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Liu, W.; Tamai, Hidesada*; Takase, Kazuyuki
Journal of Heat Transfer, 135(8), p.081502_1 - 081502_13, 2013/08
Times Cited Count：7 Percentile：34.19(Thermodynamics)For a steam generator in a commercialized sodium-cooled Fast Breeder Reactor (FBR), flow instability in the water side is one of the most important items need research. As the first step of the research, thermal-hydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steam-water pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and two-phase flow multipliers. As a result, the drift flux model implemented in the TRAC-BF1 code was confirmed to suitably predict the void fraction well, under the present high pressure conditions. For the two-phase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present high-pressure conditions.
Katono, Kenichi*; Tamai, Hidesada*; Nagayoshi, Takuji*; Ito, Takashi*; Takase, Kazuyuki
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
no abstracts in English
Nagatake, Taku; Tamai, Hidesada; Akimoto, Hajime; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai Dai-25-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.718 - 719, 2012/10
no abstracts in English
Watanabe, Hironori; Tamai, Hidesada; Sato, Takashi; Shibata, Mitsuhiko; Mitsutake, Toru*
Flow Measurement, p.95 - 106, 2012/03
In Boiling Water Reactor (BWR), reactor power, fuel conversion ratio and reactor cooling capacity changes by the void fraction in the core. We have developed a capacitance method (C method) to measure the void fraction under the condition of high temperature and high pressure of 7MPa, simulating reactor. This C method is based on the principle that the capacitance in the two-phase flow is a function of void fraction. Using this method, we can measure void fractions in real time, at all region of void fractions, and with a small error of measurement, which was not realized by usual techniques up to now.
Tamai, Hidesada; Akimoto, Hajime; Takase, Kazuyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 11(1), p.8 - 12, 2012/03
The Fukushima Daiichi Nuclear Plant Unit 1 accident was investigated with TRAC-BF1 code in order to confirm the effect of an isolation condenser (IC) on core cooling analytically. The analysis shows that it is too late to cool fuel rods, once the core is heated up because of the lack of coolant derived from discharge of steam through a safety relief valve. It also shows that early start-up of the IC is essential to avoid the core meltdown under station blackout conditions.
Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Journal of Power and Energy Systems (Internet), 5(3), p.229 - 240, 2011/04
For a steam generator with straight double-walled heat transfer tubes that used in a sodium cooled faster breeder reactor, clarification of flow instability in heat transfer tubes is one of the most important research themes. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15MPa. Several two-phase flow multiplier models were checked and then, it was found that both two-phase flow multiplier models of Chisholm and homogeneous can predict the present experimental data in high accuracy.
Liu, W.; Tamai, Hidesada; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
For steam generator with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important issues need researching. As the first step of the research, thermal hydraulics experiments with water were performed under high pressure condition in JAEA with using a circular tube with a similar inner diameter as that in the designed SG. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper summarizes the pressure drop characteristics under 15 MPa. Six models for the prediction of two-phase multiplier were evaluated. The results showed the Chisholm correlation and homogeneous model gave best predictions. Note that in the homogeneous model verification, the homogeneous model was only used in the friction loss calculation. In the calculation of void fraction, which is necessary for static head, drift flux model, instead of homogeneous model, was used.
Tamai, Hidesada; Nagayoshi, Takuji; Katono, Kenichi; Ito, Takashi; Takase, Kazuyuki
Proceedings of 7th International Conference on Multiphase Flow 2010 (ICMF 2010) (CD-ROM), 7 Pages, 2010/05
For the design the natural-circulation type BWR that utilizes the FSS concept, the development of a predictive model for the droplets entrained with the steam (carryover) from the free surface is indispensable. In this paper, the droplet quality was measured with a throttling calorimeter that could measure the droplet quality based on the isenthalpic process between wet and superheated steam through the throttle. The measurements were carried out under the conditions of a pressure of 1.5-2.5 MPa. The temperature of the superheated steam after passing through the throttle was confirmed to be strongly related to the quality of the wet steam. A modified model based on the measurements proved to be capable of predicting the droplet quality within the range of the database. Evaluating the droplet quality under BWR conditions validated the feasibility of the design of the natural-circulation type BWR that utilizes the FSS concept.
Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 10 Pages, 2009/09
For the Steam Generator (SG) in a commercialized sodium cooled faster breeder reactor, flow instability in water side is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments using water as test fluid were performed under high pressure condition at JAEA with using a circular tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper focuses on the discussion to steam - water pressure drop. We evaluated existing correlations for two-phase flow multiplier under high pressure. As a result, Chismholm correlation was confirmed being the best one for the present high pressure data.
Tamai, Hidesada; Nagayoshi, Takuji; Katono, Kenichi; Ito, Takashi; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2009 Koen Rombunshu, P. 2, 2009/08
The characteristics of carryover from free-surface in a natural-circulation BWR are an important subject to be resolved for economic and safe design of the reactor. In this study, droplet quality of the carryover in a test section with 0.12 m in diameter was measured using throttling calorimeter with pressure ranging from 1.5-2.5 MPa. The measured droplet quality increases with decrease in distance from free-surface and with increase in vapor volumetric flux, and these trends are similar to those of previous data.
Akimoto, Hajime; Anoda, Yoshinari; Takase, Kazuyuki; Tamai, Hidesada; Yoshida, Hiroyuki
Genshiryoku Kyokasho "Genshiryoku Netsuryudo Kogaku", 336 Pages, 2009/03
no abstracts in English
Misawa, Takeharu; Yoshida, Hiroyuki; Tamai, Hidesada; Takase, Kazuyuki
Journal of Power and Energy Systems (Internet), 3(1), p.194 - 202, 2009/00
Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 5 Pages, 2008/11
To discuss the feasibility of Steam Generator (SG) with a straight double-walled heat transfer tube that used in the Fast Breeder Reactor (FBR) system, we need to construct thermal hydraulic design method that can predict the flow instability accurately. To verify and to improve the correlations that used in the thermal-hydraulic design of the SG, Japan Atomic Energy Agency has started experiments under high pressure conditions. Detailed thermal hydraulic data including pressure drop data have been derived. This research does the analysis to the performed experiments with using TRAC-BF1 code. The pressure drop under high pressure condition is verified. It is found that with using the drift flux model in Track code for the void fraction calculation, Pffan's correlation for the friction pressure drop calculation in single phase flow and Martinelli-Nelson two-phase multiplier, the pressure drop can be predicted conservatively.
Tamai, Hidesada; Kureta, Masatoshi; Liu, W.; Sato, Takashi; Nakatsuka, Toru; Onuki, Akira; Akimoto, Hajime
Journal of Nuclear Science and Technology, 45(6), p.567 - 574, 2008/06
Times Cited Count：1 Percentile：10.05(Nuclear Science & Technology)The confirmation of thermal-hydraulic performance is one of the most important requirements for the design of the FLWR. A large-scale thermal-hydraulic experiment using a tight-lattice 37-rod bundle test section with a bowed rod was carried out with pressure ranging from 2-9 MPa and mass velocity at 200-1000 kg/(m
s). It was confirmed that boiling transition (BT) occurs downstream of the rod contact point, and that the wall temperature trace during the BT follows the typical BT pattern of BWR. Critical power with a bowed rod is about 10 percent lower than that without rod bowing. The critical power increases monotonically with increase in mass velocity, with decrease in inlet water temperature, and with decrease in exit pressure, and these trends are similar to those of the basic bundle without rod bowing. Thus, there is negligible effect of rod bowing on the dependence of critical power on the mass velocity, the inlet temperature and the exit pressure.
Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Akimoto, Hajime
Konsoryu Kenkyu No Shinten, 3, p.99 - 109, 2008/06
An estimation of void fraction in tight-lattice rod bundles was carried out. Five types of void fraction experiments with 7-, 14-, 19- and 37-rod and rod-gap of 1.0 - 1.3 mm bundle and spacer effect tests were conducted ranging from 0.1 to 7.2 MPa. Extensibility of a TRAC-BF1 code and one-dimensional drift-flux model to the tight-lattice rod bundle was studied. The TRAC-BF1 and the model calculated the void fraction with good agreement to data in case of relatively high quality and void fraction region. Applicability of a NASCA, ACE-3D, TPFIT codes to the tight-lattice rod bundle was verified by comparing with the three-dimensional void fraction data measured by neutron tomography. Tendency of the calculated void fraction by these codes and measured data was similar. Vapor distribution and velocity profile of water and vapor were discussed based on data.
Tamai, Hidesada; Tomiyama, Akio*
Journal of Power and Energy Systems (Internet), 2(1), p.295 - 305, 2008/00
A three-dimensional one-way bubble tracking method is a promising numerical method for calculation of time-spatial evolution of gas-liquid interfacial configuration with use of a little computing resource. Since the method has been applied to only an adiabatic air-water bubble flow, the method is developed for the analysis of a boiling flow in this study. One-dimensional Eulerian equation of energy conservation for a continuous liquid phase and an interface heat transfer equation for dispersed bubbles are introduced. Then, radial liquid temperature distribution, wall heat transfer between a heated wall and subcooled liquid, bubble generation on a heated wall and expansion or condensation of bubbles are taken into account. The developed method is applied to the boiling flow experiment and radial void fraction distribution is compared. It is confirmed that the method can give good prediction of tendency of the void fraction distribution in the boiling flow.
Liu, W.; Tamai, Hidesada; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.240 - 249, 2008/00
This paper describes the critical power characteristics in a 37-rod tight-lattice bundle with rod bowing under transient states. It is observed that transient Boiling Transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle, which is same as that under steady state. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transients are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time conservatively. Traditional quasi-steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight-lattice bundle with rod bowing.
Kureta, Masatoshi; Tamai, Hidesada; Yoshida, Hiroyuki; Onuki, Akira; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.271 - 282, 2008/00
An estimation of void fraction in a tight-lattice rod bundle was needed of the design of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa. Followings were made clear: (1) numerical analysis codes calculate the similar distribution to the data, and (2) TRAC-BF1 code and drift-flux model tends to overestimate the void fraction at lower quality region.
Onuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime
Journal of Power and Energy Systems (Internet), 2(1), p.229 - 239, 2008/00
Tamai, Hidesada
Nihon Genshiryoku Gakkai-Shi ATOMO
, 49(11-12), p.745 - 749, 2007/12
An Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core has a tight-lattice bundle structure and it is operated under low mass velocity and high void fraction conditions, for aiming at the achievement of a high conversion ratio of plutonium mixed oxide (MOX) fuel. These conditions make core cooling difficult, and the FLWR thermal-hydraulic characteristics under such conditions are not known well. The confirmation of thermal-hydraulic characteristics is, therefore, one of the most important R&D requirements for the FLWR design. We investigated the thermal-hydraulic performance of the FLWR core using a test section with 37-rod bundles under high pressure conditions simulating the FLWR operating conditions. The result obtains that the FLWR has sufficient thermal margins for cooling of the core.
