• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1215-1222
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• 2009

The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in a nuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography. The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated. The MISTEE-NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First, analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics, the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.99-109
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• 1997

A dynamic model for hydrogen generation by Fuel-Coolant Interactions(FCI) is developed with separate models for each FCI stage, coarse mixing and stratification. The model includes the physical concept of FCI, semi-empirical heat and mass transfer correlation and the concentration diffusion equation with the general non-zero boundary condition. The calculated amount of hydrogen, which is mainly generated in stratification, is compared with the FITS experiments. The model developed in this study shows a good agreement within a range of 10 % fuel oxidation rate and predicts the controlled mechanism of the chemical reaction very well. And this model predicts more accurately than the previous works. It is shown from the sensitivity study that the higher initial temperature of fuel particle is, the larger the reaction rate is. Up to 2700 K of temperature of the particle, the reaction rate increases rapid, which can lead to metal ignition.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.213-213
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.216-216
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• 1999

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.12
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• pp.1051-1060
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• 2010

The effect of materials on fuel coolant interactions (FCIs) was analyzed on the basis of a solidified particle size response for TROI experiments.$^{(1)}$ The solidified particle size response can provide an understanding of the relationship among the initial condition, the mixing, and an explosion. Through a comparison of the size distributions of the solidified particles in the case of explosive and non-explosive FCIs, it is revealed that an explosive FCI results in the production of a large amount of fine particles and a small amount of large particles. The material effect of the size of solidified particles was analyzed using non-explosive FCIs without losing the information on the mixing. This analysis indicates that an explosive melt includes large particles that participate in the steam explosion, whereas a nonexplosive melt includes smaller particles and finer particles.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.218-227
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• 2016

Background: Steam explosions may occur in nuclear power plants by molten fuel-coolant interactions when the external reactor vessel cooling strategy fails. Since this phenomenon can threaten structural barriers as well as major components, extensive integrity assessment research is necessary to ensure their safety. Method: In this study, the influence of yield criteria was investigated to predict the failure of a reactor cavity under a typical postulated condition through detailed parametric finite element analyses. Further analyses using a geometrically simplified equivalent model with homogeneous concrete properties were also performed to examine its effectiveness as an alternative to the detailed reinforcement concrete model. Results: By comparing finite element analysis results such as cracking, crushing, stresses, and displacements, the Willam-Warnke model was derived for practical use, and failure criteria applicable to the reactor cavity under the severe accident condition were discussed. Conclusion: It was proved that the reactor cavity sustained its intended function as a barrier to avoid release of radioactive materials, irrespective of the different yield criteria that were adopted. In addition, from a conservative viewpoint, it seems possible to employ the simplified equivalent model to determine the damage extent and weakest points during the preliminary evaluation stage.

• 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.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.544-554
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• 1995

Pool boiling of dilute aqueous solutions of polyethylene oxide polymer has been experimentally investigated for the purpose of understanding the physical mechanisms of the suppression of vapor explosions in this polymer solution. Tn solid spheres of 22.2mm and 9.5mm-diameter ore heat-ed and quenched in the polymer solutions of various concentrations at 3$0^{\circ}C$. The results showed that minimum film boiling temperature($\Delta$ $T_{MFB}$) in this highly-subcooled liquid rapidly decreased from over $700^{\circ}C$ for pure water to about 15$0^{\circ}C$ as the polymer concentration was increased up to 300ppm for 22.2mm sphere, and it decreased to 35$0^{\circ}C$ for 9.5mm sphere. This large decrease of minimum film boiling temperature in this aqueous polymer solution may explain its ability to suppress spontaneous vapor explosions. Also, tests with applying a pressure wave showed that the vapor film behaved more stable against an external disturbance at higher polymer concentrations. These observations together with the experimental evidences of vapor explosion suppression in dilute polymer solutions suggest that the application of polymeric additives such as polyethylene oxide as low as 300ppm to reactor emergency coolant be considered to prevent or mitigate energetic fuel-coolant interactions during severe reactor accidents.s.