JOPSS - Search Results

Journal Articles

Prediction of critical heat flux for the forced convective boiling based on the mechanism

Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10

The new prediction method of critical heat flux (CHF) of the fuel assemblies based on the mechanism is proposed in this study. The prediction method of CHF based on the mechanism has been needed for a long time to enhance the safety analysis and reduce the design cost. From several experimental findings of the liquid-vapor behavior near the heating surface from the nucleate boiling to the CHF, the authors consider that the macrolayer dryout model will be appropriate to predict the CHF under the reactor condition. The prediction method of the macrolayer thickness and the passage period of vapor mass in the fuel assemblies are needed to predict CHF from the macrolayer dryout model. In this study, the CHF under the forced convection is evaluated by combining the prediction methods for the macrolayer thickness and passage period of vapor mass, which are proposed by authors. The prediction of the CHF under the forced convection is examined and compared with the experimental data.

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, 63(12), p.820 - 824, 2021/12

https://doi.org/10.3327/jaesjb.63.12_820

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

https://doi.org/10.1080/00223131.2019.1588798

Times Cited Count：12 Percentile：72.77(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

https://doi.org/10.1016/0029-5493(93)90142-V

Times Cited Count：1 Percentile：18.58(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

Koizumi, Yasuo; ; Tasaka, Kanji

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

https://doi.org/10.3327/jnst.21.965

Times Cited Count：3 Percentile：58.93(Nuclear Science & Technology)

no abstracts in English

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Prediction of critical heat flux for the forced convective boiling based on the mechanism

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

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

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

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

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

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

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

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

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

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

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

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

Transient burnout under rapid flow reduction condition

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

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

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

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

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

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