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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.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
Journal of Nuclear Science and Technology, 46(7), p.690 - 698, 2009/07
Times Cited Count：9 Percentile：53.3(Nuclear Science & Technology)Thermal conductivity of graphite components in HTGR is reduced by neutron irradiation. The reduced thermal conductivity is expected to be recovered by thermal annealing when irradiated graphite component is heated above irradiation temperature. In this study, the thermal conductivities of IG-110 graphite for the VHTR were measured systematically and thermal annealing effect was evaluated quantitatively. As the results, the thermal conductivities of IG-110 graphite were recovered up to 80% of unirradiated ones at maximum and the thermal annealing effect of IG-110 on thermal conductivity could be evaluated quantitatively using proposed thermal annealing evaluation model based on experimental results. Moreover, the calculated thermal conductivities of IG-110 with modified thermal resistance model were good agreement with experimental ones more than irradiation temperature. It implies that modified thermal resistance model can predict the thermal conductivity of IG-110.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
JAEA-Research 2008-036, 33 Pages, 2008/03
JAEA-Research-2008-036.pdf:3.9MB
To advance the performance and safety of HTGR, heat-resistant ceramic materials are expected to be used as reactor internals of HTGR. C/C composite and superplastic zirconia are the promising materials for this purpose. In order to use these new materials as reactor internals in HTGR, it is necessary to establish a structure design method to guarantee the structural integrity under environmental and load conditions. Therefore, C/C composite expected as reactor internals of VHTR is focused and an investigation on the structural design method applicable to the C/C composite and a basic applicability of the C/C composite to representative structures of HTGR were carried out in this report. As the results, it is found that the competing risk theory for the strength evaluation of the C/C composite is applicable to design method and C/C composite is expected to be used as reactor internals of HTGR.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Iyoku, Tatsuo; Sawa, Kazuhiro
JAEA-Research 2008-007, 30 Pages, 2008/03
JAEA-Research-2008-007.pdf:1.34MB
Neutron irradiation remarkably reduces the thermal conductivity of graphite components in HTGR. The reduced thermal conductivity is expected to be recovered by annealing of irradiation-induced defects, when the graphite components are heated above the irradiation temperature. The annealing effect is not considered in the maximum fuel temperature analysis of the HTTR design from a viewpoint of conservative evaluation for the maximum fuel temperature. Therefore, it is expected that the temperature evaluation at accident conditions could be carried out more accurately with a reasonable stand point by considering the annealing effect. In order to advance the evaluation method for temperature analysis of accident in the HTGR, the annealing effect on thermal conductivity of graphite was evaluated quantitatively and the design curve on the thermal conductivity for graphite components of HTGR was proposed in this study.
Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Goto, Minoru; Takeda, Tetsuaki; Iyoku, Tatsuo
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 6 Pages, 2007/04
The Very High Temperature Reactor (VHTR) system, which is one of generation IV reactors, is the high temperature gas-cooled reactor (HTGR) with capabilities of hydrogen production and high efficiency electricity generation. The High Temperature Engineering Test Reactor (HTTR) is the first HTGR in Japan. The HTTR achieved full power of 30MW at a reactor outlet coolant temperature of about 950C in April, 2004 during the "rise-to-power tests" confirming the reactor performance. The safety demonstration tests by using the HTTR started from 2002 and are under going to demonstrate inherent safety features of HTGRs. The experimental data obtained in these tests are inevitable to design the VHTR with the high cost performance. The analytical models validated through these tests in the HTTR are applicable to precise simulation of an HTGR performance and can contribute to the research and development of the VHTR.
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
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.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08
Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.
C in HTTRIyoku, Tatsuo; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi
UTNL-R-0446, p.14_1 - 14_9, 2005/03
no abstracts in English
C in HTTRFujikawa, Seigo; Hayashi, Hideyuki; Nakazawa, Toshio; Kawasaki, Kozo; Iyoku, Tatsuo; Nakagawa, Shigeaki; Sakaba, Nariaki
Journal of Nuclear Science and Technology, 41(12), p.1245 - 1254, 2004/12
Times Cited Count：89 Percentile：97.75(Nuclear Science & Technology)A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30MW and reactor-outlet coolant temperature of 950C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.103 - 112, 2004/10
Times Cited Count：1 Percentile：10.03(Nuclear Science & Technology)no abstracts in English
Sumita, Junya; Ishihara, Masahiro; Nakagawa, Shigeaki; Kikuchi, Takayuki; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.81 - 88, 2004/10
Times Cited Count：4 Percentile：29.26(Nuclear Science & Technology)A High Temperature Gas-cooled Reactor is particularly attractive due to its capability of producing high temperature helium gas and its possibility to exploit inherent safety characteristic. To achieve high temperature helium-gas, reactor internals are made of graphite and heat resistant materials, its surroundings are composed of metals. The reactor internals of the HTTR consist of graphite and metallic core support structures and shielding blocks. This paper describes the reactor internal design of the HTTR, especially the core support graphite structures, and the program of an in-service inspection.
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：22 Percentile：79.11(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.
Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 17 Pages, 2004/10
no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Elevated Temperature Design and Analysis, Nonlinear Analysis, and Plastic Components, 2004 (PVP-Vol.472), p.39 - 44, 2004/07
The reactor pressure vessel (RPV) of the HTTR is 5.5m in inside diameter, 13.2m in inside height, and 122mm and 160mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1
10
n/cm
(E
1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X-bar. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.
Sakaba, Nariaki; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Saito, Kenji; Furusawa, Takayuki; Tochio, Daisuke; Tachibana, Yukio; Iyoku, Tatsuo
JAERI-Tech 2004-014, 24 Pages, 2004/02
JAERI-Tech-2004-014.pdf:1.06MB
Safety demonstration tests using the HTTR are in progress 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 reactors. This paper describes the reactivity insertion test and coolant flow reduction test by trip of gas circulator and partial flow loss of coolant planned in 2004 with 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.
Nakagawa, Tatsuo*; Mihara, Yorichika*; Komurasaki, Kimiya*; Takahashi, Koji; Sakamoto, Keishi; Imai, Tsuyoshi
Journal of Spacecraft and Rockets, 41(1), p.151 - 153, 2004/02
Times Cited Count：40 Percentile：88.62(Engineering, Aerospace)A launching experiment of a microwave-boosted vehicle model was carried out using the 110GHz, 1MW gyrotoron and a propulsive inpulse to lift up the vehicle was measured. The rf power and pulse was 1MW and 0.175 
0.8msec. The launching mechanism is as follows. Plasma is produced in the nozzle of the vehicle model when the rf beam is injected toward it. The plasma heated by the rf beam can produce a shock wave that gives a propulsive impulse to the vehicle. Maximum momentum coupling coefficient from the impulse to the vehicle is 395N/MW which is comparable to that of a laser boosted vehicle. The rf pulse was 0.175msec. The coupling coefficient is limitted by the gyrtron operation in pulse length and can increase if the pulse length is shorter than 0.175msec.
Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Iyoku, Tatsuo
Nuclear Engineering and Design, 224(2), p.179 - 197, 2003/09
Times Cited Count：13 Percentile：64.66(Nuclear Science & Technology)no abstracts in English
Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/09
no abstracts in English
Sakaba, Nariaki; Nakagawa, Shigeaki; Takada, Eiji*; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Tochio, Daisuke; Iyoku, Tatsuo
JAERI-Tech 2003-074, 37 Pages, 2003/08
JAERI-Tech-2003-074.pdf:1.83MB
Safety demonstration tests using HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. The first phase of the safety demonstration tests includes reactivity insertion tests by means of control-rod withdrawal and coolant flow reduction tests by tripping the gas circulators. In the second phase, accident simulation tests will be conducted. This paper describes the plan of coolant flow reduction tests by tripping of gas circulators planned in August 2003 with detailed test method, procedure and results of pre-test analysis. The analysis results of the steady state and transient behaviours of the reactor and the plant of the HTTR show that in the case of a rapid decrease of the coolant flow rate, the negative reactivity feedback effect of the core brings the reactor power safely to certain stable level without a reactor scram, and that the temperature transient of the reactor core is slow.
