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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.
H monitoring in radioactive gaseous wasteFurutani, Misa; Kometani, Tatsunari; Nakagawa, Masahiro; Ueno, Yumi; Sato, Junya; Iwai, Yasunori*
Hoken Butsuri (Internet), 55(2), p.97 - 101, 2020/06
Herein, an oxidation catalyst was introduced after heating it to 600C to oxidize tritium gas (HT) existing in exhaust into tritiated water vapor (HTO). This study aims to establish a safer H monitoring system by lowering the heating temperature required for the catalyst. In these experiments, which were conducted in the Nuclear Science Research Institute, Japan Atomic Energy Agency, cupric oxide, hydrophobic palladium/silicon dioxide (Pd/SiO
), and platinum/aluminum oxide (Pt/AlO
) catalysts were ventilated using standard hydrogen gas. After comparing the oxidation efficiency of each catalyst at different temperatures, we found that the hydrophobic Pd/SiO and Pt/AlO catalysts could oxidize HT into HTO at 25C.
Kawasaki, Masatsugu; Nakajima, Junya; Yoshida, Keisuke; Kato, Saori; Nishino, Sho; Nozaki, Teo; Nakagawa, Masahiro; Tsunoda, Junichi; Sugaya, Yuki; Hasegawa, Rie; et al.
JAEA-Data/Code 2017-004, 57 Pages, 2017/03
JAEA-Data-Code-2017-004.pdf:2.34MB
In emergency situation of nuclear facilities, we need to estimate the radiation dose due to radiation and radioactivity to grasp the influence range of the accident in the early stage. Therefore, we prepare the case studies of dose assessment for public exposure dose and personal exposure dose and contribute them to emergency procedures. This document covers about accidents of nuclear facilities in Nuclear Science Research Institute and past accident of nuclear power plant, and it can be used for inheritance of techniques of emergency dose assessment.
Meguro, Yoshihiro; Nakagawa, Akinori; Kato, Jun; Sato, Junya; Nakazawa, Osamu; Ashida, Takashi
Proceedings of International Conference on the Safety of Radioactive Waste Management (Internet), p.139_1 - 139_4, 2016/11
A variety of radioactive wastes have been generated in decommissioning of Fukushima Daiichi Nuclear Power Station. It is necessary to evaluate feasibility of conditioning methods to these wastes, because the majority of such wastes have not been solidified in Japan. The authors investigated an approach for screening of conditioning methods for the Fukushima wastes on the basis of the findings of the existing methods and results of fundamental solidification tests using synthetic Fukushima wastes. Here five solidification methods were selected, and also 13 wastes with different chemical composition are solidified, and characteristics of the solidified form are studied. A screening flow was proposed, and evaluation criteria on each step in the flow was set up. In this presentation a trial result was opened for a waste and improvements of the screening flow found in the trial evaluation was described.
C monitoring in gaseous radioactive wasteUeno, Yumi; Nakagawa, Masahiro; Sato, Junya; Iwai, Yasunori
Hoken Butsuri, 51(1), p.7 - 11, 2016/03
In the Nuclear Science Research Institute, Japan Atomic Energy Agency (JAEA), in order to oxidize C, which exists in various chemical forms in exhaust, into CO, a copper oxide (CuO) catalyst is introduced after heating to 600C. Our goal was to establish a safer C monitoring system by lowering the heating temperature required for the catalyst; therefore, we developed a new hydrophobic palladium/silicon dioxide (Pd/SiO) catalyst that makes the carrier's surface hydrophobic. In these experiments, catalysts CuO, platinum/aluminum oxide (Pt/AlO), palladium/zirconium dioxide (Pd/ZrO), hydrophobic Pd/SiO, and hydrophilic Pd/SiO were ventilated with standard methane gas, and we compared the oxidation efficiency of each catalyst at different temperatures. As a result, we determined that the hydrophobic Pd/SiO catalyst had the best oxidation efficiency. By substituting the currently used CuO catalyst with the hydrophobic Pd/SiO catalyst, we will be able to lower the working temperature from 600C to 300C and improve the safety of the monitoring process.
Yamada, Katsunori; Fujii, Katsutoshi; Kanda, Hiroshi; Higashi, Daisuke; Kobayashi, Toshiaki; Nakagawa, Masahiro; Fukami, Tomoyo; Yoshida, Keisuke; Ueno, Yumi; Nakajima, Junya; et al.
JAEA-Review 2013-033, 51 Pages, 2013/12
JAEA-Review-2013-033.pdf:2.73MB
After the accident at Fukushima Dai-ichi Nuclear Power Plant, various numerical criteria relevant to radiation protection were defined. We surveyed these criteria through internet. As a result of survey, the following 13 items were identified: (1) criteria for taking stable iodine tablets, (2) criteria for the screening of surface contamination, (3) evacuation area, sheltering area, etc., (4) activity concentrations in food, drinking water, etc., (5) dose limit for radiation workers engaged in emergency work, (6) guideline levels of radioactive substances in bathing areas, (7) criteria for use of school buildings and schoolyards, (8) restriction on planting rice, (9) acceptable activity concentrations in feedstuff, (10) acceptable activity concentrations in compost, (11) criteria for export containers and ships, (12) criteria for contaminated waste, (13) standards for radiation workers engaged in decontamination work. In this report, the basis of and issues on these criteria are summarized.
Takamatsu, Kuniyoshi; Ueta, Shohei; Sumita, Junya; Goto, Minoru; Hamamoto, Shimpei; Tochio, Daisuke; Nakagawa, Shigeaki
JAEA-Research 2010-038, 59 Pages, 2010/11
JAEA-Research-2010-038.pdf:4.6MB
Research and development and hydrogen production technologies by HTGRs will establish future hydrogen energy system. Additionally, the R&D will contribute to innovative hydrogen production technologies by nuclear powers as one of nuclear heat utilizations. We are making an effort to propose a prototype of reactor hydrogen production system until about 2020. Therefore, JAEA is promoting the R&D for confirming the technical basis of HTGRs with the HTTR in the first midterm plan. In 2007, we accomplished the rated power operation for continuous 30 days of the HTTR. In the high temperature operation for continuous 50 days, JAEA evaluated the experimental data such as core burn-up, helium purity control, performance of high temperature equipments, structural integrity in the core, etc. and demonstrated the nuclear thermal availability of heat source for thermo-chemical hydrogen production technology.
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：52.30(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.
Katayama, Hisato*; Kitamura, Harushige*; Mori, Mari*; Nakagawa, Junya*; Yoshida, Takahiro*; Kawai, Toshihiko*; Hase, Yoshihiro; Tanaka, Atsushi
JAEA-Review 2006-042, JAEA Takasaki Annual Report 2005, P. 94, 2007/02
no abstracts in English
Tochio, Daisuke; Sumita, Junya; Takada, Eiji*; Fujimoto, Nozomu; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.57 - 67, 2006/03
High Temperature Engineering Test Reactor(HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Agency(JAEA) achieved the reactor outlet coolant temperature of 950C for the first time in the world at Apr. 19, 2004. To ensure the thermal integrity of fuel in high temperature test operation, it is necessary that fuel temperature is designed appropriately by fuel temperature designing method, and that estimated maximum fuel temperature is lower than the thermal limit temperature. In this report, by constructing newly a realistic core-shape representing model, the current fuel temperature estimation model is improved. Moreover fuel temperature in high-temperature test operation is estimated with the newly-constructed model, and it is confirmed that estimated maximum fuel temperature in high temperature test operation is lower than the thermal limit temperature.
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.
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：28.88(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.
Shimazaki, Junya; Suzuki, Katsuo; Nabeshima, Kunihiko; ; Nakagawa, Shigeaki; Shinohara, Yoshikuni
JAERI-M 89-223, 55 Pages, 1990/01
no abstracts in English
He in radioactive gaseous wasteKometani, Tatsunari; Furutani, Misa; Nakagawa, Masahiro; Ueno, Yumi; Sato, Junya
no journal, ,
no abstracts in English
Kitamura, Harushige*; Katayama, Hisato*; Mori, Mari*; Nakagawa, Junya*; Yoshida, Takahiro*; Kawai, Toshihiko*; Hase, Yoshihiro; Tanaka, Atsushi
no journal, ,
no abstracts in English
Hino, Kosaku*; Katayama, Hisato*; Kitamura, Harushige*; Kawamura, Yoko*; Nakagawa, Junya*; Yoshida, Takahiro*; Mori, Mari*; Semba, Toshio*; Hase, Yoshihiro; Tanaka, Atsushi
no journal, ,
no abstracts in English
Sato, Junya; Suzuki, Shinji; Nakagawa, Akinori; Kato, Jun; Sakakibara, Tetsuro; Meguro, Yoshihiro; Nakazawa, Osamu
no journal, ,
no abstracts in English
irradiation for inorganic solidified samples containing simulated ALPS wastesSato, Junya; Suzuki, Shinji; Nakagawa, Akinori; Kato, Jun; Sakakibara, Tetsuro; Meguro, Yoshihiro; Nakazawa, Osamu
no journal, ,
no abstracts in English
