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Aoki, Takeshi; Hasegawa, Takeshi; Kurahayashi, Kaoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Sakaba, Nariaki
Proceedings of 11th International Topical Meeting on High Temperature Reactor Technology (HTR 2024), 6 Pages, 2024/10
Japan Atomic Energy Agency (JAEA) is planning to perform a test named HTTR heat application test coupling HTTR (High temperature engineering test reactor) and a hydrogen production plant. The present study reports results of the safety design and safety analysis for HTTR heat application test facility. As a safety design, safety classification of structures, systems, and components was defined in the test facility based on their safety functions. As a preliminary safety analysis, a thermal-hydraulic analysis was performed with RELAP5 code. The safety analysis revealed that newly identified events for HTTR heat application test facility except for the rupture of heat transfer tube of steam generator was enveloped by the licensing basis events in conventional HTTR. The preliminary analysis proved that the safety criteria is satisfied in the candidate of licensing basis event.
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/08
-phenylmaleimideYamazaki, Shun*; Kaneko, Naoya*; Kato, Atsuya*; Watanabe, Kohei*; Aoki, Daisuke*; Taniguchi, Tatsuo*; Karatsu, Takashi*; Ueda, Yuki; Motokawa, Ryuhei; Okura, Koki*; et al.
Polymer, 298, p.126846_1 - 126846_11, 2024/04
Times Cited Count：0 Percentile：0.00(Polymer Science)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
Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; Noguchi, Hiroki; et al.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 6 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
Aoki, Takeshi; Shimizu, Atsushi; Noguchi, Hiroki; Kurahayashi, Kaoru; Yasuda, Takanori; Nomoto, Yasunobu; Iigaki, Kazuhiko; Sato, Hiroyuki; Sakaba, Nariaki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The safety design philosophy is developed for the HTTR (High Temperature Engineering Test Reactor) heat application test facility connecting high temperature gas-cooled reactor (HTGR) and the hydrogen production plant. The philosophy was proposed to apply proven conventional chemical plant standards to the hydrogen production facility for ensuring public safety against anticipated disasters caused by high pressure and combustible gases. The present study also proposed the safety design philosophy to meet specific safety requirements identified to the nuclear facilities with coupling to the hydrogen production facility such as measures to ensure a capability of normal operation of the nuclear facility against a fire and/or explosion of leaked combustible material, and fluctuation of amount of heat removal occurred in the hydrogen production plant. The safety design philosophy will be utilized to establish its basic and detailed designs of the HTTR-heat application test facility.
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
Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki
JAEA-Review 2022-016, 193 Pages, 2022/08
JAEA-Review-2022-016.pdf:42.06MB
Aiming to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR), Japan Atomic Energy Agency (JAEA) is planning a HTTR heat application test producing hydrogen with High Temperature Engineering Test Reactor (HTTR) achieved 950
C of the highest reactor outlet coolant temperature in the world. In the HTTR heat application test, it is required to establish its safety design realizing highly safe connection of a HTGR and a hydrogen production plant by the Nuclear Regulation Authority to obtain the permission of changes to reactor installation. However, installation of a system connecting the hydrogen production plant and a nuclear reactor, and its safety design has not been conducted so far in conventional nuclear power plant including HTTR in the world. A special committee on the HTTR heat application test, established under the HTGR Research and Development Center, considered a safety design philosophy for the HTTR heat application test based on an authorized safety design of HTTR in terms of conformity to the New Regulatory Requirements taking into account new considerable events as a result of the plant modification and connection of the hydrogen production plant. This report provides materials of the special committee such as technical reports, comments provided from committee members, response from JAEA for the comments and minutes of the committee.
Aoki, Takeshi; Sato, Hiroyuki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
High temperature gas-cooled reactor (HTGR) has a potential to produce competitive and large amount of carbon-free hydrogen. It is required to establish the control method and system for the HTGR hydrogen production system to maintain its normal operation against the abnormality in the hydrogen production facility through performance evaluations of the control system by transient thermal-hydraulic analysis. In the present study, the reactor response against the disturbance in the reactor inlet coolant temperature was revealed in the HTGR hydrogen production system. The analytical results showed that the reactor outlet coolant control system enabled to control the variation of the reactor outlet coolant temperature was less than 4C against 30C of large disturbance in the reactor inlet coolant temperature and to maintain its normal operation in the HTGR hydrogen production system. Thus, the effectiveness of the control method was confirmed.
Aoki, Takeshi; Shimizu, Atsushi; Iigaki, Kazuhiko; Okita, Shoichiro; Hasegawa, Takeshi; Mizuta, Naoki; Sato, Hiroyuki; Sakaba, Nariaki
JAEA-Technology 2022-011, 60 Pages, 2022/07
JAEA-Technology-2022-011.pdf:2.08MB
Japan Atomic Energy Agency is planning a High Temperature Engineering Test Reactor (HTTR) heat application test producing hydrogen with the HTTR which achieved the highest reactor outlet coolant temperature of 950C in the world to realize a massive, cost-effective and carbon-free hydrogen production technology utilizing a high temperature gas cooled reactor (HTGR). In the HTTR heat application test, it is required to establish its safety design for coupling a hydrogen production plant to HTGR through the licensing by the Nuclear Regulation Authority (NRA). A draft of a safety design philosophy for the HTTR heat application test facility was considered taking into account postulated events due to the plant modification and coupling of the hydrogen production plant based on the HTTR safety design which was authorized through the safety review of the NRA against New Regulatory Requirements. The safety design philosophy was examined to apply proven conventional chemical plant standards to the hydrogen production plant for ensuring public safety against disasters caused by high pressure gases. This report presents a result of a consideration on safety design philosophies regarding the reasonability and condition to apply the High Pressure Gas Safety Act for the hydrogen production plant, safety classifications, seismic design classification, identification of important safety system.
Barucci, M. A.*; Reess, J.-M.*; Bernardi, P.*; Doressoundiram, A.*; Fornasier, S.*; Le Du, M.*; Iwata, Takahiro*; Nakagawa, Hiromu*; Nakamura, Tomoki*; Andr
, Y.*; et al.
Earth, Planets and Space (Internet), 73(1), p.211_1 - 211_28, 2021/12
Times Cited Count：18 Percentile：82.30(Geosciences, Multidisciplinary)The MMX InfraRed Spectrometer (MIRS) is an imaging spectrometer on board of MMX JAXA mission. MIRS is built at LESIA-Paris Observatory in collaboration with four other French laboratories, collaboration and financial support of CNES and close collaboration with JAXA and MELCO. The instrument is designed to fully accomplish MMX's scientific and measurement objectives. MIRS will remotely provide near-infrared spectral maps of Phobos and Deimos containing compositional diagnostic spectral features that will be used to analyze the surface composition and to support the sampling site selection. MIRS will also study Mars atmosphere, in particular to spatial and temporal changes such as clouds, dust and water vapor.
Atkinson, S.*; Aoki, Takeshi; Litskevich, D.*; Merk, B.*; Yan, X.
Progress in Nuclear Energy, 134, p.103689_1 - 103689_10, 2021/04
Times Cited Count：1 Percentile：12.48(Nuclear Science & Technology)This article evaluates the safety features of the designed 10 MWth U-Battery concept with respect to a control rod withdrawal and a depressurised loss of coolant accident. This article provides the evaluation methodology for both transients, using a one-dimensional heat transfer model involving point reactor kinetic model to simulate reactor feedback in the control rod withdrawal. Overall, this work has shown that during the control rod withdrawal the fuel temperature rises by 110 K and at this point the excess reactivity is compensated by the negative temperature coefficient of the fuel. During the depressurised loss of coolant accident, the maximum fuel temperature reached 1455 K after 60 hours. This concludes that during both transients the temperatures maintained well below the maximum fuel operating temperature.
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 950C, 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.
Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the HTTR construction. The guidance consists of a set of FOMs and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the AESJ with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed.
Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.
Aoki, Takeshi; Chirayath, S. S.*; Sagara, Hiroshi*
Annals of Nuclear Energy, 141, p.107325_1 - 107325_7, 2020/06
Times Cited Count：2 Percentile：20.42(Nuclear Science & Technology)The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of 
U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi; Yan, X.
Nuclear Engineering and Design, 360, p.110493_1 - 110493_8, 2020/04
Times Cited Count：11 Percentile：76.86(Nuclear Science & Technology)JAEA has been conducting research and development with a central focus on the utilization of HTTR, the first HTGR in Japan, towards the realization of industrial use of nuclear heat. On the basis of licensing experience through the HTTR construction, JAEA initiated an activity to establish an international safety standard for licensing of commercial HTGR cogeneration systems fully taking into account safety features of HTGRs. We have developed a roadmap towards licensing of commercial HTGR cogeneration systems. A test plan using the HTTR to support the establishment of safety standards and safety analysis methods are also presented. In addition, we confirmed that a vessel cooling system, a passive air-cooled decay heat removal system, satisfies the safety requirement.
Mizuta, Naoki; Aoki, Takeshi; Ueta, Shohei; Ohashi, Hirofumi; Yan, X.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Enhancement of safety and cooling performance of fuel elements are desired for a commercial High Temperature Gas-cooled Reactor (HTGR). Applying sleeveless fuel elements and dual side directly cooling structures with oxidation resistant SiC-matrix fuel compact has a possibility of improving safety and cooling performance at the pin-in-block type HTGR. The irradiated effective thermal conductivity of a fuel compact is an important physical property for core thermal design of the pin-in-block type HTGR. In order to discuss the irradiated effective thermal conductivity of the SiC-matrix fuel compact which could improve the cooling performance of the reactor, the maximum fuel temperature during normal operation of the pin-in-block type HTGR with dual side directly cooling structures are analytically evaluated. From these results, the desired irradiated thermal conductivity of SiC matrix are discussed. In addition, the suitable fabrication method of SiC-matrix fuel compact is examined from viewpoints of the sintering temperature, the purity and the mass productivity.
