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Abstract: The mechanical strength of oxide dispersion strengthened (ODS) ferritic steels depends on the dispersion and microstructural stability of the oxide particles. Yttria is normally selected as dispersion particles to strengthen ODS ferritic steels, and other additional elements such as titanium and aluminum are added to obtain finer dispersion particles for the improvement of the mechanical properties of the ODS steels. Kyoto University is involved with the development of new ODS ferritic steels. 16Cr–0.1Ti–0.35Y2O3 ODS steel is a fusion structural material. In this research, the irradiation resistance of 16Cr–0.1Ti–0.35Y2O3 (16Cr0.1Ti) ODS steel was investigated by high-resolution TEM and by performing an ion irradiation experiment. The ODS steel was consolidated by hot extrusion at 1150°C and annealed at 1050°C for 1h for recrystallization. Oxides in 16Cr0.1Ti ODS steel were characterized by HRTEM and STEM-EDX and specified as (Y, Ti) complex oxides with an atomic ratio of Y:Ti=2:1. A 1.7MeV tandem accelerator for heavy-ion irradiation at the dual-beam material irradiation facility for energy technology (DuET), Kyoto University, was used to introduce irradiation damages in materials. The comparison of the dispersion, shape, and chemical composition of the oxides between the pre- and post-irradiated materials revealed that (Y, Ti) complex oxides were stable under the ion irradiation up to 60dpa at 650°C. [Copyright &y& Elsevier]

Abstract: The possibility of reducing the concentration of 14C produced in irradiation of reduced activation ferritic steels (RAFS) intended as structural materials in first wall and blanket of fusion power plants was investigated. During RAFS irradiation 14C is mainly produced from the more abundant of the two isotopes of nitrogen present in the steel, namely 14N. The method proposed consists in increasing the enrichment of the other isotope (15N). One-dimension transport calculations show that for a typical DEMO blanket configuration an 15N enrichment from the natural value (0.37%) to 95% was sufficient to keep the end of life 14C concentration in the RAFS below the limit (3.7×107 Bq/kg) fixed by the Japanese Nuclear Safety Commission for qualifying it as a low level material (LLM) which can be disposed by shallow land burial. [Copyright &y& Elsevier]

Abstract: About one thousands of small size specimens will be irradiated in the High Flux Test Module (HFTM) with a limited irradiation volume of 0.5l in the International Fusion Materials Irradiation Facility (IFMIF). It is necessary to verify that the experimental data of these small specimens mechanical characterization can be safely extrapolated to standard specimen data, enabling a sound dimensioning of DEMO reactor. The program of small specimen test technique (SSTT) in IFMIF/EVEDA (Engineering Validation and Engineering Design Activity) phase for fatigue, fracture toughness and crack growth measurement is summarized, and recent progress and some analysis of the experiments for small size specimens have been shown. [Copyright &y& Elsevier]

Abstract: Effects of ion-irradiation on the swelling behavior were investigated for a high-chromium oxide dispersion strengthened (ODS) ferritic steel and a 9Cr–2W reduced-activation ferritic (RAF) steel up to a high dose (nominal 60 dpa) with and without helium (900 appm) by using single and dual-ion irradiation techniques at 773K. Microstructure observations revealed that the average size and number density of cavities formed in the ODS steel are half the size and twice the density of those in the RAF steel, indicating that the ODS steel has superior resistance to swelling. Trapping of helium atoms and vacancies at the boundaries between the nano-sized oxide particles and the matrix probably accelerated nucleation of bubbles and prevented the voids from growing. It is concluded that small oxide particles in high density are effective traps for helium atoms and/or vacancies by offering a high number density of trapping sites. [Copyright &y& Elsevier]

Abstract: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed. [Copyright &y& Elsevier]