• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.177-177
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• 2002

• Magazine of the Korea Concrete Institute
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• v.12 no.2
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• pp.34-40
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• 2000

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.165.2-165.2
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• 2003

• Journal of the Korean Professional Engineers Association
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• v.33 no.3
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• pp.37-41
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• 2000

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.159-159
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.370.1-370
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• 1999

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.04a
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• pp.301-308
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• 2004

The present study mainly focuses on the inelastic stress analysis of the 1/4 scale prestressed concrete containment vessel model(PCCV) under internal pressure and evaluates not only failure mode but also ultimate pressure capacity of the PCCV. Inelastic analysis is carried out 2D axisymmertic FE model and 3D FE model using four concrete material models which are Drucker-Prager Model, Chen-Chen Model, Damaged Plasticity Model and Menetrey-Willam Model. The uplift phenomenon of the basemat is considered in the 2D axisymmetric FE models. It is found from the 2D axisymmetric analysis results that both of Drucker-Prager model and Damaged Plasticity Model have a good performance and the uplift of the basemat is too small to influence on the global behavior of the PCCV. The FE analysis results on the ultimate pressure and failure mode have a good agreement with experimental results.

• Journal of Advanced Research in Ocean Engineering
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• v.3 no.3
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• pp.112-124
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• 2017

As an LNG carrier preserves and transports liquefied natural gas under minus $163^{\circ}C$, the cargo tank has to have sufficient hull strength against not only the wave loads but also against loads caused by loading and unloading and thermal expansion to keep the LNG safely. The main insulation types for a CCS are No.96 and Mark III from GTT for the membrane LNG carrier. Particularly, the invar membrane plate in No.96 is very thin and its connections could experience high local stresses owing to such dynamic loads. Therefore, it should be verified whether those connections have sufficient fatigue lives for the purpose of operation and maintenance. This research aims at performing fatigue analysis with 0.1 fatigue damage criteria for 40 years of design life to support new membrane CCS development using proper S-N curves and the associated finite element modeling technique for each connection and then propose a reasonable design methodology.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.464-470
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• 1997

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks(LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code)computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanism of calandria vessel and containment. In addition, some insights for accident management program(AMP) are also given.