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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Cluster dynamics modeling of irradiation growth in single crystal Zr

Irradiation growth of single-crystal Zr is modeled with a reduced set of cluster dynamics equations. Nucleation and growth of basal and prismatic dislocation loops are both accounted for in the model. Equations are developed for the time evolution of single point defects, a limited number of point defect clusters, interstitial and vacancy loop nucleation rates, as well as for the growth of vacancy loops on basal planes and interstitial loops on prismatic planes. Reduction of the usual infinite hierarchy of cluster dynamics equations to the simple set studied here is justified on the physical basis of the stability of small loops once nucleated. This simplified cluster dynamics model with a small number of adjustable parameters avoids the complexity of explicit representation of higher order point defect clusters. The model shows consistency with experimental observations of the following aspects: (1) the growth rates of Zr crystals along the a- and c-axes; (2) the onset dose for breakaway irradiation growth; (3) the saturation dislocation loop densities of vacancy and interstitial loops; (4) the effects of cold work and temperature on irradiation growth.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Plasma and X-ray radiation-induced damage mechanisms in a tungsten heavy alloy

Tungsten-based materials are among the main candidate plasma-facing materials for future fusion reactors where extreme conditions of thermomechanical stresses, radiation, and energetic particles must be withstood. Here a hybrid X-pinch platform combined with a pulsed power machine was used to study damage mechanisms in 90W–4Cu–6Ni tungsten heavy alloy as a result of exposure to high energy density plasma, X-ray radiation, and the resulting thermal shock. It was found that the plasma-facing surface of the material undergoes severe melting and ablation with the extent of ablation being higher in the Cu–Ni phase than the W particles. Rapid melting and solidification transform the W particles into fine dendrite morphologies embedded in the Cu–Ni solid solution. Cracking was also observed in the subsurface layers at the interfaces between the W particles and the Cu-rich phase of the Cu–Ni matrix. X-ray radiation, on the other hand, cause slight melting on the surface without significant ablation and no cracking.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Vacancy and interstitial interactions with crystal/amorphous, metal/covalent interfaces

In this work, we use atomistic simulations to investigate the interaction of vacancies and interstitials with interfaces between a crystalline metal and an amorphous, covalently-bonded solid. We select the gold (Au)/silicon (Si) binary system as a model material and construct interface models along two different facets of crystalline Au and with amorphous Si (a-Si) created at three different quench rates. We compute formation energies of vacancies, self-interstitials, and interstitial impurities as a function of position relative to the interface and find that they have markedly lower values near the interface than in the interior of the adjoining phases. We conclude that crystal/amorphous, metal/covalent interfaces may be as effective at removing radiation-induced point defects as interfaces in polycrystalline metals composites. Moreover, irrespective of interface character, the average formation energies of all point defects at all the Au/a-Si interfaces we investigated are comparable. Thus, unlike in polycrystalline metals, where an interface’s crystallographic character has a marked effect on its interactions with point defects, all interface types in crystal/amorphous, metal/covalent composites may be equally effective at absorbing all radiation-induced defects.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Dislocation loop evolution in F/M steel T91 under in-situ ion irradiation: Influence of the presence of initial dislocations

Ferritic/Martensitic (F/M) steel T91 was irradiated in-situ to 4 and 10 dpa at 470 °C using 1 MeV Kr 2+ . Additionally, the microstructure evolution under irradiation was followed and characterized by in-situ Transmission Electron Microscopy (TEM), in terms of dislocation loop Burgers vector, average size and density as a function of dose. Furthermore, the influence of the presence of dislocations in the pre-irradiated matrix was investigated by comparing the radiation induced defect evolution in grains with and without initial dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Mechanical and chemical properties of PVD and cold spray Cr-coatings on Zircaloy-4

In this paper, accident-tolerant fuel (ATF) cladding materials aim to improve fuel reliability and safety during accident scenarios in water-cooled reactors. These ATF cladding should also perform comparable or better than the current zirconium alloy cladding under reactor normal operating conditions. Surface-modified Zircaloy-4 was produced by depositing a protective coating of chromium by two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coating thickness, structure, composition, and mechanical properties of the coated Zircaloy were assessed with advanced materials characterization techniques. Results showed a strengthening effect of the Cr-coating, along with ~83% higher hardness compared to that of the Zircaloy-4 substrate. Surface topography and surface chemistry of the Cr-coated Zircaloy-4 were assessed with contact profilometry, atomic force microscopy, contact angle goniometry, and X-ray photoelectron spectroscopy. Deposited Cr-coating significantly improved surface wettability compared to the substrate Zircaloy-4. The formation of surface oxides on the coatings by prolonged exposure to ambient air showed a further increase in the hydrophobic behavior. Surface parameters and surface chemistry after flow boiling critical heat flux tests, conducted at atmospheric pressure, evidenced a slight increase in surface roughness with notable changes in wettability associated with the formation of surface oxides.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterisation of nano-scale precipitates in BOR60 irradiated T91 steel using atom probe tomography

Atom probe tomography has characterised the microstructural changes in T91 steel after BOR60 reactor irradiation at five temperatures between 376 °C and 524 °C to doses between 14.6 dpa and 35.1 dpa. Irradiation-induced precipitation and segregation to carbide/matrix interface induced by neutron irradiation has been characterised. Atom probe tomography characterisation shows that Mn, Ni, Si -rich (MNS-rich) clusters form in T91 steel irradiated in BOR60 reactor at temperatures between 376 °C and 415 °C, which is not observed at higher temperatures 460 °C and 524 °C. Additionally, the number density, volume fraction and composition of MNS-rich clusters have been characterised. Ni, Mn, Si and P is found to segregate at carbide matrix interface after irradiation at lower temperature and only P segregation is observed at 524 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Electrochemical behavior and separation of iodide in molten LiF-NaF-KF melt

This work studied the electrochemical behavior of iodide in molten LiF-NaF-KF mixture and the separation of iodide from the molten salt at 973 K using a graphite working electrode. The cyclic voltammetry (CV) and chronoamperometry (CA) measurements indicated an anodic passivation layer was formed on the graphite electrode in this melt and the layer blocked the charge transfer across the electrolyte/electrode interface. However, the added I- ions could pass through the passivation layer and were oxidized at the graphite electrode. The characterization of the obtained products proved that iodide ions was efficiently oxidized to I2 gas during electrolysis. Furthermore, the graphite anode showed excellent stability without any obvious change in dimensions after electrolysis, which showed the high current efficiency of I2 evolution reaction. Thus, the graphite electrode was possible to be served as an excellent anode to separate other anions (such as Br) from molten fluoride salts.

36 MATERIALS SCIENCE↗
DOE OSTI2020

High spatial resolution thermal conductivity mapping of SiC/SiC composites

Herein, silicon carbide (SiC) ceramic matrix composites are being investigated as a new generation of fuel cladding materials due to its higher accident tolerance compared to Zircaloy. Characterization of the thermal conductivity of SiC constituents and their interfaces and interphases in SiC/SiC composites is needed as inputs to models of cladding performance. We used time-domain thermoreflectance (TDTR) to map the thermal conductivity with a spatial resolution of 2 µm. The SiC/SiC composite is comprised of Hi-Nicalon Type S fibers in a matrix made by chemical vapor infiltration (CVI matrix). The interphase material is a pyrolytic carbon/SiC multilayer. We report thermal conductivity maps of the SiC fiber and SiC matrix at temperatures of 25 °C, 90 °C, 164 °C, and 250 °C. The fiber has a uniform and isotropic thermal conductivity of 22 W m -1 K -1 at 25 °C; the thermal conductivity of fiber is approximately independent of temperature. The matrix thermal conductivity is not uniform and varies from 50 to 120 W m -1 K -1 at 25 °C; the thermal conductivity of the matrix scales approximately with the inverse of the square root of temperature, i.e., ∝ 1/T 1/2 . The thermal conductivity of the interphase at room temperature is 6 W m -1 K -1 . The calculated spatially averaged thermal conductivity of this SiC/SiC composite at 1000°C is 29 W m -1 K -1 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Thermal stability of oxidized ultrathin Li films on TZM for plasma facing components

For improved understanding of the behavior of Li plasma facing components (PFCs) in the presence of oxygen impurities, we report on the thermal stability of ultrathin (up to 1.0 nm) LiO x films on a titanium-zirconium-molybdenum (TZM) alloy substrate over the temperature range of 310–1400 K. LiO x films were prepared by either post-oxidation of Li films or Li films deposited on an oxygen-precovered TZM surface. Thermal stability of these films was studied using a combination of temperature programmed desorption (TPD), X-ray photoelectron spectroscopy (XPS), and low-energy ion scattering (LEIS). For post-oxidized Li films, TPD showed that no Li desorption occurred until temperatures above 620 K, and then Li desorbed from the surface via at least three desorption peaks. The O 1s XPS spectra on the post-oxidized films at 310 K indicate the formation of lithium oxide (Li2O) and peroxide/hydroxide (Li 2 O 2 /LiOH). The peroxide/hydroxide converts to oxide after heating to 680 K with no desorption of Li or O 2 , and then this film decomposes to liberate Li into the gas phase while leaving oxygen at the TZM surface. Heating the LiO x films to 1070 K led to a solid-state reaction with surface Mo atoms of the substrate to form a condensed binary lithium molybdenum oxide (LixMoOy) phase, which upon further heating decomposed to liberate gas phase binary lithium molybdenum oxide species. Li deposition on an oxygen-precovered TZM surface formed a complex, highly stable, oxygen rich, Li-O-Mo interfacial oxide. Li desorption from multilayer Li films on oxygen-precovered TZM surfaces occurred in a metallic Li multilayer peak and three other oxide-derived peaks. These results support the improved thermal stability of oxidized lithium as compared to metallic Li and have implications for Mo impurity transport into the plasma at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Atom probe tomography characterization of ion and neutron irradiated Alloy 800H

Alloy 800H is a high temperature and creep resistant alloy, and is considered as a candidate alloy for use in Generation IV nuclear reactor systems. To clarify the alloy's behavior under irradiation, samples were subjected to single ion irradiations up to 20 dpa at 440°C, a dual ion/He irradiation to 17 dpa at 460°C, and a neutron irradiation to 17 dpa at 385°C. The irradiated microstructures were characterized using atom probe tomography to complement previously published transmission electron microscopy studies. Additionally, after single ion irradiation, sparse fine Al and Ti clusters were observed after 1dpa, while high number densities of nanoscale Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops developed after 10 and 20 dpa. The microstructure formed during single ion irradiation exhibited a strong depth dependence. No significant differences were observed between the single and dual ion irradiations. While Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops were also observed after neutron irradiation, the neutron-irradiated microstructure differed from those found in the ion-irradiated samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Method for identification of redox control parameters for corrosion mitigation in molten fluoride salts

Structural material corrosion is a significant concern for the development of molten salt-cooled reactors (MSR) and redox control was proposed to mitigate the corrosion. This study presents an efficient approach to identify the partial pressure of F 2 gas for fluorine driven corrosion, the concentration ratios of UF 4 /UF 3 and ZrF 4 /ZrF 2 , and the partial pressure ratio of HF/H 2 needed to limit metal corrosion and its subsequent dissolution in molten fluoride salts. The identification procedure is made by simply using auxiliary axes in Ellingham diagrams. This approach greatly simplifies the optimization of redox control methods in studying the material corrosion in molten salts.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Void shrinkage in 21Cr32Ni austenitic model alloy during in-situ ion irradiation

Austenitic 21Cr32Ni model alloy thin foils, previously irradiated with 5 MeV Fe ++ ions in bulk to create voids, were re-irradiated in-situ in the Intermediate Voltage Electron Microscope Facility (IVEM). The voids which had been formed under bulk-ion irradiation shrank and disappeared after in-situ Kr ion irradiation in the temperature range 50 K-713 K to an additional dose of 1 dpa. The voids were unaffected by eithersuccessive thermal annealing to 673 K and by prolonged exposure to the 200 keV electron beam at the irradiation temperature. The high void shrinkage rate observed did not change significantly for irradiation temperatures between 50 K and 713 K, suggesting that the void shrinkage process in thin foils during in-situ heavy-ion irradiation results from the interactions of displacement cascades with the voids. Finally, possible void shrinkage mechanisms under thin foil irradiation are discussed in this study.

36 MATERIALS SCIENCE↗
DOE OSTI2020

High Flux Helium Irradiation of Dispersion-Strengthened Tungsten Alloys and Effects of Heavy Metal Impurity Layer Deposition

Tungsten has been chosen as the plasma-facing material (PFM) for the divertor region in ITER and also a candidate PFM for future plasma-burning nuclear fusion reactors. During fusion device operation, PFMs will be exposed to low-energy He irradiation at high temperatures, resulting in sub-surface bubbles and surface morphology changes such as pores and fuzz. Carbide dispersion-strengthened W materials may enhance the ductility of W, but their behavior under high flux He irradiation remains unclear. In this work, the response of dispersion-strengthened tungsten materials to high flux, low energy He irradiation at high temperature is examined. Tungsten alloyed with 1, 5, or 10 wt. % tantalum carbide or titanium carbide exposed to these conditions result in surface pores, coral-like feature growth and sub-surface helium bubbles. Reactor-relevant helium irradiation (5 x 10 26 m -2 fluence) combined with high powered laser pulses to simulate off-normal reactor events does not significantly alter the surface morphology, as the surface nanostructures appear stable and cracks are only observed on a localized region of one sample. However, specimens show the development of an impurity layer on the surface, likely impurity deposition from the sample holder during irradiation, resulting in a mixed material layer on the surface. Helium bubbles exist in this impurity layer, and obscure conclusions about helium interactions with the carbide dispersoids. Nonetheless, it is clear that the dispersoid microstructure limits He bubble formation and subsequent surface nanostructuring, attributed to the dispersoid composition.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Corrosion behavior of ferritic FeCrAl alloys in simulated BWR normal water chemistry

Ferritic iron-chromium-aluminum (FeCrAl) alloys with varying Cr content were subjected to in-situ irradiation-corrosion with energetic protons or electron beams in pure water with various oxygen concentrations. Oxide microstructure and composition were characterized to compare the corrosion under irradiation with the thermal corrosion in the same water conditions. The dominant effect seems to be the production of H 2 O 2 at the surface by the intense proton and electron beams. The corrosion potential of FeCrAl under BWR-NWC condition is somewhat above the Cr 2 O 3 /CrO 4 2- phase boundary, oxidizing the Cr(III) to Cr(VI) in the outer part of the film, but with some Cr(III) remaining in the interior of the oxide. The beam increases the corrosion potential, significantly reducing Cr(III) content throughout the oxide. Finally, the crystallites on the surface, which form by re-precipitation from solution, do not contain Cr because at low potential Cr(III) is not soluble, and at high potential Cr(VI) is quite soluble.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High-throughput ion irradiation of additively manufactured compositionally complex alloys

Several advanced nuclear reactor designs promise efficiency and safety improvements over the current reactor fleet but are limited by the current set of ASME code-qualified materials. Novel alloys including high-entropy alloys (HEAs), and more broadly compositionally complex alloys (CCAs), have shown promising irradiation-tolerance. However, the vast range of alloy compositions adds to an already time-consuming alloy development process. In this study, to accelerate the development of novel alloys for nuclear applications, a high-throughput (HTP) methodology has been employed. Additive manufacturing has been used to produce a compositional array of unary, binary, ternary, and quaternary alloys, including several CCAs, which span the Cr-Fe-Mn-Ni composition space. Additionally, the compositional array was homogenized at 1000°C for 24 hours and each sample was irradiated using 4-MeV Ni2+ ions at room temperature to a peak damage of 50 dpa, as estimated using SRIM, at the University of Wisconsin Ion Beam Laboratory. A custom XY stage was built to accommodate the large compositional array and half of each sample was masked during irradiation enabling both the irradiated and unirradiated properties of each alloy to be characterized side-by-side. Each alloy was characterized using X-ray fluorescence (XRF), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and nanoindentation. CALPHAD simulations spanning the entire Cr-Fe-Mn-Ni composition space at 1000°C were performed to compare predicted equilibrium phases with phases identified experimentally from the unirradiated regions of each alloy. Nanoindentation measurements indicate radiation-induced hardening ranging from ~1-1.5 GPa in each of Cr-Fe-Mn-Ni CCAs, which is relatively insensitive to modest changes in alloy composition and comparable to hardening observed in neutron irradiated Cr-Fe-Mn-Ni CCAs in the literature. Overall, a substantial time savings was realized by employing HTP synthesis, irradiation, and characterization in this study compared to conventional techniques, the implications of which are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Quantifying the effect of hydride microstructure on zirconium alloys embrittlement using image analysis

Since hydride microstructure can significantly influence Zr alloy nuclear fuel cladding’s ductility, a new metric has been developed to quantify hydride microstructure in 2D micrographs and relate it to crack propagation. As cladding failure usually results from a hoop stress, this new metric, called the Radial Hydride Continuous Path (RHCP), is based on quantifying the continuity of brittle hydride particles along the radial direction of the cladding tube. Compared to previous metrics, this approach more closely relates to the propensity of a crack to propagate radially through the cladding tube thickness. The RHCP takes into account hydride length, orientation, and connectivity to choose the optimal path for crack propagation through the cladding thickness. The RHCP can therefore be more closely linked to hydride embrittlement of the Zr alloy material, thus creating a relationship between material structure, properties, and performance. The new definition, along with previously proposed metrics such as the Radial Hydride Fraction (RHF), the Hydride Continuity Coefficient (HCC), and the Radial Hydride Continuity Factor (RHCF), have been implemented and automated in MATLAB. These codes are made available with this publication as supplementary materials. These metrics were verified by comparing their predictions of hydride morphology against expected values in simple cases, and the implementation of the new metric was validated by comparing its predictions with manual measurements of hydride microstructure performed on ImageJ. The RHCP was also validated against experimental measurements of fracture behavior and it was shown to correlate with cladding failure better than previous metrics. The information provided by these metrics will help accurately assess cladding integrity during operation, transportation, and storage.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Molecular dynamics simulations of radiation cascade evolution near cellular dislocation structures in additively manufactured stainless steels

The dislocation dense cellular structures observed in additively manufactured (AM) stainless steels can allow these materials to have enhanced strength and ductility compared to conventionally manufactured materials. Effective design of radiation resistant materials often necessitates creating microstructures containing of a high density of dislocations that act as sinks for radiation induced defects. This work uses molecular dynamics simulations to study the impact of pre-existing defects on radiation damage in stainless steel 316L fabricated by the laser powder bed fusion process. The evolution of synthetically generated dislocations in a crystal system in response to multiple radiation collision cascades suggest that dislocation dense regions reduce the probability of surviving point defects forming clusters. This finding seems to result from the dislocation cores’ ability to absorb point defects, thereby limiting the formation of clusters during the recombination phase. However, after successive cascades, the ability of the dislocation entanglement to prevent defect formation is diminished and defect cluster formation trends to an equilibrium. Accordingly, it is hypothesized that in an experimental setting the dislocation cells will initially act as neutral sinks for point defects, potentially delaying the onset of radiation damage effects; but, these structures will be degraded due to radiation-enhanced diffusion and lose their effectiveness at doses above 1.5–2 displacements per atom. Nevertheless, the ability to produce complex and feature-specific microstructures suggests promise for the design of radiation tolerant materials using AM methods.

36 MATERIALS SCIENCE↗
DOE OSTI2021

The Role of Co-injected Helium on Swelling and Cavity Evolution at High Damage Levels in Ferritic-Martensitic Steels

The influence of helium co-injection at rates from 0 to 4 appm He/dpa on swelling in ferritic-martensitic alloys T91 and HT9 was explored. Irradiations with 5.0 MeV Fe++ ions and degraded He++ ions were performed at 445°C up to damage levels of 150 dpa and helium co-injection rates of 0, 0.02, 0.2 and 4 appm He/dpa in T91, and at 460°C to a damage level of 188 dpa and helium co-injection rates of 0, 0.06 and 4 appm He/dpa in HT9. Helium was observed to enhance cavity nucleation at low damage levels, resulting in the maximum swelling at the highest helium co-injection rate. As the damage level was increased, the helium content at which swelling was maximized shifted to lower helium concentrations, ultimately resulting in the highest swelling occurring with zero helium by 150 dpa. Finally, this behavior was due to the helium-stabilized bubble microstructure that increased the cavity sink strength and reduced both cavity growth rate and swelling relative to the helium-free condition.

36 MATERIALS SCIENCE↗