Steven Zinkle

A long-standing and critical issue in the field of irradiated structural materials is that void swelling is significantly higher in face-centered cubic-structured (fcc) materials (1% dpa⁻¹) as compared to that of body-centered cubic-structured (bcc) materials (0.2% dpa⁻¹). Despite extensive research in this area, the underlying mechanism of the difference in swelling resistance between these two types of materials is not yet fully understood. Here, by combining atomistic simulations and STEM imaging, we find stacking fault tetrahedra (SFTs) are the primary cause of the high swelling rate in pure fcc copper. We reveal that SFTs in fcc copper are not neutral sinks, different from the conventional knowledge. On the contrary, they are highly biased compared to other types of sinks because of the SFT-point defect interaction mechanism. SFTs show strong absorption of mobile self-interstitial atoms (SIAs) from the faces and vertices, and weak absorption of mobile vacancies from the edges. We compare the predicted swelling rates with experimental findings under varying conditions, demonstrating the distinct contributions of each type of sink. These findings will contribute to understanding the swelling of irradiated structural materials, which may facilitate the design of materials with high swelling resistance.

Critical aspects of innovative design in engineering disciplines like infrastructure, transportation, and medical applications require the joining of dissimilar materials. This study investigates the literature on solid-state bonding techniques, with a particular focus on diffusion bonding, as an effective method for establishing engineering bonds. Welding and brazing, while widely used, may pose challenges when joining materials with large differences in melting temperature and can lead to mechanical property degradation. In contrast, diffusion bonding offers a lower temperature process that relies on solid-state interactions to develop bond strength. The joining of tungsten and steel, especially for fusion reactors, presents a unique challenge due to the significant disparity in melting temperatures and the propensity to form brittle intermetallics. Here, diffusion characteristics of tungsten–steel interfaces are examined and the influence of bonding parameters on mechanical properties are investigated. Additionally, CALPHAD modeling is employed to explore joining parameters, thermal stability, and diffusion kinetics. The insights from this research can be extended to join numerous dissimilar materials for specific applications such as aerospace, automobile industry, power plants, etc., enabling advanced and robust design with high efficiency.

Nitride ceramics have been investigated for different applications in the nuclear industry, such as space nuclear power, fusion reactor diagnostics and plasma heating, inert matrix fuels, and accident tolerant fuels. Although thermal conductivity remains one of the most important properties to track following irradiation, traditional techniques such as laser flash and xenon flash are limited to bulk sample characterization, which requires lengthy and cost-consuming neutron irradiation. This work used spatial domain thermoreflectance (SDTR) for the micrometer-scale measurement of thermal conductivity in 15 MeV Ni ion-irradiated silicon nitride and zirconium nitride from 1 to 50 dpa and 300 to 700 °C. The SDTR-measured unirradiated thermal conductivity was found to be consistent with the published data on bulk samples. Electrically conductive ZrN exhibits modest reduction after irradiation which is minimal at the highest irradiation temperatures. In electrically insulating Si3N4, the reduction is more significant and unlike ZrN, the reduction remains significant even at a higher irradiation temperature. The thermal resistance evolution following irradiation was compared with lattice swelling, which was determined using grazing incidence x-ray diffraction, and radiation-induced defects were observed using transmission electron microscopy. A saturation value was observed between 15 and 50 dpa for thermal conductivity degradation in both nitride ceramics and a direct correlation with high-temperature defect recombination was observed, as well as the potential presence of additional carrier scattering mechanisms.

The pulsed linear plasma device Prototype Material Plasma Exposure eXperiment (Proto-MPEX) uses a radio frequency (RF) helicon antenna with an aluminum nitride ceramic window for plasma production. The RF sheath created under the helicon antenna is sufficient to cause ion impact energies to be greater than the sputtering threshold of the AlN helicon window material and for impurities to be created. Here, we investigate the RF sheath on the inner diameter of the helicon window and its impact on the impurity production rates in Proto-MPEX. Three models—a 3D COMSOL finite element RF model of the Proto-MPEX helicon region, a rectified DC sheath potential model, and the 3D Global Impurity TRansport code—are coupled together to study impurity production and transportation. This novel method of impurity generation and transport modeling spans length scales ranging from the sheath (millimeters or less) up to the full device (meters) and can be applied to other radio frequency sources and antennas in a wide range of plasma physics studies, including basic plasmas, low-temperature processing plasmas, plasma thrusters, and fusion plasmas.

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) is a linear plasma device being used in plasma source research and development (R&D) for the proposed MPEX. Once the R&D is completed, this device can also be used to perform plasma–material interaction studies. To perform these studies, a new materials analysis and particle probe (MAPP) has been constructed. The MAPP’s components are a sample holder and manipulator and a custom vacuum chamber with ports to facilitate surface chemistry diagnostics. The MAPP’s overall design enables rapid sample turnaround and in vacuo surface characterization. The surface analysis vacuum chamber has ports for x-ray photoelectron spectroscopy, thermal desorption spectroscopy, back-scatter ion scattering spectroscopy, forward-scatter ion scattering spectroscopy, and direct recoil spectroscopy. The sample manipulator and holder is a Lesker/UHV Multi-Centre Analytical Stage, which is used to place the samples in the exposure region of the Proto-MPEX or the analysis position in the MAPP vacuum chamber. The sample holder has a heating capability of up to 1200 °C for heated exposure and for desorption studies. In this work, we present the MAPP’s design and the first tungsten sample exposure with ex situ analysis that shows a surface deposition layer on the exposed target, highlighting the need for additional in situ measurements on the Proto-MPEX.

The intrinsic compatibility of silicon carbide (SiC) and hydrogen (H₂) at high temperatures (2000‐2473 K) and pressure near one atmosphere was evaluated through a combination of thermodynamic calculations and hot hydrogen exposure testing. Thermodynamic calculations predict the decomposition of SiC in a hydrogen environment to form free silicon (Si) and free carbon (C). Free Si is predicted to vaporize from the surface as a volatile species, while free C may interact with H₂ to form the hydrocarbons CH₄ (T < 2100 K) or C₂H₂ (T > 2100 K). Coupons of high purity chemical vapor deposition (CVD) β‐SiC were exposed to slowly flowing hydrogen at temperatures ranging between 2000 and 2473 K. SiC experienced active attack as the result of H₂ exposure, exhibiting linear weight loss kinetics and an Arrhenius dependence of weight loss on exposure temperature. The linear volatilization constant was experimentally evaluated to correspond with an activation energy of 370 ± 18 kJ/mol. Due to the dependence of observed corrosion rates on gas velocity, corrosion of SiC in flowing H₂ was determined to be governed by external mass transfer of volatile Si species through the boundary layer. Experimentally derived mass losses were in good agreement with mass losses predicted by a boundary layer limited gas diffusion model.

Atomic collision processes are fundamental to numerous advanced materials technologies such as electron microscopy, semiconductor processing and nuclear power generation. Extensive experimental and computer simulation studies over the past several decades provide the physical basis for understanding the atomic-scale processes occurring during primary displacement events. The current international standard for quantifying this energetic particle damage, the Norgett−Robinson−Torrens displacements per atom (NRT-dpa) model, has nowadays several well-known limitations. In particular, the number of radiation defects produced in energetic cascades in metals is only ~1/3 the NRT-dpa prediction, while the number of atoms involved in atomic mixing is about a factor of 30 larger than the dpa value. Here we propose two new complementary displacement production estimators (athermal recombination corrected dpa, arc-dpa) and atomic mixing (replacements per atom, rpa) functions that extend the NRT-dpa by providing more physically realistic descriptions of primary defect creation in materials and may become additional standard measures for radiation damage quantification.

High-entropy alloys, near-equiatomic solid solutions of five or more elements, represent a new strategy for the design of materials with properties superior to those of conventional alloys. However, their phase space remains constrained, with transition metal high-entropy alloys exhibiting only face- or body-centered cubic structures. Here, we report the high-pressure synthesis of a hexagonal close-packed phase of the prototypical high-entropy alloy CrMnFeCoNi. This martensitic transformation begins at 14 GPa and is attributed to suppression of the local magnetic moments, destabilizing the initial fcc structure. Similar to fcc-to-hcp transformations in Al and the noble gases, the transformation is sluggish, occurring over a range of >40 GPa. However, the behaviour of CrMnFeCoNi is unique in that the hcp phase is retained following decompression to ambient pressure, yielding metastable fcc-hcp mixtures. This demonstrates a means of tuning the structures and properties of high-entropy alloys in a manner not achievable by conventional processing techniques.

Nuclear energy provides more than 10% of electrical power internationally and the increasing engagement of nuclear energy is essential to meet the rapid worldwide increase in energy demand. A paramount challenge in the development of advanced nuclear reactors is the discovery of advanced structural materials that can endure extreme environments, such as severe neutron irradiation damage at high temperatures. It has been known for decades that high dose radiation can introduce significant void swelling accompanied by precipitation in austenitic stainless steel (SS). Here we report, however, that through nanoengineering, ultra-fine grained (UFG) 304L SS with an average grain size of ~100 nm, can withstand Fe ion irradiation at 500°C to 80 displacements-per-atom (dpa) with moderate grain coarsening. Compared to coarse grained (CG) counterparts, swelling resistance of UFG SS is improved by nearly an order of magnitude and swelling rate is reduced by a factor of 5. M₂₃C₆ precipitates, abundant in irradiated CG SS, are largely absent in UFG SS. This study provides a nanoengineering approach to design and discover radiation tolerant metallic materials for applications in extreme radiation environments.

Proposed fusion and advanced (Generation IV) fission energy systems require high-performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts of transmutant hydrogen and helium isotopes. After a brief overview of fusion reactor concepts and radiation effects phenomena in structural and functional (nonstructural) materials, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials in which vacancies are immobile at the design operating temperatures, or engineer materials with high sink densities for point defect recombination. Environmental and safety considerations impose several additional restrictions on potential materials systems, but reduced-activation ferritic/martensitic steels (including thermomechanically treated and oxide dispersion–strengthened options) and silicon carbide ceramic composites emerge as robust structural materials options. Materials modeling (including computational thermodynamics) and advanced manufacturing methods are poised to exert a major impact in the next ten years.

The stress required to deform a perfect crystal to its elastic limit while maintaining perfect periodicity, the so-called ideal strength, sets the gold standard for the strength of a given material. Materials this strong would be of obvious engineering importance, potentially enabling more efficient turbines for energy production, lighter materials for transportation applications, and more reliable materials for nuclear reactor applications. In practice, the strength of engineering materials is often more than two orders of magnitude less than the ideal strength due to easily activated deformation processes involving dislocations. For many materials, precipitate strengthening is a promising approach to impede dislocation motion and thereby improves strength and creep resistance. This observation begs the question: What are the limits of nanoparticle strengthening? Can the ideal strength of a matrix material be reached? To answer these questions, we need a detailed, atomic scale understanding of the interactions between dislocations and obstacles. Fortunately, simulations are beginning to explore this interaction.

The migration of point defects, for example, crystal lattice vacancies and self-interstitial atoms (SIAs), typically occurs through three-dimensional random walk in crystalline solids. However, when vacancies and SIAs agglomerate to form planar clusters, the migration mode may change. We observed nanometer-sized clusters of vacancies exhibiting one-dimensional (1D) fast migration. The 1D migration transported a vacancy cluster containing several hundred vacancies with a mobility higher than that of a single vacancy random walk and a mobility comparable to a single SIA random walk. Moreover, we found that the 1D migration may be a key physical mechanism for self-organization of nanometer-sized sessile vacancy cluster (stacking fault tetrahedron) arrays. Harnessing this 1D migration mode may enable new control of defect microstructures such as effective defect removal and introduction of ordered nanostructures in materials, including semiconductors.

A brief review is given of fundamental materials science concepts important for development of structural materials for fusion energy systems. Particular attention is placed on displacement damage effects associated with the unique deuterium-tritium fusion environment. Recent examples of multiscale materials modeling results (closely coupled with experimental studies) are summarized. Fundamental differences in the behavior of body centered cubic versus face centered cubic crystal structures are highlighted. Finally, a brief overview is given of the high-performance reduced-activation materials being developed by fusion.

An experimental investigation of the in situ electrical conductivity of Wesgo Al995 polycrystalline alumina at approximately 450 °C has been performed at the high flux beam reactor at Brookhaven National Laboratory. The measured radiation induced conductivity (RIC) was about 10−8 S/m at an ionizing dose rate of 6000 Gy/s. No evidence for permanent radiation induced electrical degradation was observed for an applied electric field of 147 V/mm up to a dose level of ≈1.8 displacements per atom. The effect of neutron irradiation on the electrical properties of two mineral insulated cables was also investigated. The RIC in the MgO insulation of a coaxial and a triaxial cable was measured to be in the range of 6–20 ×10−8 S/m at an ionizing dose rate of ≈ 6000 Gy/s.

This review provides a comprehensive evaluation of the state-of-knowledge of radiation effects in crystalline ceramics that may be used for the immobilization of high-level nuclear waste and plutonium. The current understanding of radiation damage processes, defect generation, microstructure development, theoretical methods, and experimental methods are reviewed. Fundamental scientific and technological issues that offer opportunities for research are identified. The most important issue is the need for an understanding of the radiation-induced structural changes at the atomic, microscopic, and macroscopic levels, and the effect of these changes on the release rates of radionuclides during corrosion.

A resonant cavity technique was used to measure the dielectric constant and loss tangent of ceramic insulators at a frequency near 100 MHz during pulsed fission reactor irradiation near room temperature. Tests were performed on single crystal and several different grades of polycrystalline Al2O3, MgAl2O4, AlN, and Si3N4. Lead shielding experiments were performed for some of the irradiations in order to examine the importance of gamma ray versus neutron irradiation effects. With the exception of AlN, the dielectric constant of all of the ceramics decreased slightly (&lt;0.2% change) during the pulsed fission reactor irradiation. The dielectric constant of AlN was observed to slightly increase during irradiation. Significant transient increases in the loss tangent to values as high as 6×10−3 occurred during pulsed reactor irradiation with peak ionizing and displacements per atom (dpa) radiation fields of 4.2×104 Gy/s and 2.4×10−6 dpa/s, respectively. The loss tangent measured during irradiation for the different ceramics did not show any correlation with the preirradiation or postirradiation values. Analysis of the results indicates that the transient increases in loss tangent are due to radiation induced increases in the electrical conductivity. The loss tangent increases were proportional to the ionizing dose rate in all materials except for AlN, which exhibited a dose rate exponent of ∼1.6.

Magnesium aluminate spinel (MgAl₂O₄) with an ¹⁷O enrichment (¹⁷O/O_Tot) of about 23 at.% was prepared by reacting fine mixtures of aluminum hydroxide (enriched with ¹⁷O) and magnesium oxide of normal isotopic content. The material was prepared for experiments in which the radiation damage produced in a fusion reactor is simulated by fission reactor exposures. The powder mixtures were obtained by hydrolyzing, with water containing the ¹⁷O iostope, a mixture of aluminum isopropoxide and magnesium oxide powder. The mixture was converted into pure spinel by a series of heat treatments and grindings. Essentially fully dense bodies, which contained about 45% of the ¹⁷O isotope initially present in the water, were successfully fabricated provided that all thermal treatments were conducted in argon or vacuum atmospheres.

The general techniques necessary to produce a high‐quality cross‐sectioned ceramic specimen for transmission electron microscope observation are outlined. A particularly important point is that the width of the glued region between faces of the ceramic specimen must be < 0.2 μm to prevent loss of the near‐surface region during ion milling. A recently developed vise for gluing ceramic cross‐section specimens is described, and some examples of the effect of glue thickness on specimen quality are shown.

Alumina enriched in ¹⁷O was successfully fabricated from aluminum isopropoxide and water containing the ¹⁷O isotope. This material was necessary for an experiment to study the radiation damage expected in alumina exposed to a nuclear fusion reactor environment. The enrichment levels of specimens subjected to different preparation schedules were measured using a nuclear reaction analysis technique. Replacement of the ¹⁷O isotope in the ceramic by atmospheric oxygen occurred readily. Therefore, successful fabrication of suitably enriched alumina specimens required that all processing steps be performed under vacuum or inertgas environments. The optimum fabrication procedure produced enriched alumina specimens of >99.5% of theoretical density, ∽10‐μm grain size, and a flexural strength of 280 MPa.

Stoichiometric polycrystalline magnesium aluminate spinel has been irradiated at 25° and 650°C with 2.4‐MeV Mg⁺ ions to a fluence of 1.4 × 10²¹ ions/m² (∼35 dpa (displacement per atom) peak damage level). Microindentation hardness measurements and transmission electron microscopy combined with energy dispersive X‐ray spectroscopy measurements were used to characterize the irradiation effects. The room‐temperature hardness of spinel increased by about 5% after irradiation at both temperatures. There was no evidence for amorphization at either irradiation temperatures. Interstitial‐type dislocation loops lying on {110} and {111} planes with Burgers vectors along 〈110〉 were observed at intermediate depths (∼1 μm) along the ion range. The 〈110〉{111} loops are presumably formed from 〈111〉{111} loops as a result of a shear on the anion sublattice. Only about 0.05% of the calculated displacements were visible in the form of loops, which indicates that spinel has a high resistance to aggregate damage accumulation. The peak damage region contained a high density of dislocation tangles. There was no evidence for the formation of voids or vacancy loops. The specimen irradiated at 650°C was denuded of dislocation loops within ∼1 μm of the surface.