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Shibata, Hiroki; Saito, Hiroaki; Hayashi, Hirokazu; Takano, Masahide
JAEA-Data/Code 2019-023, 138 Pages, 2020/03
Transmutation of minor actinides in the form of nitride fuel by the accelerator driven system has been developed to reduce the radiotoxicity and volume in the radioactive wastes. Nitride fuel behavior under irradiation condition is necessary for its design and development. Nitride fuel performance analysis module based on light water reactor fuel performance code, FEMAXI-7, was developed by introducing fundamental properties of nitride pellet, 9Cr-1Mo ferrite cladding, and Pi-Bi coolant. As a result of test analysis with this module, we have understood that the nitride fuel shows excellent behavior under irradiation due to its high thermal conductivity. We found that, however, it may be a main concern that fuel cladding integrity is maintained during irradiation in which pellet-cladding mechanical interaction is increased by He gas release, low creep rate of nitride pellet at low temperatures, and high creep rate of cladding above 873 K.
Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09
Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 99 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.
Furumoto, Kenichiro*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Teshima, Hideyuki*; Yamashita, Shinichiro; Saito, Hiroaki; Shirasu, Noriko
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Since 2015, Mitsubishi Nuclear Fuel (MNF) has joined in a Japanese R&D project of ATF founded by the Ministry of Economy, Trade and Industry (METI) as a subcontractor to Japan Atomic Energy Agency (JAEA) which is the prime contractor to METI. In this program, MNF plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). This paper reports the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding. As out-of-pile tests, thermal diffusivity measurement and autoclave corrosion test for SiC composite samples were performed. Test apparatus were developed for evaluation of performance of SiC composite under the condition simulated design basis accident (DBA).
Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03
Wakai, Eiichi; Kikuchi, Takayuki; Kogawara, Takafumi; Kimura, Haruyuki; Yokomine, Takehiko*; Kimura, Akihiko*; Nogami, Shuhei*; Kurishita, Hiroaki*; Saito, Masahiro*; Nishimura, Arata*; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 6 Pages, 2011/03
Japanese activities of test facilities in IFMIF-EVEDA (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities) project have three subjects of engineering design of post irradiation examination (PIE) facilities, small specimen test technique (SSTT), and engineering design of high flux test module (HFTM), and this paper is summarized about present status. Functional analysis and design of 2-D and 3-D models of PIE facility were performed. In HFTM, as materials of heater, W-3Re alloy and/or SiC/SiC composite were selected in the points of high temperature materials, fabrication technology and some suitable properties such as resistance of thermal shock, high temperature re-crystallization, ductility, resistance of irradiation degradation, and low-activation. In SSTT, a test machine of fracture toughness was designed and developed for small specimens with 10 mm square, and it had high accuracy controllability for stress and displacement.
Wakai, Eiichi; Kogawara, Takafumi; Kikuchi, Takayuki; Yamamoto, Michiyoshi; Molla, J.*; Kimura, Akihiko*; Kasada, Ryuta*; Kim, B.*; Nogami, Shuhei*; Hasegawa, Akira*; et al.
no journal, ,
no abstracts in English
Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Kikuchi, Takayuki; Ito, Yuzuru*; Hoashi, Eiji*; Yoshihashi, Sachiko*; Horiike, Hiroshi*; et al.
no journal, ,
no abstracts in English
Hayashi, Hirokazu; Saito, Hiroaki; Kurata, Masaki
no journal, ,
no abstracts in English
Takeuchi, Yutaka*; Kakiuchi, Kazuo*; Sato, Hisaki*; Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Furumoto, Kenichiro*; Teshima, Hideyuki*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro
no journal, ,
Mitsubishi Nuclear Fuel (MNF) plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). In this presentation, the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR will be reported. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding.
Horie, Hideki*; Takeuchi, Yutaka*; Kakiuchi, Kazuo*; Sato, Hisaki*; Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Fukahori, Tokio
no journal, ,
The problem in the accident tolerance fuel development was considered in comparison the fuel behavior of SiC/SiC composite cladding tube with current Zry cladding tube. In this study, Reactivity Initiated Accident (RIA) was evaluated using the plant excess security analysis code TRACTTM and the fuel performance code FEMAXI.
Yamashita, Shinichiro; Shirasu, Noriko; Saito, Hiroaki
no journal, ,
The FEMAXI-ATF code has been developed to analyze the SiC-cladding-fuel behaviors as an extended version of the LWR fuel-analysis code FEMAXI-7. The analysis using the FEMAXI-ATF code indicated that the SiC cladding shows large swelling by irradiation and it causes the decrease of cladding thermal conductivity and the gap conductance, and consequently increase of the fuel temperature. Furthermore, the mechanical integrity for SiC cladding and influence of pellet-cladding mechanical interaction (PCMI) were evaluated.