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Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950
C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Uchida, Shunsuke; Chimi, Yasuhiro; Kasahara, Shigeki; Hanawa, Satoshi; Okada, Hidetoshi*; Naito, Masanori*; Kojima, Masayoshi*; Kikura, Hiroshige*; Lister, D. H.*
Nuclear Engineering and Design, 341, p.112 - 123, 2019/01
Times Cited Count：5 Percentile：48.99(Nuclear Science & Technology)Improvement of plant reliability based on reliability-centered-maintenance (RCM) is going to be undertaken in NPPs. RCM is supported by risk-based maintenance (RBM). The combination of prediction and inspection is one of the key issues to promote RBM. Early prediction of IGSCC occurrence and its propagation should be confirmed throughout the entire plant systems which should be accomplished by inspections at the target locations followed by timely application of suitable countermeasures. From the inspections, accumulated data will be applied to confirm the accuracy of the code, to tune some uncertainties of the key data for prediction, and then, to increase their accuracy. The synergetic effects of prediction and inspection on application of effective and suitable countermeasures are expected. In the paper, the procedures for the combination of prediction and inspection are introduced.
by tuning nematicity and magnetism in FeSe
S
superconductorsMatsuura, Kohei*; Mizukami, Yuta*; Arai, Yuki*; Sugimura, Yuichi*; Maejima, Naoyuki*; Machida, Akihiko*; Watanuki, Tetsu*; Fukuda, Tatsuo; Yajima, Takeshi*; Hiroi, Zenji*; et al.
Nature Communications (Internet), 8, p.1143_1 - 1143_6, 2017/10
Times Cited Count：80 Percentile：91.57(Multidisciplinary Sciences)Inaba, Yoshitomo; Lee, T.*; Ueta, Shohei; Kasahara, Seiji; Honda, Yuki; Lee, H.*; Kim, E.*; Cho, M.*; Bae, K.*; Sakaba, Nariaki
JAEA-Review 2015-043, 96 Pages, 2016/03
JAEA-Review-2015-043.pdf:79.27MB
The information exchange meeting on HTGR and hydrogen production technology between Korea Atomic Energy research Institute (KAERI) and Japan Atomic Energy Agency (JAEA) was held in the Oarai Research and Development Center of JAEA on November 5th - 6th, 2015 based on the cooperative research program of the KAERI-JAEA implementation of "Development of HTGR and Nuclear Hydrogen Technology" under "The Implementation of Cooperative Program in the Field of Peaceful Uses of Nuclear Energy between KAERI and JAEA." In order to facilitate efficient technology development on the HTGR and nuclear hydrogen by the IS process, both sides mutually showed the present status and future plan of the research and development on the HTGR and nuclear hydrogen technology, respectively. This proceeding summarizes all materials of the presented technical discussions on the HTGR and hydrogen production technology based on the open documents as well as the meeting briefing including collaboration items.
SiTonegawa, Sho*; Kasahara, Shigeru*; Fukuda, Tatsuo; Sugimoto, Kunihisa*; Yasuda, Nobuhiro*; Tsuruhara, Yugo*; Watanabe, Daiki*; Mizukami, Yuta*; Haga, Yoshinori; Matsuda, Tatsuma*; et al.
Nature Communications (Internet), 5, p.4188_1 - 4188_7, 2014/06
Times Cited Count：52 Percentile：89.07(Multidisciplinary Sciences)Shibauchi, Takasada*; Kasahara, Shigeru*; Matsuda, Yuji*; Fukuda, Tatsuo; Sugimoto, Kunihisa*
Nihon Kessho Gakkai-Shi, 55(2), p.128 - 134, 2013/04
no abstracts in English
(AsP)Kasahara, Shigeru*; Shi, H. J.*; Hashimoto, Kenichiro*; Tonegawa, Sho*; Mizukami, Yuta*; Shibauchi, Takasada*; Sugimoto, Kunihisa*; Fukuda, Tatsuo; Terashima, Takahito*; Nevidomskyy, A. H.*; et al.
Nature, 486(7403), p.382 - 385, 2012/06
Times Cited Count：385 Percentile：99.31(Multidisciplinary Sciences)Isobe, Nobuhiro*; Sukekawa, Masayuki*; Nakayama, Yasunari*; Date, Shingo*; Otani, Tomomi*; Takahashi, Yukio*; Kasahara, Naoto; Shibamoto, Hiroshi*; Nagashima, Hideaki*; Inoue, Kazuhiko*
Nuclear Engineering and Design, 238(2), p.347 - 352, 2008/02
Times Cited Count：21 Percentile：78.82(Nuclear Science & Technology)The effect of ratcheting on fatigue strength was investigated in order to rationalize the strain limit as a design criterion of commercialized fast reactor systems. Ratcheting fatigue tests were conducted at 550C. Duration of the ratchet straining was set for a certain number of strain cycles taking the loading condition of fast reactors into account, and the number of cycles for strain accumulation was defined as the ratchet-expired cycle. Fatigue lives decrease as the accumulated strain by ratcheting increases. Fatigue life reduction was negligible when the maximum mean stress was less than 25 MPa, corresponding to an accumulated strain of 2.2 percent. Accumulated strain is limited to 2 percent in the present design guidelines and this strain limit is considered effective to avoid reducing fatigue life by ratcheting. Micro-crack growth behaviors were also investigated in these tests in order to discuss the life reduction mechanisms in ratcheting conditions.
Sakaba, Nariaki; Kasahara, Seiji; Onuki, Kaoru; Kunitomi, Kazuhiko
International Journal of Hydrogen Energy, 32(17), p.4160 - 4169, 2007/12
Times Cited Count：59 Percentile：79.27(Chemistry, Physical)Japan Atomic Energy Agency (JAEA) launched a preliminary design of the hydrogen production system by using heat from the Japan's first high-temperature gas-cooled reactor HTTR from 2005. The thermochemical water-splitting iodine sulphur (IS) process is the progressive candidate for its hydrogen production. The conceptual design of the HTTR-IS system is described in this paper. Since the secondary helium of the HTTR will be utilised in this hydrogen production system, the possibility of utilisation of non-nuclear class IS process as a chemical plant is investigated and available structure of the HTTR-IS system with its approved heat mass balance is proposed. Hydrogen production rate of about 1100 Nm
/h and its thermal efficiency of 44% were shown by flowsheet evaluation of the HTTR-IS system.
Kasahara, Seiji; Kubo, Shinji; Sato, Hiroyuki; Sakaba, Nariaki
JAEA-Technology 2007-040, 31 Pages, 2007/07
JAEA-Technology-2007-040.pdf:13.35MB
Purification methods of hydrogen and oxygen as products of thermochemical hydrogen production iodine sulphur (IS) process thermally connected with the High Temperature Engineering Test Reactor (HTTR) was investigated and evaluated. Present state of R&D of membrane separation method, pressure swing adsorption (PSA) method and cryogenic distillation method was researched and their applicability to the HTTR-IS system was evaluated. At present, PSA method was the most effective due to its feasibility, soundness and reliability with past performance. In addition, hydrogen purification systems by using PSA and membrane separation method were described in this paper.
Homma, Hiroyuki; Sato, Hiroyuki; Kasahara, Seiji; Ohashi, Hirofumi; Hara, Teruo; Kato, Ryoma; Sakaba, Nariaki
JAEA-Technology 2007-006, 60 Pages, 2007/02
JAEA-Technology-2007-006.pdf:15.91MB
At the present time, we are alarmed by depletion of fossil energy and adverse effect of rapid increase in fossil fuel burning on environment such as climate changes and acid rain, because our lives depend still heavily upon fossil energy. It is thus widely recognized that hydrogen is one of important future energy carriers in which it is used without emission of carbon dioxide greenhouse gas and atmospheric pollutants and that hydrogen demand will increase greatly as fuel cells are developed and applied widely in the near future. To meet massive demand of hydrogen, hydrogen production from water utilizing nuclear, especially by thermochemical water-splitting Iodine-Sulphur (IS) process utilizing heat from High-Temperature Gas-cooled Reactors (HTGRs), offers one of the most attractive zero-emission energy strategies and the only one practical on a substantial scale. However, to establish a technology based for the HTGR hydrogen production by the IS process, we should close several technology gaps through R&D with the High-Temperature Engineering Test Reactor (HTTR), which is the only Japanese HTGR built and operated at the Oarai Research & Development Centre of Japan Atomic Energy Agency (JAEA). We have launched design studies of the IS process hydrogen production system coupled with the HTTR (HTTR-IS system) to demonstrate HTGR hydrogen production. In designing the HTTR-IS system, it is necessary to consider preventive and breakdown maintenance against accidents occurred in the IS process as a chemical plant. This report describes case study on chemical plant accidents relating to the IS process plant and shows a proposal of accident protection measures based on above case study, which is necessary for flow-sheet design of the HTTR-IS System.
Sakaba, Nariaki; Kasahara, Seiji; Ohashi, Hirofumi; Sato, Hiroyuki; Kubo, Shinji; Terada, Atsuhiko; Nishihara, Tetsuo; Onuki, Kaoru; Kunitomi, Kazuhiko
Proceedings of 16th World Hydrogen Energy Conference (WHEC-16) (CD-ROM), 11 Pages, 2006/06
High-temperature reactors (HTRs) are particularly attractive due to their wide industrial application from electricity generation to hydrogen production. The Japan Atomic Energy Agency's (JAEA's) HTTR, which is the first HTR in Japan, attained its maximum reactor-outlet coolant temperature and successfully delivered 950C coolant helium outside its reactor vessel. A hydrogen production system based on the thermochemical water-splitting iodine sulphur (IS) process is planned to be connected to the HTTR in the near future. This will establish the hydrogen production technology with an HTR, including the system integration technology for connection of hydrogen production system to HTRs. It will probably be the world's first demonstration of hydrogen production directly using heat supplied from an HTR. The HTTR-IS system design was launched from a conceptual design in 2005. This paper shows the summary of the HTTR, plan for developing the IS process in JAEA, thermal efficiency evaluation for the HTTR-IS system, etc. The verification of the hydrogen production by the HTTR-IS system by using heat from a nuclear reactor is greatly expected to contribute to the commercialization of nuclear hydrogen in coming hydrogen society.
Sakaba, Nariaki; Kasahara, Seiji; Ohashi, Hirofumi; Terada, Atsuhiko; Kubo, Shinji; Onuki, Kaoru; Kunitomi, Kazuhiko
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), p.2238 - 2245, 2006/06
Japan Atomic Energy Agency (JAEA) launched a preliminary design of the hydrogen production system by using heat from Japan's first high-temperature gas-cooled reactor HTTR from fiscal year 2005. The thermochemical water-splitting iodine sulphur (IS) process is the progressive candidate for its hydrogen production. This paper describes the conceptual design of the HTTR-IS system and its evaluated thermal efficiency for the hydrogen production.
Iyoku, Tatsuo; Sakaba, Nariaki; Nakagawa, Shigeaki; Tachibana, Yukio; Kasahara, Seiji; Kawasaki, Kozo
Nuclear Production of Hydrogen, p.167 - 176, 2006/00
no abstracts in English
Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto
JNC TN9400 2003-021, 205 Pages, 2003/04
JNC-TN9400-2003-021.pdf:8.86MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. Aming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1)Multi-language (SoftWIRE.NET, Visual Basic .NET and Fortran) (2)Fortran90 and (3)Python to make a prototype of the next generation code system. As this result, the followings were comfirmed. (1)It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using visual Basic .NET. (2)The maintenanability of the existing code written by Fortran77 can be improved by using the new features of Fortran90. (3)The toolbox-type code system can be built by using Python.
Yokoyama, Kenji; Uto, Nariaki; kasahara, Naoto; ; Ishikawa, Makoto
Nihon Genshiryoku Gakkai 2003-Nen Aki No Taikai, 2(E64), 343 Pages, 2003/00
None
Yokoyama, Kenji; Hosogai, Hiromi*; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; *; *; Ishikawa, Makoto
JNC TN9420 2002-004, 309 Pages, 2002/11
JNC-TN9420-2002-004.pdf:11.4MB
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomina to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical propaties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the phisical propaties and engineering models are replesented as a programming languare or a diagams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language(EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML.
Kasahara, Naoto; *; *; Suril, F.*; Yoshimura, Shinobu*
ASME PVP Vol.365, 0 Pages, 1998/00
None
Kawasaki, Nobuchika; Felix, S.; Kasahara, Naoto; Furukawa, Tomonori*; Komura, Shinobu*; Yagawa, Genki*
PNC TY9602 97-001, 26 Pages, 1997/04
None
; *; *; Kunieda, Osamu*; Osaki, Hideki; Anoda, Yoshinari; Kukita, Yutaka
JAERI-Tech 96-004, 74 Pages, 1996/02
no abstracts in English
