• Journal of the Korean Society of Visualization
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• v.18 no.3
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• Transactions of the Korean Society of Mechanical Engineers
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• v.10 no.1
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• pp.7-14
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• 1986

To predict the fuel clad temperature during the reflooding phase of a LOCA, one may need a knowledge of reflood heat tranfer mechanism in a rod bundle. For this purpose reflooding experiments have been carried out with an electrically heated 3*3 rod bundle. Using the method for the determination of local heat transfer coefficient from the measured wall temperature the parametric effects of coolant flow rate, initial wall temperature, coolant subcooling and heat generation rate on the propagation of rewetting front were investigated. Prediction of the wall temperature histories for these experiments was discussed using REFLUX code with modification of the rewetting temperature correlation. Through this modification, better agreement between experiment and prediction was obtained.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.409-414
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• 1997

Reflood model in RELAP5/MOD3.1 are modified to improve the unrealistic prediction results of the model. In the new method, the modified Zuber pool boiling critical heat flux (CHF) correlation is adopted. The reflood drop size is characterized by the use of We=1.5 and the minimum drop size of 0.0007 m for $p^{*}\;{\leq}\;0.025$. To describe the wall to vapor heat transfer at low pressure and low flow condition, the Webb-Chen correlation is utilized . The suggested method has been verified through the simulations of the Lehigh University rod bundle reflood tests. Through sensitivity study it is shown that the effect of drag coefficients is dominant in the reflood model. It is proved that the present modifications result in much more improved quench behavior and accurate wan and vapor temperature predictions.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.521-529
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• 2000

A modified interfacial heat transfer correlation between a dispersed water droplet and ambient superheated steam is proposed and compared with available experimental data and other correlations. Modified one overcomes the inherent deficiencies of Lee and Ryley's interfacial heat transfer correlation that ignored the effects of steam superheating which can not be neglected especially in the reflood situation of a loss-of-coolant accident. Modified one is represented by (equation omitted) In the present correlation the effect of possible subcooling of a water droplet is not taken into consideration. Comparison of the above correlation with currently available measurement data for a water droplet in high temperature gas flow shows that the proposed one correlates well with the measurement data where the degree of superheating is negligible and considerable.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.22 no.5
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• pp.554-560
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• 2021

CRUD (chalk river unidentified deposits) is a porous material deposited on the surface of nuclear fuel during nuclear power plant operation. The CRUD is composed of metal oxides, such as iron, nickel, and chromium. It is essential to investigate the effects of the CRUD layer on the wall heat transfer between the nuclear fuel surface and the coolant in the event of a nuclear accident. CRUD only negatively affects the temperature of the nuclear fuel due to heat resistance because the effects of the CRUD layer on two-phase boiling heat transfer are not considered. In this study, the physical property models for the porous CRUD layer were developed and implemented into the SPACE code. The effects of boiling heat transfer models on the peak cladding temperature and quenching were investigated by simulating a reflood experiment. The calculation results showed some positive effects of the CRUD layer.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.265-276
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• 1994

Several improvements to the RELAP5/MOD3 reflood model hate been made. These improvement were made to correct deficiencies in the reflood model identified by the assessment of the RELAP5/MOD3 code against FLECHT-SEASET experiments. The improvements consist of modification of reflood wall heat transfer package and adjusting the droplet size in dispersed flow regime. The time smoothing of wall vaporization and level tracking of transition flow are also added to eliminate the pressure spikes and level oscillation during reflood process. Assessment of the improved model against FLECHT-SEASET experimental data and application of LBLOCA analysis for plant shows that the deficiencies have been corrected.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.32-38
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• 2024

The best estimate plus uncertainty methodologies for loss-of-coolant accident analyses make use of the best-estimate codes and relevant experimental databases. Inherently, best-estimate codes have various uncertainties in the model parameters, which can be quantified by the dedicated experimental database. Therefore, this study was devoted to establishing procedures for identifying the input parameters of predictive models and quantifying their uncertainty ranges. The rod bundle heat transfer experiments were employed as a representative reflood separate effect test, and the TRACE code was utilized as a best-estimate code. In accordance with the present procedure for uncertainty quantification, the integrated list of the influential input parameters and their uncertainty ranges was obtained through local sensitivity calculations and screening criteria. The validity of the procedure was confirmed by applying it to uncertainty analyses, which checks whether the measured data are within computed ranges of the variables of interest. The uncertainty quantification procedure proposed in this study is anticipated to provide comprehensive guidance for the conduct of uncertainty analyses.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.326-336
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• 1990

Recent experiments show the existence of spacer grid improves the heat removal from the fuel rods during the reflood phase of LOCA. The local heat transfer within and downstream of the grid is increased due to the earlier quenching than rod surface, shattering of the entrained droplets into smaller ones which can be more easily evaporated and enhanced turbulent effect. Therefore, the consideration of these phenomena is necessary for the DFFB regime which prevails above the water level during the reflood. In this paper, droplet breakup model at spacer grid has been developed and incorporated into RELAP5/MOD2. Verification calculations are carried out for FEBA tests which examine the thermalhydraulic performance of grid spacer during reflood.

• Journal of Energy Engineering
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• v.9 no.3
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• pp.261-268
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• 2000

A heat transfer correlation to predict the vaporization of a water droplet in highly superheated steam during a loss-of-coolant accident(LOCA) of a nuclear power plant is provided. Vaporization of liquid fuel or water droplets in superheated air or steam and subsequent interface heat transfer between a liquid droplet and superheated gas is typically correlated by way of a Nusselt number as a function of Reynolds number, Prantl number, and in some cases including mass transfer number. Presently available correlations and experimental data of the evaporation of liquid droplets in air or steam are analyzed and a new Nusselt number correlation is proposed taking Schmidt number into consideration in order to account for binary diffusion of the vapor as well, Nu$\_$f/(1+B)$\^$0.7/=2+0.53Sc$\_$f/$\^$-1/5/Re$\_$M/$\^$$\sfrac{1}{2}$/Pr$\_$f/$\^$$\sfrac{1}{3}$/ for which properties are evaluated at film condition except the density of Reynolds number evaluated at ambient condition. Diverse correlations for various combinations of liquid and gas species are put into single equation. The blowing correction factor of (1+B)$\^$0.7/ is confirmed appropriate, and a criterion to distinguish so-called high- and low-temperature condition of ambient gas is set forth.

• Nuclear Engineering and Technology
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• v.12 no.1
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• pp.9-18
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• 1980

Reflood experiments under atmospheric pressure have been conducted with a single heated tube to investigate basically the rewetting phenomena following a LOCA. Experimental conditions are 180cm length of test tube, wall temperature range of 300-80$0^{\circ}C$, coolant flooding rate of 5-30cm/sec. and subcooling of 35-85$^{\circ}C$. Experiments show that the rewetting velocity is dependent on the initial wall temperature of test tube, coolant flow rate and coolant subcooling. It is required to develop the proper method to evaluate the rewetting temperature and the heat transfer coefficient.