Oxidation Resistant Graphite
Oxidation Resistant Graphite, current and prior research.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Irradiation test about oxidation-resistant graphite in WWR-K research reactor - 18533
Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). It is desirable to enhance oxidation resistance of graphite to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies, Toyo Tanso, Ibiden, Tokai Carbon and Nippon Techno-Carbon are studying to develop oxidation-resistant graphite for the in-core components. Neutron irradiation test with two capsules was carried out by WWR-K research reactor of Institute of Nuclear Physics of the Republic of Kazakhstan (INP) through the International Science and Technology Center (ISTC) partner project. The total irradiation cycles of WWR-K operation was 10 cycles and the total duration was 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E>0.18 MeV) for the gas-swept capsule irradiated at a central irradiation hole was preliminary calculated as 1.2x10{sup 25} m{sup -2}, and for the sealed capsule at a peripheral irradiation hole as 4.2x10{sup 24} m{sup -2}. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Oxidation and Material Property Effects on Graphite
Status of Graphite Oxidation Research Chronic Oxidation Oxidation of Irradiated Graphite Oxidation Resistant Graphite Material Property Changes with Oxidation
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE ART Graphite R&D Introduction
Introduction of ART Graphite R&D to include oxidation activities, oxidation resistant graphite, model development, ASME component failure, ASME code development (design rules), ceramic composites, AGC update, molten salt intrusion, split-disk studies, wear testing, and concluding remarks.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Oxidation Activities
Slides for Oxidation Activities, oxidation rate, penetration/lathing, and strength after oxidation work. Also includes recent and current oxidation studies, background for rate behavior, strength behavior, density profile, and penetration measurements, oxidation resistant graphite coating development, and irradiated and unirradiated graphite oxidation rate behavior.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Determining the oxidation behavior of matrix graphite
This work presents the oxidation behavior of matrix graphite in air. Matrix graphite, graphite powder/flakes bonded by a small amount of non-graphitic carbon, surrounds coated fuel particles in order to form cylindrical fuel compacts (in prismatic core designs) or spheres (in pebble-bed reactor designs). This work focuses on oxidation tests conducted on two matrix graphite materials, one provided by Kairos Power and the other A3 matrix graphite. Some of the tests followed American Society for Testing and Materials (ASTM) oxidation testing standards using a vertical furnace system and others were performed in a thermogravimetric analyzer (TGA). It was determined that, at temperatures of 450 °C–700 °C, the oxidation rate of the Kairos matrix graphite follows the Arrhenius equation. In comparison with A3 matrix graphite, the Kairos matrix graphite shows better oxidation resistance at high temperatures (≥550 °C), but also a higher oxidation rate at low temperatures. Both the A3 matrix graphite and the Kairos matrix graphite materials may experience preferential oxidation of the partially graphitized binder. An oxygen penetration gradient was also observed when using the three characterization methods (i.e., optical microscope, x-ray tomography [XCT], and density profile by the lathe) enlisted in this research. In conclusion, the oxygen penetration depth increases with decreasing isothermal oxidation temperature, while the center of the oxidized samples (10% weight loss) remains almost untouched even at 500 °C.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Experimental study on kinetic oxidation of graphite IG-110 by steam
Graphite is proposed for use in High-temperature Gas-cooled Reactors (HTGRs) as the fuel matrix, neutron moderator/reflector, and core structural material. One important property of nuclear grade graphite is their resistance to oxidation in high-temperature environment. Extensive investigation has been performed in the literature for graphite oxidation by air. However, available experimental data are still limited for graphite oxidation by steam under conditions comparable to a postulated steam ingress accident in HTGRs. In this study, the oxidation rate of graphite IG-110 by steam was measured at temperatures from 850 to 1100 °C with the steam partial pressure varying from 0.5 to 20.0 kPa and the hydrogen partial pressure varying from 0 to 2.0 kPa. Further analysis confirms the oxidation process in this present study is dominated by the chemical kinetics, which lends credit to the data for being used to develop numerical models. It was observed that the increase of the kinetic oxidation rate with the steam partial pressure tends to become less apparent if the steam partial pressure keeps increasing. In addition, it was found that the partitioning of hydrogen inhibits the graphite-steam reaction process even with the steam partial pressure up to 20.0 kPa. However, this inhibiting effect starts to become saturated when the hydrogen partial pressure exceeds 1.0 kPa. The oxidation rates were fitted to the conventional Langmuir-Hinshelwood (LH) and Boltzmann-enhanced Langmuir-Hinshelwood (BLH) models by a multivariable optimization algorithm. The BLH model exhibits a better accuracy than the LH model within the specified experimental conditions. The predicted oxidation rate using the BLH model shows a mean relative difference of about 24% with the maximum difference of about 55% when compared with our experimental data.
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Computational Modeling of Molten Salt Infiltration and Oxidation in Nuclear Graphite
Graphite is utilized as a moderator and reflector in advanced nuclear reactor designs due to its high thermal conductivity, neutron moderation properties, and resistance to radiation damage. However, its longterm performance and reliability are challenged by degradation mechanisms such as molten salt infiltration in molten salt reactors (MSRs) and oxidation in gas-cooled reactors (GCRs). These mechanisms can compromise the structural integrity and operational lifetime of graphite components, necessitating a more detailed assessment of their physical behavior. This report focuses on the development of computational models for molten salt infiltration and oxidation of graphite to aid the design and performance analysis of graphite components. For molten salt infiltration, a computational framework is developed that couples incompressible Navier-Stokes and phase-field model to simulate the penetration of molten salt into graphite?s interconnected pore structure. Initial model verification is performed using two-phase flows in two dimensions, demonstrating the models ability to capture fundamental physical behavior and agree with analytical solution. This framework is then applied to a realistic IG110 nuclear graphite , where a computed tomography extracted pore geometry is used to analyse the infiltration behavior of FLiNaK molten salt. This model provides insights into how the microstructure and other relevant parameters influence the transport pathways of molten salt into graphite, potentially offering a means to rapidly evaluate a graphite grade?s resistance to infiltration. For oxidation, the report details pore-scale mass and heat transport models, describing the diffusion of gases, reaction kinetics, and thermal effects. Additionally, this report highlights inconsistencies in the existing volume-averaged macroscopic model, particularly in upscaling of reaction kinetics and flux terms, and surface to volume transformations. These inconsistencies suggest that current formulations may not accurately capture the experimentally observed graphite oxidation process, highlighting the need for improved model development. This work advances the development of physics-based computational models for graphite degradation, contributing to improved predictive models for next-generation nuclear reactor designs. Future efforts will focus on refining the infiltration model to address non-physical behaviors and enhance its robustness. Additionally, for oxidation, further studies will employ the principles of volume averaging to rigorously derive the upscaled equations, potentially in collaboration with subject matter experts.
Computational Modeling of Molten Salt Infiltration↗
Nanoscale electrostatic control in ultraclean van der Waals heterostructures by local anodic oxidation of graphite gates
In an all-van der Waals heterostructure, the active layer, gate dielectrics and gate electrodes are assembled from two-dimensional crystals that have a low density of atomic defects. This design allows two-dimensional electron systems with very low disorder to be created, particularly in heterostructures where the active layer also has intrinsically low disorder, such as crystalline graphene layers or metal dichalcogenide heterobilayers. A key missing ingredient has been nanoscale electrostatic control, with existing methods for fabricated local gates typically introducing unwanted contamination. Here we describe a resist-free local anodic oxidation process for patterning sub-100 nm features in graphite gates, and their subsequent integration into an all-van der Waals heterostructure. We define a quantum point contact in the fractional quantum Hall regime as a benchmark device and observe signatures of chiral Luttinger liquid behaviour, indicating an absence of extrinsic scattering centres in the vicinity of the point contact. In the integer quantum Hall regime, we demonstrate in situ control of the edge confinement potential, a key requirement for the precision control of chiral edge states. In conclusion, this technique may enable the fabrication of devices capable of single anyon control and coherent edge-state interferometry in the fractional quantum Hall regime.
36 MATERIALS SCIENCE↗
Electrically Conductive Amine Functionalized Reduced Graphite Oxide Foam for CO 2 Removal from the Air
Rapid regeneration of CO 2 adsorbents is critical to improving the productivity of direct air capture (DAC) systems. In this study, we codesigned a material to have appropriate electrical conductivity and CO 2 adsorption properties to enable efficient CO 2 capture from air. Specifically, we present a poly(ethylenimine) (PEI)-impregnated thermally annealed graphite oxide (TAGO900) foam adsorbent tailored for vacuum-assisted electrically driven thermal swing adsorption (V-ETSA). This structured adsorbent leverages the high electrical conductivity of the reduced graphite oxide framework to enable fast and direct heating of the adsorbent material by electrical resistance heating (Joule heating). An optimal sample, 40 wt % PEI (molecular weight 25k) impregnated TAGO900, shows the best balance between adsorption capacity (1.54 mmol g –1 ) and adsorption rates (0.016 mmol g –1 min –1 ) using fixed bed breakthrough experiments at 25 °C and 70% RH using 50 sccm 400 ppm of CO 2 /N 2 flow. Compared to conventional temperature vacuum swing adsorption (TVSA), the V-ETSA approach achieves substantially faster CO 2 desorption, achieving average desorption rates (including cooling time) of 0.09 mmol g –1 min –1 ─approximately 2.5 times faster than TVSA under similar operating conditions. The maximum desorption rate reaches 0.23 mmol/g/min during the desorption stage. These results underscore the importance of the direct heating strategy, such as Joule heating, for fast and highly productive vacuum swing adsorption in DAC systems.
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Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength
Matrix graphite (MG) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 MG at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized MG specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of MG. The transition temperature between Regimes I and II is ~700°C and the activation energy ( E a ) in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates MG is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of MG than oxidation at 900°C. Comparing with the strength of pristine MG specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. Microstructure images of SEM and porosity measurement by Mercury Porosimetry indicate that the significant compressive strength loss of MG oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Synthesis and Thermal Oxidation Resistance of Boron-Rich Boron–Carbide Material
A boron-rich boron–carbide material (B4+δC) was synthesized by spark plasma sintering of a ball-milled mixture of high-purity boron powder and graphitic carbon at a pressure of 7 MPa and a temperature of 1930 °C. This high-pressure, high-temperature synthesized material was recovered and characterized by X-ray diffraction, X-ray photoelectron spectroscopy, Raman spectroscopy, Vickers hardness measurements, and thermal oxidation studies. The X-ray diffraction studies revealed a single-phase rhombohedral structure (space group R-3m) with lattice parameters in hexagonal representation as a = 5.609 ± 0.007 Å and c = 12.082 ± 0.02 Å. The experimental lattice parameters result in a value of δ = 0.55, or the composition of the synthesized compound as B4.55C. The high-resolution scans of boron binding energy reveal the existence of a B-C bond at 188.5 eV. Raman spectroscopy reveals the existence of a 386 cm−1 vibrational mode representative of C-B-B linear chain formation due to excess boron in the lattice. The measured Vickers microhardness at a load of 200 gf shows a high hardness value of 33.8 ± 2.3 GPa. Thermal gravimetric studies on B4.55C were conducted at a temperature of 1300 °C in a compressed dry air environment, and its behavior is compared to other high-temperature ceramic materials such as high-entropy transition metal boride. The high neutron absorption cross section, high melting point, high mechanical strength, and thermal oxidation resistance make this material ideal for applications in extreme environments.
36 MATERIALS SCIENCE↗
Preparation of Nano-SiO{sub 2}-Coated Graphite Films by a Laser-Assisted Sol–Gel Process
Here, we prepared nano-SiO{sub 2}-coated graphite films using a novel laser-assisted sol–gel method. The SiO{sub 2} sol–gel prepositioned on the graphite substrate was irradiated by the fiber laser at different energy densities. The growth mechanism of the coatings under laser irradiation was studied and the effects of processing parameters on the coating structure were systematically investigated. In addition, the high-temperature oxidation resistance of these samples was examined. Surface morphology and elemental composition were examined using scanning electron microscopy and energy-dispersive spectrometer. Phase constituents and microstructure were identified by x-ray diffraction and transmission electron microscopy. The oxidation resistance was analyzed by thermal gravimetric analysis. The results found that laser irradiation significantly modifies the sol–gel-deposited nano-SiO{sub 2} layer. A uniform and compact nano-SiO{sub 2} coating was obtained when irradiated by laser energy density of 14.4 kJ/cm{sup 2}. Optimum high-temperature oxidation resistance was achieved for the two-layer SiO{sub 2}-coated films.
36 MATERIALS SCIENCE↗
Oxidation Behavior and Property Degradation of Nuclear Graphites
During its multidecade operation in the core of nuclear reactors, graphite components are subjected to aggressive and continuous exposure to a high field of ionizing and neutron irradiation, high temperature, and various types of present and postulated chemical attacks. High density, high crystallinity polygranular synthetic graphite is unique among other materials for its extraordinary capacity of resisting and adapting to the aggression inflicted by high temperature, high energy neutron bombardment and ionizing gamma radiation. But, as a carbonaceous material, even though of very high purity, graphite is reactive towards common oxidizing agents: oxygen, carbon dioxide, water. Safe operation of HTGRs relies, among other aspects, on engineered safeguard systems for efficient and continuous protection of graphite components against oxidation. Graphite oxidation behavior was, and continues to be, an important direction of theoretical and experimental research, engineering analyses, models and simulations, and design and safety regulations. The avalanche of publications, reports, experimental data, computer codes, and regulatory documents related to oxidation behavior of nuclear graphite is now accelerating to new levels, prompted by the increased interest for nuclear energy as a clean, carbon-free energy source. Even though public’s perception of nuclear energy advantages may still be influenced by the memories of past accidents of nuclear reactors from generations II and III, the community of informed scientists and engineers, regulators and statemen knows that generation IV of nuclear reactors is designed at very high safety standards, doubled by great advances of scientific knowledge and technological progress. One of routes of these recent advances is directed at better understanding of graphite oxidation behavior, its relationship with graphite manufacturing and microstructural properties, along with the effects of various environmental factors and process variables. Together, the recent progress in manufacturing, properties characterization, and modeling of intricated physical and chemical processes that concur to the oxidation behavior led to development of powerful simulation codes able to analyze various scenarios of normal operation and hypothetical off-normal events, and thus to clearly specify the allowable parameters envelopes for the designers, constructors, and operators of current and future modular HTGRs. This review begins with an introduction on manufacturing methods, structure, and properties of nuclear graphite, including basic requirements that this specialty graphite type must satisfy for nuclear use. It continues with a chapter on environmental effects on nuclear graphite, where emphasis is placed less on irradiation and much more on oxidation phenomena, their safety implications, and the basic traits of chronic and acute oxidation by air (oxygen) and water (humidity, steam). Particular attention is placed on the three graphite grades of interest for this document (IG-110, NBG-18, PCEA). A chapter on properties degradation induced by oxidation follows, with focus on density, dimensional, and mechanical properties changes. The next chapter is intended as a brief review of various approaches used for modeling of graphite oxidation behavior. It summarizes the progress of oxidation models, from the early attempts to complex computational approaches interfaced with specialized computer codes designed for nuclear reactor simulations. Last, a list is presented of knowledge gaps where more research is needed. A short summary concludes the review.
36 MATERIALS SCIENCE↗
Mn versus Al in Layered Oxide Cathodes in Lithium-Ion Batteries: A Comprehensive Evaluation on Long-Term Cyclability
Nickel-rich layered oxide cathodes with the composition LiNi 1-x-y Co x Mn y O 2 (NCM, (1-x-y) ≥ 0.6) are under intense scrutiny recently to contend with commercial LiNi 0.8 Co 0.15 Al 0.05 O 2 (NCA) for high-energy-density batteries for electric vehicles. However, a comprehensive assessment of their electrochemical durability is currently lacking. Herein, two in-house cathodes, LiNi 0.8 Co 0.15 Al 0.05 O 2 and LiNi 0.7 Co 0.15 Mn 0.15 O 2 , are investigated in a high-voltage graphite full cell over 1500 charge-discharge cycles (≈5–10 year service life in vehicles). Despite a lower nickel content, NCM shows more performance deterioration than NCA. Critical underlying degradation processes, including chemical, structural, and mechanical aspects, are analyzed via an arsenal of characterization techniques. Overall, Mn substitution appears far less effective than Al in suppressing active mass dissolution and irreversible phase transitions of the layered oxide cathodes. The active mass dissolution (and crossover) accelerates capacity decline with sustained parasitic reactions on the graphite anode, while the phase transitions are primarily responsible for cell resistance increase and voltage fade. With Al doping, on the other hand, secondary particle pulverization is the more limiting factor for long-term cyclability compared to Mn. Finally, these results establish a fundamental guideline for designing high-performing Ni-rich NCM cathodes as a compelling alternative to NCA and other compositions for electric vehicle applications.
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Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis
Molten-salt reactors (MSRs) are Generation IV nuclear reactors that use liquid salt as a coolant and/or fuel. In several MSR designs, graphite serves as a moderator and/or reflector. However, due to limited experimental data and operational experience, our understanding of graphite behavior in molten salt environments remains incomplete. This report aims to identify the degradation mechanisms of nuclear graphite in MSRs, detail the mechanisms of each factor, and provide an initial assessment of their impact on the structural integrity of graphite components. This assessment is based on an extensive literature review and insights from subject matter experts. Furthermore, given the limited data, a modeling strategy using existing Grizzly software is proposed for a more thorough analysis where appropriate. Additionally, it presents mitigation strategies where applicable. The report covers physical degradation mechanisms such as infiltration, erosion, and abrasion, as well as chemical degradation mechanisms including fluorination, intercalation, corrosion, and oxidation. Molten salt can infiltrate the porous structure of graphite, leading to several detrimental effects. Entrapment of fissile products within the graphite pores can cause radiation damage and could pose challenges in the handling and disposal of contaminated components. The differential thermal expansion between the infiltrated salt and graphite, along with internal stress from pressurized molten salt and volumetric heating, can compromise the structural integrity of graphite. To mitigate these effects, employing ultra-fine graphite grades and applying sealants and coatings are effective strategies. A computational model based on coupled solid mechanics and heat transfer phenomena could be used to predict the internal stresses using Grizzly software. In pebble-bed MSRs, graphite fuel pebbles can cause abrasion against reactor components due to friction and wear. The severity of wear is influenced by various factors such as temperature, environment, and the presence of lubricants. Tribological studies reveal that higher temperatures and molten salt environments, such as FLiBe, significantly reduce wear rates compared to dry conditions. Additionally, the chemical composition of the salt can further optimize graphite's tribological performance. Long-term wear effects can be modeled by incorporating surface defects into the geometry and predict stresses under thermal and radiation effects using Grizzly software. Chemical degradation of graphite in a molten salt environment can occur through fluorination and intercalation. Fluorination can occur via replacement of hydrogen or oxygen atoms, or at the active sites, but does not cause structural degradation. Intercalation, on the other hand, can lead to exfoliation, where layers of graphite separate and peel away, damaging the graphite. Protective coatings can enhance graphite's resistance to intercalation. Graphite generally exhibits good chemical stability in molten salt environments, though it can corrode under specific conditions, particularly in the presence of impurities or oxidants. Studies have shown that protective coatings, such as plasma-sprayed partially stabilized zirconia (PSZ), can effectively prevent such degradation. Corrosion behavior varies significantly with different graphite grades and coating applications, underscoring the need for detailed studies on uncoated and coated graphite to understand and mitigate corrosion mechanisms in MSRs. Research indicates that the presence of oxidants and impurities can accelerate graphite degradation in molten salts, making it essential to explore acceptable impurity limits. Oxidation is another critical degradation mechanism, leading to weight loss and structural damage due to the formation of CO and CO 2 from the reaction of carbon atoms with oxygen. This process creates new porosity and compromises graphite's integrity. While extensive research on graphite oxidation has been conducted for gas-cooled reactors, studies specific to MSRs are limited. Findings from the coal industry suggest that molten alkali metal salts can significantly accelerate graphite oxidation, a hypothesis worth exploring for fluoride salts in MSRs. Understanding oxidation behavior in MSRs is vital for developing protective measures. The analysis of post-irradiated graphite from the MSRE experiment demonstrated exceptional chemical compatibility with molten fluoride salt, suggesting that the extent of chemical attack on graphite largely depends on the salt's infiltration capability. Therefore, the use of ultra-fine grade graphite could help mitigate chemical degradation effects. Existing oxidation modeling capabilities in Grizzly, which use reaction-diffusion equations to model graphite-air interactions, could be adapted to simulate the chemical degradation effects of graphite in molten salt environments.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Surface normal emissivity measurements in uncoated and oxidized B-coated plasma facing component materials in the long wavelength infrared spectrum
Accurate surface emissivity measurements are integral to the fidelity of the infrared thermography diagnostic evaluation in fusion reactors. The emissivities of ATJ TM graphite, Sigrafine® (R6510) graphite, and TZM alloy were measured as a function of temperature in accordance with the Contact Thermometer Method in the ASTM E1933 standard. Samples were heated resistively in a high-vacuum chamber, and the surface temperature was monitored using a surface thermocouple and a Telops long-wavelength infrared camera. The surfaces of Sigrafine® graphite and TZM alloy samples were coated with 10 nm and 20 nm layers of oxidized boron and were also measured in the uncoated condition. The thickness was assessed from measurement by a quartz crystal microbalance (QCM). ATJ TM graphite was uncoated. The measured emissivity of ATJ TM Graphite ranged from 0.82 to 0.83, uncoated Sigrafine® graphite ranged from 0.76 to 0.98, and uncoated TZM alloy ranged from 0.11 to 0.13. Oxidized boron coatings increased the emissivity of the TZM alloy to 0.18–0.23 but decreased the emissivity of Sigrafine® graphite to 0.56–0.66 for a 10 nm coating and to 0.65–0.74 for a 20 nm coating. In contrast to the expected blackbody radiance, the emissivity of uncoated Sigrafine® graphite and TZM alloy did not monotonically increase with temperature from 100 to 500 ℃.
Emissivity↗
Control of High-Temperature Static and Transient Thermomechanical Behavior of SiMo Ductile Iron by Al Alloying
Abstract Silicon and molybdenum ( SiMo ) ductile iron is commonly used for exhaust manifolds because these components experience thermal cycling in oxidizing environment, which requires resistance to fatigue during transient thermomechanical loads. Previous studies have demonstrated that alloying elements, such as Al , to SiMo ductile iron reduces the amount of surface degradation during static high-temperature exposure. However, deterioration of sphericity of the graphite nodules and a decrease in ductility could affect the tendency of cracking during thermal cycling. In this article, the effect of Al alloying on static and transient thermomechanical behavior of SiMo ductile iron was investigated to optimize the amount of Al alloying. A thermodynamic approach was used to confirm the effect of the Al alloying on the phase transformations in two SiMo cast irons, alloyed by 1.8% Al and 3% Al . These two alloys were cast in a laboratory along with the baseline SiMo ductile iron. Several experimental methods were used to evaluate the dimensional stability, physical properties, static oxidation, and failure resistance during constrained thermal cycling testing to compare their high-temperature capability. Experimental results verified that Al alloying increases the temperature range and decreases volume change during eutectoid transformation, which together with enhancement of oxidation protection improved the dimensional stability. Thermocycling tests showed that the number of cycles to failure depends on the amount of Al alloying and the applied high-temperature exposure during each cycle. SEM/EDX, high-resolution TEM and µCT analysis were used to verify the mechanism resulting from the Al alloying protection. It was shown that an optimal level of Al alloying for balancing oxidation and thermal cracking resistance depends on thermomechanical conditions of application.
36 MATERIALS SCIENCE↗