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Journal Articles

Critical heat flux prediction for subcooled flow boiling in annulus

Liu, W.; Podowski, M. Z.*

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2016 Koen Rombunshu (USB Flash Drive), 2 Pages, 2016/10

Prediction of Critical Heat Flux (CHF) is important for nuclear reactor safety. However, the CHF prediction for subcooled flow boiling in complicated geometry such as fuel assembly still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a fuel bundle. We performed the CHF prediction by using liquid sublayer dryout model, combining with ANSYS CFX code to get the single phase velocity distribution inside the annulus. The results show that the CHF in annulus can be predicted in an accuracy of about $$pm$$20%.

Journal Articles

Prediction of heater surface temperature change at subcooled flow boiling DNB

Liu, W.; Podowski, M. Z.*

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2015 Koen Rombunshu (CD-ROM), 2 Pages, 2015/10

This paper gives prediction to the transient heat transfer at Departure of Nucleate Boiling (DNB) point for subcooled flow boiling. The prediction is carried out by solving the heat conduction equations in cylindrical coordinates with convective boundary condition, which changes with the change of the heat transfer mode on the heated surface. DNB is assumed to happen at the complete dryout of liquid sublayer trapped between the heated wall and an elongated vapor clot, during the passing time of the vapor clot. Important parameters including initial thickness of the liquid sublayer, vapor clot length, vapor clot velocity and void fraction etc., are calculated from the Liu - Nariai model. The initial heater surface temperature is derived from the Jens-Lottes correlation. The transient changes of liquid sublayer thickness, surface temperature at DNB are reported. No obvious temperature jumping is observed at DNB. To predict temperate excursion at Critical Heat Flux (CHF), more simulations to the transient boiling and film boiling processes are needed.

Journal Articles

Experimental examination of heat removal limitation of screw cooling tube at high pressure and temperature conditions

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 81(1-7), p.347 - 354, 2006/02

 Times Cited Count:10 Percentile:52.29(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Critical heat flux testing on screw cooling tube made of RAFM-steel F82H for divertor application

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 75-79, p.313 - 318, 2005/11

 Times Cited Count:8 Percentile:47.08(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Planning outline of CHF experiment for small diameter tube in reactor multiple irradiation environment performed in JMTR

Shibamoto, Yasuteru; Yonomoto, Taisuke; Nakamura, Hideo; Nishikizawa, Tomotoshi

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.210 - 214, 2004/11

no abstracts in English

JAEA Reports

CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu; Kureta, Masatoshi; Yoshida, Hiroyuki; Ishikawa, Nobuyuki; Sato, Takashi; Watanabe, Hironori; Okubo, Tsutomu; Iwamura, Takamichi; et al.

JAERI-Research 2002-018, 37 Pages, 2002/10

JAERI-Research-2002-018.pdf:2.62MB

no abstracts in English

Journal Articles

Development of Reduced-Moderation Water Reactor (RMWR) for sustainable energy supply

Iwamura, Takamichi; Okubo, Tsutomu; Kureta, Masatoshi; Nakatsuka, Toru; Takeda, Renzo*; Yamamoto, Kazuhiko*

Proceedings of 13th Pacific Basin Nuclear Conference (PBNC 2002) (CD-ROM), 7 Pages, 2002/10

In order to ensure sustainable energy supply in Japan, the reduced-moderation water reactor (RMWR) has been developed by JAERI since 1998. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio. In order to establish negative void reactivity coefficient, the core should be short and flat to increase neutron leakage from the core. The core designs were accomplished to a large core with 1,356MWe and a small core with 330MWe. For both cores, negative void coefficient and natural circulation cooling of the core were realized. To confirm thermal-hydraulic feasibility, critical heat flux experiments were performed using 7-rod bundles with the gap width of 1mm and 1.3mm. The results indicated that enough cooling was assured for the tight lattice core. Further R&D studies, including large scale thermal-hydraulic experiments, reactor physics experiments, development of high burn-up fuel cladding material and simplified reprocessing technology, are necessary to realize commercial introduction of RMWR by 2020's for the replacement of current generation LWRs.

JAEA Reports

Experimental result of BWR post-CHF tests; Critical heat flux and post-CHF heat transfer coefficient (Contract research)

Iguchi, Tadashi; Iwaki, Chikako*; Anoda, Yoshinari

JAERI-Research 2001-060, 91 Pages, 2002/02

JAERI-Research-2001-060.pdf:6.34MB

no abstracts in English

Journal Articles

Development of ITER divertor plate with annular swirl tube and tungsten rods

Sato, Kazuyoshi; Ezato, Koichiro; Taniguchi, Masaki; Suzuki, Satoshi; Akiba, Masato

Journal of Plasma and Fusion Research SERIES, Vol.5, p.556 - 560, 2002/00

no abstracts in English

JAEA Reports

Data report of BWR post-CHF tests; Transient core thermal-hydraulic program (Contract research)

Iguchi, Tadashi; Ito, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru*; Anoda, Yoshinari

JAERI-Data/Code 2001-013, 502 Pages, 2001/03

JAERI-Data-Code-2001-013.pdf:32.38MB

no abstracts in English

Journal Articles

Experimental study on the critical heat flux (CHF) in a rectangular channel with micro ribs for a solid target and proton beam window design

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

Proceedings of International Workshop on Current Status and Future Directions in Boiling Heat Transfer and Two-Phase Flow, p.135 - 141, 2000/00

no abstracts in English

Journal Articles

Dimensional analysis of critical heat flux in subcooled water flow under one-side heating conditions for fusion application

J.Boscary*; Araki, Masanori; J.Schlosser*; Akiba, Masato; F.Escorbiac*

Fusion Engineering and Design, 43(2), p.147 - 171, 1998/00

 Times Cited Count:28 Percentile:88.9(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

L.Zheng*; Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime

JAERI-Research 97-054, 85 Pages, 1997/08

JAERI-Research-97-054.pdf:2.18MB

no abstracts in English

Journal Articles

Status of transient thermal-hydraulic demonstration test program at JAERI

Iguchi, Tadashi; Onuki, Akira; Iwaki, Chikako*; Kureta, Masatoshi; Akimoto, Hajime

Proc. of 5th Int. Conf. on Nuclear Engineering (ICONE-5), p.1 - 9, 1997/00

no abstracts in English

JAEA Reports

Benchmark calculation of subchannel analysis codes

JAERI-Data/Code 96-004, 109 Pages, 1996/02

JAERI-Data-Code-96-004.pdf:2.3MB

no abstracts in English

JAEA Reports

Critical heat flux for rod bundle under high-pressure boil-off conditions

Guo, Z.*; Kumamaru, Hiroshige; Kukita, Yutaka

JAERI-M 93-238, 20 Pages, 1993/12

JAERI-M-93-238.pdf:0.67MB

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:1 Percentile:19.22(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

JAEA Reports

Evaluation of mechanistic DNB models using HCLWR CHF data

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

JAERI-M 92-033, 66 Pages, 1992/03

JAERI-M-92-033.pdf:1.37MB

no abstracts in English

Journal Articles

Application of subchannel code to DNB analysis of HCLWR

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

Subchannel Analysis in Nuclear Reactors, p.281 - 301, 1992/00

no abstracts in English

35 (Records 1-20 displayed on this page)