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Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.
Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02
Times Cited Count：1 Percentile：11.8(Nuclear Science & Technology)The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO
) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.
Goto, Minoru; Demachi, Kazuyuki*; Ueta, Shohei; Nakano, Masaaki*; Honda, Masaki*; Tachibana, Yukio; Inaba, Yoshitomo; Aihara, Jun; Fukaya, Yuji; Tsuji, Nobumasa*; et al.
Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) (USB Flash Drive), p.507 - 513, 2015/09
A concept of a plutonium burner HTGR named as Clean Burn, which has a high nuclear proliferation resistance, had been proposed by Japan Atomic Energy Agency. In addition to the high nuclear proliferation resistance, in order to enhance the safety, we propose to introduce PuO-YSZ TRISO fuel with ZrC coating to the Clean Burn. In this study, we conduct fabrication tests aiming to establish the basic technologies for fabrication of PuO-YSZ TRISO fuel with ZrC coating. Additionally, we conduct a quantitative evaluation of the security for the safety, a design of the fuel and the reactor core, and a safety evaluation for the Clean Burn to confirm the feasibility. This study is conducted by The University of Tokyo, Japan Atomic Energy Agency, Fuji Electric Co., Ltd., and Nuclear Fuel Industries, Ltd. It was started in FY2014 and will be completed in FY2017, and the first year of the implementation was on schedule.
Tsuji, Nobumasa*; Ohashi, Kazutaka*; Tazawa, Yujiro*; Tachibana, Yukio; Ohashi, Hirofumi; Takamatsu, Kuniyoshi
FAPIG, (190), p.20 - 24, 2015/07
In a loss of forced cooling accident, decay heat in HTGRs must be removed by radiation, thermal conduction and natural convection. Passive heat removal performance is of primary concern for enhancing inherent safety features of HTGRs. Therefore, the thermal hydraulic analyses for normal operation and a loss of forced cooling accident are conducted by using thermal hydraulic CFD code. And further, a multi-hole type fuel block of MHTGR is also modeled and the flow and heat transfer characteristics are compared with a pin-in-block type fuel block.
Tsuji, Nobumasa*; Nakano, Masaaki*; Takada, Eiji*; Tokuhara, Kazumi*; Ohashi, Kazutaka*; Okamoto, Futoshi*; Tazawa, Yujiro; Inaba, Yoshitomo; Tachibana, Yukio
Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 9 Pages, 2012/10
Passive heat removal performance of the reactor vessel cavity cooling system (RCCS) is of primary concern for enhanced inherent safety of HTGR. In a loss of forced cooling accident, decay heat must be removed by radiation and natural convection of RCCS. Thus thermal hydraulic analysis of reactor internals and RCCS is powerful means for evaluation of the heat removal performance of RCCS. The thermal hydraulic analyses using CFD computation tools are conducted for normal operation of the High Temperature Engineering Test Reactor (HTTR) and are compared to the temperature distribution of measured data. The calculated temperatures on outer faces of the permanent side reflector (PSR) blocks are in fair agreement with measured data. The transient analysis for decay heat removal mode in HTTR is also conducted.
C high temperature gas-cooled reactorNakano, Masaaki*; Takada, Eiji*; Tsuji, Nobumasa*; Tokuhara, Kazumi*; Ohashi, Kazutaka*; Okamoto, Futoshi*; Tazawa, Yujiro; Tachibana, Yukio
Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 6 Pages, 2012/10
The conceptual core design study of High Temperature Gas-cooled Reactor (HTGR) is performed. The major specifications are 600 MW thermal output, 950C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, 
Ag and 
Cs from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.
Ohashi, Kazutaka*; Nishihara, Tetsuo; Tazawa, Yujiro; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2008-093, 56 Pages, 2009/03
JAEA-Technology-2008-093.pdf:2.31MB
As HTGR hydrogen production systems, such as HTTR-IS system or GTHTR300C currently being developed by Japan Atomic Energy Agency, consists of nuclear reactor and chemical plant, which are without a precedent in the world, safety design philosophy and regulatory framework should be newly developed. In this report, phenomena to be considered and events to be postulated in the safety evaluation of the HTGR hydrogen production systems were investigated and basic principles to establish acceptance criteria for the explosion and toxic gas release accidents were provided. Especially for the explosion accident, quantitative criteria to the reactor building are proposed with relating sample calculation results. It is necessary to treat abnormal events occurred in the hydrogen production system as external events to the nuclear plant in order to classify the hydrogen production system as no-nuclear facility and basic policy to meet such requirement was also provided.
Yan, X.; Takeda, Tetsuaki; Nishihara, Tetsuo; Ohashi, Kazutaka; Kunitomi, Kazuhiko; Tsuji, Nobumasa*
Nuclear Technology, 163(3), p.401 - 415, 2008/09
Times Cited Count：12 Percentile：61.64(Nuclear Science & Technology)A rupture of primary piping in HTGR represents a design basis event. In such a loss of coolant event a safety issue remains graphite oxidation damage to fuel and core should major air ingress take place through the breached primary boundary. The present study deals with the two most probable cases of air ingress. The first results from rupture of a standpipe. A design change proposed in the vessel top structure intends to rule out any probability of a standpipe rupture. The feasibility of the modified structure is evaluated. The second case results from rupture of a main coolant pipe. Experiment and analysis are performed to gain understanding of the multi-phased air ingress phenomena and accordingly a new mechanism of sustained counter-air diffusion is proposed that is fully passive and effective in preventing major air ingress in the event of main coolant pipe rupture. The results of the present study may lead to improved safety and economic design of the HTGR.
Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.
JAEA-Technology 2008-019, 57 Pages, 2008/03
JAEA-Technology-2008-019.pdf:8.59MB
The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.
Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko; Nakano, Masaaki*; Tazawa, Yujiro*; Okamoto, Futoshi*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(1), p.32 - 43, 2008/03
Interests on the development of the Very High-Temperature Gas-Cooled Reactor (VHTR), of which the reactor outlet temperature is 950C or much higher, are recently increasing world-widely and it was selected as one of the candidate reactor types of the GIF. Japan Atomic Energy Agency has already initiated R&D efforts on the electricity and hydrogen co-generation plant with VHTR system, GTHTR300C. Although technical feasibility of its VHTR reactor using Pin-in-block fuel, which has experience to be already used in the HTTR, has been shown fundamentally, more improvements of the core performances, such as decrease of the occupational exposure doses during the plant maintenance, are desired. This report presents the results of the conceptual core design study using Multi-hole type fuel and the study on the occupational exposure doses. The latter results shows much better plant maintainability compared to the previous results of the GTHTR-300.
Sakaba, Nariaki; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko
JAEA-Research 2007-058, 31 Pages, 2007/08
JAEA-Research-2007-058.pdf:16.44MB
Since hydrogen produced by nuclear should be competitiveness economically compared by other method in a hydrogen society, it is important to built hydrogen production system to be coupled with the reactor as a conventional chemical plant. Japan Atomic Energy Agency started the safety study to establish a new safety philosophy with safety requirements and with considering non-nuclear grade hydrogen production system for the nuclear coupling system. Also, structural concepts with integrating functions for the Bunsen reactor and sulphuric acid decomposer were proposed to reduce construction cost of the IS process hydrogen production system. In addition, HI decomposer which enabled the process condition to be eased and could adopt conventional materials and technologies was studied. Moreover, basic approval of the HTTR-IS system in which the hydrogen production rate of 1,000 Nm
/h by using the supplied heat of 10 MW from the intermediate heat exchanger of the HTTR was confirmed. This paper describes the conceptual design of the HTTR-IS hydrogen production system.
Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(1), p.46 - 57, 2007/03
Japan Atomic Energy Agency (JAEA) has been conducting an R&D work on the VHTR reactor system and IS hydrogen production system to realize hydrogen production using nuclear heat. As a part of this activity, JAEA is planning to connect an IS test system to the High Temperature Engineering Test Reactor (HTTR) to demonstrate its technical feasibility. This paper proposes a fundamental philosophy on the safety design of the HTTR-IS hydrogen production system including the methodology to select postulated abnormal events and its event sequences and to define safety functions of the IS system to ensure the reactor safety. Also the measure to clarify the IS system as non-reactor system is proposed.
Nishihara, Tetsuo; Ohashi, Kazutaka; Murakami, Tomoyuki; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(4), p.325 - 333, 2006/12
A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) based on the achievement of gas turbine high temperature reactor design has been carried out in Japan Atomic Energy Agency. The safety design philosophy of the GTHTR300C to keep hydrogen economy and to attract a lot of interest from non nuclear industries is discussed. The hydrogen production system which is coupled to the secondary helium loop of the intermediate heat exchanger installed upstream of the gas turbine system shall be designed as a non nuclear graded system. General nuclear safety shall be ensured by the items installed in the reactor system. Functions of secondary helium loop which are primary helium cooling and pressure control and purification of secondary helium are required to continue normal operation. Means to maintain these functions are proposed by using equipment of the reactor system and the gas turbine system without the hydrogen production system so that the power generation can continue independently of operational state of the hydrogen production system. Means of protection against external event of flammable and/or toxic gas release are also considered.
Nakagawa, Shigeaki; Sawa, Kazuhiro; Ohashi, Kazutaka*
Journal of Nuclear Science and Technology, 30(6), p.579 - 588, 1993/06
Times Cited Count：1 Percentile：18.76(Nuclear Science & Technology)no abstracts in English
*; Ohashi, Kazutaka*; Iyoku, Tatsuo
FAPIG, 0(129), p.13 - 21, 1991/11
no abstracts in English
*; Ohashi, Kazutaka*; *; *; Sawa, Kazuhiro; Nakagawa, Shigeaki; Kunitomi, Kazuhiko
FAPIG, 0(127), p.11 - 19, 1991/03
no abstracts in English
Nakagawa, Shigeaki; Sawa, Kazuhiro; Ohashi, Kazutaka*
Proc. of the 1st JSME/ASME Joint Int. Conf. on Nuclear Engineering,Vol. 2, p.299 - 304, 1991/00
no abstracts in English
Nakagawa, Shigeaki; Mitake, Susumu*; Ohashi, Kazutaka*; Hirano, Mitsumasa
JAERI-M 89-013, 44 Pages, 1989/02
no abstracts in English
Ohashi, Kazutaka*; Mitake, Susumu; Suzuki, Katsuo; Tazawa, Yujiro*
FAPIG, (116), p.11 - 17, 1987/07
no abstracts in English
Tazawa, Yujiro*; Okamoto, Futoshi*; Ohashi, Kazutaka; Kunitomi, Kazuhiko
no journal, ,
no abstracts in English
Tachibana, Yukio; Ohashi, Hirofumi; Sato, Hiroyuki; Suzuki, Tetsu*; Ohashi, Kazutaka*; Mori, Yuichiro*; Furihata, Noboru*
no journal, ,
no abstracts in English
