• 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 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.

• 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.9 no.4
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• pp.463-470
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• 1985

In this study a parametric analysis for the rewetting temperature was made with 572 data obtained from the single tube experiment. The rewetting temperature was also evaluated by measuring the vaporization time of a liquid drop on a hot surface at the elevated pressures. The results showed that the rewetting temperature increased with flooding rate, inlet subcooling pressure and initial wall temperature, and decreased with increasing axial elevation. Based on the results obtained, the rewetting temperature correlation was suggested. From the comparison of correlated rewetting temperatures with measured values, it showed that the correlated values fell within .+-.5% error from the measured values.

• Journal of Energy Engineering
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• v.17 no.2
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• pp.67-76
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• 2008

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). Recently several integral effect tests for the reflood period of a LBLOCA (Large Break LOss of Coolant Accident) of the APR1400 have been performed with the ATLAS. In the APR1400 a high flow condition is changed to a low flow condition due to an fluidic device during an operation of the SIT. As the self-controlled fluidic device was not installed in the ATLAS, a set of characterization tests was performed to simulate its injection capability from the SIT for the APR1400 simulation. In the ATLAS the required SIT flow rate in the high flow condition was acquired by installing orifices with an optimized flow area to throttle the SIT discharge line and the low flow condition was achieved by changing the opening of the flow control valve in the SIT injection line. The test results showed that the safety injection systems of the ATLAS could simulate the required high and low flow rates of the SIT for the APR1400 simulation efficiently.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.821-827
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• 2001

As one of the advanced design features of the Korea next generation reactor, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLl) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, a new scaling method, using time and velocity reduced linear scaling law, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in PWR downcomer.

• Journal of Life Science
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• v.8 no.5
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• pp.557-562
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• 1998

This study was conducted to investigate the influence of soil moisture on the photosynthetic rate, transpiration rate, stomatal conductance, leaf water content and its any correlation in soybean. Followings were achieved as a conclusion. At the soil moisture was moved from drought condition to saturation condition, the photosynthetic rate, transpiration rate, stomatal conductance and leaf water content were gradually increased on the period of re-watering treatment. The recovery of photosynthetic rate was faster than others. There were positive correlation with the photosynthetic rate and leaf water content, the stomatal conductance and leaf water content, the photosynthetic rate and stomatal conductance respectively, but the correlation coefficient of photosynthetic rate and leaf water content was high.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• KOREAN JOURNAL OF CROP SCIENCE
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• v.42 no.2
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• pp.189-195
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• 1997

To cope with increasing importance of water stress in food crop production, characteristics and cultivar differences in physiological responses of winter barley plants to water stress during seedling stage were studied employing three covered-barley, Milyang 12, Durubori, and Olbori, one naked-barley, Baekdong, and one two-row beer-barley, Hyangmaeg. Water stress was given to barley plants by withholding watering in soil-culture pots or by adding P.E.G. (polyethylene glycol, M. W. 4,000) to the nutrient culture solution. When water stress was imposed by withholding watering for 10 days to the seedlings grown in pot-soil for 10 days after emergence, leaf water potential of 5 cultivars was decreased almost to the same degree, from control -3.5 bar to the water stressed -26.6 bar. However, growth and some physiological responses were differently affected by cultivars: The plant height was decreased by 16-26％ : seedling dry weight by 25∼42％ ; chlorophyll content by 21∼39％ ; second leaf survival rate by 8∼18％ ; and free proline content per gram of leaf dry weight was increased from control 0.2∼0.5mg to the water stressed 5.8∼13.2mg. Drought resistances of 5 barley cultivars, based on the leaf survival rate and the decreased rate in seedling dry weight at the end of stress treatment, were high in the order of 3 covered-barley＞naked-barley＞two-row barley.