Ozawa, Kazumi; Tanigawa, Hiroyasu; Morisada, Yoshiaki*; Fujii, Hidetoshi*
Fusion Engineering and Design, 98-99, p.2054 - 2057, 2015/10
Reduced activation ferritic/martensitic steel, as typified by F82H, is a promising candidate for structural material of DEMO fusion reactors. To prevent plasma sputtering, tungsten (W) coating was essentially required. This study aims to examine the irradiation effects on hardness and microstructure of vacuum-plasma-spray coated W-F82H steel, with a special emphasis on the impacts of grain-refining induced by frictional stir processing (FSP). It was revealed that the hardness of the VPS-FSP W after ion-irradiation to 5.4 dpa at 800C were not remarkably changed, where bulk W usually exhibited significant irradiation hardening.
Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu; Wakai, Eiichi; Stoller, R. E.*; Myers, J.*
Fusion Science and Technology, 68(3), p.648 - 651, 2015/10
Pressurized tubes of F82H and B-doped F82H irradiated at 573 and 673 K up to 6dpa have been measured by a laser profilometer. The irradiation creep strain in F82H irradiated at 573 and 673 K was almost linearly dependent on the effective stress level for stresses below 260 MPa and 170 MPa, respectively. The creep strain of BN-F82H was similar to that of F82H IEA at each effective stress level except 294 MPa at 573 K irradiation. For 673 K irradiation, the creep strain of some BN-F82H tubes was larger than that of F82H tubes. It is suggested that a swelling caused in each BN-F82H because small helium babbles might be produced by a reaction of B(n, ) Li.
Wakai, Eiichi; Ando, Masami; Okubo, Nariaki
Journal of Plasma and Fusion Research SERIES, Vol.11, p.104 - 112, 2015/03
The reduced-activation ferritic/martensitic (RAFM) steels for the fusion DEMO reactor have been developing from around the 1980s. RAFM steels are the first candidate materials for the first wall and blanket structure of fusion DEMO reactors, the target back-plate and the target assembly of IFMIF. In this study, two subjects had been examined and are summarized as below: (1) Effect of initial heat treatment on the microstructures and mechanical properties of RAFM steels, including irradiation damage, is very important to design the fusion DEMO reactors and also control the changes of mechanical properties after the irradiation. (2) Effects of He and H production on the microstructures and mechanical properties of RAFM steels, including irradiation damage, are essential in the evaluation of design of fusion DEMO reactor, and we have to check and evaluate them in Fusion irradiation environment like IFMIF.
Furuya, Kazuyuki; Wakai, Eiichi; Miyamoto, Kenji*; Akiba, Masato; Sugimoto, Masayoshi
Journal of Nuclear Materials, 367-370(1), p.494 - 499, 2007/08
A partial mock-up of a breeding blanket structure made of F82H steel has been successfully fabricated. In this study, microstructural observation and EDX analysis of the HIP interfaces were performed, and effects of irradiation on mechanical properties of the HIP-bonded region were also examined. Neutron irradiation was performed up to about 2 dpa at about 523 K. After the irradiation, tensile test was performed at temperatures of 295 and 523 K. The HIP interfaces possessed many precipitates, and enriched peak spectrum of chromium was detected from the precipitates. In addition, aspect of the spectrum was qualitatively equivalent to that of MC in grain boundaries of F82H steel. In result, the HIP boundary has many MC which were generally seen in grain boundaries of F82H steel. Rupture did not occur in the HIP interface. In result, it can be mentioned that bondability is maintained under the irradiation and testing conditions. The strength and elongation of the HIP-bonded region decreased somewhat in comparison with the results of an IEA standard steel.
Kulsartov, T. V.*; Hayashi, Kimio; Nakamichi, Masaru*; Afanasyev, S. E.*; Shestakov, V. P.*; Chikhray, Y. V.*; Kenzhin, E. A.*; Kolbaenkov, A. N.*
Fusion Engineering and Design, 81(1-7), p.701 - 705, 2006/02
no abstracts in English
Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Ikeda, Kazuki*; Nakamori, Yuko*; Orimo, Shinichi*; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1285 - 1290, 2006/02
Neutron transport calculations were carried out to evaluate the capability of metal hydrides and borohydrides as an advanced shielding material. Some hydrides indicated considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability to the conventional materials. From the temperature dependence of dissociation pressure, ZrH and TiH can be used without releasing hydrogen at the temperature of less than 640 C at 1 atm. ZrH and Mg(BH) can reduce the thickness of the shield by 30% and 20% compared to a combination of steel and water, respectively. Mixing some hydrides with F82H produces considerable effects in -ray shielding. The neutron and -ray shielding capabilities decrease in order of ZrH Mg(BH) and F82H TiH and F82H water and F82H.
Wakai, Eiichi; Otsuka, Hideo*; Matsukawa, Shingo; Furuya, Kazuyuki*; Tanigawa, Hiroyasu; Oka, Keiichiro*; Onuki, Somei*; Yamamoto, Toshio*; Takada, Fumiki; Jitsukawa, Shiro
Fusion Engineering and Design, 81(8-14), p.1077 - 1084, 2006/02
no abstracts in English
Suzuki, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Hirose, Takanori; Hayashi, Kimio; Tanigawa, Hisashi; Ochiai, Kentaro; Nishitani, Takeo; Tobita, Kenji; Akiba, Masato
Nuclear Fusion, 46(2), p.285 - 290, 2006/02
This paper presents significant progress in R&D of key technologies on the water-cooled solid breeder blanket for the ITER-TBM in JAERI. By the improvement of heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H, can be obtained by homogenizing it at 1150 C followed by normalizing at 930 C after the HIP process. Moreover, a promising bonding process for a tungsten armor and an F82H structural material was developed by using a uniaxial hot compression without any artificial compliant layer. Also, it has been confirmed that a fatigue lifetime correlation, which was developed for ITER divertor, can be applicable for F82H first wall mock-up. As for R&D on a breeder material, LiTiO, the effect of compression loads on thermal conductivity of pebble beds has been clarified. JAERI have extensively developed key technologies for ITER-TBM, and now steps up into an engineering R&D stage, where integrated performance of TBM structures will be demonstrated by scalable mock-ups.
Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato
Fusion Engineering and Design, 75-79, p.313 - 318, 2005/11
no abstracts in English
Wakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.
Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08
The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with B, B and B+B irradiated at 250C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He irradiation was also performed to implant about 85 appm He atoms at 120C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam DBTT for the F82H+B, F82H+B+B, and F82H+B were 40, 110, and 155C, respectively. The shifts of DBTT due to He production were evaluated as about 70C by 150 appmHe and 115C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50C.
Wakai, Eiichi; Sato, Michitaka*; Okubo, Nariaki; Sawai, Tomotsugu; Shiba, Kiyoyuki; Jitsukawa, Shiro
Nihon Kinzoku Gakkai-Shi, 69(6), p.460 - 464, 2005/06
no abstracts in English
Yoshida, Hajime; Kosaku, Yasuo*; Enoeda, Mikio; Abe, Tetsuya; Akiba, Masato
JAERI-Research 2005-003, 13 Pages, 2005/03
Hydrogen permeation fluxes of the reduced activation ferritic steel F82H were quantitatively measured by a newly proposed method, vacuum thermo-balance method, for a precise estimation of tritium leakage in a fusion reactor. We prepared sample capsules made of F82H, which enclosed hydrogen gas. The hydrogen in the capsules permeated through the capsule wall, and subsequently desorbed from the capsule surface during isothermal heating. The vacuum thermo-balance method allows simultaneous measurement of the hydrogen permeation flux by two independent methods, namely, the net weight reduction of the sample capsule and exhaust gas analysis. Thus the simultaneous measurements by two independent methods increase the reliability of the permeability measurement. The ratio of the hydrogen permeation fluxes obtained by the net weight reduction to that measured by the exhaust gas analysis was in the range from 1/4 to 1/1 in this experiment. It has been demonstrated that the vacuum thermo-balance method is effective for the measurement of hydrogen permeation rate of F82H.
Wakai, Eiichi; Taguchi, Tomitsugu; Yamamoto, Toshio*; Tomita, Hideki*; Takada, Fumiki; Jitsukawa, Shiro
Materials Transactions, 46(3), p.481 - 486, 2005/03
no abstracts in English
Taguchi, Tomitsugu; Jitsukawa, Shiro; Sato, Michitaka*; Matsukawa, Shingo*; Wakai, Eiichi; Shiba, Kiyoyuki
Journal of Nuclear Materials, 335(3), p.457 - 461, 2004/12
F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtained from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400C.
Ando, Masami; Wakai, Eiichi; Sawai, Tomotsugu; Matsukawa, Shingo; Naito, Akira*; Jitsukawa, Shiro; Oka, Keiichiro*; Tanaka, Teruyuki*; Onuki, Somei*
JAERI-Review 2004-025, TIARA Annual Report 2003, p.159 - 161, 2004/11
The objectives of this study are to evaluate radiation hardening on ion-irradiated F82H up to 100 dpa and to examine the extra component of radiation hardening due to implanted helium atoms (up to 3000 appmHe) in F82H under ratio of 0, 10, 100 appmHe/dpa.The ion-beam irradiation experiment was carried out at the TIARA facility of JAERI. Specimens were irradiated at 633 K by 10.5 MeV Fe ions with/without 1.05 MeV He ions. Micro-indentation tests were performed at loads to penetrate about 0.40 mm in the irradiated specimens using an UMIS-2000. The results are summarized as follows:1) As a result of the single irradiated F82H, the micro-hardness tended to increase about 30 dpa. 2) The extra radiation hardening was obviously caused by co-implanted helium atoms more than 1000 appm in F82H irradiated at 633 K. 3) In the dual-beam (100 appmHe/dpa) irradiated microstructure, nano-voids and fine defects were observed. It is suggested that the formation of nano-voids causes the extra radiation hardening by helium co-implantation.
JAERI-Research 2004-013, 165 Pages, 2004/09
no abstracts in English
Department of Materials Science; Department of Fusion Engineering Research (Tokai Site)
JAERI-Review 2004-018, 97 Pages, 2004/08
Extensive efforts for evaluating the irradiation performances of a reduced activation ferritic/martensitic steel (RAF/M) of F82H* and other several RAF/Ms have been made in recent several years. They are, examinations of the effects of neutron irradiation on (1) Ductile to brittle transition temperature (DBTT) up to a damage level of 20 dpa to explore lower temperature limit, (2) Enhanced He effect on DBTT shift for Ni/B doped heats (isotopic tailoring method was used for B doping), (3) Susceptibility to environmentally assisted cracking by the slow strain rate tensile tests (SSRT) in a high temperature pressurized water and (4) Flow stress-plastic strain relation obtained by measuring the profile of the specimen during tensile testing, together with the activities of (5) the development of the test methods after neutron irradiation and (6) other supporting researches. Results are summarized in the present report. They clearly indicate the good applicability of RAF/Ms to fusion machines.
Tanigawa, Hiroyasu; Hashimoto, Naoyuki*; Sakasegawa, Hideo*; Klueh, R. L.*; Sokolov, M. A.*; Shiba, Kiyoyuki; Jitsukawa, Shiro; Koyama, Akira*
Journal of Nuclear Materials, 329-333(1), p.283 - 288, 2004/08
Reduced-activation ferritic/martensitic steels (RAFs) were developed as candidate structural materials for fusion power plants. In a previous study, it was reported that ORNL9Cr-2WVTa and JLF-1 (Fe-9Cr-2W-V-Ta-N) steels showed smaller ductile-brittle transition temperature (DBTT) shifts compared to IEA modified F82H (Fe-8Cr-2W-V-Ta) after neutron irradiation up to 5 dpa at 573K. This difference in DBTT shift could not be interpreted as an effect of irradiation hardening, and it is also hard to be convinced that this difference was simply due to a Cr concentration difference. To clarify the mechanisms of the difference in Charpy impact property between these steels, various microstructure analyses were performed.
Wakai, Eiichi; Matsukawa, Shingo; Yamamoto, Toshio*; Kato, Yoshiaki; Takada, Fumiki; Sugimoto, Masayoshi; Jitsukawa, Shiro
Materials Transactions, 45(8), p.2641 - 2643, 2004/08
no abstracts in English
Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Hashimoto, Naoyuki*; Zinkle, S. J.*; Klueh, R. L.*; Koyama, Akira*
Fusion Materials Semiannual Progress Report for the Period Ending (DOE/ER-0313/35), p.33 - 36, 2004/04
Extraction replica samples were prepared from F82H-IEA, F82H HT2, JLF-1 and ORNL9Cr to analyze the precipitate distribution. The samples were examined to obtain precipitate size distribution with TEM and to analyze chemical composition distribution with SEM. Back-scattered electron imaging was found to be the effective way to separate Ta-rich precipitate from other precipitates. Results showed that most of the precipitates were M23C6, and the shape is a round ellipsoid in F82H-IEA and HT2, but was a long ellipsoid in JFL-1 and ORNL9Cr. It was also found that MX precipitates were few and large and contain Ti in F82H-IEA and HT2, but a lot of fine MX precipitates without Ti were observed in JLF-1 and ORNL9Cr.