Hirose, Takanori; Sokolov, M. A.*; Ando, Masami; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Stoller, R. E.*; Odette, G. R.*
Journal of Nuclear Materials, 442(1-3), p.S557 - S561, 2013/11
Hirose, Takanori; Okubo, Nariaki; Tanigawa, Hiroyasu; Ando, Masami; Sokolov, M. A.*; Stoller, R. E.*; Odette, G. R.*
Journal of Nuclear Materials, 417(1-3), p.108 - 111, 2011/10
This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels irradiated at 573 K. The materials used in this research were F82H-IEA and its modified steels. Post irradiation mechanical tests revealed that irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated proof stress is less than 1 GPa. The deterioration of total elongation was also saturated by 9 dpa. Irradiation response of F82H-mod3, which is stable to temperature instability during material production and HIP treatment, was very similar to that of F82H-IEA, and negative impacts of extra tantalum was not observed. Therefore it can be an attractive option for the structural materials for blanket components manufactured by HIP.
Okubo, Nariaki; Sokolov, M. A.*; Tanigawa, Hiroyasu; Hirose, Takanori; Jitsukawa, Shiro; Sawai, Tomotsugu; Odette, G. R.*; Stoller, R. E.*
Journal of Nuclear Materials, 417(1-3), p.112 - 114, 2011/10
Irradiation hardening and fracture toughness of reduced-activation ferritic/martensitic steel F82H after irradiation were investigated with a focus on changing the fracture toughness transition temperature as a result of several heat treatments. The specimens were standard F82H-IEA (IEA), F82H-IEA with several heat treatments (Mod1 series) and a higher tantalum containing (0.1%) heat of F82H (Mod3). The specimens were irradiated up to 18 dpa at 300 C in High Flux Isotope Reactor under a collaborative research program between JAEA/US-DOE. The results of hardness tests showed that irradiation hardening of IEA was comparable with that of Mod3. However, the fracture toughness transition temperature of Mod3 was lower than that of IEA. The transition temperature of Mod1 was also lower than that of the IEA heat. These results suggest that tightening of specifications on the heat treatment condition and modification of the minor alloying elements seem to be effective to reduce the fracture toughness transition temperature after irradiation.
Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Mslang, A.*; Stoller, R. E.*; Lindau, R.*; Sokolov, M. A.*; Odette, G. R.*; Kurtz, R. J.*; Jitsukawa, Shiro
Journal of Nuclear Materials, 417(1-3), p.9 - 15, 2011/10
ITER construction was started, and R&D toward DEMO shifted to more practical stage. On this stage, the candidate material for DEMO blanket have to be the one which have sound engineering bases to be ready for engineering designing activity for DEMO reactor in 10 years. Reduced activation ferritic/martensitic (RAFM) steels, such as F82H (Fe-8Cr-2W-0.2V-0.04Ta) or EUROFER97 (Fe-9Cr-1W-0.2V- 0.12Ta), is the only material which currently have enough potential to meet this requirement, and selected as the target material in the R&D on materials engineering for DEMO blanket under the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between EU and Japan. In this paper, current status of RAFM R&D is overviewed especially on fabrication technology, inspection/testing technology, and material database. Overview on irradiation effect study is also provided.
Baluc, N.*; Gelles, D. S.*; Kimura, Akihiko*; Jitsukawa, Shiro; Klueh, R. L.*; Odette, G. R.*; Van der Schaaf, B.*; Jinnan, Y.*
Journal of Nuclear Materials, 367-370(1), p.33 - 41, 2007/08
Recent research results obtained in Europe, Japan, China and the USA on reduced-activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen.
Jitsukawa, Shiro; Tamura, Manabu*; Van der Schaaf, B.*; Klueh, R. L.*; Alamo, A.*; Petersen, C.*; Schirra, M.*; Spaetig, P.*; Odette, G. R.*; Tavassoli, A. A.*; et al.
Journal of Nuclear Materials, 307-311(Part1), p.179 - 186, 2002/12
Reduced activation ferritic/martensitic steel is the primary candidate structural material for ITER Test Blanket Modules and DEMOnstration fusion reactor because of its excellent dimensional stability under irradiation and lower residual activity as compared with the Ni bearing steels such as the austenitic stainless steels. In this paper, microstructural features, tensile, fracture toughness, creep and fatigue properties of a reduced activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) are reported before and after irradiation, in addition to the design concept used for development of this alloy. A large number of collaborative test results including those generated under the IEA working group implementing agreements are collected and are used to evaluate the feasibility of use of F82H steel as one of the reference alloys. The effect of metallurgical variables on the irradiation hardening is reviewed and compared with the results obtained from irradiation experiments.