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Morimoto, Kyoichi; Ono, Takahiro; Kakutani, Satomi; Yoshida, Moeka; Suzuki, Soichiro
Journal of Robotics and Mechatronics, 36(1), p.125 - 133, 2024/02
The Naraha Center for Remote Control Technology Development was established for the purpose of developing and verifying remote control equipment for promoting the decommissioning of the Fukushima Daiichi Nuclear Power Station and the external use of this center was started in 2016. The mission of this center is to contribute to the decommissioning of the Fukushima Daiichi Nuclear Power Station and for the reconstruction of Fukushima Prefecture. In this review, we describe the equipment related to the full-scale mock-up test, the component test for a remote-control device and the virtual reality system in this center. In addition, the case examples for usage of these equipment are introduced.
Naraha Center for Remote Control Technology Development, Fukushima Research Insitute
JAEA-Review 2018-014, 52 Pages, 2018/12
JAEA-Review-2018-014.pdf:5.62MB
The Naraha Remote Technology Development Center (Naraha Center) consists of a mock-up test building and a research management building, and various test facilities necessary for the decommissioning work after the accident of TEPCO Fukushima Daiichi Nuclear Power Station are installed. Using these test facilities, a wide range of users, such as companies engaged in decommissioning work, research and development institutions, educational institutions, etc., can efficiently develop robots through characterization and performance evaluation of remote-controlled robots. Furthermore, it is possible to make various uses such as exhibitions that many companies have met together, experts' meetings on decommissioning. This report summarizes the activities of the Naraha Center such as development of remote control technologies, maintenance and training of remote control equipment for emergency response, use of component test areas, and so on in FY2016.
Seki, Masakazu; Maekawa, Tomoyuki; Izawa, Kazuhiko; Sono, Hiroki
JAEA-Technology 2017-038, 52 Pages, 2018/03
JAEA-Technology-2017-038.pdf:4.6MB
The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY core is controlled by the water level fed in the core tank as well as the present STACY. In order to verify the basic design of the water feed and drain system of the modified STACY, we constructed a mockup test apparatus with almost the same structure and specifications as the modified STACY. In the mockup test, performance checks were pursued regarding limitation of maximum flow of water feeding, adjustment of the flow rate of water feeding, stop of water feeding and others. This report describes the outline and results of the mock-up test of the water feed and drain system of the modified STACY.
Daido, Hiroyuki; Kawatsuma, Shinji; Kojima, Hisayuki; Ishihara, Masahiro; Nakayama, Shinichi
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 8 Pages, 2017/00
Tochio, Daisuke; Nakagawa, Shigeaki; Furusawa, Takayuki*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.147 - 155, 2005/06
High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at JAERI achieved the reactor outlet coolant temperature of 950
C for the first time in the world at Apr. 19, 2004. To remove of generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value.
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki*; Emori, Koichi; Tachibana, Yukio
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.388 - 395, 2004/12
Chemistry control is important for the helium coolant of High Temperature Gas-cooled Reactors (HTGRs) because impurities cause oxidation of the graphite used in the core and corrosion of high temperature materials used in the heat exchanger. In the High Temperature Engineering Test Reactor (HTTR) which is the first HTGR in Japan, the chemical impurity concentration is restricted and its behaviour is monitored during reactor operations. The impurity is reduced by the helium purification system and the concentration is measured by the helium sampling system installed to the primary and secondary helium system, continuously. This paper describes the impurity behaviour during the rise-to-power test which is the initial power-up of the HTTR. Also, the amount of the emitted impurity to the primary circuit from the graphite component and insulator used at the concentric hot gas duct are evaluated. During the power up, any abnormal impurity increases were not obtained and the chemical composition of the primary circuit is sufficiently in the stability area to avoid carbon deposition.
JAERI Working Group for Examination of the Ruptured Pipe at Hamaoka-1
JAERI-Tech 2002-045, 253 Pages, 2002/03
JAERI-Tech-2002-045.pdf:60.08MB
no abstracts in English
Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo
JAERI-Tech 2001-090, 69 Pages, 2002/01
JAERI-Tech-2001-090.pdf:7.88MB
HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.
JAERI Working Group for Examination of the Ruptured Pipe at Hamaoka-1
JAERI-Tech 2001-094, 60 Pages, 2001/12
JAERI-Tech-2001-094.pdf:13.99MB
no abstracts in English
Kato, Takashi; Nakajima, Hideo; Isono, Takaaki; Hamada, Kazuya; Kawano, Katsumi; Sugimoto, Makoto; Nunoya, Yoshihiko; Koizumi, Norikiyo; Matsui, Kunihiro; Oshikiri, Masayuki*; et al.
Teion Kogaku, 36(6), p.315 - 323, 2001/06
no abstracts in English
Saito, Kenji; Homma, Fumitaka; Omata, Toru; Aono, Tetsuya; Kawaji, Satoshi; Kawasaki, Kozo; Iyoku, Tatsuo
Proceedings of International Topical Meeting on Nuclear Plant Instrumentation, Controls, and Human-Machine Interface Technologies (NPIC&HMIT 2000) (CD-ROM), 8 Pages, 2000/00
no abstracts in English
Sakaba, Nariaki; Emori, Koichi; Saruta, Toru
JAERI-Tech 99-072, p.125 - 0, 1999/10
no abstracts in English
; Miyoshi, Yoshinori; ; ; ;
Proc. of the Int. Conf. on Nuclear Power Plant Operations; Ready for 2000, p.435 - 441, 1992/00
no abstracts in English
;
JAERI 1272, 32 Pages, 1981/06
no abstracts in English
