Current status of high temperature gas-cooled reactor development in Japan

Nagatsuka, Kentaro; Noguchi, Hiroki; Nagasumi, Satoru; Nomoto, Yasunobu; Shimizu, Atsushi; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki

Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08

https://doi.org/10.1016/j.nucengdes.2024.113338

HTGR has a potential to contribute to decarbonization of hard-to-abate industries by supplying a large amount of hydrogen and high temperature heat or steam without carbon dioxide emission. JAEA has been conducting R&Ds for HTGR technologies with High Temperature Engineering Test Reactor (HTTR). This paper shows that HTTR's tests including the loss of core cooing test as a joint the OECD/NEA international research project and a HTTR heat application test plan which demonstrate hydrogen production by coupling the HTTR with a hydrogen production test facility. Additionally, aiming for operation start from the latter half of 2030s, the basic design of the HTGR demonstration reactor has been shown. The Japan's HTGR technology capabilities established by the HTTR project will be fully utilized for the construction of HTGR demonstration reactor.

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

Takamatsu, Kuniyoshi; Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Shoji; Yan, X.; Sawa, Kazuhiro; Sakaba, Nariaki; Kunitomi, Kazuhiko

Journal of Nuclear Science and Technology, 51(11-12), p.1427 - 1443, 2014/11

https://doi.org/10.1080/00223131.2014.967324

In a safety demonstration test involving a loss of both reactor reactivity control and core cooling, HTGRs such as the HTTR, which is the only HTGR in Japan, demonstrate that the reactor power would stabilize spontaneously. In the test at an initial power of 30%, when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero, a reactor transient was initiated and examined. The results confirmed that the reactor power would decrease immediately and become effectively zero.

Annual report of Naka Fusion Research Establishment from April 1, 2004 to March 31, 2005

Naka Fusion Research Establishment

JAERI-Review 2005-046, 113 Pages, 2005/09

JAERI-Review-2005-046.pdf:24.35MB

This annual report provides an overview of research and development activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2004 to 31 March, 2005. The activities in the Naka Fusion Research Establishment are highlighted by researches in JT-60 and JFT-2M, theoretical and analytical plasma researches, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR (Contract research)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*

JAERI-Tech 2005-015, 26 Pages, 2005/03

JAERI-Tech-2005-015.pdf:1.77MB

Safety demonstration tests using the HTTR are in progress since 2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.

Safety demonstration tests using high temperature engineering test reactor

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.08.016

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.

Systems engineering for decommissioning the Japan Power Demonstration Reactor

Yanagihara, Satoshi; ; ; Fujiki, Kazuo

1st JSME/ASME Joint Int. Conf. on Nuclear Engineering, p.65 - 70, 1991/00

no abstracts in English

JSME Series in Thermal and Nuclear Power Generation Vol. 5; HTGR, 2; Development of HTGR demonstration reactor based on Japanese HTGR technology

Tachibana, Yukio; Noguchi, Hiroki; Sumita, Junya; Ohashi, Hirofumi; Sato, Hiroyuki; Sakaba, Nariaki

no journal, , 

For the High-Temperature Gas-cooled Reactor (HTGR) demonstration reactor, establishment of an annular core, enlargement of isolation valves, etc., establishment of codes and standards, establishment of technology to connect the hydrogen production system to the HTGR, etc. are required. In FY2022, JAEA started a project to connect hydrogen production system to the HTTR. JAEA will advance safety evaluations to ensure sufficient safety, and confirm the hydrogen production technology using nuclear heat for the first time in the world. JAEA has been requested by the UK and Poland for technical cooperation for the introduction of HTGRs, and has started technical cooperation to make the best use of the technology for the future demonstration reactor in Japan. This report introduces the current status of development for the HTGR demonstration reactor.