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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

Lattice Boltzmann modeling and simulation of forced-convection boiling on a cylinder

Saito, Shimpei*; De Rosis, A.*; Fei, L.*; Luo, K. H.*; Ebihara, Kenichi; Kaneko, Akiko*; Abe, Yutaka*

Physics of Fluids, 33(2), p.023307_1 - 023307_21, 2021/02

 Times Cited Count:2 Percentile:90.05(Mechanics)

A Boiling phenomenon in a liquid flow field is known as forced-convection boiling. We numerically investigated the boiling system on a cylinder in a flow at a saturated condition. To deal with such a phenomenon, we developed a numerical scheme based on the pseudopotential lattice Boltzmann method. The collision was performed in the space of central moments (CMs) to enhance stability for high Reynolds numbers. Furthermore, additional terms for thermodynamic consistency were derived in a CMs framework. The effectiveness of the model was tested against some boiling processes, including nucleation, growth, and departure of a vapor bubble for high Reynolds numbers. Our model can reproduce all the boiling regimes without the artificial initial vapor phase. We found that the Nukiyama curve appears even though the focused system is the forced-convection system. Also, our simulations support experimental observations of intermittent direct solid-liquid contact even in the film-boiling regime.

Journal Articles

Sea water flow boiling heat transfer involving sea salt deposition; Role of deposited sea salt

Koizumi, Yasuo*; Uesawa, Shinichiro; Ono, Ayako; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2019 Koen Rombunshu (USB Flash Drive), 1 Pages, 2019/10

no abstracts in English

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

Ultrahigh CHF prediction for subcooled flow boiling based on homogenous nucleation mechanism

Liu, W.; Nariai, Hideki*

Journal of Heat Transfer, 127(2), p.149 - 158, 2005/02

 Times Cited Count:9 Percentile:41.85(Thermodynamics)

Homogeneous nucleation, although being discounted as a mechanism for vapor formation for water in most conditions, is found being possible to occur under some extreme conditions in subcooled flow boiling. In this paper, firstly, the existence of the homogeneous nucleation governed condition is indicated. Followed, a criterion is developed to judge a given working condition is the conventional one or the homogeneous nucleation governed one. With the criterion, subcooled flow boiling data are categorized and typical homogeneous nucleation governed datasets are listed. CHF triggering mechanism for the homogeneous nucleation governed condition is proposed and verified. Parametric trends of the CHF, in terms of mass flux, pressure, inlet subcooling, channel diameter and the ratio of heated length to diameter are also studied.

Journal Articles

Master plan and current status for feasibility study on thermal-hydraulic performance of reduced-moderation water reactors

Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime

Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12

We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.

Journal Articles

Experimental study on cooling limit under flow instability in boiling flow channel

Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04

Authors investigated the cooling limit under flow instability, by conducting THYNC experiments using a 2$$times$$2 bundle test section of electrical rod heaters、whose heated lengths and diameters were 3.71m and 12.3mm. The experimental result indicated periodic rise and rapid drop of the rod temperature under flow oscillation, indicating periodic film boiling. When the heating power increased further, the rod temperature indicated continuous film boiling. The power at the onset of continuous film boiling (cooling limit) under flow oscillation was about 50%-80% of the cooling limit under steady flow condition in THYNC. The ratio of both cooling limits almost agreed with the Umekawa model prediction in cases of P$$<$$2MPa and G$$<$$400kg/m2s. For high pressure and high mass flux conditions, the ratio almost agreed with the empirical model based on the heat balance during one cycle of flow oscillation. TRAC-BF1 code simulated periodic film boiling qualitatively, but the cooling limit under the flow oscillation was not predicted well probably due to inaccurate rewetting prediction.

Journal Articles

Study of critical heat flux mechanism in flow boiling using bubble crowding model; Application to CHF in short tube and in tube with twisted tape under non-uniform heating conditions

Kinoshita, Hidetaka; Nariai, Hideki*; Inasaka, Fujio*

JSME International Journal, Series B, 44(1), p.81 - 89, 2001/01

no abstracts in English

Journal Articles

Analysis of steam-blanket effect in PCCS with horizontal heat exchanger

Onuki, Akira; Nakamura, Hideo; Anoda, Yoshinari

Dai-7-Kai Doryoku Enerugi Gijutsu Shimpojiumu Koen Rombunshu (00-11), p.258 - 263, 2000/00

no abstracts in English

Journal Articles

Numerical investigation of heat transfer enhancement phenomenon during the reflood phase of PWR-LOCA

Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 36(11), p.1021 - 1029, 1999/11

 Times Cited Count:1 Percentile:13.53(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Application of neutron radiography to visualization of cryogenic fluid boiling two-phase flows

Takenaka, Nobuyuki*; Asano, Hitoshi*; Fujii, Terushige*; Ushiro, Toshihiko*; Iwatani, Junji*; Murata, Yutaka*; Mochiki, Koichi*; Taguchi, Akira*; Matsubayashi, Masahito; Tsuruno, Akira

Nuclear Instruments and Methods in Physics Research A, 377(1), p.174 - 176, 1996/07

 Times Cited Count:1 Percentile:23.76(Instruments & Instrumentation)

no abstracts in English

Journal Articles

Analysis of density wave instability in a boiling flow using a characteristic method

Transient Thermal Hydraulics,Heat Transfer,Fluid-structue Interaction,and Structural Dynamics,ASME94, 0, p.65 - 73, 1994/00

no abstracts in English

JAEA Reports

Assessment of TRAC-PF1/MOD1 code for core thermal hydraulic behavior during reflood with CCTF and SCTF data

; Onuki, Akira; *; Murao, Yoshio

JAERI-M 93-032, 190 Pages, 1993/03

JAERI-M-93-032.pdf:3.0MB

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

Journal Articles

Effect of liquid flow rate on film boiling heat transfer during reflood in rod bundle

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 27(6), p.535 - 546, 1990/06

no abstracts in English

JAEA Reports

Reflood experiments in single rod channel under high-pressure condition

G.Xu*; Kumamaru, Hiroshige; Tasaka, Kanji

JAERI-M 89-178, 35 Pages, 1989/11

JAERI-M-89-178.pdf:0.74MB

no abstracts in English

38 (Records 1-20 displayed on this page)