R&D status of HTGR heat utilization system and thermochemical H production IS process

Kubo, Shinji

Shokubai, 67(2), p.71 - 77, 2025/04

no abstracts in English

Forefront of development of next-generation innovative nuclear reactors (fast reactor and high-temperature gas-cooled reactor), 1; Latest trends of development of next-generation innovative nuclear reactors in Japan and foreign countries

Yamano, Hidemasa; Toyooka, Junichi; Sato, Hiroyuki; Sakaba, Nariaki

Nihon Genshiryoku Gakkai-Shi ATOMO, 66(12), p.607 - 611, 2024/12

https://doi.org/10.3327/jaesjb.66.12_607

This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries.

Development of nuclear instruments to measure power distribution of HTGR, 1; Development of ex-core detector

Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Terao, Tsuyoshi*; Koike, Akifumi*

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

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 for 3 years from 2021 supported by "Nuclear Energy System R&D Project" in MEXT. In the project, there are two R&Ds for "Development of ex-core detector" and "Development of in-core detector". The part of "Development of ex-core detector" is reported in this presentation. The "Development of ex-core detector" is innovative technology by virtue of long flight length neutron of graphite moderated HTGR core and Computed Tomography (CT) technologies. These technologies is expected to be applied to other reactors.

Investigation of deposits on filter element of primary gas circulators in HTTR

Hasegawa, Toshinari; Nagasumi, Satoru; Nemoto, Takahiro; Nakajima, Kunihiro; Yokoyama, Keisuke; Fujiwara, Yusuke; Arakawa, Ryoki; Iigaki, Kazuhiko; Inoi, Hiroyuki; Kawamoto, Taiki

Proceedings of 2024 International Congress on Advanced in Nuclear Power Plants (ICAPP 2024) (Internet), 10 Pages, 2024/06

The filter element of the primary gas circulators (PGC) in High Temperature engineering Test Reactor (HTTR) and its deposits were investigated by Scanning Electron Microscope (SEM) observation and Energy Dispersive X-ray spectroscopy (EDX) analysis to find the cause of the increase of the filter differential pressure during the operation in 2021. SEM observation showed that the clumpy deposits and fibrous deposits smaller than the filtration pore size and the rod-shaped deposits larger than the pore size were present on the filter element. EDX analysis showed that the clumpy deposits and fibrous deposits could include silicone oil in the primary helium purification system (PHPS) gas circulators and that the rod-shaped deposits were thermal insulators inside of the co-axial double pipes in the primary cooling system. It is considered that silicone oil leaked from the PHPS gas circulators due to deterioration in the absorption performance of the activated charcoal filter. Next, it could be vaporized and reach PGC's filter element after passing through the reactor core. Since those deposits including silicone oil were present over the entire surface of the filter element, the filter differential pressure could be increased due to a reduction in the pore size and a rise in its flow resistance. The thermal insulator was unrelated to filter clogging because it was present mainly in the lower part of the filter element. Therefore, silicone oil could increase the filter differential pressure, and the graphite powder, which is the cause of the previous issue was unrelated.

Technology information on High Temperature Gas-cooled Reactor (HTGR)

HTGR Design Group, HTGR Project Management Office

JAEA-Technology 2023-019, 39 Pages, 2024/01

JAEA-Technology-2023-019.pdf:1.34MB

In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.

Fuel cycle scenarios and back-end technologies of HTGR in Japan

Fukaya, Yuji; Goto, Minoru; Shibata, Taiju

IAEA-TECDOC-2040, p.133 - 136, 2023/12

https://www-pub.iaea.org/MTCD/Publications/PDF/TE-2040web.pdf

Japan has developed back-end technologies to establish a multi-recycling fuel cycle with fast breeder reactors (FBRs) to ensure energy resources. Even though the development of FBR has been retreated to one of fundamental research, the reprocessing technologies for uranium fuel and disposal technologies had been completed for Light Water Reactor (LWR) fuel cycle on the process. These technologies were inherited to utilities and are about to be practical. Now, Japan had been completed High Temperature Engineering Test Reactor (HTTR) a prototype and research reactor, a commercial High Temperature Gas-cooled Reactor (HTGR) design Gas Turbine High Temperature Reactor 300 (GTHTR300) with related reprocessing technologies, and is planning domestic demonstration reactor project. In this context, a representative fuel cycle policy is reprocessing in Japan. However, Japan has investigated various fuel cycle scenarios to expand the usage of the commercial HTGR. Then, we would like to introduce the scenarios and development status of related technologies in the present study.

Study on disposal of waste from reprocessing for commercial HTGR spent fuel

Fukaya, Yuji; Maruyama, Takahiro; Goto, Minoru; Ohashi, Hirofumi; Higuchi, Hideaki

JAEA-Research 2023-002, 19 Pages, 2023/06

JAEA-Research-2023-002.pdf:1.48MB

A study on disposal of waste derived from commercial High Temperature Gas-cooled Reactor ("HTGR") has been performed. Because of significant difference between the reprocessing of Light Water Reactor ("LWR") and that of HTGR due to difference in structures of the fuel, adoptability of the laws relating to reprocessing waste disposal, which is enacted for LWR, to HTGR waste should be confirmed. Then, we compared the technologies and waste of reprocessing and evaluated radioactivity concentration in graphite waste by activation and contamination based on whole core burn-up calculation. As a result, it was found that SiC residue waste should be disposed of into a geological repository as 2nd class designated radioactive waste in the Designated Radioactive Waste Final Disposal Act (Act No.117 of 2000), by way of amendment of the applicable order, same as hull and end-piece of LWR, and graphite waste should be shallowly disposed of than geological disposal as 2nd class waste for pit disposal in the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Act No.166 of 1957) same as a channel box of LWR.

Changing society and the potential of nuclear energy

Koito, Yuko

Nihon Genshiryoku Gakkai-Shi ATOMO, 65(5), P. 298, 2023/05

https://doi.org/10.3327/jaesjb.65.5_298

no abstracts in English

Feasibility study on reprocessing of HTGR spent fuel by existing PUREX plant and technology

Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi

Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02

https://doi.org/10.1016/j.anucene.2022.109534

Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.

Study on evaluation method of kernel migration of TRISO fuel for High Temperature Gas-cooled Reactor

Fukaya, Yuji; Okita, Shoichiro; Sasaki, Koei; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi; Yan, X.

Nuclear Engineering and Design, 399, p.112033_1 - 112033_9, 2022/12

https://doi.org/10.1016/j.nucengdes.2022.112033

Kernel migration of TRi-structural ISOtropic (TRISO) fuel for High Temperature Gas-cooled Reactor (HTGR) has been analyzed to investigate the potential dominating effects. Kernel migration is a major fuel failure mode and dominant to determine the lifetime of the fuel for High Temperature engineering Test Reactor (HTTR). However, this study shows that the result and reliability depend on the evaluation method. The evaluation method used in this study takes into account of actual distribution of Coated Fuel Particles (CFPs) and the resulting heterogeneous fuel temperature calculation with such distribution. The result shows that the Kernel Migration Rate (KMR) is predicted to be about 10% less compared with the most conservative evaluation.