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AbstractAbstract

[en] Analytical solutions of the magnetic pressure drop and the heat transfer coefficient were obtained for liquid metal flow in an annular channel with perfectly conducting walls under transverse magnetic field. Numerical calculations of the magnetic pressure drop and the heat transfer coefficient were also conducted for liquid metal flow in an annular channel with finite wall conductivity under transverse magnetic field. It was assumed in the analysis and the numerical calculations that the flow was fully laminarized due to the effect of strong magnetic field and the velocity and temperature fields were fully developed. It became clear from the results of numerical calculations that the Poiseuille number increased with the Hartmann number and the wall conductivity numbers, and that the Nusselt numbers at 0 and π from the direction of magnetic field were lower than the average Nusselt number, but higher at π/2. The Poiseuille number and the Nusselt number obtained by the numerical calculations for large values of the wall conductivity numbers almost agreed with the analytical solutions for the case of walls with infinite conductivity. (author)

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Journal Article

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 21(5); p. 393-400

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Kumamaru, Hiroshige

Report of study meeting on heat transfer and structure problems in nuclear facilities, 1979

Report of study meeting on heat transfer and structure problems in nuclear facilities, 1979

AbstractAbstract

[en] Brief review of the studies on the flow and heat transfer of two-component, two-phase flow of liquid metal in magnetic field is presented. R.J. Thome measured the distribution of void rate, slip ratio and pressure loss for the two-phase flow of NaK-N

^{2}under vertical magnetic field. The void rate distribution became even and the slip ratio increased with the increasing magnetic field. The experimental results of pressure loss was compared with the calculation by an equation derived from the homogeneous flow model. R.G. Owen et al. made the analytical studies of the MHD friction loss of two phase flow. Michiyoshi et al. made experimental studies on the hydrodynamic local properties of Hg-Ar two-phase flow of slug region in a vertically ascending tube under magnetic field, and Kimi et al. also made studies on the heat transfer of Hg-Ar flow under magnetic field. Saito et al. measured the slip ratio and pressure loss of NaK-N^{2}flow. As a whole, it can be said that the average void rate decreases, and its distribution becomes even under magnetic field. The slip ratio increases, and the friction loss factor becomes nearly one. It was hard to make clear the heat transfer characteristics. (Kato, T.)Primary Subject

Source

Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab; 114 p; May 1980; p. 46-60; Study meeting on heat transfer and structural problems in nuclear facilities; Tokai, Ibaraki (Japan); 4 Feb 1980

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Report

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AbstractAbstract

[en] Based on simple models, magnetic pressure frop and heat transfer coefficient correlations were derived for two-component two-phase liquid metal flow in an annular channel under magnetic field. Magnetic pressure drop and heat transfer experiments of Na-Ar two-phase flow in an annular channel under magnetic field were conducted. It became clear from the experimental results that the magnetic pressure drop of two-component two-phase liquid metal flow was equal to or smaller than that of single-phase liquid metal flow under conditions of the same liquid flow rate and magnetic field strength. The measured magnetic friction factor multiplier was smaller than predictions by the correlations derived in this study. It was also revealed from the experimental results that the heat transfer coefficient of two-component two-phase liquid metal flow was equal to or slightly smaller than that of single-phase liquid metal flow under conditions of the same liquid flow rate and magnetic field strength. The tendency that the measured heat transfer coefficient ratio decreased from 1.0 with increasing the void fraction agreed with trends of predictions by the correlations derived in this study. (author)

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Journal Article

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; v. 21(6); p. 438-449

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Koizumi, Yasuo; Yagi, Junji; Kumamaru, Hiroshige.

Japan Atomic Energy Research Inst., Tokyo (Japan)

Japan Atomic Energy Research Inst., Tokyo (Japan)

AbstractAbstract

[en] In a countercurrent two-phase flow, where gas phase flows in the upward direction against a gravity-driven liquid downflow, the liquid downflow rate begins to be limited when the gas flow rate exceeds a certain threshold value. This phenomenon, termed 'flooding', may occur during a loss-of-coolant accident (LOCA) at such locations in reactor coolant system as steam generator (SG) U-tubes in a pressurized water reactor (PWR). Flooding generally tends to reduce the amount of water available for core cooling in emergency situations. Flooding has been studied for various flow conditions and geometries, in particular for vertical channels. Most of these studies were concerned with those situations where the lower entry of the channel is exposed to the gas phase or a gas-continuous two-phase flow, and scarcely dealt with such situations where the liquid is the continuous phase at the channel lower entry. However, in a PWR small-break LOCA, where the reactor coolant inventory is depleted only slowly, the latter situations would be encountered more frequently than the former. The present study is concerned with flooding in a vertical channel whose lower entry is facing to a liquid-continuous two-phase flow. Experiments were conducted using Freon R-113 as a simulant of high-pressure steam-water two-phase flow. Experimental results indicate that flooding for this situation initiates when the two-phase mixture swell level in the channel, which indicates large fluctuations with time, reaches the channel top entry at the peaks of level fluctuations. It was also found that the flooding correlation developed formerly by the authors for air-water flows can be applied to the present R-113 case if the difference in fluid properties are considered appropriately. (author)

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Oct 1993; 57 p

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AbstractAbstract

[en] Numerical calculations have been performed on magnetohydrodynamic (MHD) two-phase annular flow in a rectangular channel with a small aspect ratio, i.e.a small ratio of the channel side perpendicular to the applied magnetic field and the side parallel to the field. Results of the present calculation agree nearly with Inoue et al.'s experimental results in the region of large liquid Reynolds numbers and large Hartmann numbers. Calculation results also show that the pressure drop ratio, i.e. the ratio of pressure drop of two-phase flow to that of single-phase flow under the same liquid flow rate and applied magnetic field, becomes lower than ∼0.02 for conditions of a fusion reactor plant. (author)

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Journal Article

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Nippon Kikai Gakkai Ronbunshu, B Hen; ISSN 0387-5016; ; v. 65(633); p. 1535-1541

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AbstractAbstract

[en] Numerical calculations on a magnetohydrodynamic (MHD) flow in a rectangular channel have been performed for thc cases of very-large Hartmann numbers (greater than ∼10

^{4}), finite wall conductivities and small aspect ratios (i.e. small length ratios of the channel side perpendicular to the applied magnetic field and the side parallel to the field), simulating fusion-reactor blanket conditions. The MHD pressure drops for the cases of aspect ratios considerably lower than 1.0 become fairly smaller than those for the cases of aspect ratios larger than 1.0. The calculation results also show that a two-dimensional numerical calculation is required for estimating exactly the MHD pressure drop in the cases of aspect ratios lower than 1.0. Also, numerical calculations on an MHD flow in a circular pipe have been carried out for the cases of very-large Hartmann numbers (greater than ∼10^{4}) and finite wall conductivities, simulating fusion-reactor blanket conditions. Calculation results show that the MHD pressure drops for very-large Hartmann numbers and finite wall conductivities can be predicted by an approximate equation derived by Chang-Lundgren for a circular pipe, though the equation did not cover originally these parameter ranges. (author)Primary Subject

Source

Japan Society of Mechanical Engineers, Tokyo (Japan); 4252 p; 1999; [10 p.]; ICONE-7: 7. international conference on nuclear engineering; Tokyo (Japan); 19-23 Apr 1999; This CD-ROM can be used for WINDOWS 95/98/NT, MACINTOSH and UNIX; Acrobat Reader 3.0.1 is included; Data in PDF format, Track No. 07, ICONE-7041

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AbstractAbstract

[en] Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5 ∼ 10 % of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1 % break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5 % or less, is unnecessary for breaks of approximately 5 % or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5 %. (author)

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Journal Article

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Numerical Data

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Journal of Nuclear Science and Technology (Tokyo); ISSN 0022-3131; ; CODEN JNSTAX; v. 29(12); p. 1162-1172

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ACCIDENTS, CONTROL EQUIPMENT, COOLING SYSTEMS, DATA, ECCS, ENGINEERED SAFETY SYSTEMS, ENRICHED URANIUM REACTORS, EQUIPMENT, FAILURES, FLOW REGULATORS, INFORMATION, NUMERICAL DATA, POWER REACTORS, REACTOR ACCIDENTS, REACTOR COMPONENTS, REACTOR COOLING SYSTEMS, REACTOR PROTECTION SYSTEMS, REACTORS, STRUCTURAL MODELS, THERMAL REACTORS, VALVES, WATER COOLED REACTORS, WATER MODERATED REACTORS

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AbstractAbstract

[en] Experiments were performed in a 5 . 5 rod bundle under conditions of total mass fluxes from 80 to 320 kg/m

^{2}s, inlet qualities from 0.1 to 0.8, heat fluxes from 3 to 26 W/cm^{2}and a pressure of 3 MPa. Heater rod surface temperatures or heat transfer coefficients predicted by several correlations were compared with experimental data with emphasis on the applicability of the correlations to the present experimental conditions which were pertinent to thermal-hydraulic conditions during a LOCA in a nuclear reactor. The Chen and Bjorge et al. correlations underestimated heat transfer coefficients in the pre-dryout region. The Varone-Rohsenow prediction which accounted for the thermal nonequilibrium effect, calculated heater rod surface temperatures relatively well in the post-dryout region over the whole region of the present experimental conditions. The Dittus-Boelter and Groeneveld correlations predicted heater rod surface temperatures relatively well in the post-dryout region under high total mass flux conditions, but underestimated considerably under low total mass flux conditions. (orig./HP)Primary Subject

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AbstractAbstract

[en] Free surface fluctuations at the neighboring two locations on a water jet have been evaluated using the developed non-intrusive optical technique. The free-surface slope angles are measured by detecting the two dimensional trajectories of laser beams refracted at the jet free surface. The obtained time series data of slope angles are divided into each wave period in accordance with the zero-up-crossing method. Then, the individual wave speed is evaluated from the dominant time lag of cross-correlation coefficient. The spatial shape of each wave is reconstructed by integrating the slope angle data. By comparing the analysis of wave shape statistics and the visual observation of free-surface photograph, it is found that the wave steepness reaches a maximum in the free surface region where a periodic wave train breaks down into less regular three dimensional patterns. In addition, the probability density functions of non-dimensional wave height are close to the Rayleigh distribution in the downstream region of wave breaking. (author)

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Journal Article

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Nippon Kikai Gakkai Ronbunshu, B Hen; ISSN 0387-5016; ; v. 76(763); p. 475-477

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Kumamaru, Hiroshige; Tasaka, Kanji

Japan Atomic Energy Research Inst., Tokyo

Japan Atomic Energy Research Inst., Tokyo

AbstractAbstract

[en] The relation between the liquid level signal obtained by a ROSA-III type conductivity probe and the liquid level, flow pattern or void fraction was clarified by visualized two-phase flow experiment at atmospheric pressure. It was also confirmed that the experimental data for potential pressure drop versus two-phase mixture level height agreed well with the simple theoretical equation. (author)

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Feb 1981; 34 p

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