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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Toward a high-fidelity tritium transport modeling for retention and permeation experiments

We report Tritium Migration Analysis Program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE). Created at Idaho National Laboratory (INL), MOOSE is an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. Using TMAP8, tritium transport in a divertor monoblock was analyzed to elucidate the effects of pulsed operation (up to fifty 1,600 s plasma discharge and cool-down cycles) on the tritium in-vessel inventory source term and ex-vessel release term (i.e., tritium retention and permeation) for safety analysis. With its built-in Message Passing Interface capability, TMAP8 can, in under 2 h, simulate tritium transport in three different layered materials (i.e., tungsten, copper, and copper-chromium-zirconium alloy) in 2D geometry, using a single device/computer with 10 cores. The MOOSE-based TMAP8 code can leverage other MOOSE tools developed under the Nuclear Energy Advanced Modeling and Simulation program to perform tritium and thermal transport in complex geometries and multiphysics environments. And via its massively parallel computation, MOOSE will enable the fusion pilot plant designers to conduct high-fidelity multiphysics modeling for the design of the divertor and blanket systems as well as for the safety analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Hydrothermal catalytic conversion of oleic acid to heptadecane over Ni/ZrO 2

There is growing interest in the substitution of expensive noble metal catalysts with low-cost earth-abundant metals in applications targeting biofuels production from waste organic feedstocks. Here, nickel (Ni) catalysts supported on zirconium dioxide (ZrO 2 ), both with and without copper (Cu) as a co-metal, were tested in hydrothermal reactions of unsaturated and saturated C18 fatty acids as models for waste oil feedstocks. In contrast to recent reports, this study showed no enhancement of nickel's activity for fatty acid conversion to alkane products when Cu was present. Ni/ZrO 2 prepared by metal coprecipitation methods converted 100% of oleic acid with 25% selectivity to heptadecane after 5h of reaction at 350 °C using methanol as a hydrogen donor source, increasing to 41% heptadecane after 20 h. Lower yields were observed with catalysts prepared by wet impregnation and using alternative hydrogen donor sources (glycerol, formic acid). Greater selectivity to heptadecane was also observed at higher temperatures (370 °C) and when the initial fatty acid had greater saturation. Longer term exposure to hydrothermal media led to metal sintering and catalyst deactivation. Findings support a path forward to the development of earth-abundant metal catalysts for the upgrading of waste organic feedstocks.

09 BIOMASS FUELS↗
DOE OSTI2025

Investigation of residual stress distribution in wire-arc directed energy deposited refractory molybdenum alloy utilizing numerical thermo-mechanical analysis and neutron diffraction method

Directed energy deposition (DED), a metal additive manufacturing (AM) technique, offers higher deposition rates and energy efficiency, making it suitable for fabricating components from refractory molybdenum alloys, such as molybdenum-titanium-zirconium (TZM). However, large thermal gradients and non-equilibrium thermal cycles in DED could generate high residual stress in the component, potentially deteriorating quality and performance. Thus, this study aims to investigate residual stress generation and its distribution in wire-arc DED of TZM thin-wall, utilizing thermo-mechanical analysis and high-fidelity neutron diffraction (ND) method. Two interpass temperatures (50°C and 200°C) have been considered to investigate their impact on residual stress formation. During experiments, in-situ thermal data has been recorded using thermocouples, which have been utilized for calibrating the thermal model. Thermocouple data shows a good agreement with the simulation results, having a difference of less than 10 %. Post-deposition part deformation has been observed, which is measured using a coordinate measuring machine, showing maximum values of 0.93 mm and 0.78 mm for interpass temperatures of 50°C and 200°C, respectively. Numerical predictions of distortion deviated by less than 15% from the experimental results. ND measurement and simulation results indicate that residual stress magnitude and evolution vary across the TZM deposits, revealing microstructural anisotropy in both conditions. Notably, lower interpass temperatures resulted in higher residual stresses, confirmed by experimental and simulation data. Further, this study demonstrated that an integrated experimental and thermo-mechanical analysis can potentially reveal the temperature history, part deformation, and residual stress formation in wire-arc DED TZM alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Unveiling the nucleation and growth of Zr oxide precipitates in internally oxidized Nb3Sn superconductors

We report on atomic-scale analyses of nucleation and growth of Zr oxide precipitates and the microstructural evolution of internally oxidized Nb3Sn wires for high-field superconducting magnet applications, utilizing atom probe tomography (APT), transmission electron microscopy (TEM), and first-principles calculations. APT analyses reveal that oxygen and zirconium are already segregated at grain boundaries (GBs) in the unreacted Nb-1Zr-4Ta (at%) alloy prior to forming Nb3Sn through reacting the Nb alloy with Sn and SnO2. After forming Nb3Sn, Zr oxide precipitates nucleate both at the Nb3Sn/Nb heterophase interfaces and in the Nb3Sn grains, driven by the small solubilities of Zr and O in Nb3Sn compared to their value in Nb. A high number density (Nv) of Zr oxide nanoprecipitates is observed in the Nb3Sn layers, ∼10 23 m −3 , with a mean diameter <10 nm for a heat treatment at 625 °C. Quantitative APT and TEM analyses of the Zr oxide precipitates in the reacted Nb3Sn layers elucidate details of the nucleation, growth, and coarsening processes of the Zr oxide precipitates in Nb3Sn. First-principles calculations and classical nucleation theory are employed to study the nucleation of Zr oxide precipitates in Nb3Sn and to estimate the maximum energy barrier and critical radius for nucleation. Our research unveils the kinetic pathways for nucleation and growth of Zr oxide precipitates and the microstructural evolution of Nb3Sn layers, which helps to understand and improve the superconducting properties of internally oxidized Nb3Sn wires for use in high-field superconducting magnets. •We investigated the nucleation and growth of Zr oxide nanoprecipitates in Nb3Sn superconductors for magnet applications.•APT and TEM analyses reveals the nucleation and growth process of Zr oxide precipitates in Nb3Sn layers.•Classical nucleation theory with DFT is employed to describe the nucleation process of Zr oxide precipitates in Nb3Sn.•We find that the interfacial reactions at Nb/Nb3Sn interface play a critical role in the nucleation of Zr oxide precipitates.•Current study provides a pathway to improve the properties of internally oxidized Nb3Sn superconducting wires.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Tetragonal structure and uniaxial magnetic anisotropy in the arc-melted (Ce,Zr) 2 (Fe,M) 17 (M = Mo, W, Co) alloys

The tetragonal ThMn 12 -type structure is stabilized in (Ce 1-x Zr x ) 2 T 16 M (x = 0.2 – 0.3; T = Fe or Fe/Co, M = Mo or W) arc-melted alloys. Approximately 5 at% of Mo or W admixture is sufficient to transform the hexagonal Th 2 Ni 17 -type structure of (Ce 1-x Zr x ) 2 Fe 17 into the tetragonal ThMn 12 -type structure of nearly single-phase (Ce 1-x Zr x ) 2 Fe 16 M and/or (Ce 1-x Zr x ) 2 Fe 15 CoM bulk alloys. X-ray Rietveld refinements reveal that Zr and Mo (W) substitute different sites in the tetragonal crystal structure. Zirconium preferentially replaces Ce, whereas Mo (W) substitutes Fe. At room temperature, the tetragonal phases exhibit strong ferromagnetism and a uniaxial magneto-crystalline anisotropy with anisotropy fields of 15 – 17 kOe. These materials possess room-temperature saturation magnetizations of 95 – 105 emu/g and Curie temperature of 420 – 505 K. At some concentrations, concurrent Zr and Mo (W) site occupancy facilitates 1:12 structure formation in bulk alloys with minimal presence of non-magnetic element in the Fe sublattice, thus securing the highest known magnetic moment per Fe atom (∼ 1.37 µ B ) in this type of materials. The intrinsic magnetic characteristics, as well as the absence of critical rare earths (RE), make these compounds interesting for development as low-cost permanent magnets. In conclusion, small Co additions improve the Curie temperature, especially if combined with W.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Onion-Like nanoparticles of the metal–organic framework UiO-66 synthesized by sequential spike crystal growth

Reproducible size and morphology of metal–organic framework (MOF) particles are essential for tuning these materials for applications like heterogeneous catalysis. Particle size measurements from electron micrographs provide information about the impact of changing synthetic parameters on MOF growth, but reported data are usually limited to averages and statistical distributions. Further elucidating synthetic control parameters could facilitate studies of facet-dependent and non-classical growth mechanisms. Here, in this work, phase-pure nanoparticles of the MOF UiO-66 were synthesized using alternating spikes of zirconium and hafnium-based nodes, forming an onion-like structure with alternating layers of Zr-UiO-66 and Hf-UiO-66. In conventional, bright-field transmission electron micrographs, Hf-UiO-66 and Zr-UiO-66 layers appeared darker and lighter, respectively. Alternating layers of Hf-UiO-66 and Zr-UiO-66 were also apparent in elemental maps obtained using scanning transmission electron microscopy with energy-dispersive X-ray spectroscopy. Average particle sizes increased monotonically with each spike, and no Zr-only or Hf-only particles were observed after the first spike, indicating that new material was consistently incorporated onto existing seeds. Surface area of the onion-like particles and their uptake of Ni 2+ from solution were both between the values measured for Zr-UiO-66 and Hf-UiO-66. Results demonstrate the use of sequential spikes of nodes in monitoring crystal growth in MOFs.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Isotropic nanocrystalline Sm(Fe,Co) 11.3 Ti 0.7 magnets modified with B and Zr

Rare-earth-lean Sm(Fe,Co,Ti) 12 alloys with the ThMn 12 crystal structure and less than one Ti atom per formula unit have the potential of exceptionally powerful permanent magnets, but all prior attempts to develop high coercivity in bulk alloys, especially coercivity combined with crystallographic texture, have fallen short of the expectations. This study was aimed at improvement of the currently best Sm(Fe,Co,Ti) 12 magnets prepared through melt-spinning which are inherently isotropic. Modifications of the alloys with B and Zr, already demonstrated in earlier studies to be effective separately, have been implemented simultaneously. Here, a systematic study of Sm 1.1-x (Fe,Co) 11.3-y Ti 0.7 B y alloys melt-spun at a tangential speed of 50 m/s and annealed at 600–950 °C allowed for monitoring the continuous evolution of the two consecutive crystal structures, those of the TbCu 7 and ThMn 12 types. Zirconium was found to facilitate the formation of the 1:12 structure at the expense of the 1:7, whereas boron has the opposite effect, at certain concentrations completely suppressing the 1:12. When the two alloying elements are introduced simultaneously, they inhibit growth of the 1:12 crystallites at annealing temperatures higher than 800 °C, thus allowing for the development of a higher coercivity. Because of instrumental limitations, bulk magnets were prepared through a two-step process – compaction of the melt-spun ribbons at 650 °C and additional treatment at a higher temperature – and they were characterized by a reduced, 90–93%, density. Nevertheless, an isotropic Sm 0.9 Zr 0.2 (Fe,Co) 10.8 Ti 0.7 B 0.5 magnet exhibited fair values of the remanence (7.4 kG), maximum energy product (8.5 MGOe) and coercivity (5.4 kOe), as well as high Curie temperature of 525 °C and remarkably small temperature coefficient of the coercivity, -0.25%/°C.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effect of alloying with Sc, Nb and Zr on reduction-diffusion synthesis of magnetically hard Sm(Fe,Co,Ti) 12 -based monocrystalline powders

Powders of Sm(Fe,Co) 11.2 Ti 0.8 alloys modified with Sc, Nb and Zr, as well as with additional Ti were prepared by reducing mechanically activated raw oxides with Ca metal in the furnace preheated to 990–1250 °C. Expansion of the crystal lattice upon introduction of Nb or additional Ti implies that atoms of these elements replace the smaller Fe atoms in the tetragonal ThMn 12 -type structure. On the other hand, contraction of the lattice upon introduction of Sc or Zr was smaller than what was expected for replacement of the Sm atoms, which suggests that the Sc and Zr atoms replace both the Sm and Fe atoms. Washing away the reduction byproducts expands the crystal lattice of the 1:12 particles and increases their coercivity. The lattice expansion associated with the washing is believed to be caused by interstitial H atoms; more research, however, is needed to establish the mechanism(s) of the washing effect on the coercivity. The earlier reported development of a high coercivity in zirconium-modified monocrystalline particles achieved by increasing the reduction annealing temperature to ≈1200 °C was similarly characteristic of the particles modified with Sc (the coercivity reaches 11.5 kOe) and Nb (8.1 kOe), but not for the particles prepared with additional Ti where the maximum coercivity of 8.3 kOe develops for a lower annealing temperature. Furthermore, it is concluded that Sc, Nb and Zr modify the high-temperature phase equilibria of the Sm(Fe,Co) 11.2 Ti 0.8 alloys allowing for an effective high-temperature processing, whereas the alloy coercivity increases with the synthesis temperature through a different, still unknown mechanism which may involve suppression of the defects specific to the 1:12 crystals.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A multiscale model to understand the interface chemistry, contacts, and dynamics during lithium stripping

A reversible Li-metal electrode, paired with a solid electrolyte, is critical for attaining higher energy density and safer batteries beyond the current lithium-ion cells. A stable stripping process may be even harder to attain as the stripping process will remove Li-atoms from the surface, and naturally reduce surface contact area, if not self-corrected by other mechanisms, such as diffusion and plastic deformation under an applied external stack pressure. Here, we capture these mechanisms occurring at multiple length- and time- scales, i.e., interface interactions, vacancy hopping, and plastic deformation, by integrating density functional theory (DFT) simulations, kinetic Monte Carlo (KMC), and continuum finite element method (FEM). By assuming the self-affine nature of multiscale contacts, we predict the steady-state contact area as a function of stripping current density, interface wettability, and stack pressure. We further estimate the exponential increase of overpotential due to contact area loss to maintain the same stripping current density. We demonstrate that a lithiophilic interface requires less stack pressure to reach the same steady-state contact area fraction than a lithiophobic interface. A “tolerable steady-state” contact area loss for maintaining stable stripping is estimated at 20 %, corresponding to a 10 % increase in overpotential. To constrain contact loss within the tolerance, the required stack pressure is 0.1, 0.5, and 2 times the yield strength of lithium metal for three distinct interfaces, lithiophilic Li/lithium oxide(Li2O), Li/lithium lanthanum zirconium oxide(LLZO), and lithiophoblic Li/lithium fluoride(LiF), respectively. The modeling results agree with experiments on the impact of the stack pressure quantitatively, while the discrepancy in stripping rate sensitivity is attributed to the simplifying interface interaction in our simulations. Overall, this multiscale simulation framework demonstrates the importance of electrochemical-mechanical coupling in understanding the dynamics of the Li/SE interface during stripping.

Feng, Min↗
DOE OSTI2022

Multimodal analysis and characterization of the boehmite layer formed on AA6061 before and after alkaline etching

Low-enriched uranium (LEU) alloyed with 10% Mo (U-10Mo) is being considered as a promising alternative to oxide-based dispersion fuel with high-enriched uranium for use in research reactors. The configuration of this proposed LEU monolithic LEU fuel plate consists of a U-10Mo plate-type fuel foil with a 25 µm zirconium interlayer barrier clad with an aluminum alloy (AA6061). In certain research reactors, the clad AA6061 is coated with a boehmite layer to prevent corrosion. The boehmite layer has a high-pH passivation range, which makes it resistant to oxidation. Boehmite is usually formed on the AA6061 surface by autoclave processing. Before the boehmite layer is added, the surface of the AA6061 is cleaned using techniques such as polishing and wet etching. In this study, we use multimodal analysis to examine how pretreatment of AA6061 using polishing followed by alkaline etching affects the chemical composition of the boehmite layer. X-ray photoelectron microscopy (XPS), transmission electron microscopy (TEM), and x-ray diffraction (XRD) were used to study the chemical changes in the boehmite layer caused by alkaline etching pretreatment. XPS provides quantitative analysis for the Al:O ratio as well as oxidation states present on the surface, which suggests slight oxidation of the boehmite surface after alkaline etching of the AA6061 surface. We further explored this suggested oxidation of the boehmite surface using high-resolution transmission electron microscopy with selected area electron diffraction (SAED) and grazing incidence x-ray diffraction (GI-XRD), which suggested only a small amount of aluminum oxide at the surface. The multimodal analysis and imaging yielded new insights for optimizing boehmite growth on AA6061 for research reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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

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 OSTI2021

Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide

To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.

Uranium Dioxide, Gadolinium Oxide, Thermal Air Oxi↗
DOE OSTI2021

Evaluation of the anisotropic grain boundaries and surfaces of α -U via molecular dynamics

Alloys based on uranium-zirconium are gaining renewed interest as fuels for the Versatile Test Reactor and a number of microreactor designs. Implementing metallic fuel in reactors creates the need for robust descriptive and predictive fuel performance modeling. Here, the current state of metallic fuel performance modeling relies on empirical equations derived from historical experiments, which may be unreliable when applied outside of their temperature, power, and composition phase space. One area where such data is lacking is the irradiation behavior of alpha -U, specifically tearing and porosity formation at the early stages of irradiation. While grain boundaries likely play a key role in this fuel behavior, relatively little is known about grain boundaries in alpha-U. Thus, we evaluate the grain boundary, surface energy, and work of adhesion of alpha -U utilizing molecular dynamics. Symmetric tilt grain boundaries (STGBs) are analyzed with the tilt plane oriented along each major crystallographic axis, for a total of eighty unique grain boundaries. The effect of temperature, tilt plane, and misorientation angle on interfacial energies are analyzed. The interfacial energies typically increase with temperature and there is significant variance as a function of misorientation angle, irrespective of the tilt plane. At 500 K, the average surface energy (1.23 J/m 2 ) is approximately 1.5 times the grain boundary energy (0.79 J/m 2 ), and the work of adhesion is approximately twice the grain boundary energy (1.68 J/m 2 ). Orientations for the likely formation of twins and likely failure planes are identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Layer dissolution in accumulative roll bonded bulk Zr/Nb multilayers under heavy-ion irradiation

In this work, the heavy-ion irradiation behavior of bulk zirconium-niobium multilayered composites was investigated up to large doses. Multilayers with an average individual layer thicknesses ranging between 15 and 80 nm were synthesized by accumulative roll bonding technique. The irradiation was performed with a defocused 7 MeV Zr 2+ ion beam at 500 °C. The maximum dose achieved was ~145 dpa at the depth of ~1.5 μm from the irradiated surface. Sub-surface microstructural damage and the chemical redistribution were characterized by transmission electron microscopy and energy dispersive spectroscopy, respectively. Irrespective of the layer thicknesses, the irradiation condition caused layer instability and the extent of damage varied with the dose levels. Doses lesser than ~60 dpa caused layer fragmentation and greater than ~60 dpa resulted in layer dissolution. The chemical mixing of layers occur to a depth of ~1 μm, consuming multiple bi-layer periods. Despite the elevated irradiation temperature (500 °C) and a slightly positive heat of mixing (+6 kJ/mol), no phase separation was observed and single-phase was retained in the mixed region. The results demonstrate that chemical mixing was facilitated by the liquid phase miscibility of Zr and Nb, which propelled the interdiffusion within the thermal spikes towards mixing.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Determination of the hydrogen heat of transport in Zircaloy-4

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. Here, the hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Hydrogen effects on thermal diffusivity and electrical resistivity of zircaloy cladding

Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. Here, in this work, we first used laser flash analysis method and four-probe configuration method to measure the thermal diffusivity and electrical resistivity of Zircaloy tubes respectively, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the laser flash analysis setup and four-probe configuration method for investigating thermal diffusivity and electrical resistivity of Zircaloy tubes. We further investigated the hydrogen effect on thermal diffusivity and electrical resistivity of Zircaloy. Hydrogen plays significant roles in the thermal diffusivity of Zircaloy, which depends on the hydrogen concentration. For higher hydrogen contents (1130 and 1820 wppm in this work), where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature at the α-Zr + δ-hydride phase regime, while increase at higher temperature at the α-Zr + β-Zr phase regime. For low hydrogen concentration, where hydride dissolved into α-Zr matrix phase at higher temperature, the thermal diffusivity is lower than non-hydrided Zircaloy-4 with similar temperature-dependent trends. Such observation demonstrates the hydrogen effects on reducing thermal diffusivity of α-Zr phase. Hydrogen increases the electrical resistivity of Zircaloy. Similar to thermal diffusivity results, the phase transformation causes a reversal of temperature-dependent trends in electrical resistivity results in Zircaloy-4 with higher hydrogen contents.

36 MATERIALS SCIENCE↗