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Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi*; Eto, Masao*; Miyagawa, Takayuki*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In the frame work of generation IV international forum, safety design criteria and safety design guideline for the generation IV sodium-cooled fast reactors have been developing. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC. In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX was modified for the primary heat exchanger, which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator, protective wall tube type design is under investigation as an option with less R&D risks.
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio
Transactions of the American Nuclear Society, 91, P. 377, 2004/00
Carbon deposition occurred occasionally in the graphite-moderated gas-cooled reactors was evaluated for the reactor pressure vessel, intermediate heat exchanger, etc. using the measured chemical impurity data for the initial condition of the safety demonstration test. By the evaluated result, it is confirmed that the high-temperature components keep their structural integrity during the any temperature transients in safety demonstration tests.
JNC TN4400 2000-002, 33 Pages, 2000/06
JNC-TN4400-2000-002.pdf:5.22MB
An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.
Futakawa, Masatoshi; ; Takada, Shoji;
Nuclear Technology, 118(1), p.83 - 88, 1997/04
Times Cited Count：1 Percentile：14.29(Nuclear Science & Technology)no abstracts in English
Takeda, Takeshi; Kunitomi, Kazuhiko; ;
Nucl. Eng. Des., 168, p.11 - 21, 1997/00
Times Cited Count：49 Percentile：93.69(Nuclear Science & Technology)no abstracts in English
Kaji, Yoshiyuki; Ioka, Ikuo; Fukaya, Kiyoshi
Nihon Genshiryoku Gakkai-Shi, 38(10), p.47 - 56, 1996/10
no abstracts in English
Kunitomi, Kazuhiko; Takeda, Takeshi; Shinozaki, Masayuki; Okubo, Minoru; ; Koikegami, Hajime*
Nihon Genshiryoku Gakkai-Shi, 37(4), p.316 - 326, 1995/00
Times Cited Count：1 Percentile：17.36(Nuclear Science & Technology)no abstracts in English
Kaji, Yoshiyuki; Ioka, Ikuo;
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.363 - 368, 1995/00
no abstracts in English
; Hada, Kazuhiko
Doryoku, Enerugi Gijutsu No Saizensen : Shimpojiumu Koen Rombunshu 1994, 0, p.311 - 316, 1994/00
no abstracts in English
Kunitomi, Kazuhiko; Shinozaki, Masayuki; Okubo, Minoru; Koikegami, Hajime*;
Proc. of ARS 94 Int. Topical Meeting on Advanced Reactors Safety,Vol. 1, 0, p.188 - 192, 1994/00
no abstracts in English
Kunitomi, Kazuhiko; Shinozaki, Masayuki; ; Okubo, Minoru; Baba, Osamu; ; Otani, Akihito*
JAERI-M 92-147, 77 Pages, 1992/10
no abstracts in English
Hada, Kazuhiko; Okubo, Minoru; Baba, Osamu
Nucl. Eng. Des., 132, p.13 - 21, 1991/00
Times Cited Count：1 Percentile：19.78(Nuclear Science & Technology)no abstracts in English
Nishiguchi, Isoharu; Kaji, Yoshiyuki; Ioka, Ikuo; ;
J. Pressure Vessel Technol., 112, p.233 - 239, 1990/08
Times Cited Count：8 Percentile：53.81(Engineering, Mechanical)no abstracts in English
Saito, Shinzo; Sudo, Yukio; Fukuda, Kosaku; Nakajima, Hajime; Oku, Tatsuo; ; Tanaka, Toshiyuki
Genshiryoku Kogyo, 36(4), p.20 - 62, 1990/04
no abstracts in English
Okubo, Minoru; Hada, Kazuhiko; Baba, Osamu
High Temperature Metallic Materials for Gas-Cooled Reactors, p.42 - 49, 1989/00
no abstracts in English
Osakabe, Masahiro
Journal of Nuclear Science and Technology, 24(6), p.498 - 500, 1987/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
; ; ; ; ; ; ;
JAERI-M 85-182, 304 Pages, 1985/11
no abstracts in English
JAERI-M 84-080, 40 Pages, 1984/04
no abstracts in English
;
JAERI-M 83-212, 54 Pages, 1983/12
no abstracts in English
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JAERI-M 83-150, 27 Pages, 1983/09
no abstracts in English
