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Journal Articles

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

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 950$$^{circ}$$C, 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.

Journal Articles

Nuclear heat supply fluctuation test by non-nuclear heating using HTTR

Takada, Shoji; Sekita, Kenji; Nemoto, Takahiro; Honda, Yuki; Tochio, Daisuke; Inaba, Yoshitomo; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

To investigate the safety design criteria of heat utilization system for the HTGRs, it is necessary to evaluate the effect of fluctuation of thermal load on the reactor. The nuclear heat supply fluctuation test by non-nuclear heating was carried out to simulate the nuclear heat supply test which is carried out in the nuclear powered operation. The test data is used to verify the numerical code to calculate the temperature of core bottom structure to carry out the safety evaluation of abnormal events in the heat utilization system. In the test, the helium gas temperature was heated up to 120$$^{circ}$$C. A sufficiently high temperature disturbance was imposed on the reactor inlet temperature. It was found that the response of temperatures of metallic components such as side shielding blocks was faster than those of graphite blocks in the core bottom structure, which was significantly affected by the heat capacities of components, the level of imposed disturbance and heat transfer performance.

Journal Articles

Intelligible seminar of fusion reactors, 10; Remote maintenance robot for in-vessel components, Advanced robot technology for handling large-heavy components with high positioning accuracy

Shibanuma, Kiyoshi

Nihon Genshiryoku Gakkai-Shi, 47(11), p.761 - 767, 2005/11

In-vessel components such as blanket and divertor of the fusion reactor are activated by neutron produced during fusion reaction. Gamma radiation will be about 500 MGy/h in maximum after fusion reaction. When the components are failed or troubled in the vessel, the maintenence has to be carried out by the robot because the human cannot be close inside the vessel. The required functions and present R&D status of the typical robots applied to ITER are introduced as examples of robots maintaining the in-vessel components of the fusion reactor.

Journal Articles

Present research status on divertor and plasma facing components for fusion power plants

Suzuki, Satoshi; Ueda, Yoshio*; Tokunaga, Kazutoshi*; Sato, Kazuyoshi; Akiba, Masato

Fusion Science and Technology, 44(1), p.41 - 48, 2003/07

 Times Cited Count:26 Percentile:84.5(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Introduction of ductile crack extension analysis model based on R6 method into PFM code PASCAL

Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*

Proceedings of 4th International Workshop on the Integrity of Nuclear Components, p.31 - 41, 2002/00

no abstracts in English

Journal Articles

Acceptance test of graphite components in nuclear reactor

Ishihara, Masahiro; Hanawa, Satoshi; Iyoku, Tatsuo; Shiozawa, Shusaku

Tanso, 2001(196), p.39 - 48, 2001/02

no abstracts in English

JAEA Reports

Design pressure differences and design velocities for core components of the JRR-3 silicide core

Kaminaga, Masanori; Murayama, Yoji; ;

JAERI-Tech 97-043, 63 Pages, 1997/09

JAERI-Tech-97-043.pdf:1.64MB

no abstracts in English

Journal Articles

Progress of LWR structural safety research at JAERI

Shibata, Katsuyuki

Nucl. Eng. Des., 174(1), p.79 - 90, 1997/00

 Times Cited Count:1 Percentile:14.79(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Water cooled plasma facing componets for ITER

Akiba, Masato

Proceedings of Japan-U.S. Workshop on Fusion High Power Density Devices and Design, 2, p.86 - 96, 1997/00

no abstracts in English

JAEA Reports

Critical element development of double seal door for tritium containment

*; Kakudate, Satoshi; Oka, Kiyoshi; Nakahira, Masataka; *; Obara, Kenjiro; Tada, Eisuke; Shibanuma, Kiyoshi; Seki, Masahiro

JAERI-Tech 94-012, 17 Pages, 1994/08

JAERI-Tech-94-012.pdf:1.14MB

no abstracts in English

Journal Articles

Runaway electron effects

H.-W.Bartels*; Kunugi, Tomoaki; A.J.Russo*

Atomic and Plasma-Material Interaction Data for Fusion, Vol. 5, 0, p.225 - 244, 1994/00

no abstracts in English

Journal Articles

10th Topical Meeting on the Technology of Fusion Energy,Boston,7-12,June,1992

*; *; Seki, Yasushi

Purazuma, Kaku Yugo Gakkai-Shi, 68(5), p.511 - 515, 1992/11

no abstracts in English

Journal Articles

The Latest achievements of fusion technology development

Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Akiba, Masato; Okumura, Yoshikazu

Kaku Yugo Kenkyu, 68(3), p.249 - 267, 1992/09

no abstracts in English

JAEA Reports

An Explication of design data of the graphite structural design code for core support components of High Temperature Engineering Test Reactor

Ishihara, Masahiro; Iyoku, Tatsuo; *; ; Shiozawa, Shusaku

JAERI-M 91-154, 39 Pages, 1991/10

JAERI-M-91-154.pdf:0.73MB

no abstracts in English

JAEA Reports

An Explication of design data of the graphite structural design code for core components of High Temperature Engineering Test Reactor

Ishihara, Masahiro; Iyoku, Tatsuo; *; ; Shiozawa, Shusaku

JAERI-M 91-153, 51 Pages, 1991/10

JAERI-M-91-153.pdf:1.0MB

no abstracts in English

JAEA Reports

An Inspection standard of graphite for the High Temperature Engineering Test Reactor

Toyota, Junji; Iyoku, Tatsuo; Ishihara, Masahiro; Takikawa, Noboru; Shiozawa, Shusaku

JAERI-M 91-102, 61 Pages, 1991/07

JAERI-M-91-102.pdf:1.49MB

no abstracts in English

JAEA Reports

An Explication of the graphite structural design code of core components for the High Temperature Engineering Test Reactor

Iyoku, Tatsuo; Ishihara, Masahiro; *; Shiozawa, Shusaku

JAERI-M 91-083, 31 Pages, 1991/05

JAERI-M-91-083.pdf:1.04MB

no abstracts in English

JAEA Reports

An Explication of the graphite structural design code of core support components for the High Temperature Engineering Test Reactor

Iyoku, Tatsuo; Ishihara, Masahiro; *; Shiozawa, Shusaku

JAERI-M 91-070, 32 Pages, 1991/05

JAERI-M-91-070.pdf:0.94MB

no abstracts in English

Journal Articles

Construction of in-core structure test section in HENDEL, (II); Testing devices and their performance test

; ; ; ; ; ; *;

Nihon Genshiryoku Gakkai-Shi, 30(5), p.427 - 433, 1988/05

 Times Cited Count:1 Percentile:19.96(Nuclear Science & Technology)

no abstracts in English

27 (Records 1-20 displayed on this page)