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Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*; Harai, Yasutaka*; Sasaki, Gaku*; Shingeya, Hideshi*; Yamashita, Toshiyuki*; Yoneda, Jiro*; Okabayashi, Kazuki*; Sakamoto, Hiroyuki*; et al.
JAEA-Data/Code 2025-012, 151 Pages, 2025/12
JAEA-Data-Code-2025-012.pdf:9.69MB
Based on the lessons learned from the Fukushima Daiichi Nuclear Power Station accident, we have highly paid attention to the advancement of the fundamental technologies which are indispensable in timely response to hydrogen safety measures and assessments especially in both nuclear reactors and decommissioning. Focusing on this attention, we developed an analysis system that predicts the behavior of hydrogen from generation to diffusion, combustion and explosion. The system utilizes the commercial computational fluid dynamics software (FLUENT, AUTODYN), and incorporates new modules and pre/post-processors in order to withstand the general practical use. We also developed a system by utilizing open-source code (OpenFOAM) that can be used in hydrogen disaster prevention plans for nuclear facilities. So far, we have expanded the system to deal with the phenomena that should be considered from the practical point of view for PWR (Pressurized Water Reactor) in nuclear power plants. This report summarizes the overview of the integrated analysis system for hydrogen behavior, the handling method, and real scale analysis examples.
Motegi, Kosuke; Matsumoto, Toshinori; Shiotsu, Hiroyuki
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 10 Pages, 2025/08
Tanaka, Hirohisa*; Matsumura, Daiju; 8 of others*
International Journal of Hydrogen Energy, 141, p.1317 - 1327, 2025/06
Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*
JAEA-Review 2024-049, 400 Pages, 2025/03
JAEA-Review-2024-049.pdf:13.94MB
In the aftermath of the Fukushima Daiichi Nuclear Power Station accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. As one of efforts to form a common knowledge base between nuclear engineers and experts on combustion and explosion, we issued the "Handbook of Advanced Nuclear Hydrogen Safety (1st edition)" in 2017. For improvement of the rational advancement of hydrogen safety measures and further reliability of hydrogen safety evaluation, a CFD analysis is highly expected to produce more precisely and quantitative results. We have been developing an integrated CFD analysis code system which can analyze hydrogen diffusion, explosion-combustion and structural integrity at the severe accident especially for pressurized water reactors (PWRs). We organized the role of LP and the CFD analyses and their utilization examples of hydrogen safety validation. Based on these results, we made the "Handbook of Advanced Nuclear Hydrogen Safety (2nd volume)". The analysis results of real scale PWR described in 2nd volume are confirmed by cross-analysis models and existing data obtained through representative small, medium and large-scale tests.
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/11
Cantarel, V.; Lambertin, D.*; Labed, V.*; Yamagishi, Isao
Journal of Nuclear Science and Technology, 58(1), p.62 - 71, 2021/01
Times Cited Count：7 Percentile：51.06(Nuclear Science & Technology)The gas production of wasteforms is a major safety concern for encapsulating active nuclear wastes. For geopolymers and cements, the H
produced by radiolytic processes is a key factor because of the large amount of water present in their porous structure. Herein, the gas composition evolution around geopolymers was monitored on line under 
Co gamma irradiation. Transient evolution of the hydrogen production yield was measured for samples with different formulations. The rate of its evolution and the final values are consistent with the presence of a chemical reaction of the pseudo-first order consuming hydrogen in the samples. The results show this phenomenon can significantly reduce the hydrogen source term of geopolymer wasteform provided their diffusion constant remains low. Lower hydrogen production rates and faster kinetics were observed with geopolymers formulations in which pore water pH was higher. Besides hydrogen production, a steady oxygen consumption was observed for all geopolymers samples. The oxygen consumption rates are proportional to the diffusion constants estimated in the modelization of hydrogen recombination by a pseudo first order reaction.
Trianti, N.; Motegi, Kosuke; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08
Kunitomi, Kazuhiko; Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(4), p.236 - 240, 2018/04
High temperature gas-cooled reactor (HTGR) is a graphite-moderated and helium-gas-cooled thermal-neutron reactor that has excellent safety features and can produce high temperature heat of 950
C. It is expected to use for various heat applications as well as for electricity generation to reduce carbon dioxide emission. Japan Atomic Energy Agency (JAEA) has been promoted research and development to demonstrate the HTGR safety features using High temperature engineering test reactor (HTTR) and it's heat application. JAEA are also conducting the action to international deployment of Japanese HTGR technologies in cooperation with industries-government-academia. This paper reports status of the research and development of HTGR and domestic and international collaborations.
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.
Tachibana, Yukio
Genshiryoku Nenkan 2005-Nen Ban, p.279 - 287, 2005/00
no abstracts in English
C in HTTRFujikawa, Seigo; Hayashi, Hideyuki; Nakazawa, Toshio; Kawasaki, Kozo; Iyoku, Tatsuo; Nakagawa, Shigeaki; Sakaba, Nariaki
Journal of Nuclear Science and Technology, 41(12), p.1245 - 1254, 2004/12
Times Cited Count：98 Percentile：97.85(Nuclear Science & Technology)A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30MW and reactor-outlet coolant temperature of 950C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.
Shimizu, Akira; Nishihara, Tetsuo; Moriyama, Koichi*
JAERI-Tech 2004-051, 69 Pages, 2004/06
JAERI-Tech-2004-051.pdf:4.68MB
HTTR of JAERI will be connected with a hydrogen production system by steam reforming of methane for development of nuclear heat utilization technology. This facility will handle much inflammable gas near the nuclear reactor so that special safety consideration is necessary. This report describes the Probabilistic Safety Assessment (PSA) of inflammable gas leakage in the HTTR hydrogen production system. Vessels and pipes, which contain flammable gas, were divided into several systems. Probability of gas leakage were calculated at all candidate places. As a result of assessment, the counter measures such as double-covered inflammable gas pipes, small diameter instrument pipes, leakage detector and emergency shut off valves, are confirmed to be very effective to minimize the scale of explosion and to prevent the damage on nuclear plant.
Tachibana, Yukio
Genshiryoku Nenkan 2004-Nen Ban, p.79 - 87, 2003/11
no abstracts in English
Nishihara, Tetsuo; Shimizu, Akira; Tanihira, Masanori*; Uchida, Shoji*
JAERI-Tech 2002-101, 46 Pages, 2003/01
no abstracts in English
Shiozawa, Shusaku
Nihon Kikai Gakkai Doryoku Enerugi Shisutemu Bumon Nyusu Reta, (24), p.2 - 3, 2002/05
no abstracts in English
Nishihara, Tetsuo; Hada, Kazuhiko
Nihon Genshiryoku Gakkai-Shi, 41(5), p.571 - 578, 1999/05
Times Cited Count：5 Percentile：39.73(Nuclear Science & Technology)no abstracts in English
; Miyamoto, Yoshiaki; Akino, Norio; Shiina, Yasuaki; Hishida, Makoto*; Ogawa, Masuro; Fumizawa, Motoo; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Takada, Shoji; et al.
JAERI-Review 98-024, 403 Pages, 1999/01
JAERI-Review-98-024.pdf:17.17MB
no abstracts in English
Nishihara, Tetsuo; Hada, Kazuhiko; Shiozawa, Shusaku
JAERI-Research 97-022, 110 Pages, 1997/03
JAERI-Research-97-022.pdf:4.1MB
no abstracts in English
Fuketa, Toyoshi; Ishijima, Kiyomi;
Journal of Nuclear Science and Technology, 33(1), p.43 - 51, 1996/01
Times Cited Count：4 Percentile：38.57(Nuclear Science & Technology)no abstracts in English
Saito, Shinzo
Sekiyu Gakkai-Shi, 34(6), p.486 - 499, 1991/00
no abstracts in English
