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Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya
Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12
Times Cited Count：20 Percentile：80.18(Materials Science, Multidisciplinary)The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.
Saito, Shinzo*; Tanaka, Ryuichi*; Kume, Tamikazu; Inoue, Tomio*; Takahashi, Shoji*
Genshiryoku eye, 54(5), p.34 - 41, 2008/05
no abstracts in English
Ninomiya, Hiromasa; Akiba, Masato; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko; Hosogane, Nobuyuki; Ikeda, Yoshitaka; Inoue, Nobuyuki; et al.
Journal of the Korean Physical Society, 49, p.S428 - S432, 2006/12
To contribute DEMO and ITER, the design to modify the present JT-60U into superconducting coil machine, named National Centralized Tokamak (NCT), is being progressed under nationwide collaborations in Japan. Mission, design and strategy of this NCT program is summarized.
Kikuchi, Mitsuru; Tamai, Hiroshi; Matsukawa, Makoto; Fujita, Takaaki; Takase, Yuichi*; Sakurai, Shinji; Kizu, Kaname; Tsuchiya, Katsuhiko; Kurita, Genichi; Morioka, Atsuhiko; et al.
Nuclear Fusion, 46(3), p.S29 - S38, 2006/03
Times Cited Count：13 Percentile：41.76(Physics, Fluids & Plasmas)The National Centralized Tokamak (NCT) facility program is a domestic research program for advanced tokamak research to succeed JT-60U incorporating Japanese university accomplishments. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility and mobility is pursued in aspect ratio and shape controllability, feedback control of resistive wall modes, wide current and pressure profile control capability for the demonstration of the high-b steady state.
Tsuchiya, Katsuhiko; Akiba, Masato; Azechi, Hiroshi*; Fujii, Tsuneyuki; Fujita, Takaaki; Fujiwara, Masami*; Hamamatsu, Kiyotaka; Hashizume, Hidetoshi*; Hayashi, Nobuhiko; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 81(8-14), p.1599 - 1605, 2006/02
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)no abstracts in English
Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji
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no abstracts in English
Nishinoiri, Kenji; Akasaka, Naoaki; Ogawa, Ryuichiro; Inoue, Toshihiko
no journal, ,
In fast reactor, deformation behavior and failure strength of fuel cladding tube (C/T) under loss of coolant flow (LOF) event are important evaluation items of reactor safeties. To evaluate C/T behavior under the primary phase of LOF event, transient bust examination was conducted by neutron irradiated C/T. Specimens of C/T made of PNC316 were irradiated in experimental fast reactor JOYO. In this paper reported the transient burst techniques and the results of the post irradiated examination. In the results, the failure temperature of irradiated C/T has no extreme degradation by comparison of the failure temperature of un-irradiated C/T.
Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji
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no abstracts in English
Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji
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no abstracts in English
Inoue, Toshihiko; Ogawa, Ryuichiro; Inoue, Masaki; Yoshitake, Tsunemitsu; Nishinoiri, Kenji
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Yano, Yasuhide; Ogawa, Ryuichiro; Inoue, Masaki; Yoshitake, Tsunemitsu
no journal, ,
Oxide dispersion strengthened (ODS) steel is a prospective cladding material for the advanced fuel claddings of fast reactors, owing to their excellent radiation resistance and high temperature strength capability. JAEA has been developed two types of 9Cr martensitic and 12Cr ferritic ODS steel claddings, and conducted the irradiation test for the accumulation of irradiation data as well as for the understanding of irradiation behavior of ODS steel claddings. In this study, post irradiation examination data on metallurgical examination, ring-tensile test, hardness measurement, microstructural observation, and chemical analysis were obtained, indicating that there were no significant degradations in mechanical property and also no changes in microstructure due to irradiation and that ODS steel claddings had good irradiation torelance.
Hayashi, Takehiro; Sasaki, Shinji; Mashiko, Shinichi; Yamagata, Ichiro; Ogawa, Ryuichiro; Inoue, Masaki; Yamashita, Shinichiro; Maeda, Koji
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no abstracts in English
Sekio, Yoshihiro; Yamagata, Ichiro; Yamashita, Shinichiro; Sasaki, Shinji; Ogawa, Ryuichiro; Mashiko, Shinichi; Hayashi, Takehiro; Inoue, Toshihiko; Inoue, Masaki; Maeda, Koji
no journal, ,
Corrosion and mechanical property tests utilizing spent fuel cladding made of zircaloy-2 were performed as a tentative test of the project for the purpose of simulating an environment at the very early stage after the accident in the SFP of unit 4 in the Fukushima Dai-ichi Nuclear Power Plant. The result of metallurgical investigation after corrosion test showed that no obvious changes in oxide film formed on the outer surface of cladding such as stripping occurred. In addition to that, the ring-tensile test results of samples after corrosion test was obtained and compared with that of samples before corrosion test, indicating that no significant degradation in mechanical property was confirmed. These results would have indicated that integrity of FAs was kept to be high as same as what it was before conducting corrosion test.
Hayashi, Takehiro; Sasaki, Shinji; Mashiko, Shinichi*; Yamagata, Ichiro; Ogawa, Ryuichiro; Inoue, Masaki; Yamashita, Shinichiro
no journal, ,
In the Great East Japan Earthquake, the Fukushima Dai-ichi Nuclear Power Plants (1F) were in station blackout and dropped their cooling power. To cool the plants, seawater was poured into 1F buildings included SFP. In this study, the immersion tests were carried out with artificial seawater to evaluate the effect on the strength properties of Zircaloy-2 cladding tubes in seawater. The results indicated the growth of corrosion was not observed in microstructures on the surface of Zircaloy-2 tubes, and there were no significant changes in the tensile properties.
-ray irradiationInoue, Hiroyuki*; Hata, Kuniki; Idehara, Ryuichi*; Kojima, Takao*; Kasahara, Shigeki; Iwase, Akihiro*; Ueno, Fumiyoshi
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no abstracts in English
Tanigawa, Hisashi; Ueno, Kenichi; Inoue, Ryuichi; Takeda, Nobukazu; Kakudate, Satoshi
no journal, ,
The shield blanket in ITER has an active cooling structure necessitating hydraulic connections to the cooling water manifold. To maintain or replace the blanket, welding the hydraulic connection by remote handling is necessary. Access for the welding is limited in a small hole in the first wall because of spatial constraints related to neutron and heat fluxes. A bore welding tool is required. Laser and TIG welding tools have been developed, and the welding conditions have been optimized for all position welding to horizontally located pipes. Additionally capability of re-welding between as-cut and new pipes has been confirmed. Based on the results, applicability of laser and TIG welding are comparatively assessed.
Ueno, Kenichi; Tanigawa, Hisashi; Noguchi, Yuto; Inoue, Ryuichi; Anzai, Katsunori; Kazawa, Minoru; Takeda, Nobukazu; Kakudate, Satoshi
no journal, ,
At the ITER blanket, there are cooling pipes for cooling. They will be cut and welded by dedicated tools. For the cooling pipe cutting, remaining the dust and swarf at the cooling pipe are prohibited and suitable cutting surface for rewelding are required. To implement these requirement, swage cutter cutting system were made as a mockup and tested for simulated cooling pipes. There were no cutting dust and suitable cutting surface by the swage cutter cutting system. For the blanket cooling pipe welding, relaxation for the positioning of welding groove, improvement for optical system durability of LASER welding are required. The mockup of LASER welding tool head were made and tested for simulated cooling pipes. As the result of welding condition improvement, relaxed welding condition with low sputter generation were taken. For the cooling pipe welding groove alignment, alignment tool was made and tested for correction of simulated cooling pipe misalignment. For less than 1.5mm of linear misalignment and 0.5 degree angular misalignment, required alignment accuracy were confirmed. Positioning accuracy and stability improvement at the tool operation were issued for the ITER blanket cooling pipe tool heads development.
Inoue, Ryuichi; Noguchi, Yuto; Maruyama, Takahito; Tanigawa, Hisashi; Takeda, Nobukazu; Kakudate, Satoshi
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no abstracts in English
Takeda, Nobukazu; Noguchi, Yuto; Maruyama, Takahito; Inoue, Ryuichi; Komai, Masafumi; Kozaka, Hiroshi; Tanigawa, Hisashi; Kakudate, Satoshi
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In general, nuclear fusion device requires remote maintenance system to avoid human access because of -ray emitted from structural material, which is activated by neutron of fusion reaction. The remote maintenance system was first introduced in the Joint European Torus (JET) which was constructed in UK based on international cooperation in Europe. The JET used so-called "Boom type" remote handling system which introduces articulated arm from a port. The arm is supported from the port with canti-levered and therefore the capacity is relatively low: 300 kg in JET. On the contrary, the ITER uses different type of remote handling system. The JT-60SA, which is under construction in Japan, also considers remote maintenance. This paper describes outline of remote maintenance systems for the international fusion experimental reactor, ITER.
