Japanese activities on fusion structural materials R&D have been well organized under the coordination of university programs and JAERI/NIMS programs more than two decades. Where, two categories of structural materials have been studied, those are; reduced activation ferritic/martensitic steels (RAFs) as reference material and vanadium alloys and SiC/SiC composite materials as advanced materials. The R&D histories of these candidate materials and the present status in Japan are reviewed with the brief explanation of Japanese strategy and current status of fusion reactor engineering R&D.

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Reduced-activation ferritic/martensitic (RAF/M) steels have been considered to be the prime candidate for the fusion blanket structural material. The irradiation data obtained up to now indicates rather high feasibility of the steels for application to fusion reactors because of their high resistance to degradation of material performance caused by both the irradiation-induced displacement damage and transmutation helium atoms. The martensitic structure of RAF/M steels consists of a large number of lattice defects before the irradiation, which strongly retards the formation of displacement damage through absorption and annihilation of the point defects generated by irradiation. Transmutation helium can be also trapped at those defects in the martensitic structure so that the growth of helium bubbles at grain boundaries is suppressed. The major properties of the steels are well within our knowledge, and processing technologies are mostly developed for fusion application. RAF/M steels are now certainly ready to proceed to the next stage, that is, the construction of International Thermo-nuclear Experimental Reactor Test Blanket Modules (ITER-TBM).
Oxide dispersion strengthening (ODS) steels have been developed for higher thermal efficiency of fusion power plants. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening accompanied by no loss of ductility. High-Cr ODS steels whose chromium concentration was in the range from 14 to 19 mass% showed high resistance to corrosion in supercritical pressurized water. It is shown that the 14Cr-ODS steel is susceptible to neither hydrogen nor helium embrittlement.
A combined utilization of ODS steels with RAF/M steels will be effective to realize fusion power early at a reasonable thermal efficiency.

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Vanadium alloys are attractive candidate structural materials for breeding blanket of fusion reactors because of their low activation properties and high temperature strength. Studies on various blanket designs are being carried out using vanadium alloys as structural materials and liquid lithium or molten salt Flibe as breeding and coolant materials. Recently significant progress in fabrication technology has been made for vanadium alloys through a program of producing large ingots of high purity V-4Cr-4Ti (NIFS-HEATs). Fundamental understandings on the effects of interstitial impurities (C, N, O) on mechanical properties, radiation effects on microstructure and mechanical properties, corrosion and compatibility in various environments were enhanced. New promising candidates were identified for the MHD insulator coating. Among the remaining critical issues are the effects of transmutant helium on mechanical properties and development of long life MHD coating. Development of vanadium alloys will be carried out in coordination with IFMIF and ITER-TBM schedules as essential tools for verification of their performance in fusion blanket environments.

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m² in the divertor targets and around 0.5 MW/m² at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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For the development of materials in a fusion reactor environment without a fusion reactor, it is important to understand the generation and accumulation of point defects. This paper discusses five important factors in metals that influence the initial stage of defect clustering under cascade damage conditions. The effect of the PKA energy spectrum on damage evolution was explained from the viewpoint of subcascade formation. Two origins of damage rate dependence of defect cluster formation, direct mutual annihilation of point defects and annihilation of point defects at cascade-induced defect clusters, were mentioned. As the effect of motion of interstitial clusters, an example was given to illustrate the strong correlation between the mobility of interstitial clusters and void growth. Thermal activation processes depend on irradiation temperature. Varying irradiation temperature experiment was evaluated as a technique for investigating the point defect processes during irradiation. Migration of alloying elements during irradiation was reviewed from the standpoint of the interaction of alloying elements with point defects.

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In-situ TEM observations were carried out in copper and gold under 100 keV C⁺, 240 keV Cu⁺, 600 keV Kr²⁺, and 900 keV Xe³⁺ irradiations from 573 to 823 K, in order to obtain direct experimental insights into the defect accumulation processes. Defect clusters corresponding to displacement cascades were observed to be unstable depending on the temperature, ion species, and fluence. Multiple (2 or 3) defect clusters showing up with their contrast in the same video frames were considered to be features associated with subcascades and high mobility of these clusters when located within 20 nm and from 20 to 140 nm, respectively. Instability and diffusion of defect clusters were also detected under ion irradiation. The directions of the cascade-driven and instability-driven motions of the defect clusters were strongly related to crowdion directions, suggesting that the underlying mechanism is based on the motion of crowdion-related glissile defects. This instability is interpreted as the transition of sessile defects into glissile ones. The effects of intra- and inter-cascades on the formation of glissile defects are suggested based on their dependence on the ion species and flux.

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The influence of undersized solute additions on radiation-induced segregation (RIS) at grain boundaries was examined for austenitic model alloys. High-purity Fe-15Cr-20Ni alloys, doped with either 0.5 mass% silicon or 0.025 mass% phosphorus, were exposed to electron irradiation at several temperatures in a high-voltage electron microscope. The grain boundary depletion of chromium and segregation of nickel were significantly suppressed by the addition of silicon and phosphorus, and the addition of phosphorus significantly reduced the RIS at a grain boundary. It was shown by numerical simulation that the present results stem from the strong interaction between the additives and interstitial atoms produced by the electron irradiation.

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Vacancies and interstitial type dislocation loops of two fusion reactor candidate materials (V-4Cr-4Ti and F82H) after fission and fusion neutron irradiation were studied by positron annihilation lifetime spectroscopy. Fusion neutron irradiation was performed in the FNS of JAERI, and fission neutron irradiation was performed in the KUR of Kyoto University. The neutron irradiation dose was about 1×10⁻⁶–1×10⁻⁴ dpa, and the irradiation temperatures were room temperature and 673 K. In the irradiation at room temperature, the defects (mainly dislocation loops) in both alloys were detected even at a low irradiation dose of 10⁻⁶ dpa and the mean lifetime of positrons increased as the irradiation dose increased. The effects of the fission and fusion neutron irradiation on the point defect production were almost the same if they were compared at the same dpa. This can be explained by the fact that the number of subcascades, which is an important factor for the defect formation at room temperature, is proportional to dpa in these metals. In contrast, an effect of cascade size was found for the irradiation at 673 K. Dislocation loops were detected only in the fusion-neutron-irradiated F82H at 673 K.

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Since it has recently become an important subject to study the interrelation between a dislocation loop and a straight edge dislocation because of dynamic behaviors of small dislocation loops under irradiation condition, the simulation studies were made in model Fe lattice. It was found that dislocation cores both for a loop and a dislocation can be expressed by the regular array of three types of crowdions and have stable positions of b/3 periodicity, but this regularity and periodicity are disturbed as the decrease of the loop size. The relation between a dislocation core and crowdion types are also studied. Furthermore, this change gives rise to the increase of the inherent lattice resistance (Peierls stress) for the dislocation loops.

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We report thermally activated transport of highly mobile dislocation loops in terms of a line tension model where the dislocation loops are assumed to be a flexible string. The activation energy and transition rate are calculated on the basis of a classical rate theory. The activation energy merely increases with the length of the dislocation loops. However, the activation process and temperature dependence of the transition rate qualitatively change at a critical length L_c. If the dislocation loops are longer than the critical length, the thermal activation occurs through the conventional double-kink formation process on the dislocation lines. On the other hand, if the dislocation loops are shorter than that, the saddle point configuration is not the double-kink type but non-deformation one. Therefore, the critical length L_c is a plausible criterion for the dislocation loops to distinguish dislocation like from point-defect like in size.

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Self-interstitial atom (SIA) type dislocation loop is one of the possible candidates of the so-called matrix damage that causes hardening and embrittlement of blanket structural materials of fusion reactors and/or pressure vessel materials of light water reactors. We present in this paper molecular dynamics computer simulation results on the interactions between an edge dislocation and a SIA loop with Burgers vectors of b=\\fraca₀2[1\\bar11] and b=\\fraca₀2[\\bar111], respectively, which are introduced in bcc-Fe crystal. Then shear stresses of several different magnitudes are applied so that the dislocation moves to meet the SIA loop. General observation is that the SIA loops with diameter of ∼2 nm can be obstacles to dislocation motion, and the strength as obstacles to dislocation motion depends on applied stress. The origin of the stress dependent strength can be explained athermally using the elastic theory of dislocation interaction. In most cases, the SIA loops are absorbed by the edge dislocations to form a large super-jog after the interactions. This suggests a possibility of localized deformation of irradiated bcc-Fe due to the formation of dislocation channeling.

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It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni-doped F82H, show a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573 K up to 5 dpa. These differences could not be interpreted as an effect of irradiation hardening caused by dislocation loop formation. In this paper, the effects of irradiation on precipitation of RAFs were investigated to determine how these effects might affect the mechanical properties. The precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. These analyses suggested that irradiation caused (1) an increase of the amount of precipitates (mainly M₂₃C₆), (2) an increase of Cr and decrease of W contained in precipitates, and (3) the disappearance of MX (TaC) in ORNL9Cr and JLF-1.

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Validity of nickel isotope tailoring (NIT) method to be able to generate a large amount of transmutation helium in 9%Cr-2%W reduced activation ferritic (RAF) steel for fusion reactor structural material was investigated and discussed through mechanical properties and microstructural evolution after fission neutron irradiation. Miniature tensile and Charpy impact specimens of RAF steels and 1% nickel added RAF steel were irradiated in the ATR-A1 to evaluate irradiation hardening and shift in ductile-brittle transition temperature (ΔDBTT). The amount of transmutation helium in all the RAF steels with and without nickel addition was calculated as only about 0.6 appm. After irradiation at 621 K, the ΔDBTT for the 1% nickel added steel was similar for the JLF-1. After irradiation at 543 K, however, the ΔDBTT for the 1% nickel added steel was significantly larger than that for the JLF-1. Microstructure observations revealed that irradiation-induced dislocation loops in the 1% nickel added steel were finer and denser than in the RAF steel without nickel addition, suggesting that the nickel addition to the RAF steels directly affected nucleation and growth processes of dislocation loops and enhanced irradiation hardening and embrittlement. Therefore, there is a limit in the NIT method as a simulation method for understanding effects of helium on irradiation embrittlement of RAF steels.

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The influcence of addition of ¹⁰B and ¹¹B on radiation hardening as a function of irradiation temperature was examined in reduced-activation martensitic F82H+B steels (8Cr-2W-0.2V-0.04Ta-0.1C-0.006B). Specimens used were ¹⁰B doped, ¹⁰B+¹¹B doped and ¹¹B doped F82H steels. Helium concentration produced in the specimens through ¹⁰B(n, α)⁷Li reaction was measured from about 15 to about 330 appm. Radiation hardening of the specimens irradiated at 300°C to 2.3 dpa and 150°C to 1.9 dpa was observed. Increment of radiation hardening possibly due to helium in the specimen irradiated at 300°C tended to increase slightly with increasing helium production in tensile test at 20°C, but it was not observed for 150°C irradiation. Therefore, it can be mentioned that the enhancement of radiation hardening in the ¹⁰B(¹¹B) doped steels depends on irradiation temperature. The dependence of radiation hardening on tempering time was also examined for F82H-std irradiated at 150°C to 1.9 dpa. Increment of yield strength of F82H-std tended to increase with increasing tempering time. The relation between the shift of DBTT and the increment of yield strength due to irradiation suggest that the extension of tempering time or the increase of tempering temperature would be beneficial for mechanical properties of F82H-std.

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The effects of chemical compositions (titanium, oxygen) and consolidation temperature on high-temperature mechanical properties of 9Cr-oxide dispersion strengthened steel (9CrODS steel) were investigated. A possible high-temperature strengthening mechanism of 9CrODS steel was discussed based on the experimental results. Creep strength of 9CrODS steel at 973 K was remarkably improved when titanium concentration was 0.35 mass%. A higher amount of added titanium than 0.2 mass% was effective for providing consistently reliable manufacturing of high strength 9CrODS steel because it reduced the effect of oxygen contamination on high-temperature strength. The fraction of elongated δ-ferrite grains, which had an ultra-fine oxide particle dispersion, tended to increase with increasing titanium. The elongated grains were considered to improve creep strength of 9CrODS steel. It was also found that creep strength was degraded by elevating the consolidation temperature from 1423 K to 1473 K.

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Oxide dispersion strengthened (ODS) steels, which are candidate materials for water-cooled solid breeder blankets, were fabricated with several manufacturing parameters, and then irradiated in the experimental fast reactor JOYO to evaluate their irradiation behavior.
Engineering stress strain curves of ODS steels irradiated at 673 K exhibited superior material response, i.e., increased tensile strength due to irradiation hardening and no loss of total elongation. Also, their temperature dependence of tensile properties indicated that degradation of the tensile properties at elevated temperature, which is closely related to phase stability during irradiation, could be avoided due to optimal combination of manufacturing parameters, such as chemical composition, types of inert gas during mechanical alloying, heat-treatment temperature and initial phases of the matrix.

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Weld samples of fusion reactor vanadium alloy (NIFS-HEAT-2) were neutron-irradiated at 563 K up to 4.5×10²³ neutrons/m² (0.08 dpa). The recovery of irradiation hardening, degradation of impact properties, and damage structures in the weld metal were investigated after post-irradiation isothermal annealing at temperatures between 673 and 1073 K. Irradiation hardening and the decrease in impact absorbed energy at 77 K were larger for the weld metal than for the base metal. Recovery of the hardening by post-irradiation annealing was coincident with recovery of the impact absorbed energy in both the weld metal and the base metal. Recovery of the mechanical properties required post-irradiation annealing at 1073 K for 1 h for the weld metal, which was 100 to 200 K higher than that for the base metal. The dislocation loops introduced by the neutron irradiation are likely to account for the hardening. The dislocation loops were observed even after annealing at 973 K in the weld metal, whereas they disappeared in the base metal. The characteristics of the radiation defects in the weld metal and the mechanisms for irradiation hardening and embrittlement are discussed.

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Microstructural examination and microhardness test were carried out for unalloyed vanadium, V–5Ti, V–4Cr–4Ti alloys irradiated with 4 MeV Cu ions at temperatures from 200 to 600°C. In order to investigate the impurity effect, high purification technique was used. Impurity effects can be seen in the nucleation and growth processes of voids in unalloyed vanadium and those of Ti(OCN) precipitates in V–5Ti. The changes in microstructure due to interstitial impurity could not be seen in V–4Cr–4Ti after the irrodiations below 300°C. The effect of impurity for irradiation hardening in ion-irradiated V–4Cr–4Ti alloys could be seen at 400°C, but it could not be seen at 200 and 300°C. On the other hand, dislocation channels were observed at 200 and 300°C, and the total length of dislocation channels became larger as the impurity level increased. It can be considered that there was practically no correlation between irradiation hardening and total length of dislocation channels due to impurity effects.

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The change in the chemical surface state of a V-4Cr-4Ti alloy after heat treatments in vacuum at temperatures from 573 to 1273 K was investigated by means of X-ray photoelectron spectroscopy. Before heating, the surface of the as-polished alloy was covered with a V oxide film. Diffusion of oxygen into the bulk started at around 673 K, and the alloy surface gradually became metallic. Oxygen, however, did not disappear from the surface completely. Surface segregation of Ti was observed as the temperature increased. The concentration of Ti reached 20 at% at 1273 K. No significant segregation of Cr or other impurities were observed. This surface segregation of Ti significantly reduced the surface reaction rates of H₂ and D₂. The isotope effect on the surface reaction rates of H₂ and D₂ was not observed under the present experimental conditions.

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Corrosion tests in pressurized and vaporized water were conducted for V-based high Cr and Ti alloys and V-4Cr-4Ti type alloys containing minor elements such as Si, Al and Y. Weight losses were observed for every alloy after corrosion tests in pressurized water. It was apparent that addition of Cr effectively reduced the weight change in pressurized water. The weight loss of V-4Cr-4Ti type alloys in corrosion tests in vaporized water was also reduced as Cr content increased. The V-20Cr-4Ti alloy had a slight weight gain, almost same as that of SUS316, which had the best corrosion properties in the tested alloys. The elongation of alloys with in excess of 10% Cr was reduced as Cr content increased. The elongations of the V-12Cr-4Ti and the V-15Cr-4Ti alloys were significantly reduced by corrosion and cleavage fracture was observed reflecting hydrogen embrittlement. The reduced elongations of the alloys were recovered to the same level of as annealed conditions after hydrogen degassing. After corrosion, the V-15Cr-4Ti-0.5Y alloy still kept enough elongation, suggesting that the addition of Y is effective to reduce the hydrogen embrittlement.

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In fusion reactors, structural materials are expected to experience non-steady histories of irradiation temperature, neutron flux and other parameters during reactor start-up/shut-down, plasma disruptions, etc. The objective of the present study is to clarify the effects of the downward temperature change during irradiation.
Vanadium that is the one of candidate materials for the first wall of fusion reactors was irradiated with heavy ions during the downward temperature change. TEM observation, nano-indentation and HVEM observation were carried out.
It was reported that the growth of defect clusters including cavities and precipitates occurred under neutron irradiation and heavy ion irradiation during the downward temperature change. In the present paper, the detailed mechanism of growth of defect clusters was investigated by changing the lower temperature as a parameter. The growth of defect clusters occurs just after the downward temperature change followed by re-nucleation of defect clusters at lower temperatures.

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Fast fracture properties of chemically vapor-infiltrated silicon carbide matrix composites with Hi-Nicalon^TM Type-S near-stoichiometric silicon carbide fiber reinforcements and thin pyrolytic carbon interphase were studied. The primary emphasis was on preliminary assessment of the applicability of a very thin pyrolytic carbon interphase between fibers and matrices of silicon carbide composites for use in nuclear environments. It appears that the mechanical properties of the present composite system are not subject to strong interphase thickness effects, in contrast to those in conventional non-stoichiometric silicon carbide-based fiber composites. The interphase thickness effects are discussed from the viewpoints of residual thermal stress, fiber damage, and interfacial friction. A preliminary conclusion is that a thin pyrolytic carbon interphase is beneficial for fast fracture properties of stoichiometric silicon carbide composites.

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Cavity formation was investigated in a SiC/SiC composite under multi-ion beam irradiation up to 10 dpa at 1073 K, 1273 K and 1573 K by transmission electron microscopy. The cavity formation behavior of each component of the composite was dependent on the component’s grain structure, the helium and/or hydrogen implantation mode, and irradiation temperature. It was found that helium rather than hydrogen is likely to enhance cavity formation or cavity swelling. The contributions of helium and hydrogen to cavity formation in the composite components are discussed in detail.

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Mechanical testing after neutron irradiation is a critical research tool for evaluating materials for fusion systems, such as silicon carbide fiber silicon carbide matrix (SiC/SiC) composites. However, single-axis tensile testing, which is required to build a fundamental database, requires large specimens. Therefore miniaturization of tensile test specimens has long been pursued as a method to reduce the irradiation volume to fit the capsule size limitation. The objective of this study is to identify specimen size effects on tensile properties of SiC/SiC composites from the viewpoints of the influences of fabric architecture and tensile loading axis, with a final goal to establish a small specimen test technique for tensile testing of the composites. The axial fiber volume fraction plays an important role in achieving good tensile properties. However the size dependent change of the axial fiber volume fraction gives specimen size effect. The composites with much fiber volume content tended to have superior tensile strength, elastic modulus and proportional limit stress. Contrarily, the tensile properties of the composites with the same axial fiber volume fraction were almost independent of the specimen size. This type of size effect is generally common in any types of architecture. The size-relevant fracture mode in off-axis tension: detachment in shorter widths vs. in-plane shear at larger widths, also gives specimen size effect on tensile properties, resulting in strict limitation of miniaturization of the tensile specimen. Finally we proposed a miniature tensile specimen for the composites.

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The elucidation of the trapping and detrapping mechanisms of hydrogen isotopes in SiC is one of the most critical issues for future fusion reactors if SiC is used as the first wall and structure material. In this study, 1 keV deuterium (D₂⁺) ions were implanted into SiC and the chemical states of C and Si were evaluated by X-ray photoelectron spectroscopy (XPS). The deuterium desorption and retention were also analyzed by thermal desorption spectroscopy (TDS). The deuterium desorption behavior for SiC was compared to that for Si and graphite, and it was found that deuterium is preferentially trapped by C and, after the saturation of the C-D bond, it is trapped by Si in SiC. Deuterium desorption was found to consist of two stages, namely deuterium desorptions bound to Si and C. Their trapping mechanisms were influenced by the damaged structures produced by the D₂⁺ ion implantation. Finally, deuterium retention in SiC at temperatures above 700 K was higher than that in graphite, indicating that tritium retention in SiC may be high compared to that in graphite during plasma operation.

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Tungsten mixed carbon dust was prepared using a deuterium arc plasma discharge with tungsten and carbon electrodes. The concentration of tungsten with an atomic ratio of W/(C+W) was in the range from 0 to 0.3. The Raman spectra showed that the crystal structure of carbon became amorphous with the addition of tungsten, although two broad peaks owing to graphite and defective graphite appeared. The structure also became amorphous when the substrate temperature decreased. The retained deuterium desorbed mainly at approximately 900 K, which corresponds to the desorption of deuterium trapped by carbon atoms. Namely, most of the co-deposited deuterium is retained in the carbon atoms. The amount of retained deuterium increased with an increase in the tungsten concentration. The amount of retained deuterium was larger than that of the co-deposited carbon dust which was similarly prepared in the arc plasma discharge with only carbon electrodes. The increase in deuterium retention is due to the enhanced amorphous structure by the mixture of tungsten.

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Surface modification of polycrystal tungsten due to low-energy and high-flux helium plasma exposure was investigated. Micron-sized holes and bubbles were formed on the surface at a surface temperature of 2200 K even when the incident ion energy of helium was 10 eV. Hole formation was significantly reduced on the surface when the incident ion energy of helium was around 5 eV and no surface modification was seen at less than 5 eV. Coalescence of micron-sized helium bubbles was observed in the cross section of a W sample.

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Surface modification and internal damage of tungsten irradiated with 8 keV Helium (He) ions at 1273 K was observed by the complementary use of a scanning electron microscope (SEM), an atomic force microscope (AFM), and a transmission electron microscope (TEM). It was found that an irradiated surface was roughened by the formation of dense large bulges of 100–200 nm and small ones of a few nm in size. It was also observed that fine He bubbles ranging from a few to 100 nm were formed in the subsurface layer. These results indicate that roughening of a He⁺ ion irradiated surface at a high temperature results from the formation of bubbles just beneath the surface. Their thermal migration and coalescence under irradiation play an important role in the formation of various sized bubbles.

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Low temperature embrittlement, recrystallization embrittlement and radiation embrittlement in tungsten and its alloys are critical issues for use as high heat flux components and high-power density structural materials. In order to establish a process for microstructural control to improve the resistance to such embrittlement, modification of powder-metallurgical processing is proposed to avoid three microstructural factors giving detrimental effects on the ductility: (1) precipitation of the brittle W₂C phase, (2) heterogeneity in grain size and particle distributions, and (3) loss of carbon which is a constituent of transition metal carbides. The processing was applied to fabricate W-0.3 mass%TiC alloys with a microstructure of fine grains and nano-sized dispersoids of TiC. Transmission electron microstructural observations and three-point bending tests at room temperature were performed on the alloys in the unirradiated state. It is demonstrated that the developed alloys are almost free from the three microstructural factors and exhibit appreciable room-temperature ductility before fracture in the as-forged and as-rolled states, but not in the as-HIPed state. This beneficial effect of plastic working on ductility improvement is strongly dependent on grain size and becomes prominent with decreasing grain size. Success in fabricating consolidated bodies with grain sizes as small as 0.17∼0.4 μm and a high relative density of around 99% is presented.

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The spark plasma sintered in situ TiB reinforced titanium metal matrix composite was investigated by EBSD and TEM. The sintered composite consists of 30.8% α-Ti phase, 55.7% β-Ti phase and 12.9% in situ TiB reinforcements. Most of the TiB whiskers distributed along phase boundaries between α-Ti and β-Ti, others located within α-Ti and β-Ti grains. Some β-Ti grains with parallel TiB whiskers were observed in the orientation imaging micrograph (OIM). The TiB grows along [010] direction and forms whisker with a hexagonal cross-section. The [010] direction of paralleled TiB whiskers is parallel to [111] direction of cubic β-Ti. High density stacking faults are formed in (100)_TiB planes to minimizing the lattice mismatch between TiB and the Ti matrix.

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The formation of austenite during the heating of high-silicon steels with a microstructure of bainitic ferrite plates separated by carbon-enriched regions of retained austenite has been studied. As the steel is heated, retained austenite decomposes into a mixture of ferrite and carbides before a temperature is reached where it becomes the more stable phase and hence can grow. The decomposition makes it necessary for the austenite to nucleate from the mixture of ferrite and carbides. In this work, dilatometric analysis, transmission electron microscopy and X-ray diffraction have helped to identify the processes that take place during the heating experiment.

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Lotus-type silver was fabricated by unidirectional solidification in mixture gas of oxygen and argon at high pressures. Two solidification methods were adopted in this study: the unidirectional solidification casting method and the Czochralski method. The directional pores were evolved by insoluble oxygen when silver melt dissolving oxygen was solidified. Non-uniform size distributions of pores in distorted shapes were observed in lotus-type silver fabricated by the unidirectional solidification casting method. On the other hand, round shape pores with good circularity and more uniform pore size were found in the cross section of the lotus-type silver made by the Czochralski method. The distortion of pore shape is attributed to the re-dissolution of oxygen gas in the pores to the molten silver during solidification.

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This work was carried out to investigate the effect of welding variables (includes wire feeding techniques, wire feeding rates and heat inputs) on the cooling rate, Δt_8⁄5 (cooling time from 800 to 500°C), in the weld and heat affected zone (HAZ) areas of multi-pass weldments in a super duplex stainless steel. Furthermore, changes in thermal condition caused by welding variables can affect the microstructure and consequently the material properties. Therefore, supporting tests, including pitting corrosion resistance and microstructure analysis were investigated.
To establish the effect of welding variables on the cooling rate, a series of weldments using a semi-automatic welding process which is enable the control of welding variables such that one parameter can be varied at a time were completed. The method used for pitting resistance tests was the basic ASTM G48-76 standard, with reference to the ASTM A923-94 standard and recommended practice for pitting corrosion tests for duplex stainless steel weldments by the use of ferric chloride solution from The Welding Institute (TWI). Microstructure Analysis was carried out using a Buehler Omnimet Image Analysis System and a manual point counting technique with reference to ASTM E562-89 (Standard test method for determining volume fraction by systematic manual point count).
From these tests results, various empirical correlations have been proposed to relate the cooling rates and pitting corrosion resistance to the welding variables.

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Single crystals of pure copper were tensile-deformed along ⟨011⟩- and ⟨001⟩ direction and then annealed to study their deformation and recrystallization behaviors. In the ⟨011⟩ single crystals, no large-scale deformation band was formed. The areas where the secondary slip was activated were small and were randomly distributed throughout the deformed crystals. Large rotation with respect to the initial orientation was found. After annealing, many recrystallized grains were formed. In the ⟨001⟩ single crystal, because of almost equal activation of eight slip systems on four slip planes, orientation change associated with deformation was very small. Large-scale cross-slip was completely suppressed. As a result, the work-hardening of the ⟨001⟩ single crystal was very large. After annealing, large primary recrystallized grains were formed. The above results are compared with those of aluminum single crystals having the same tensile orientations.

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Electron irradiation was performed in an amorphous phase of Fe₇₁Zr₉B₂₀ metallic glass. Although the amorphous phase showed high thermal stability with a wide supercooled liquid region of 71 K, it could not maintain the original structure under electron irradiation and electron irradiation induced crystallization occurred at 103 K and 298 K. There was a great difference in phase selection between thermal crystallization and electron irradiation induced crystallization. The irradiation temperature strongly affected the phase selection and transformation kinetics. The dominant factors of phase selection in electron irradiation induced crystallization were discussed on the basis of phase stability against electron irradiation and the necessity of redistribution of constituent atoms.

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Phase transformation behavior and phase stability of Zr_66.7Cu_33.3 metallic glass were examined during thermal annealing and electron irradiation. Metastable f.c.c.-Zr₂Cu phase precipitated in amorphous matrix and formed nanostructure by electron irradiation induced crystallization. With further thermal annealing, the nano grains of the f.c.c.-Zr₂Cu coarsened accompanied by precipitation of thermally stable b.c.t.-Zr₂Cu phase from amorphous matrix. The thermal equilibrium b.c.t.-Zr₂Cu crystalline phase was transformed to the nanocrystalline f.c.c.-Zr₂Cu phase through the amorphous state during electron irradiation. The unique solid state amorphization and crystallization behavior by electron irradiation can be explained by the thermodynamical model based on the change in relative phase stability among amorphous and crystalline phases by electron irradiation. The thermal stability of the f.c.c.-Zr₂Cu phase and the effect of dose rate on electron irradiation induced phase transformation were investigated in order to confirm the validity of the thermodynamical model.

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A finite element analysis was employed to simulate the deformation behavior of Al-Mg alloy foams of 92% porosity during compressive testing and the simulated results were compared with experimental ones. The crushable foam model developed in ABAQUS software for compressible foam materials could well reproduce the uniaxial compressive behavior in terms of stress-strain response and deformed geometry. In particular, the effect of friction between the dies and the foam specimen was discussed.

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The rate of cavitation with superplastic strain was investigated for a superplastic AZ61 magnesium alloy at a strain rate of 2×10⁻⁴ s⁻¹ and temperature of 648 K, under the conditions of which an elongation of more than 250% has been found. Cavities initiated at grain boundaries. The cavitation showed a growth perpendicular to the applied stress direction after the initial strains. The subsequent growth and coalescence of cavities invariably leads to failure of the material. The experimental growth rates are in good agreement with the rate predicted by the plasticity-controlled growth mechanism.

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Modified Small Punch (MSP) test techniques have been used to evaluate the strength of Mo/PSZ (partially stabilized zirconia) composites from room temperature to 1573 K. The dependence of deformation, strength and fracture behavior of the composites on temperature, composition and microstructure is discussed in detail. The results show that the high temperature strength, which depends on the composition as well as the microstructure, can be simply measured by the MSP tests. Linear relationship is obtained between the MSP strengths and 4-point bending strengths.

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In-situ X-ray diffraction measurements by synchrotron radiation and conventional X-ray source have been carried out for identifying constituent species of corrosion products, which were formed on the surface of a pure iron and an iron-5 mass% nickel alloy by reaction with aqueous solutions containing sodium chloride or sodium sulfate. A cell has been prepared for in-situ diffraction measurements of the corrosion products. Diffraction patterns from the corrosion products showed that major constituent species of the corrosion products was γ-FeOOH, and the fraction of minor species of α-FeOOH and Fe₃O₄ in the corrosion products depended on corrosion conditions, such as wetting and drying processes, anion species in aqueous solutions and nickel in the sample. An oxide scale thermally formed on the iron surface was also found to affect the formation of corrosion products.

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The thermodynamic properties for the CaO-CaF₂, BaO-CaO and BaO-CaF₂ systems were calculated by molecular dynamics (MD) simulation using the simple Born-Mayer-Huggins type potential model. The interatomic potential parameters were determined by fitting the thermodynamic properties of pure CaO, BaO and CaF₂. The calculated thermodynamic properties for CaO, BaO and CaF₂ were in good agreement with measured results, and the superionic conductivity on the solid-solid phase transition of CaF₂ has also been successfully assessed by MD simulation. The ΔH^M, ΔS^M and ΔG^M for each binary system were calculated based on the thermodynamic parameters obtained by MD simulation and thermodynamic solution model. The calculated enthalpy interaction parameters for the BaO-CaF₂ system represented the possibility of formation of the compounds such as BaO·CaF₂ in the BaO-CaF₂ system. The calculated phase diagrams for the CaO-CaF₂ and BaO-CaO systems were in good agreement with experimentally measured and CALPHAD method results. The calculated eutectic points for the CaO-CaF₂ and BaO-CaO systems were about 20 mol% CaO at 1650 K and about 20 mol% CaO at 2050 K, respectively. The BaO-CaF₂ system has also been estimated the liquidus lines in the CaF₂-rich and BaO-rich region by MD simulation.

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An interface between stratified two liquids contained in a cylindrical vessel rotated around the vertical vessel axis when a bottom blown jet of the lower liquid impinged on the interface. This swirl motion has high mixing intensity, and hence, is useful for enhancing mass transfer at the interface. A critical condition for the occurrence of the swirl motion was experimentally investigated. Empirical equations proposed originally for predicting the critical occurrence condition of a swirl motion in a single-liquid bath were applicable to this case by replacing the gravitational acceleration constant by a modified gravitational acceleration one.

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From the viewpoint of recycling and recovery of metal values from used lead-batteries, especially from lead anode slime, recovery of antimony has been studied experimentally. First, oxidation kinetics has been investigated for pure liquid antimony in the temperature range between 973 and 1373 K to elucidate the reaction mechanism. Since lead anode slime generally consists of antimony, lead and bismuth, oxidation experiments have also been carried out using antimony-lead-bismuth alloy. It was found that gas phase mass transfer step mainly controls the overall oxidation rate. The overall rate was expressed as the sum of antimony oxide and metallic antimony evaporation rates. The oxide evaporation is dominant at lower temperature around 1073 K with higher oxygen partial pressure. In the experiment of the antimony-lead-bismuth alloy simulated for anode slime, only the oxide Sb₄O₆ evaporation was observed, indicating that antimony was preferentially oxidized followed by evaporation of antimony oxide. The oxidation rate of the alloy was substantially identical with that of pure antimony. This is the advantage of oxidation treatment for anode slime.

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Lithium ferrite powder was prepared at quite a low temperature by solid-state reaction, and its performance as a CO₂ absorbent was tested. In order to evaluate its performance efficiently, the idea of combinatorial-material development was introduced and combined with novel X-ray fluorescence (XRF) imaging analysis. The XRF of multiple samples on a combinatorial substrate were observed in parallel—and therefore in a short space of time—thus enabling the X-ray absorption fine structure (XAFS) of the samples to be compared with one another. Thermogravimetry (TG) was also used to estimate the absorption speed and absorbing mass of CO₂ quantitatively. The X-ray diffraction pattern showed that the obtained lithium ferrite was mainly α-LiFeO₂, and the broadening of the lines suggested a nano-crystalline product. LiFeO₂ reacted with CO₂ to form Li₂CO₃ and γ-Fe₂O₃. Changes in the XAFS spectra and the weight changes observed by TG with CO₂ exposure showed quick and voluminous absorbance by the low-temperature synthesized sample, and a higher performance was confirmed compared with samples prepared by the conventional ceramic method.

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Zircon is widely used in industrial field as a refractory material, because of its excellent mechanical and chemical properties at high temperature. Authors have been studied on the effect of the heat treatment on the structure and properties of plasma sprayed zircon coating as a candidate for an environmental barrier coating (EBC). In this study, very dense coating with excellent adhesive strength was successfully obtained by optimizing the spray process. Substrate temperature is one of the dominant factors to control porosity and crack formation in the coating. The higher substrate temperature, obtained by plasma pre-heating, resulted in lower porosity and less cracks. Open porosity of the coating was quite low, about 2%, in the coatings obtained especially above 1473 K of the substrate temperature.

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The devitrification of Ni₆₀Nb₂₅Ti₁₅ glassy alloy has been investigated by X-ray diffraction (XRD), differential scanning and isothermal calorimetry and transmission electron microscopy (TEM). The primarily crystallized phase was a cubic Ni(Ti,Nb) phase with a lattice parameter of 0.293 nm. A Ni₃Nb phase follows precipitation of the Ni(Ti,Nb) phase. The cubic Ni(Ti,Nb) phase undergoes partial transformation to Ni₄Ti₃ phase. These cubic Ni(Ti,Nb) phases disappear at the equilibrium conditions.
The supercooled liquid region (ΔT_x) of the Ni₆₀Nb₂₅Ti₁₅ glassy alloy was extended to 64 K with the addition of Pt. The Ni₅₅Nb₂₅Ti₁₅Pt₅ glassy alloy rods with diameters up to 2 mm were formed by mold casting. Pt and the other noble metals additions do not alter devitrification of Ni₆₀Nb₂₅Ti₁₅ glassy alloy.

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This study seeks to clarify the effects of gas contamination from milling atmospheres of mechanical alloying (MA) on mechanical properties. An iron-based dispersion alloy of Fe-13Cr-3W-0.5Ti-0.5Y₂O₃ (mass%) was selected as the experiment material. We prepared MA powders by milling mixed powders in atmospheres of argon, helium, hydrogen, nitrogen, and vacuum; then we made bulk alloys by groove rolling the MA powders. We then examined atmospheric elements trapped in the MA powders and their releasing processes with heat treatment in vacuum. For bulk alloys, we also examined the high-temperature behavior of residual atmospheric elements and their effects on impact strength as a function of heat-treatment time at 923 K.
Experiment results demonstrated that some of the atmospheric element trapped by MA powder was unexpectedly difficult to remove with heat treatment. The content of widely used argon in MA powder, for example, was 0.013 mass%, and the argon was difficult to remove even with treatment at 1473 K, which is considered the maximum allowable temperature. Therefore, most of the argon was introduced into bulk alloy as gas contamination. Nitrogen was effectively reduced with treatment at 1323 K; however, hydrogen could not be sufficiently removed with this treatment. Residual argon and helium in bulk alloys formed bubbles at elevated temperatures and caused density decrease (swelling). Impact strengths of the bulk alloys obtained through milling in argon, hydrogen, nitrogen, and vacuum decreased with increased treatment time; more remarkable decreases were observed in the alloys including argon and hydrogen. We concluded that nitrogen is the most suitable milling atmosphere to be applied as high-temperature materials.

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The nitriding of the AA5052 aluminum alloy was carried out using an electron beam excited plasma (EBEP) technique. The specimen was characterized with respect to the following properties: crystallographic structure (X-ray powder diffraction) and the surface and cross sectional microstructures of the nitrided layer (AlN layer) observed by means of scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The AlN layer was uniformly formed on the AA5052 alloy with the thickness of around 4–5 μm. In the AlN layer, pillar-shaped grains were formed perpendicular to the surface with different orientations. The average grain size near the interface between the substrate and the AlN layer was smaller than that near the surface of the AlN layer. On the surface of the AlN layer, the nitrogen concentration was high, and in the middle of the AlN layer, it had a constant concentration like aluminum, and the concentration decreased as it approached the interface. The magnesium concentrates at the interface due to the formation of MgAl₂O₄.

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The present paper proposes an active SiC-fiber/aluminum composite utilizing its thermal deformation caused by non-uniform distribution of reinforcement fibers. A laminate of a continuous and unidirectional SiC fiber reinforced aluminum plate and an unreinforced aluminum plate was fabricated by the interphase forming/bonding method using a copper insert foil and its thermal deformation characteristics were investigated. The fabricated composite curved unidirectionally in the fiber direction by cooling from the hot pressing temperature, which is different from the behavior of a bimetal. Curvature of the composite at room temperature was maximized by investigating the effect of thickness of the aluminum plate, distance between the fibers and length of the composite in the experimented range. Under the optimum condition, its dependence on fiber length was clarified. It was also revealed that copper diffused but concentrated around the fibers, and that the interphase forming/bonding method with a copper insert increased the curvature of the composite. In a thermal cycling test, the curvature of the composite at room temperature reduced by heating from that to zero at the temperature of about 580 K. The zero-curvature temperature and the curvature at room temperature were reproducible for ten thermal cycles. These results suggest an availability of this composite as an active material.

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Measurement of electrical properties of fiber-reinforced plastics (FRPs) is an attractive method for predicting fatigue life and recording strain in FRPs. In this study, we prepared ferroelectric specimens laminated with woven carbon fabric and BaTiO₃ particulate epoxy. We measured those specimens’ electrical properties during tensile testing. Electrical resistance increased slightly and the electrical capacitance decreased as the tensile stress increased up to 500 MPa. As the tensile stress increased above 500 MPa, the electrical resistance and electrical capacitance increased. The electrical resistance and capacitance may be dependent on the applied tensile stress because of delamination between an epoxy layer and a carbon fabric layer, and fiber breakage in carbon fabric layers. Capacitance degradation from the beginning of tensile testing indicates that the addition of BaTiO₃ particles into epoxy layers induced the delamination. Subsequently, piezoelectric specimens laminated with woven carbon fabric and poled lead zirconate titanate (PZT) particulate epoxy were prepared for repeated loading tests. Thereby, we investigated the relationship between loading and piezoelectric signals. Variation in the capacitance of epoxy layers rather than that in the polarization of PZT particles may be the generation mechanism of measured signals in the piezoelectric specimens. The peak interval in the piezoelectric signal waveforms was deeply related to the applied tensile stress. The peak intensity tended to increase with loading cycles.

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