Katsunori Abe

Microstructural development of neutron irradiated W–Re alloys

Abstract Tungsten (W) alloys are candidate materials to be used as high-heat-flux materials in fusion reactors. In our previous work, W–26 wt% Re showed drastic hardening and embrittlement after the neutron irradiation. In this study, to clarify the irradiation hardening and embrittlement behavior of W–26 wt% Re, from the viewpoint of microstructural development, the microstructure observation of the neutron irradiated W–26 wt% Re was carried out using transmission electron microscope (TEM). The specimens were irradiated at the materials open test assembly of the fast flux test facility (FFTF/MOTA-2A cycle 11) up to ∼1×10 27 n/m 2 , (En>0.1 MeV). The irradiation temperatures were 646, 679, 792, 873 and 1073 K. In all neutron irradiated W–26 wt% Re samples, sigma-phase precipitates and chi-phase precipitates were observed, while in the thermally aged specimen, only sigma-phase precipitates were observed. Irradiation effects on microstructural development are discussed.

Tensile behavior and microstructure of neutron-irradiated Mo-5% Re alloy

Abstract This work reports the effect of heat treatment on the tensile behavior and and microstructure of neutron-irradiated Mo-5% Re alloy. Stress-relived and recrystallized specimens conditions were irradiated at five temperatures between 646 and 1073 K in FFTF/MOTA. The exposure levels were in the range of 6.8 to 34 dpa depending on the irradiation temperatures. Tensile tests were carried out at room temperature and 673 K and microstructures of the irradiated specimens were observed by TEM. The Mo-5% Re alloy irradiated at high temperatures shows ductile behavior even at room temperature. The total elongation of stress-relived specimens irradiated at 873 and 1073 K ranged from 5 to 10%, and that of recrystallized specimens irradiated at 1073 K was 5%. The fracture modes of these specimens were transgranular type. Voids were observed in all of the irradiated specimens, but precipitates were found only in specimens irradiated above 792 K. It is important for the Mo Re alloy to be used in high-heat flux components of fusion reactors that the alloy showed the ductility after neutron exposures of relatively high fluences.

R&D of a MW-class solid-target for a spallation neutron source

Abstract R&D for a MW-class solid target composed of tungsten was undertaken to produce a pulsed intense neutron source for a future neutron scattering-facility. In order to solve the corrosion of tungsten, tungsten target blocks were clad with tantalum by means of HIP’ing, brazing and electrolytic coating in a molten salt bath. The applicability of the HIP’ing method was tested through fabricating target blocks for KENS (spallation neutron source at KEK). A further investigation to certify the optimum HIP conditions was made with the small punch test method. The results showed that the optimum temperature was 1500 °C at which the W/Ta interface gave the strongest fracture strength. In the case of the block with a hole for thermocouple, it was found that the fabrication preciseness of a straight hole and a tantalum sheath influenced the results. The development of a tungsten stainless-steel alloy was tried to produce a bare tungsten target, using techniques in powder metallurgy. Corrosion tests for various tungsten alloys were made while varying the water temperature and velocity. The mass loss of tungsten in very slow water at 180 °C was as low as 0.022 mg/y, but increased remarkably with water velocity. Simulation experiments for radiation damage to supplement the STIP-III experiments were made to investigate material hardening by hydrogen and helium, and microstructures irradiated by electrons. Both experiments showed consistent results on the order of the dislocation numbers and irradiation hardness among the different tungsten materials. Thermal-hydraulic designs were made for two types of solid target system of tungsten: slab and rod geometry as a function of the proton beam power. The neutronic performance of a solid target system was compared with that of mercury target based on Monte Carlo calculations by using the MCNP code.

Analysis and measurement of residual stress distribution of vanadium/ceramics joints for fusion reactor applications

Abstract Vanadium alloys are considered as candidate structural materials for fusion reactor system. When vanadium alloys are used in fusion reactor system, joining with ceramics for insulating is one of material issues to be solved to make component of fusion reactor. In the application of ceramics/metal jointing and coating, residual stress caused by difference of thermal expansion rate between ceramics and metals is an important factor in obtaining good bonding strength and soundness of coating. In this work, residual stress distribution in direct diffusion bonded vanadium/alumina joint (jointing temperature: 1400°C) was measured by small area X-ray diffraction method. And the comparison of Finite Element Method (FEM) analysis and actual stress distribution was carried out. Tensile stress concentration at the edge of the boundary of the joint in alumina was observed. The residual stress concentration may cause cracks in alumina, or failure of bonding. Actually, cracks in alumina caused by thermal stress after bonding at 1500°C was observed. The stress concentration of the joint must be reduced to obtain good bonded joint. Lower bonding temperature or to devise the shape of the outer surface of the joint will reduce the stress concentration.

Tensile properties of a series of V–4Ti–4Cr alloys containing small amounts of Si, Al and Y, and the influence of helium implantation

Abstract V–Ti–Cr type alloys containing Si, Al and Y have been developed to improve proof-oxidation properties and high temperature strength. In order to optimize the composition of Si, Al and Y, seven V–Ti–Cr type alloys containing Si, Al and Y up to 0.5 wt% were prepared. Tensile tests were carried out at temperatures from 300 to 1123 K. From the results of the tests, the dependence of yield stress and ultimate tensile strength on Si, Al and Y concentration is low at these temperatures. Total elongation of the alloys tested at 923 and 1123 K increase with increasing concentration of Si, Al and Y. Helium implantation up to about 50 appm by an accelerator was also carried out. The loss of total elongation at 923 and 1123 K depends on the concentration of Si, Al and Y. From the results of helium implantation, it is suggested that the appropriate concentration of Si, Al and Y for V–4Ti–4Cr alloys is between 0.1 and 0.5 wt%, respectively.

Oxidation and hardness profile of V–Ti–Cr–Si–Al–Y alloys

Abstract Several alloys of composition V–4Ti–4Cr–(0–0.5)Si–(0–0.5)Al–(0–0.5)Y were selected for oxidation in air. In addition a V–4Ti–4Cr–0.5Si–0.5Al–0.5Y alloy was pre-implanted with helium to study the effect of helium on oxidation behavior. Rapid oxidation experiments in air were conducted on tensile specimens and disc bend specimens at 300°C, 500°C, 600°C and 700°C for 1 h. After oxidation, tensile specimens were mounted in cross-section, and micro-indentation hardness tests were conducted. The depths of increased hardness obtained by indentation tests and the thickness of cleavage fracture zones measured by using scanning electron microscopy (SEM) micrographs were similar. The hardness profile of the cross-section in the depth direction corresponded to a model of oxygen diffusion in vanadium. Three-point bend tests were conducted on helium-implanted and air- exposed disc specimens. The bending stress of the specimen with 700 appm helium was larger than that with 70 appm helium.

Fabrication using a levitation melting method of V–4Cr–4Ti–Si–Al–Y alloys and their mechanical properties

Reduction of interstitial impurities such as O and N is a potential method to improve various properties of vanadium alloys. It was shown that addition of Si, Al and Y was useful to reduce the oxygen concentration and to improve post-irradiation ductility at relatively low temperatures for V-Cr-Ti alloys. Several 2.5 kg alloys of V-4Cr-4Ti-Si-Al-Y type were fabricated by using a levitation melting method. Charpy impact test by an instrumented testing machine has been conducted using miniaturized specimens. Tensile tests have been carried out before and after neutron irradiation. The miniaturized specimens were irradiated up to 8 x 10 22 n/m 2 (E > 1 MeV) at 290 °C in Japan Materials Testing Reactor. By adopting a levitation melting method, high-purity kg-scale ingots of V-4Cr-4Ti-Si-Al-Y alloys with ∼80 ppm C, <170 ppm O and ∼110 ppm N were obtained. The V-4Cr-4Ti-0.1Si-0.1Al-0.1Y alloy fabricated in this study showed good impact properties compared with a previous laboratory-scale alloy. This alloy showed good tensile properties even after neutron irradiation at 290 °C. Levitation melting can be adopted to produce large ingots of V-Cr-Ti-Si-Al-Y type alloys by controlling the amount of yttrium addition. In this study, the technology for fabrication of high-purity kg-scale ingots of V-4Cr-4Ti-Si-Al-Y alloy has been demonstrated, and has made it possible to investigate systematically various properties of the alloy.

Behavior of helium gas atoms and bubbles in low activation 9Cr martensitic steels

Abstract The behavior of helium-gas release from helium-implanted 9Cr martensitic steels (500 appm implanted at 873 K) during tensile testing at 873 K was studied. Modified 9Cr-1Mo, low-activation 9Cr-2W and 9Cr-0.5V were investigated. Cold-worked AISI 316 austenitic stainless steel was also investigated as a reference which was susceptible helium embrittlement at high temperature. A helium release peak was observed at the moment of rupture in all the specimens. The total quantity of helium released from these 9Cr steels was in the same range but smaller than that of 316CW steel. Helium gas in the 9Cr steels should be considered to remain in the matrix at their lath-packets even if deformed at 873 K. This is the reason why the martensitic steels have high resistance to helium embrittlement.

Molecular dynamics simulation of vanadium using an interatomic potential fitted to finite temperature properties

Abstract Elastic constants of vanadium are obtained by means of molecular dynamics simulation of stress–strain response. Results at finite temperature are compared with experiments to provide validation of the interatomic potential. Previously, experimental elastic constants and lattice parameter, obtained at finite temperature, were used for parameter fitting at 0 K, leading to poor agreement in the temperature variation. In this work four out of the 11 parameters in modified embedded atom method potential of Baskes have been re-fitted. The calculated elastic properties and the lattice parameter now agree well with experiment.

Defect microstructure and deformation behavior of V–Ti–Cr–Si–Al–Y alloy irradiated in ATR

Abstract Low-temperature irradiation performance of vanadium alloys, especially after neutron irradiation below 400°C, is a major concern. Defect microstructures and deformation behavior of V–Ti–Cr type alloys containing small amounts of Si, Al and Y after neutron irradiation below 400°C were studied. Neutron irradiation was conducted in the advanced test reactor (ATR) to fluences of 0.7–4.7 displacements per atom (dpa) at temperatures from 141°C to 293°C. A large amount of hardening was observed in the specimens irradiated at 290°C or below, which corresponded to a high density of small dislocation loops. The density and the size of loops depended on the irradiation temperature and the pre-irradiation annealing temperature. The loops in the specimen annealed at 900°C had a lower density and a larger size than those annealed at 1000°C and 1100°C. The specimen annealed at 900°C showed very small hardening and retained good work-hardening capability after irradiation at 290°C.