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Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).
Tochio, Daisuke; Nakagawa, Shigeaki
JAERI-Tech 2005-041, 109 Pages, 2005/08
JAERI-Tech-2005-041.pdf:4.48MB
In High Temperature Engineering Test Reactor (HTTR) of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary pressurized water cooler and the intermediate heat exchanger. To remove generated heat at reactor core and to hold reactor inlet coolant temperature to specified temperature, heat exchangers in main cooling system of HTTR should have designed heat exchange performance. In this report, heat exchange performance for ACL in main cooling system is evaluated with previous operation data, and evaluated values are compared with designed value. Moreover, heat exchange performance at full power operation is estimated for the air temperature. As the result, ACL has heat exchange performance removing generated heat at reactor core under the designe ACL inlet air temperature of 33
C.
Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Eiji*; Sakaba, Nariaki; Takamatsu, Kuniyoshi
JAERI-Tech 2003-097, 55 Pages, 2004/01
JAERI-Tech-2003-097.pdf:3.34MB
In high temperature engineering test reactor (HTTR) of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler (ACL) by way of the heat exchangers of the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). Therefore, air temperature (secondary-side condition at ACL) is important factor for the heat removal capability of the reactor. Coping with the air temperature, stable reactor inlet temperature control is achieved by adjusting of ACL coolant temperature with coolant (pressurized water and air) flow rate. ACL heat removal characteristic was based on the previous operation data in rise-to-power test and in-service operation at HTTR. And evaluate heat removal capability at summertime air temperature as the most severe condition was estimated. As the result, it was confirmed that the rated power of 30 MW can be removed at the condition of summertime air-temperature.
Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Katanishi, Shoji; Kosugiyama, Shinichi; Minatsuki, Isao*; Miyoshi, Yasuyuki*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.341 - 351, 2002/12
no abstracts in English
Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Katanishi, Shoji; Kosugiyama, Shinichi; Shiozawa, Shusaku
Nihon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.189 - 192, 2002/00
no abstracts in English
Katagiri, Masaki
Denki Gakkai Genshiryoku Kenkyukai Shiryo (NE-01-23), p.9 - 14, 2001/10
no abstracts in English
Katagiri, Masaki; Kobayashi, Yoshii; Takahashi, Koji*; *; *
KEK-Proceedings 96-4, 0, p.209 - 213, 1996/07
no abstracts in English
Takada, Shoji; Shiina, Yasuaki; Inagaki, Yoshiyuki; Hishida, Makoto; Sudo, Yukio
JAERI-Research 95-056, 40 Pages, 1995/08
JAERI-Research-95-056.pdf:1.17MB
no abstracts in English
Takada, Shoji; Shibata, Mitsuhiko; ; Fujisaki, Katsuo; Ota, Yukimaru; ;
JAERI-M 94-013, 89 Pages, 1994/02
no abstracts in English
Hino, Ryutaro; *;
Nihon Genshiryoku Gakkai-Shi, 33(1), p.73 - 75, 1991/01
Times Cited Count：0 Percentile：0.3(Nuclear Science & Technology)no abstracts in English
*; Himeno, Yoshiaki; *
PNC TN9410 87-168, 65 Pages, 1987/12
In the event of a design basis sodium leak accident in the reactor building of the FBRs, the leaked sodium is collected and stored in a sodium storage room until the accident comes to an end. By placing the emphasis on the heat release characteristics of the sodium storage room, the following tests were performed. (1)Determination of thermal properties of materials for the floor of the room; perlite concrete, structural concrete, etc. (2)Partial engineering tests of heat conduction from a hot sodium to structural concrete simulating the accident conditions of the floor. In the former tests, the reliable data of the thermal properties were obtained. In the latter tests, transient temperature distributions were obtained.Comparison of these distributions with the calculated results of the code indicated that the code gives conservative results.
*; *; Kuroha, Mitsuo; *; *; *; *
PNC TN941 85-53, 144 Pages, 1985/03
SWAC13E is a one-dimensional thermo-hydraulic computer program to analyze large scale sodium-water reaction accidents in an LMFBR steam generator. The code is the advanced version of SWAC13, the long-term hydraulic analysis module of SWACS; the energy conservation is taken into consideration in the new version to add the function to analyze the temperature behavior of the reaction. The present document covers the validation study of the code by using the large leak data of the Steam Generator Safety Test Facility (SWAT-3). The analytical parameters are as follows: (1)Model of relative velocity. (2)Void/droplet density. (3)The number of nodes where water leaks. (4)Reaction heat. It is concluded that the code can analyze the phenomena with a reasonable conservatism by choosing the proper value of the parameters.
loop; ; ; ; ;
Nihon Genshiryoku Gakkai-Shi, 26(4), p.318 - 326, 1984/00
Times Cited Count：2 Percentile：29.96(Nuclear Science & Technology)no abstracts in English
+A loop; ; ; ; ;
Nihon Genshiryoku Gakkai-Shi, 26(5), p.410 - 420, 1984/00
Times Cited Count：3 Percentile：38.16(Nuclear Science & Technology)no abstracts in English
; ; ; ; ; ; ; *
Nihon Genshiryoku Gakkai-Shi, 26(3), p.239 - 248, 1984/00
Times Cited Count：4 Percentile：45.52(Nuclear Science & Technology)no abstracts in English
Kobayashi, Jun; Ezure, Toshiki; Onojima, Takamitsu; Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
Preliminary confirmation tests were conducted by using PLANDTL-2 sodium experimental facility for evaluation of decay heat removal in a sodium-cooled fast reactor. Fundamental performance was confirmed in steady state and transient state of PLANDTL-2.
