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JAEA Reports

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR), FY2003

Department of HTTR Project

JAERI-Review 2005-010, 83 Pages, 2005/03

JAERI-Review-2005-010.pdf:5.18MB

The HTTR (High Teperature Engineering Test Reactor) with the thermal power of 30MW and the reactor outlet coolant temperature of 850/950$$^{circ}$$C is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components and helium gas for primary coolant. December 2001, the thermal power of 30MW and the reactor outlet coolant temperature of 850$$^{circ}$$C was attained. JAERI received the certificate of pre-operation test, that is, the commissioning licence for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R&D on HTGRs for FY2003.

Journal Articles

Validation of the TAC-NC code through the experimental results of safety demonstration test using the HTTR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.369 - 380, 2004/12

This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30% (9MW). By upgrading the code, the analytical result can evaluate accurately the measured value of the transient temperature distribution within 20$$^{circ}$$C. Also, the improved code can analyze the maximum fuel temperature and temperature distributions of the test by tripping all the three gas circulators. The result of this study can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) such as one of the Generation IV reactors.

Journal Articles

NUCEF in the stage of initiation of research

Takeshita, Isao

Enerugi, 28(10), p.63 - 66, 1995/00

no abstracts in English

JAEA Reports

Safety demonstration test plan in the High Temperature Engineering Test Reactor(HTTR) and their safety evaluation

Kunitomi, Kazuhiko; Maruyama, So; Shindo, Masami; Sudo, Yukio

JAERI-M 90-070, 46 Pages, 1990/04

JAERI-M-90-070.pdf:1.29MB

no abstracts in English

Oral presentation

Role of research reactor for development of next generation nuclear reactor; Role of HTTR

Inagaki, Yoshiyuki

no journal, , 

In this presentation, the outline of High Temperature engineering Test Reactor (HTTR) as role of a research reactor for development of a next generation nuclear reactor is described. The presentation includes features of High Temperature Gas-cooled Reactor (HTGR) such as inherent safety, research results with HTTR such as 50-days operation with a reactor outlet temperature of 950$$^{circ}$$C and safety demonstration tests on loss of forced cooling, and a demonstration test plan with HTTR and heat utilization facilities such as a hydrogen production facility.

Oral presentation

Proposition for the future development of the advanced fast reactor cycle (from our experience), 3; Proposition for the high-temperature gas-cooled reactor development

Iyoku, Tatsuo

no journal, , 

Since I got employment in the Japan Atomic Energy Agency in 1979, I have consistently engaged in high-temperature gas-cooled reactor development and experienced from HTTR design to its in-service operation. Based on this experience, I will make recommendations for future high-temperature gas-cooled reactor development at the conference. Typical recommendations are as follows: In future HTGR research and developments, HTTR should be positioned as a COF (center of facility).

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