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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Amorphization and siliconization of silicon carbide as a first wall material

The understanding and prediction of silicon carbide (SiC) material evolution exposed to SOL plasma conditions is of prime interest because SiC represents a promising main chamber wall plasma-facing material for next-step fusion devices (low hydrogenic diffusion, good mechanical and thermal properties under neutron irradiation). Gross and net Si erosion rates from SiC surfaces in contact with a well-diagnosed L-mode plasma in the DIII-D tokamak have been simulated and the surface concentrations of impurities have been tracked as a function of time. Coupled simulation of surface model and impurity transport demonstrates amorphization of crystalline SiC exposed to L-mode plasma due to the accumulation of displacement damages under ion irradiation. This affects the lifetime of SiC plasma facing components. Surface evolution is tightly coupled to impurity transport in the plasma and therefore needs to be integrated with impurity transport simulations to effectively predict Si erosion rates and sub-surface concentrations as a function of time. The simulation workflow couples a semi-analytical surface model to the impurity transport code GITR. The surface model is a homogeneous mixed-material model that tracks physical & chemical sputtering and reflection of impurities. Gross erosion is primarily influenced by the background plasma parameters and redeposition patterns are mainly influenced by the prompt redeposition due to the gyro-orbits of impurity ions. Although crystalline form of SiC is preferable for fusion wall applications because of resistance to neutron irradiation, this work indicates that crystalline SiC will undergo amorphization under D plasma contact with implications of higher sputtering and fuel retention. These results direct us to explore the effects of amorphization on crystalline SiC and further the physics basis of SiC usage as first wall material for fusion environments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Thermal diffusivity recovery and defect annealing kinetics of self-ion implanted tungsten probed by insitu transient grating spectroscopy

Tungsten is a promising candidate material for plasma-facing armour components in future fusion reactors. A key concern is irradiation-induced degradation of its normally excellent thermal transport properties. In this comprehensive study, thermal diffusivity degradation in ion-implanted tungsten and its evolution from room temperature (RT) to 1073 K is considered. Five samples were exposed to 20 MeV self-ions at RT to achieve damage levels ranging from 3.2 x 10 -4 to 3.2 displacements per atom (dpa). Transient grating spectroscopy with in-situ heating was then used to study thermal diffusivity evolution as a function of temperature. Using a kinetic theory model, an equivalent point defect density is estimated from the measured thermal diffusivity. The results showed a prominent recovery of thermal diffusivity between 450 K and 650 K, which coincides with the onset of mono-vacancy mobility. After 1073 K annealing samples with initial damage of 3.2 x 10 -3 dpa or less recover close to the pristine value of thermal diffusivity. For doses of 3.2 x 10 -2 dpa or higher, on the other hand, a residual reduction in thermal diffusivity remains even after 1073 K annealing. Transmission electron microscopy reveals that this is associated with extended, irradiation-induced dislocation structures that are retained after annealing. A sensitivity analysis shows that thermal diffusivity provides an efficient tool for the assessing total defect content in tungsten up to 1000 K.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Influence of preheating on tungsten damage under pulsed hydrogen plasma exposure

Preheated tungsten plate behavior under multiple pulsed hydrogen plasma irradiations in the PW-7 plasma accelerator was investigated. The preheating temperature was 973 K, which is above the ductile to brittle transition temperature (DBTT ≈ 673 K). The plate was exposed to three consecutive series of 13 plasma pulses each. Preheating sufficiently suppressed surface cracking at the early irradiation stage: after 13 pulses, cracks were rare, localized, and did not form extensive fractures. During subsequent irradiation series, the area fraction of fine cracks increased by ∼ 2% due to accumulation of dislocations around the pre-existing defects. However, their width remained nearly constant, averaging 0.2–0.3 μm, with rare cracks up to 1 μm in size. Repeated plasma exposure resulted in pronounced surface roughening associated with multiple melting cycles, crater formation and blister rupture, with roughness reaching ∼ 204 nm and a recorded sample mass loss of ∼ 2.8 mg. These results indicate that controlling tungsten preheating well above DBTT effectively reduces damage. However, the surface condition during initial irradiation significantly influences its subsequent morphological evolution. Preheating to 773 K, selected regarding the minimum DBTT value, has previously been demonstrated to be an effective approach for limiting crack growth. A further increase in temperature of approximately ∼ 1000 K did not qualitatively improve results: crack still formed and widths did not decrease. At the same time, the absence of further crack coarsening indicates that the limiting effect of preheating has been reached. Considering the increased energy costs, further increasing the preheating temperature seems unjustified.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

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 (UO 2 ), and one advanced fuel candidate material, thorium dioxide (ThO 2 ). 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Dynamics of particle network in composite battery cathodes

We report that improving composite battery electrodes requires a delicate control of active materials and electrode formulation. The electrochemically active particles fulfill their role as energy exchange reservoirs through interacting with the surrounding conductive network. We formulate a network evolution model to interpret the regulation and equilibration between electrochemical activity and mechanical damage of these particles. Through statistical analysis of thousands of particles using x-ray phase contrast holotomography in a LiNi 0.8 Mn 0.1 Co 0.1 O 2 -based cathode, we found that the local network heterogeneity results in asynchronous activities in the early cycles, and subsequently the particle assemblies move toward a synchronous behavior. Our study pinpoints the chemomechanical behavior of individual particles and enables better designs of the conductive network to optimize the utility of all the particles during operation.

25 ENERGY STORAGE↗
DOE OSTI2025

Mesoscale Modeling of Dislocation Cell Structure Evolution and Radiation-Induced Segregation in Additively Manufactured Austenitic Stainless Steel

Structural alloys under irradiation develop radiation-induced segregation (RIS) at point defect sinks, which lead to undesired changes in the alloy's properties. Additively manufactured austenitic stainless steels are expected to yield a distinct response to irradiation damage owing to their unique as-printed dislocation cell structure. In this talk, we present the development of a mesoscale model for dislocation cell structure evolution coupled with RIS using the MOOSE framework. The evolution of network dislocations within spatially distinct cells and cell walls are modeled using climb-mediated edge annihilation and generation processes. Starting with initial microstructures of as-printed non-equilibrium segregations at dislocation cell walls, our 2D and 3D simulations explore the effects of irradiation temperature, dose rate, dislocation sink bias, and specimen thickness on microstructure evolution. The results will be compared against experimental characterizations of in-situ and ex-situ ion-irradiated samples, and the implications of evolving cell structure on irradiation damage response will be discussed.

additive manufacturing↗
DOE OSTI2020

From suppressed void growth to significant void swelling in NiCoFeCr complex concentrated solid-solution alloy

Void swelling can result in dimensional instability and undermine the safe operation of nuclear reactors. Current strategies to inhibit void swelling mainly focus on enhancing defect absorption and recombination by introducing high-density defect sinks. Complex concentrated solid-solution alloys (CSAs), including high-entropy alloys, can withstand severe radiation damage due to their inherent chemical complexity without interfaces. However, the underlying mechanisms for void suppression in CSAs are far from clear. In this research, we studied the void evolution with respect to irradiation depths, doses, and temperatures in equiatomic NiCoFeCr under 3 MeV Ni ion irradiations. At relatively low doses (16 and 54 displacements per atom, dpa), voids form mainly outside of the ion-damaged region, and void formation in the peak damage region is suppressed, leading to negligible swelling. However, with further increase of dose (86 up to 250 dpa), significant void growth occurs in the peak damage region and extended dislocation lines dominate instead of short dislocation lines and loops formed at lower doses. From 500 to 700 °C, the dislocation density decreases while dislocations grow. Although the overall void swelling increases dramatically from 500 to 580 °C at 54 dpa, void growth in the peak damage region is still suppressed. The transition from suppressed void growth to significant void swelling is attributed to dislocation evolution and local chemical inhomogeneity (enrichment of Fe/Cr in the matrix) at higher doses. Our study shows that controlling element diffusion and defect evolution through tuning chemical complexity can further enhance the swelling resistance of CSAs.

36 MATERIALS SCIENCE↗
DOE OSTI2022

In vitro investigation of protein assembly by combined microscopy and infrared spectroscopy at the nanometer scale

The nanoscale structure and dynamics of proteins on surfaces has been extensively studied using various imaging techniques, such as transmission electron microscopy and atomic force microscopy (AFM) in liquid environments. These powerful imaging techniques, however, can potentially damage or perturb delicate biological material and do not provide chemical information, which prevents a fundamental understanding of the dynamic processes underlying their evolution under physiological conditions. Here, we use a platform developed in our laboratory that enables acquisition of infrared (IR) spectroscopy and AFM images of biological material in physiological liquids with nanometer resolution in a cell closed by atomically thin graphene membranes transparent to IR photons. In this work, we studied the self-assembly process of S-layer proteins at the graphene-aqueous solution interface. The graphene acts also as the membrane separating the solution containing the proteins and Ca 2+ ions from the AFM tip, thus eliminating sample damage and contamination effects. The formation of S-layer protein lattices and their structural evolution was monitored by AFM and by recording the amide I and II IR absorption bands, which reveal the noncovalent interaction between proteins and their response to the environment, including ionic strength and solvation. Our measurement platform opens unique opportunities to study biological material and soft materials in general.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Operation and performance of the ATLAS semiconductor tracker in LHC Run 2

The semiconductor tracker (SCT) is one of the tracking systems for charged particles in the ATLAS detector. It consists of 4088 silicon strip sensor modules. During Run 2 (2015–2018) the Large Hadron Collider delivered an integrated luminosity of 156 fb ₋1 to the ATLAS experiment at a centre-of-mass proton-proton collision energy of 13 TeV. The instantaneous luminosity and pile-up conditions were far in excess of those assumed in the original design of the SCT detector. Due to improvements to the data acquisition system, the SCT operated stably throughout Run 2. It was available for 99.9% of the integrated luminosity and achieved a data-quality efficiency of 99.85%. Detailed studies have been made of the leakage current in SCT modules and the evolution of the full depletion voltage, which are used to study the impact of radiation damage to the modules.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

4D evolution of Cr 23 C 6 precipitates in neutron-irradiated and annealed HT-UPS steel observed via synchrotron micro-computed tomography

High-temperature-ultrafine precipitate strengthened (HT-UPS) steel is a potential structural material for advanced nuclear reactors; however, its irradiation response is not well understood. This research provides insight into irradiation-induced effects, such as precipitate evolution mechanisms and four-dimensional morphological evolution, in HT-UPS steel using synchrotron micro-computed tomography. Identical specimens were characterized pre-irradiation and post-irradiation following neutron exposure up to 0.3 displacements per atom at 600 °C. Irradiation effects were also differentiated from the annealing response of precipitates. Following neutron irradiation, the average Cr 23 C 6 precipitate size reduced, affected by the synergy of nucleation and growth, ballistic dissolution, and inverse coarsening, which was observed at fluences an order of magnitude lower than previously observed. Annealing at 600 °C for 32 h increased the average Cr 23 C 6 precipitate size and decreased the phase fraction, attributed to precipitate coarsening. The precipitate morphology evolution and resultant mechanisms can be utilized to parameterize and validate microstructural models simulating radiation damage or annealing.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Experimental Evaluation of Deformation and Fracture Mechanisms in Highly Irradiated Austenitic Steels

The present report documents recent experimental results of analysis using scanning electron microscopy/electron backscatter diffraction (SEM-EBSD) of plastic deformation mechanisms and strain localization phenomena in austenitic steels irradiated by neutrons. Experiments were performed with specimens irradiated to 125 dpa and, additionally, with specimens that experienced radiation-induced swelling up to 3%. Section 1 briefly analyzes the deformation localization in irradiated steels and its consequences on the material performance. The section describes the advantages and importance of the SEM-EBSD approach combined with in situ mechanical testing capability. Section 2 briefly introduces the experimental tools and methods (i.e., SEM/EBSD in situ tensile frame, electric discharge machine to manufacture irradiated specimens) and describes the investigated materials (i.e., element composition, irradiation conditions, and general microstructure). Section 3 describes the key experimental results and provides a brief analysis and comparison with the datasets obtained earlier within the same task (i.e., low-dose specimens). The discussion focuses on EBSD microstructure maps with strain localization features, misorientation evolution as a function of strain, and observed deformation mechanisms. Section 4 evaluates data collected in recent years on highly irradiated steel and estimates the possible misorientation evolution under irradiation. The section introduces and discusses the concept of in-service-induced damage as an irradiation-assisted stress-corrosion cracking precursor. Section 5 summarizes the work performed. As expected, the present work results are beneficial for exploring and understanding degradation mechanisms in highly irradiated in-core materials found in light water reactors after long-term in-service life.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

In situ TEM annealing of neutron-irradiated Ti reveals a two-stage mechanism for elevated temperature radiation damage recovery

Understanding how irradiation-induced defects evolve at elevated temperatures is of critical importance to predicting materials' behavior under steady-state and accident scenarios. However, such mechanistic insight into microstructural evolution is limited by the nature of ex situ annealing and subsequent imaging. Here we show direct observation and quantification of defect recovery in neutron-irradiated Ti using in situ transmission electron microscopy (TEM) annealing experiments. In agreement with our prior work, and at temperatures below the irradiation temperature (T irr = 300 °C), dislocation loops are observed to glide. At elevated temperatures (>500 °C), dislocation lines become mobile and promote significant recovery of the microstructure. These mechanisms challenge the established electron irradiation-based model for radiation damage recovery, which originally suggests dissolution of static defect clusters, and demonstrates the importance of in situ characterization in understanding defect evolution in irradiated materials.

Hirst, Charles A. [Univ. of Michigan, Ann Arbor, M↗
DOE OSTI2022

Synergies between H, He and radiation damage in dual and triple ion irradiation of candidate fusion blanket materials

Three ferritic/martensitic alloys were studied to understand the synergistic effect between single ion beam (Fe 2+ ), dual ion beam (Fe 2+ +He 2+ and Fe 2+ +H + ), and triple ion beam (Fe 2+ +He 2+ +H + ) irradiations on cavity evolution. A commercial alloy, F82H, a castable nanostructured alloy, CNA3, and a high purity model alloy, Fe8Cr2W, were irradiated at 400°C to 600°C to a damage level of 50 dpa at a damage rate of 1 × 10 –3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field transmission electron microscopy and high-angle annular dark-field scanning transmission electron microscopy was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium resulted in an increased cavity number density and maximum cavity size, producing an increase in swelling over that from helium injection alone. Swelling in F82H appears to peak between 450°C and 500°C. At 600°C, swelling was minimal and cavities of high density and small size were confined to grain boundaries and dislocations while at 400°C, swelling is also low with a nearly homogeneous, high density, distribution of very small cavities throughout. Swelling was least in the commercial alloy F82H due to the high sink strength. The CNA3 alloy underwent dissolution of precipitates that lowered the sink strength and resulted in higher swelling than F82H, but less than the model alloy. Electron energy loss spectroscopy (EELS) elemental mapping revealed hydrogen forming a halo-like structure about the periphery of the cavities and helium residing within the cavities themselves. Finally, this observation suggests that hydrogen reduces the surface energy of helium-filled cavities which results in both increased cavity number density and cavity size in triple beam irradiation over dual beam irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Evolution of Rev7 interactions in eukaryotic TLS DNA polymerase Polζ

Translesion synthesis (TLS) DNA polymerase Polζ is crucial for the bypass replication over sites of DNA damage. The Rev7 subunit of Polζ is a HORMA (Hop1, Rev7, Mad2) protein that facilitates recruitment of Polζ to the replication fork via interactions with the catalytic subunit Rev3 and the translesion synthesis scaffold protein Rev1. Human Rev7 (hRev7) interacts with two Rev7-binding motifs (RBMs) of hRev3 by a mechanism conserved among HORMA proteins whereby the safety-belt loop of hRev7 closes on the top of the ligand. The two copies of hRev7 tethered by the two hRev3-RBMs form a symmetric head-to-head dimer through the canonical HORMA dimerization interface. Recent cryo-EM structures reveal that Saccharomyces cerevisiae Polζ (scPolζ) also includes two copies of scRev7 bound to distinct regions of scRev3. Surprisingly, the HORMA dimerization interface is not conserved in scRev7, with the two scRev7 protomers forming an asymmetric head-to-tail dimer with a much smaller interface than the hRev7 dimer. Here, we validated the two adjacent RBM motifs in scRev3, which bind scRev7 with affinities that differ by two orders of magnitude and confirmed the 2:1 stoichiometry of the scRev7:Rev3 complex in solution. However, our biophysical studies reveal that scRev7 does not form dimers in solution either on its own accord or when tethered by the two RBMs in scRev3. These findings imply that the scRev7 dimer observed in the cryo-EM structures is induced by scRev7 interactions with other Polζ subunits and that Rev7 homodimerization via the HORMA interface is a mechanism that emerged later in evolution.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

FFT-based model for irradiated aggregate microstructures in concrete

The concrete biological shield of light water reactors is exposed to neutron and gamma irradiation throughout its lifetime, which results in the long-term degradation of the concrete’s mechanical properties. Under neutron irradiation, the concrete’s aggregates are subjected to radiation-induced volumetric expansion (RIVE), which strongly depends on the mineral content of the aggregate and exhibits the largest expansion in silicate-bearing minerals. In this work, the authors used the fast Fourier transform-based code Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in 2D to model the expansion of five different aggregates provided by the Japan Concrete Aging Management Program (JCAMP). Comparable rock specimens were irradiated at the JEEP-II test reactor. The model uses realistic aggregate microstructure reconstruction based on high-resolution characterization images. The model accounts for anisotropic RIVE, thermal expansion, and the associated initiation and propagation of damage. The RIVE models are calibrated based on expansion data in the literature. The authors assume that damage occurs exclusively at interfaces between the particles that compose an aggregate and that these interfaces also exhibit swelling. Using a micromechanical model, the evolution of Young’s modulus with RIVE is calculated for each aggregate and compared with Russian irradiation data. The modeled linear expansion agrees well with the experimentally measured expansion. Furthermore, the model also predicts that anisotropic RIVE and thermal expansion result in an earlier onset of damage with neutron fluence than in the isotropic case.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effects of internal swelling on residual elasticity of a quasi-brittle material through a composite sphere model

This work describes the development of a micromechanical-based constitutive model accounting for the effect of internal expansion on the residual elasticity of a Hashin composite material. This material is made of spherical inclusions that are subjected to gradual swelling within a quasi-brittle matrix. The main focus of this work is to describe and analyze the material mechanical response, with an additional focus on the internal swelling’s effect on the stress–strain response and residual elasticity. Microstructural features and parameters of major importance for the mechanical responses were identified. The innovative characteristics of the proposed approach are summarized as follows: (1) a full determination of the physics of a complete-damage problem throughout the whole process of inclusion swelling with upscaling techniques, which transfers the microcrack-related properties from the lower scale to upper scale; and (2) an evolution of the mechanical fields and the corresponding residual elasticity for various inclusion swelling levels. Concerning the matrix–inclusion composite, it was hypothesized that only the matrix was susceptible to cracking, with varied degrees of damage, whereas the inclusions behave elastically and the elastic modulus of the expanding inclusions remains constant. It is to emphasize that the gradual swelling of inclusions is modeled by an increasing strain in the current micromechanical-based constitutive model, and the microcracks are represented by a set of randomly oriented penny-shaped microcracks with identical radii (namely, closed cracks). The main contribution of the current research is to establish the exact mathematical solutions for the mechanical fields (stress, strain) caused by the swelling of the inclusions (mechanical loading) and derive the effective residual elasticity of a composite (structural response) subject to internal expansion and quasi-brittle damage. Based on the assumption of closed cracks that could be extended to open cracks in the upcoming work, the results proposed by the present paper help to understand the non-linear mechanical behavior of quasi-brittle materials subject to microcracking and provide a theoretical framework to be used as academical benchmark for numerical simulations.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Micromechanics of Fracture Propagation During Multistage Stress Relaxation and Creep in Brittle Rocks

Time-dependent rock deformation caused by the initiation and growth of fractures leads to the weakening of the rock mass. Understanding the fracturing mechanisms involved in the time-dependent behavior in brittle rocks is very important and to achieve this goal, a systematic series of three types of experiments was performed on double-flawed prismatic Barre granite specimens under unconfined compression. The first series aimed to identify the failure mechanism in the short-term failure mode under monotonic loading, the, second series involved multistage relaxation (constant strain) experiments to analyze the damage at different strain levels, and the third series explored the fracture propagation under multistage creep (constant load) experiments. The spatial and temporal evolution of cracking mechanisms were evaluated using the acoustic emission (AE) and two-dimensional digital image correlation (2D-DIC) techniques to observe the whole crack growth process as well as the accumulated inelastic strain at the specified region of interest. Results suggest that in the case of multistage creep experiments, the time to failure was less compared to the multistage relaxation, when loaded above the crack damage threshold (CD) estimated from the monotonic testing. Furthermore, the frequency magnitude distribution of the AE events generated in the three loading conditions followed the Gutenberg Richter model. A relatively lower b-value was obtained for the creep experiments, indicative of high energy AE events and faster crack growth. In addition, the AE and DIC results also revealed high evolution of tensile cracks at-different stages of creep and relaxation compared to shear and mixed-mode cracks.

58 GEOSCIENCES↗
DOE OSTI2021

Effect of stacking fault energy on damage microstructure in ion-irradiated CoCrFeNiMn concentrated solid solution alloys

In order to understand the effect of stacking fault energy on microstructural evolution in concentrated solid solution alloys, Kr 2+ ion irradiation was performed for FCC-type CoCrFeNiMnx alloys. During Kr 2+ ion irradiation at 500 degrees C the formation of black dots, self-interstitial atom faulted loops and perfect loops were observed, but no observable voids were found in all the alloys. The comparison of microstructural evolution among CoCrFeNiMn x (x = 0.7, 1, 1.3) alloys revealed less faulted loop number density and size in CoCrFeNiMn 1.3 . The stacking fault energy of each alloy was investigated using the deformed CoCrFeNiMn x alloys, and it was revealed that CoCrFeNiMn 1.3 had the highest stacking fault energy, which is higher value than that of 316SS. From these results, it is suggested that FeCrNiMn-based concentrated solid solution alloys could be well-designed by controlling the stacking fault energy with optimized Mn concentration as one of high radiation resistance structure materials. (C) 2020 Elsevier B.V. All rights reserved.

36 MATERIALS SCIENCE↗