Kiyohiro Yabuuchi

Anisotropy in the Mechanical Properties of Potassium and Rhenium Doped Tungsten Alloy Plates for Fusion Reactor Applications

Abstract The effects of K-bubble dispersion and 3 wt.% Re addition on the tensile properties and their anisotropy in W were investigated in this work. K-doped W and K-doped W-3%Re showed ∼45 and ∼65% higher tensile strengths than pure W, respectively. The ultimate tensile strength and its temperature dependence in pure W, K-doped W, and K-doped W-3%Re showed anisotropy. However, the effects of K-bubble dispersion and 3% Re addition on the anisotropic tensile strength were not clearly observed. K-doped W and K-doped W-3%Re showed better deformation abilities than pure W. K-doped W-3%Re showed better tensile properties than pure W under non-irradiation conditions used in this work. Since irradiation hardening is suppressed by adding 3% Re, K-doped W-3%Re is expected to be more advantageous as a plasma facing material in a fusion reactor than pure W and K-doped W.

Neutron Irradiation Hardening of Fe-Based Binary Alloys

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.

Joint inhomogeneity in dissimilar friction stir welded martensitic and nanostructured ferritic steels

ABSTRACT Dissimilar welding between oxide dispersion strengthened ferritic (ODS) steel and reduced activation martensitic steel would be required for constructing the advanced blanket of progressive fusion reactors. In this study, we achieved dissimilar joints by friction stir welding, and aimed to characterise and ameliorate joint inhomogeneity. Main results reveal that the joint inhomogeneity is generated from discrepant microstructural evolutions within the martensitic and ODS ferritic steels. The ODS steel achieves evolution by the dynamic recrystallisation, while the martensitic steel undergoes phase transformation that drastically hardens the stir zone. By a proper post-weld heat treatment, the joint inhomogeneity can be effectively ameliorated due to carbide reprecipitation and stress relief in the joint.

Evaluation of Irradiation Hardening of Fe-Ion Irradiated F82H by Nano-Indentation Techniques

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.

Development of High Irradiation Resistant and Corrosion Resistant Oxide Dispersion Strengthened Austenitic Stainless Steels

The next generation of light water reactors, resource renewable BWR (RBWR), which can be burned trans uranium (TRU) is currently under development at Hitachi. The RBWR requires a high flux of fast neutron for efficient burning of the TRU, which is four times as large as that of the ABWR. Therefore, structural materials require both a high resistance to corrosion and to irradiation. In this study, oxide dispersion strengthened austenitic stainless steels (ODS-ASUS) with high corrosion resistance have been developed. The objective of this research is to evaluate irradiation resistance and SCC susceptibility in a simulated reactor water environment for the ODS-ASUS. The materials were irradiated with 6.4 MeV Fe3+ at 673 K up to 8.0 dpa using the DuET facility at Kyoto University. The creviced bent beam (CBB) test is conducted to assess the SCC susceptibility in the hot water, 288 °C, 8 MPa with a dissolved oxygen of 8 ppm.