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Okita, Shoichiro; Mizuta, Naoki; Takamatsu, Kuniyoshi; Goto, Minoru; Yoshida, Katsumi*; Nishimura, Yosuke*; Okamoto, Koji*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Takada, Eiji*; Fujimoto, Nozomu; Nojiri, Naoki; Umeta, Masayuki; Kokusen, Shigeru; Ashikagaya, Yoshinobu
JAERI-Data/Code 2002-009, 83 Pages, 2002/05
JAERI-Data-Code-2002-009.pdf:3.51MB
Dose equivalent rate around the fuel handling machine, the control rod handling machine, stand pipe compartment, maintenance pit were measured during gamma ray measurements from HTTR fuel, which was called as “power distribution measurements". The power distribution measurement was the first time to handle the fuel blocks irradiated in the core. Dose equivalent rate measurement aiming the check of shielding performance of components, the check of unexpected streaming path. The radiation monitoring during operation was carried out. As the results, there was no problem on shielding. The measured data at operation condition were also obtained. The data will be useful to expect operation circumstance in the future.
Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nagatomi, Hideki; Kaminaga, Masanori; Funayama, Yoshiro
JAERI-Tech 2002-034, 40 Pages, 2002/03
JAERI-Tech-2002-034.pdf:1.97MB
JRR-4, a swimming-pool type research reactor with a thermal power of 3.5MW, attained criticality in July 1998, after replacing its 90% enrichment fuel with a 20% enrichment fuel under the Reduced Enrichment Program. As a part of the program, safety analysis on thermo-hydraulics of the reactor core was conducted on cases including single channel blockage accident. With the conclusion that a certain margin on thermo-hydraulics was necessary, investigation and experiments were carried out with an aim to increase the core flow rate. To increase the core flow, it was carried out to reduce the bypass flow in the core and to increase the primary coolant flow rate from 7m
/min to 8m/min. After flow measurements using a mock-up fuel element, flow velocity of the fuel channel was determined as 1.45m/s as opposed to the designed value of 1.44m/s, and the ratio of core flow to total flow was 0.88, exceeding the value 0.86 used for the safety analysis.This report describes the JRR-4 core flow increase plan as well as the results of the channel flow rate measurement
Kaminaga, Masanori
JAERI-Tech 97-015, 74 Pages, 1997/03
no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi
JAERI-Tech 96-039, 72 Pages, 1996/09
no abstracts in English
Yamashita, Kiyonobu
Nihon Genshiryoku Gakkai-Shi, 37(3), p.213 - 216, 1995/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro*;
Nucl. Eng. Des., 150, p.69 - 80, 1994/00
Times Cited Count：5 Percentile：47.29(Nuclear Science & Technology)no abstracts in English
Takase, Kazuyuki; Hino, Ryutaro;
Nihon Genshiryoku Gakkai-Shi, 33(6), p.564 - 573, 1991/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
Shindo, Ryuichi; Yamashita, Kiyonobu; Murata, Isao
JAERI-M 90-048, 225 Pages, 1990/03
no abstracts in English
Yamashita, Kiyonobu; Fumizawa, Motoo;
JAERI-M 89-222, 74 Pages, 1990/01
no abstracts in English
;
Nuclear Technology, 78(9), p.207 - 215, 1987/09
no abstracts in English
; ; ; Minato, Kazuo; ; ; ; Iwamoto, K.; Ikawa, Katsuichi
JAERI-M 86-092, 286 Pages, 1986/07
no abstracts in English
;
Journal of Nuclear Science and Technology, 23(2), p.176 - 178, 1986/00
Times Cited Count：2 Percentile：42.25(Nuclear Science & Technology)no abstracts in English
; ; ; ; ;
Nihon Genshiryoku Gakkai-Shi, 28(6), p.527 - 533, 1986/00
Times Cited Count：2 Percentile：32.21(Nuclear Science & Technology)no abstracts in English
; ; ;
JAERI-M 85-187, 98 Pages, 1985/11
no abstracts in English
Minato, Kazuo; ; ; Ikawa, Katsuichi
JAERI-M 83-167, 24 Pages, 1983/10
no abstracts in English
Minato, Kazuo; ; ; ; Ikawa, Katsuichi; Iwamoto, K.; ; ;
JAERI-M 83-055, 77 Pages, 1983/03
no abstracts in English
Tsuji, Nobumasa*; Ohashi, Kazutaka*; Tazawa, Yujiro*; Ohashi, Hirofumi; Takamatsu, Kuniyoshi; Tachibana, Yukio
no journal, ,
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 study is performed using thermal hydraulic CFD code with one column-model of fuel blocks which is represented explicitly as individual coolant channels in the 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. The 30
sector model of limited core region is also developed for CFD calculation of natural convection flow pattern in core. The mass flow rate of upward flow is considerably reduced from one column model. And further, multi-hole type fuel blocks of MHTGR are also modeled and compare the flow and heat transfer characteristics between multi-hole type and pin-in-block type fuel block. The multi-hole type fuel block raise more natural convection flow in core region than pin-in-block type.
