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Takamatsu, Kuniyoshi; Funatani, Shumpei*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 11 Pages, 2024/11
Our research objectives are to develop a VCS that utilizes radiative cooling to passively remove decay heat and residual heat from the RPV during expected and unexpected natural phenomena and accidents. To solve the back pressure problem around the inlet and outlet, it is necessary to minimize reliance on fluid actuation, such as water, air, etc., and to avoid using natural circulation or natural convection as much as possible to improve safety against external hazards. In this presentation, we present the structural concept of the proposed VCS integrated with the reactor building and report the results of the cooling performance evaluation based on the results of experimental and analytical studies conducted to date.
Department of HTTR
JAEA-Review 2024-034, 70 Pages, 2024/10
JAEA-Review-2024-034.pdf:3.22MB
This report summarizes the activities carried out in the fiscal year 2022 about the operation and maintenance of the High Temperature Engineering Test Reactor (HTTR), the R&Ds using the HTTR and so on. The HTTR is the first Japanese test reactor of High Temperature Gas-cooled Reactor (HTGR) type with 30MW in thermal power and whose maximum outlet coolant temperature achieved 950
C. HTGRs are regarded as the promising candidates of the Next Generation Nuclear Plants conformed to the future decarbonized society because of the inherent safety characteristics as well as high temperature heat supply capability for not only a power generation but for wide-ranging industrial uses such as a hydrogen production and so on. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2022, we conducted maintenance of the HTTR such as countermeasures of differential pressure rise event for the primary helium gas circulator's filters occurred at an operation in the year 2021.
Takeda, Takeshi; Shibata, Taiju
JAEA-Review 2024-040, 29 Pages, 2024/09
JAEA-Review-2024-040.pdf:1.33MB
An important theme of Japan's 6th strategic energy plan is to indicate the energy policy path towards carbon neutrality by 2050. Policy responses for Japan's nuclear energy research and development (R&D) towards 2030 contain the demonstrations of technologies for small modular reactors (SMRs) through international cooperation by 2030. In light of this energy plan, basic policy initiatives over the next 10 years have been compiled to realize Green Transformation (GX), which simultaneously achieves decarbonization and economic growth. Looking overseas, activities of SMR R&D are active internationally, mainly in the US, Canada, Europe, China, and Russia. These activities are not only by heavy industry manufactures and R&D institutes, but also by venture companies. Under these circumstances, the NEA CSNI has gathered an Expert Group on SMRs (EGSMR) to help estimate the safety effects of SMRs. The EGSMR efforts required the submission of responses to several questionnaires whose main purpose was to collect the latest information on the efforts of SMR deployment and research. The first author of this report responded to this based on information from Hitachi-GE Nuclear Energy, Ltd. and Mitsubishi Heavy Industries, Ltd. as well as JAEA. Most of the responses from Japan to the questionnaires are the information that serves as the basis of CSNI Technical Opinion Paper No. 21 (TOP-21). In this report, the Japan's publicly available responses to the questionnaires arranged and additional information are explained, which complements some of the content of the TOP-21. In this manner, the investigation results of R&D related to SMR in Japan, focusing on the EGSMR activities (2022-2023), are summarized. The target of this report is to provide useful information for future discussions on international cooperation concerning SMR as well as nuclear power field human resources development internationally and domestically.
Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki
Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/08
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/08
Okita, Shoichiro; Sakurai, Tatsuhiro*; Kudo, Ryohei*; Takagi, Katsuyuki*; Nakano, Takayuki*; Hino, Masahiro*
KURNS Progress Report 2023, P. 97, 2024/07
Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Sawahata, Hiroaki; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; et al.
Nuclear Engineering and Design, 417, p.112795_1 - 112795_6, 2024/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Takamatsu, Kuniyoshi
Kakushinteki Reikyaku Gijutsu; Mekanizumu Kara Soshi, Shisutemu Kaihatsu Made, p.179 - 183, 2024/01
The HTGR has excellent safety, and even in the event of an accident where the reactor coolant is lost, the decay heat and residual heat in the core can be dissipated from the outer surface of the RPV, so the fuel temperature never exceeds the limit value, and the core stabilizes. On the other hand, regarding the cooling system that transports the heat emitted from the RPV to the final heat sink, an active cooling system using forced circulation of water by a pump, etc., and a passive cooling system using natural circulation of the atmosphere have been proposed. However, there is a problem that the cooling performance is affected by the operation of dynamic equipment and weather conditions. This paper presents an overview of a new cooling system concept using radiative cooling, which has been proposed to solve the above problem, and introduces the results of analysis and experiments aimed at confirming the feasibility of this concept.
Department of HTTR
JAEA-Review 2023-016, 82 Pages, 2023/09
JAEA-Review-2023-016.pdf:2.31MB
The High Temperature Engineering Test Reactor (HTTR) is the first Japanese High Temperature Gas-cooled Reactor (HTGR) with 30MW in thermal power and 950C of maximum outlet coolant temperature that is constructed by the Japan Atomic Energy Agency located at Oarai-machi, Higashiibaraki-gun, Ibaraki-ken, Japan. The purpose of the HTTR is establishment of basic HTGR technologies, demonstration of HTGR safety characteristics and so on. The HTTR has had a lot of experience of HTGRs' operation and maintenance throughout rated power operations, safety demonstration tests, long-term high temperature operations and demonstration tests relevant to HTGRs' R&Ds. In the fiscal year 2021, as the HTTR completed activities to conform to the New Regulatory Requirements of Nuclear Regulation Authority, The HTTR restarted since the 2011 off the Pacific coast of Tohoku Earthquake and carried out the Loss-of-forced cooling test without Vessel Cooling System (VCS) operational at 9MW (Three gas circulators trip and VCS is stopped.) as the safety demonstration test. This report summarizes the activities carried out in the fiscal year 2021, which were the situation of the New Regulatory Requirements screening of the HTTR, the operation and maintenance of the HTTR, R&Ds relevant to commercial-scale HTGRs, the international cooperation on HTGRs and so on.
Ishii, Katsunori; Morita, Keisuke; Noguchi, Hiroki; Aoki, Takeshi; Mizuta, Naoki; Hasegawa, Takeshi; Nagatsuka, Kentaro; Nomoto, Yasunobu; Shimizu, Atsushi; Iigaki, Kazuhiko; et al.
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2023/09
Kawano, Takahiro*; Mizuta, Naoki; Ueta, Shohei; Tachibana, Yukio; Yoshida, Katsumi*
JAEA-Technology 2023-014, 37 Pages, 2023/08
JAEA-Technology-2023-014.pdf:2.35MB
Fuel compact for High Temperature Gas-cooled Reactor (HTGR) is fabricated by calcinating a matrix consisting of graphite and binder with the coated fuel particle. The SiC-matrixed fuel compact uses a new matrix made of silicon carbide (SiC) replacing the conventional graphite. Applying the SiC-matrixed fuel compact for HTGRs is expected to improve their performance such as power densities. In this study, the sintering conditions for applying SiC as the matrix of fuel compacts for HTGR are selected, and the density and thermal conductivity of the prototype SiC are measured.
Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*
Nuclear Science and Engineering, 197(8), p.2251 - 2257, 2023/08
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)Sawa, Kazuhiro*; Haseda, Masaya*; Aihara, Jun
Nihon Kikai Gakkai Rombunshu (Internet), 89(921), p.22-00314_1 - 22-00314_6, 2023/05
In high temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In the high burnup coated fuel particle, stress due to fission gas pressure and irradiation-induced pyrolytic carbon (PyC) shrinkage is introduced into the coating layers and consequently the stress could cause failure of coating layers under high burnup irradiation condition. A failure model has developed to predict failure fraction of TRISO-coated particle under high burnup irradiation. In the model, failure probability is strongly dependent on the irradiation characteristics of PyC. This paper describes the outline of the failure model and evaluation result of high burnup fuel irradiation experiment by the model.
Nomoto, Yasunobu; Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; et al.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
Ishii, Katsunori; Aoki, Takeshi; Isaka, Kazuyoshi; Noguchi, Hiroki; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
Okita, Shoichiro; Mizuta, Naoki; Takamatsu, Kuniyoshi; Goto, Minoru; Yoshida, Katsumi*; Nishimura, Yosuke*; Okamoto, Koji*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Banno, Masaki*; Funatani, Shumpei*; Takamatsu, Kuniyoshi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
A fundamental study on the safety of a passive cooling system for the RPV with radiative cooling is conducted. The object of this study is to demonstrate that passive RPV cooling system with radiative cooling is extremely safe and reliable even in the event of natural disasters. Therefore, an experimental apparatus, which is about 1/20 scale of the actual cooling system, was fabricated with several stainless steel containers. The surface of the heating element in the experimental apparatus simulates the surface of the RPV, and the heating element generates natural convection and radiation. A comparison of the Grashof number between the actual cooling system and the experimental apparatus confirmed that both were turbulent, and the experimental results as a scale model are valuable. Moreover, the experimental results confirmed that the heat generated from the surface of the RPV during the rated operation can be removed.
Takamatsu, Kuniyoshi; Funatani, Shumpei*
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 17 Pages, 2023/04
The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.
Simanullang, I. L.*; Nakagawa, Naoki*; Ho, H. Q.; Nagasumi, Satoru; Ishitsuka, Etsuo; Iigaki, Kazuhiko; Fujimoto, Nozomu*
Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)