• 대한조선학회논문집
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• 제61권5호
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• pp.334-342
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• 2024

In this study, sloshing effects on the heat transfer coefficient of the vertical walls in a membrane type LNG cargo containment system (CCS) were analyzed. To develop correlations of the heat transfer coefficient incorporating sloshing effects, experiments were conducted to measure the heat transfer coefficient of an oscillating vertical plate. Based on experiments, two different correlations were developed which are each applicable to vertical and horizontal oscillations. Using the developed correlations, the heat flow rate penetrated into the vertical walls in KC-1 CCS was investigated under sloshing conditions. The results showed that the penetrated heat flow rate increases with the sloshing speed, which is defined as the product of the sloshing amplitude and the frequency. The maximum increase ratio of the heat flow rate was found to be about 7.5% at the sloshing speed of 10m/s. Also, it was found that the penetrated heat flow rate more increases when the CCS oscillates in the perpendicular direction to the vertical walls, than the parallel direction to the vertical walls. This study suggests that the increase in wall heat transfer coefficients has to be considered when evaluating the boil-off rate of CCSs that are shaking due to sloshing.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.735-740
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• 2001

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, it occurs thermal stratification phenomena in case that there is the mixing of cooling water and high temperature water due to valve leakage in ECCS. This thermal stratification phenomena raises excessive thermal stresses at pipe wall. Therefore, this phenomena causes the accident that reactor coolant flows in reactor containment in the nuclear power plant due to the deformation of pipe and thermal fatigue crack(TFC) at the pipe wall around the place that it exists. Hence, in order to fundamental identification of this phenomena, it requires the experimental research of modeling test in the pipe flow that occurs thermal stratification phenomena. So, this paper models RCS and ECCS pipe arrangement and analyzes the mechanism of thermal stratification phenomena by measuring of temperature in variance with leakage flow rate in ECCS modeled pipe and Reynold number in RCS modeled pipe. Besides, results of this experiment is compared with computational analysis which is done in advance.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• Structural Engineering and Mechanics
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• 제60권2호
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• pp.337-350
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• 2016

Even though extensive researches have been performed for steam explosion due to their complex mechanisms and inherent uncertainties, establishment of severe accident management guidelines and strategies is one of state-of-the arts in nuclear industry. The goal of this research is primarily to examine effects of vessel failure modes and locations on nuclear facilities under typical steam explosion conditions. Both discrete and integrated models were employed from the viewpoint of structural integrity assessment of steel components and evaluation of the cracking and crushing in reinforced concrete structures. Thereafter, comparison of systematic analysis results was performed; despite the vessel failure modes were dominant, resulting maximum stresses at the all steel components were sufficiently lower than the corresponding yield strengths. Two failure criteria for the reinforced concrete structures such as the limiting failure ratio of concrete and the limiting strains for rebar and liner plate were satisfied under steam explosion conditions. Moreover, stresses of steel components and reinforced concrete structures were reduced with maximum difference of 12% when the integrated model was adopted comparing to those of discrete models.