Papers by Author: Akihiko Kimura

Abstract: A novel GBD treatment with Dy-Ni-Al eutectic alloy powder enhanced the coercivity of the sintered Nd-Fe-B magnet plate as thick as 5mm to 1760 kA/m (22 kOe) without reducing the remanence. The results of wavelength dispersive X-ray spectroscopy (WDS) indicated that this industrially epoch-making treatment spread Dy, which is a coercivity enhancing element, from the surface to the centre of the magnet through Nd-rich phase. Microstructural observations suggested that Ni and Al, which are the melting point depressants of Nd and Dy, enabled the high diffusivity of Dy.

Abstract: The effects of small amount (1 or 2 wt.%) of Ni additionson the irradiation hardening of the reduced-activation ferritic/martensitic steel, F82H, used as fusion reactor blanket structural materials were investigated by means of Fe-ion irradiation experimental test method and nano-indentation technique. The ion-irradiation hardening of Ni-added F82H is larger than that of the steel without Ni addition. The methodology to derive the irradiation hardening of ion-irradiated F82H steel was proposed from the results of hardness depth profile.

Abstract: We investigated mechanical properties of neutron irradiated Fe based binary alloys in order to extract roles of each alloying element in reactor pressure vessel (RPV) steels on irradiation hardening and annealing recovery behavior. Materials used were Pure-Fe, Fe-1Cr, Fe-1Mn, Fe-1Ni, Fe-1Cu and Fe-1Mo in at.%. Neutron irradiations were carried out at various irradiation doses from 0.3 to 8.5 × 1019 n/cm2 ( > 1.0 MeV) at 290 °C. Irradiation hardening of Fe-1Cu showed a tendency of saturation at a low dose. Irradiation hardening of Pure-Fe and the other binary alloys increased with increasing in irradiation dose. Especially, Fe-1Mn irradiated over 4.3 × 1019 n/cm2 showed significant irradiation hardening which is comparable to Fe-1Cu. However, the post-irradiation annealing recovery behavior of the irradiation hardening in Fe-Mn showed one-stage recovery at around 450 °C, which was completely different from the two-stages recovery behavior of Fe-1Cu.

Abstract: In order to estimate the long life integrity of vessel steels with considering various material compositions and irradiation conditions, it is necessary to understand physical mechanisms of the degradation of mechanical properties. In this research, chemical composition effects were investigated for Reactor Pressure Vessel Steels (RPVS) to apply small specimen test technique to surveillance test method. All specimens used in this study were machined from the A533B plate material, which are standard, low Mn, high Cu, high P, and high Cu and high P steels. Tensile strength is increased by phosphorous and copper additions. Charpy tests were carried out at temperature from 73 K to 473 K. The ductile to brittle transition temperature (DBTT) is shifted to higher temperatures with phosphorus additions accompanied by the reduction of the upper shelf energy (USE). The fracture mode of P-added A533B steels at temperatures in the lower shelf energy (LSE) region is intergranular cracking. Test results were discussed in view of the differences on elements of Cu, Mn and P.

Abstract: Liquid phase diffusion bonding between ODS steel and pure W was carried out and its joint strength was investigated for fusion applications. A block of high-Cr ODS ferritic steel and a W plate were diffusion bonded at 1240 °C for 1h with/without an insert material under an uni-axial compression load and a high vacuum atmosphere. Cross sectional microstructures of joint region were observed by scanning electron microscope and the mechanical properties of the joint region were evaluated by hardness test and torsion tests. Microstructure analysis revealed that high Cr ODS ferritic steel block and W plate with insert material was successfully diffusion bonded with free of voids. Shear strength of liquid phase diffusion bonded ODS steel and W was higher than that of directly solid state diffusion bonded ODS steel and W. This was attributed to residual strain which is resulted from the difference of thermal conductivity between the ODS steel and W.

Abstract: The SSRT behavior in hydrogen dissolved hot water was investigated for cold worked SUS316L at a strain rate of 5 x 10-7/sec. The cold work to 75% thickness reduction of the as-annealed steel resulted in the hardness increase from 150 HV to 420 HV. The tensile yield stress of the cold worked specimens (CW=75%) was about 1000 MPa and the total tensile elongation was significantly reduced from 0.8% of annealed specimen to 0.14% of the 75%CW specimen. The results of EPR tests on SUS316L steel indicated that the EPR-DOS increased with increasing sensitization period at 700°C and decreased with increasing degree of cold work or reduction in thickness. In the water with hydrogen dissolution of 0.4 ppm, many IGSCC type cracks were nucleated on the specimen side surfaces, while the fractured surface was almost TGSCC. No such a SCC as observed in hydrogen dissolved water was observed after the test in oxygen dissolved water. The susceptibility to SCC increased with increasing hydrogen content in hot water. Cold work caused the reduction of the number of surface cracks and disappearance of IGSCC.

Abstract: The oxide dispersion strengthened (ODS) ferritic steel and non-ODS reduced-activation ferritic (RAF) steel were irradiated at 773 K by means of a dual-beam ion irradiation technique to a dose of 0.4 dpa with simultaneous helium implantation up to 1000 appm. Microstructural changes were investigated by transmission electron microscopy. The RAF steel showed a preferential formation of cavities at grain boundaries, precipitate interfaces and dislocations. In contrast, the ODS ferritic steel showed a homogeneous and fine distribution of cavities in the matrix. This paper discusses the superior resistance of the ODS ferritic steel against development of cavities in terms of the effects of nano-oxide particles dispersed in the matrix.

Abstract: Two types of oxide dispersion strengthened (ODS) ferritic steels have been produced by mechanical alloying (MA) either in argon or in hydrogen atmosphere, and vacuum hot pressing (VHP). A drastic reduction in the oxygen and nitrogen contents after VHP was strongly affected by hydrogen gas used as the MA atmosphere. MA in hydrogen was found to be effective for refining the steel matrix and enhancing the tensile ductility of the ODS ferritic steels.

Abstract: Effects of neutron irradiation and thermal aging on the tensile properties and Charpy impact properties of oxide dispersion strengthened (ODS) ferritic steels for advanced nuclear systems were investigated and discussed with the results of microstructural observation. After the neutron irradiation in JMTR, significant hardening after irradiation at 290 and 400 °C as well as thermal aging at 500 °C, while no effect was observed after irradiation at 600 °C. While the irradiation hardening was not accompanied by a reduction of total elongation in tensile tests, Charpy impact energy at room temperature was reduced after the irradiation. The hardening after the irradiation at 400 °C and the aging at 500 °C was probably due to the formation of Cr-rich phases. The irradiation hardening observed in the ODS ferritic steels irradiated at 290 °C was well explained by the formation of dislocation loops.

Abstract: Oxide dispersion strengthened (ODS) materials is leading candidates for blanket/first-wall structures of the fusion reactor. ODS materials for structure application in fusion rector would allow to increase the operating temperature to approximately 650
. Therefore, this work focused on the optimization of metallurgical features to improve high temperature strength and elongation through understanding of contents of Cr and Al. In the study, the three kinds of ODS steels such as 19Cr-ODS (K1), 13Cr-Al-ODS (K2) and 19Cr-Al-ODS (K4) with Y2O3 content of 0.37wt% have been produced. And tensile test were performed on three ODS ferritic/martensitic steel between RT, 300, 400 and 600
Dispersion hardening represents an interesting approach to improve the mechanical properties at elevated temperature, as they are foreseen in the future fusion reactor It has been successfully demonstrated that it is possible to expanse the temperature range for the application of fusion reactor.