Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Minari, Eriko*; Kabasawa, Satsuki; Mihara, Morihiro; Makino, Hitoshi; Asano, Hidekazu*; Nakase, Masahiko*; Takeshita, Kenji*
Journal of Nuclear Science and Technology, 60(7), p.793 - 803, 2023/07
Times Cited Count：3 Percentile：47.03(Nuclear Science & Technology)Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(4), p.221 - 225, 2018/04
no abstracts in English
Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*
Nihon Genshiryoku Gakkai-Shi ATOMO, 58(3), p.161 - 166, 2016/03
no abstracts in English
Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.668 - 677, 2005/10
A systematic research program on high burnup fuel behavior under LOCA conditions is being conducted at JAERI. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence were conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44 GWd/t at a PWR, to investigate behavior and condition of cladding fracture during quenching for safety evaluation. Differences were not clearly observed between irradiated and unirradiated claddings at similar hydrogen concentrations in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases. Ductility of pre-hydrided, oxidized and quenched claddings was also evaluated by using ring-tensile and ring-compression tests. Embrittlement criteria (zero-ductility limits) from both the tests were lower than the fracture conditions in the integral thermal shock tests. This indicates that loading conditions should be well simulated to evaluate cladding performance under LOCA conditions.
Fuketa, Toyoshi; Nagase, Fumihisa; Sasahara, Akihiro*
Nihon Genshiryoku Gakkai-Shi, 47(2), p.112 - 119, 2005/02
Behavior of LWR fuel during reactivity-initiated accident (RIA) and loss-of-coolant accident (LOCA) is described.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Sasajima, Hideo; Nagase, Fumihisa
NUREG/CP-0185, p.161 - 172, 2004/00
To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Research Institute (JAERI). A series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the NSRR power burst experiments with irradiated PWR fuels with ZIRLO and MDA claddings, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Separate-effect studies including tube-burst and ring-tensile tests on Zircaloy cladding also described.
Yonomoto, Taisuke; Okubo, Tsutomu; Iwamura, Takamichi; Ishida, Toshihisa
Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04
An innovative concept of the pressure suppression system having functions of water injection and non-condensable gas confinement is developed for the next generation light water reactors (LWRs). The use of the system is advantageous for the mitigation of effects of the loss-of-coolant accidents (LOCAs) in (1) keeping the containment pressure as low as for the conventional LWRs, (2) injecting water to the containment for cooling the reactor pressure vessel (RPV) and/or flooding a break, and (3) confining the non-condensable gas in the drywell. The gas confinement function makes the system considerably suitable for reactor designs with passive cooling systems utilizing heat exchangers, such as the steam generator (SG) secondary side cooling system for an integral reactor, and the passive containment cooling system (PCCS), because it avoids adverse effects of non-condensable gas on the heat transfer performance during LOCAs. The usefulness of the developed concept is confirmed in the RELAP5/MOD3 code calculation.
Fuel Safety Research Laboratory
JAERI-Conf 2002-009, 491 Pages, 2002/08
JAERI-Conf-2002-009.pdf:102.1MB
Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's "Advanced LWR Fuel Performance & Safety Research Program". A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting.
Fuel Safety Research Laboratory
JAERI-Conf 2001-010, 303 Pages, 2001/09
JAERI-Conf-2001-010.pdf:59.22MB
The 24th NSRR Technical Review Meeting was held at Tranomon Pastoral, Tokyo, on November 13 and 14, 2000. The purpose of the meeting was to present and discuss the recent progress of the NSRR program and other LWR fuel safety researches at JAERI. Twenty-one papers, including five by foreign institutes, were presented and discussed regarding fuel behavior during normal operation, reactivity initiated accident (RIA) and loss-of-coolant accident (LOCA) and FP release behavior during severe accident. The meeting was a great help in planning future research and promoting research cooperation. This proceeding contains the papers presented in the meeting.
Subcommittee on Improvement of Reactor Thermal-Hydraulic Analysis Codes
JAERI-Review 2000-002, p.105 - 0, 2000/03
JAERI-Review-2000-002.pdf:6.24MB
no abstracts in English
Aritomi, Masanori*; Onuki, Akira; Arai, Kenji*; ; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime
Nihon Genshiryoku Gakkai-Shi, 41(7), p.738 - 757, 1999/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
JAERI-Review 98-006, 194 Pages, 1998/03
JAERI-Review-98-006.pdf:8.23MB
no abstracts in English
Onuki, Akira
Dai-1-Kai Oganaizudo Konsoryu Foramu Koen Rombunshu, p.73 - 82, 1997/00
no abstracts in English
Murao, Yoshio; Ochiai, Masaaki
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 2, p.1075 - 1085, 1997/00
no abstracts in English
Kukita, Yutaka*; Arai, Kenji*;
Nihon Genshiryoku Gakkai-Shi, 39(2), p.151 - 153, 1997/00
no abstracts in English
Nakamura, Takehiko; Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi
Journal of Nuclear Science and Technology, 33(12), p.924 - 935, 1996/12
Times Cited Count：11 Percentile：68.15(Nuclear Science & Technology)no abstracts in English
; ; ; Kunieda, Osamu*; Osaki, Hideki; Anoda, Yoshinari; Kukita, Yutaka
JAERI-Tech 96-004, 74 Pages, 1996/02
no abstracts in English
Kosaka, Atsuo; ; Sugimoto, Jun
10th Pacific Basin Nuclear Conf. (10-PBNC), 1, p.341 - 346, 1996/00
no abstracts in English
Hoshi, Tatsuo
Nihon Genshiryoku Gakkai-Shi, 37(9), p.792 - 794, 1995/00
no abstracts in English
Shibata, Katsuyuki; Isozaki, Toshikuni; Ueda, Shuzo; Kurihara, Ryoichi; Onizawa, Kunio; Kosaka, Atsuo
Nucl. Eng. Des., 153, p.71 - 86, 1994/00
Times Cited Count：11 Percentile：68.64(Nuclear Science & Technology)no abstracts in English
