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The present study reports the corrosion behavior of a reduced-activation ferritic martensitic steel F82H exposed to high temperature pressurized water for 28 and 100 h using a rotating disk apparatus at rotation speeds of 500 and 1000 rpm at a temperature of 573 K under a water pressure of 15 MPa with <20 ppb dissolved oxygen. Static tests without rotation increased the weight of sample with time, whereas rotating tests decreased the weight of sample with the exposure time and rotating speed. Oxide particles in the size of several micron meters on the surface of static samples were not observed on the rotated samples which became covered with a thin Cr-rich oxide layer. These results suggest an occurrence of erosion–corrosion and/or flow-accelerated corrosion of F82H under the rotating condition. [ABSTRACT FROM AUTHOR]

Abstract: The effects of alloying elements and temperature on corrosion behavior in supercritical water (SCW) were investigated for oxide dispersion strengthened (ODS) steels to design alloy compositions for corrosion-resistant ODS ferritic steels. Corrosion tests of the ODS steels were carried out at 400, 500, and 600°C under 25MPa with 8ppm of dissolved oxygen. The addition of 4wt.%Al is effective to improve the corrosion resistance of 16Cr-ODS steel, suggesting that an adequate combination between Cr and Al content ranges (14–16)Cr and (3.5–4.5)Al. The addition of Ce or Zr to W-added ODS steel is expected to increase its corrosion resistance. At higher temperatures, the addition of 4wt.%Al to 16Cr-ODS steel induces better corrosion resistance rather than the temperature of 400°C. It is considered that for 16Cr-ODS steels with Al, the continuous alumina mainly suppresses the inward oxygen diffusion and the Fe–Cr spinel adds to the suppression of oxygen diffusion. [Copyright &y& Elsevier]

Abstract: Anisotropic fracture behavior of SOC-1 oxide dispersion strengthened (ODS) ferritic steel has been investigated for a hot-extruded bar by tensile tests and Charpy impact tests. These mechanical properties are better in the longitudinal direction than in the transverse directions against extrusion direction (ED). Fracture surface observations by scanning electron microscopy and auger electron spectroscopy indicated bundle-like morphology with existence of segregation/precipitation/inclusions along ED. Pole figures of the hot-extruded bar characterized using electron back scattering diffraction (EBSD) technique and X-ray diffraction exhibited 〈110〉 fiber texture formation along ED. The EBSD orientation map showed a complex bundle-like grain morphology which consists of elongated grains having a specific orientation 〈110〉// ED and relatively isotropic and small grains having other orientation. The results conclude that the combined effects of observed elongated grain morphology and these small grains with segregation/precipitation/inclusions along ED can explain the anisotropic fracture behavior of the hot-extruded ODS ferritic steel. [Copyright &y& Elsevier]

Abstract: The effect of aging on the ductile to brittle transition behavior of recrystallized high-Cr ODS ferritic steels were investigated using small-punch test technique. In order to reduce the anisotropy of microstructure in the as-received ODS steels, they were recrystallized after cold rolling. The ODS steel without Al hardly recrystallized at 1200°C for 90min, while the ODS steel with Al significantly recrystallized. Ductile–brittle transition curves were drawn for as received, recrystallized, and thermally aged (at 450°C for 1440h). Before aging, both materials had good impact properties with low ductile–brittle transition temperatures and high USEs. The ductile–brittle transition temperatures are not dependent of the grain size. The aging degraded their impact properties for ODS steel with Al. However, almost no degradation occurred for the ODS steel without Al. SEM observation showed an influence of the cold-rolling direction on the fracture behavior especially at low temperatures. [Copyright &y& Elsevier]

Abstract: A successful example of high-Cr oxide dispersion strengthened (ODS) steels development is introduced with showing key technologies to overcome the issues to meet material requirements for next generation nuclear systems as well as fusion blanket systems. Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires it less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead–bismuth eutectics (LBE), while it is detrimental to high-temperature strength. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 973K in Al-added ODS steels. Feasibility of high-Cr ODS steel without Al addition is assessed for fusion application in terms of corrosion resistance in SCW. [Copyright &y& Elsevier]

Abstract: The corrosion behaviour of 16%Cr and 16%Cr–4%Al ODS ferritic steels in different heat treatment conditions has been investigated in a supercritical water environment. The exposed coupons were analyzed using scanning electron microscopy (SEM), electron probe micro analysis (EPMA), Auger and X-ray diffraction analysis (XRD). It was found that the formation of oxides depends on the chemical composition and not on the metallurgical condition. The Al-free alloys formed a monolayer oxide film composed of (Cr,Fe)2O3. The Al-containing alloys formed an oxide film composed of an outer layer of hematite and magnetite and an inner layer of Al2O3. The oxidation mechanisms are discussed. [ABSTRACT FROM AUTHOR]

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 effects of neutron irradiation on the tensile properties of oxide dispersion strengthened (ODS) ferritic steels have been investigated. Various ODS steels containing 14–22wt% Cr and 4.5wt% Al or no Al were produced by mechanical alloying. In order to investigate their susceptibility to hardening and embrittlement induced by neutron irradiation, the ODS steels were irradiated in JMTR at 290, 400 and 600°C up to doses of 0.01, 0.3 and 0.75dpa, respectively. Tensile tests were performed at RT before and after irradiation. Microstructure observations were conducted in TEM. It was found that high-Cr-ODS steels show a significant hardening after irradiation at 290 and 400°C, while no effect was observed after irradiation at 600°C. The irradiation hardening, however, is not accompanied by a reduction of total elongation. After irradiation at 400°C the measured hardening increases with Cr concentration, while it is almost constant versus Cr concentration after irradiation at 290 and 600°C. [Copyright &y& Elsevier]

Abstract: Pre- and post-deformation microstructures and tensile properties of the newly-developed oxide dispersion strengthened (ODS) ferritic steels, K1 (Fe–19Cr–0.3W–0.3Ti–0.3Y2O3) and K4 (Fe–19Cr–4Al–2W–0.3Ti–0.3Y2O3) for nuclear fusion reactors and Generation IV nuclear fission systems were investigated. The nano-oxides of the K1 were cubic pyrochlore Y2Ti2O7, while those of K4 were mainly perovskite AlYO3, which was confirmed by transmission electron microscopy and X-ray diffraction (XRD) analyses. K1 was found to have much higher strength than K4 and the difference in the nano-oxides of K1 and K4 can account for this difference. Most of the extracted residues in the K1 and K4 identified by XRD consisted of M23C6 carbides. [Copyright &y& Elsevier]

Abstract: The reduced activation ferritic steel (RAFS) has been selected as structural material for a variety of blanket systems for ITER test blanket modules (TBM). In the evaluation of integrated performance of ferritic steels as structural components of blanket systems, there are unique issues as well as common issues for each blanket system. One of the unique issues for each system is the compatibility of ferritic steels with the coolant materials. The corrosion rate of ferritic steels in hot water, super critical pressurized water (SCPW), humid air, Pb–17Li, lithium and Flibe at various temperatures is reviewed in this work. Efforts to improve corrosion resistance have been made, taking the alloy design into account. A dispersion of yttria was effective to improve corrosion resistance of a RAFS. The compatibilities of RAFSs with hot water, Pb–17Li, lithium and Flibe are considered to be good enough for the TBM applications. The liquid metal embrittlement (LME) is considered to be a critical issue for the utilization of RAFSs for the lithium systems. Several issues towards DEMO and beyond are shown from the compatibility point of view. [Copyright &y& Elsevier]

Four different sizes of compact tension (CT) specimens of a reduced activation ferritic steel, JLF-1 steel, were fabricated and the fracture toughness was measured at room temperature by means of the unloading compliance method referring to in the ASTM E1820-99a. The JQ values obtained for the 1T–1CT and 1/2T–1CT specimens were 404 and 623 kJ/m2, respectively, indicating that the JQ value increased with decreasing the specimen thickness. On the other hand, the JQ value decreased when specimens were miniaturized while maintaining the similar proportions. The JQ values for the 1/2CT and 1/4CT specimens were 371 and 321 kJ/m2, respectively. Specimen size effects are interpreted in terms of an increase in the plain stress state region and plastic zone size at the crack tip in the specimen. [Copyright &y& Elsevier]

Abstract: The irradiation hardening behavior of reduced-activation ferritic steels after single Fe-ion beam irradiation and dual-ion (Fe ion and He ion) beam irradiation experiments was investigated with nanoindentation tests. The ion-irradiation experiments were conducted at 563K with 6.4MeV Fe3+ ions up to 3dpa at a 600nm depth from the irradiated surface. Furthermore, these experiments were conducted with and without simultaneous energy-degraded 1MeV He+ ions up to 300appm. The materials used were F82H, F82H+1Ni, and F82H+2Ni to investigate the effect of Ni addition on the irradiation hardening behavior. The measured nanoindentation hardness was converted to the bulk-equivalent hardness based on a combination of the Nix–Gao model to explain the indentation size effect and the composite hardness model to explain the softer substrate effect of the nonirradiated region beyond the irradiated depth range. It is clearly shown that the Ni addition enhances the irradiation hardening of F82H. The bulk-equivalent hardness is compared with the experimentally obtained Vickers hardness of F82H steels after neutron irradiation. The effect of simultaneously implanted helium on the irradiation hardening is negligible in the investigated irradiation conditions. [Copyright &y& Elsevier]

Abstract: In the present work, microstructures of dispersoids in two Al-added high-Cr ODS ferritic steels (including the commercial MA956, and another ODS steel with a similar composition recently developed in Kyoto University) were studied. The study with transmission electron microscopy (TEM) of both the thin-foil specimens and the extraction replica specimens showed that the nano-sized oxides in the two ODS steels consist mainly of a complex Y–Al–O compound (YAlO3) in a perovskite structure (i.e. YAP), while other oxide particles with different chemical composition or crystalline structures are in significantly lower number density. The comparison between the two kinds of ODS steels shows that the size distribution and the number density of the oxide particles significantly depends on the conditions of fabrication. Mechanisms underlying the processes were discussed. [Copyright &y& Elsevier]

Abstract: The mechanical strength of oxide dispersion strengthened (ODS) ferritic steels depends on the oxide particles, such as Yttria. Other elements such as titanium and aluminum are added to refine the dispersed particles of ODS steels. The addition of elements results in the formation of complex oxides such as (Y, Al) and (Y, Ti) oxides. Because the ODS steels are proposed for the higher temperature operation of fusion applications, the stability of the complex oxides under irradiation environments is important. In this work, a microstructural stability investigation under heavy-ion irradiation was carried out for ODS steels focusing on the morphology of (Y, Ti) and (Y, Al) complex oxides. The 19Cr ODS steel strengthened by (Y, Ti) complex oxides and 19Cr–2W–4Al steel with (Y, Al) complex oxides were ion-irradiated and investigated using TEM. The results revealed that the complex oxides were stable after the ion irradiation at 670°C and 150dpa. [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]