New market opened up by advanced nuclear reactors (Chapter 3, 4, 5, 7)

Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.

Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10

Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

https://doi.org/10.1016/C2018-0-05383-9

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

An Experimental setup for creating and imaging He excimer cluster tracers in superfluid helium-4 via neutron-He absorption reaction

Sonnenschein, V.*; Tsuji, Yoshiyuki*; Kokuryu, Shoma*; Kubo, Wataru*; Suzuki, So*; Tomita, Hideki*; Kiyanagi, Yoshiaki*; Iguchi, Tetsuo*; Matsushita, Taku*; Wada, Nobuo*; et al.

Review of Scientific Instruments, 91(3), p.033318_1 - 033318_12, 2020/03

https://doi.org/10.1063/1.5130919

Excellent feature of Japanese HTGR technologies

Nishihara, Tetsuo; Yan, X.; Tachibana, Yukio; Shibata, Taiju; Ohashi, Hirofumi; Kubo, Shinji; Inaba, Yoshitomo; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; et al.

JAEA-Technology 2018-004, 182 Pages, 2018/07

JAEA-Technology-2018-004.pdf:18.14MB

Research and development on High Temperature Gas-cooled Reactor (HTGR) in Japan started since late 1960s. Japan Atomic Energy Agency (JAEA) in cooperation with Japanese industries has researched and developed system design, fuel, graphite, metallic material, reactor engineering, high temperature components, high temperature irradiation and post irradiation test of fuel and graphite, high temperature heat application and so on. Construction of the first Japanese HTGR, High Temperature engineering Test Reactor (HTTR), started in 1990. HTTR achieved first criticality in 1998. After that, various test operations have been carried out to establish the Japanese HTGR technologies and to verify the inherent safety features of HTGR. This report presents several system design of HTGR, the world-highest-level Japanese HTGR technologies, JAEA's knowledge obtained from construction, operation and management of HTTR and heat application technologies for HTGR.

Examination on small-sized cogeneration HTGR for developing countries

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, X.; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; et al.

JAEA-Technology 2008-019, 57 Pages, 2008/03

JAEA-Technology-2008-019.pdf:8.59MB

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors.

Conceptual design of the HTTR-IS hydrogen production system; Dynamic simulation code development for advanced process heat exchanger in the HTTR-IS system

Sato, Hiroyuki; Kubo, Shinji; Sakaba, Nariaki; Ohashi, Hirofumi; Sano, Naoki; Nishihara, Tetsuo; Kunitomi, Kazuhiko

Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.812 - 819, 2007/09

The objective of this study is to confirm the availability of proposed mitigation methodology against thermal load increase events initiated by the thermochemical water splitting IS process hydrogen production system coupling with the HTTR. JAEA has been performing the development of dynamic simulation code which can evaluate complex phenomena in the HTTR-IS system all at one once to achieve the requirement. The notable feature of the developed code is the APHX module which enables to estimate the IS process thermal load variation considering phase change and chemical reaction behavior assumed in the APHX. In this paper, two cases of dynamic calculation for the thermal load increase events were performed using the newly developed APHX module. The results of the analytical studies clearly show the availability of the developed model for dynamic simulation of the HTTR-IS system and the thermal load increase mitigation methodology.

Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

Goto, Yoshitaka*; Tanabe, Tetsuo*; Ishimoto, Yuki*; Masaki, Kei; Arai, Takashi; Kubo, Hirotaka; Tsuzuki, Kazuhiro*; Miya, Naoyuki

Journal of Nuclear Materials, 357(1-3), p.138 - 146, 2006/10

https://doi.org/10.1016/j.jnucmat.2006.06.005

no abstracts in English

Hydrogen production by thermochemical water-splitting IS process utilizing heat from high-temperature reactor HTTR

Sakaba, Nariaki; Kasahara, Seiji; Ohashi, Hirofumi; Sato, Hiroyuki; Kubo, Shinji; Terada, Atsuhiko; Nishihara, Tetsuo; Onuki, Kaoru; Kunitomi, Kazuhiko

Proceedings of 16th World Hydrogen Energy Conference (WHEC-16) (CD-ROM), 11 Pages, 2006/06

High-temperature reactors (HTRs) are particularly attractive due to their wide industrial application from electricity generation to hydrogen production. The Japan Atomic Energy Agency's (JAEA's) HTTR, which is the first HTR in Japan, attained its maximum reactor-outlet coolant temperature and successfully delivered 950C coolant helium outside its reactor vessel. A hydrogen production system based on the thermochemical water-splitting iodine sulphur (IS) process is planned to be connected to the HTTR in the near future. This will establish the hydrogen production technology with an HTR, including the system integration technology for connection of hydrogen production system to HTRs. It will probably be the world's first demonstration of hydrogen production directly using heat supplied from an HTR. The HTTR-IS system design was launched from a conceptual design in 2005. This paper shows the summary of the HTTR, plan for developing the IS process in JAEA, thermal efficiency evaluation for the HTTR-IS system, etc. The verification of the hydrogen production by the HTTR-IS system by using heat from a nuclear reactor is greatly expected to contribute to the commercialization of nuclear hydrogen in coming hydrogen society.

Study of global wall saturation mechanisms in long-pulse ELMy H-mode discharges on JT-60U

Takenaga, Hidenobu; Nakano, Tomohide; Asakura, Nobuyuki; Kubo, Hirotaka; Konoshima, Shigeru; Shimizu, Katsuhiro; Tsuzuki, Kazuhiro; Masaki, Kei; Tanabe, Tetsuo*; Ide, Shunsuke; et al.

Nuclear Fusion, 46(3), p.S39 - S48, 2006/03

https://doi.org/10.1088/0029-5515/46/3/S06

no abstracts in English

Application of glow discharges for tritium removal from JT-60U vacuum vessel

Nakamura, Hirofumi; Higashijima, Satoru; Isobe, Kanetsugu; Kaminaga, Atsushi; Horikawa, Toyohiko*; Kubo, Hirotaka; Miya, Naoyuki; Nishi, Masataka; Konishi, Satoshi*; Tanabe, Tetsuo*

Fusion Engineering and Design, 70(2), p.163 - 173, 2004/02

https://doi.org/10.1016/j.fusengdes.2003.11.001

In order to establish the effective and conventional in-vessel tritium removal method, glow discharge methods, usually used as wall conditioning, have been applied and examined in vacuum vessel of JT-60U for tritium removal characteristics and kinetics. Release rates of all hydrogen isotopes as well as hydrocarbons from JT-60U vacuum vessel induced by Glow Discharge Cleaning (GDC) with He and H were measured. Release characteristics of hydrogen isotopes were classified into three different release processes each of which is well described by a simple exponential decay with time. It was found that H GDC showed the superior hydrogen isotope release characteristics than the He GDC, probably because chemical processes, such as isotope exchanges assisted by the chemical sputtering process between discharged hydrogen and hydrogen isotopes plasma facing carbon tiles are enhanced by the H glow discharge. Based on the release kinetics observed in the present work, it is estimated that it will take several days to reduce tritium inventory in the surface area of JT-60U to a half by continuous H GDC at 573 K.