• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2003년도 추계학술대회 논문집(II)
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• pp.271-276
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• 2003

In this paper, various hazard factors are reviewed on the train fire accident in railway tunnel and subway. In case studies of systematic risk evaluation on the tunnel fire accident, we have learned the critical fire safety points for accident prevention and damage reduction such as fire-endurance of infrastructure, mortality of heat & toxic smoke, emergency situation control and management of escape requirements etc.. These hazard analysis study will contribute for improving the railway fire-safety and establishing the long-tenn safety management plan.

• 한국전산유체공학회지
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• 제21권3호
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• pp.8-14
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• 2016

A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.

• 한국가스학회지
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• 제19권2호
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• pp.1-4
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• 2015

본 연구는 벌크로리 이 충전중 가스누출로 인해 화재가 발생하여 주변의 식당 등 2차적인 피해를 준 가스사고에 대해 위험성평가를 DNV사의 PHAST-RISK v6.7 program을 활용하여 제트화재시 영향면적, 복사열 농도 및 피해영향거리 등 가스 누출 확산 피해범위에 대해 도출하여 충전중 가스사고를 사전에 예방하고자 한다.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.180-189
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• 2023

We measured the heat load to a reactor vessel with and without the in-vessel debris bed under an IVR-ERVC condition. Mass transfer methodology was adopted based on heat and mass transfer analogy to achieve high Ra'_H of order ~10¹⁵ with compact test rigs. We postulated the in-vessel debris bed has a flat top and particulate debris was simulated as an identical diameter spheres. We conducted experiments varying the height of the debris bed and the results showed that Nusselt numbers decreased in both uppermost and curved surfaces with the increasing bed height. Once the debris bed is formed, it acts as an obstacle to the natural convective flow, which reduces the buoyancy. The reduction of driving force results in the impaired heat transfer in both upward and downward heat transfers.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.636-643
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• 2024

The coolability of the debris bed with a simulant of solidified corium is experimentally studied, focusing on the effects of the structure of the axial stratified debris bed on the dryout heat flux (DHF). DHF was obtained for the four structures with different particle sizes for the axial stratified debris bed under top flooding. The experimental results show that the dryout position of the axial stratified debris bed is formed at the stratified interface indicated by the temperature rise, and the DHF of the axial stratified bed is much lower than that of the homogeneous bed packed with the upper small particles. To predict the dryout heat flux of the stratified debris beds, by considering the properties of the mixed area, a one-dimensional dryout heat flux model of the porous medium is derived from a water and vapor momentum equation for porous medium, two-phase permeability modifications, interfacial drag, and the correlation between capillary pressure and liquid saturation and verified with the experimental data. The modified model can give reasonable results under different structures.

• 한국가스학회지
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• 제25권6호
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• pp.35-44
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• 2021

반도체 공정 중 PR (Photo Resist) 공정은 여러 인화성 물질을 혼합하여 사용하는 공정으로, 공정 장비가 Clean Room 내 설치되어 인화성 물질이 누출되는 경우 질식, 화재, 폭발 사고로 이어질 위험성이 크므로, 물질 누출시 발생할 수 있는 사고에 대한 영향을 분석하고, 작업자들의 안전을 보장할 수 있는지 평가하는 것이 필요하다. 본 연구는 FLACS CFD - Fire Module을 이용하여 10개의 누출, 화재 시나리오에 대한 CFD Simulation을 통해 Clean Room 내부 가상으로 설정한 Monitor Point에서의 복사열 및 온도 변화 값을 확인하였다. Clean Room 내부에서 발생한 화재는 층간 구조물에 높은 복사열을 전달하지만, 그 범위가 상당히 제한적이며, 단일 화재 사고로 인해 붕괴될 가능성이 희박하다. 화재 사고로 인해 탈출구로 이어지는 2 곳의 계단이 동시에 높은 복사열에 노출되는 시나리오는 없어, 화재 발생시 작업자가 탈출 가능하였다. 또한 복사열 및 온도 상승의 수준은 아래층으로 이동하면서 급격하게 낮아지는 것을 확인하였으며, API 520의 기준에 따라 작업자가 30초 동안 버틸 수 있는 복사열인 6.31 kW/m²에 노출된 작업자도 화재가 발생한 Clean Room 내부에서 충분히 탈출할 수 있음을 확인하였다.