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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

A Literature Review on the Study of Moisture in Polymers

This literature review covers the main chemical and physical interactions between moisture and the polymer matrix. Fickian versus Non-Fickian diffusion behaviors are discussed in approximating the characteristics of moisture sorption. Also, bound water and free water sorbed in polymers are distinguished. Methods to distinguish between bound and free water include differential scanning calorimetry, infrared spectroscopy, and time-domain nuclear magnetic resonance spectroscopy. The difference between moisture sorption and water sorption is considered, as well as the difficulties associated with preventing moisture sorption. Furthermore, specific examples of how moisture sorption influences polymers include natural fiber-polymer composites, starch-based biodegradable thermoplastics, and thermoset polyurethane and epoxies.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a “fuel body”, in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 8-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both to be focused on obtaining fission product transport data to support improvements in modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These goals were combined in the AGR-3/4 irradiation experiment, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body, in which the outer graphite ring included a floor and cap to fully enclose the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to obtain data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 1-3, 4-3, 10-1, and 10-2

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support modeling improvement. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Determination of Average Burnup in AGR-3/4 Compacts 1-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite. These two goals were combined in the AGR-3/4 irradiation, which consisted of 12 independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. Two capsule types were used: a standard capsule and a so-called fuel body, in which the outer graphite ring included a floor and cap that fully enclosed the fuel. The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from the matrix and graphite at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

An experimental study of the thermal history of fragment-laden 'basalt' 77115

Phase equilibrium and dynamic crystallization experiments using a synthetic analog of the 77115 matrix composition restrict interpretations of the thermal history of the lunar sample. Rapid early cooling (10-25 G/hr) of melt containing a large number of a nuclei and slow late stage cooling (less than 7 C/hr) are required to crystallize the textures and mineralogy of the 77115 matrix. This result is in agreement with that predicated by the nature and extent of diffusive cation exchange between various xenocrysts phases and the 77115 matrix.

Thornber, C. R.↗
NASA NTRSDate not supplied

Development of Multilayered Coatings for Nerva-Type Graphite Fuel Elements- An Overview

“Mid-passage corrosion” (MPC) was observed in NbC and ZrC-coated graphite (Gr)-based nuclear fuels during the NERVA/Rover programs. A new multilayered coating concept was proposed to solve this problem. The concept envisions designing a compliant multilayered coatings architecture to accommodate the thermal expansion mismatch strains between the ZrC outer coating and the Gr/(U,Zr)C fuel matrix. Additionally, the coating architecture uses diffusion barriers to minimize carbon diffusion from the substrate, and hydrogen diffusion from the flowing gas stream, into the multilayer coating. The present paper provides a brief overview of the development of multilayered coatings deposited on Gr disks, as well as within the inner channels of a19-hole hexagonal rod 152.4 mm long, by chemical vapor deposition (CVD) in proof-of concept studies. The cross-sectional microstructural observations are discussed. These trials have established that it is possible to deposit multilayered coatings by CVD in 19-hole hexagonal rods similar to the NERVA fuel rod design although the quality of the coatings and the uniformity of their thicknesses require further improvement.

Multilayer coatings, graphite, nuclear thermal pro↗
NASA NTRS2023

Ultrasonic Inspection During Autoclave Cure of Reflowable-Interface Composite Joints

In structural bonds, the interface between adherend and adhesive is nearly two-dimensional making the interface susceptible to minute quantities of contamination, which can cause weak bonds. Regulatory organizations such as the Federal Aviation Administration (FAA) often require redundant load paths in secondary-bonded primary-structures due to uncertainty in bond performance. To address this issue, the NASA Convergent Aeronautics Solutions (CAS): Adhesive Free Bonding of Composites (AERoBOND) project is investigating reformulated aerospace epoxy-matrix resins to enable reflow and diffusion of the resin at the joint interface during a secondary bonding and cure process that can eliminate the material discontinuity at the interface. Implementing in-situ process monitoring enables assessing the bond quality during processing rather than waiting for post-fabrication mechanical testing to be completed. In this paper, an in-situ ultrasonic inspection system is used to monitor the joining of three composite laminates assembled using the AERoBOND technique. For each panel, the amplitude of the wave reflection at the joint was measured throughout the cure cycle. The results indicate the timing and extent of reflow and cure of the epoxy resin at the joint. Factors indicating end of cure were ascertained and a metric was developed to qualitatively predict acceptable mode-II fracture toughness based on the ultrasonic amplitude at the joint line during cure. In addition, since the inspection system scans an area of the joint, high-resolution localized results can be obtained across the joint.

Tyler B Hudson↗
NASA NTRS2024

Offset-Stoichiometric Reflowable Composite Bonding Method with Adhesive for Mitigating Strict Faying Surface Tolerances

Inherent susceptibility of adhesive bonds to miniscule quantities of contamination can cause undetectable weakened bonds. For this reason, the Federal Aviation Administration (FAA) places strict regulations on adhesively bonded joints in primary aircraft structures. To meet certification requirements aircraft manufactures resort to redundant load paths in the form of fasteners which inherently add weight to the structure and increase manufacturing time. In prior work, a secondary bonding technique called AERoBOND was developed, which utilized off-stoichiometric epoxymatrix resins to facilitate reflow and diffusion of the resin within the joint interface during a secondary bonding/cure process, thus achieving a bond similar to a co-cured joint. However, the AERoBOND process required tight spatial tolerances between the two parts being joined. This study examined the utilization of conventional adhesive with the AERoBOND method to act as a filler in the joint line, effectively reducing the need for tight tolerances on the joining parts and serving as a flexible alternative for existing manufacturing processes. Ultrasonic inspection, optical microscopy, and ASTM International standard tests were performed to analyze the joints for defects and to quantify the mode-I and mode-II interlaminar fracture toughness and short beam strength of the proposed methodology with varying manufacturing parameters. The comprehensive results indicate that the AERoBOND+ method with and without surface preparation performs comparably to co-cured and conventional, adhesively bonded joints when secondary cured in an autoclave with 791 kPa of pressure. As an example, the AERoBOND+ panel without surface preparation bonded with 791 kPa of pressure (referred to as AB+3 throughout paper) had a mode-I fracture toughness (G Ic ) of 0.643 kJ/m 2 and a mode-II fracture toughness (G IIc ) of 4.000 kJ/m 2 in the non-precracked condition and 4.218 kJ/m 2 in the precracked condition at the adhesive-to-prepreg interface. These results were 96%, 142%, and 217%, respectively, of a co-cured baseline panel (referred to as C1 throughout paper).

Composites↗
DOE OSTI2018

Modeling the Long-Term Behavior of Cementitious Barriers - 18188

Cementitious materials are commonly included as backfill/barrier materials in geological repositories for radioactive waste. An understanding of the long-term behavior of cement-based barriers is therefore required to provide confidence in their safety functions within an environmental safety case. Geochemical modelling has been undertaken on several different radioactive waste repository concepts (and their natural/industrial analogues), to explore the nature of cement degradation over timescales relevant to postclosure safety (many tens of thousands of years to c. 1 Ma). In radionuclide transport calculations undertaken as part of performance assessments, it is typically considered that the properties of cementitious engineered barrier materials change as they degrade, potentially resulting in an increase in porosity hydraulic conductivity, and effective diffusion coefficients. However, reactive transport modelling (supported by some analogue data) suggests that, in some repository systems, a degree of clogging of the cement pore space could occur and that some advective flow paths (cracks in the barriers, fractures in the host rock) could become (to some extent) infilled by cement and rock alteration products. This could result in a reduction in water flow and contaminant transport in the vicinity of the cement-rock interface, but some uncertainties remain (especially process couplings and the nature of water flow in complex fracture networks that are becoming clogged). In addition, the armoring of fracture surfaces could result in decreased potential for diffusion of solutes into the rock matrix. It is envisaged that ongoing work will continue to improve reactive-transport model capabilities, such that the implications cement degradation has for contaminant transport can be predicted with greater confidence. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Simulating the Effect of Fracture Connectivity on Repository Performance with GDSA Framework - 18589

This work examines the possible migration of radionuclides from a deep geologic nuclear waste repository sited in fractured crystalline host rock. The key safety concern is the potential effect on waste isolation of the inter-connectivity of the fracture network, which is primarily established by the temporal evolution of the temperature and stress fields at the time of original rock intrusion and cooling. Two end members are considered in these simulations, one with a high degree of connectivity to the biosphere, such that advective transport through the fracture and fault network controls radionuclide migration, and the other with a low degree of connectivity, such that slow diffusive transport through the crystalline rock matrix is the controlling process for radionuclide migration to the biosphere. Both end members have several areally extensive, high-transmissivity deformation zones, whose distance to the repository is an important factor for waste isolation capability. Uncertainties in fracture properties (transmissivity, orientation, radius) and fracture distribution also give rise to uncertainty in waste isolation capability (i.e., the overall connectivity of the repository with the surface biosphere). These types of natural system uncertainties are generally present in any geologic site-characterization program and thus represent an important factor in assessing repository performance in hard rock environments. The waste considered in these simulations are spent fuel rod assemblies from the U.S. commercial nuclear reactor fleet. Thermal output of the waste must be considered in the simulations and can influence the rate and timing of waste package failure, waste form degradation, and fluid flux due to thermal expansion around the repository horizon. To achieve a high-fidelity representation of radionuclide transport in fractures and rock matrix, combined with thermal energy transport and fluid flow in fractures and matrix, the mathematical model is solved numerically in a parallel high-performance computing (HPC) environment on a finite volume unstructured grid consisting of approximately 5 million cells, using Geologic Disposal Safety Assessment (GDSA) Framework (https://pa.sandia.gov), an open-source performance assessment tool for deep underground disposal of nuclear waste. GDSA Framework uses PFLOTRAN to solve the balance equations on a three-dimensional grid with heterogeneous properties, using multiple processors in a parallel configuration based on domain decomposition. The fracture networks in these simulations are originally generated as discrete fracture networks (DFNs), which are sets of two-dimensional planes distributed in a three-dimensional domain. The method used in GDSA Framework maps the stochastically generated DFN to an equivalent continuous porous medium (ECPM) domain that allows for the simulation of coupled heat flow, fluid flow, and radionuclide transport, including heat conduction through the matrix of the fractured rock, which is a process not easily modeled in a DFN representation. Computational efficiency is also greatly enhanced using the ECPM method, allowing for a realistic representation and analysis of uncertainties in a multi-realization performance assessment of a deep geologic repository. The effect of fracture connectivity on the waste isolation safety function, as brought to light by these GDSA Framework simulations, points to the importance of including a realistic representation of uncertainties in fracture properties and distribution (effectively, uncertainty in spatial heterogeneity) in repository safety assessment simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Towards High Performance Oxide-Dispersion-Strengthened Alloys

The current first-generation candidate swelling-resistant alloys envisioned to replace austenitics are ferritic and ferritic-martensitic steels, with second-generation alloys being yttria dispersion-hardened variants of these steels. In ferritic-martensitic alloys, ferrite grains always swell earlier than martensite. Currently, most oxide dispersion-strengthened (ODS) alloys have only a ferrite matrix, thus giving up the first line of defense against swelling provided by tempered martensite. Towards development of better swelling-resistant ODS alloys, dual phase (ferrite + martensite) ODS alloys with oxide particles uniformly distributed in both ferrite and martensite phases is needed. Furthermore, fundamental understanding on mechanisms governing dispersoid stability and its correlation with void swelling is greatly needed. We reported here further alloy improvement by exploring new 9Cr ODS alloy with majority of oxide dispersoids uniformly distributed in tempered martensite grains. The alloy was irradiated by 3.5 MeV Fe self ions up to 600 dpa (displacements per atom) and at temperatures ranging from 325 to 625 deg. C. According to SRIM code, the dpa profile peaks at about 1000 nm. First of all, ion irradiation reduces the dispersoid sizes. The dispersoid sizes begin to increase beyond ∼1000 nm. Second, higher irradiation temperatures result in larger dispersoids on average. Third, there is no significant dispersoid size difference between 100 and 200 dpa irradiation. The equilibrium sizes are quickly saturated. Size and coherency of a dispersoid is closely related. In martensite grains, fine dispersoids tend to be coherent while coarse dispersoids tend to incoherent. We found that at a low irradiation temperature, coherent small dispersoids become smaller and their interface coherency remains. In a comparison, incoherent large dispersoids disappear after irradiation. At an intermediate irradiation temperature, the final equilibrium dispersoid sizes increase. At a high irradiation temperature, the final dispersoid sizes further increase and some large semi-coherent dispersoids survive. In ferrite grains, as-received dispersoids are relatively smaller in comparison with martensite grains. Their equilibrium size changes after irradiation follow a similar trend but large dispersoids completely disappear even under high temperature ion irradiation. We propose a model to explain the above dispersoid size changes under various irradiation conditions. The model considers the competing effects between ballistic-collision- assisted dispersoid dissolution into matrix and thermally driven back-diffusion for recovery. When the competing reaches a balance, an equilibrium size distribution curve is determined. For dispersoids smaller than the equilibrium size, they grow upon ion irradiation. Large dispersoids may shrink and reach the equilibrium size but ultra-large ones will completely disappear if they shrinkage paths are far away from the curve. The model further considers the complexity from the dpa rates, and damage cascade morphology.

36 MATERIALS SCIENCE↗
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↗
DOE OSTI2019

Relative Size of the Polymer and Nanoparticle Controls Polymer Diffusion in All-Polymer Nanocomposites

The dynamics of polymers in an all-polymer nanocomposite that are composed of soft cross-linked polystyrene nanoparticles and linear polystyrene have been examined. In this article, we describe how the relative size of the nanoparticle to that of the polymer chain and its rigidity impact the linear polymer chain diffusion. The results of the in situ neutron reflectivity experiments show three distinct regimes in the linear polymer diffusion. The results indicate that the inclusion of soft nanoparticles increases the amount of topological constraints and confinement effects for low matrix molecular weight polymer. At modest molecular weights where the size of the nanoparticle and polymer chain are similar, the soft nanoparticles neither inhibit nor enhance the linear polymer diffusion, while at the highest polymer matrix molecular weight, the linear polymer diffusion increases due to an increase in the constraint release mechanism by the soft nanoparticles. Thermal analysis demonstrates that the nanoparticles do not increase the free volume of the system nor do they behave as a plasticizing agent. These results are interpreted to indicate that the competition between a topological barrier effect and enhancing constraint release defines the behavior of a given all-polymer nanocomposite.

36 MATERIALS SCIENCE↗
NASA NTRS1993

Effect of metamorphism on isolated olivine grains in CO3 chondrites

The presence of a metamorphic sequence in the CO3 chondrite group has been shown previously to result in changes in properties of chondrule silicates. However, the role of isolated olivine grains during metamorphism of these chondrites has not been addressed. Isolated olivine grains in two metamorphosed CO3 chondrites, Lance and Isna, have been investigated in this study in order to assess the compositional properties of isolated olivine grains that may be attributable to metamorphism. Compositional changes in isolated olivines with increasing petrologic subtype are very similar to changes in chondrule olivines in the same chondrites. Olivine compositions from all occurrences (chondrules, isolated grains, and matrix) converge with increasing petrologic subtype. The degree of equilibration of minor elements is qualitatively related to the diffusion rate of each element in olivine, suggesting that diffusion-controlled processes are the most important processes responsible for compositional changes within the metamorphic sequence. The data are consistent with metamorphism taking place in a closed system on the CO3 chondrite parent body. Fe-poor olivine grains in metamorphosed chondrites are characterized by an Fe-rich rim, which is the result of diffusion of Fe into the grains from Fe-rich matrix. In some instances, 'complex', Fe-rich rims have been identified, which appear to have originated as igneous overgrowths and subsequently to have been overprinted by diffusion processes during metamorphism. Processes experienced by CO3 chondrites are more similar to those experienced by the ordinary chondrites than to those encountered by other carbonaceous chondrites, such as the CV3 group.

Jones, Rhian H.↗
DOE OSTI2018

Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation

Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86 x 10 17 ions/cm 2 , or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction (ARD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, D n - D$n\atop{0}$ = K phi, with n = 5.1. The Al-ZrN interaction products (Al 3 Zr and AIN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance.

36 MATERIALS SCIENCE↗