Structural Materials For Fusion Reactors Research Articles

The temperature effect on the formation and growth of helium-vacancy clusters in bcc iron: A comprehensive molecular dynamics simulation

Bubble generation can significantly degrade the performance of fusion reactor structural materials. In this work, the temperature effects on the initial clustering of helium (He) atoms, the formation of helium-vacancy clusters (HenVm) and the subsequent growth of the clusters into bubbles in bcc iron under different conditions were simulated by molecular dynamics. The results show that He behaviors are primarily dominated by coalescence at temperature range from 400 K to 500 K, which leads to a preference for He to bind together. Conversely, when temperature is higher than 500 K, dissociation become more prominent, causing large He clusters to be hard to form. Under the conditions of increasing He concentration and irradiation dose, the higher the temperature, the higher the density of HenVm clusters at He concentration around 100 appm. However, the temperature dependence reversed at high He concentration. The detailed analysis of formation cases revealed the percentage of nucleation sites for He bubbles at defects during He segregation. In addition, there is no significant difference among the He/V ratio of HenVm clusters formed at vacancies, interstitial dislocation loops (IDLs) and interstitial sites during the cluster growth. However, the He/V ratio of He bubbles growing on edge dislocation (ED) is overall lower than those growing at other formation sites, since the growth of HenVm clusters on ED requires a lower pressure to kick out self-interstitial atoms (SIAs) or SIA clusters. The average He/V ratio of He bubbles in bulk decreases as temperature rises from 400 K to 800 K, while that of He bubbles growing on ED is independent of temperature.

Superior radiation resistance of ODS-RAFM steels revealed by Positron annihilation spectroscopy study

Oxide-dispersion-strengthened (ODS) reduced-activation ferritic/martensitic (RAFM) steels are considered as most promising structural materials for fusion reactors and other advanced nuclear systems due to its excellent irradiation resistance under high dose irradiations. However, the extreme complex synergistic effect between transmuted helium/hydrogen and displacement damage which most seriously affects the structure and mechanical properties of ODS-RAFM steels is very difficult to understand, especially the role of hydrogen still remains unclear. Taking advantage of very sensitive to small vacancy clusters of Positron annihilation spectroscopy (PAS) method, we conducted irradiation defect study on ODS-RAFM, RAFM and its corresponding model alloy Fe–9Cr and α-Fe to investigate their anti-swelling properties at the early stage of irradiation with hydrogen-helium synergism and the mechanism. Three distinct irradiation schemes including (1) single Fe ion beam, (2) simultaneous Fe and He (named Fe + He) ion beam, (3) subsequent H ion injection after Fe + He (named Fe + He/H) ion beam irradiation were performed. Hydrogen-helium synergistic effects on vacancy evolution within these steels were explored combining the first-principles calculation. The implantation of He was observed to significantly increase defect concentration among all four steels, while H exhibited interestingly complexity on affecting the evolution of helium bubbles. The H post-injection increased defect concentration obviously in the materials with abundant sinks, but improved poorly in ones with rare sinks. Through the systematic study of the interactions between H, He, and vacancies, it was revealed that even at the early irradiation stage, ODS-RAFM steels has already exhibited excellent irradiation resistance compared to other alloys across three irradiation schemes.

Evolution of interfacial microstructure and oxides of CLAM steel by hot compression bonding

The development of the joining process is essential for the application of China low activation martensitic (CLAM) steel as the structural material for fusion reactors. In the present study, the hot compression bonding (HCB) for CLAM steel was carried out under different parameters. The effects of different strain rates, temperatures, strains and holding time on the evolution of interfacial microstructure and oxides were investigated. Compressed at 1000 °C with 5% deformation, some micron-size continuous oxides distributed along the flat interface. The interfacial oxides were broken and decomposed as the deformation increased to 20%, their average size was reduced to 400 nm. With the bonding temperature increasing, more interfacial grain boundaries (IGBs) could cross the original bonding interface and flat interface became wavy. Compressed at 1000 °C/20%, the interfacial oxides consisted of fcc-structured spinel MnCr2O4 and enriched in Si and Ta oxides. The MnCr2O4 oxides would decompose preferentially after 0.5 h holding treatment at 1100 °C. Only a small number of oxides rich in Si and Ta elements remained at the original bonding area after 2 h holding treatment. Compressed at 1100 °C/20%, the bonding interface was completely replaced by the migrated grains after 2 h holding treatment and no large-size oxides could be observed. Moreover, the tensile strength and elongation at room temperature of the joint were comparable to those of the base material.

Liquid lithium corrosion of SiC/SiC composites

Silicon carbide fibre reinforced silicon carbide matrix composite (SiC/SiC) is a key structural material for fusion reactors to allow high temperature operation of liquid lithium breeder blankets. This research investigated the corrosion behaviour of two SiC/SiC composites (using Tyranno SA3 and SA4 fibres) immersed in static liquid lithium at 600 °C for 100 h. Utilizing immersion tests and advanced microstructural analysis, the study reveals considerable corrosion in SiC/SiC composites, particularly in local areas enriched with residual carbon or oxygen. The interface carbon layer, while enhancing mechanical properties, induces preferential sites for corrosion cracking, diminishing the material's corrosion resistance. This study offers critical insights into the corrosive interaction between SiC/SiC composites and liquid lithium, highlighting the need for manufacturing process optimization and exploration of alternative interface layers, protective coatings, or avoiding contact between SiC/SiC and molten lithium in breeder blanket system.

Impact of trace silicon on irradiation hardening and embrittlement of RAFM steel subjected to neutron irradiation

The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required. The additional refining processes significantly increase the cost of steel fabrication. Currently, the role of residual or trace Si in neutron irradiation of steel remains unclear. In the presented study, two 9Cr-2 W RAFM steels with Si contents of 0.1 wt.% and 0.01 wt.% were fabricated and subjected to a neutron irradiation at 563 K with a neutron fluence of 4.8 × 1023m−2. The main results show that 0.1 wt.% Si can contribute to the suppression of irradiation hardening and embrittlement, as evidenced by the reduction in irradiation-increased hardness and yield strength, while maintaining preferable elongation. Microscopic analysis suggests that an appropriate content of Si helps refine the grains in RAFM steel, increases the ductile fracture region of the fracture surface, and effectively reduces the density of dislocation loops induced by neutron irradiation. The findings evidence that retaining a certain amount of trace Si should be an informed and rational choice in the design of RAFM steel.

Determination of the Surface-Rate Coefficients of Deuterium Through Nickel by Gas-Driven Permeation Experiments in Surface-Limited Regime

Tritium permeation through structural materials of fusion reactors is an important issue from both safety and tritium self-sufficiency points of view. Mutual influences among hydrogen isotopes are also important in assessments of tritium permeation fluxes, and although some theoretical and experimental works have been carried out so far—which have all focused on steady-state permeation—their conclusions are insufficiently clear about how one isotope affects the permeation of another. This was motivation to further investigate this problem, both theoretical and experimental in a non-steady-state approach and also in the surface-limited regime (SLR) of permeation. After initial theoretical work, a dedicated experimental installation for gas-driven permeation experiments was designed and assembled. Then, initial experimental work was completed for testing monoisotope permeation of deuterium through a very thin (0.075-mm) nickel membrane in the temperature range of 473 to 773 K and for a driving pressure of deuterium gas in the range of 10−2 to 1 Pa. This work was done in preparation for subsequent multi-isotope experiments, with the proposed goal to observe how hydrogen affects the permeation of deuterium. The main objectives of the work were to confirm experimentally the achievement of the SLR of permeation for the selected conditions of testing; to determine experimentally the kinetic coefficients of surface transport for deuterium in nickel; and to test and validate the experimental setup, procedures, and methods used and their reliability to more stringent requirements of the multi-isotope experiments. Within this paper, the experimental setup and all the operating procedures used both in calibration operations and in permeation experiments are presented in detail, as well as their results. The obtained results confirmed that permeation occurred in the SLR in the tested range of pressure for each testing temperature. The surface-rate coefficients of dissociation and recombination were both determined experimentally. The values obtained for the dissociation coefficient were in very good agreement with other similar experimental data available in the literature. For the recombination coefficient, agreement was not quite satisfactory, but comparison could be made with values calculated based only on the dissociation and solubility coefficients, as data for these coefficients for the nickel-hydrogen system (determined directly from gas-driven permeation experiments) were not found in the literature. However, these results indicated good reliability of the installation, its calibration, and operating procedures in order to proceed to experimental testing of multi-isotope permeation.

Microstructural changes in He-irradiated V-Cr-Ti alloys with low Ti addition at 700 °C

V-Cr-Ti alloys with reduced Ti contents are expected to be good candidates as recyclable structural materials for fusion reactors. Herein, He irradiation was applied to V-4Cr-xTi (x = 0 to 4) alloys to investigate the additional effect of Ti and gas interstitial impurities on their microstructural evolution and irradiation hardening. Following He irradiation of the specimens at 700 °C with a maximum damage depth of 0.5 dpa, transmission electron microscopy and nanoindentation hardness tests were conducted. The microstructural evolution of the titanium-oxycarbonitride (Ti(CON)) precipitates formed during He ion irradiation was observed. The behavior of irradiation hardening is explained by the formation process of Ti(CON) precipitates and the magnitude of irradiation hardening depends on the amount of gas interstitial impurities that contribute to precipitate formation.

Accelerated design of low-activation high entropy alloys with desired phase and property by machine learning

Low-activation high-entropy alloys (HEAs) have been regarded as novel candidate structural materials for fusion reactors due to their excellent mechanical and radiation resistant properties. Nevertheless, the potential vast composition space brings a prominent challenge in the design of low-activation HEAs. Herein, a new strategy based on machine learning (ML) was proposed to accelerate the exploitation of low-activation HEAs with desired phase and property. Two optimized classification models with accuracy of > 85% were developed to identify single-phase body-centered-cubic (BCC) solid-solution (SS) HEAs. One regression model with correlation coefficient (R) of > 0.9 was constructed to predict the hardness of HEAs. The phase and hardness prediction models were combined in accordance with an integrated design strategy to identify the low-activation HEAs with desired phase and property from 284,634 candidates. With this design strategy, a new desired single-phase BCC Fe35Cr30V20Mn10Ti5 low-activation HEA with 555.9 ± 15.3 HV was designed and fabricated via only two experimental iterations. Besides, two new phase selection rules with accuracy of > 95% were given to efficiently screen SS and BCC HEAs, respectively. Our research work not only provides new ideas to search for low-activation HEAs, but also can be extended to the integrated design of structure and performance of other advanced materials.

Effect of tungsten doping on the irradiation resistance of FeCrV-based refractory medium entropy alloy for potential nuclear applications

Given the increasing demand for energy and the rising cost of fossil fuels, the development of nuclear fission and fusion reactors has become a key priority. Recently, High Entropy Alloys (HEAs) and Medium Entropy Alloys (MEAs) have emerged as promising materials for use in nuclear reactors due to their unique properties, which include significant lattice distortion, high configurational entropy, reduced diffusivity, and the cocktail effect. In this study, we designed and analysed a series of FeCrV-based Refractory Medium-entropy Alloys (RMEAs) as structural materials in fusion reactors, using Density Functional Theory. By calculating the empirical parameters, we have verified the solid solution's structure stability of the FeCrV-based RMEA. We examined the mechanical, electronic, and irradiation resistance properties of these alloys and found that the incorporation of tungsten dopants could enhance both the mechanical properties and irradiation resistance of the materials, with the optimal doping level being approximately 8% of Tungsten. Furthermore, electronic structure calculations were performed to elucidate the underlying mechanisms. This knowledge can guide the development of new HEA/MEA materials for use in nuclear reactors.

First-principles studies of behavior of hydrogen and mechanical properties of TiCrTaV high-entropy alloys

This study investigates the dissolution and diffusion of H atom in TiCrTaV of two body-centered cubic (BCC) high-entropy alloys and their effects on the mechanical properties and electronic structure of the alloys by first-principles calculations. Results show that the solution energy of H atom and local lattice distortion vary depending on the local environment. Meanwhile, the site occupancy of H atoms in tetrahedral interstitials are slightly advantageous in the two structures. Diffusion of H also presents a sluggish effect. The dramatic increases and decreases of diffusion barrier disrupted the periodicity and symmetry of H diffusion in the two BCC structures, thereby slowing down the H accumulation. After hydrogenation, TiCrTaV high-entropy alloys (HEAs) still maintain resistance to deformation, indicating that they are the potential candidate structural materials for fusion reactors. Our investigation indicates that this low-activation HEA may exhibit promising irradiation resistance due to its special H behavior.

Experimental results of deuterium/hydrogen co-current permeation through Nickel membrane, in non-steady-state surface limited regime

A better understanding of mutual influence among hydrogen isotope species in multi-isotope permeation is important for accurate assessments and predictions on tritium permeation through structural materials of fusion reactors. Moreover, this is important as the injection of H2 gas into the helium coolant of HCPB breeding blanket is proposed for mitigation of the permeation flux of tritium from the purge gas into the coolant. Although various modeling and experimental works have been carried out so far on this topic of multi-isotope permeation, unresolved issues or even controversies still persist. This work presents the results of some experimental tests of co-current permeation of deuterium with hydrogen through a very thin (0.075 mm) nickel membrane, performed in non-steady-state and surface limited regime (SLR), with the purposes to observe the effect of hydrogen on the permeation effective flux of deuterium, as well as to verify experimentally the predictions of a previous theoretical work. Thus, two deuterium/hydrogen gas mixtures of different ratios, as well as pure deuterium gas, were alternately introduced into the high-pressure side of the membrane by keeping the effective partial pressure of deuterium as close as possible to the same value (0.1 Pa), so that the comparison between multi- and mono-isotope permeation be relevant. The results indicate that the presence of (co-permeation) hydrogen does not lead to the decrease of deuterium permeation, but, on the contrary, to its increase. This effect is clearer revealed well before steady state is reached, where the effective permeation flux of deuterium has a constantly higher rise-rate when (co-permeation) hydrogen is present, and this rise-rate is even higher when the effective hydrogen pressure is higher. Also, a quite good consistency between experimental results and the theoretical predictions has been obtained, as well as a good repeatability of the results.

Heterogeneous microstructure and anisotropic mechanical properties of reduced activation ferritic/martensitic steel fabricated by wire arc additive manufacturing

• Reduced activation ferritic/martensitic steel was fabricated by wire arc additive manufacturing. • The as-deposited microstructure exhibited the heterogeneous band characteristics. • Microstructure differences caused the periodical hardness and anisotropic tensile properties. • A potentially detrimental softening zone, was found at the heat-affected zone of the as-deposited RAFM steel. Reduced activation ferritic/martensitic (RAFM) steel is an iron-based alloy as a candidate structural material in fusion reactor. This paper evaluates the compatibility of RAFM steel as a choice material for wire arc additive manufacturing (WAAM). Two specimens of RAFM steels with high and low heat input were fabricated by WAAM. The effects of the heat input on the microstructure, microhardness and tensile properties of samples were investigated. The fusion boundaries are spaced uniformly in the whole sample. Three distinctive zones were present in the periodic region, including heat-affected zone (HAZ), columnar grains zone and fine-grained zone occurred alternatively. The HAZ was affected by the heat input. The fully γ-annealed top region consisted of epitaxial elongated grains without HAZ. The periodic pattern in microhardness along the building direction was found which was related to the periodic microstructure featured. The tensile properties presented anisotropic characteristics due to the heterogeneous microstructure. Further analysis indicated that the grain coarsening in the HAZ and C precipitates distributed at the grain boundaries caused substantial softened in the HAZ, resulted in the lower localized microhardness and tensile strength. Compared to the high heat input specimen, the low heat input specimen had smaller grain sizes, higher microhardness and tensile properties.

Effect of Damage Rate on the Cavity Swelling of Pure Nickel Irradiated with Triple Ion Beams

He-H synergistic effects influence the performance of structural materials in fusion reactors. Due to the lack of high-intensity fusion neutron sources, multiple ion beam irradiation has been widely used as an emulation method to study its synergistic effects. However, the damage rate under multiple ion beam irradiation is three to four orders of magnitude higher than that under fusion neutron irradiation, and its effect on the cavity swelling is still unclear. In this study, pure nickel was irradiated with single and triple ion beams to ~1 displacements per atom (dpa) at 450 °C. The damage rate ranged from 1.4 × 10−4 to 1.4 × 10−3 dpa/s, with the identical gas-dose ratios of ~400 H appm/dpa and 100 He appm/dpa. Large and isolated cavities formed under single ion irradiation, while triple ion irradiation induced smaller and denser cavities and higher swelling. As the damage rate increased, the cavity size, density, and swelling decreased, due to the constraint of cavity nucleation and growth processes. The effect of damage rate on cavity evolution under triple ion irradiation strongly depends on two competing factors: the enhancement of aggregation and binding of H/He/vacancies, and the enhancement of vacancies–interstitials recombination with increasing damage rate.

A property-oriented design strategy of high-strength ductile RAFM steels based on machine learning

Property-oriented design of RAFM steels can greatly enhance the opportunity to discover high-performance structural materials for fusion reactors, which has always been a big challenge. In the present work, the forward and reverse models are established, which are used to capture the mutual relationship of compositions and heat treatment conditions to tensile properties. The intelligent design model, combining the forward model with the reverse model, is developed to design the compositions and heat treatment parameters for RAFM steels with the targeted tensile properties. The validity of the intelligent design model is verified by the experimental data of three RAFM steels reported in the relevant literatures. Using this intelligent design model, a new type of RAFM steel was designed and prepared. In the test temperature range of 25–600 °C, the ultimate tensile strength of the new RAFM steel is ∼100–400 MPa higher than the conventional RAFM steels while maintaining comparable elongation. Therefore, this strategy is suitable for the property-oriented design of RAFM steels and can also be considered as a very promising approach to develop high-performance structural materials.

Study of liquid tin-lithium alloy interaction with structural materials of fusion reactor at high temperatures

The paper presents the experimental studies to determine the corrosion compatibility of various base materials of the capillary-porous structures (CPS) with a liquid tin-lithium alloy under high temperatures. The experiments were carried out with Li-Sn alloy and the samples of vanadium, tungsten, molybdenum, tantalum, and stainless steel. These materials were selected as the candidate ones for the matrix of Li-Sn CPS prototypes. Manufacture of Li-Sn alloy and experiments were carried out at test bench VIKA. The ranges of the temperatures was: 600 °C–800 °C for stainless steel, and 850 °C–1000 °C for other metals; exposure time for every temperature step was up to 100 h.The implemented activities included manufacture of Sn-Li alloys with various lithium contents (25 at.% and 27 at.%), characterization of the samples, experiments to study compatibility of liquid tin-lithium with vanadium, tungsten, molybdenum, tantalum at temperatures of 850 °C–1000 °C, and stainless steel at temperatures of 600–800 °C. As a result of the performed activities, molybdenum and tungsten were selected as the most corrosion resistant materials in respect to Li-Sn alloys.

Fe+ ion irradiation effects in Fe-10at%Cr films irradiated at 300 °C

Fe-Cr alloys constitute the model systems for the investigation of radiation damage effects in ferritic-martensitic steels which are candidate structural materials for fusion reactors. In the current study Fe-10at%Cr alloy films of 70 nm thickness were irradiated by 490 keV Fe + ions at 300 °C at doses ranging from 0.5 up to 20 displacements per atom (dpa). The Fe + ion energy chosen corresponds to the energy of primary Fe(Cr) knock-on atoms from 14 MeV neutrons. The irradiation effects were investigated employing X-ray diffraction and X-ray and polarized neutron reflectivity. The irradiation produced dose dependent: a) lattice constant increase, b) grain size growth and c) Cr depletion in the matrix. These changes occur largely up to 4 dpa and afterwards the system attains a dynamic equilibrium.

High temperature zirconium alloys for fusion energy

This review considers current Zr alloys and opportunities for advanced zirconium alloys to meet the demands of a structural material in fusion reactors. Zr based materials in the breeder blanket offer the potential to increase the tritium breeding ratio above that of Fe, Si and V based materials. Current commercial Zr alloys might be considered as a material in water-cooled breeder blanket designs, due to the similar operating temperature to fission power plants. For breeder blankets designed to operate at higher temperatures, current commercial Zr alloys will not meet the high temperature strength and thermal creep requirements. Hence, Zr alloys with an operational temperature capability beyond that of current commercial fission alloys have been reviewed, specifically: binary Zr alloy systems Zr-Al, Zr-Be, Zr-Cr, Zr-Nb Zr-Ti, Zr-Si, Zr-Sn, Zr-V and Zr-W; as well as higher order Zr alloys Zr-Mo-Ti, Zr-Nb-Ti, Zr-Ti-Al-V and Zr-Mo-Sn. It is concluded that, with further work, higher order Zr alloys could achieve the required high temperature strength, alongside ductility, while maintaining a low thermal neutron cross-section. However, there is limited data and uncertainty regarding the structural performance and microstructural stability of the majority of advanced Zr alloys for temperatures 500–700 °C, at which they would be expected to operate for helium- and liquid metal-cooled breeder blanket designs.

Evolution of dislocation loops and effect of annealing temperature on hydrogen-ion-implanted Fe-based binary alloys

Reduced activation ferritic/martensitic (RAFM) steels have been considered as a family of prime candidate structural materials for fusion reactors due to low radioactivity and good resistance to irradiation swelling. Various types of defects such as dislocation loops can form in these materials during irradiation. Effects of alloying elements in iron on the formation and migration of dislocation loops have been widely investigated. However, most studies dealt with interstitial-type dislocation loops in iron alloys, while very few focused on vacancy-type dislocation loops. Previous high voltage electron microscope (HVEM) studies from the authors' group have shown that interstitial loops are fully eliminated in hydrogen-ion-implanted α-Fe at 500 ℃, only vacancy loops remain and can achieve up to 100 nm in size. The addition of Ni in α-Fe can reduce the formation temperature of vacancy-type dislocation loops (<i>T</i><sub>c</sub>) to ~450 ℃, while the addition of Cr can increase the temperature to above 600 ℃. However, these experiments are usually difficult to perform due to the scarce resource of HVEM facilities. In this work, in-situ observations by conventional transmission electron microscope (CTEM, 200 kV) are systematically carried out on the hydrogen-ion-implanted α-Fe and Fe-based binary alloys (Fe-3wt.%Cr, Fe-1.4wt.%Ni and Fe-1.4wt.%Mn). The evolutions of morphology and average size of dislocation loops under different annealing temperatures are investigated. The formation temperatures of vacancy-type dislocation loops are determined from the change of average loop size with annealing temperature. The results are consistent with previous studies by HVEM. The effect of Mn atoms in α-Fe is similar to that of Cr atoms, which leads to <i>T</i><sub>c</sub> increase, and the addition of Ni in α-Fe can reduce <i>T</i><sub>c</sub>. Furthermore, the results of D thermal desorption spectrum analysis show that <i>T</i><sub>c</sub> is affected by the binding and release process of hydrogen isotopes to vacancies in α-Fe. Alloying element Ni promotes the binding and release of hydrogen isotopes to vacancies, which leads to<i> T</i><sub>c</sub> decrease. Cr and Mn inhibit the binding and release of hydrogen isotopes to vacancies, causing <i> T</i><sub>c</sub> to increase.

Finnis–Sinclair-type potential for atomistic simulation of defects behaviour in V-Ti-Ta ternary system

• Interatomic potential for V-Ti-Ta ternary system based on F-S formalism has been developed. • The Ti potential well reproduce the point-defect properties of vacancy and self-interstitial atoms and the pressure-volume relation. • The typical defect energies of solute atoms Ti and Ta in the V matrix and Ti and V in the Ta matrix predicted by the present potential were in good agreement with DFT results. Vanadium (V)-based alloys are potential candidates for structural materials of fusion reactors because of their excellent properties; recently, the V–Ti–Ta alloy has received considerable attention. The irradiation-resistant performance is significant for the development and application of fusion reactor materials. In this study, combining with our previously developed potentials of V and Ta, we developed the V-Ti-Ta interatomic potential based on the Finnis–Sinclair (FS) formalism, which is an important step towards an atomic-scale investigation of V-based alloys. The potential parameters were determined by fitting to a large database of experimental data and density functional theory (DFT) calculations. The formation energies of vacancy and self-interstitial atoms, as well as the pressure-volume relation obtained from the present Ti potential, were well reproduced. The typical defect energies of solute atoms Ti and Ta in the V matrix and Ti and V in the Ta matrix, such as the formation energies of substitutional solute atoms, binding energies between solute atoms and point defects, agree well with DFT results. Thus, the developed potentials are expected to be suitable for atomistic simulations of point defects evolution in the V-Ti-Ta ternary system.

Effect of Thermo-Mechanical Treatment on the Microstructure and Tensile Properties of the Fe-22Cr-5Al-0.1Y Alloy.

Oxide dispersion strengthened ferritic steel is considered an important structural material in fusion reactors due to its excellent resistance to radiation and oxidation. Fine and dispersed oxides can be introduced into the matrix via the powder metallurgy process. In the present study, large grain sizes and prior particle boundaries (PPBs) formed in the FeCrAlY alloy prepared via powder metallurgy. Thermo-mechanical treatment was conducted on the FeCrAlY alloy. Results showed that microstructure was optimized: the average grain diameter decreased, the PPBs disappeared, and the distribution of oxides dispersed. Both ultimate tensile strength and elongation improved, especially the average elongation increased from 0.5% to 23%.