Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2024)

Department of HTTR

JAEA-Review 2025-053, 86 Pages, 2026/02

JAEA-Review-2025-053.pdf:3.06MB

This report summarizes the activities carried out in the fiscal year 2024 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 950C. 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 power generation but for wide-ranging industrial uses such as hydrogen production and so on. The HTTR achieved its reactor outlet coolant temperature of 950C under full thermal power of 30MW on April 19, 2004. And since then, 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 2024, we conducted heat load variation tests simulating heat load fluctuations due to equipment abnormalities at thermal utilization facilities (hydrogen production facilities) planned to be connected to HTTR, as well as radioactive iodine quantitative evaluation tests to assess the amount of radioactive iodine deposited in the pipes, assuming a primary double-pipe high temperature gas duct rupture accident of the HTGR. Additionally, to confirm hydrogen production technology using the high-temperature gas reactor, we applied to Nuclear Regulation Authority for a reactor installation change permit to connect a hydrogen production facility to HTTR.

Development of the three-dimensional CFD analysis model of helium heat exchanger type steam reformer for hydrogen production by nuclear heat

Ishii, Katsunori; Ono, Masato; Noguchi, Hiroki; Shimizu, Atsushi; Nomoto, Yasunobu; Sato, Hiroyuki; Sakaba, Nariaki

Proceedings of World Hydrogen Technologies Convention 2025 (WHTC 2025) (Internet), p.26 - 28, 2025/10

https://www.whtc2025.com/conference-proceedings

Reactor response during thermal load fluctuation test using HTTR

Hasegawa, Toshinari; Nagasumi, Satoru; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Nakagawa, Shigeaki; Shimazaki, Yosuke; Nakajima, Kunihiro; Sakurai, Yosuke

Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 6 Pages, 2025/09

JAEA has planned a hydrogen production test using the High-Temperature Engineering Test Reactor (HTTR) to demonstrate hydrogen production utilizing the heat from a high-temperature gas-cooled reactor (HTGR). To realize the coupling of a hydrogen production facility with an HTGR, one of the key issues is to confirm the effect of thermal load fluctuations in the facility on the reactor. In this study, a thermal load fluctuation test was conducted during HTTR operation to investigate the reactor's response. The test was performed at 90% reactor power, during which the reactor inlet coolant temperature was increased by 11C to simulate a thermal load fluctuation. As a result, the reactor outlet coolant temperature remained almost unchanged, and the heat corresponding to the inlet temperature increase was absorbed by the core graphite blocks. Furthermore, due to the negative reactivity feedback effect associated with the rise in graphite block temperature, the reactor power decreased to 88% and stabilized without any control rod operation. These findings indicate that disturbances in the reactor inlet coolant temperature are mitigated by the heat storage capacity of the core graphite blocks.

Feasibility study on installation of a new vessel cooling system for a high temperature gas-cooled reactor

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.

JENDL-5 benchmarking for advanced test reactor for preparing burnup analysis using isotopic data from HTGR type fuel irradiation tests

Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio

Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11

https://doi.org/10.1115/ICONE31-131748

Survey on research and development status of Japanese small modular reactors in OECD/NEA activities (2022-2023)

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.

High-temperature test for BGaN semiconductor neutron detectors

Okita, Shoichiro; Sakurai, Tatsuhiro*; Ezaki, Iwao*; Takagi, Katsuyuki*; Nakano, Takayuki*; Hino, Masahiro*

KURNS Progress Report 2023, P. 97, 2024/07

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

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

https://doi.org/10.1016/j.nucengdes.2023.112795

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR) (FY2021)

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.

Study of fabrication of SiC-matrixed fuel compact for HTGR

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.