• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.351-361
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• 2015

External reactor vessel cooling (ERVC) for in-vessel retention (IVR) has been considered one of the most useful strategies to mitigate severe accidents. However, reliability of this common idea is weakened because many studies were focused on critical heat flux whereas there were diverse uncertainties in structural behaviors as well as thermal-hydraulic phenomena. In the present study, several key factors related to molten corium behaviors and thermal characteristics were examined under multi-layered corium formation conditions. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative reactor pressure vessel (RPV) to figure out the possibility of high temperature induced failures. From the sensitivity analyses, it was proven that the reactor cavity should be flooded up to the top of the metal layer at least for successful accomplishment of the IVR-ERVC strategy. The thermal flux due to corium formation and the relocation time were also identified as crucial parameters. Moreover, three-layered corium formation conditions led to higher maximum von Mises stress values and consequently shorter creep rupture times as well as higher damage factors of the RPV than those obtained from two-layered conditions.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.21-28
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• 2013

The flow and heat transfer characteristics of the ex-vessel core melt (corium) were investigated using a commercial CFD code along with the experimental data on the spreading of corium available in the literature (VULCANO VE-U7 test). In the numerical simulation of the unsteady two-phase flow, the volume-of-fluid model was applied for the spreading and interfacial surface formation of corium with the surrounding air. The effects of the key parameters were evaluated for the corium spreading, including the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The results showed a reasonable trend of corium progression influenced by the changes in the radiation, decay heat, temperature-dependent viscosity and initial temperature of corium. The modeling of the viscosity appropriate for corium and the radiative heat transfer was critical, since the front progression and temperature profiles were strongly dependent on the models. Further development is required for the code to consider the formation of crust on the surfaces of corium and the interaction with the substrate.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.4
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• pp.425-429
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• 2011

In the unlikely of nuclear reactor meltdown, the leaked core melt or corium must be contained in a device called core-catcher so that the corium can be cooled and stabilized. The ex-vessel behavior of corium involves complex physical and chemical mechanisms of flow propagation, heat transfer, and reactions with sacrificial substrates. In this study, the detailed characteristics of corium flow and heat transfer were investigated by using a commercial CFD code for VULCANO VE-U7 test reported in the literature. The volume-of-fluid (VOF) model was used to predict the interfacial surface formation of corium and the surrounding air, and the discrete ordinate model was adopted to calculate radiation between corium and the surroundings. It was found that cooling via radiation through the top surface of corium had a dominant effect on the temperature and viscosity profiles at the front of the corium flow.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.66-73
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• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.26-32
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• 2015

During the course of a severe accident in a light water nuclear reactor, large amounts of hydrogen can be generated and released into the containment during reactor core degradation. Additional burnable gases [hydrogen ($H_2$) and carbon monoxide (CO)] may be released into the containment in the corium/concrete interaction. This could subsequently raise a combustion hazard. As the Fukushima accidents revealed, hydrogen combustion can cause high pressure spikes that could challenge the reactor buildings and lead to failure of the surrounding buildings. To prevent the gas explosion hazard, most mitigation strategies adopted by European countries are based on the implementation of passive autocatalytic recombiners (PARs). Studies of representative accident sequences indicate that, despite the installation of PARs, it is difficult to prevent at all times and locations, the formation of a combustible mixture that potentially leads to local flame acceleration. Complementary research and development (R&D) projects were recently launched to understand better the phenomena associated with the combustion hazard and to address the issues highlighted after the Fukushima Daiichi events such as explosion hazard in the venting system and the potential flammable mixture migration into spaces beyond the primary containment. The expected results will be used to improve the modeling tools and methodology for hydrogen risk assessment and severe accident management guidelines. The present paper aims to present the methodology adopted by Institut de Radioprotection et de $S{\hat{u}}ret{\acute{e}}$ $Nucl{\acute{e}}aire$ to assess hydrogen risk in nuclear power plants, in particular French nuclear power plants, the open issues, and the ongoing R&D programs related to hydrogen distribution, mitigation, and combustion.