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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：32 Percentile：96.97(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.
Kurikami, Hiroshi; Niizato, Tadafumi; Tsuruta, Tadahiko; Kato, Tomoko; Kitamura, Akihiro; Kanno, Mitsuhiro*; Kurosawa, Naohiro*
JAEA-Research 2016-020, 50 Pages, 2017/01
JAEA-Research-2016-020.pdf:6.02MB
In this report, we developed a compartment model of radionuclide migration in environment based on exposure pathways in a river basin scale and performed a preliminary calculation. The results showed good agreement with some measurement, although the comparison of bed sediment, transportation to outer sea and to agricultural products with the measurement was not enough. We continue to validate the model.
Oi, Motoki; Kai, Tetsuya; Meigo, Shinichiro; Kinoshita, Hidetaka; Sakai, Kenji; Sakamoto, Shinichi; Kaminaga, Masanori; Kato, Takashi; Kato, Tadahiko*
Europhysics Conference Abstracts, 29J, 6 Pages, 2005/00
The 3GeV proton beam transport facility (3NBT) is a high-intensity proton beam line from the 3GeV Rapid Cycling Synchrotron (RCS) to the Material and Life science Facility (MLF) at Japan Proton Accelerator Research Complex (J-PARC). In order to allow hands-on maintenance, a design criterion has been that the average beam loss at 3NBT be less than 1W/m. The systems for beam monitoring and magnet power control play an important role. In J-PARC, the Experimental Physics and Industrial Control System (EPICS) [1] will be used for the main control system. For the proton beam monitor system of 3NBT, EPICS is used and it has to work at 25Hz. In this study, a data acquisition system for the proton beam monitors is presented that has been developed with EPICS. Its performance has been investigated under 25Hz frequency condition.
Tamura, Fumihiko; Yoshikawa, Hiroshi; Yoshii, Masahito*; Chiba, Junsei*; Kato, Tadahiko*; Takagi, Akira*
Proceedings of 1st Annual Meeting of Particle Accelerator Society of Japan and 29th Linear Accelerator Meeting in Japan, p.677 - 679, 2004/08
We describe the overvier of the J-PARC timing system. J-PARC accelerators consists of the linac, the RCS and the MR, which have different repetition rates. The beam destinations of the linac and RCS are different in each pulse. We present the precise timing system which governs the accelerator timing.
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
Kurikami, Hiroshi; Niizato, Tadafumi; Tsuruta, Tadahiko; Kato, Tomoko; Kitamura, Akihiro; Kanno, Mitsuhiro*; Kurosawa, Naohiro*
no journal, ,
To understand environmental fate of radioactive materials released from nuclear accidents is important for recovery measures. We have developed a compartment model including all important compartments and processes to describe environmental fate of radioactive materials and have applied to the Fukushima environment. The results showed a good agreement with field investigations.
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.
