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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 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Magnetic phase formation in irradiated austenitic alloys

Austenitic alloys are often observed to develop magnetic properties during irradiation, possibly associated with radiation-induced acceleration of the ferrite phase. Some of the parametric sensitivities of this phenomenon have been addressed using a series of alloys irradiated in the BOR-60 reactor at 593K. The rate of development of magnetic phase appears to be sensitive to alloy composition. To the first order, the largest sensitivities to accelerate ferrite formation, as explored in this experiment, are associated with silicon, carbon and manganese and chromium. Si, C, and Mn are thought to influence diffusion rates of point defects while Cr plays a prominent role in defining the chromium equivalent and therefore the amount of ferrite at equilibrium. Pre-irradiation cold working was found to accelerate ferrite formation, but it can play many roles including an effect on diffusion, but on the basis of these results the dominant role or roles of cold-work cannot be identified. Based on the data available, ferrite formation is most probably associated with diffusion.

Gussev, Maxim N↗
DOE OSTI2015

Multiscale simulation of xenon diffusion and grain boundary segregation in UO₂

In light water reactor fuel, gaseous fission products segregate to grain boundaries, resulting in the nucleation and growth of large intergranular fission gas bubbles. The segregation rate is controlled by diffusion of fission gas atoms through the grains and interaction with the boundaries. Based on the mechanisms established from earlier density functional theory (DFT) and empirical potential calculations, diffusion models for xenon (Xe), uranium (U) vacancies and U interstitials in UO₂ have been derived for both intrinsic (no irradiation) and irradiation conditions. Segregation of Xe to grain boundaries is described by combining the bulk diffusion model with a model for the interaction between Xe atoms and three different grain boundaries in UO₂ (Σ5 tilt, Σ5 twist and a high angle random boundary), as derived from atomistic calculations. The present model does not attempt to capture nucleation or growth of fission gas bubbles at the grain boundaries. The point defect and Xe diffusion and segregation models are implemented in the MARMOT phase field code, which is used to calculate effective Xe and U diffusivities as well as to simulate Xe redistribution for a few simple microstructures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Dose dependence of helium bubble formation in nano-engineered SiC at 700 °C

Knowledge of radiation-induced helium bubble nucleation and growth in SiC is essential for applications in fusion and fission environments. Here we report the evolution of microstructure in nano-engineered (NE) 3C SiC, pre-implanted with helium, under heavy ion irradiation at 700 °C up to doses of 30 displacements per atom (dpa). Elastic recoil detection analysis confirms that the as-implanted helium depth profile does not change under irradiation to 30 dpa at 700 °C. While the helium bubble size distribution becomes narrower with increasing dose, the average size of bubbles remains unchanged and the density of bubbles increases somewhat with dose. These results are consistent with a long helium bubble incubation process under continued irradiation at 700 °C up to 30 dpa, similar to that reported under dual and triple beam irradiation at much higher temperatures. The formation of bubbles at this low temperature is enhanced by the nano-layered stacking fault structure in the NE SiC, which enhances point defect mobility parallel to the stacking faults. Here, this stacking fault structure is stable at 700 °C up to 30 dpa and suppresses the formation of dislocation loops normally observed under these irradiation conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Temperature dependence of underdense nanostructure formation in tungsten under helium irradiation

Recently, tungsten has been found to form a highly underdense nanostructured morphology (“W fuzz”) when bombarded by an intense flux of He ions, but only in the temperature window 900–2000 K. Furthermore, using object kinetic Monte Carlo simulations (pseudo-3D simulations) parameterized from first principles, we show that this temperature dependence can be understood based on He and point defect clustering, cluster growth, and detrapping reactions. At low temperatures (<900 K), fuzz does not grow because almost all He is trapped in very small He-vacancy clusters. At high temperatures (>2300 K), all He is detrapped from clusters, preventing the formation of the large clusters that lead to fuzz growth in the intermediate temperature range.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Influence of solidification structures on radiation-induced swelling in an additively-manufactured austenitic stainless steel

Metal additive manufacturing offers potential advantages for producing structural materials, such as austenitic stainless steels, in nuclear power systems. However, the microstructure developed during metal additive processing is notably different from the one developed in conventional processing, and the influence of the microstructural differences on performance in radiation environments has not been fully quantified. Using heavy ion irradiation and transmission electron microscopy, the radiation-induced swelling response of a laser powder-bed fusion-manufactured austenitic stainless steel was investigated at high doses. The influence of solidification-induced dislocation and precipitate structures was studied by comparing the radiation-induced swelling response of a 316L stainless steel in three microstructural states: as-fabricated, solution annealed, and fully recrystallized. Void swelling was approximately twice as pronounced in the as-fabricated state compared to post-processed states. In the framework of the rate theory for radiation effects, the higher swelling in the as-fabricated state can be explained by the strong sink bias for interstitial point defects exerted by the intermediate density of pre-existing dislocations. Void swelling was inhibited in the vicinity of pre-existing precipitates, but the density of precipitates in the as-fabricated material was not enough to compensate for the increase in swelling caused by dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗
DOE OSTI2021

A density functional theory and neutron diffraction study of the ambient condition properties of sub-stoichiometric yttrium hydride

Several mechanical and thermophysical properties are required as a function of non-stoichiometry for the successful implementation of YH 2-x for nuclear reactor moderator applications. Density functional theory calculations, in combination with neutron diffraction experiments, were used to study the structural and mechanical properties of YH 2-x . Point defect analysis indicated H occupation primarily at the tetrahedral site within an fcc Y sub-lattice, confirming a fluorite YH 2 structure. The small positive formation energy for H vacancies under Y-rich conditions predicted that hypo-stoichiometry is accommodated by Y+YH 2 at ambient conditions and by H vacancies in the YH 2-x single phase that is relevant to high temperatures. Neutron diffraction studies were used to confirm both the occupation of H on tetrahedral sites and the near-stoichiometric composition of the hydride phase in the two-phase Y+YH 2 region of the phase diagram that dominates at room temperature. Energy minimized special-quasirandom-structures of H vacancies were used to calculate lattice parameters, elastic constants, and several other properties as a function of composition for the single phase YH 2-x . The lattice parameter of YH 2-x decreased by only 0.004 Å with increasing H/Y for the range 1.31 ≤ H / Y ≤ 2.0 indicating a negligible effect on lattice parameters due to vacancy formation. In the two-phase region, however, calculations predicted the density of z Y + ( 1 - z ) Y H 2 - x to increase with decreasing H/Y at lower temperatures due to the increased fraction of high-density Y metal. Additionally, for the high temperature single phase, decreasing H/Y reduced the density as a consequence of the lattice expansion associated with vacancy formation. All elastic constants and moduli increased with increasing hydrogen content in single-phase YH 2-x .

36 MATERIALS SCIENCE↗
DOE OSTI2022

Understanding radiation effects in friction stir welded MA956 using ion irradiation and a rate theory model

We report an outstanding challenge in the manufacturing and joining of oxide dispersion strengthened steels is retaining the nanofeatures in the alloy throughout the fabrication and welding process. MA956 was friction stir welded with two different sets of welding parameters, resulting in a medium and high heat input. After welding, 5 MeV Fe ++ ion irradiations were performed at doses ranging from 50 to 200 dpa in the temperature range of 400 to 500°C. Post-irradiation characterization was performed with scanning transmission electron microscopy and energy-dispersive x-ray spectroscopy to investigate the Y-Al-O dispersoids, voids, and dislocations. After welding, the dispersoid microstructure coarsened, resulting in fewer and larger dispersoids regardless of heat input. After irradiation, the dispersoid behavior in the welded material was sensitive to temperature, exhibiting growth behavior attributed to Ostwald coarsening at 500 °C but a mixture of nucleation and more muted growth at 400 and 450 °C, attributed to competing mechanisms of radiation-enhanced diffusion and Ostwald coarsening. Void swelling correlated to heat input; being more prevalent in the welded conditions occurring at lower doses and in higher values relative to the base material. The low values of swelling despite microstructure coarsening caused by welding demonstrate the excellent swelling resistance of MA956, even after welding with the highest swelling values of 0.5% noted in the stir zone high heat input condition at 450 °C, 200 dpa. The dislocation behavior was inconsistent: the strongest trend was that network density was higher for welded versus base material, and an increase in loop diameter with temperature was observed. A rate theory model based on the observed microstructure suggests at high temperature interstitial loss to sinks was more likely to be dominant compared to mutual annihilation via point defect recombination, because of an increase of the radiation diffusion coefficient with temperature regardless of initial welded microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO) and uranium carbide mixture. The addition of UC helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. Additionally, the presence of UC or UC in the UCO fuel kernels is shown to have a substantial impact on the UO non-stoichiometry by inducing oxygen vacancies and driving UO sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Depth-resolved thermal conductivity and damage in swift heavy ion irradiated metal oxides

We investigated thermal transport in swift heavy ion (SHI) irradiated insulating single crystalline oxide materials: yttrium aluminum garnet- Y 3 Al 5 O 12 (YAG), sapphire (Al 2 O 3 ), zinc oxide (ZnO) and magnesium oxide (MgO) irradiated by 167 MeV Xe ions at 10 12 – 10 14 ions/cm 2 fluences. Depth profiling of the thermal transport on nano- and micro- meter scales was assessed by time-domain thermoreflectance (TDTR) and modulated thermoreflectance (MTR) methods, respectively. This combination allowed us to isolate the conductivities of different sub-surface damage-regions characterized by their distinct microstructure evolution regimes. Thermal conductivity degradation in SHI irradiated YAG and Al 2 O 3 is attributed to formation of ion tracks and subsequent amorphization, while in ZnO and MgO it is mostly due to point defects. Additionally, notably lower conductivity when probed by very low penetrating thermal waves is consistent with surface hillock formation. An analytical model based on Klemens-Callaway method for thermal conductivity coupled with a simplified microstructure evolution capturing saturation in defect concentration was used to obtain depth dependent damage across the ion impacted region. The studies showed that YAG has the highest damage profile resulting in the less dependence of thermal conductivity with the depth, while MgO on the contrary has the strongest dependence. The presented work sheds new light on how SHI induced defects affect thermal transport degradation and recovery of oxide ceramics as promising candidates for next generation nuclear reactor applications.

36 MATERIALS SCIENCE↗
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

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 OSTI2023

The effect of secondary phases on microstructure and irradiation damage in an as-built additively manufactured 316 L stainless steel with a hafnium compositional gradient

Additive manufacturing (AM) or rapid prototyping has become a crucial tool for reducing both cost and time while increasing efficiency in qualifying structural materials for reactor use. In this study, directed energy deposition (DED) was used to develop an as-built 316 L stainless steel sample with three regions of increasing Hf-dopant to study the effects of Hf on the irradiation response of the material. Morphological and microstructural changes were analyzed before and after 2 MeV proton irradiation at 360 °C to a damage of 2.5 dpa at ~ 10 µm below the surface. The addition of Hf effectively suppressed radiation-induced damage (dislocation loops, radiation-induced segregation) due to enhanced point defect recombination. The radiation damage seen in the as-built sample was further compared to a thermo-mechanically treated counterpart of the same fabrication and irradiation parameters which was found to behave superiorly. In conclusion, the increased radiation resistance of this material may be attributed to the as-built microstructure, which includes undissolved Hf particles, delta ferrite grains and cellular sub-grain boundaries that can hinder defect motion.

316L Stainless Steel↗
DOE OSTI2023

Improving charge transport in integrated MoO 3 /C electrode materials for water-in-salt energy storage systems by incorporating oxygen vacancies

Improvements in the charge storage properties of α-MoO 3 used as an electrode with a 30m ZnCl 2 water-in-salt electrolyte have been achieved by enhancements in electron and ion transport enabled by an inventive synthesis route. Electron transport was improved through the integration of MoO 3 with dopamine-derived carbon via a chemical preintercalation route, and enhanced ion transport was achieved by incorporating oxygen vacancies in MoO 3 structure through ethanol 2 reduction under hydrothermal conditions. Here, the presence of carbon was confirmed by corresponding D and G bands observed in Raman spectroscopy measurements. The presence of oxygen vacancies was proven through correlated XPS, TGA, Raman spectroscopy and XRD analyses, with the introduction of oxygen vacancies leading to an expanded interlayer region. Four-point probe measurements provided evidence of increased electronic conductivity due to the incorporation of carbon, and cyclic voltammetry-based charge storage mechanism analyses revealed increases in ion transport kinetics due to oxygen vacancy formation. Tuning the oxygen vacancy concentration is critical, as excessive concentrations of these point defects leads to structural instability and poor capacity retention. This work demonstrates the combined potential of carbon and oxygen vacancies in moderate concentrations to enhance the charge storage properties of transition metal oxides. The strategies developed in this study offer a path to the development of promising materials for high-rate, high-capacity, and long-duration electrochemical energy storage technologies.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2013

Correlation of Schottky constants with interatomic distances of selected I–VII and II–VI compounds

The observed linear (Na-, K-halides) and near-linear (Mg-, Sr-, Zn-, Cd-, and Hg-chalcogenides) dependences of Schottky constants on reciprocal interatomic distances yield the relation logK{sub S}=((s{sub s}1/T)+i{sub s})1/d{sub (A−B)}+(s{sub i}1/T)+i{sub i}, where K{sub S} is the product of metal and non-metal thermal equilibrium vacancy concentrations, and s{sub s}, i{sub s}, s{sub i} and i{sub i} are the group specific slope and intercept values obtained from an extended analysis of the above log K{sub S} versus 1/d{sub (A−B)} data. The previously reported linear dependences of log K{sub S} on the Born–Haber lattice energies [1] are the basis for combining the earlier results [1] with the Born–Mayer lattice energy equation to yield a new thermodynamic relationship, namely logK{sub S}=−(2.303nRT){sup −1}(c{sub (B−M)}/d{sub (A−B)}−I{sub e}), where c{sub (B−M)} is the product of the constants of the Born–Mayer equation and I{sub e} is the metal ionization energy of the above compounds. These results establish a correlation between point defect concentrations and basic thermodynamic, coulombic, and structural solid state properties for selected I–VII and II–VI semiconductor materials. - Graphical abstract: Display Omitted.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Investigation on hydrogenation performance of Mg{sub 2}Ni+10 wt.% NbN composite

The Mg{sub 2}Ni+10 wt.% NbN composite was prepared by mechanical milling and its hydrogen absorption/desorption properties and microstructure were systematically investigated. XRD results indicated that NbN was stable during ball milling process while partly decomposed into NbN{sub 0.95} and NbH during hydriding/dehydriding cycles irreversibly. The composite exhibited excellent hydrogenation/dehydrogenation kinetics performance with 2.71 wt.% hydrogen absorbed in 60 s at 423 K and 0.81 wt.% hydrogen released in 2 h at 523 K, respectively. The H diffusion constant of the composite reached 14.98×10{sup −5} s{sup −1} which was more than twice increased than that of pure Mg{sub 2}Ni powder. The superior hydrogen storage properties of the composite were ascribed to the refined grain size and abundant N-defect points provided by NbN and NbN{sub 0.95} in the composite. - Graphical abstract: The Mg{sub 2}Ni+10 wt.% NbN composite displays improvements on particle size distribution as well as hydrogen storage properties compared with that of pure Mg{sub 2}Ni. - Highlights: • NbN is introduced into Mg{sub 2}Ni hydride by Ar protected ball-milling. • Surfaces of the additive NbN particle are reduced by Mg{sub 2}NiH{sub 4}. • Hydrogenation kinetic property at 423 K is double improved. • Dehydrogenation capacity at 523 K of composites is beyond double improved.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗