• Nuclear Engineering and Technology
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• 제41권9호
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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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• 제29권2호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.213-213
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.216-216
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• 1999

• Nuclear Engineering and Technology
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• 제56권1호
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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.

• 대한기계학회논문집B
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• 제34권12호
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• pp.1051-1060
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• 2010

고화 입자 크기의 관점에서 TROI 용융물-냉각수 반응 실험의 결과에 대한 물질 효과를 분석하였다. 고화 입자 크기를 분석하면 용융물-냉각수 반응에서 초기 조건, 혼합, 폭발을 기적으로 해석할 수 있다. 증기 폭발이 발생한 경우와 폭발이 발생하지 않는 경우의 고화 입자 크기를 분석한 결과 증기 폭발이 발생한 경우에는 미세 입자가 많고 비교적 큰 입자는 적은 것으로 나타났다. 또한, 혼합 과정에 대한 정보를 보존할 수 있는 증기 폭발이 발생하지 않은 용융물-냉각수 반응을 이용하여 용융물 입자 크기에 대한 물질 효과를 분석하였다. 증기 폭발이 잘 발생하는 용융물은 증기 폭발에 참여할 수 있는 큰 입자를 많이 포함하고 있었고, 증기 폭발이 잘 발생하지 않는 용융물은 증기 폭발보다는 냉각되기 쉬운 작은 입자 혹은 미세 입자를 많이 포함하고 있었다.

• Nuclear Engineering and Technology
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• 제48권1호
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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.

• 대한기계학회논문집A
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• 제38권4호
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• pp.355-361
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• 2014

신규 원전에서 추진중인 외벽침수냉각 방식의 적용이 실패할 경우 노심용융물과 원자로공동 내유체의 상호작용으로 인해 증기폭발이 발생하며, 이는 격납건물 및 관통부 배관을 포함한 각 구조물의 건전성을 위협할 수 있다. 본 논문에서는 선행연구 분석결과를 토대로 증기폭발 현상을 모사할 수 있는 개선된 해석기법을 도출하고 알루미나 실험 모사를 통해 타당성을 확인하였다. 또한 동일한 기법을 원자로공동 해석에 적용하여 가상 파손위치에 따른 증기폭발 압력이력을 예측하였으며, 측면파손에 의한 최대압력 값이 하부파손에 의한 것보다 최대 70% 정도 높음을 보였다.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.544-554
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• 1995

폴리머 수용액의 증기폭발 억제 효과에 대한 물리적 현상을 이해하기 위해 폴리에틸렌옥사이드 수용액에서의 풀비등 특성을 실험적으로 관찰하였다. 본 실험에서는 22.2mm와 9.5mm 직경의 두 구형 체를 가열하여 여러가지 농도의 3$0^{\circ}C$ 수용액에서 냉각시켰다. 그 결과, 순수한 물에서는 7$0^{\circ}C$ 이상인 최소막비등온도($\Delta$ $T_{MFB}$)가 300ppm농도의 폴리머 수용액에서 22.2mm구의 경우 15$0^{\circ}C$ 까지, 9.5mm구의 경우 35$0^{\circ}C$까지 낮아짐을 알 수 있었다. 이러한 폴리머 수용액에서 최소막비등온도가 크게 낮아지는 현상은 이 수용액에서 중기폭발이 억제되는 이유로 해석될 수 있다. 또한, 외부 압력파의 막비등에 대한 영향을 관찰한 결과, 수용액의 농도가 클수록 증기막의 안정도가 커짐을 알 수 있었다. 이러한 폴리머 수용액에서의 비등 특성과 증기폭발 억제에 대한 실험 결과들은 원자로 비상냉각수에 폴리에틸렌옥사이드와 같은 폴리머를 최소 300ppm 정도 소량 첨가하는 방법으로 중대사고시 폭발적 FCI 반응을 방지 또는 완화할 수 있음을 제시한다.다.