• 대한조선학회 특별논문집
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• 대한조선학회 2017년도 특별논문집
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• pp.93-104
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• 2017

IMO type C independent tank is one of the cargo containment system specified on IGC code. It is normally adopted for small and medium size liquefied gas carrier's cargo containment system and it can be applied to fuel tank of LNG fueled vessel. This study focuses on rule scantling process and evaluation methods in early design stage of type C independent tank. Actual design results of 22K LPG/Ammonia/VCM carrier's No.2 cargo tank are demonstrated. This paper presents the calculation methods of design acceleration and liquid height for internal design pressure as defined on IGC code. And this paper shows the applied results of classification rules about shell thickness requirement and buckling strength. Additionally this paper deals with evaluation methods of structural strength and cumulative fatigue damage using FE analysis.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3194-3201
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• 2023

In water-cooled power reactor, hydrogen is generated in case of steam zirconium reaction during severe accident condition and later on in addition to hydrogen; CO is also generated during molten corium concrete interaction after reactor pressure vessel failure. Passive Autocatalytic Recombiners (PARs) are provided in the containment for hydrogen management. The performance of the PARs in presence of hydrogen and carbon monoxide along with air has been evaluated. Depending on the conditions, CO may either react with oxygen to form carbon dioxide (CO2) or act as catalyst poison, reducing the catalyst activity and hence the hydrogen conversion efficiency. CFD analysis has been carried out to determine the effect of CO on catalyst plate temperature for 2 & 4% v/v H₂ and 1-4% v/v CO with air at the recombiner inlet for a reported experiment. The results of CFD simulations have been compared with the reported experimental data for the model validation. The reaction at the recombiner plate is modelled based on diffusion theory. The developed CFD model has been used to predict the maximum catalyst temperature and outlet species concentration for different inlet velocity and temperatures of the mixture gas. The obtained results were used to fit a correlation for obtaining removal rate of carbon monoxide inside PAR as a function of inlet velocity and concentrations.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.257-264
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• 1994

원자력발전소 중대사고시 용융된 노심과 잔류냉각수가 증기폭발을 일으켜 원자로 격납용기의 건전성을 위협할 수 있다. 본 연구에서는 증기폭발을 모사할 수 있는 실험 장치를 제작하고, 물과 프레온을 사용하여 증기폭발실험을 수행하였다. 이때 고속카메라를 사용하여 폭발현상을 관측하였고, 동압측정기와 압력분출관을 이용하여 생성되는 폭발압력과 기계적인 에너지를 계측하였다. 이를 토대로 증기폭발의 중요인자들(물의 온도, 물의 주입속도, 물의 주입 시간, 그리고 냉매의 깊이)에 대한 민감도 분석을 수행하였다. 그리고, 압력용기 바닥의 구조물이 용융/냉각재의 반응에 미치는 영향을 살펴보기위하여 실험용기 내부에 그리드를 설치하여 폭발실험을 실시하였다. 물/프레온의 폭발실험에서 계측된 기계적에너지를 이용한 에너지효율은 0.5∼l.6%인 것으로 계산되었다.

• 한국가스학회지
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• 제4권1호
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• pp.69-76
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• 2000

BLEVE로 인한 위험을 평가하기 위한 프로그램, BLEVE ESTIMATOR를 이용하여 온도에 따른 폭발량과 Flashing mass를 계산하였고, 부천 가스충전소의 사고를 모델로 피해를 예측하였으며, 상용프로그램인 Dupont의 SAFER 프로그램과 비교하였다. 폭발량과 Flashing mass는 폭발온도의 증가에 따라 지수함수로 증가하였으며, Propane이 n-Butane보다 상대적으로 높게 나타났다. 용기의 온도, 압력, 충전비가 높을수록 Overpressure가 높게 나타났다.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1994년도 가을 학술발표회 논문집
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• pp.291-300
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• Coupled systems mechanics
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• 제13권3호
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• pp.231-245
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• 2024

In the present work, a microelogated thermoelastic model based on Lord-Shulman (1967) and Green-Lindsay (1972) theories of thermoelasticity has been constructed. The governing equations for the simulated model are converted into two-dimensional case and made dimensionless for further simplification. Laplace and Hankel transforms followed by eigen value approach has been employed to solve the problem. The use of eigen value approach hasthe advantage of finding the solution of governing equationsin matrix form notations. This approach is straight forward and convenient for numerical computation and avoids the complicate nature of the problem. The components of displacement,stress and temperature distribution are obtained in the transformed domain. Numerical inversion techniques have been used to invert the resulting quantities in the physical domain. Graphical representation of the resulting quantities for describing the effect of microelongation are presented. A special case is also deduced from the present investigation. The problem find application in many engineering problems like thick-walled pressure vesselsuch as a nuclear containment vessel, a cylindricalroller etc.

• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.