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Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki
Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08
Times Cited Count：2 Percentile：68.64(Nuclear Science & Technology)HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.
Ohno, Shuji; Maeda, Seiichiro
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09
Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.
Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. The local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1
C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Takamatsu, Kuniyoshi; Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Shoji; Yan, X.; Sawa, Kazuhiro; Sakaba, Nariaki; Kunitomi, Kazuhiko
Journal of Nuclear Science and Technology, 51(11-12), p.1427 - 1443, 2014/11
Times Cited Count：15 Percentile：72.02(Nuclear Science & Technology)In a safety demonstration test involving a loss of both reactor reactivity control and core cooling, HTGRs such as the HTTR, which is the only HTGR in Japan, demonstrate that the reactor power would stabilize spontaneously. In the test at an initial power of 30%, when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero, a reactor transient was initiated and examined. The results confirmed that the reactor power would decrease immediately and become effectively zero.
Inoue, Takashi; Hanada, Masaya; Kashiwagi, Mieko; Nishio, Satoshi; Sakamoto, Keishi; Sato, Masayasu; Taniguchi, Masaki; Tobita, Kenji; Watanabe, Kazuhiro; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1291 - 1297, 2006/02
Times Cited Count：12 Percentile：62.01(Nuclear Science & Technology)Requirement and technical issues of the neutral beam inejctor (NBI) is discussed for fusion DEMO plant. The NBI for the fusion DEMO plant should be high efficiency, high energy and high reliability with long life. From the view point of high efficiency, use of conventional electrostatic accelerator is realistic. Due to operation under radiation environment, vacuum insulation is essential in the accelerator. According to the insulation design guideline, it was clarified that the beam energy of 1.5
2 MeV is possible in the accelerator. Development of filamentless, and cesium free ion source is required, based on the existing high current/high current density negative ion production technology. The gas neutralization is not applicable due to its low efficiency (60%). R&D on an advanced neutralization scheme such as plasma neutralization (efficiency: 
80%) is required. Recently, development of cw high power semiconductor laser is in progress. The paper shows a conceptual design of a high efficiency laser neutralizer utilizing the new semiconductor laser array.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.
Naka Fusion Research Establishment
JAERI-Review 2005-046, 113 Pages, 2005/09
JAERI-Review-2005-046.pdf:24.35MB
This annual report provides an overview of research and development activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2004 to 31 March, 2005. The activities in the Naka Fusion Research Establishment are highlighted by researches in JT-60 and JFT-2M, theoretical and analytical plasma researches, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
JAERI-Data/Code 2005-003, 31 Pages, 2005/06
JAERI-Data-Code-2005-003.pdf:4.83MB
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.
Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*
JAERI-Tech 2005-015, 26 Pages, 2005/03
JAERI-Tech-2005-015.pdf:1.77MB
Safety demonstration tests using the HTTR are in progress since 2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.
Naka Fusion Research Establishment
JAERI-Review 2004-023, 126 Pages, 2004/11
JAERI-Review-2004-023.pdf:28.0MB
no abstracts in English
Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10
Times Cited Count：23 Percentile：79.53(Nuclear Science & Technology)Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.
Naka Fusion Research Establishment
JAERI-Review 2003-035, 129 Pages, 2003/11
JAERI-Review-2003-035.pdf:13.55MB
This annual report provides an overview of research and development (R&D) activities at Naka Fusion Research Establishment in collaboration with other research establishment of JAERI, research institutes, and universities during the period from 1 April, 2002 to 31 March, 2003. The activities in the Naka Fusion Research Establishment are highlighted by high performance plasma researches in JT-60 and JFT-2M, R&D of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.
Nishitani, Takeo; Ochiai, Kentaro; Klix, A.; Verzilov, Y. M.; Sato, Satoshi; Yamauchi, Michinori*; Nakao, Makoto*; Hori, Junichi; Enoeda, Mikio
Proceedings of 20th IEEE/NPSS Symposium on Fusion Engineering (SOFE 2003), p.454 - 457, 2003/10
no abstracts in English
Research Committee for Fusion Reactor; Research Committee for Fusion Materials
JAERI-Review 2003-015, 123 Pages, 2003/05
JAERI-Review-2003-015.pdf:24.89MB
no abstracts in English
Seki, Masahiro; Yamanishi, Toshihiko; Shu, Wataru; Nishi, Masataka; Hatano, Toshihisa; Akiba, Masato; Takeuchi, Hiroshi; Nakamura, Kazuyuki; Sugimoto, Masayoshi; Shiba, Kiyoyuki; et al.
Fusion Science and Technology, 42(1), p.50 - 61, 2002/07
Times Cited Count：5 Percentile：33.92(Nuclear Science & Technology)Latest status on development of long-term fusion nuclear technologies at JAERI is overviewed. A tritium processing system for the ITER and DEMO reactors was designed and basic technologies for each component of this system was demonstrated successfully by an operation of the integrated system for one month. An ultra-violet laser with a wave length of 193 nm was found quite effective for removing tritium from in-vessel components of D-T fusion reactors. Blanket technologies have been developed for the Test Blanket Module of the ITER and for advanced blankets for DEMO reactors. This blanket is composed of Li
TiO
breeder pebbles and neutron multiplier Be pebbles, contained in a box structures made of a reduced activation ferritic steel F82H. Mechanical properties of F82H under neutron irradiation up to 50 dpa were obtained in a temperature range from 200 to 500C. Design of the International Fusion Materials Irradiation Facility (IFMIF) has been developed so as to obtain engineering data for candidate materials for DEMO reactors, under neutron irradiation up to 100-200 dpa.
Research Committee for Fusion Reactor; Research Committee for Fusion Materials
JAERI-Review 2002-008, 79 Pages, 2002/03
JAERI-Review-2002-008.pdf:9.92MB
Joint research committee for fusion reactor and materials was held in Tokyo on July 16, 2001. In the committee, a review of the development programs and the present status on the blanket technology, materials and IFMIF(International Fusion Materials Irradiation Facility) in JAERI and Japanese Universities was reported, and the direction of these R&D was discussed. Moreover, the progress of the collaboration between JAERI and Japanese Universities was discussed. This report consists of the summaries of the presentations and the viewgraphs which were used at the committee.
Takatsu, Hideyuki; Kawamura, Hiroshi; Tanaka, Satoru*
Fusion Engineering and Design, 39-40, p.645 - 650, 1998/09
Times Cited Count：17 Percentile：77.20(Nuclear Science & Technology)no abstracts in English
Hoshi, Tatsuo
Nikkei Konsutorakushon, 0(42), p.24 - 29, 1991/06
no abstracts in English
Yanagihara, Satoshi; ; ; Fujiki, Kazuo
1st JSME/ASME Joint Int. Conf. on Nuclear Engineering, p.65 - 70, 1991/00
no abstracts in English
