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Journal Articles

Advanced reactor experiments for sodium fast reactor fuels (ARES) project; Transient irradiation experiments for metallic and MOX fuels

Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hiroka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

Journal Articles

Application of transient pulse method to permeability measurement for clay

Kato, Masaji*; Nara, Yoshitaka*; Okazaki, Yuki*; Kono, Masanori*; Sato, Toshinori; Sato, Tsutomu*; Takahashi, Manabu*

Zairyo, 67(3), p.318 - 323, 2018/03

To ensure the safe geological disposal of radioactive wastes, it is important to determine the permeability of clays. The transient pulse test is suitable to apply to the low permeability materials, because it takes relatively short term to determine the permeability. Usually we increase the upstream pore pressure in the measurement with the transient pulse test. However, it is impossible to determine the permeability of clay in this procedure because of the increase of pore pressure. Therefore, the transient pulse test has never been applied to the determination of permeability of clays. In this study, we tried to apply the transient pulse test to a clay obtained in Mizunami Underground Research Laboratory to determine the permeability with decreasing the downstream pore pressure. It was clarified that the transient pulse test with decreasing downstream pore pressure is appropriate from the measurements of granite and sandstone. It was shown that the permeability of a clay was determined by the transient pulse test with decreasing the downstream pore pressure, which agreed with the permeability determined from the falling head test. The measurement time of the transient pulse test is much shorter than that of the falling head test. It is concluded that the transient pulse test is appropriate for the determination of the permeability of clays.

Journal Articles

Temperature transient analysis of gas circulator trip test using the HTTR

Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. 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). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.

JAEA Reports

Data on loss of off-site electric power simulation tests of the High Temperature Engineering Test Reactor

Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

JAERI-Data/Code 2002-015, 39 Pages, 2002/07

JAERI-Data-Code-2002-015.pdf:1.53MB

no abstracts in English

JAEA Reports

Outline of operation and control system and analytical investigation of transient behavior of an out-of-pile hydrogen production system for HTTR heat utilization system

Inagaki, Yoshiyuki; Hada, Kazuhiko; Nishihara, Tetsuo; Takeda, Tetsuaki; Hino, Ryutaro; Haga, Katsuhiro

JAERI-Tech 97-050, 125 Pages, 1997/10

JAERI-Tech-97-050.pdf:2.96MB

no abstracts in English

Journal Articles

CHF experiments under steady-state and transient conditions for tight lattice core with non-uniform axial power distribution

Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(5), p.413 - 424, 1993/05

 Times Cited Count:2 Percentile:30.22(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Accident simulation tests for high conversion light water reactor using a high pressure water loop

Iwamura, Takamichi; Watanabe, Hironori; Araya, Fumimasa; Okubo, Tsutomu; Murao, Yoshio

JAERI-M 92-050, 46 Pages, 1992/03

JAERI-M-92-050.pdf:1.24MB

no abstracts in English

Journal Articles

Dimensional stability of low enriched uranium silicide plate-type fuel for research reactors at transient conditions

Yanagisawa, Kazuaki; ; Horiki, Oichiro; Soyama, Kazuhiko; Ichikawa, Hiroki; Kodaira, Tsuneo

Journal of Nuclear Science and Technology, 29(3), p.233 - 243, 1992/03

no abstracts in English

Journal Articles

ROSA-IV program; Major outcomes and future development

Kukita, Yutaka

Genshiryoku Kogyo, 38(5), p.8 - 16, 1992/00

no abstracts in English

Journal Articles

Pressurized water reactor station blackout; Experimental simulation in the ROSA-IV LSTF

Kukita, Yutaka; Anoda, Yoshinari; ; F.Serre*

Power Plant Transients; 1990, p.7 - 14, 1991/00

no abstracts in English

JAEA Reports

Loss of off-site power test of ROSA-III; Run 971 with HPCS failure

Nakamura, Hideo; Tasaka, Kanji; Suzuki, Mitsuhiro; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Murata, Hideo

JAERI-M 89-130, 127 Pages, 1989/09

JAERI-M-89-130.pdf:3.19MB

no abstracts in English

JAEA Reports

JAEA Reports

ROSA-IV Large Scale Test Facility(LSTF)System Description

*; Tasaka, Kanji; ; ; ; Koizumi, Yasuo; ; ; ; ; et al.

JAERI-M 84-237, 300 Pages, 1985/01

JAERI-M-84-237.pdf:7.57MB

no abstracts in English

JAEA Reports

Journal Articles

Results of LOFT program

; ; Koizumi, Yasuo; *; Katsuragi, Satoru

Nihon Genshiryoku Gakkai-Shi, 26(5), p.375 - 383, 1984/00

no abstracts in English

JAEA Reports

Conceptual Design of Large Scale Test Facility(LSTF) of ROSA-IV Program for PWR Small Break LOCA Integral Experiment

Tasaka, Kanji; ; ; *; *; C.P.Fineman*; D.R.Bosley*;

JAERI-M 9849, 67 Pages, 1981/12

JAERI-M-9849.pdf:1.58MB

no abstracts in English

Oral presentation

Development of fuel transient test facility in JMTR

Kimura, Nobuaki; Matsui, Yoshinori; Tobita, Masahiro*; Nakamura, Jinichi; Yamaura, Takayuki

no journal, , 

Power transient tests for the fuels of light water reactors were to be carried out in the Japan Materials Testing Reactor (JMTR) using capsule-type irradiation test facilities, and the integrity of the fuels had been to be investigated by the tests. Prior to the irradiation tests of the fuels, evaluation of thermal-hydraulic behavior in the capsule using ACE-3D code were carried out. And more, the out-pile test facility, which had an electric heater pin instead of a test fuel pin, was designed and fabricated to simulate the capsules used in the power transient tests. Using this facility, necessary tests for the realization of the irradiation tests and the planning of the test methods were carried out. From these analytical and experimental results, it was confirmed that the power ramping tests for 10$$times$$10 BWR fuel are able to achieve linear heat rate of 600W/cm in JMTR.

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