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Journal Articles

Challenge next-generation nuclear system; Development of oxide dispersion strengthened ferritic steel

Otsuka, Satoshi; Kaito, Takeji

Enerugi Rebyu, 39(1), p.44 - 46, 2019/01

For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.

Journal Articles

Effect of helium on irradiation creep behavior of B-doped F82H irradiated in HFIR

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

 Times Cited Count:1 Percentile:86.4(Nuclear Science & Technology)

Pressurized tubes of F82H and B-doped F82H irradiated at 573 and 673 K up to $$sim$$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 $$^{10}$$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 $$^{10}$$BN-F82H tubes was larger than that of F82H tubes. It is suggested that a swelling caused in each $$^{10}$$BN-F82H because small helium babbles might be produced by a reaction of $$^{10}$$B(n, $$alpha$$) $$^{7}$$Li.

Journal Articles

Effects of irradiation on mechanical properties of HIP-bonded reduced-activation ferritic/martensitic steel F82H first wall

Furuya, Kazuyuki; Wakai, Eiichi; Miyamoto, Kenji*; Akiba, Masato; Sugimoto, Masayoshi

Journal of Nuclear Materials, 367-370(1), p.494 - 499, 2007/08

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

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 M$$_{23}$$C$$_{6}$$ in grain boundaries of F82H steel. In result, the HIP boundary has many M$$_{23}$$C$$_{6}$$ 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.

Journal Articles

Review of recent steady-state advanced tokamak research and its further pursuit by reduction of TF ripple on JT-60U

Shinohara, Koji; JT-60 Team

Journal of the Korean Physical Society, 49, p.S56 - S62, 2006/12

The recent results of JT-60, such as the long discharge with the high normalized beta of 2.3 and the pulse length of 22.3 s which was 13 times longer than the current profile relaxation, and the observation of the increase of D$$alpha$$ emission and electron density as results of wall saturation will be reviewed. Additionally, the motivation and the design work will be reviewd on an on-going recent project of the ferritic insertion for the reduction of the toroidal field ripple. The pioneering works of JFT-2M in the ferritic insertion will be also reported.

Journal Articles

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

 Times Cited Count:111 Percentile:0.77(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Consideration on blanket structure for fusion DEMO plant at JAERI

Nishio, Satoshi; Omori, Junji*; Kuroda, Toshimasa*; Tobita, Kenji; Enoeda, Mikio; Tsuru, Daigo; Hirose, Takanori; Sato, Satoshi; Kawamura, Yoshinori; Nakamura, Hirofumi; et al.

Fusion Engineering and Design, 81(8-14), p.1271 - 1276, 2006/02

 Times Cited Count:15 Percentile:25.19(Nuclear Science & Technology)

The design guideline for the blanket is decided to meet the mission of the DEMO plant which is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. To moderate the technological extrapolation, the structural material of reduced activation ferritic steel (F82H), ceramic tritium breeder of Li$$_{2}$$TiO$$_{3}$$ and neutron multiplier of Be are introduced. To improve the economical aspect, the coolant material of the supercritical water with inlet/outlet temperatures of 280/510$$^{circ}$$C, coolant pressure of 25 MPa is chosen. Resultantly the thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirement (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

Journal Articles

Ripple reduction with ferritic insert in JFT-2M

Shinohara, Koji; Sato, Masayasu; Kawashima, Hisato; Tsuzuki, Kazuhiro; Suzuki, Sadaaki; Urata, Kazuhiro*; Isei, Nobuaki; Tani, Takashi; Kikuchi, Kazuo; Shibata, Takatoshi; et al.

Fusion Science and Technology, 49(2), p.187 - 196, 2006/02

 Times Cited Count:7 Percentile:50.82(Nuclear Science & Technology)

In JFT-2M, the toroidal field ripple was reduced by ferritic insert. Two kinds of ripple reduction were carried out. In the first case, ferritic steel was installed between toroidal field coil and vacuum vessel, just under toroidal field coil, outside vacuum vessel. In the second one, ferritic steel was installed inside vacuum vessel covering almost whole inside wall. The ripple was successfully reduced in the both cases. The temperature increment on the first wall measured by infrared TV was also reduced. A new version of OFMC code was also developed to analyze fast ion behavior in the complex structure of the toroidal field. The TF ripple reduction with ferritic insert in JFT-2M is summarized in this article.

JAEA Reports

Development of the device for 3D-measurement of the magnetic field profile in the toroidal direction

Yamamoto, Masahiro*; Tsuzuki, Kazuhiro; Kimura, Haruyuki; Sato, Masayasu; Shibata, Takatoshi; Okano, Fuminori; Suzuki, Sadaaki

JAERI-Tech 2005-060, 16 Pages, 2005/09


The low activation ferritic steel is one of the candidate for structural material of a demo-reactor. However, it was afraid that the plasma confinement and stability might degrade due to the error field by the ferromagnetic property of the ferritic material. So, on JFT-2M tokamak Advanced Material Tokamak EXperiment program (AMTEX) has been carried out to investigate about the conformity with the plasma and ferritic steel. AMTEX was performed by introducing low activation ferritic plates (FPs) step by step. At the third stage, the FPs were installed to cover almost whole inside wall (FIW) of the vacuum vessel (VV) as a simulation of the blanket wall.In this experiment the accurate measurement of the distribution of the magnetic fields strength was required before and after the installation of the FIW.Therefore, the device for 3D-measurement of the magnetic field profile along the toroidal direction was developed.

Journal Articles

Tempering treatment effect on mechanical properties of F82H steel doped with boron and nitrogen

Okubo, Nariaki; Wakai, Eiichi; Matsukawa, Shingo; Tanigawa, Hiroyasu; Sawai, Tomotsugu; Jitsukawa, Shiro; Onuki, Somei*

Materials Transactions, 46(8), p.1779 - 1782, 2005/08

 Times Cited Count:1 Percentile:82.83(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Recent technological progress for advanced tokamak research in JT-60U and JFT-2M

Hosogane, Nobuyuki; JT-60 Team; JFT-2M Group

Fusion Science and Technology, 47(3), p.363 - 369, 2005/04

 Times Cited Count:3 Percentile:73.52(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Heat treatment effects on microstructures and DBTT of F82H steel doped with boron and nitrogen

Okubo, Nariaki; Wakai, Eiichi; Matsukawa, Shingo*; Furuya, Kazuyuki; Tanigawa, Hiroyasu; Jitsukawa, Shiro

Materials Transactions, 46(2), p.193 - 195, 2005/02

 Times Cited Count:1 Percentile:100(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Extra radiation hardening and microstructural evolution in F82H by high-dose dual ion irradiation

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 $$sim$$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.

JAEA Reports

Recent accomplishment for the development of reduced activation ferritic/martensitic steels; Interim report for HFIR phase 4 with results of relating activities

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.

Journal Articles

Effects of heat treatment process for blanket fabrication on mechanical properties of F82H

Hirose, Takanori; Shiba, Kiyoyuki; Sawai, Tomotsugu; Jitsukawa, Shiro; Akiba, Masato

Journal of Nuclear Materials, 329-333(Part1), p.324 - 327, 2004/08

 Times Cited Count:53 Percentile:4.43(Materials Science, Multidisciplinary)

Reduced activation ferritic/martensitic steel, RAFs is the leading candidates for the structural materials of breeding blankets. HIP is examined as a near-net-shape fabrication process for this structure. The HIP requires heating above the normalizing temperature and the final microstructural features depends on the HIP processing conditions. Conventional HIP process caused a prior-austenite grain (PAG) coarsening of RAFs and subsequent increase of ductile brittle transition temperature. Japanese RAFs F82H and its modified steels were investigated by metallurgical method after isochronal heat treatment up to 1473K simulating HIP equivalent thermal hysteresis. Although Conventional F82H IEA heat showed significant grain growth after conventional solid HIP conditions (1313K $$times$$ 2hr.), F82H with 0.1wt.% tantalum kept fine grain after the same heat treatment. On the other hands, conventional RAF/Ms with coarse grain were recovered by the post HIP normalizing at temperature below TaC dissolution temperature. This process can refine the PAG size of F82H more than ASTM grain size number 7.

Journal Articles

Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

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

 Times Cited Count:18 Percentile:23.11(Materials Science, Multidisciplinary)

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.

Journal Articles

Impurity release and deuterium retention properties of a ferritic steel wall in JFT-2M

Ogawa, Hiroaki; Yamauchi, Yuji*; Tsuzuki, Kazuhiro; Kawashima, Hisato; Sato, Masayasu; Shinohara, Koji; Kamiya, Kensaku; Kasai, Satoshi; Kusama, Yoshinori; Yamaguchi, Kaoru*; et al.

Journal of Nuclear Materials, 329-333(Part1), p.678 - 682, 2004/08

 Times Cited Count:4 Percentile:68.32(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Investigation of compatibility of low activation ferritic steel with high performance plasma by full covering of inside vacuum vessel wall on JFT-2M

Tsuzuki, Kazuhiro; Shinohara, Koji; Kamiya, Kensaku; Kawashima, Hisato; Sato, Masayasu; Kurita, Genichi; Bakhtiari, M.; Ogawa, Hiroaki; Hoshino, Katsumichi; Kasai, Satoshi; et al.

Journal of Nuclear Materials, 329-333(1), p.721 - 725, 2004/08

 Times Cited Count:7 Percentile:52.27(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Reduced activation martensitic steels as a structural material for ITER test blanket

Shiba, Kiyoyuki; Enoeda, Mikio; Jitsukawa, Shiro

Journal of Nuclear Materials, 329-333(Part1), p.243 - 247, 2004/08

 Times Cited Count:47 Percentile:5.63(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Effect of initial heat treatment on tensile properties of F82H steel irradiated by neutrons

Wakai, Eiichi; Taguchi, Tomitsugu; Yamamoto, Toshio*; Kato, Yoshiaki; Takada, Fumiki

Materials Transactions, 45(8), p.2638 - 2640, 2004/08

 Times Cited Count:0 Percentile:100(Materials Science, Multidisciplinary)

no abstracts in English

JAEA Reports

Integrity of the first wall in fusion reactors

Kurihara, Ryoichi

JAERI-Tech 2004-052, 39 Pages, 2004/07


The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel.

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