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Journal Articles

Adefining the mechanism of the gas-bubble AE characteristics by two-phase flow test

Niunoya, Sumio*; Hata, Koji*; Uyama, Masao*; Aoyagi, Kazuhei; Tanai, Kenji

Dai-47-Kai Gamban Rikigaku Ni Kansuru Shimpojiumu Koenshu (Internet), p.92 - 97, 2020/01

Since underground water at the Horonobe Underground Research Laboratory site includes the dissolved gas, it is important to understand the quantitative behavior of AE signal waveform clearly and to develop the criteria of sorting technique. In this report, we tried to perform two types of laboratory tests (Small pipe test and Flat-plate test) in order to obtain detail data of AE signal wave form under two-phase flow. As the result, we could understand that there exists the relationship between the pressure breathing and AE generation, and that the diameter of pipe did not affect the AE behavior.

Journal Articles

Development of ITER divertor vertical target with annular flow concept, 2; Development of brazing technique for CFC/CuCrZr joint and heating test of large-scale mock-up

Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*

Fusion Science and Technology, 46(4), p.530 - 540, 2004/12

 Times Cited Count:14 Percentile:68.23(Nuclear Science & Technology)

The first fabrication and heating test of a large-scale CFC monoblock divertor mock-up using annular flow concept have been performed to demonstrate its manufacturability and thermo-mechanical performance. Prior to the fabrication of the mock-up, brazed joint tests between the CFC monoblock and the CuCrZr tube have been carried out to find the suitable heat treatment mitigating loss of the high mechanical strength of the CuCrZr material. Basic mechanical examination on CuCrZr undergoing the brazing heat treatment and FEM analyses are also performed to support the design of the mock-up. High heat flux tests on the large-scale divertor mock-up have been performed in an ion beam facility. The mock-up has successfully withstood more than 1,000 thermal cycles of $$rm 20 MW/m^2$$ for 15 s and 3,000 cycles more than $$rm 10 MW/m^2$$ for 15 s, which simulates the heat load condition of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed throughout the thermal cycle test.

Journal Articles

Demonstration of inherent safety features of HTGRs using the HTTR

Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 17 Pages, 2004/10

no abstracts in English

JAEA Reports

Structural integrity assessment of helium component during safety demonstration test using HTTR, 1 (Contract Research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio

JAERI-Tech 2004-045, 67 Pages, 2004/04

JAERI-Tech-2004-045.pdf:4.74MB

Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.

Journal Articles

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Sawa, Kazuhiro; Iyoku, Tatsuo

Nuclear Engineering and Design, 224(2), p.179 - 197, 2003/09

 Times Cited Count:13 Percentile:66.42(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Safety demonstration tests using High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Iyoku, Tatsuo

Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/09

no abstracts in English

JAEA Reports

Safety demonstration test plan of the High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Nishihara, Tetsuo; Sawa, Kazuhiro; Iyoku, Tatsuo

JAERI-Tech 2002-059, 42 Pages, 2002/08

JAERI-Tech-2002-059.pdf:1.63MB

no abstracts in English

JAEA Reports

Cause and countermeasure for heat up of HTTR core support plate at power rise tests

Fujimoto, Nozomu; Takada, Eiji*; Nakagawa, Shigeaki; Tachibana, Yukio; Kawasaki, Kozo; Saikusa, Akio; Kojima, Takao; Iyoku, Tatsuo

JAERI-Tech 2001-090, 69 Pages, 2002/01

JAERI-Tech-2001-090.pdf:7.88MB

HTTR has carried out many kinds of tests as power rise tests in which reactor power rises step by step after attained the first criticality. In the tests, temperature of a core support plate showed higher results than expected value at each power level, the temperature was expected to be higher than the maximum working temperature at 100% power level. Therefore, tests under the high temperature test operation mode, in which the core flow rate was different, were carried out to predict the temperature at 100% power precisely, and investigate the cause of the temperature rise. From the investigation, it was clear that the cause was gap flow in a core support structure. Furthermore, it was estimated that the temperature of the core support plate rose locally due to change in gap width between the core support plate and a seal plate due to change in core pressure drop. The maximum working temperature of the core support plate was revised. The integrity of core support plate under the revised maximum working temperature condition was confirmed by stress analyses.

Journal Articles

Air cooling design on stand pipe used for control rod of HTTR

Takeda, Takeshi; Kunitomi, Kazuhiko; Okubo, Minoru

Nihon Genshiryoku Gakkai-Shi, 38(4), p.307 - 314, 1996/00

no abstracts in English

JAEA Reports

Core void fraction distribution under high-temperature high-pressure boil-off conditions; Experimental study with two-phase flow test facility(TPTF)

; Kumamaru, Hiroshige; Murata, Hideo; Anoda, Yoshinari; Kukita, Yutaka

JAERI-M 93-200, 56 Pages, 1993/10

JAERI-M-93-200.pdf:1.85MB

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

BETHSY/LSTF counterpart test on natural circulation in a pressurized water reactor

P.Bazin*; R.Deruaz*; Yonomoto, Taisuke; Kukita, Yutaka

ANS Proc. of the 1992 National Heat Transfer Conf., p.301 - 308, 1992/00

no abstracts in English

Journal Articles

Estimation of hot streak in core bottom structure of high temperature gas-cooled reactor; Thermal mixing test of coolant in core bottom structure of HENDEL

Inagaki, Yoshiyuki; Suzuki, Kunihiro; Ioka, Ikuo*; Kunitomi, Kazuhiko;

Nihon Kikai Gakkai Rombunshu, B, 57(542), p.3520 - 3525, 1991/10

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:20.09(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Journal Articles

Recirculation pump discharge line break test at ROSA-III for a boiling water reactor

; Tasaka, Kanji; ; ; ;

Nuclear Technology, 70, p.189 - 203, 1985/00

 Times Cited Count:2 Percentile:38.26(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Core Hydrodynamic Tests and Determination of Core Flow of JRR-3 to Be Renewed ata 20MWt

Sudo, Yukio; ; *;

JAERI-M 84-119, 108 Pages, 1984/06

JAERI-M-84-119.pdf:2.16MB

no abstracts in English

28 (Records 1-20 displayed on this page)