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Chikazawa, Yoshitaka; Okano, Yasushi; Konomura, Mamoru; Sato, Koji; Ando, Masato*; Nakanishi, Shigeyuki*; Sawa, Naoki*; Shimakawa, Yoshio*
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 8 Pages, 2006/06
A diversified or modular power source is attractive since it requires a low construction cost per unit and can be demonstrated in small scale experimental facilities. In this study, a new metal fuel sodium cooled reactor with 300MW electric has been developed enhancing cost reduction. And economical potential at demonstration stage with first of a kind (FOAK) is emphasized. A minimum configuration with a compact reactor vessel, a one-loop main cooling system and a simple fuel handling system are adopted enhancing cost reduction within safety requirement. Besides, construction cost of a demonstration plant with a first kind of reactor and a small reprocessing and fuel manufacturing facility is also evaluated. A major feature of the present concept is that the demonstration facilities can be appropriated for commercialized ones since they can be easily commercialize by increasing reactor and electrorefiner modules. A FBR cycle commercialization scenario using the present concept is thought to give low risk and high cost performance since the total demonstration plant cost is relatively small and the facilities are directly appropriated to commercial use.
Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 9 Pages, 2006/06
The sodium heated steam generator (SG) of the commercialized FBR being designed in the Feasibility Study is a straight double-wall-tube type and it is large-sized to reduce the manufacturing cost by economics of scale. This paper addresses the temperature and flow multi-dimensional distributions at steady state to obtain the prospect of the SG. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These phenomena might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs. The flow adjustment devices installed in the SG are optimized to prevent these issues, and the temperature distribution properties are uncovered by analysis methods. The analysis model of the SG consists of two parts, a sodium inlet distribution plenum (the plenum) and a heat transfer tubes bundle region (the bundle). The flow and temperature distributions in the plenum and the bundle are evaluated by the three-dimensional flow code "FLUENT" and the two dimensional thermal-hydraulic code "MSG". The MSG code is particularly developed for sodium heated SGs in JAEA. These codes have revealed that the sodium flow is distributed uniformly by the flow adjustment devices, and that the lateral tube temperature distributions remain within the allowable temperature range for the structural integrity of the tubes and the tube to tube-sheet junction.
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.
Sakai, Takaaki; Monji, Hideaki*; Eguchi, Yuzuru*; Iwasaki, Takashi*; Ohshima, Hiroyuki
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 7 Pages, 2006/06
Design method for gas entrainment (GE) from liquid surfaces in a sodium-cooled fast breeder reactor (FBR) was studied. As one of trials to develop the design method by a computation fluid dynamics (CFD) technique, prediction of the GE condition from a surface vortex dimple in a cylindrical tank was performed by using a conventional CFD method. The CFD results showed reasonable agreements with the measured velocity profiles in experiments. Non-dimensional numbers by using the CFD results were proposed for the criteria of GE prevention. The GE design map was drawn by the CFD non-dimensional numbers for the existing CFD database. It was prospective that the gas entrainment can be evaluated by using the map of CFD non-dimensional numbers.
Noguchi, Hiroki; Ota, Hiroyuki; Terada, Atsuhiko; Kubo, Shinji; Onuki, Kaoru; Hino, Ryutaro
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 8 Pages, 2006/06
The Japan Atomic Energy Agency (JAEA) has been conducting R&D on thermo-chemical Iodine-Sulfur (IS) process, which is one of most attractive water-splitting hydrogen production methods using nuclear heat of a high-temperature gas-cooled reactor (HTGR). In the IS process, sulfuric acid is evaporated and decomposed into H
O and SO
in a sulfuric acid decomposer operated under high temperature condition up to 500
C. Necessary heat is supplied by high temperature helium gas from the HTGR. Since the sulfuric acid decomposer will be exposed to severe corrosion condition, we have proposed a new decomposer concept of a block type heat exchanger made of SiC ceramic which has excellent corrosion and mechanical strength performance. To verify the concept, integrity of new type gaskets applied for boundary seal of the decomposer was examined as a first step. Pure gold gaskets coupled with absorption mechanism against thermal expansion showed good seal performance under 500C. Based on this result, a mock-up model for a IS pilot-plant with 30 m
/h-hydrogen production rate was test-fabricated as the next step. Through the fabrication and gastight tests, fabricability and structural integrity were confirmed. Also, the decomposer showed good mechanical strength and seal performances against horizontal loading simulating earthquake motion.
Kubo, Shinji; Futakawa, Masatoshi; Tanaka, Nobuyuki; Iwatsuki, Jin; Yamaguchi, Akihisa*; Tsukada, Ryuji*; Onuki, Kaoru
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), 6 Pages, 2006/06
no abstracts in English
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Misawa, Takeharu; Akimoto, Hajime
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), p.1619 - 1625, 2006/06
R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the FLWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and summarize experimental studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility. Most important objective of the large-scale test is to resolve a fundamental subject whether the core cooling under a tight-lattice configuration is feasible. We have confirmed the thermal-hydraulic feasibility from the experimental results.
Yoshida, Hiroyuki; Onuki, Akira; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), p.1593 - 1600, 2006/06
R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. In FLWR core, there is no sufficient information about the effects of the gap width and grid spacer configuration on the flow characteristics yet. Then, we started development of qualitative analytical procedures on thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this paper, we describe the outline of the simulation technology and examples of numerical results.
Coste, P.*; Quemere, P.*; Roubin, P.*; Emonot, P.*; Tanaka, Masaaki; Kamide, Hideki
Proceedings of 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06) (CD-ROM), p.1626 - 1635, 2006/06
The WATLON experiment, a water facility about fluid mixing in a T-pipe, is calculated with a finite element volume method and a Large Eddy Simulation (LES) approach, with TRIO-U code. Its unstructured tetrahedron grids do not lead to the same noteworthy disagreements previously mentioned with Cartesian grids. Branch and main pipe inlet velocity fluctuations due to turbulence are simulated with the use of "periodic boxes". These more realistic inlet fluctuations allow physical instabilities to develop, improving the predictions. When an elbow is added upstream of the injection, the influence of the secondary flow on temperature averaged values and fluctuations is underlined.
Shiozawa, Shusaku
no journal, ,
First, the expected demand for the nuclear hydrogen is introduced. And the importance and advantage of HTGR hydrogen production is described, followed by the HTTR Project which is underway at JAEA. The recent fruits from the HTTR Project are given with an emphasis of the IS process material development. The possible internal cooperation with the HTTR Project will be suggested in the presentation. Finally the future plan of the HTTR Project is introduced with personal view on perspective of the upcoming hydrogen economy.