buildup of damage that can be described by the phenomenological zero cascade overlap model. Compared to the behavior of other group-III nitride semiconductors, radiation damage in hBN is influenced by polymorphism, involving transitions between bonding configurations of hexagonal, cubic, and amorphous phases.

Seo, M. [Lawrence Livermore National Laboratory (L↗
DOE OSTI2023

Nanoprecipitates to Enhance Radiation Tolerance in High-Entropy Alloys

The growth of advanced energy technologies for power generation is enabled by the design, development, and integration of structural materials that can withstand extreme environments, such as high temperatures, radiation damage, and corrosion. High-entropy alloys (HEAs) are a class of structural materials in which suitable chemical elements in four or more numbers are mixed to typically produce single-phase concentrated solid solution alloys (CSAs). Many of these alloys exhibit good radiation tolerance like limited void swelling and hardening up to relatively medium radiation doses (tens of displacements per atom (dpa)); however, at higher radiation damage levels (>50 dpa), some HEAs suffer from considerable void swelling limiting their near-term acceptance for advanced nuclear reactor concepts. In this study, we developed a HEA containing a high density of Cu-rich nanoprecipitates distributed in the HEA matrix. The Cu-added HEA, NiCoFeCrCu 0.12 , shows excellent void swelling resistance and negligible radiation-induced hardening upon irradiation up to high radiation doses (i.e., higher than 100 dpa). The void swelling resistance of the alloy is measured to be significantly better than NiCoFeCr CSA and austenitic stainless steels. Density functional theory simulations predict lower vacancy and interstitial formation energies at the coherent interfaces between Cu-rich nanoprecipitates and the HEA matrix. The alloy maintained a high sink strength achieved via nanoprecipitates and the coherent interface with the matrix at a high radiation dose (~50 dpa). From our experiments and simulations, the effective recombination of radiation-produced vacancies and interstitials at the coherent interfaces of the nanoprecipitates is suggested to be the critical mechanism responsible for the radiation tolerance of the alloy. Furthermore, the materials design strategy based on incorporating a high density of interfaces can be applied to high-entropy alloy systems to improve their radiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Chemical Design of Pb-Free Relaxors for Giant Capacitive Energy Storage

Dielectric capacitors have captured substantial attention for advanced electrical and electronic systems. Developing dielectrics with high energy density and high storage efficiency is challenging owing to the high compositional diversity and the lack of general guidelines. Herein, we propose a map that captures the structural distortion ($\delta$) and tolerance factor (t) of perovskites to design Pb-free relaxors with extremely high capacitive energy storage. Our map shows how to select ferroelectric with large $\delta$ and paraelectric components to form relaxors with a t value close to 1 and thus obtaining eliminated hysteresis and large polarization under a high electric breakdown. Taking the Bi 0.5 Na 0.5 TiO 3 -based solid solution as an example, we demonstrate that composition-driven predominant order-disorder characteristic of local atomic polar displacements endows the relaxor with a slushlike structure and strong local polar fluctuations at several nanoscale. This leads to a giant recoverable energy density of 13.6 J cm -3 , along with an ultrahigh efficiency of 94%, which is far beyond the current performance boundary reported in Pb-free bulk ceramics. In conclusion, our work provides a solution through rational chemical design for obtaining Pb-free relaxors with outstanding energy-storage properties.

25 ENERGY STORAGE↗