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Takano, Kazuya; Kato, Atsushi; Uchita, Masato*; Okazaki, Hitoshi*; Ikari, Risako*
no journal, ,
Primary Heat Transport System (PHTS) flow rate is the most promising signal to detect the pump/diagrid link rupture in pool type SFR. The electromagnetic flowmeter is considered to measure the PHTS flow rate and is designed to install nearby outlet of the PHTS circulation pump in the current design of pool-type SFR with 600 MWe. The flowmeter could be significantly affected by the noise from pump motor. In order to improve the signal strength, the structure of flowmeter is drastically modified with the D-shape yokes and the magnetic circuits above and below an electrode.
Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi
no journal, ,
no abstracts in English
Iwamoto, Hiroki; Iwamoto, Osamu; Kunieda, Satoshi
no journal, ,
no abstracts in English
Shikaze, Yoshiaki
no journal, ,
no abstracts in English
Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa; Demachi, Kazuyuki*
no journal, ,
In order to evaluate the effectiveness of the measures for improving resilience against excessive earthquake, this study assumed to improve seismic safety margin of components of heat removal system as the measures against loss of heat removal systems event after reactor shutdown. The core damage frequency was calculated and the reduction effect of it was estimated by comparing it before and after the introduction of the measures for improving resilience.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa; Demachi, Kazuyuki*
no journal, ,
In order to evaluate the effectiveness of the measures for improving resilience at ultra-high temperatures, a concept of evaluation focusing on core damage frequency was proposed. Assuming loss of heat removal systems event after reactor shutdown which may result in core damage in sodium-cooled fast reactors, the measures for improving resilience which enable to recover the safety functions at ultra-high temperatures are identified: one is to retain the primary coolant using failure mitigation technology, and the other is to add a heat removal system that can be used under ultra-high temperature conditions. The core damage frequencies were calculated preliminarily and their reduction effect was estimated by comparing them before and after the introduction of the measures for improving resilience.
Okamura, Shigeki*; Kinoshita, Takahiro*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*
no journal, ,
no abstracts in English
Kinoshita, Takahiro*; Okamura, Shigeki*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*
no journal, ,
no abstracts in English
Demachi, Kazuyuki*; Kuwabara, Yuto*; Chen, S.*; Kasahara, Naoto*; Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi
no journal, ,
no abstracts in English
C-SS eutectic molten materialsFukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
no journal, ,
The purpose of this study is to develop the thermophysical property model needed to simulate the behavior of the eutectic compound of the control rod material (boron carbide (BC)) and the structural material of a reactor vessel (stainless steel: SUS316L (SS)) during a core disruptive accident in a sodium-cooled fast reactor. This paper describes the liquidus temperature, density and surface tension of the molten compound of 15mass% BC-SS systematically measured with the High-Temperature Thermophysical Property Measurement System (PROSPECT).
Suzuki, Risa; Nomi, Takayoshi; Nagatani, Taketeru; Shiromo, Hideo; Shiba, Tomooki; Kaburagi, Masaaki; Okumura, Keisuke; Kosuge, Yoshihiro*; Takada, Akira*; Nauchi, Yasushi*
no journal, ,
no abstracts in English
Suzuki, Toru*; Honda, Haruki*; Takahashi, Yu*; Kawada, Kenichi
no journal, ,
no abstracts in English
Takatsuka, Daichi*; Nakamura, Takeshi*; Zhang, T.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
no journal, ,
no abstracts in English
Sato, Daisuke; Watanabe, So; Yano, Kimihiko; Kitawaki, Shinichi; Arai, Tsuyoshi*; Shibata, Atsuhiro; Takeuchi, Masayuki
no journal, ,
no abstracts in English
Yamagishi, Isao; Kato, Tomoaki; Horita, Takuma
no journal, ,
no abstracts in English
Hata, Kuniki; Hanawa, Satoshi; Chimi, Yasuhiro; Uchida, Shunsuke
no journal, ,
no abstracts in English
Akie, Hiroshi; Tada, Kenichi; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro
no journal, ,
Japan Atomic Energy Agency is developing an advanced neutronics/thermal-hydraulics coupling simulation system for the design advancement and the safety improvement of light water reactors. In this presentation, an impact of the fine-scale void distribution in sub-channels on neutronics calculations is studied.
Kakuda, Ayaka; Taniguchi, Takumi; Namiki, Masahiro*; Kikuchi, Michio*; Yamamoto, Takeshi*; Kaneda, Yoshihisa*; Osawa, Norihisa*; Osugi, Takeshi; Sone, Tomoyuki; Kuroki, Ryoichiro
no journal, ,
no abstracts in English
Kikuchi, Michio*; Yamamoto, Takeshi*; Otsuka, Taku*; Kawato, Takaya*; Kaneda, Yoshihisa*; Sakamoto, Ryo*; Haga, Kazuko*; Kakuda, Ayaka; Osugi, Takeshi; Sone, Tomoyuki; et al.
no journal, ,
no abstracts in English
Yoshimura, Kazuya; Fujiwara, Kenso; Nakama, Shigeo; Abe, Tomohisa
no journal, ,
no abstracts in English