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Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju
JAEA-Data/Code 2017-002, 74 Pages, 2017/03
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Inaba, Yoshitomo; Nishihara, Tetsuo
Annals of Nuclear Energy, 101, p.383 - 389, 2017/03
Times Cited Count:8 Percentile:58.10(Nuclear Science & Technology)In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Shimizu, Akira; Ohashi, Hirofumi; Kato, Michio; Hayashi, Koji; Aita, Hideki; Nishihara, Tetsuo; Inaba, Yoshitomo; Takada, Shoji; Morisaki, Norihiro; Sakaki, Akihiro*; et al.
JAERI-Tech 2005-031, 174 Pages, 2005/06
Connection of hydrogen production system by steam reforming of methane to the High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Research Institute (JAERI) has been surveyed until now. Mock-up test facility of this steam reforming system with full-scale reaction tube was constructed in FY 2001, and a lot of operational test data on heat exchanges were obtained in these tests.In this report specifications, structures and heat transfer formulae of steam reformer, steam superheater, steam generator, condenser, helium gas cooler, feed gas heater and feed gas superheater were described. Evaluation codes were newly made to evaluate heat transfer characteristics from measured test data. Overall heat-transfer coefficient obtained from the experimental data were compared and evaluated with the prospective value calculated with heat transfer formulae. As a result, heat transfer performance and thermal efficiency of these heat exchangers were confirmed to be appropriate.
Tada, Eisuke; Hada, Kazuhiko; Maruo, Takeshi; Safety Design/Evaluation Group
Purazuma, Kaku Yugo Gakkai-Shi, 78(11), p.1145 - 1156, 2002/11
no abstracts in English
Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi; Hada, Kazuhiko; Tada, Eisuke
Proceedings of 10th International Conference on Nuclear Engineering (ICONE 10) (CD-ROM), 8 Pages, 2002/00
The Japan Atomic Energy Research Institute (JAERI) has conducted to develop a new design code for construction and operation/maintenance of the International Thermonuclear Experimental Reactor (ITER), which will be formed as a code case of ASME B&PV Code Section III, division 4, in collaboration with the ASME international. The new design code will also include the several new techniques and materials developed for each components of ITER. This paper describes the new cryogenic steels used in the magnet system and the design approach with taking account of unique features of the ITER superconducting magnets.
Okumura, Keisuke
JAERI-Data/Code 98-025, 243 Pages, 1998/10
no abstracts in English
Yamashita, Kiyonobu; Nojiri, Naoki; Fujimoto, Nozomu; Nakano, Masaaki*; Ando, Hiroei; Nagao, Yoshiharu; Nagaya, Yasunobu; Akino, Fujiyoshi; Takeuchi, Mitsuo; Fujisaki, Shingo; et al.
Proc. of IAEA TCM on High Temperature Gas Cooled Reactor Applications and Future Prospects, p.185 - 197, 1998/00
no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki;
Journal of Nuclear Science and Technology, 34(8), p.760 - 770, 1997/08
Times Cited Count:2 Percentile:22.91(Nuclear Science & Technology)no abstracts in English
Okumura, Keisuke; ;
JAERI-Data/Code 96-015, 445 Pages, 1996/03
no abstracts in English
Araya, Fumimasa; Murao, Yoshio; Iwamura, Takamichi
Journal of Nuclear Science and Technology, 32(10), p.1039 - 1046, 1995/10
Times Cited Count:5 Percentile:49.21(Nuclear Science & Technology)no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki
Transactions of the American Nuclear Society, 73, p.207 - 208, 1995/00
no abstracts in English
Hishinuma, Akimichi
Purazuma, Kaku Yugo Gakkai-Shi, 70(7), p.697 - 703, 1994/07
no abstracts in English
Onuki, Akira; ; Murao, Yoshio
JAERI-M 94-026, 60 Pages, 1994/03
no abstracts in English
Asano, Yoshihiro; Sasamoto, Nobuo
Radiation Physics and Chemistry, 44(1-2), p.133 - 137, 1994/00
no abstracts in English
Asano, Yoshihiro; Sasamoto, Nobuo
Radiation Physics and Chemistry, 44(1-2), p.133 - 137, 1994/00
Times Cited Count:17 Percentile:79.67(Chemistry, Physical)no abstracts in English
; Minato, Kazuo; Eto, Motokuni; Oku, Tatsuo*;
JAERI-M 92-085, 28 Pages, 1992/06
no abstracts in English
; Okumura, Keisuke
JAERI-M 92-068, 107 Pages, 1992/05
no abstracts in English
Yamashita, Kiyonobu; Murata, Isao; Shindo, Ryuichi
Nuclear Science and Engineering, 110, p.177 - 185, 1992/02
Times Cited Count:3 Percentile:51.97(Nuclear Science & Technology)no abstracts in English
Ishihara, Masahiro; Iyoku, Tatsuo; ; ; Shiozawa, Shusaku
JAERI-M 91-154, 39 Pages, 1991/10
no abstracts in English
Ishihara, Masahiro; Iyoku, Tatsuo; ; ; Shiozawa, Shusaku
JAERI-M 91-153, 51 Pages, 1991/10
no abstracts in English