• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1547-1554
• /
• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• 콘크리트학회논문집
• /
• 제15권1호
• /
• pp.60-68
• /
• 2003

한국원자력연구소(KAERI)의 프로그램 일환으로 콘크리트 격납건물 벽체 부재의 half-thickness 모델을 대상으로 인장실험을 수행하였다. KAERI의 이번 실험연구 목적은 격납건물 내부에서 예기치 못한 사고로 인하여 극한 내압이 작용할 때 콘크리트 격납건물의 성능을 평가할 수 있는 실험적으로 규명된 해석방안을 마련하는데 있다. 여기에 수록된 실험으로부터 얻은 데이터는 콘크리트의 균열거동 및 철근/콘크리트 사이의 상호작용 등을 포함한 재료모델을 요하는 해석방법을 검증하는데 유용할 것이다. 주요 실험 변수는 콘크리트의 압축강도로써 2축 인장을 받는 프리스트레스트 콘크리트 패널 부재의 균열거동에 미치는 영향을 살펴보았다.

• Nuclear Engineering and Technology
• /
• 제54권8호
• /
• pp.2960-2973
• /
• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• 한국강구조학회 논문집
• /
• 제16권6호통권73호
• /
• pp.747-757
• /
• 2004

강-콘크리트 합성보에서 강과 콘크리트 사이의 불완전 합성거동은 완전 합성보에 비해서 처짐이 매우 증가하게 된다. 특히, 춤이 깊은 데크 및 속 빈 PC슬래브 등을 사용한 매립형 합성보의 경우, 자체의 형상에 기인하여 처짐에 매우 취약하다. 본 연구에서는 기존 연구에서 유도한 슬립효과를 고려한 처짐 계산법 및 실험으로부터 구한 하중-슬립 관계로부터 매립형 합성보의 전단부착응력 및 슬립에 의한 추가 처짐 값을 제시하였다. 매립형 합성보의 처짐에 미치는 슬립의 영향은 완전 합성보의 강성 값에 비해 약 30%정도 감소함을 알 수 있었다. 또한, 실험 및 예측 값의 비교결과, 6%내외의 오차로 비교적 좋은 결과를 나타내었다.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.575-602
• /
• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Steel and Composite Structures
• /
• 제37권3호
• /
• pp.371-390
• /
• 2020

One of the most common strategies for retrofitting as-built reinforced concrete (RC) columns is to enlarge the existing section through the application of a new concrete layer reinforced by both steel transverse and longitudinal reinforcements. The present study was dedicated to developing a comprehensive model to predict the seismic behavior of as-built RC jacketed columns. For this purpose, a new sectional model was developed to perform moment-curvature analysis coupled by the plastic hinge method. In this analysis-oriented model, new methodologies were suggested to address the impacts of axial, flexural and shear mechanisms, variable confining pressure, eccentric loading, longitudinal bar buckling, and varying axial load. To consider the effective interaction between core and jacket, the monolithic factor approach was adopted to extent the response of the monolithic columns to that of a respective RC jacket strengthened column. Next, parametric studies were implemented to examine the effectiveness of the main parameters of the RC jacket strategy in retrofitting as-built RC columns. Ultimately, the reliability of the developed analytical model was validated against a series of experimental results of as-built and retrofitted RC columns.

• 국제초고층학회논문집
• /
• 제9권2호
• /
• pp.167-173
• /
• 2020

W-Project is 60-story mixed-use residential building complex project in Daegu, the third biggest city in South Korea. There are lots explorable items to be solved to secure structural safety and meet the serviceability requirements. This paper describes what kind of structural system is optimized based on the architectural requirements and structural components design and the grade of concrete strength altered on floors. The defining process of lateral resisting system of outrigger compared to the core ratio of typical plan is illustrated in detail.

• Nuclear Engineering and Technology
• /
• 제32권4호
• /
• pp.299-308
• /
• 2000

A safety assessment of reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The core melt relocation parameters were chosen within the ranges of physically realizable bounds. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were peformed using ANSYS code. Then, the calculated strain results and the established failure criteria were used in determining the failure probability of the lower head, In the explosion analyses, it is shown that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations. Strain analyses show that the vapor explosion-induced lower head failure is not possible under the present framework of assessment. The result of static analysis using the conservative explosion-end pressure of 50 MPa also supports the conclusion. It is recommended, however, that an assessment of fracture mechanics for preexisting cracks be also considered to obtain a more concrete conclusion.

• Computational Structural Engineering : An International Journal
• /
• 제1권1호
• /
• pp.11-20
• /
• 2001

The paper describes the conceptual design, structural modelling and wind load analysis of tall buildings. The lateral stiffness of the building can be obtained economically through the interaction of core walls with peripheral frame tube and/or bundle of frame tubes and integrated design of the basement. The main structural components should be properly distributed such that the building will deflect mainly in the direction of the applied force without inducing significant response in other directions and twist. The cost effectiveness can be further enhanced through close consultation between architects and engineers at an early stage of conceptual design. Simplified structural modelling of the building and its response in three principal directions due to wind load are included. Effects of the two main structural components on the performances of a 70-story reinforced concrete building in terms of peak drift and maximum acceleration under wind load are discussed.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.177-185
• /
• 2022

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.