Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 21

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Toward mechanistic evaluation of critical heat flux in nuclear reactors, 2; Recent studies and future challenges toward mechanistic and reliable CHF evaluation

Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(12), p.820 - 824, 2021/12

The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.

Journal Articles

An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

Sato, Ikken

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

 Times Cited Count:5 Percentile:75.46(Nuclear Science & Technology)

Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.

Journal Articles

Numerical study on effect of nucleation site density on behavior of bubble coalescence by using CMFD simulation code TPFIT

Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10

The mechanism of Critical Heat Flux (CHF) remains to be clarified, even though it is important to evaluate the CHF for super high heat flux components such as light water reactors (LWRs). Some theoretical models to predict the CHF is proposed so far. A macrolayer formation model which is proposed in order to predict the CHF based on the macrolayer dryout model. In this model, it is assumed that the liquid is captured inside vapor mass at coalescence. In this study, the verification of the assumption of a macrolayer formation model by the numerical simulation of CMFD code, TPFIT, from the view point of hydrodynamics.

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

JAEA Reports

Improvement of COBRA-TF code models for liquid entrainments in film-mist flow

Ezzidi, A.*; Okubo, Tsutomu; Murao, Yoshio

JAERI-M 93-133, 39 Pages, 1993/07

JAERI-M-93-133.pdf:0.99MB

no abstracts in English

JAEA Reports

Assessment of models in COBRA-TF code for liquid entrainments in film-mist flow

Okubo, Tsutomu; Ezzidi, A.*; Murao, Yoshio

JAERI-M 93-069, 115 Pages, 1993/03

JAERI-M-93-069.pdf:2.16MB

no abstracts in English

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions in PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

Nucl. Eng. Des., 144, p.257 - 268, 1993/00

 Times Cited Count:1 Percentile:19.22(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Critical heat flux and heat transfer above mixture level under high-pressure boil-off conditions for PWR type and tight-lattice type fuel bundles

Kumamaru, Hiroshige; Kukita, Yutaka

ANP 92: Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants,Vol. 3, p.24.4-1 - 24.4-7, 1992/00

no abstracts in English

Journal Articles

Post-dryout heat transfer of steam-water two-phase flow in rod bundle under high-pressure and low-flow conditions

Kumamaru, Hiroshige; Kukita, Yutaka

ANS Proc. 1991 National Heat Transfer Conf., Vol. 5, p.22 - 29, 1991/00

no abstracts in English

JAEA Reports

Fundamental experiment on the concentric tube type closed circuit thermosyphon for JRR-3 cold neutron source, II

Kumai, Toshio; *; ; Akutsu, Cho; Takahashi, Hidetake

JAERI-M 89-114, 32 Pages, 1989/09

JAERI-M-89-114.pdf:1.16MB

no abstracts in English

Journal Articles

Critical heat flux for uniformly heated rod bundle under high-pressure, low-flow and mixed inlet conditions

Kumamaru, Hiroshige; Koizumi, Yasuo; Tasaka, Kanji

Journal of Nuclear Science and Technology, 26(5), p.544 - 557, 1989/05

no abstracts in English

Journal Articles

Critical heat flux for rod bundle under high-pressure,low-flow and mixed inlet conditions

Kumamaru, Hiroshige; Koizumi, Yasuo; Tasaka, Kanji

Journal of Nuclear Science and Technology, 25(2), p.207 - 209, 1988/00

no abstracts in English

Journal Articles

Transient burnout under rapid flow reduction condition

Journal of Nuclear Science and Technology, 24(10), p.811 - 820, 1987/10

 Times Cited Count:4 Percentile:45.75(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Core liquid level depression due to manometric effect during PWR small bleak LOCA; Effect of bleak area

Osakabe, Masahiro; Christian Chauliac*; ; Koizumi, Yasuo; ; Tasaka, Kanji

Journal of Nuclear Science and Technology, 24(2), p.103 - 110, 1987/02

 Times Cited Count:9 Percentile:67.55(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Effect of heat generation difference among fuel bundles on core thermal-hydraulics during 200% and 5% loss-of-coolant accident experiments at ROSA-III

Koizumi, Yasuo; ; Tasaka, Kanji

Journal of Nuclear Science and Technology, 24(1), p.61 - 74, 1987/01

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle

; Koizumi, Yasuo; Tasaka, Kanji

Nucl.Eng.Des., 102, p.71 - 84, 1987/00

 Times Cited Count:18 Percentile:83.98(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Post-dryout heat transfer of high-pressure steam-water two-phase flow in single rod channel and multi rod bundle

Koizumi, Yasuo; ; ; Tasaka, Kanji

Nucl.Eng.Des., 99, p.157 - 165, 1987/00

 Times Cited Count:10 Percentile:70.71(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Prediction of dryout heat flux for particle bed simulating degraded core in LWR severe core damage accident

Abe, Yutaka; Sudo, Yukio

Journal of Nuclear Science and Technology, 21(12), p.962 - 964, 1984/00

 Times Cited Count:7 Percentile:82.14(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation of pre- and post-dryout heat transfer in upward steam-water two-phase flow at low flow rate with improved surface temperature measurement

Koizumi, Yasuo; ; Tasaka, Kanji

Journal of Nuclear Science and Technology, 21(12), p.965 - 968, 1984/00

 Times Cited Count:3 Percentile:60.12(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation of pre- and post-dryout heat transfer in upword steam-water two-phase flow at low flow rate

Koizumi, Yasuo; Tasaka, Kanji

Journal of Nuclear Science and Technology, 19(12), p.965 - 984, 1982/00

 Times Cited Count:8 Percentile:66.47(Nuclear Science & Technology)

no abstracts in English

Oral presentation

Macrolayer formation model for prediction of critical heat flux in saturated and subcooled pool boiling

Ono, Ayako; Sakashita, Hiroto*

no journal, , 

A macrolayer formation model for saturated and subcooled pool boiling is proposed, which is developed based on the visual observation of boiling behavior. The distribution of nucleation sites is assumed to distribute based on Poisson distribution. The predicted macrolayer thicknesses are used for the prediction of CHFs in saturated and subcooled pool boiling based on the macrolayer dryout model. The combination of the macrolayer dryout model and macrolayer formation model is able to predict the CHF well at least up to subcooling 40K.

21 (Records 1-20 displayed on this page)