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At least 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

The Water-Fe-Pressure dependent single-crystal elastic properties of wadsleyite: Implications for the seismic anisotropy in the upper Mantle Transition Zone

Recent seismic studies suggest an anisotropic Mantle Transition Zone (MTZ) in areas adjacent to subducted slabs. Wadsleyite is the main anisotropy contributor in the upper MTZ, therefore the interpretation of these seismic observations requires the knowledge of single-crystal elastic moduli (C ij s) and the deformation-induced lattice preferred orientation (LPO) of wadsleyite. Wadsleyite can host up to 3 wt% water in its crystal structure as point defects in the form of hydroxyl groups, however, the combined effect of water content, Fe content, and pressure on the C ij s of wadsleyite remains unclear. In this study, we measured the high-pressure single-crystal elasticity of a synthetic hydrous Fe-bearing wadsleyite (0.14(5) wt% water, Fe#=9.4, Fe 3+ /ΣFe=0.3) up to 18.2 (2) GPa. In combination with previous experimental data, we separated the effects of pressure, water, and Fe contents on the Cijs and intrinsic elastic anisotropy of wadsleyite. Overall, our results suggest that the intrinsic elastic anisotropy of wadsleyite decreases with pressure, water, and Fe contents. At 15 GPa, increasing the water content by 0.1 wt% or Fe# by 1 decreases the V P and V S anisotropy of wadsleyite by ~1.1-1.3%, and ~0.8-1.3% in average, respectively. Combined with the LPO determined in previous deformation experiments, we modeled the seismic anisotropy in the upper MTZ generated by a sub-vertical mantle flow near cold subducted slabs and a sub-horizontal mantle flow in the ambient mantle. In both scenarios, the LPO of wadsleyite leads to V SV (vertically polarized shear wave velocity) > V SH (horizontally polarized shear wave velocity). Our results suggest that wadsleyite may account for a weak anisotropic MTZ (<1%) on the global scale. Considering the fact that water decreases the elastic anisotropy but promotes LPO of wadsleyite, seismic anisotropy may not be a good water sensor.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Mechanism-based modeling of thermal and irradiation creep behavior: An application to ferritic/martensitic HT9 steel

In this work, the creep behavior of HT9 steel in both thermal and irradiation environments is predicted using an integrated modeling framework. Multiple physical mechanisms such as diffusional creep and dislocation climb are incorporated into crystal plasticity calculations using the Visco-Plastic Self-Consistent (VPSC) approach. Climb velocities are informed by mean field rate theory laws in place of empirical power law formulations. More interestingly, the climb velocities explicitly consider the contribution of irradiation-induced point defects, i.e., stress induced preferential absorption (SIPA) effect. The developed expressions are shown to apply under conventional thermal creep and to the more complex irradiation conditions as well. This physically-informed, mechanism-based model is used to simulate the creep strain evolution of HT9 pressurized tubes under various loading conditions. Furthermore, it is demonstrated that the experimental behavior of this material reported in the literature is well described by this theoretical framework. The role of each relevant mechanism is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Local structure study of Fe dopants in Ni-deficit Ni 3 Al alloys

We found that the local electronic and magnetic structure, hyperfine interactions, and phase composition of polycrystalline Ni–deficient Ni 3-x FexAl (x = 0.18 and 0.36) were investigated by means of 57 Fe Mössbauer spectroscopy. The samples were characterized by X–ray diffraction and magnetization measurements. The ab initio calculations performed with the projector augmented wave method and the calculations of the energies of iron point defects were done to elucidate the electronic structure and site preference of Fe doped Ni 3 Al. Moreover, the value of calculated electric field gradient tensor V zz =1.6 10 21 Vm -2 matches well with the results of Mössbauer spectroscopy and indicates that the Fe atoms occupy Ni sites.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

Lattice response to ambient self-irradiation of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures

Self-irradiation in plutonium is a continuous process that introduces radiation damage ranging from point defects to nanoscale radiation damage as well as decay daughter product impurities that require to be accommodated by the host material. These changes are affecting the structure and properties of the material by lattice swelling, mechanical strengthening, and other alterations. X-ray diffraction and dilatometry studies have shown that the room temperature self-irradiation of δ-phase Pu-Ga alloys (2 and 7 at%Ga) results in swelling of the lattice that saturates after 0.1–0.2 displacement per atom (dpa) of accumulated lattice damage. Here, for a better understanding of these self- irradiation processes, defect accumulation and damage evolution, this time-of-flight neutron diffraction study investigates the impact of ambient self-irradiation effects on the structural evolution of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures (<120 K), associated with a varied history of storage time and thermal treatments. Two key behaviors were observed in the lattice evolution after exposure to cryogenic temperatures with respect to room-temperature accumulated self-irradiation damage. Samples with no self- irradiation damage accumulation under ambient conditions showed swelling of the lattice at the end of the measurement, while samples that accumulated significant self-irradiation damage under ambient conditions showed only a partial recovery of the lattice parameter value upon completion of the measurements (cooling followed by heating back to room temperature). Possible causes of the behavior are proposed, and their associated effects are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of substrate orientation on CdS homoepitaxy by molecular dynamics

CdS homoepitaxy growth was performed by molecular dynamics using different substrate orientations and structures in order to analyze the CdS crystallinity. As anticipated from thermodynamics of homoepitaxy, highly crystalline films with only point defects were obtained on substrates with rectangular surface geometries, including View the MathML source[112¯] zinc blende (ZB), [101¯0] wurtzite (WZ), [112¯0] WZ, [110][110] ZB, [010][010] ZB, and View the MathML source[1101110] ZB. In contrast, films grown on substrates with hexagonal surface geometries, corresponding to the [0001][0001] WZ and [111][111] ZB growth directions, showed structures with a large number of defects including; anti-sites, vacancies, stacking faults, twinning, and polytypism. WZ and ZB transitions and grain boundaries are identified using a lattice identification algorithm and represented graphically in a structural map. A dislocation analysis was performed to detect, identify, and quantify linear defects within the atomistic data. Systematic simulations using different temperatures, deposition rates, and substrate polarities were perform to analyze the trends of dislocation densities on [0001][0001] WZ direction and showed persistent polytypism. As a result, the polytypism observed in the films grown on the substrates with hexagonal surface geometry is attributed to the similar formation energies of the WZ and ZB phases.

36 MATERIALS SCIENCE↗
DOE OSTI2015

In-situ luminescence monitoring of ion-induced damage evolution in SiO 2 and Al 2 O 3

Real-time, in-situ ionoluminescence measurements provide information of evolution of emission bands with ion fluence, and thereby establish a correlation between point defect kinetics and phase stability. Using fast light ions (2 MeV H and 3.5 He MeV) and medium mass-high energy ions (8 MeV O, E=0.5 MeV/amu), scintillation materials of a-SiO 2 , crystalline quartz, and Al 2 O 3 are comparatively investigated at room temperature with the aim of obtaining a further insight on the structural defects induced by ion irradiation and understand the role of electronic energy loss on the damage processes. For more energetic heavy ions, the electronic energy deposition pattern offers higher rates of excitation deeper into the material and allows to evaluate the competing mechanisms between the radiative and non-radiative de-excitation processes. Irradiations with 8 MeV O ions have been selected corresponding to the electronic stopping regime, where the electronic stopping power is dominant, and above the critical amorphization threshold for quartz. Lastly, the usefulness of IBIL and its specific capabilities as a sensitive tool to investigate the material characterization and evaluation of radiation effects are demonstrated.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermal conductivity suppression in uranium-doped thorium dioxide due to phonon-spin interactions

In this work, impact of low level of uranium (U) atom substitution on thermal conductivity of thorium dioxide (ThO 2 ) is investigated. ThO 2 is an electronic insulator with a wide optical band-gap and no unpaired electrons whose thermal transport is governed by phonons. U-substitution introduces unpaired f-electrons resulting in paramagnetic behavior of U-ThO 2 at room temperature, which significantly suppresses its thermal conductivity. A single crystal of U-ThO 2 with graded composition of U is grown using a hydrothermal synthesis method, and thermal conductivity measurements are performed in regions with uniform composition of U at levels of 0%, 6%, 9% and 16%. Measured thermal conductivity profiles over 77–300 K temperature range are analyzed using an analytical expression for phonon-mediated thermal transport based on Klemens-Callaway model. Temperature dependent thermal conductivity is found to deviate significantly from the Rayleigh scattering trend expected for a simple substitutional point defect with a small perturbation to mass and interatomic forces. With the resonant scattering term, observed large suppression of thermal conductivity at low temperatures can be closely reproduced. Additionally, the extracted phonon-spin coupling constants imply a nonlinear relation of phonon-spin interaction intensity with respect to U doping percentage. Our study reveals how phonon-spin scattering contributed by unpaired f-electrons in U atoms influences thermal transport in the U-ThO 2 system.

36 MATERIALS SCIENCE↗
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↗