Radiation effects on materials for electrochemical energy storage systems.

Batteries and electrochemical capacitors (ECs) are of critical importance for applications such as electric vehicles, electric grids, and mobile devices. However, the performance of existing battery and EC technologies falls short of meeting the requirements of high energy/high power and long durability for increasing markets such as the automotive industry, aerospace, and grid-storage utilizing renewable energies. Therefore, improving energy storage materials performance metrics is imperative. In the past two decades, radiation has emerged as a new means to modify functionalities in energy storage materials. There exists a common misconception that radiation with energetic ions and electrons will always cause radiation damage to target materials, which might potentially prevent its applications in electrochemical energy storage systems. But in this review, we summarize recent progress in radiation effects on materials for electrochemical energy storage systems to show that radiation can have both beneficial and detrimental effects on various types of energy materials. Prior work suggests that fundamental understanding toward the energy loss mechanisms that govern the resulting microstructure, defect generation, interfacial properties, mechanical properties, and eventual electrochemical properties is critical. We discuss radiation effects in the following categories: (1) defect engineering, (2) interface engineering, (3) radiation-induced degradation, and (4) radiation-assisted synthesis. We analyze the significant trends and provide our perspectives and outlook on current research and future directions in research seeking to harness radiation as a method for enhancing the synthesis and performance of battery materials.

Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation.

Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni-Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5-9 displacements per atom (dpa) and temperatures ranging from 430-683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni3Si stoichiometry in a Ni core-Si shell structure. This core-shell structure provides insight into a potentially unique nucleation and growth mechanism-nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels.

Mechanical testing data from neutron irradiations of PM-HIP and conventionally manufactured nuclear structural alloys.

This article presents the comprehensive mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300 and 400 °C. This article contains the data collected from post-irradiation uniaxial tensile tests following ASTM E8 specifications, fractography of these tensile bars, and nanoindentation. By making this systematic and valuable neutron irradiated mechanical behavior dataset openly available to the nuclear materials research community, researchers may now use this data to populate material performance databases, validate material performance and hardening models, design follow-on experiments, and enable future nuclear code-qualification of PM-HIP techniques.

Irradiation-Induced Amorphous-to-Crystalline Phase Transformations in Ceramic Materials.

Amorphous ceramics are a unique class of materials with unusual properties and functionalities. While these materials are known to crystallize when subjected to thermal annealing, they have sometimes been observed to crystallize athermally when exposed to extreme irradiation environments. Because irradiation is almost universally understood to introduce disorder into materials, these observations of irradiation-induced ordering or crystallization are unusual and may partially explain the limited research into this phenomenon. However, the archival literature presents a growing body of evidence of these irradiation-induced amorphous-to-crystalline (a-to-c) phase transformations in ceramics. In this perspective, the summary and review of examples from the literature of irradiation-induced a-to-c transformations for various classifications of ceramics are provided. This work will highlight irradiation conditions and material parameters that appear most influential for activating a-to-c transformations, identify trends, examine possible mechanisms, and discuss the impact of a-to-c transformations on material properties. Finally, future research directions that will enable researchers to harness a-to-c transformations to tailor materials behaviors will be provided.

Crystallographic orientation data from chloride-induced stress corrosion crack (CISCC) paths in gas tungsten arc welded (GTAW) austenitic stainless steel 304L.

This Data in Brief article presents crystallographic data collected along chloride-induced stress corrosion cracks (CISCC) in a gas tungsten arc welded (GTAW) austenitic stainless steel (AuSS) 304L. The experimental setup involved a welded stainless steel 304L coupon of dimensions 105 mm × 18.5 mm × 3 mm, loaded in a 4-point bending fixture with a maximum tensile stress of 380 MPa. The fixtured specimen was immersed in boiling magnesium chloride (MgCl2) solution until a through-crack was observed on the specimen surface after 17 hours of boiling. The cross-section was subsequently polished, and 37 cracks of interest in the heat affected zone (HAZ) and weld zone (WZ) were selected for crystallographic characterization. Scanning electron microscopy (SEM) based electron backscatter diffraction (EBSD) was used to map the grain orientations along and surrounding each crack path. The obtained orientation imaging microscopy (OIM) datasets were post-processed using EDAX OIM V8 proprietary software to generate inverse pole figures (IPF), image quality (IQ) figures, detector signal (SEM) images, and to determine the Taylor factor and Schmid factor of mapped grains. This dataset can be used to understand CISCC crack initiation, propagation, and termination behaviors, as has been reported in the accompanying original research article. This data article providing the raw EBSD OIM datasets and processed images formatted for accessibility in future studies. This comprehensive EBSD dataset can further be used to extract grain boundary misorientation information; benchmark comparative studies of SCC/CISCC in AuSS and other Fe or Ni alloys; and provide critical validation data on grain morphology, misorientation, and crystallography for GTAW and CISCC models.

Thermal Energy Transport in Oxide Nuclear Fuel.

To efficiently capture the energy of the nuclear bond, advanced nuclear reactor concepts seek solid fuels that must withstand unprecedented temperature and radiation extremes. In these advanced fuels, thermal energy transport under irradiation is directly related to reactor performance as well as reactor safety. The science of thermal transport in nuclear fuel is a grand challenge as a result of both computational and experimental complexities. Here we provide a comprehensive review of thermal transport research on two actinide oxides: one currently in use in commercial nuclear reactors, uranium dioxide (UO2), and one advanced fuel candidate material, thorium dioxide (ThO2). In both materials, heat is carried by lattice waves or phonons. Crystalline defects caused by fission events effectively scatter phonons and lead to a degradation in fuel performance over time. Bolstered by new computational and experimental tools, researchers are now developing the foundational work necessary to accurately model and ultimately control thermal transport in advanced nuclear fuels. We begin by reviewing research aimed at understanding thermal transport in perfect single crystals. The absence of defects enables studies that focus on the fundamental aspects of phonon transport. Next, we review research that targets defect generation and evolution. Here the focus is on ion irradiation studies used as surrogates for damage caused by fission products. We end this review with a discussion of modeling and experimental efforts directed at predicting and validating mesoscale thermal transport in the presence of irradiation defects. While efforts in these research areas have been robust, challenging work remains in developing holistic tools to capture and predict thermal energy transport across widely varying environmental conditions.

Influence of field conditions on quantitative analysis of single crystal thorium dioxide by atom probe tomography.

Atom probe tomography (APT), a 3D microscopy technique, has great potential to reveal atomic scale compositional variations, such as those associated with irradiation damage. However, obtaining accurate compositional quantification by APT for high bandgap materials is a longstanding challenge, given the sensitivity to field evaporation parameters and inconsistent behaviors across different oxides. This study investigates the influence of APT laser energy and specimen base temperature on compositional accuracy in single crystal thoria (ThO2). ThO2 has a broad range of applications, including advanced nuclear fuels, sensors, lasers and scintillators, electrodes, catalysis, and photonics and optoelectronics. The expected stoichiometry of ThO2 is achieved at APT base temperature of 24 K and laser energy of 100 pJ. To overcome mass resolution limitations associated with significant thermal tails, Bayesian methods are applied to deconvolute ion identity within the mass spectra. This approach affirms that the parameters chosen are appropriate for APT analysis of ThO2.

Comparative Thermal Aging Effects on PM-HIP and Forged Inconel 690.

This study compares thermal aging effects in Inconel 690 (IN690) produced by forging and powder metallurgy with hot isostatic pressing (PM-HIP). Isothermal aging is carried out over 400-800°C for up to 1000 h and then metallography and nanoindentation are utilized to relate grain microstructure with hardness and yield strength. The PM-HIP IN690 maintains a constant grain size through all aging conditions, while the forged IN690 exhibits limited grain growth at the highest aging temperature and longest aging time. The PM-HIP IN690 exhibits comparable mechanical integrity as the forged material throughout aging: hardness and yield strength are unchanged with 100 h aging, but increase after 1000 h aging at all temperatures. In both the PM-HIP and forged IN690, the Hall-Petch relationship for Ni-based super-alloys predicts yield strength for 0-100 h aged specimens, but underestimates yield strength in the 1000 h aged specimens because of thermally induced precipitation.