• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3086-3091
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled-down as the half height and 1/238 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.308-311
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• 2008

As a severe accident mitigation strategy in a nuclear power plant, ERVC(External Reactor Vessel Cooling) has been proposed. Under ERVC conditions, where a molten corium is relocated in a reactor vessel lower head, a natural circulation two-phase flow is driven in the annular gap between the reactor vessel wall and its insulation. This flow should be sufficient to remove the decay heat of the molten corium and maintain the integrity of the reactor vessel. Preliminary experimental study was performed to estimate the natural circulation two-phase flow. The experimental facility which is one dimensional, the half height, and the 1/238 channel area of APR1400, was prepared and the experiments were carried out to estimate the natural circulation two-phase flow with varying the parameters of the coolant inlet area, the heat rate, and the coolant inlet subcooling. In results, the periodic circulation flow was observed and the characteristics were varied from the experimental parameters. The frequency of the natural circulation flow rate increased as the wall heat flux increased.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.427-434
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• 1996

OECD-NEA 주관으로 수행된 TMI-2의 압력용기 변형연구의 결과, 하반부의 creep해석에 많은 문제점이 제기되어 있다. 본 논문은 TMI-2 노심용융 사고에 대한 기존 구조해석에서 creep 상관식의 형태, 적용방법 및 FEM 해석절차상의 상이점을 밝혀내고 이에 따라 압력용기 하반부의 파손확률이 크게 다르게 결정됨을 보였다. 기존의 TMI-2 구조해석에서 주 오차의 요인으로서 시간의 변화에 따른 국부열점 및 이를 포함한 재배치된 용융노심의 열경계조건의 불확실도와 압력용기강의 creep strain을 시간 및 온도에 대하여 불충분하게 묘사한 점을 밝혔다. 또한 creep-rupture 예측에 사용된 Larson-Miller Parameter도 해석을 지나치게 보수적인 결과로 유도하였다. 중대사고시 압력용기 하반부 천공방어를 위한 방안인 용기하부 외벽 냉각방식을 적용하였을 때 TMI-2 사고를 재해석한 결과, 압력용기의 건전성이 충분한 보수성을 가지고 유지됨을 보였다.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2626-2631
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• 2007

Air-water two phase natural circulation flow in the T-HERMES (Thermo-Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow)-1D experiment has been evaluated to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5 results have shown that an increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not effective on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. The water level is not effective on the water circulation mass flow rate. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it is not effective on the local pressure.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.355-361
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• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.1
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• pp.23-31
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• 2011

At supercritical pressure, the physical properties of fluid change substantially and the heat transfer at a temperature similar to the critical or pseudo-critical temperature improves considerably; however, the heat transfer may deteriorate due to a sudden increase in the wall temperature at a certain condition of a mass and heat flux. In this study, the heat transfer rates in $CO_2$ flowing vertically upward and downward in a circular tube with a diameter of 4.57 mm under various conditions were calculated by measuring the temperature of the outer wall of the tube. The published heat transfer correlations were analyzed by comparing their prediction values with 7,250 experimental data. By introducing a buoyancy parameter, a heat transfer correlation, which could be applied only to a normal heat transfer regime, was extended such that it can be applied to regime of heat transfer deterioration. The published criteria for heat transfer deterioration were evaluated against the conditions obtained from the experiment in this study.