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Ishii, Katsunori; Ono, Masato; Noguchi, Hiroki; Shimizu, Atsushi; Nomoto, Yasunobu; Sato, Hiroyuki; Sakaba, Nariaki
International Journal of Hydrogen Energy, 275, p.157448_1 - 157448_9, 2026/10
A helium-heated multi-tube steam reformer is required for large-scale hydrogen production utilizing high temperature gas-cooled reactors (HTGRs). However, scale-up of the reformer causes axial and radial temperature distributions, which affect reaction rates and hydrogen production performance. In this study, a three-dimensional thermal-fluid and reaction analysis model of a single tube of a helium-heated steam reformer was developed by ANSYS Fluent and was validated using the experimental data from a mock-up test. Comparison of the Yagi-Wakao correlation and Kunii-Suzuki correlation showed that the Yagi-Wakao correlation provided better agreement with the measured temperature and hydrogen production rates. Under certain operating conditions, most of the calculated internal temperatures in the steam reformer agreed with measurements within 
5%. The validated model incorporating the Yagi-Wakao correlation can serve as a basis for future scale-up studies of HTGR-based hydrogen production systems.
Takamatsu, Kuniyoshi; Tachibana, Yukio; Kawabuchi, Soichiro*; Aihara, Jun; Fukaya, Yuji; Okita, Shoichiro*; Okamoto, Koji*
Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-15) (Internet), 14 Pages, 2026/09
Our proposed high-power-density HTGR, HTR60S-SiC, employs a core using SiC-based fuel compacts, achieving 60 MWt and an average power density of 4.2 MW/cm
. The HTR60S-SiC demonstrates a 1.2-fold performance improvement compared to the conventional HTR50S reactor, which has 50 MWt and 3.5 MW/cm. Therefore, it is necessary to demonstrate that the HTR60S-SiC can ensure passive safety features. As part of a feasibility study for commercial HTGRs, a depressurization accident analysis was conducted using the TAC-NC code, which was employed in HTGR safety assessments. This analysis derived the temperature distributions in the core of the HTR60S-SiC during a depressurization accident. The evaluation confirmed that the maximum fuel and RPV temperatures could remain and fall below these temperature limits during a depressurization accident. This demonstrated that the HTR60S-SiC can maintain passive safety features even in the event of a depressurization accident.
Liu, X.*; Funatani, Shumpei*; Takamatsu, Kuniyoshi
Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-15) (Internet), 13 Pages, 2026/09
A fundamental study on a new passive radiation cooling system for HTGRs has been conducted. A heat transfer test equipment called 'a scale model' generates natural convection and radiation inside the scale model. The heat flux (approximately 3 kW/m
) needs to be removed from the RPV as heat loss during rated operation for practical HTGRs with a maximum thermal reactor power of 600 MWt. The experimental results demonstrated that the passive radiation cooling system effectively removed heat dissipation from the RPV, confirming its feasibility. Additionally, downsizing the passive radiation cooling system by shortening the overall length did not significantly affect its performance. Shortening the overall length increases its temperature; therefore, the overall length should be adjusted to stay within the structural temperature limits. Moreover, an increase of outside temperature surrounding the passive radiation cooling system did not significantly affect the performance.
Marufuji, Takato; Ueta, Shohei; Kawaguchi, Koichi; Watanabe, Masashi; Yan, X.
JAEA-Technology 2026-011, 22 Pages, 2026/08
JAEA-Technology-2026-011.pdf:1.84MB
In microcapsule production employing a novel double-layer coating technique in which the fuel core is pre-sealed, the required droplet size is larger than that in conventional methods, necessitating a re-examination of the optimal manufacturing conditions. However, conventional equipment involves multiple operating parameters that vary simultaneously, making it difficult to isolate and evaluate the effect of each parameter and thus hindering the identification of optimal conditions. In addition, the economic performance of the equipment is an important issue for practical application. To address these challenges, this study developed a new apparatus that enables independent control and evaluation of each parameter while also reducing cost.
Fukaya, Yuji; Asano, Kazuhito; Sato, Hiroyuki; Ohashi, Hirofumi; Sakaba, Nariaki
JAEA-Review 2026-015, 10 Pages, 2026/06
JAEA-Review-2026-015.pdf:0.96MB
The Japan Atomic Energy Agency (JAEA) is developing SiC matrix fuel as a fuel to improve oxidation resistance. It is a simple manufacturing method called the slurry method, in which raw materials are dissolved in water, molded, and sintered, and are designed to pursue manufacturability in consideration of mass production. On the other hand, there is a patent in the U.S. for SiC matrix fuel, and it was expressed that there is a risk that the JAEA's SiC matrix fuel technology may infringe this patent, and it should be confirmed. In that case of infringement, it could become an obstacle to the social implementation of that technology. Therefore, in order to determine whether the SiC-based fuel for HTGR proposed by the JAEA would constitute a patent infringement against the aforementioned U.S. patent, JAEA requested a formal infringement evaluation from a major U.S. patent law firm. As a result, it was determined that the SiC-based fuel for high-temperature gas reactors proposed by the JAEA does not constitute a patent infringement against the aforementioned U.S. patent. This removed obstacles to the future deployment of the SiC matrix fuel proposed by the JAEA for high temperature gas-cooled reactors.
Fujiwara, Yusuke; Nakajima, Kunihiro; Urabe, Kohei; Nagatsuka, Kentaro; Asano, Kazuhito; Shimizu, Atsushi; Noguchi, Hiroki; Sato, Hiroyuki; Ohashi, Hirofumi; Sumita, Junya; et al.
JAEA-Technology 2026-001, 43 Pages, 2026/05
JAEA-Technology-2026-001.pdf:2.93MB
High Temperature Gas-cooled Reactors (HTGRs) have excellent safety features and can supply high-temperature heat without emitting carbon dioxide, and therefore are expected to stably produce large amounts of hydrogen to contribute carbon neutrality by 2050. The pertinent material for the "Basic Policy for GX Realization" shows the development process for the HTGR demonstrator with the goal of starting operation in the 2030s. Meanwhile, to achieve net-zero, the UK government has started the Advanced Module Reactor (AMR) Research, development and demonstration (RD&D) programme with the aim of starting operations of the HTGR demonstrator in the early 2030s. Against this background, with the aim of early deployment of HTGR technology, Japan Atomic Energy Agency (JAEA), in collaboration with the United Kingdom National Nuclear Laboratory (UKNNL), aims to demonstrate Japanese HTGR technology outside Japan and reflect the development of HTGR demonstrator in the UK to Japan. This document summarizes the results of a market survey of HTGR products and surveys of industrial infrastructure in the UK, the environment and social conditions of Hartlepool, a candidate construction site, with the aim of contributing to the design study of the HTGR demonstrator in the UK (UKJ-HTR).
Tada, Kenichi; Katano, Ryota; Kondo, Ryoichi; Nagaya, Yasunobu; Chiba, Go*; Yamamoto, Akio*
Proceedings of International Conference on the Physics of Reactors (PHYSOR 2026) (Internet), 8 Pages, 2026/04
A High-Temperature Gas-cooled Reactor (HTGR) is a type of Generation IV reactor. HTGRs have a good advantage in preventing the core meltdown in the event of a severe accident, such as loss of forced cooling. We are developing a new core analysis code system, ARCANUM, to accurately calculate HTGR cores within a reasonable calculation time. It consists of two parts, i.e., the 2D lattice calculation part and the 3D triangular prism core calculation part. The MOC-based transport solver GENESIS was used for 2D lattice calculations. The other calculations, such as 3D core calculation code, are performed using newly developed codes. This paper shows an overview of the ARCANUM code system and verification results of this system. Two fuel elements, which are obtained from the IRPhE benchmark, were used to verify the lattice calculation part. Takeda-benchmark Model 4 was used to verify the core calculation part. The calculation results showed good agreement with the reference results.
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 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 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.
Myagmarjav, O.; Noguchi, Hiroki; Tanaka, Nobuyuki; Kamiji, Yu; Sugimoto, Chihiro; Ishii, Katsunori; Nomura, Mikihiro*; Takegami, Hiroaki
International Journal of Hydrogen Energy, 198, p.152495_1 - 152495_9, 2026/01
Times Cited Count：1 Percentile：13.61(Chemistry, Physical)Nagasumi, Satoru; Hasegawa, Toshinari; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Kubo, Shinji; Shimazaki, Yosuke; Nakajima, Kunihiro; Sakurai, Yosuke; Shinohara, Masanori; Saito, Kenji; et al.
Nuclear Engineering and Design, 446(Part A), p.114542_1 - 114542_14, 2026/01
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)To demonstrate HTGR's safety features, a loss-of-forced-cooling (LOFC) test was conducted using the HTTR. In this test, the forced cooling in the reactor core was intentionally lost by shutting down all helium gas circulators (HGCs) without reactor scram. During steady-state operation at 100% reactor power (30 MW), after the LOFC, the reactor power spontaneously decreased. This power reduction occurred due to the negative reactivity feedback effect triggered by an increase in core temperature. The power stabilized at a lower value of 1.2% after re-criticality. Additionally, the measured radioactivity concentration in the primary coolant remained nearly unchanged during this LOFC operation and during an immediately subsequent HTTR operation. This indicates no failure of the coated particle fuel, even after the increase in core temperature associated with the LOFC event. These results provide experimental evidence of the safety features of HTGRs.
Makuuchi, Etsuyo; Aizawa, Kosuke; Imai, Yoshiyuki; Kamiji, Yu; Akasaka, Naoaki; Yan, X.; Sakuma, Wataru*; Tanihira, Masanori*
Nuclear Technology, 11 Pages, 2026/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Saijo, Tomoaki; Shimazaki, Yosuke; Ishihara, Masahiro
JAEA-Technology 2025-010, 126 Pages, 2025/12
JAEA-Technology-2025-010.pdf:12.52MB
During the operation of the High Temperature Engineering Test Reactor (HTTR), thermal stress is generated in the graphite components. In addition, graphite exhibits dimensional shrinkage and creep deformation under neutron irradiation. As a result, residual stress remains in the graphite components during reactor shutdown. Therefore, in the design of the HTTR core graphite structures, stress analyses of the graphite components have previously been performed using the finite element analysis code VIENUS. In the HTTR, the graphite components are exposed to a wide range of temperature, from approximately 400C to 1200C, depending on their location. Consequently, irradiation-induced behaviors such as material property changes and irradiation shrinkage vary among the graphite components. On the other hand, since VIENUS code evaluates stress based on thermal fluid and heat conduction analysis results, it is not suitable for parametric studies. In this study, the influence of irradiation behavior on the stress behavior of graphite components in the wide temperature range (400C to 1200C) was analyzed using simplified viscoelastic evaluation model, consisting of two beam elements, to conduct efficient parametric studies. Operational stress exhibits two distinct patterns depending on whether the irradiation temperature is below or above 800C, due to irradiation shrinkage. Residual stress approaches the thermal stress, preventing excessive increase even when irradiation shrinkage is large. Moreover good agreement in stress behavior trends was observed between the stress analysis results by the simplified viscoelastic evaluation model and VIENUS code. These results indicate that the simplified viscoelastic evaluation model is beneficial in simulating stress behavior.
Hirota, Noriaki
Material Science and Technology of Japan, 62(6), p.192 - 196, 2025/12
In a high-temperature sulfuric acid environment, Alloy600 showed a high corrosion rate of 0.40 mm/year, while Alloy800H and 3Al-Ferrite exhibited much lower rates of 0.03 to 0.01 mm/year. Four-point bending tests revealed the stress to reach 0.2 % yield stress followed the order: 3Al-Ferrite 
Alloy600 Alloy800H. Bending strain at 80 % of 0.2 % yield stress was highest in Alloy800H and lowest in 3Al-Ferrite, indicating its low deformability. Post-corrosion microstructural analysis revealed that Alloy800H and 3Al-Ferrite formed thin oxide films with no cracks, whereas Alloy600, which developed a thick multilayered oxide film composed of Ni, Fe and Cr, exhibited open cracks. Electron Backscattered Diffraction (EBSD) and Grain Reference Orientation Deviation (GROD) maps confirmed intergranular crack growth and residual tensile stress in Alloy600. These findings indicate that Alloy600, primarily composed of Ni and Cr, formed a spallation and degradation oxide film and subsequently generated oxides that weakened the grain boundaries, thereby promoting the occurrence of Stress Accelerated Grain Boundary Oxidation (SAGBO).
Toyooka, Junichi; Kubo, Shigenobu; Nishimura, Masahiro; Kamide, Hideki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 67(11), p.644 - 648, 2025/11
Global nuclear power development is accelerating against the backdrop of achieving decarbonization goals and ensuring energy security. This paper outlines international development trends for innovative reactors, including innovative light water reactors, fast reactors as Generation IV reactors, and high-temperature gas-cooled reactors, with a focus on small modular reactors (SMRs), regulatory standards, and new manufacturing technologies.
Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki
Nihon Genshiryoku Gakkai-Shi ATOMO, 67(10), p.593 - 597, 2025/10
no abstracts in English
Kamiji, Yu; Tanaka, Nobuyuki; Sugimoto, Chihiro; Myagmarjav, O.; Takegami, Hiroaki
Proceedings of World Hydrogen Technologies Convention 2025 (WHTC 2025) (Internet), p.88 - 89, 2025/10
no abstracts in English
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
demand and HTGR development potential in the industrial complex in JapanNoguchi, Hiroki; Ishii, Katsunori; Ono, Masato; Kasahara, Seiji; Sato, Hiroyuki; Sakaba, Nariaki
Proceedings of World Hydrogen Technologies Convention 2025 (WHTC 2025) (Internet), p.50 - 52, 2025/10
Achieving carbon neutrality in Japan in 2050, hydrogen is expected to be used as an alternative to fossil fuels in the hard-to-abate sectors. In steelmaking, hydrogen-based reduction process has been developed as a substitute for the conventional blast furnace steelmaking process, which involves the reduction of iron ore by coke. In chemical industry, a novel olefin production process has been developed using hydrogen and CO, through methanol as an intermediate chemical. A large amount of hydrogen is required for these novel processes. Nuclear energy is well-suited to large-scale low-carbon hydrogen production. High temperature gas cooled reactor (HTGR) is a type of nuclear reactor featuring extraction of high temperature heat. The heat can be applicable to hydrogen production. This study predicts hydrogen demand in five industrial complexes in Japan in 2050 and estimates the potential for introducing HTGR to meet the demand. The introduction of HTGR could be a promising solution for decarbonizing industrial complexes due to their large-scale hydrogen supply capacity.
Department of HTTR
JAEA-Review 2025-032, 75 Pages, 2025/09
JAEA-Review-2025-032.pdf:3.15MB
This report summarizes the activities carried out in the fiscal year 2023 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 30 MW 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 power generation but for wide-ranging industrial uses such as 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 2023, the HTTR was confirmed its inherent safety of HTGR due to carry out the safety demonstration test (Loss of forced cooling test at the 100% power) as the international joint research of Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA).
Aoki, Takeshi; Shimizu, Atsushi; Ishii, Katsunori; Morita, Keisuke; Mizuta, Naoki; Kurahayashi, Kaoru; Yasuda, Takanori; Noguchi, Hiroki; Nomoto, Yasunobu; Iigaki, Kazuhiko; et al.
Annals of Nuclear Energy, 220, p.111503_1 - 111503_7, 2025/09
Times Cited Count：2 Percentile：61.39(Nuclear Science & Technology)Aiming to establish coupling technologies between a high temperature gas cooled reactor and a hydrogen production plant, JAEA has initiated the HTTR Heat Application Test Project and is conducting the safety design and the safety analysis for the licensing of the HTTR Heat Application Test Facility. The present study proposed a relative evaluation methodology for the demarcation of applicable laws and design standards for the nuclear hydrogen production system and applied it to the HTTR Heat Application Test Facility. The evaluation results showed that a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior demarcation option for the HTTR Heat Application Test Facility.
