Sato, Hiroyuki; Nomoto, Yasunobu*; Horii, Shoichi*; Sumita, Junya; Yan, X.
Nuclear Engineering and Design, 329, p.247 - 254, 2018/04
This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H plant) so as to confirm that the design meets the requirements with respect to the demonstration test objective. Start-up and shut down operation sequences as well as operability of load following operation were investigated. In addition, system dynamic and control analyses for the test plant in the events of loss of generator load and upset of H plant were performed. The simulation results presented in the paper show that the test plant is suitable for the test bed to validate control schemes against postulated transients in the GTHTR300C. The results also lead us to the conclusion that HTTR-GT/H plant can be used to test operational procedures unique to HTGR direct-cycle gas turbine cogeneration.
Yan, X. L.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu*; Horii, Shoichi*; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.
Nuclear Engineering and Design, 329, p.223 - 233, 2018/04
The pre-licensing design of an HTGR cogeneration test plant to be coupled to JAEA's existing test reactor HTTR is presented. The plant is designed to demonstrate the system of JAEA commercial plant design GTHTR300C. With construction planned to be completed around 2025, the test plant is expected to be the first-of-a-kind nuclear system operating on two of the advanced energy conversion systems attractive for the HTGR closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.
Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Sato, Hiroyuki; Yan, X.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
This paper presents the cost performance design of heat transport piping systems for GTHTR300C plant and HTTR-GT/H plant. Two types of pipe structure are designed and compared in terms of cost performance. Relative to the coaxial double-pipe structure, the insulated single pipe structure is found to have the advantage in overall cost performance considering both the material quantity and the heat loss because it reduces the quantity of steel used for construction. Furthermore it is possible to reduce the heat loss and temperature reduction of hot helium gas by the attachment of the external insulation. The pressure tube made of type-316 stainless steel with high-temperature strength is possible to achieve the same temperature reduction by smaller diameter than that made of 2 1/4Cr-1Mo steel. It contributes to the reduction of the quantity of steel. Specifications of heat transport piping systems for both plants are determined according to these study results.
Sato, Hiroyuki; Nomoto, Yasunobu; Horii, Shoichi; Sumita, Junya; Yan, X.; Ohashi, Hirofumi
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.759 - 766, 2016/11
This paper presents the system performance evaluation for HTTR gas turbine cogeneration test plant (HTTR-GT/H plant) so as to confirm that the design meets the requirements with respect to the demonstration test objective. Start-up and shut down operation sequences as well as operability of load following operation were investigated. In addition, system dynamic and control analyses for the test plant in the events of loss of generator load and upset of H plant were performed. The simulation results presented in the paper show that the test plant is suitable for the test bed to validate control schemes against postulated transients in the commercial Gas Turbine High Temperature Reactor Cogeneration (GTHTR300C). The results also lead us to the conclusion that HTTR-GT/H plant can be used to test operational procedure unique to HTGR direct-cycle gas turbine cogeneration.
Yan, X.; Sato, Hiroyuki; Sumita, Junya; Nomoto, Yasunobu; Horii, Shoichi; Imai, Yoshiyuki; Kasahara, Seiji; Suzuki, Koichi*; Iwatsuki, Jin; Terada, Atsuhiko; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.827 - 836, 2016/11
Pre-licensing basic design for a cogenerating HTGR test plant system is presented. The plant to be coupled to existing 30 MWt 950C test reactor HTTR is intended as a system technology demonstrator for GTHTR300C plant design. More specifically the test plant of HTTR-GT/H aims to (1)demonstrate the licensability of the GTHTR300C for electricity production by gas turbine and hydrogen cogeneration by thermochemical process and (2) confirm the operation control and safety of such cogeneration system. With construction and operation completion by 2025, the test plant is expected to be the first of a kind HTGR-powered cogeneration plant operating on the two advanced energy conversion systems of closed cycle helium gas turbine for power generation and thermochemical iodine-sulfur water-splitting process for hydrogen production.
Imai, Yoshiyuki; Sato, Hiroyuki; Nomoto, Yasunobu; Yan, X.
JAEA-Data/Code 2016-007, 27 Pages, 2016/08
This report provides design database for helium gas turbine for HTGR. The design details for secondary helium cooling system and helium gas turbine in a heat application system to be coupled to the HTTR are described. In addition, experimental data for fission product isotope diffusion through turbine blade alloy is compiled.
Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi
JAEA-Data/Code 2015-008, 39 Pages, 2015/06
HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.
Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi
JAEA-Technology 2014-038, 51 Pages, 2014/12
The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.
Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2013-017, 71 Pages, 2014/02
Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.
Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio
JAEA-Technology 2013-016, 176 Pages, 2013/09
JAEA has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750C to 900C and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S.
Ohashi, Hirofumi; Sato, Hiroyuki; Goto, Minoru; Yan, X.; Sumita, Junya; Tazawa, Yujiro*; Nomoto, Yasunobu; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; et al.
International Journal of Nuclear Energy, 2013, p.918567_1 - 918567_18, 2013/00
Japan Atomic Energy Agency (JAEA) has conducted a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for multiple heat applications, named HTR50S, with the reactor outlet coolant temperature of 750 C and 900 C. It is first-of-a-kind of the commercial plant or a demonstration plant of a small-sized HTGR system to deploy it in developing countries in the 2020s. The design concept of HTR50S is to satisfy the user requirements for multipurpose heat application, to upgrade its performance compared to that of HTTR without significant R&D utilizing the knowledge obtained by the HTTR design and operation, and to fulfill the high level of safety by utilizing the inherent features of HTGR and a passive decay heat removal system.
Imaizumi, Tomomi; Miyauchi, Masaru; Ito, Masayasu; Watahiki, Shunsuke; Nagata, Hiroshi; Hanakawa, Hiroki; Naka, Michihiro; Kawamata, Kazuo; Yamaura, Takayuki; Ide, Hiroshi; et al.
JAEA-Technology 2011-031, 123 Pages, 2012/01
The number of research reactors in the world is decreasing because of their aging. However, the planning to introduce the nuclear power plants is increasing in Asian countries. In these Asian countries, the key issue is the human resource development for operation and management of nuclear power plants after constructed them, and also the necessity of research reactor, which is used for lifetime extension of LWRs, progress of the science and technology, expansion of industry use, human resources training and so on, is increasing. From above backgrounds, the Neutron Irradiation and Testing Reactor Center began to discuss basic concept of a multipurpose low-power research reactor for education and training, etc. This design study is expected to contribute not only to design tool improvement and human resources development in the Neutron Irradiation and Testing Reactor Center but also to maintain and upgrade the technology on research reactors in nuclear power-related companies. This report treats the activities of the working group from July 2010 to June 2011 on the multipurpose low-power research reactor in the Neutron Irradiation and Testing Reactor Center and nuclear power-related companies.
Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Suzuki, Satoshi; Ezato, Koichiro; Hatano, Toshihisa; Nomoto, Yasunobu; Nishi, Hiroshi; Homma, Takashi; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
This paper presents progresses of the test strategy development, design and supporting R&Ds of solid breeder Test Blanket Modules (TBMs) in Japan. Japan is proposing to test its unique designs of Water Cooled Solid Breeder (WCSB) TBM and Helium Cooled Solid Breeder (HCSB) TBM from the beginning of the ITER operation. As for the design work, structure design showed steady progress and clarified detailed structure taking into account the fabrication procedure. As for supporting R&Ds, the corrosion characteristics of the structural material by high temperature and pressure water was clarified as one of critical structure integrity issues. Also, important design data of the breeder pebble bed has been clarfied. Along with the development progress, the test strategy has been investigated to obtain the most effective results of TBM test program.
Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2006/10
no abstracts in English
Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.
Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02
no abstracts in English
Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; et al.
JAERI-Research 2004-008, 383 Pages, 2004/06
The present report contains the achievement of "Research and Development on Reduced-Moderation Light Water Reactor with Passive Safety Features", which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies.
Kuroda, Toshimasa*; Sato, Kazuyoshi; Akiba, Masato; Ezato, Koichiro; Enoeda, Mikio; Osaki, Toshio*; Kosaku, Yasuo; Sato, Satoshi; Sato, Shinichi*; Suzuki, Satoshi*; et al.
JAERI-Tech 2002-044, 25 Pages, 2002/03
no abstracts in English
Takahashi, Koji; Imai, Tsuyoshi; Mori, Kensuke*; Mori, Seiji*; Nomoto, Yasunobu*
JAERI-Research 2000-036, 26 Pages, 2000/09
no abstracts in English
Hirose, Takanori; Nomoto, Yasunobu; Enoeda, Mikio; Akiba, Masato
no journal, ,
161.5mm3500mm cylindrical tubes and 11mm1.5mm3500mm square tubes were successfully fabricated from F82H bullet by cold rolling. The tube is long enough to fabricate the first wall without any joint in the cooling path. The surface roughness (Rz) is less than 1m and it is smooth enough to HIP bonding without further surface finish. As for the HIP pretreatment, the degassing conditions and surface roughness were investigated as parameters of pretreatment conditions in this work. F82H was HIPped with the conditions of 1373K, 150MPa and 2h holding time. Normalizing at 1233K for 0.5h and tempering at 1023K were performed as post-HIP heat treatment. Although tensile property of the HIP joint was not affected by the surface roughness and degassing conditions, the impact property was significantly degraded by lack of degassing. The degradation in impact property was caused by the micro dispersed oxide on the bonding surface. The oxide was decomposed by thermal aging at 1313K and the aging recovered the impact property. Therefore, to complete HIP bonding, degassing conditions must be strictly controlled to avoid the oxidation of bonding surface. And the inappropriately degassed joint should be followed by thermal aging to decompose the oxides.
Nomoto, Yasunobu; Hirose, Takanori; Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Enoeda, Mikio; Akiba, Masato
no journal, ,
A detailed study was conducted for water cooled, solid breeder, ferritic steel structure blanket as a Japanese primary candidates for ITER Test Blanket Module (TBM). The wall structure of TBM has built-in cooling channels, which is fabricated by Hot Isostatic Pressing (HIP) method using square tubes and thin plates. This paper presents a structural concept of water cooled, solid breeder, ferritic steel structure TBM and the key issues of structural material and the joining process for demonstration of the first wall structure.