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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Soliton motion induced along ferromagnetic skyrmion chains in chiral thin nanotracks

Using atomistic magnetic simulations we investigate the soliton motion along a pinned skyrmion chain containing an interstitial skyrmion. We find that the soliton can exhibit stable motion along the chain without a skyrmion Hall effect for an extended range of drives. Under a constant drive the solitons have a constant velocity. We also measure the skyrmion velocity–current curves and identify the signatures of different phases including a pinned phase, stable soliton motion, and quasi-free motion at higher drives where all of the skyrmions depin from the pinning centers and move along the rigid wall. In the quasi-free motion regime, the velocity is oscillatory due to the motion of the skyrmions over the pinning sites. For increasing pinning strength, the onset of soliton motion shifts to higher values of current density. We also find that for stronger pinning, the characteristic velocity–current shape is affected by the annihilation of single or multiple skyrmions in the drive interval over which the soliton motion occurs. Our results indicate that stable skyrmion soliton motion is possible and that the solitons could be used as information carriers instead of the skyrmions themselves for technological applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

What causes the variation in superconducting properties of UTe 2 ?

Reaching a consensus on the superconducting order parameter of unconventional superconductors remains a central challenge in the field of magnetically-mediated superconductivity. Though UTe 2 is largely accepted as a rare example of an odd-parity superconductor, its precise order parameter remains highly debated, even at ambient conditions. A key underlying issue is the large sample-to-sample variation in superconducting properties at zero applied pressure and magnetic field. Here, we investigate the origin of the observed variation by means of single crystal x-ray diffraction (SC-XRD) and scanning transmission electron microscopy (STEM) measurements. Our results reveal highly ordered crystalline lattices, in agreement with the expected Immm structure, and no signs of uranium vacancies. Tiny amounts of interstitial defects, however, are observed on the Te2 layers that host Te chains along the b axis. We argue that these defects give rise to slightly enhanced atomic displacement parameters observed in SC-XRD data and are enough to disrupt the unconventional superconducting state in UTe 2 . Our findings highlight the need to focus future order parameter determination efforts on single crystals of UTe 2 with minimal amounts of structural disorder.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2017

Scale transition using dislocation dynamics and the nudged elastic band method

Microstructural features such as precipitates or irradiation-induced defects impede dislocation motion and directly influence macroscopic mechanical properties such as yield point and ductility. In dislocation-defect interactions both atomic scale and long range elastic interactions are involved. Thermally assisted dislocation bypass of obstacles occurs when thermal fluctuations and driving stresses contribute sufficient energy to overcome the energy barrier. The Nudged Elastic Band (NEB) method is typically used in the context of atomistic simulations to quantify the activation barriers for a given reaction. In this work, the NEB method is generalized to coarse-grain continuum representations of evolving microstructure states beyond the discrete particle descriptions of first principles and atomistics. The method we employed enables the calculation of activation energies for a View the MathML source glide dislocation bypassing a [001] self-interstitial atom loop of size in the range of 4-10 nm with a spacing larger than 150nm in α-iron for a range of applied stresses and interaction geometries. This study is complemented by a comparison between atomistic and continuum based prediction of barriers.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Validating continuum theory for Cottrell atmosphere solute drag by molecular dynamics simulations

When a dislocation moves through a field of mobile solute atoms, solutes segregate to the dislocation and form Cottrell atmospheres which exert solute drag forces. Over the last 70 years, continuum theory has been used extensively to estimate these drag forces and their dependence on the dislocation velocity, however few prior works have validated the accuracy of continuum theories. Here, in this work, molecular dynamics (MD) simulations of dislocation motion in face-centered cubic Ni containing interstitial H atoms were performed in order to test the accuracy of continuum theory predictions. Our results demonstrate that continuum theory provides an accurate estimate of the solute drag force for velocities below the critical velocity at which the solute drag force is maximized. Above the critical velocity, continuum theory systematically deviates from MD. Additional analysis reveals that this deviation results from the multi-valued and unstable nature of solute drag under load control, and also from the transition to random solute drag which occurs at high velocities.

Dislocations↗
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

MeV per Nucleon Ion Irradiation of Nuclear Materials with High Energy Synchrotron X-ray Characterization

The combination of MeV/Nucleon ion irradiation (e.g. 133 MeV Xe) and high energy synchrotron x-ray characterization (e.g. at the Argonne Advanced Photon Source, APS) provides a powerful characterization method to understand radiation effects and to rapidly screen materials for the nuclear reactor environment. Ions in this energy range penetrate ~10 μm into materials. Over this range, the physical interactions vary (electronic stopping, nuclear stopping and added interstitials). Spatially specific x-ray (and TEM and nanoindentation) analysis allow individual quantification of these various effects. Hard x-rays provide the penetration depth needed to analyze even nuclear fuels. Here, this combination of synchrotron x-ray and MeV/Nucleon ion irradiation is demonstrated on U-Mo fuels. A preliminary look at HT-9 steels is also presented. We suggest that a hard x-ray facility with in situ MeV/nucleon irradiation capability would substantially accelerate the rate of discovery for extreme materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Dimensional isotropy of 6H and 3C SiC under neutron irradiation

This investigation experimentally determines the as-irradiated crystal axes dimensional change of the common polytypes of SiC considered for nuclear application. Single crystal α-SiC (6H), β-SiC (3C), CVD β-SiC, and single crystal Si have been neutron irradiated near 60 °C from 2 × 10 23 to 2 × 10 26 n/m 2 (E > 0.1 MeV), or about 0.02–20 dpa, in order to study the effect of irradiation on bulk swelling and strain along independent crystalline axes. Single crystal, powder diffractometry and density measurement have been carried out. For all neutron doses where the samples remained crystalline all SiC materials demonstrated equivalent swelling behavior. Moreover the 6H–SiC expanded isotropically. The magnitude of the swelling followed a ~0.77 power law against dose consistent with a microstructure evolution driven by single interstitial (carbon) mobility. Extraordinarily large ~7.8% volume expansion in SiC was observed prior to amorphization. Above ~0.9 × 10 25 n/m 2 (E > 0.1 MeV) all SiC materials became amorphous with an identical swelling: a 11.7% volume expansion, lowering the density to 2.84 g/cm 3 . As a result, the as-amorphized density was the same at the 2 × 10 25 and 2 × 10 26 n/m 2 (E > 0.1 MeV) dose levels.

36 MATERIALS SCIENCE↗
DOE OSTI2016

In situ transmission electron microscopy He + implantation and thermal aging of nanocrystalline iron

Due to their high density of interfaces, nanostructured material are hypothesized to show a higher tolerance to radiation damage compared to conventional coarse-grained materials and are on interest for use in future nuclear reactors. In order to investigate the roles of vacancies, self-interstitials, and helium during defect accumulation, and the thermal evolution of such defects, a complex set of in situ TEM experiments were performed in nanocrystalline iron.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Revisiting the diffusion mechanism of helium in UO 2 : A DFT+ U study

The understanding of migration properties of helium atoms after their generation through α-decay of actinides in spent nuclear fuels is important for the safety of nuclear fuel storage and disposal. The diffusion of helium in UO 2 is revisited by using the DFT+U simulation methodology employing the “U-ramping” method to address the issue of metastable energy states. A novel diffusion mechanism by helium interstitials, the “asymmetric hop” mechanism, is reported and compared to other diffusion mechanisms including an oxygen vacancy mediated mechanism and available experimental diffusion data. We show that the new mechanism is the dominant one over a wide temperature range.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Irradiation-induced amorphization in the zirconium suboxide on Zr-0.5Nb alloys

For the first time, the observation of irradiation-induced amorphization of the zirconium suboxide formed during aqueous corrosion of Zr-0.5Nb alloys is discussed. High-resolution transmission electron microscopy results reveal amorphization of the hexagonal-ZrO suboxide under heavy ion irradiation at cryogenic temperatures. This irradiation-induced amorphization behaviour is discussed in relation to the arrangement of oxygen interstitials and the formation of stable superlattices. The sensitivity of the suboxide to irradiation damage can lead to phase changes and the accumulation of defects near the oxide/metal interface, which needs to be considered in the development of mechanistic models addressing radiation-assisted acceleration of corrosion rates in zirconium alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Mechanisms of radiation-induced segregation around He bubbles in a Fe-Cr-Ni crystal

We report that Fe-15Cr-15Ni (wt.%) austenitic stainless steel was irradiated with 200 keV 4 He + ions to a fluence of 5 × 10 16 /cm 2 at 500 °C. Radiation-induced segregation (RIS) at He bubbles in the irradiated steel was measured using Energy Dispersive X-ray Spectroscopy (EDS) coupled with Scanning Transmission Electron Microscopy (STEM). The RIS measurements exhibited an enrichment of Ni and a depletion of Cr and Fe at the bubbles. To determine the underlying mechanisms of the RIS, the prediction of the Marwick model was compared with the RIS measurements. The model with parametric input in accordance with the Inverse Kirkendall effect (IKE), i.e. a preferential flux of elements associated with vacancy flux, was inadequate in explaining the observed amount of segregation of Fe, Cr and Ni. More specifically, a considerably lower ratio of the concentration gradient between Cr and Fe, $\left(\frac{∇Cr}{∇Fe}\right)$, of 0.53 was obtained from the RIS measurements versus the model-predicted value of 2.3, indicating an overestimation of Cr depletion at the bubbles by the Marwick model. To explain this difference between the model prediction and the RIS measurements, in addition to the Cr depletion at the bubbles by IKE due to the fastest diffusion via vacancy flux, a preferential interstitial flux of Cr from the matrix to the bubbles is discussed as a possible concurrently operating mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration

Fast reactor fuel cladding candidate materials require proficiency in extreme environments consisting of high temperatures and irradiation doses in excess of 150 displacements per atom (dpa). Nanostructured oxide dispersion strengthened (ODS) alloys have been developed extensively for this purpose due to their notable high temperature strength, creep resistance, and irradiation resistance. However, their properties can deteriorate if interstitial impurities such as C and N are not well controlled during the fabrication process. A new Fe-12Cr nanostructured ODS alloy OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding) with solute additions of Mo, Ti, and Nb has been developed to provide the desired properties mentioned above while simultaneously sequestering impurities within the matrix. After extrusion at 850°C, the as-extruded microstructure consists of an average 490 nm grain size and a high number density (6.8 × 10 23 m -3 ) of 2.2 nm diameter (Y,Ti,O) nanoclusters distributed homogeneously in the matrix. Atom probe tomography investigations suggest non-stochiometric compositions for the smallest nanoclusters. In addition, a second population of nanometer scale (Nb,Ti) rich carbonitrides is also present in the microstructure that captures the potentially detrimental C and N impurity atoms present in the matrix. Atom probe tomography results indicate elemental segregation of Cr, Mo, and Nb to grain boundaries in the as-extruded material, consistent with previous investigations of solid solution strengthening by solute additions. In conclusion, the ability of OFRAC to sequester impurities introduced from the powder metallurgical approach to nanostructured ferritic alloy development, compounded with its beneficial mechanical properties, makes this alloy a competitive candidate for fast reactor applications.

36 MATERIALS SCIENCE↗
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 OSTI2019

Cluster dynamics simulation of uranium self-diffusion during irradiation in UO 2

While fission fragments pass through UO 2 nuclear fuel, a considerable concentration of Frenkel pair defects (i.e. vacancies and interstitials) are created. The steady-state concentration of these defects leads to enhanced uranium self-diffusion, one of several fundamental kinetic parameters that control key engineering properties such as creep and fission gas swelling in UO 2 nuclear fuel. A cluster dynamics method to track point defects and defect clusters has been implemented in the MARMOT phase-field code in order to predict as-measured out-of-pile and irradiation enhanced thermal diffusivity. Here, the calculated uranium self-diffusion coefficient compares well with non-irradiated fuel measurements, and shows similar trends to those observed in irradiated fuel, which is a good result given the complexities introduced by non-stoichiometric compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Grand potential sintering simulations of doped UO 2 accident-tolerant fuel concepts

Chromium-doped UO is a widely-studied near-term deployable accident-tolerant fuel concept because it results in a dense, large-grain structure that increases the fuel resistance to densification, swelling, and fission gas release. A new charged-interstitial mechanism was recently proposed to describe the behavior of dopants like chromium in sintered UO 2 . Based on that mechanism, manganese was suggested as an even stronger dopant than chromium. In the current work we use mesoscale sintering simulations in an effort to validate the new mechanism. We compare the relative behavior of Cr-doped and undoped UO 2 against experimental data in the literature. We also make predictions of the relative behavior of Mn-doped UO 2 . Simulations are done using the phase field-based grand potential sintering model. Dopants have two effects on sintered UO 2 . They marginally increase the densification rate and greatly increase the average grain size. Both of these effects are individually tested in small-scale simulations. Then large 3D sintering simulations are used to test the combined behavior of both effects. The results for the densification rate simulations are consistent with experiments for Cr-doped fuels. However, we note the grain growth rates are lower than what is found in the literature for Cr-doped fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Irradiation effects on Al 0.3 CoCrFeNi and CoCrMnFeNi high-entropy alloys, and 316H stainless steel at 500 °C

To evaluate the potential of high entropy alloys (HEAs) for nuclear applications, two HEAs, Al 0.3 CoCrFeNi and CoCrFeMnNi, and a conventional Type 316H stainless steel (SS) were irradiated with 1 MeV krypton ions at 500 degrees C up to 1 dpa, and examined in-situ with a transmission electron microscope (TEM). After irradiation, a high density of ordered L1 2 nanoparticles was observed in Al 0.3 CoCrFeNi. In contrast, no phase transformation was observed in CoCrFeMnNi and 316H SS. In the thin foil regions of TEM samples, stacking-fault tetrahedra were observed in the HEAs. In the thick foil regions, interstitial dislocation loops were observed for all alloys, and the areal loop density increased linearly with foil thickness. The Al 0.3 CoCrFeNi had the largest loop size (the lowest density), followed by the CoCrFeMnNi and then the 316H SS. The higher loop nucleation rate in the 316H SS was attributed to carbon content. Finally, the degree of irradiation hardening was slightly lower for the HEAs than for the 316H SS, which is a promising sign for the nuclear application of HEAs at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Determination of neutron irradiation temperatures of SiC using electrical resistivity method

The irradiation temperature of silicon carbide (SiC) was determined post-irradiation by examination of the recovery of the electrical resistivity due to thermal annealing in a rapid heating/cooling optical furnace. High-purity, high-resistivity grade SiC is routinely used as a passive temperature monitor in neutron irradiation studies at the High Flux Isotope Reactor (HFIR), and this paper presents an alternative automated technique for determination of the irradiation temperature the SiC experienced. Neutron collisions with the atoms results in displaced lattice atoms (interstitials) that act as electron donors yielding a significant decrease in electrical resistivity. The irradiation defects become thermodynamically unstable and start to recombine, when annealed above the irradiation temperature, resulting in a recovery of the electrical resistivity. The resistivity is measured at a fixed elevated temperature above ambient, which is below the target irradiation temperature. When the resistivity is plotted as a function of annealing temperatures, a clear increase is observed due to the recovery of irradiation defects. We have demonstrated that this electrical resistivity measurement of SiC is effective to determine irradiation temperature of SiC. In conclusion, energy levels of various defects in SiC were calculated from Arrhenius plots of electrical conductivity versus inverse temperature.

36 MATERIALS SCIENCE↗