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Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09
After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07
In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Proceedings of 2005 ASME International Mechanical Engineering Congress and Exposition (CD-ROM), 8 Pages, 2005/11
no abstracts in English
Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao
JAERI-Review 2005-029, 119 Pages, 2005/09
JAERI-Review-2005-029.pdf:11.01MB
The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.
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
Takase, Kazuyuki
JAERI-Research 94-034, 28 Pages, 1994/11
JAERI-Research-94-034.pdf:1.12MB
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：47.13(Nuclear Science & Technology)no abstracts in English
Kaminaga, Masanori; Sudo, Yukio
Nihon Kikai Gakkai Rombunshu, B, 58(553), p.2799 - 2804, 1993/09
no abstracts in English
Sudo, Yukio; Kaminaga, Masanori
J. Heat Transfer, 115, p.426 - 434, 1993/05
Times Cited Count：54 Percentile：92.62(Thermodynamics)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
Kaminaga, Masanori; Sudo, Yukio
Proc. of the 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering,Vol. 1, p.73 - 79, 1991/00
no abstracts in English
Takase, Kazuyuki; Hino, Ryutaro;
Nihon Genshiryoku Gakkai-Shi, 32(11), p.1107 - 1110, 1990/11
Times Cited Count：12 Percentile：86.08(Nuclear Science & Technology)no abstracts in English
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Nihon Genshiryoku Gakkai-Shi, 29(2), p.133 - 140, 1987/02
Times Cited Count：1 Percentile：19.15(Nuclear Science & Technology)no abstracts in English
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Nucl.Eng.Des., 102, p.11 - 20, 1987/00
Times Cited Count：7 Percentile：55.18(Nuclear Science & Technology)no abstracts in English
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JAERI-M 85-187, 98 Pages, 1985/11
no abstracts in English
Kaminaga, Masanori; ; ; Sudo, Yukio
JAERI-M 85-071, 65 Pages, 1985/06
no abstracts in English
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Journal of Nuclear Science and Technology, 20(3), p.246 - 253, 1983/00
Times Cited Count：4 Percentile：44.38(Nuclear Science & Technology)no abstracts in English
