THE FlRST ATOMlC POWER INDUSTRY GROUP
FAPIGÌ@\ ( 33 )
Cover DesignFNozomu Watanabe
Kazuhiro Koga, Norikazu Oohara, Hiroichi Ino, Katsumi Kondo, Hideaki
Ito, Takashi Ashida, Toshiyuki Nakamura
In the experimental fast reactor "Joyo", the incident that the MARICO-2
test subassembly was bent in the reactor vessel occurred in 2007.
KEYWORDSFfast reactor, Joyo, MARICO-2, test subassembly, retrieval works
Shingo Hirata,@Tomoaki Mohri, Minoru Igarashi, Manabu Satoh, Yu Kamiji,
Yusuo Nishihata, Ryutaro Hino
A new type passive autocatalytic recombiner (PAR) to mitigate released hydrogen gas at the accident of nuclear power plants has been developed. In order to realize easy and anywhere installation, this new PAR has advanced automotive catalysts with features such as light and small, high durability, and high performance. Downsizing and light-weighting of recombiner has been conducted with thermal hydraulic and structural analysis, and the PAR concept proposed is revealed feasible for hydrogen mitigation in nuclear power plants.
KEYWORDSFpassive autocatalytic recombiner, nuclear power plant, hydrogen mitigation
Naoki Uchiyama, Norikazu Kuchiki, Kouji Satou
JAEA has established SERF (Sodium Engineering Research Facility) facility for research and development of advanced sodium handling technology in Tsuruga City, Fukui Prefecture. We introduce design, fabrication and installation of SERF carried out by KHI through fiscal year 2011 to 2014.
KEYWORDSFsodium handling technology, FBR, MONJU
Nobumasa Tsuji, Kazutaka Ohashi, Yujirou Tazawa, Yukio Tachibana, Hirofumi
Ohashi, Kuniyoshi Takamatsu,
Passive heat removal performance is of primary concern for enhanced inherent safety of HTGR. In a loss of forced cooling accident, decay heat of fuels must be removed to graphite blocks by radiation, thermal conduction and natural convection in block-type HTGR. Because the temperature of fuels is strongly affected by natural convection of coolant in core region, it becomes important to estimate the behavior of natural convection in core region precisely. The numerical studies are performed using thermal hydraulic CFD code for one column-model of fuel blocks which is represented explicitly as individual coolant channels in fuel block. The thermal hydraulic analyses are conducted for normal operation and loss of forced cooling accident conditions, as results, the flow and heat transfer characteristics of fuel blocks are quantitatively evaluated both in forced convection mode and natural convection mode.
KEYWORDSFHTGR, HTTR, fuel block, natural convection, CFD code
This paper propose an evaluation method for social infrastructures
and the applicability of the proposed evaluation method is verified
by the trial application of the method to a nuclear power plant which
is one of social infrastructures.
KEYWORDSFsocial infrastructure, environment, maintenance, hazard, maturity model, evaluation figures, ISO