• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.190-190
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• 1999

CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.209-209
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.189.2-189
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• 2002

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.410-423
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• 2000

A two-dimensional numerical model is developed and applied to the LAVA-EXV tests performed at the Korea Atomic Energy Research Institute (KAERI) to investigate the external cooling effect on the thermal margin to failure of a reactor pressure vessel (RPV) during a severe accident. The computational program was written to predict the temperature profile of a two-dimensional spherical vessel segment accounting for the conjugate heat transfer mechanisms of conduction through the debris and the vessel, natural convection within the molten debris pool, and the possible ablation of the vessel wall in contact with the high temperature melt. Results of the sensitivity analysis and comparison with the LAVA-EXV test data indicated that the developed computational tool carries a high potential for simulating the thermal behavior of the RPV during a core melt relocation accident. It is concluded that the main factors affecting the RPV failure are the natural convection within the debris pool and the ablation of the metal vessel, The simplistic natural convection model adopted in the computational program partly made up for the absence of the mechanistic momentum consideration in this study. Uncertainties in the prediction will be reduced when the natural convection and ablation phenomena are more rigorously dealt with in the code, and if more accurate initial and time-dependent conditions are supplied from the test in terms of material composition and its associated thermophysical properties.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.227.2-227
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.160-160
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.159-159
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• 2002

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• 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.