• 대한기계학회논문집A
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• 제38권7호
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• pp.719-725
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• 2014

본 연구에서는 사출성형 충전공정 중 금형의 변형을 예측하기 위하여 비뉴턴 유동, 열전달, 구조해석이 함께 고려된 수치해석 연구를 수행하였다. 정밀 사출성형 금형을 설계/제작하기 위해서는 충전공정 중에 발생하는 금형의 변형을 정확하게 예측하는 것이 중요하다. 이와 같은 금형의 국부적인 변형은 다양한 요인에 의해 발생할 수 있으나, 용융된 고분자 수지의 유동에 의한 압력이 가장 큰 원인 중의 하나로 여겨지고 있다. 따라서, 본 연구에서는 2 차원 축대칭 형상의 단순 원형 디스크 제품의 금형을 모델링하고 이에 대한 수치해석을 수행하였다. 이를 바탕으로 금형 내부의 고분자 수지의 유동 특성과 금형 변형량, 온도 분포에 대한 분석을 수행하였다. 또한 다구치 방법을 기반으로 한 실험계획법을 도입하여 유동 속도, 금형 온도, 고분자 수지의 온도가 금형 변형에 미치는 영향을 파악하였다.

• Nuclear Engineering and Technology
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• 제45권1호
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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.

• 대한기계학회논문집B
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• 제35권4호
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• pp.425-429
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• 2011

원자로의 노심 손상에 따른 노심 용융물의 노외 유출시 코어캐처라고 불리는 설비를 통해 용융물을 억제하고 냉각시키게 된다. 이 때 노외 노심용융물의 거동은 희생물질과의 반응을 포함한 복잡한 물리적, 화학적 현상에 의해 결정된다. 이 연구는 기존의 용융물 거동 실험결과에 대해 용융물의 유동과 열전달의 세부적인 특성을 상용코드를 이용해 해석하여 검증함으로써 코어캐처의 설계에 활용할 수 있도록 하기 위한 것이다. 단순화된 채널에서 시간에 따른 용융물과 공기의 이상유동과 복사열전달을 VOF 모델과 구분종좌법을 적용하여 비정상상태에서 해석한 결과, 열전달에 따른 용융물 내부의 온도 변화 및 이에 따른 점성 변화 등을 예측할 수 있음을 확인하였다. 이러한 접근방식을 기초로 향후 용융물의 조성, 유량 및 용도 등의 조건에 따른 용융물의 거동에 대한 자세한 평가가 필요하다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

• 유변학
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• 제11권2호
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• pp.159-167
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• 1999

샌드위치 사출성형 공정은 기존의 사출성형 공정이 가지지 못하는 여러 장점들로 인해 최근 산업적으로 주목 받고 있는 고분자 가공 공정이다. 이 공정의 해석적인 접근은 거의 불가능하므로, 본 연구에서는 수치모사를 통해서 샌드위치 사출성형의 충전 공정을 연구하였다. 수치모사는 기본적으로 유한요소법을 사용하였고 Flow Analysis Network(FAN)/관할체적(Control Volume)법 등을 함께 이용하였다. 그리고 skin polymer의 선단을 확인할 수 있는 기존의 충전율 변수와 함께 skin polymer와 core polymer의 경계를 표시하는 새로운 충전율 변수를 도입하였고 이것을 이용하여 core polymer의 선단을 추적하였다. 새로운 충전율 변수는 두께 방향으로 온도장을 풀기 위해 나눈 각 층에서 정의되었다. 수치모사에 사용된 skin polymer와 core polymer로는 물성이 다른 두 고분자 물질을 주입시켜서 나타나는 충전 형태를 비교했다. 즉, 점도 상수, power-law 지수 등과 같은 유변 물성이 다른 두 고분자 물질을 충전시키기 위해 공정상 필요한 입구에서의 압력 등을 계산했으며 나중에 들어가게 되는 core polymer의 충전 완료 후 금형 내에서의 두께 방향과 흐름 방향으로의 분포 등을 구하였다. 또한 실제 공정 상에서 가공조건에 해당되는 switchover time과 벽 온도 등의 조건을 바꿔가면서 수치모사를 진행하였다. 사례 연구를 통하여 얻어진 물성과 가공 조건에 따른 core polymer의 충전 형태와 입구에서의 압력 등은 샌드위치 사출성형의 산업적 이용에 매우 유용하게 사용될 수 있다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.127-132
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• 2001

An experimental study on penetration integrity of the reactor vessel has been performed under external vessel cooling during a core melt accident. In this study a series of experiments are performed for the verification of the effects of coolant in the annulus between the ICI(In-Core Instrumentation) nozzle and the thimble tube and also the effects of external vessel cooling on the integrity of the penetration using the test section including only one penetration and $Al_{2}O_{3}$ melt as a corium simulant. The experimental results have shown that penetration is more damaged in the case of no external vessel cooling compared with the case of external vessel cooling. It is preliminarily concluded that the external vessel cooling is very effective measure for the improvement of the penetration integrity. Also it is confirmed from the experimental results that the coolant in the annulus reduces the melt penetration distance through the annulus and enhance the integrity of the reactor vessel penetration in the end.

• Nuclear Engineering and Technology
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• 제32권4호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3491-3496
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• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2511-2521
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• 2020

The Fukushima accident is characterized by the fact that three reactors at the same site experienced reactor vessel failure and the accident resulted in significant radiological release to the environment, which was about 1/10 of the Chernobyl releases. The safe removal of fuel debris in the reactor vessel and Primary Containment Vessel (PCV) and treatment of huge amount of contaminated water are the major issues for the decommissioning in coming decades. Discussions on the new researches efforts being carried out in the area of investigation of the end state of fuel debris and Boling Water reactor (BWR) specific core melt progression, development of technologies for the mitigation of radiological releases to comply with the strengthened safety requirement set after the Fukushima accident are discussed.

• Nuclear Engineering and Technology
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• 제38권8호
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• pp.733-752
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• 2006

Severe accidents in PWRs are characterized by a continuously changing geometry of the core due to chemical reactions, melting and mechanical failure of the rods and other structures. These local variations of the porosity and other parameters lead to multi-dimensionnal flows and heat transfers. In this paper, a comprehensive set of multi-dimensionnal models describing heat transfers, thermal-hydraulics and melt relocation in a reactor vessel is presented. Those models are suitable for the core description during a severe accident transient. A series of applications at the reactor scale shows the benefits of using such models.