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Sr/Y inside the Fukushima Daiichi Nuclear Power Station Unit 3 reactor building using a liquid light guide Cherenkov counterTerasaka, Yuta; Sato, Yuki; Furuta, Yoshihiro*; Kubo, Shin*; Ichiba, Yuta*
Nuclear Instruments and Methods in Physics Research A, 1070(2), p.170021_1 - 170021_9, 2025/01
Times Cited Count：0 Percentile：0.00(Instruments & Instrumentation)Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01
Times Cited Count：1 Percentile：72.25(Nuclear Science & Technology)A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.
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.
Morita, Keisuke; Aoki, Takeshi; Shimizu, Atsushi; Sato, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 6 Pages, 2024/11
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
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.
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transitionsYang, Q.*; Yang, X.*; Wang, Y.*; Fei, Y.*; Li, F.*; Zheng, H.*; Li, K.*; Han, Y.*; Hattori, Takanori; Zhu, P.*; et al.
Nature Communications (Internet), 15, p.7778_1 - 7778_9, 2024/09
Times Cited Count：4 Percentile：72.72(Multidisciplinary Sciences)Luminescent materials that simultaneously embody bright singlet and triplet excitons hold great potential in optoelectronics, signage, and information encryption. However, achieving high-performance white-light emission is severely hampered by their inherent unbalanced contribution of fluorescence and phosphorescence. Herein, we address this challenge by pressure treatment engineering via hydrogen bonding cooperativity effect to realize the mixture of n-- transitions, where the triplet state emission was boosted from 7% to 40% in isophthalic acid (IPA). A superior white-light emission based on hybrid fluorescence and phosphorescence was harvested in pressure-treated IPA, and the photoluminescence quantum yield was increased to 75% from the initial 19% (blue-light emission). In-situ high-pressure IR spectra, X ray diffraction, and neutron diffraction reveal continuous strengthening of the hydrogen bonds with the increase of pressure. Furthermore, this enhanced hydrogen bond is retained down to the ambient conditions after pressure treatment, awarding the targeted IPA efficient intersystem crossing for balanced singlet/triplet excitons population and resulting in efficient white-light emission. This work not only proposes a route for brightening triplet states in organic small molecule, but also regulates the ratio of singlet and triplet excitons to construct high-performance white-light emission.
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：72.25(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.
Ishitsuka, Etsuo; Nagasumi, Satoru; Hasegawa, Toshinari; Kawai, Hiromi*; Wakisaka, Shinji*; Nagase, Sota*; Nakamura, Kento*; Yaguchi, Hiroki*; Ishii, Toshiaki; Nakano, Yumi*; et al.
JAEA-Technology 2024-008, 23 Pages, 2024/07
JAEA-Technology-2024-008.pdf:1.69MB
Five people from three universities participated in the 2023 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the analysis of HTTR core, the analysis of behavior on loss of forced cooling test, the analysis of Iodine deposition behavior in primary cooling system and the feasibility study of energy storage system for HTGRs. In the questionnaire after this training, there were impressions such as that it was useful as a work experience and some students found it useful for their own research. These impressions suggest that this training was generally evaluated as good.
Okita, Shoichiro; Sakurai, Tatsuhiro*; Ezaki, Iwao*; Takagi, Katsuyuki*; Nakano, Takayuki*; Hino, Masahiro*
KURNS Progress Report 2023, P. 97, 2024/07
Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 11(4), p.24-00188_1 - 24-00188_12, 2024/07
Japan Atomic Energy Agency (JAEA) is developing the evaluation method for a two-phase flow in the reactor core using simulation codes based on the Volume Of Fluid (VOF) method. JAEA started developing a Simplified Boiling Model (SBM) for the large-scale two-phase flow in the fuel assemblies. In the SBM, the motion and growth equations of the bubble are solved to obtain their diameter and time length at the detachment, of which size scale is within/around the calculation grid size of the numerical simulation. JUPITER calculates the bubble behavior with a scale of more than several 
m. In this study, the convection boiling on a vertical heating surface is simulated using the developed SBM. The comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.
Li-glass detector to gamma rays by a coincidence methodIto, Fumiaki*; Lee, J.; Hironaka, Kota; Takahashi, Tone; Suzuki, Satoshi*; Mochimaru, Takanori*; Hori, Junichi*; Terada, Kazushi*; Koizumi, Mitsuo
Nuclear Instruments and Methods in Physics Research A, 1064, p.169465_1 - 169465_9, 2024/07
Times Cited Count：0 Percentile：0.00(Instruments & Instrumentation)Nemoto, Takahiro; Fujiwara, Yusuke; Arakawa, Ryoki; Choyama, Yuya; Nagasumi, Satoru; Hasegawa, Toshinari; Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Kawamoto, Taiki; et al.
JAEA-Technology 2024-003, 17 Pages, 2024/06
JAEA-Technology-2024-003.pdf:1.91MB
In order to investigate the cause of the increase in differential pressure in the primary helium circulator filter that occurred during the RS-14 cycle, a clogged filter was investigated. As a result of the investigation, deposits caused by silicone oil were confirmed on the surface of the filter element. These results revealed that the cause of filter clogging was silicone oil mixed into the primary system due to performance deterioration of the charcoal filter in the gas circulator of primary helium purification system. As a measure to prevent the recurrence of this event, in addition to the conventional management based on operating hours for replacing of charcoal filter in the gas circulator of primary helium purification system, we have established a new replacement plan for every three years.
Saijo, Tomoaki; Mizuta, Naoki; Hasegawa, Toshinari; Suganuma, Takuro; Shimazaki, Yosuke; Ishihara, Masahiro; Iigaki, Kazuhiko
JAEA-Technology 2024-002, 96 Pages, 2024/06
JAEA-Technology-2024-002.pdf:22.18MB
Nuclear-grade graphite is used for core components of High Temperature Engineering Test Reactor (HTTR) due to excellent heat resistant properties. The physical properties of this graphite change with temperature and neutron irradiation, as well as exhibit complex behavior such as irradiation deformation and creep deformation. Then, stress analysis code has been developed for the graphite. In previous study, the code has been used to evaluate the shutdown stress by residual strain that accumulates with neutron irradiation. However, the effects of change in physical properties such as Young's modulus and thermal expansion-coefficient on shutdown stress have not been fully understood. Therefore, an evaluation model based on a simplified beam model was developed to clarify the effects of changes in physical properties and complex deformations on stresses occurring during operation and reactor shutdown, and to contribute to the development of graphite structures with longer lifetimes. As an application example, the effects of changes on various physical properties on operational and shutdown stresses were clarified for graphite components in the temperature range from 600 to 800C.
Inaba, Yoshitomo; Sato, Hiroyuki; Sumita, Junya; Ohashi, Hirofumi; Nishihara, Tetsuo; Sakaba, Nariaki
Nihon Kikai Gakkai-Shi, 127(1267), p.25 - 28, 2024/06
Aiming to contribute to net-zero emissions through early social implementation of HTGRs, JAEA promote five projects: HTTR-Heat Application Test, HTGR Domestic Demonstration Reactor, UK HTGR Demonstration Program, UK HTGR Fuel Development Program, and Poland HTGR Research Reactor Basic Design. In addition to these five projects, this article provides an overview of the safety demonstration tests using HTTR.
Noguchi, Hiroki; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki
Kagaku Kogaku, 88(5), p.211 - 214, 2024/05
High temperature gas-cooled reactor (HTGR), one of the next-generation innovative reactors, has an inherent safety and can generate very high-temperature heat which can be used for various heat application including hydrogen production. In Japan, Green Growth Strategy for Carbon Neutrality in 2050 and Basic Policy for the Realization of GX state the promotion of technology development necessary for mass and low-cost carbon-free hydrogen production and development and construction of next-generation innovative reactors including the HTGR for the decarbonization of industrial sectors. Based on these policies, JAEA has been conducted the world's first hydrogen production test using nuclear heat from an HTGR, in addition to verifying the excellent safety features of HTGR, and has also started to study the construction of an HTGR demonstration reactor in cooperation with the industrial community. This paper shows the current status of R&D of HTGR in Japan.
TaEndo, Shunsuke; Kimura, Atsushi; Nakamura, Shoji; Iwamoto, Osamu; Iwamoto, Nobuyuki; Rovira Leveroni, G.; Toh, Yosuke; Segawa, Mariko; Maeda, Makoto
Nuclear Science and Engineering, 198(4), p.786 - 803, 2024/04
Times Cited Count：1 Percentile：31.89(Nuclear Science & Technology)Takamatsu, Kuniyoshi; Funatani, Shumpei*
Nuclear Engineering and Technology, 56(3), p.832 - 845, 2024/03
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.
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)