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B
( =Yb, La) observed by multiple-wavelength neutron holographyUechi, Shoichi*; Oyama, Kenji*; Fukumoto, Yohei*; Kanazawa, Yuki*; Happo, Naohisa*; Harada, Masahide; Inamura, Yasuhiro; Oikawa, Kenichi; Matsuhra, Wataru*; Iga, Fumitoshi*; et al.
Physical Review B, 102(5), p.054104_1 - 054104_10, 2020/08
Times Cited Count：7 Percentile：54.81(Materials Science, Multidisciplinary)Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
Times Cited Count：30 Percentile：96.17(Materials Science, Multidisciplinary)Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400
C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with 
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transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.
Hayashi, Kenta; Nakamura, Takahisa; Takagi, Hisatsugu; Horie, Kaoru; Nakayama, Tamotsu; Hashimoto, Kazuhiko; Hayashi, Shoichi; Nakamura, Shinji; Takenaka, Shigeki; Ishizuka, Nobuo; et al.
Proceedings of INMM 54th Annual Meeting (CD-ROM), 10 Pages, 2013/07
JAEA and USDOE (Los Alamos National Laboratory (LANL)) have been collaborating on spent fuel measurements with a PNAR/SINRD NDA instrument at Fugen, in the course of the NGSI Spent Fuel Nondestructive Assay Project. In this collaboration, LANL's role has been to design and fabricate the detector (integrated PNAR and SINRD system), while JAEA's role has been to undertake the installation of the detector at the appropriate position in the spent fuel pool and to prepare for the actual measurements. In this paper we report the rehearsal of the measurement using a mock-up detector and a dummy fuel assembly in December 2012 and the plan of actual measurements in June 2013 (at the time of submission of this paper).
Uchida, Shinichi; Yuasa, Wataru; Hayashi, Akihiko; Inose, Shoichi; Ouchi, Yuichiro; Asakawa, Kenichi*; Uchikoshi, Yuta*
Kaku Busshitsu Kanri Gakkai (INMM) Nihon Shibu Dai-32-Kai Nenji Taikai Rombunshu (Internet), 7 Pages, 2011/11
JAEA has developed a TMS which can monitor the movement of nuclear material convoys to make safe and proper transport of nuclear materials. The TMS mainly consists of the location information monitoring system to monitor the location of the convoys and the visual information monitoring system to survey around the convoys. The TMS can send information in real-time to the TCC located at the shipper site. The JAEA has operated the TMS for ground transportation of MOX fuels since 2005, and the JAEA solved visual control problems that were observed during the operational experience and upgraded the system by adding the automatic communication control system, etc. In the case of emergency during transport, the TMS can send much more detailed visual information of the accident site to the TCC, which is useful for planning and executing an effective response. This paper reports the overview of the upgraded TMS and its effectiveness.
Yuasa, Wataru; Uchida, Shinichi; Hayashi, Akihiko; Inose, Shoichi; Ouchi, Yuichiro
Proceedings of INMM 52nd Annual Meeting (CD-ROM), 6 Pages, 2011/07
The Japan Atomic Energy Agency (JAEA) has developed a transport monitoring system (TMS), which tracks the movement of nuclear material convoys in real time, and which is consistent with the physical protection requirements for transport of Category 1 nuclear materials. The TMS has been used for land transport of MOX fuels since 2005. JAEA solved some problems that were observed during operational experience and upgraded the system by adding the following features. In the case of emergency during transport, the upgraded TMS can send much more detailed visual information of the accident site to the TCC, which is useful for planning and executing an effective response. This paper presents the features of the TMS and the results obtained from operational experiences.
Yano, Yasuhide; Inoue, Toshihiko; Otsuka, Satoshi; Furukawa, Tomohiro; Kato, Shoichi; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*
no journal, ,
no abstracts in English
Kato, Shoichi; Furukawa, Tomohiro; Otsuka, Satoshi; Yano, Yasuhide; Inoue, Toshihiko; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*
no journal, ,
In order to evaluate the fracture limit of the cladding material made by ODS at the severe accident condition, the mechanical strength tests have been performed at elevated temperature. In this meeting, the research plan and the progress on the mechanical strength under this research project is presented. In addition, the technical development result concerning the 1000C creep apparatus prepared for this research is also reported.
Yano, Yasuhide; Kato, Shoichi; Otsuka, Satoshi; Inoue, Toshihiko; Tanno, Takashi; Oka, Hiroshi; Furukawa, Tomohiro; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; et al.
no journal, ,
no abstracts in English
Kato, Shoichi; Furukawa, Tomohiro; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Oka, Hiroshi; Inoue, Toshihiko; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
no journal, ,
Oxide dispersion strengthened (ODS) steel is a prime candidate for cladding tubes of Japan Sodium-cooled Fast Reactor (JSFR) due to the high temperature and radiation resistances. One of the safety design of JSFR for Design Extension Condition (DEC) is the control of severe plant conditions, including prevention of severe accidents and mitigation of severe-accident consequences. Therefore, it is necessary to acquire the mechanical properties at ultra-high temperature conditions for core materials to evaluate safety design. There are, however, no data for ODS claddings at ultra-high temperature condition for the reflecting to the design criteria. In this study, creep rupture tests of 9Cr-ODS, 12Cr-ODS and FeCrAl-ODS steel claddings have been done at elevated temperatures, and the effect of minor elements such as Al, Zr and O on the mechanical strength and the creep rupture curve for the safety design were evaluated. The effect of minor elements was estimated based on the data at 700C and 1000C. As the results, it was confirmed that the addition of Zr had an effect on the improvement of creep strength at elevated temperature for the FeCrAl-ODS steel claddings.
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
no journal, ,
no abstracts in English
Yano, Yasuhide; Kato, Shoichi; Otsuka, Satoshi; Uwaba, Tomoyuki; Sekio, Yoshihiro; Inoue, Toshihiko; Furukawa, Tomohiro; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
no journal, ,
no abstracts in English
Kato, Shoichi; Furukawa, Tomohiro; Otsuka, Satoshi; Yano, Yasuhide; Inoue, Toshihiko; Kaito, Takeji; Kimura, Akihiko*; Torimaru, Tadahiko*; Hayashi, Shigenari*; Ukai, Shigeharu*
no journal, ,
An evaluation on tolerance to failure of existing ODS ferritic steel claddings at the accident condition is important from the viewpoint of the reactor safety. This paper describes the high temperature strength of the 9/12Cr-ODS steels for fast reactors and the FeCrAl-ODS steels for light water reactors.
Hayashi, Shoichi; Matsushima, Akira; Kawagoe, Shinji; Soejima, Goro; Awatani, Yuto; Isomi, Kazuhiko
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no abstracts in English
Nomoto, Yasunobu; Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; et al.
no journal, ,
no abstracts in English
