Search

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: 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: A dual-ion irradiation method is one of the most popular techniques for the research of microstructural evolution of materials under fusion environmental conditions. The application of oxide dispersion strengthened (ODS) ferritic steels is anticipated for realization of high temperature operation of fusion reactors above 600°C. In this research, the investigation of the microstructural evolution of high corrosion resistant ODS steels was performed using the ion irradiation methods. Dual ion-irradiation experiments were performed at the DuET facility, Kyoto University using 6.4MeV Fe self-ion and 1MeV helium ion. The irradiation temperatures were 300 and 500°C. The dose, dose rate, and helium over dpa ratio were 20dpa, 5×10−4 dpa/s, and 15appm/dpa, respectively. The post-irradiation experiments were performed using SEM, TEM and the nano-indentation methods. Significant microstructural evolution of grains and oxides were not observed after the single-ion irradiation at 20dpa, 500°C. Fine helium cavities were observed in the dual-ion irradiated 19Cr–2W–4Al ODS steel. Fine cavities of approximately several nanometers diameter were densely and uniformly formed in dual-ion irradiated range. [Copyright &y& Elsevier]

Abstract: The effects of He on the fracture behavior of reduced-activation ferritic/martensitic steels, including oxide dispersion-strengthened (ODS) steels and F82H, was determined by characterizing the microstructural evolution in and fracture behavior of these steels after He implantation up to 1000appm at around 550°C. He implantation was carried out by a cyclotron with a beam of 50MeV α-particles. In the case of F82H, the ductile-to-brittle transition temperature (DBTT) increase induced by He implantation was about 70°C and the grain boundary fracture surface was only observed in the He-implanted area of all the ruptured specimens in brittle manner. By contrast, no DBTT shift or fracture mode change was observed in He-implanted 9Cr-ODS and 14Cr-ODS steels. Microstructural characterization suggested that the difference in the bubble formation behavior of F82H and ODS steels might be attributed to the grain boundary rupture of He-implanted F82H. [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]