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Watanabe, Yuichi*; Muramatsu, Ken; Oikawa, Tetsukuni
Nuclear Engineering and Design, 235(23), p.2495 - 2512, 2005/12
Times Cited Count：2 Percentile：80.93(Nuclear Science & Technology)This paper presents an evaluation of seismic capacity of a group of vertical U-tube type heat exchangers(HXs) with support frames for residual heat removal systems of BWRs for seismic Probabilistic Safety Assessment in Japan. The median capacity was evaluated by a time history response analysis with a detailed model for a representative HX selected from four HXs. The logarithmic standard deviation(LSD) for uncertainty due to lack of knowledge was evaluated with consideration of the variabilities in three influential parameters, i.e., diameter of anchor bolt, weight of HX and position of center of gravity of HX. The dominant failure mode of HXs was the failure of anchor bolts of lugs mainly due to shearing stress. The capacity expressed in terms of zero period acceleration on the foundation of HX was evaluated to be 4,180 Gal(4.3 g) for median, LSD for uncertainty due to randomness was 0.11 from the variability in material property and LSD due to lack of knowledge was 0.21 to 0.53 depending on combination of the variability in design parameters to be considered.
Horiguchi, Hironori; Oyama, Koji; Ishikuro, Yasuhiro; Hirane, Nobuhiko; Ito, Kazuhiro; Kameyama, Iwao
JAERI-Tech 2005-001, 38 Pages, 2005/02
JAERI-Tech-2005-001.pdf:2.79MB
After JRR-4 heat exchanger was renewed in made of stainless steel from carbon steel, it was examined how to manage the heat exchanger. The main subject is the cleaning technology of the heat exchanger. The recovery of old heat exchanger cooling performance has been by only chemical cleaning. Now we use chemical and dry cleaning as a new technique. It helps prevent of corrosions of secondary pipes and reduce of management costs. This report describes the performance management and cleaning technology of the JRR-4 heat exchanger and the management of the JRR-4 coolant.
Ishiyama, Shintaro
Nippon Kinzoku Gakkai-Shi, 68(7), p.490 - 497, 2004/07
Times Cited Count：0 Percentile：100(Metallurgy & Metallurgical Engineering)SUS304 multi stage blazed fin core body for GTHTR300 was tested its high temperature strength, fatigue and creep and it was found that the body behaves as same strength of SUS304 material.
Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Minatsuki, Isao*
Dai-31-Kai Gasu Tabin Teiki Koenkai Rombunshu, p.55 - 60, 2003/06
no abstracts in English
Matsui, Shingo*; Muto, Yasushi*; Shiina, Yasuaki
Nippon Genshiryoku Gakkai Wabun Rombunshi, 2(2), p.175 - 186, 2003/06
no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi
Nippon Kikai Gakkai Rombunshu, A, 69(679), p.682 - 688, 2003/03
no abstracts in English
Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Katanishi, Shoji; Kosugiyama, Shinichi; Minatsuki, Isao*; Miyoshi, Yasuyuki*
Nippon Genshiryoku Gakkai Wabun Rombunshi, 1(4), p.341 - 351, 2002/12
no abstracts in English
Takeda, Tetsuaki
Nippon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.199 - 202, 2002/00
There are several methods for enhancement of heat transfer such as attaching various type fins on a channel surface of a heat exchanger tube and a steam reforming tube. The objective of this study is to clarify characteristics of heat transfer and pressure drop in the channel inserted metallic wire with high porosity. A heat transfer experiment was performed using a horizontal circular tube to obtain the heat transfer characteristics of the channel inserted copper wire. From the results obtained in this experiment, it seems to be probable that an enhancement of heat transfer method using metallic wire inserts is further improved under the high temperature conditions in efficiency as compared with the other methods.
Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, X.; Katanishi, Shoji; Kosugiyama, Shinichi; Shiozawa, Shusaku
Nippon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.189 - 192, 2002/00
no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi; Ogata, Hiroshi*; Yamada, Seiya*
Nippon Genshiryoku Gakkai-Shi, 43(11), p.1136 - 1148, 2001/11
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi; Tanihira, Masanori*; Ogata, Kan*; Kamito, Yoshimi*; Igari, Toshihide*
Nippon Genshiryoku Gakkai-Shi, 43(6), p.603 - 611, 2001/06
Times Cited Count：5 Percentile：64.97(Nuclear Science & Technology)no abstracts in English
Ishiyama, Shintaro; Muto, Yasushi; Nishii, Takashi*; Kasaba, Michihiro*
Nippon Kinzoku Gakkai-Shi, 65(6), p.524 - 533, 2001/06
no abstracts in English
Matsui, Shingo*; Muto, Yasushi; Shiina, Yasuaki
JAERI-Tech 2000-083, 107 Pages, 2001/01
JAERI-Tech-2000-083.pdf:3.46MB
no abstracts in English
Muto, Yasushi; Ishiyama, Shintaro
Nippon Genshiryoku Gakkai-Shi, 42(10), p.1020 - 1027, 2000/10
Times Cited Count：1 Percentile：87.37(Nuclear Science & Technology)no abstracts in English
Okamoto, Koji*; *
JNC-TY9400 2000-016, 90 Pages, 2000/06
JNC-TY9400-2000-016.pdf:2.53MB
no abstracts in English
JNC-TN9410 2000-010, 72 Pages, 2000/03
JNC-TN9410-2000-010.pdf:2.14MB
The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 16 years from 1982 to 1997. During the MK-II core operation, extensive data were accumulated from the plant characteristic tests. Tests conducted at JOYO included operating characteristic tests for confirming operational safety, performance tests for confirming design performance of the MK-II core, and special tests for research and development ofthe plant. In this report, the outline and the results of each test item are shown. These test data can be provided by the magnet-optical disk.
; ; Saikawa, Takuya*; Sukegawa, Kazuya*
JNC-TN9410 2000-008, 66 Pages, 2000/03
JNC-TN9410-2000-008.pdf:1.39MB
The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.
Tsukimori, Kazuyuki; Furuhashi, Ichiro*
JNC-TN9400 2000-049, 93 Pages, 2000/03
JNC-TN9400-2000-049.pdf:2.82MB
lt is one of the important key points to reduce thermal stress of the primary piping system in the design of sodium coolant loop-type FBR plants. The objectives of this study are to understand the characteristics of the thermal stresses in the simple S-shaped hot leg piping systems which run from the outlet nozzle of the reactor vessel (R/V) to the inlet nozzle of the intermediate heat exchanger (IHX), and to propose some recommendable routings of piping systems. Results are summarized as follows. (1)Generally, the thermal stresses in elbows are severer than those at nozzles. The tendency was observed that the stress in elbow decreases with the increase of the distance between the outlet nozzle of R/V and the inlet nozzle of IHX and also the distance between the outlet nozzle of R/V and the liquid surface level. (2)lt is expected to reduce thermal stresses in elbow to big extent by adopting super 90 degree elbows. Therefore, in these cases the dimension region which satisfies the allowable stress is broad compared with that in the case of the conventional 90 degree elbow. (3)The stress estimations in elbow based on 'MITl notice No.501' become excessively large compared with the results by FEA using shell elements, when the maximum stress occurs at the end of elbow. ln these cases, the estimation can be rationalized by replacing the maximum stress by the mean of stresses at the end and at the middle of the elbow. (4)Two routings with 105 degree elbows are recommended. 0ne has the advantage from the view point of reduction of length of pipe and the other does from the view point of reduction of thermal stresses, compared with the routing with 90 degree elbows.
Ishiyama, Shintaro; Muto, Yasushi; Kasaba, Michihiro*; Kaneta, Takayoshi*
Nippon Genshiryoku Gakkai-Shi, 42(3), p.196 - 203, 2000/03
Times Cited Count：1 Percentile：87.37(Nuclear Science & Technology)no abstracts in English
; Miyamoto, Yoshiaki; Tanaka, Toshiyuki
Karyoku Genshiryoku Hatsuden, 50(9), p.1123 - 1130, 1999/09
no abstracts in English
