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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.
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：0 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).
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
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：1 Percentile：59.09(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.
Tanaka, Nobuyuki; Sawada, Shinichi*; Koshikawa, Hiroshi*; Yamaki, Tetsuya*
Material Stage, 25(6), p.76 - 80, 2025/09
A thermochemical water-splitting iodine-sulfur process enables us to provide the Carbon-free hydrogen (H
) at high-efficiency levels, and it uses high-temperature heat sources, including high-temperature gas-cooled reactors, solar heat, and more. The cation exchange membranes (CEMs) for the HI mediated electro-electrodialysis (EED) were developed using a radiation grafted polymerization method in order to improve the process efficiency of the IS process. High proton (H
) conductivity and selectivity are required for the performance of CEMs to reduce the consumption energy for EED. The H conductivity of the radiation grafted CEMs were successfully improved by controlling the grafting amount, comparing with that of Nafion. Moreover, the H selectivity and water transport of the developed CEMs was improved by introducing the crosslinker. Currently, the further improvement of the membrane performance is underway by using the ion-track grafting technic.
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.
Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Terao, Tsuyoshi*; Koike, Akifumi*
Progress in Nuclear Science and Technology (Internet), 8, p.116 - 120, 2025/09
Japan Atomic Energy Agency, ANSeeN, and Shizuoka University has been conducted a joint-research to develop nuclear instrument for High Temperature Gas-cooled Reactor (HTGR) core power distribution. In the project, we developed the ex-core detector system to avoid heigh temperature environment of the HTGR core. The system achieves the measurement by taking advantage of the HTGR core feature of long flight length neutron due to the graphite moderated core with Computed Tomography (CT) technologies. The theory is demonstrated by calculation in HTTR (High Temperature Engineering Test Reactor) and criticality experiment in KUCA (Kyoto University Criticality Assembly). These technologies are expected to be applied to other reactors.
Hasegawa, Toshinari; Nagasumi, Satoru; Ishitsuka, Etsuo; Egashira, Keiichiro*; Furuya, Aoi*; Ando, Ryota*; Sakaguchi, Akira*; Sakurai, Yosuke; Nakano, Yumi*; Iigaki, Kazuhiko
JAEA-Technology 2025-004, 20 Pages, 2025/07
JAEA-Technology-2025-004.pdf:1.67MB
Four people from three universities participated in the 2024 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the analysis of the HTTR core, the analysis of 
Cs deposition behavior in the primary cooling system, and the feasibility study of nuclear rockets using HTGR. In the questionnaire after this training, there were comments from participants that it was beneficial as a work experience and that it was meaningful because of many opportunities to communicate with staff. These impressions suggest that this training was generally evaluated as good.
Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.
JAEA-Research 2025-005, 23 Pages, 2025/07
JAEA-Research-2025-005.pdf:2.68MB
A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.
production IS processKubo, Shinji
Shokubai, 67(2), p.71 - 77, 2025/04
no abstracts in English
Sato, Hiroyuki; Yan, X.
Progress in Nuclear Science and Technology (Internet), 7, p.293 - 298, 2025/03
Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Ono, Masato; Nomura, Mikihiro*; Takegami, Hiroaki
Progress in Nuclear Science and Technology (Internet), 7, p.235 - 242, 2025/03
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO (PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
Kubo, Shinji
Kinzoku, 95(1), p.25 - 33, 2025/01
no abstracts in English
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 Technology Convention 2025 (WHTC 2025) (Internet), p.50 - 52, 2025/00
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.
Sugimoto, Chihiro; Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Takegami, Hiroaki; Kubo, Shinji
International Journal of Hydrogen Energy, 95, p.98 - 107, 2024/12
Times Cited Count：0 Percentile：0.00(Chemistry, Physical)